ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 92

Document Title:

Oak Ridge National Laboratory – 

Occupational External Dose

Document Number: ORAUT-TKBS-0012-6 

Revision: 01

Effective Date: 09/10/2007

Type of Document TBD

Supersedes: Revision 00

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

Site Expert(s): Joseph Mohrbacher 

Approval: Signature on File Approval Date: 08/28/2007

Robert E. Burns, Jr., Document Owner

Concurrence: Signature on File Concurrence Date: 08/28/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/28/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 09/06/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 09/10/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 2 of 92

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/11/2004 00 New technical basis document for the Oak Ridge National Laboratory 

– occupational external exposures. Incorporates formal internal and 

NIOSH review comments. First approved issue. Initiated by Robert 

E. Burns, Jr.

09/10/2007 01 Approved Revision 01 initiated to include Attributions and 

Annotations, Purpose, and Scope sections, and additional 

references. Made editorial changes. Incorporates formal internal 

and NIOSH review comments. Training required: As determined by 

the Task Manager. Initiated by Robert E. Burns, Jr.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 3 of 92

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

8

6.1 Introduction ............................................................................................................................10

6.1.1 Purpose......................................................................................................................11

6.1.2 Scope.........................................................................................................................11

6.2 Basis of Comparison..............................................................................................................11

6.3 Dose Reconstruction Parameters........................................................................................... 11

6.3.1 Historical Administrative Practices..............................................................................12

6.3.1.1 Interpretation of Reported Data.................................................................... 18

6.3.2 Dosimetry Technology................................................................................................20

6.3.2.1 Beta-Gamma Dosimeters............................................................................. 20

6.3.2.2 Neutron Dosimeters ..................................................................................... 20

6.3.3 Dosimeter Calibration Practices..................................................................................21

6.3.3.1 Beta-Gamma Dosimeters............................................................................. 21

6.3.3.2 Neutron Dosimeters ..................................................................................... 22

6.3.4 Workplace Radiation Fields ........................................................................................23

6.3.4.1 Beta-Gamma Exposures.............................................................................. 23

6.3.4.2 Neutron Exposures ...................................................................................... 25

6.3.4.2.1 Health Physics Research Reactor and Graphite 

Reactor....................................................................................... 26

6.3.4.2.2 Workplace Neutron Field Characterizations 

Performed by Pacific Northwest Laboratory................................ 27

6.3.4.2.2.1 Building 3038: Isotope Materials 

Research Laboratory.............................................. 28

6.3.4.2.2.2 Building 3100: Source and Special 

Nuclear Materials Vault .......................................... 29

6.3.4.2.2.3 Building 7910: High Flux Isotope 

Reactor .................................................................. 30

6.3.4.2.2.4 Building 6000: Holifield Heavy Ion 

Research Facility.................................................... 30

6.3.4.2.2.5 Building 6010: Oak Ridge Electron 

Linear Accelerator .................................................. 31

6.3.4.2.2.6 Building 7735: Radiation 

Calibration Laboratory ............................................ 32

6.3.4.2.2.7 Building 2007: Radiation 

Standards and Calibration 

Laboratory.............................................................. 35

6.3.4.2.2.8 Buildings 7920 and 7930: 

Radiochemical Engineering 

Development Center .............................................. 36

6.3.4.2.2.9 Solid Waste Storage Area 5.................................... 43

6.3.4.2.3 Data from Neutron Dosimetry Records ....................................... 44

6.3.4.2.3.1 REDC Workers: Groups 2, 3, 4, 

and 17 .................................................................... 47

6.3.4.2.3.2 Group 7 .................................................................. 52

6.3.4.2.3.3 Unshielded Source Handlers: 

Group 9 .................................................................. 52
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6.3.4.2.3.4 SWSA 5 workers: Groups 11, 12, 

and 21 .................................................................... 54

6.3.4.2.3.5 Holifield and ORELA Facilities: 

Groups 6, 18, and 19.............................................. 57

6.3.4.2.3.6 Shielded Source Users: Group 13 ......................... 57

6.3.4.2.3.7 Neutron Source Make-up Workers: 

Group 14 ................................................................ 58

6.3.4.2.3.8 Other Workers: Groups 1, 5, 10, 

and 22 .................................................................... 59

6.3.4.2.3.9 Summary for all groups........................................... 60

6.4 Reported Quantities and Adjustments to Reported Dose ....................................................... 61

6.4.1 Reported Quantities and Associated Algorithms: Beta-Gamma................................. 61

6.4.2 Reported Quantities and Associated Algorithms: Neutrons........................................ 65

6.4.3 Adjustments to Recorded Dose: Beta-Gamma ..........................................................68

6.4.3.1 Deep Dose ................................................................................................... 68

6.4.3.2 Shallow Dose ............................................................................................... 68

6.4.4 Adjustments to Recorded Dose: Neutron...................................................................69

6.5 Limits of Detection and Missed Dose ..................................................................................... 70

6.5.1 Deep Dose .................................................................................................................70

6.5.2 Shallow Dose .............................................................................................................70

6.5.3 Neutron Dose .............................................................................................................71

6.6 Uncertainty in Photon and Neutron Dose ............................................................................... 73

6.7 Attributions and Annotations .................................................................................................. 73

References ........................................................................................................................................81

ATTACHMENT A, SELECTION OF IREP ENERGY GROUPS FOR BETA-GAMMA 

EXPOSURES FOR ORNL FACILITIES ...............................................................84

ATTACHMENT B, SELECTION OF IREP ENERGY GROUPS FOR NEUTRON 

EXPOSURES FOR ORNL FACILITIES ...............................................................87

ATTACHMENT C, NEUTRON-TO-PHOTON DOSE RATIOS FOR ORNL FACILITIES 

FOR 100% OCCUPANCY ...................................................................................89

ATTACHMENT D, NEUTRON-TO-PHOTON DOSE RATIOS FOR ORNL NEUTRON 

WORKER GROUPS............................................................................................90

ATTACHMENT E, NEUTRON DOSE DATA FOR ORNL NEUTRON WORKER 

GROUPS 1990–2004 ..........................................................................................91

ATTACHMENT F, FRACTIONS OF DOSE EQUIVALENT ABOVE ENERGY 

CUTOFF FOR NTA FILM FOR ORNL FACILITIES ............................................. 92
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 5 of 92

LIST OF TABLES

TABLE TITLE

6-1 ORNL TLDs used from 1975 to 1979 .....................................................................................

PAGE

17

6-2 ORNL dosimeter characteristics from 1944 to present ........................................................... 18

6-3 HPRR shield materials and corresponding neutron energies used to 

characterize TLDs..................................................................................................................23

6-4 Dose equivalent fractions for IREP neutron energy groups for 238Pu:Be 

and 252Cf sources in a water-shielded drum in Building 3038 ................................................. 29

6-5 Dose equivalent fractions for IREP neutron energy groups for drums of 

241Am-oxide in Building 3100.................................................................................................. 30

6-6 Dose equivalent fractions for IREP neutron energy groups for the west 

entrance to the ORELA room ................................................................................................. 32

6-7 Dose equivalent fractions for IREP neutron energy groups for the

RADCAL beta/X-ray irradiation room with a bare 252Cf source exposed in 

the adjacent neutron irradiation room..................................................................................... 34

6-8 Dose equivalent fractions for IREP neutron energy groups for the 

RADCAL beta/X-ray irradiation room with a moderated 252Cf source 

exposed in the adjacent neutron irradiation room................................................................... 34

6-9 Dose equivalent fractions for IREP neutron energy groups for the hallway 

outside the RADCAL neutron irradiation room with a moderated 252Cf 

source exposed......................................................................................................................35

6-10 Dose equivalent fractions for IREP neutron energy groups for the 

RADCAL equipment room with a moderated 252Cf source exposed in the 

adjacent neutron irradiation room........................................................................................... 36

6-11 Neutron-to-gamma dose ratios for several locations in RADCAL ........................................... 36

6-12 Fraction of dose equivalent above NTA cutoff for several locations in 

RADCAL ................................................................................................................................36

6-13 Dose equivalent fractions for IREP neutron energy groups for RASCAL 

Room 102 at 130 cm from the 238Pu:Be neutron calibration source........................................ 37

6-14 Dose equivalent fractions for IREP neutron energy groups for REDC 

Room 111 at 37 cm from a glovebox containing 252Cf ............................................................ 38

6-15 Dose equivalent fractions for IREP neutron energy groups for REDC 

Room 211 at 32 cm from the end of a glovebox bank ............................................................ 39

6-16 Dose equivalent fractions for IREP neutron energy groups for the REDC 

TDF containing 12 mg 252Cf.................................................................................................... 40

6-17 Dose equivalent fractions for IREP neutron energy groups for a waste 

cask in the REDC Limited Access Area.................................................................................. 41

6-18 Dose equivalent fractions for IREP neutron energy groups for

contamination on the walls of the REDC WTA transfer tunnel................................................ 42

6-19 Neutron-to-gamma dose ratios for several locations in the REDC.......................................... 43

6-20 Fraction of dose equivalent above NTA cutoff for several locations in the 

REDC.....................................................................................................................................43

6-21 Dose equivalent fractions for IREP neutron energy groups for waste 

casks in the SWSA 5 TRU storage bunker............................................................................. 44

6-22 Dose equivalent fractions for IREP neutron energy groups at a wall 

shared between two adjacent TRU storage bunkers at SWSA 5............................................ 45

6-23 Group-weighted NCRP 38 neutron quality factors and ICRP 60 radiation 

weighting factors ....................................................................................................................69

6-24 LODs for deep dose and dosimeter exchange frequencies for ORNL 

radiation workers....................................................................................................................70
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 6 of 92

6-25 Limits of detection for neutron dose and dosimeter exchange frequencies 

for radiation workers from 1975 to the present ....................................................................... 73

LIST OF FIGURES

FIGURE TITLE

6-1 Comparison of Hp(10) for photons with the energy responses for a 

sensitive DuPont 502 emulsion in a two-element film badge and a 

sensitive DuPont 555 emulsion in a multielement film badge .................................................

PAGE

21

6-2 Neutron spectrum measured by PNL for 238Pu:Be and 252Cf sources in a 

water-shielded drum in Building 3038..................................................................................... 28

6-3 Neutron energy spectrum measured by PNL for drums of 241Am-oxide in 

Building 3100 .........................................................................................................................30

6-4 Neutron energy spectrum at the west entrance to the ORELA room ...................................... 31

6-5 Neutron energy spectrum in the RADCAL beta/X-ray irradiation room with 

a bare 252Cf source exposed in the adjacent neutron irradiation room .................................... 33

6-6 Neutron energy spectrum in the RADCAL beta/X-ray irradiation room with 

a moderated 252Cf source exposed in the adjacent neutron irradiation 

room ......................................................................................................................................33

6-7 Neutron energy spectrum in the hallway outside the RADCAL neutron 

irradiation room with a moderated 252Cf source exposed........................................................ 34

6-8 Neutron energy spectrum in the RADCAL equipment room with a 

moderated 252Cf source exposed in the adjacent neutron irradiation room ............................. 35

6-9 Neutron energy spectrum in RASCAL Room 102 at 130 cm from the 

238Pu:Be neutron calibration source........................................................................................ 37

6-10 Neutron energy spectrum measured by PNL in REDC Room 111 at 37 

cm from a glovebox containing 252Cf ...................................................................................... 38

6-11 Neutron energy spectrum in REDC Room 211 at 32 cm from the end of a 

glovebox bank........................................................................................................................39

6-12 Neutron energy spectrum for the REDC TDF containing 12 mg 252Cf..................................... 40

6-13 Neutron energy spectrum for a waste cask in the REDC Limited Access 

Area .......................................................................................................................................41

6-14 Neutron energy spectrum for contamination on the walls of the REDC 

WTA transfer tunnel ...............................................................................................................42

6-15 Neutron energy spectrum for waste casks in the SWSA 5 TRU storage 

bunker....................................................................................................................................44

6-16 Neutron energy spectrum at a wall shared between two adjacent TRU 

storage bunkers at SWSA 5................................................................................................... 45

6-17 Normal probability plot for Group 2 neutron-to-photon dose ratios ......................................... 48

6-18 Normal probability plot for Group 2 neutron doses ................................................................. 48

6-19 Normal probability plot for Group 4 neutron-to-photon dose ratios ......................................... 49

6-20 Normal probability plot for Group 4 neutron doses ................................................................. 50

6-21 Normal probability plot for Group 3 neutron-to-photon dose ratios ......................................... 50

6-22 Normal probability plot for Group 3 neutron doses ................................................................. 51

6-23 Normal probability plot for Group 17 neutron-to-photon dose ratios ....................................... 51

6-24 Normal probability plot for Group 17 neutron doses ............................................................... 52

6-25 Normal probability plot for Group 7 neutron-to-photon dose ratios ......................................... 53

6-26 Normal probability plot for Group 7 neutron doses ................................................................. 53

6-27 Normal probability plot for Group 9 neutron-to-photon dose ratios ......................................... 54

6-28 Normal probability plot for Group 9 neutron doses ................................................................. 54
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6-29 Normal probability plot for Group 11 neutron doses ............................................................... 55

6-30 Normal probability plot for Group 21 neutron-to-photon dose ratios ....................................... 56

6-31 Normal probability plot for Group 21 neutron doses ............................................................... 56

6-32 Normal probability plot for Group 13 neutron-to-photon dose ratios ....................................... 58

6-33 Normal probability plot for Group 13 neutron doses ............................................................... 58

6-34 Normal probability plot for Group 14 neutron-to-photon dose ratios ....................................... 59

6-35 Normal probability plot for Group 14 neutron doses ............................................................... 59

6-36 Normal probability plot for neutron-to-photon dose ratios derived from 

ORNL personnel neutron monitoring records for 1990 – 2004 ............................................... 61

6-37 Normal probability plot for neutron doses ≥10 mrem for ORNL workers in 

the period from 1990 – 2004 .................................................................................................. 61
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 8 of 92

ACRONYMS AND ABBREVIATIONS

γ gamma (photon)

AEC U.S. Atomic Energy Commission

CEDS Centralized External Dosimetry System 

cm centimeter

D-T deuterium-tritium

DC ORNL designation for critical organ dose, analogous to deep or penetrating dose

DL ORNL designation for dose to the lens of the eye at a dose depth of 300 mg/cm2

DM ORNL designation for moderately penetrating dose at a dose depth of 130 mg/cm2

DOE U.S. Department of Energy

DOELAP Department of Energy Laboratory Accreditation Program

DP ORNL designation for penetrating dose at a dose depth greater than or equal to 

1,000 mg/cm2

DS ORNL designation for skin or superficial (nonpenetrating) dose at a dose depth of 

7 mg/cm2

ft foot

g gram

GSD geometric standard deviation 

HFIR High Flux Isotope Reactor

HPRR Health Physics Research Reactor

Hp(d) personal dose equivalent at depth d in tissue (d in centimeters)

hr hour

ICRP International Committee on Radiological Protection

ICRU International Committee on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

keV kilovolt-electron, 1,000 electron volts

LEN ORNL designation for dose to the lens of the eye (see DL)

LOD limit of detection

m meter

MeV megavolt-electron, 1 million electron volts

mg milligram

mm millimeter

MOP ORNL designation for moderately penetrating dose (see DM)

mR milliroentgen

mrep millirep

MW megawatt

1

n neutron

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 9 of 92

NTA nuclear track emulsion, type A 

ORELA Oak Ridge Electron Linear Accelerator

ORIC Oak Ridge Isochronous Cyclotron

ORNL Oak Ridge National Laboratory

OW designation for film dosimeter response behind the open window with no filter present, 

just the inherent filtration of the dosimeter materials (see W)

PEN ORNL designation for penetrating dose (see DP) 

PME probable maximum exposure

PNL Pacific Northwest Laboratory

PTR probable total reading

R roentgen

RADCAL Radiation Calibration Laboratory

RASCAL Radiation Standards and Calibration Laboratory

REDC Radiochemical Engineering Development Center

S designation for penetrating dose behind 1-mm-thick cadmium filter

SWSA solid waste storage area

TDF Transuranium Decontamination Facility

TLD thermoluminescent dosimeter

TLD-100 lithium-fluoride TLD chip with lithium at its natural enrichment 

TLD-600 lithium-fluoride TLD chip with lithium enriched in 6

Li (to enhance neutron response)

TLD-700 lithium-fluoride TLD chip with lithium enriched in 7

Li (to minimize neutron response)

TRU transuranic

TSR Tower Shielding Reactor

UCC-ND Union Carbide Corporation Nuclear Division

U.S.C. United States Code)

W alternate designation for film dosimeter response behind the open window (see OW). 

W watt

WEAF Waste Examination and Assay Facility 

WTA Waste Transfer Area
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 10 of 92

6.1 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 historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility 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 Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 

U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures 

• Radiation from diagnostic X-rays received in the treatment of work-related injuries 

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the probability of causation 

(POC). 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 11 of 92

6.1.1

The purpose of this TBD is to describe the external dosimetry systems and practices at Oak Ridge 

National Laboratory (ORNL) from late 1943 to the present. This document discusses historical and 

current practices in relation to the evaluation of external radiation exposure of monitored and 

unmonitored workers.

Purpose

6.1.2

ORNL began operations in early 1943. The startup of the Graphite Reactor in November 1943 

introduced the potential for personnel exposures to ionizing radiation. Operations with radioactive 

materials and radiation-generating devices at ORNL expanded over subsequent years with the 

Laboratory functioning in the role of technology development and pilot testing of new processes. 

These activities introduced increasing numbers of radiation sources and increasing potential for 

personnel exposure. 

Scope

In early 1943, ORNL instituted external dose measurement techniques using paired pocket ionization 

chambers. At that time, the Laboratory used photographic film dosimeters experimentally, which 

became the primary dosimeter of record at ORNL on June 25, 1944. Film dosimeters, augmented 

with nuclear track emulsion, type A (NTA) film for neutron monitoring, were the principal means of 

personnel monitoring at ORNL until the advent of thermoluminescent dosimeter (TLD) technology in 

the mid-1970s. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.7.

6.2 BASIS OF COMPARISON

Since the initiation of the Manhattan Engineer District project in the early 1940s, various radiation 

dose concepts and quantities have been used to measure and record occupational dose. The basis 

of comparison for reconstruction of dose is the personal dose equivalent, Hp(d), where d identifies the 

depth (in millimeters) and represents the point of reference for dose in tissue. For weakly penetrating 

radiation of significance to skin dose, d is 0.07 mm and the dose equivalent is noted as Hp(0.07). For 

penetrating radiation of significance to whole-body dose, d is 10 mm and the dose equivalent is noted 

as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities recommended by the International 

Commission on Radiation Units and Measurements (ICRU) for use as the operational quantities to be 

recorded for radiological protection purposes (ICRU 1993). In addition, Hp(0.07) and Hp(10) are the 

radiation quantities used since the 1980s in the U.S. Department of Energy (DOE) Laboratory 

Accreditation Program (DOELAP) to accredit personnel dosimetry systems (DOE 1986). The 

International Agency for Research on Cancer selected Hp(10) as the quantity to assess error in 

historical recorded whole-body dose for workers in Agency nuclear worker epidemiologic studies 

(Thierry-Chef et al. 2002). The basis for comparison for neutron radiation is more complicated 

because the calibration of dosimeters to measure neutron dose was based historically on different 

dose quantities such as first collision dose, multiple collision dose, dose equivalent index, and so 

forth. The numerical difference in using these dose quantities compared to the Hp(10) dose used in 

current DOELAP performance testing could be evaluated by using the relative values of the dose 

conversion factors for the respective dose quantities in conjunction with characteristics of the 

respective ORNL neutron dosimeter response characteristics and workplace radiation fields. 

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray and gamma ray), and neutron radiation types, energies and 

exposure geometries, and their respective dosimeter response characteristics are crucial for the 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 12 of 92

assessment of bias and uncertainty in the original recorded dose in relation to the radiation quantity 

Hp(10). Bias and uncertainty for current dosimetry systems are typically well documented for Hp(10) 

(ORAUT 2007). The performance of current dosimeters is often comparable to performance 

characteristics of historic dosimetry systems in the same, or highly similar, facilities or workplaces. In 

addition, the application of current performance testing techniques to earlier dosimetry systems can 

result in a consistent evaluation of all dosimetry systems. Dosimeter response characteristics for 

radiation types and energies in the workplace are crucial to the overall analysis of error in recorded 

dose. 

Overall, the accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2006) guidelines depend on:

• Administrative practices adopted by facilities to calculate and record personal dose based on 

technical, administrative, and statutory compliance considerations

• The use of dosimetry technology that includes the physical capabilities of the dosimetry 

system, such as the response to radiation type and energy, especially in mixed radiation fields

• Calibration methods used for the respective monitoring systems and the similarity of the 

methods of calibration to sources of exposure in the workplace

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions

An evaluation of the original recorded doses based on these parameters is likely to provide the best 

estimate of Hp(10) and, as necessary, Hp(0.07) for individual workers, with the least relative overall 

uncertainty. 

6.3.1

Beginning in October 1943, ORNL issued paired pocket dosimeters to workers assigned to duties in 

the restricted areas associated with Graphite Reactor operations. Pocket dosimeters were used in 

pairs because of reliability issues associated with early designs. ORNL issued two dosimeters so 

there was a backup if one of them failed, went off the scale, or was otherwise unusable. An individual 

received these dual dosimeters if restricted area entries were likely more than three times per week. 

If entry into restricted areas was likely to be less than three times per week or on a random basis, 

workers received dosimeters for each entry. Workers received dosimeters at the beginning of the 

shift at the entry portal and deposited them at the same portal at the end of the shift. 

Historical Administrative Practices

The dosimeters were read, data were recorded, and the dosimeters were recharged before the 

beginning of the next shift. Both readings were recorded, but the lower was identified as primary and 

maintained as the dose of record. Each dosimeter was matched and assigned to an individual and 

replaced at the portal for the next day’s operation. ORNL investigated irregularities with the pocket 

dosimeters or their readings and made administrative notations on the dose record if a justification for 

the irregularity was evident. Follow-up was initiated if the dosimeter results exceeded 30 mR for the 

day. In addition, notations were made if a dosimeter was lost, damaged, returned late, or not worn. 

In 1944, exposure controls became more complicated due to the introduction of the 

bismuth-phosphate process for plutonium separation and other operations involving short-decay 

irradiated fuel slugs from the Graphite Reactor. The accumulation of fresh fission products and 

process upsets started to produce significant dose rates in the new process areas. As a result, 

steady improvements in radiation exposure control and technological improvements in dosimetry 

systems were required. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 13 of 92

ORNL experimentally used a beta-gamma film dosimeter based on the design developed at the 

Metallurgical Laboratory at the University of Chicago during the first half of 1944 (Mitchell et al. 1993). 

On June 25, 1944, ORNL introduced this dosimeter as the primary beta-gamma dosimeter of record. 

It used a DuPont Type 552 film packet consisting of two emulsions, one sensitive (Type 502) and the 

other insensitive (Type 510). The insensitive film was included in case there was a large exposure 

beyond the range of the sensitive film. The film dosimeter consisted of two windows (elements): One 

open (i.e., no filtration other than that of the film packet), and one under a 1-mm-thick cadmium filter 

with a density thickness of approximately 900 mg/cm2

. This two-element film dosimeter was used 

only for beta-gamma monitoring. It was assigned initially only to employees required to work in 

restricted areas more than 3 days per week. 

Neutron personnel monitoring using NTA film was formally introduced at ORNL around 1949. 

However, NTA film was apparently used on a limited basis to supplement field measurements as early 

as February 1945 (Wirth, Morgan, and Curtis 1945) and, beginning on December 28, 1947, the cards 

used to record personnel exposures included a column for neutron dose. Neutron dose data were 

recorded for both the open window (OW) of the dosimeter and for the region behind the cadmium 

shield (S). The two readings recorded as a fraction of tolerance values for fast and thermal neutron 

exposures. The neutron exposures and tolerance were expressed in terms of number of tracks per 

field (of view) or, more specifically, number of tracks in a certain number of fields. For example, fast 

and thermal neutron tolerance levels were both 20 tracks per 12 fields in early 1949. The previous 

value (from late 1947 through the eleventh week of 1949) was 16, although it is not clear if this value 

was also based on 12 fields. It appears that the fast and thermal neutron tolerance level increased to 

22 tracks as of the 38th week of 1949. Once it became available, NTA film was used in all assigned 

employee badges and exchanged on a weekly basis. However, the film was not processed unless 

the health physicist recommended it. The practice of recording beta-gamma and neutron film badge 

readings on the same data card was discontinued as of December 30, 1951, although separate 

neutron exposure records were in use before then. 

The film dosimeters were exchanged and processed on a weekly basis, until a quarterly exchange 

was introduced in the third quarter of 1956. The use of paired pocket dosimeters continued for 

exposure control purposes. Film dosimeters were initially stored at the entrance portal to the 

restricted area for employees to pick up and return each day. There is evidence that strong 

administrative controls to ensure that workers wore dosimetry as prescribed were not in effect, and 

that wearing dosimetry was a matter of the honor system. There are no indications that film 

dosimeters were used at the portal storage locations for monitoring for or subtracting ambient 

exposure. Employees were not required to wear film dosimeters while working outside the restricted 

area. Those entering the restricted area on an occasional basis (less than three times per week) 

were considered visitors and were assigned limited badges. The limited badges were processed at 

the end of the assignment or weekly, whichever was shorter. Nonradiation workers who did not enter 

the restricted area did not receive film dosimetry. These included, but were not limited to, visitors, 

clerical personnel, accountants, and construction personnel. 

As of November 20, 1951, the wearing of a combination security badge and film dosimeter, called a 

photobadge, was required of all regular workers on the ORNL site regardless of work area. This 

marked a transition period between the practice of storing personnel film dosimeters at the entry 

portals and the adoption of take-home badges. It appears the transition to the take-home photobadge 

was completed by late September 1953. It is important to note that take-home badges were issued 

mainly to ORNL employees and other workers with routine access to the site (i.e., that had a 

photobadge issued). Other personnel, such as temporary employees, some construction workers, 

services personnel, or visitors, did not receive take-home dosimeters. The weekly exchange of 

photobadge dosimeters continued until the third quarter of 1956, when the frequency was changed to 

quarterly. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 14 of 92

A new four-element film dosimeter was introduced in late September 1953. This dosimeter had four 

shields (plastic, copper, lead, and cadmium). The intent of the design was to enable depth-dose 

measurements in accordance with National Bureau of Standards (NBS) Handbook 59 (NBS 1954); 

however, “no serious attempt was made to utilize the full capability of the multi-filter badge until the 

beginning of the second half of 1956” (Hart 1966). Beginning then, four depth-dose quantities were 

considered: A skin dose, a moderately penetrating dose, a lens-of-the-eye dose, and the penetrating 

dose. However, no routine determination of skin dose was made because the element was behind an 

effective density-thickness of approximately 80 mg/cm2 (Hart 1966). ORNL did not report skin (or 

superficial) dose until the second half of 1961. 

A film dosimeter designated ORNL Model I was introduced in the late 1950s, and ORNL Model II was 

introduced in July 1960. These dosimeters employed filters consisting of an OW, plastic, aluminum, 

and cadmium. Gold foils were included in the cadmium element for use in the event of an 

accident-level neutron exposure (Hart 1966). The ORNL Model II film dosimeter used a DuPont Type 

544 film packet, which consisted of Type 555 sensitive emulsion and Type 834 insensitive emulsion. 

ORNL switched from the DuPont Type 552 film packet to the Type 544 sometime in 1959. 

Beginning in 1961, only two depth dose assignments were made from personnel dosimetry data: The 

skin (or superficial) dose (designated DS) and the critical organ dose (DC). Techniques were 

developed to allow assignment of a skin dose based on extrapolation of the response of the film 

element under the OW of the dosimeter (Hart 1966). 

As of 1968, different classes of personnel dosimetry were assigned to workers based on their duties 

and access requirements. Three classes of personnel dosimetry components were used: Classes A, 

B, and C. The class assigned to a given worker depended on the assigned photobadge. 

Class A dosimetry consisted of the ORNL Model II four-element film dosimeter described above. This 

photobadge was assigned to workers with routine access to the site, including ORNL employees, 

loaned persons, consultants, and others who required frequent access. The combination badges for 

ORNL employees and loaned personnel were take-home. The Model II film dosimeter was also used 

in so-called gate badges, which were left at the entry portals (i.e., not taken home). Exchange 

frequencies for both the beta-gamma film packet and the NTA packet were quarterly for radiation 

workers. The film packets were renewed annually in the dosimeters issued to personnel for whom 

monitoring was not required (i.e., nonradiation workers) (Gupton 1968). 

Class B dosimetry consisted of a DuPont Type 544 film packet, an NTA film packet, and components 

for accident dosimetry. The beta-gamma film packet was renewed monthly. The NTA film was 

renewed annually, or any time it was removed for processing. Class B dosimetry was issued to 

persons with casual access to ORNL facilities that included radiation zones, including U.S. Atomic 

Energy Commission (AEC) employees and other visitors with access to radiation zones. In addition, 

Class B dosimetry was used in temporary ORNL employee badges issued to employees as 

replacements for their photobadges (Gupton 1968). 

Class C dosimetry consisted of a beta-gamma film packet, an NTA film packet, and accident 

dosimetry components. The beta-gamma film packet was renewed quarterly and the NTA film was 

renewed annually. Persons issued Class C dosimetry were not allowed to enter radiation zones. 

Class C dosimetry was issued to some construction workers, services visitors, tour groups, and other 

visitors for whom entry to radiation zones was prohibited. It is important to note that not all 

construction workers were issued Class C dosimetry. ORNL maintained a construction photobadge 

for workers who required frequent access to the site. This badge was the same as the Class A 

dosimetry issued to ORNL employees (Gupton 1968). 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 15 of 92

Gupton (1968) stated that as of 1968 the beta-gamma film packet used in the Class C dosimeter was 

an Eastman-Kodak Type 2. However, Gupton (1969) stated it was a DuPont Type 544, the same as 

the Class A and Class B dosimeters. ORNL switched from the DuPont Type 544 film packet to the 

Eastman Kodak Type 2 at some point, but the date is not clear. However, the Type 544 packet 

appears to have been in use through at least 1969. The Eastman Kodak Type 2 packet had sensitive 

and insensitive emulsions bonded to opposite sides of the film base. In a 1979 interview, Gupton 

indicated that the change from the DuPont Type 544 film packet to the Eastman Kodak Type 2 packet 

resulted in lowering the dosimeter limit of detection (LOD) from 30 mR to 20 mR (Beck, Stansbury, 

and Watson 1979). The Type 2 film packet was in use in 1975 as a backup to the newly adopted 

TLDs, but because the date when the Type 2 packet replaced the Type 544 is unknown, the LOD for 

all ORNL film badges has been assumed to be 30 mR. 

Film dosimeter technology was used for beta-gamma and neutron personnel monitoring at ORNL until 

TLD technology replaced it in 1975. In fact, review of the claims files shows the use of TLDs actually 

began for at least some radiation workers at ORNL as early as 1974 to assign deep dose. The 

practice of combining the security badge and personnel dosimeter continued, and workers received a 

specific type of TLD depending on their duties. At first, each TLD contained a Type 2 film packet as a 

backup. NTA film was used as an integral part of the ORNL neutron dosimeter until elimination of 

routine use sometime in the mid-1980s. 

The practice of assigning personnel dosimetry based on different classes (A, B, and C) of security 

badges apparently continued following the initial widespread use of TLDs at ORNL in January 1975. 

At first, TLDs were issued only to those with Class A photobadges. Individuals with Class B and C 

badges were issued film dosimetry as described above (Gupton 1974). Workers with Class A badges 

received different models of TLDs depending on their potential for radiation exposure and the types of 

radiation to which they were likely to be exposed. For instance, neutron-sensitive dosimeters were 

assigned only to individuals whose work involved potential neutron exposures. 

As of January 1975, workers with Class A photobadges were assigned one of four different TLDs. 

The four versions were distinguished with color-coded labels: Red, yellow, green, and none. Red 

TLDs went to radiation workers with the potential for neutron exposure. Only 40 to 50 red dosimeters 

were available (Gupton 1974). This was a four-element dosimeter with TLD-600 and TLD-700 

elements (two of each) and NTA film. Yellow dosimeters, which were assigned to radiation workers 

for which neutron exposure was not anticipated, contained one TLD-100 and one TLD-700. 

Approximately 130 of these dosimeters were available (Gupton 1974). Uncolored dosimeters were 

the same as the yellow version, but the two TLD chips were under different filters. In 1975 or 1976, 

the use of thin (0.015-in.) TLD-700 chips began in yellow and red dosimeters (or Type 2 and 3 

dosimeters described below). It is not clear if the thin chips were used from the outset, but the 

uncolored (or Type 1) badge appears always to have used a thicker 0.035-in. TLD-700 chip. 

Uncolored dosimeters went to ORNL employees who were not classified as radiation workers. 

Approximately 4,500 of these dosimeters were available (Gupton 1974). Green dosimeters, which 

contained a single TLD-700 chip, were issued to non-ORNL employees. Approximately 700 of these 

dosimeters were available (Gupton 1974). These initial TLDs included beta-gamma film packets as 

backup in the event of a problem with the TLD elements. 

As of August 1975, the system of Class A, B, and C photobadges had been redefined. There were 

still three classes of badges, but they were now referred to as:

• Employee take-home

• Nonemployee take-home

• Nonemployee gate or exchange (not take-home) (Gupton 1975)
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 16 of 92

Persons with photobadges were designated as monitored or nonmonitored, referring to whether they 

received personnel dosimetry. All ORNL employees were monitored, as were persons with 

take-home photobadges for whom monitoring was requested by their Divisional Radiation Control 

Officer. All others were designated as “monitoring not required” (Gupton 1975). Monitored persons 

were assigned one of three types of personnel dosimetry, depending on their work assignments and 

exposure potential. The three dosimeters were designated Type 1, Type 2, and Type 3. 

The Type 1 dosimeter was the same as the uncolored dosimeter described above: A TLD-100 chip 

under cadmium and a TLD-700 chip under the OW. The Type 2 dosimeter contained a TLD-100 chip 

under an aluminum filter and two thin (0.015-in.) TLD-700 chips, one under the OW and the other 

under plastic (Gupton 1975). This dosimeter had the same role as the yellow dosimeter described 

above (i.e., issued to radiation workers for whom neutron exposure was unlikely). The Type 3 (or red) 

dosimeter was a four-element design with a TLD-600 and a TLD-700 under a cadmium filter, a 

TLD-600 and a TLD-700 under an aluminum filter, and two thin (0.015-in.) TLD-700 chips under the 

OW and plastic filters, respectively (Gupton 1975). A vapor-sealed NTA film monitored fast neutrons. 

The Type 3 dosimeter was issued to radiation workers for whom neutron monitoring was appropriate, 

as was the red dosimeter described above. 

Gupton (1978) describes the different models of TLDs used for personnel monitoring at ORNL as 

being the same as the Type 1, 2, and 3 dosimeters above. However, by that time they were 

designated Class 1, Class 2, and Class 3 rather than Type 1, 2, and 3. 

The Class 1 dosimeter was assigned to individuals for whom little or no radiation exposure was 

anticipated (Gupton 1978). This dosimeter was used for nonradiation workers, including but not 

limited to clerical, administrative, and construction workers as well as visitors. These were functions 

for which ORNL anticipated that an inconsequential fraction of the potential dose would be from 

neutrons, betas, or photons of energy below 30 keV. The exchange period for these dosimeters was 

annual. Positive readings, which were rare, were investigated, and a notational dose was made as 

necessary.

The Class 2 dosimeter was used for whole-body and skin dose, and for indicating neutron exposure 

(i.e., it was not used for neutron dosimetry but would indicate that an unexpected neutron exposure 

had occurred). This dosimeter was assigned to radiation workers where there was low anticipation of 

neutron exposure. 

The Class 3 dosimeter was used for measurement of beta, gamma, and neutron dose. The analysis 

of results from Class 3 dosimeters in accordance with established algorithms (see Section 6.4.2) 

allowed ORNL dose evaluators to assign dose equivalents for fast, thermal, and intermediate neutron 

energy groups. The total neutron dose equivalent was the sum of the three components (Gupton 

1978). 

ORNL continued to use the three dosimeters described above after the two-element Union Carbide 

Corporation Nuclear Division (UCC-ND) TLD came into use at the four Oak Ridge Reservation sites 

(ORNL, Y-12, K-25, and the Paducah Gaseous Diffusion Plant) in 1981. ORNL TLDs were used 

either in lieu of or in addition to the UCC-ND TLD. The UCC-ND TLD had two LiF chips: one under an 

OW and the other under an aluminum filter. The unit provided beta-photon discrimination in 

determining skin and whole-body doses (McLendon 1980). A 1-g foil of indium was included in the 

UCC-ND badge for accident dosimetry. 

Gupton (1981) describes the ORNL Class 1, 2, and 3 dosimeters as being called green dot, yellow 

dot, and red dot, respectively. He also describes an H dosimeter that was assigned to persons who 

did not have a UCC-ND dosimeter for short-term use where doses between 1% and 10% of DOE 

annual limits were likely during the period of use. By this time, the yellow dot (formerly yellow or 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 17 of 92

Class 2) dosimeter had evolved to a four-element design with three thin TLD-700 chips and one 

TLD-100 chip of normal thickness. The green dot dosimeter was the same as the yellow dot design, 

but had an annual exchange frequency. The ORNL neutron dosimeter (Class 3 or red dot) remained 

unchanged. Yellow dot and red dot dosimeters were referred to collectively as HP meters. In 

addition, a blue dot dosimeter for visitors and other nonemployees was the same as the green dot 

dosimeter (Berger and Lane 1985). 

Table 6-1 summarizes the various ORNL TLDs used from 1975 to 1979. 

Sometime around the mid-1980s, ORNL began using four-element Panasonic TLDs to augment the 

two-element UCC-ND dosimeters. Two Panasonic dosimeters were employed for this purpose, one 

for beta-gamma monitoring and the other for neutrons (MMES 1992). 

In 1989, ORNL began implementing the Centralized External Dosimetry System (CEDS), which used 

two dosimeters for whole-body monitoring: A four-element beta-gamma dosimeter (blue holder) and a 

four-element neutron-sensitive dosimeter (red holder). The CEDS beta-gamma dosimeter became 

the dosimeter of record as of January 1989. However, the use of the Panasonic neutron dosimeter 

continued as the neutron dosimeter of record until January 1990. The exchange frequency for the 

CEDS dosimeters is quarterly; however, neutron dosimeters can be exchanged independently of the 

beta-gamma dosimeters if an individual’s duties involve neutron exposure from different energy 

spectra. Workers may be issued more than one neutron dosimeter during a monitoring period so 

appropriate TLD response-to-neutron-dose factors can be applied. Only one neutron dosimeter is 

issued at a time, and workers are required to wear them even if they were not working in neutron 

areas. In addition, there are sometimes workers for whom dose evaluation is performed more 

frequently than quarterly, as ORNL health physics staff can request that an individual’s dosimeter be 

processed at any time. These individuals include workers for whom higher doses are anticipated, and 

sometimes pregnant workers are monitored more frequently. If an interim dose evaluation is 

requested for a neutron worker, in general both the beta-gamma and neutron dosimeters are 

processed. From review of claims files it appears such cases will be apparent to the dose 

reconstructors, as results for each dosimeter issued to an individual appear to be provided. 

The CEDS dosimetry system is DOELAP-accredited, so reported deep and shallow dose values can 

be treated as representative of Hp(10) and Hp(0.07) for dose reconstructions. 

Table 6-2 summarizes personnel dosimeters worn by radiation workers at ORNL from 1944 to the 

present. 

Table 6-1. TLDs used from 1975 to 1979. 

Approximate date range TLD designation TLD elements (filters) Usage

Jan. 1975–Aug. 1975 Uncolored TLD-100 (Cd)

TLD-700 (OW)

ORNL employees, nonradiation 

workers

Green TLD-700 (OW) Non-ORNL employees

Yellow TLD-100 (Al)

TLD-700 (OW)

Radiation workers in non-neutron 

areas

Red TLD-600 (Al)

TLD-600 (Cd)

TLD-700 (OW)

TLD-700 (plastic)

Radiation workers with potential 

for neutron exposure

Aug. 1975–1978 Type 1 TLD-100 (Cd)

TLD-700 (OW)

Nonradiation workers

Type 2 TLD-100 (Al)

TLD-700 (OW)

TLD-700 (plastic)

Radiation workers in non-neutron 

areas
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 18 of 92

Type 3 TLD-700 (OW)

TLD-700 (plastic)

TLD-600 + TLD-700 (Al)

TLD-600 + TLD-700 (Cd)

Radiation workers with potential 

for neutron exposure

1978–1981 Class 1 Same as Type 1 Nonradiation workers

Class 2 Same as Type 2 Radiation workers in non-neutron 

areas

Class 3 Same as Type 3 Radiation workers with potential 

for neutron exposure

1981–1989a Green Dot Same as Yellow Dot Nonradiation workers

Yellow Dot TLD-100 (Al)

TLD-700 (OW)

TLD-700 (plastic)

TLD-700 (Cd)

Radiation workers in non-neutron 

areas

Red Dot Same as Type 3 Radiation workers with potential 

for neutron exposure

a. UCC-ND and Panasonic TLDs were also used in this period. 

6.3.1.1 Interpretation of Reported Data

The following items are pertinent to the interpretation of personnel dosimetry data recorded for ORNL 

workers. 

• Film dosimeter readings (results) encountered in ORNL personnel monitoring histories before 

June 25, 1944, could be unreliable because the use of dosimeters during this time was 

experimental (Hart 1966). 

• It appears that the reporting limit for personnel exposures based on film dosimeter readings 

was 10 mR from 1944 through 1946, with exposures reported to the nearest 5 mR. It also 

appears that the reporting limit was revised up to 30 mR (in most cases) around 1947. NOTE: 

The 10 mR reporting limit apparently used in the 1944-to-1946 period was not taken to imply 

that the LOD for the film dosimeters was that low. 

• Film density results less than 30 mR were originally recorded as 0 (zero) or less than 30 mR 

(Hart and Martin 1949). This same reference asserts that approximately 85% of all film badge 

readings at that time were in the range from zero to 30 mR(ep). (Readings from behind the 

shielded element were reported in mR, and those from the open window were reported in 

Table 6-2. Dosimeter characteristics from 1944 to present. 

Period Dosimeter type Filters Film packets used

1944–1953 Two-element film badge OW, Cd DuPont Type 552,

NTA (in routine use around 

1949)

1953–June 

1956

Four-element film badge OW, plastic, Cu, 

Cd, Pb

DuPont Type 552,

NTA

July 1956–

1974

Four-element film badge OW, plastic, Al, 

Cd 

DuPont Type 552 (until 1959), 

DuPont Type 544 (until 

replaced with Eastman Type 

2),

Eastman Type 2 (early 1970s),

NTA
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 19 of 92

1975–1988 

(TLDs were in 

use on limited 

basis in 1974). 

ORNL TLDs - different designs assigned 

to workers based on duties and 

employment status. Used LiF TLD 

chips. 

Panasonic TLDs used from mid-to-late 

1980s; these used lithium-borate and 

calcium-sulfate elements. 

OW, plastic, Al, 

Cd (not all 

elements used in 

every design).

Eastman Type 2,

NTA

(Type 2 film was used as 

backup, NTA was used in 

neutron dosimeters only. Use 

of film was discontinued in 

mid-1980s.)

1981–1988 Two-element UCC-ND TLD (some 

workers might have worn these rather 

than ORNL dosimeters)

OW, Al N/A

1989–present Commercial Harshaw four-element TLD 

system (DOELAP accredited); 

Panasonic neutron dosimeters used 

from 1989 to 1990. 

OW, plastic, 

Teflon, Cu

N/A

mrep.) It appears that ORNL maintained a consistent practice of differentiating dosimeter 

results that were zeros or less than LODs from blank fields in dosimetry records. The 

Laboratory’s practice was to attempt always to mark results that were less than established 

reporting limits with a zero or some other indication to distinguish these entries from blank 

fields. Blanks apparently implied that monitoring of a given type was not performed for that 

period. However, the practice of recording zeros for values less than reporting levels did not 

necessarily apply to all fields for a given record (e.g., if the OW reading for a monitoring period 

was recorded as zero, the S field would typically be blank – zeros were not duplicated across 

all fields). In addition, the distinctions between blanks and zeros described here apply only to 

handwritten data on the various dosimetry data cards maintained by ORNL; they do not apply 

to summary-level information or information from databases such as the so-called IBM system 

installed in 1949. A review of claim files verifies that blanks on semiannual exposure summary 

records from the IBM system correspond to monitoring periods during which values of zero 

were recorded for weekly exposures on the handwritten data cards. Thus, blanks in the 

computer-generated summary records do not mean the same thing as they do on the 

handwritten data cards. Dose reconstructors should always use individual data cards if they 

are available rather than summary-level data. 

• It appears that information on neutron and beta-gamma dose was always either differentiated 

on the data cards or reported on separate cards. Therefore, dose reconstructors should not 

have to attempt to separate reported dose values into different components as long as the 

data cards (and not summary-level information) are available. However, sometimes neutron 

dose was included in the deep and shallow dose totals reported on the Beta-Gamma 

Dosimetry Record cards, though annotations were apparently made in such cases to identify 

when neutron dose was added. Neutron Dosimetry Record cards were used in addition to 

beta-gamma cards. NOTE: It does not appear that neutron doses were included in the 

semiannual summary reports from the IBM System beginning in 1949. Review of claims data 

indicates that these reports include only beta-gamma exposures. 

• Neutron exposure data were originally expressed as a fraction of tolerance, where both the 

exposure and the tolerance were expressed in terms of tracks per field, or tracks for a number 

of fields. Section 6.4.2 provides information on interpreting neutron exposure results 

expressed in this manner. Later, neutron exposure data were expressed in terms of tracks per 

scan, where a scan was a unit of volume (of emulsion). It appears conversion factors (in dose 

equivalent per tracks per scan) were recorded directly on the dosimetry data cards for neutron 

exposure data expressed in this way. 

• From the mid-1970s through the late 1980s, ORNL issued different TLD models to workers, 

depending on their duties. The type of dosimeter can be inferred from the individual elements 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 20 of 92

listed on the Beta-Gamma Dosimetry Record cards. NOTE: The September 1975 version of 

these cards for the yellow dosimeter incorrectly listed the TLD-100 element as 100C rather 

than 100A, which implied that it was under a cadmium filter, when in fact it was under 

aluminum. Dosimetry technicians would sometimes correct the label to read 100A by hand, 

but often it was left as 100C. 

6.3.2

Dosimetry methods at ORNL evolved over the years with the development of improved technology 

and better understanding of the complex radiation fields in the workplace. The following sections 

discuss performance characteristics of the various personnel dosimetry methods employed. 

Dosimetry Technology

6.3.2.1 Beta-Gamma Dosimeters

Figure 6-1 shows the response of the ORNL film badge to photon radiation of different energies as 

well as the Hp(10) response. The figure shows two responses for film badges: One for a sensitive 

DuPont 502 emulsion in a two-element film badge (Pardue, Goldstein, and Wollan 1944), and one for 

a sensitive DuPont 555 emulsion in a multielement film badge (Thornton, Davis, and Gupton 1961). 

The response of the sensitive Eastman Kodak Type 2 film in a multielement film badge should be 

similar to that of the sensitive DuPont 555 emulsion. The film badges show an under-response at 

lower photon energies and an over-response at photon energies around 100 keV. This over-response 

is due primarily to the silver and bromine in the film emulsions. The response of the newer TLD 

badges provided a better match to the Hp(10) response in the soft tissues of the body due to the 

lower atomic numbers of the lithium and fluorine in the TLD chips (Horowitz 1984; Cameron, 

Sunthanalingham, and Kennedy 1968). 

6.3.2.2 Neutron Dosimeters

The two general types of neutron dosimeters that have been used at ORNL (NTA film and 

neutron-sensitive TLDs) differ significantly in their responses to neutrons of different energies. In 

general, the response of NTA film decreases with decreasing neutron energy that is greater than a 

threshold energy estimated to be about 500 keV (IAEA 1990), while TLD response increases with 

decreasing neutron energy. Results reported at the first AEC Neutron Dosimetry Workshop in 1969 

indicated that laboratory dose measurements made with NTA film were about one-half to one-fourth of 

those measured with other methods, including neutron-sensitive TLDs (Vallario, Hankins, and Unruh 

1969). The response of both dosimeters is highly dependent on the neutron energy spectra, and both 

10|-2 10|-1 10|0

Photon Energy (MeV)

0.1

1

10

Relative Response

Hp(10)

Two-Element Film Badge

Multi-Element Film Badge
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 21 of 92

Figure 6-1. Comparison of Hp(10) for photons with the 

energy responses for a sensitive DuPont 502 emulsion 

in a two-element film badge (Pardue, Goldstein, and 

Wollan 1944) and a sensitive DuPont 555 emulsion in a 

multielement film badge (Thornton, Davis, and Gupton 

1961). 

dosimeter types require matching the laboratory calibration neutron spectra to the workplace neutron 

spectra for reliable results.

NTA film would not, in general, have been very sensitive to the leakage neutron spectra associated 

with ORNL reactor facilities. 

6.3.3

Potential error in recorded dose is dependent on the method of calibration for the dosimeters and the 

extent of the similarity between the radiation fields used for calibration and those encountered in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters for low-energy photon radiation and both the NTA and TLD dosimeters for 

neutron radiation. 

Dosimeter Calibration Practices

6.3.3.1 Beta-Gamma Dosimeters

A 226Ra photon source was used to calibrate film badges for response to photons for both the OW and 

cadmium-shielded elements. The radium source was used to develop calibration curves for the 

sensitive film for both the open and filtered elements of the dosimeters, and to develop a curve for the 

insensitive film under the shielded element. The use of 226Ra for the calibration of personnel 

dosimetry at ORNL continued until the late 1980s, when it changed to 137Cs. Such isotopic photon 

sources were not used exclusively for photon calibrations, as the Laboratory made efforts to 

understand the performance of various personnel dosimeters in use to different types of radiation. 

For instance, film dosimeters were calibrated to various X-ray spectra as of the mid-1950s to generate 

calibration curves used for evaluating special cases involving X-ray exposures (West 1992). 

In the ninth week of 1947, natural uranium plate sources were introduced and used as beta calibration 

sources to augment calibrations from the radium gamma source (Hart 1966). At that time, the 

assumed dose rate on the surface of the uranium slab was 270 mrep/hr. On March 1, 1951, this 

value was revised to 240 mrep/hr (Hart 1966). The uranium slab was used to develop calibration 

curves for converting film density from under the OW of the dosimeters to dose in millirep. Before the 

ninth week of 1947, readings from under the OW were expressed in milliroentgen based on calibration 

to 226Ra photons. Therefore, OW readings for this period must be adjusted if they are to be compared 

with results from later periods. Similarly, the OW results for the period between the ninth week of 

1947 and March 1, 1951, require adjustment to enable consistent comparisons. Section 6.4.3.2 

discusses these adjustments. 

Initially, ORNL felt the OW of the film badge was sensitive to any beta energy able to penetrate the 

skin and “cause damage to underlying tissue,” and that “[a]ny exposure which the meters will not 

record is thought to be non-dangerous” (Roberson 1947). The disparity between the response of the 

OW to low-energy photons versus that for beta radiation was well understood, as was the difference 

in response between low- and high-energy photons. The ORNL choice to evaluate the OW readings 

using a calibration curve derived from natural uranium beta was based on taking advantage of the 

conservatism associated with evaluating the photon response of the OW using the uranium calibration 

curve. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 22 of 92

Thornton, Davis, and Gupton (1961) discussed the results of experiments that showed film density per 

unit absorbed dose from natural uranium beta for the OW was equal to that for the cadmium-filtered 

element for 226Ra photons. That document also discussed results that showed the film density per 

unit absorbed dose from beta sources varied slowly with the average beta energy over a wide range 

and concluded that film dosimeters were equally sensitive to beta and gamma radiation. NOTE: 

These results compare the beta response of the OW to the photon response of the cadmium-filtered 

element. Hart (1966) asserted that the film density per unit dose from uranium beta for the OW was 

2.2 times lower than that from 226Ra gamma radiation. By extension, this suggests the OW response 

of the dosimeter to 226Ra photons was 2.2 times that of the cadmium-shielded element if the assertion 

that the uranium beta response of the OW and the 226Ra photon response of the shielded element 

were equivalent is accurate. 

As of the late 1980s, calibration and performance testing of CEDS dosimeters has been performed 

using both the ORNL Radiation Standards and Calibration Laboratory (RASCAL) and third-party 

facilities (ORNL 1994). RASCAL has provided irradiations for calculating element correction 

coefficients, developing algorithms, performing TLD response studies, ensuring reader reproducibility, 

and conducting blind audits. However, RASCAL was not the primary source of information for the 

development of neutron correction factors, which were determined from in situ characterization of 

workplace spectra via tissue-equivalent proportional counters and other instrumentation. 

Facilities at RASCAL include a low-scatter room for performing neutron and high-energy photon 

irradiations, a beta irradiation facility, and an X-ray facility. Isotopic photon sources include 137Cs and 

241Am. Neutron sources include bare and D2O-moderated 252Cf as well as a 238Pu:Be source. Beta 

sources include 90Sr/Y and 204Tl, along with 147Pm point sources and a depleted uranium slab source. 

The X-ray facility is capable of providing a number of standard spectra. Dosimeters can be calibrated 

using a variety of geometries, including free-in-air and DOELAP torso, finger, and extremity 

phantoms. RASCAL has participated in intercomparisons with the National Institute of Standards and 

Technology (ORNL 1994). 

6.3.3.2 Neutron Dosimeters

The NTA film was calibrated initially using the Graphite Reactor and indium foil (Davis, Gupton, and 

Hart 1954). Thornton, Davis, and Gupton (1961) assert that a fast neutron exposure of 100 mrem 

(the weekly tolerance at that time) would produce approximately 1,000 recognizable tracks per square 

centimeter in the film, and that “[t]he figure for thermal neutrons is about the same.” The small track 

lengths to be discerned in the emulsion required high (950 power) magnification, resulting in a field of 

view of 2 × 10-4/cm2

. Thus, 100 mrem of fast or thermal neutron exposure gave a track density of 

0.2 track per field (Thornton, Davis, and Gupton 1961). 

As of the late 1960s, calibration of NTA film dosimeters used either unmoderated Po:Be or 238Pu:Be 

sources, with the resultant delivered dose equivalent expressed in terms of equivalent exposure from 

a Po:Bo source. Specifically, the use of a fluence-to-dose-equivalent conversion factor of 

2.23 × 10-7 n/cm2

/rem was based on cross-referencing between the calibration sources and a Po:Bo 

spectrum, because the latter was “more suited for calibration purposes” (Gupton 1968, 1969). NTA 

film dosimeters were exposed to 1 rem to obtain a calibration factor in terms of average number of 

tracks per scan per rem. A scan was a unit of volume defined by the depth of the emulsion, the width 

of the field of view, and a length of 2 mm. 

NTA film was not capable of registering a response to neutrons having energies below about 500 keV. 

Nonetheless, when it first came into use, ORNL described a means of computing a thermal neutron 

dose based on results from NTA film; this dose was the difference between readings under the OW 

and cadmium-filtered element. Presumably, the calibration factors applied to get from track density to 

dose were determined from exposure of the dosimeters to thermal neutron beams from the Graphite 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 23 of 92

Reactor. However, it is unclear if neutron dose was actually reported in this way (i.e., as separate fast 

and thermal components). Personnel neutron exposure data reviewed to date have been in terms of 

either OW and cadmium-filtered element readings or single dose equivalent values. NOTE: Early 

tolerance values for fast and thermal neutron exposure, which were expressed in terms of tracks per a 

number of fields, were equal (e.g., the tolerance values were both 20 tracks per 12 fields in 

mid-1949). 

By the late 1960s, ORNL had apparently abandoned the idea of assigning a thermal neutron dose on 

the basis of results from NTA film dosimeters and, in fact, Gupton (1968) states “[p]ersonnel 

monitoring devices are inadequate for dosimetry of intermediate energy neutrons.” However, he goes 

on to say that “relative blackening” of the region under the cadmium filter of the beta-gamma 

dosimeter provided an indication that “nominal” exposure to intermediate-energy neutrons had 

occurred. Dose estimations could therefore be made based on appropriate survey data if exposure to 

intermediate-energy neutrons was indicated (Gupton 1968). 

Neutron calibration practices improved over time to where neutron doses were assigned based on 

characterizing the personnel neutron dosimetry to known neutron fields [e.g., from the Health Physics 

Research Reactor (HPRR)], and combining these response characteristics with knowledge of 

workplace neutron spectra. As such, reporting of neutron dose evolved from terms of OW and S 

(corresponding to response of the NTA film from under the OW and cadmium-filtered dosimeter 

elements, respectively) to single values representing the contribution from all energies. Isotopic 

sources (238Pu:Be and 252Cf) were used for routine performance testing once TLDs came into use 

(Gupton 1981). 

In 1985, the neutron response of the red dot TLD (which was the dosimeter used at ORNL for neutron 

monitoring) was determined by irradiation of badges at the HPRR. The HPRR was a small, 

unmoderated, fast reactor fueled by enriched uranium. Different neutron energies were obtained by 

employing different shield materials, as summarized in Table 6-3. 

Table 6-3. HPRR shield materials and corresponding 

neutron energies used to characterize TLDs (Berger 

and Lane 1985).

Shield material

Median neutron 

energy (keV)

Mean neutron 

energy (keV)

None (bare) 780.0 1,280

Steel 340.0 580

Concrete 3.3 560

Lucite 0.07 640

TLDs were irradiated to different neutron energies obtained from HPRR to determine 

response-to-dose calibration factors as well as relative energy response for development of neutron 

dose algorithms. Neutron dose algorithms and dose assignments were based on these 

characterizations and on characterizations of workplace neutron spectra. The same practices were 

applied to the CEDS program when it was adopted in 1989. A significant effort was begun at this time 

to further characterize the workplace neutron spectra at ORNL (see Section 6.3.4.2). 

6.3.4

6.3.4.1 Beta-Gamma Exposures

Workplace Radiation Fields

To date, quantitative assessments of beta-gamma radiation fields encountered in ORNL facilities have 

not been identified. This is not surprising, because dosimetrists generally attempt to differentiate and 

account for different radiation qualities in beta-gamma personnel monitoring using the response of 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 24 of 92

individual dosimeter elements or ratios between them. Further, ORNL’s technical basis document for 

its current dosimetry system states, “Actual field characterizations of photon spectra have not been 

performed at facilities served by CEDS” (ORNL 1994). Indeed, attempting to develop defendable 

characterizations of photon and beta fields at ORNL facilities over its history could prove impractical. 

Therefore, dose reconstructors should apply favorable to claimant assumptions for radiation types and 

energies when converting deep and shallow beta-gamma dosimeter dose for ORNL workers to organ 

dose for input to the Interactive RadioEpidemiological Program (IREP). The method and 

energy-group assumptions to use for organ dose calculations from beta-gamma dosimeter dose 

readings are discussed below. NOTE: Dose reconstructors should use exposure-to-organ-dose 

conversion factors when computing organ dose from photon exposures measured before January 

1989 for cases where the organ of interest is something other than skin. 

For a given beta-gamma dosimeter result recorded before 1989, organ dose should be computed 

using the reported deep dose plus any net shallow dose for cases where exposure from transuranic 

(TRU) material was likely or the organ of interest is the skin. Net shallow dose is the difference 

between the reported shallow and deep (or OW and S) results. Net shallow dose should be 

considered in cases of potential TRU exposure to account for dose that would not have penetrated 

the shielded dosimeter elements. (It is assumed that TRU exposure would be accurately evaluated 

after 1989, when the ORNL external dosimetry system became DOELAP accredited.) 

The recommended energy groups to be applied to the deep component are given in Attachment A, 

which provides a listing of principal ORNL facilities where workers could have received occupational 

beta-gamma exposure and their approximate dates of operation. For most facilities, the 

recommended energies are a mixture of 75% more-than-250-keV photons and 25% 30- to 250-keV 

photons [1]. Exceptions are certain accelerator facilities and those where TRU materials were 

handled, but it is not known if fission products were also present. The net shallow component, if 

necessary, should be treated as either 100% less-than-30-keV photons or 100% greater-than-15-keV 

electrons. This choice should be made based on the worker’s job function and exposure potential, 

e.g., the less-than-30-keV photon group should be selected for an individual that worked in a TRU 

facility, while the greater-than-15-keV electrons would be an appropriate choice for a uranium worker 

or one who worked with mixed fission products for whom dose to the skin is needed [2]. The net 

shallow component is not used for deep organs unless exposure from TRU materials is likely. NOTE: 

Because the OW element of early film dosimeters was not capable of differentiating between beta and 

photon dose, ORNL employed a calibration curve based on calibration to a uranium beta spectrum, 

knowing this would provide conservative results for photon exposure. This, plus the over-response of 

the film to low-energy photons, resulted in what ORNL felt were conservative assessments of 

nonpenetrating dose. In fact, this conservatism led to the introduction of the quantity probable total 

reading in the second half of 1951, followed in subsequent years by the development of methods and 

algorithms for assigning shallow dose based on the response of the individual elements of the 

multielement dosimeters. The energy group recommendations given above and in Attachment A 

should be adjusted as appropriate if more specific information is available for a worker’s reported 

dose for a given monitoring period or exposure incident. 

Examples illustrating the method for computing organ dose from beta-gamma dosimeter results are 

given below. For the first two examples, consider an individual that worked in Building 3026-D and 

had a reported deep dose of 80 mR and a reported shallow dose of 100 mrep. For the last example, 

consider the same reported dose values for an individual that worked in Building 7920 instead (see 

Attachment A for energy group selections). 

Example 1: Compute the dose to the stomach. The dose values and energy groups to be entered 

into IREP would be: 

>250 keV photons: (0.080)(0.75)(0.885) = 0.053 rem (acute)
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 25 of 92

30 keV – 250 keV photons: (0.080)(0.25)(1.251) = 0.025 rem (acute)

where the fractions 0.75 and 0.25 are those given in Attachment A for the corresponding energy 

groups, and the values 0.885 and 1.251 are the exposure-to-organ-dose conversion factors from the 

External Dose Reconstruction Implementation Guidelines (NIOSH 2006) (100% AP exposure 

geometry has been assumed). The net shallow component is not needed, since it would be treated 

as electron exposure for this facility, and the organ of interest is not the skin. 

Example 2: Compute the dose to the skin. The dose values and energy groups to be entered into 

IREP would be: 

>250 keV photons: (0.080)(0.75) = 0.060 rem (acute)

30 keV – 250 keV photons: (0.080)(0.25) = 0.020 rem (acute)

>15 keV electrons: (0.100 – 0.080) = 0.020 rem (acute)

where the net shallow component is treated as >15 keV electrons for the 3026-D facility. 

Example 3: Compute the dose to the stomach. The dose values and energy groups to be entered 

into IREP would be: 

>250 keV photons: (0.080)(0.75)(0.885) = 0.053 rem (acute)

30 keV – 250 keV photons: (0.080)(0.25)(1.251) = 0.025 rem (acute)

<30 keV photons: (0.100 – 0.080)(0.182) = 0.004 rem (acute)

where the net shallow component has been treated as <30 keV photons because Building 7920 is a 

TRU facility. The factor 0.182 is the exposure-to-organ-dose conversion factor, as in Example 1 

above. 

6.3.4.2 Neutron Exposures

With few exceptions, information on workplace neutron energy spectra or neutron exposure data for 

ORNL facilities before the late-1980s is sparse. In general, neutrons were not a large source of 

personnel exposures at ORNL (relative to photon sources), and this could be part of the reason that 

little information seems to be available on workplace neutron fields or exposures. Relatively few 

ORNL staff members were designated as neutron workers for a given monitoring period. Information 

is particularly lacking for many of the reactors that operated at ORNL early in its history. However, 

neutron dose rate information is available for the HPRR and the Graphite Reactor (see Section 

6.3.4.2.1). Section 6.3.4.2.2 details measurements performed by Pacific Northwest Laboratory (PNL) 

to characterize the response of the CEDS neutron dosimeters to workplace neutron spectra at ORNL 

and Y-12. These characterizations took place between 1989 and 1991. Section 6.3.4.2.3 describes 

the use of ORNL personnel neutron dosimetry records from 1990 through early 2004 to extract 

distributions of personnel neutron dose and neutron-to-photon dose ratios for different groups of 

ORNL neutron workers. ORNL defines different groups of neutron workers for the purpose of 

assigning appropriate response-to-dose conversion factors for their TLD results. The different groups 

are defined based on the worker’s duties and occupancy in different neutron fields. 

Attachment B contains a list of principal ORNL facilities where workers could have received 

occupational neutron exposure, their approximate dates of operation, and the radiation energy 

selections that should be made when entering personnel exposure data in IREP. The energy 

selections are based on characterizations of ORNL workplace neutron spectra where available, as 

discussed in the following sections. The favorable to claimant default of 0.1 to 2.0 MeV neutron 

energy was assigned for facilities for which characterization data were unavailable [3]. Dose 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 26 of 92

reconstructors should adjust the neutron energy group data in Attachment B as appropriate if 

information that is more specific is available for a given worker or exposure incident. 

Attachment C contains a summary of neutron-to-gamma dose ratio data derived from the workplace 

radiation field characterizations discussed in Sections 6.3.4.2.1 and 6.3.4.2.2. Dose reconstructors 

should use these ratios with caution, as they are highly dependent on location, shielding, and 

scattering characteristics and, in the case of freshly separated material, can be a function of material 

age relative to the time required for gamma-emitting decay products to reach equilibrium. The 

importance of ingrowth on neutron-to-gamma dose ratios depends on the proportion of photon dose 

that is due to radioactive decay versus that from neutron interactions with surrounding materials. 

With the exception of those for the HPRR and the Graphite Reactor, all of the values given in 

Attachment C were derived from the PNL characterization measurements. These values reflect fixed 

measurements performed in conservatively chosen locations. In addition, all of the neutron-to-photon 

dose ratios discussed in Sections 6.3.4.2.1 and 6.3.4.2.2 and Attachment C reflect 100% occupancy 

(i.e., they do not account for photon dose not associated with the neutron field). Thus, applying these 

values to an individual’s reported photon dose could result in a significant overestimation if the 

individual received photon exposure from other sources or geometries and relative occupancy is not 

considered. Section 6.3.4.2.3 addresses neutron-to-photon dose ratios for different groups of ORNL 

neutron workers where occupancy in all photon fields (i.e., total photon dose) is inherently included. 

These ratios were obtained from personnel neutron dosimetry records provided by ORNL and are 

summarized in Attachment D. Attachment E summarizes neutron dose data for ORNL neutron 

workers obtained from the neutron dosimetry records. 

Attachment F contains a summary of available workplace neutron field characterization data in terms 

of the fraction of neutron dose equivalent from neutrons with energies above the 500-keV cutoff for 

NTA film. These data are included to provide some basis for estimating the sensitivity of NTA film for 

personnel monitoring in similar neutron fields, as necessary. The values were derived from the 

information in Sections 6.3.4.2.1 and 6.3.4.2.2. 

The neutron energy spectra, doses, and neutron-to-gamma dose ratios discussed in the following 

sections and Attachments B through F are derived primarily from measurements performed since 

1989. However, in general, significant differences between conditions at the time of these 

measurements and those before or after are not expected. Conditions at reactor, accelerator, and 

calibration facilities are not likely to have changed significantly over time, and operations at the 

Radiochemical Engineering Development Center (REDC) have remained consistent over its history. 

Changes in the TRU mix at the REDC are not expected to have a significant impact on the 

characterization and dosimetry data because the fission neutron and associated gamma spectra from 

the different neutron-emitting nuclides are essentially identical. Similarly, conditions in the TRU waste 

and storage areas are not expected to vary considerably over time.

6.3.4.2.1 Health Physics Research Reactor and Graphite Reactor 

Kerr and Johnson (1968) reported radiation survey measurements performed for the HPRR control 

building. Their results were applicable to both steady-state and burst operation of the reactor (it was 

capable of both), and included dose equivalent rates (per unit reactor power) for fast and thermal 

neutrons as well as photons. The results showed neutron dose equivalent rates inside the control 

building (Building 7710) were dominated by fast neutrons, with approximately 90% of the total neutron 

dose equivalent rates due to fast neutrons and 10% to thermals (Kerr and Johnson 1968). The 

neutron quality factors applied in this study were 10 for fast neutrons and 2.5 for thermals. The Kerr 

and Johnson data showed the neutron-to-photon dose equivalent ratio for the HPRR control building 

to be approximately unity. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 27 of 92

A study of the radiation attenuation properties of the Graphite Reactor’s concrete shield (ORNL 1958) 

provided insight on fast and thermal neutron dose equivalent rates at the surface of the shield and the 

corresponding neutron-to-photon dose equivalent ratio. Measurements of absorbed dose rates from 

fast neutrons and photons and thermal neutron flux were made as a function of position moving from 

the inner face to the outer face of the shield. The fast neutron and photon dose rates were stated in 

terms of ergs per gram of tissue per hour. Kerr (2003) converted these data to absorbed dose rate in 

rad per hour, then to dose equivalent employing a fast neutron quality factor of 10. The thermal 

neutron flux data were converted to dose equivalent rate using a neutron quality factor of 2.5. The 

Kerr (2003) reevaluation of the 1958 measurements shows the neutron dose equivalent rate at the 

surface of the shield to be approximately 1 mrem/hr, with essentially all of the neutron dose from 

thermal neutrons. The data also show a neutron-to-photon dose equivalent ratio of approximately 0.1 

at the surface of the shield. 

Dose rate measurements performed in the shield of the Graphite Reactor indicated the neutron 

spectrum at the exterior face of the shield was quite thermal, which is not surprising given its 7-ft 

(2.1-m) thickness. Thus, if a particular neutron exposure occurred in the general work area of the 

Graphite Reactor, the thermal (less than 0.01 MeV) neutron energy group would be a defendable 

choice. However, if the dose reconstructor is uncertain that all of the exposure was received in the 

general work area, or if exposure from experimental ports is possible, the favorable to claimant default 

0.10- to 2.00-MeV neutron energy is advised [4]. 

6.3.4.2.2 Workplace Neutron Field Characterizations Performed by Pacific Northwest 

Laboratory

After the termination of most reactor operations at ORNL [about 1987, with the exception of the High 

Flux Isotope Reactor (HFIR)], the primary locations where workers could have been exposed to 

neutrons were:

• The REDC (Buildings 7920 and 7930), from operations involving separation and purification of 

TRU materials produced in the HFIR

• Accelerator facilities [the Oak Ridge Electron Linear Accelerator (ORELA) and Holifield Heavy 

Ion Research Facility, Buildings 6010 and 6000, respectively]

• The HFIR beam room (Building 7900)

• TRU waste storage facilities [Solid Waste Storage Area (SWSA) 5 and Building 3100, the 

Source and Special Nuclear Materials Vault]

• Calibration facilities: RASCAL, Building 2007; Radiation Calibration Laboratory (RADCAL), 

Building 7735

• The Isotope Materials Research Laboratory (Building 3038)

From 1989 through 1991, Pacific Northwest Laboratory (PNL) performed measurements to 

characterize the response of the CEDS neutron TLDs to various workplace neutron spectra at ORNL 

and Y-12. The measurements consisted of characterization of workplace neutron fields using 

tissue-equivalent proportional counters, Bonner spheres, 3

He detectors, and other instruments 

(MMES 1992). The characterization of the neutron field in each location generally consisted of 

measurements of dose equivalent rates, spectral measurements, and TLD exposures. Spectral 

measurements were not always performed, however, and in some cases TLD exposures were 

omitted. In general, both beta-gamma and albedo neutron TLDs were exposed on phantoms at each 

measurement location. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 28 of 92

Neutron spectral data, where available, allow description of the neutron fields present at the time of 

the characterizations in terms of the relative contribution to the total dose equivalent from each 

neutron energy group used in IREP. In addition, the TLD data can be used to establish approximate 

neutron-to-gamma dose ratios. The following sections summarize characterization measurements 

performed for ORNL facilities and associated data. In addition, these sections contain information on 

worker exposure scenarios assumed by ORNL for determining personnel neutron dose for some of 

the facilities. 

The fluence-to-dose-equivalent conversion factors and neutron quality factors used in the PNL 

measurements are similar to those from National Council on Radiation Protection and Measurements 

(NCRP) Report 38 (NCRP 1971) and International Commission on Radiological Protection (ICRP) 

Publication 21 (ICRP 1971). Sims and Killough (1983) contains a comparison of 

fluence-to-dose-equivalent conversion factors from NCRP Report 38, ICRP Publication 21, and 

several other commonly used information sources on fluence-to-dose equivalent conversion factors 

and neutron quality factors. 

6.3.4.2.2.1 Building 3038: Isotope Materials Research Laboratory 

At the time of the characterization measurements, Building 3038 was used to store isotopic neutron 

sources such as 238Pu:Be and 252Cf before shipment. Two sets of measurements were performed: 

One with a 238Pu:Be source in a 1-in.-thick Lucite glovebox, and one with both a 238Pu:Be source and 

a 252Cf source in a water-shielded drum. Measurements with the neutron sources in the 

water-shielded drum included a characterization of the neutron energy spectrum. Only TLD 

irradiations were performed with the 238Pu:Be source in the glovebox. 

Figure 6-2 shows the neutron energy spectrum determined for the two sources in the water-shielded 

drum in terms of dose equivalent rate. Table 6-4 lists the contribution to the total dose equivalent 

from the four IREP neutron energy groups of interest for ORNL workers. The table lists the data as 

measured as well as rebinned into two groups for convenience to the dose reconstructors. Combining 

the contributions from lower energy neutrons into the 0.1-to-2.0-MeV group is favorable to claimants 

given the difference in neutron quality factors for these energies. 

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-2. Neutron spectrum measured by PNL for 238Pu:Be and 

252Cf sources in a water-shielded drum in Building 3038 [5]. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 29 of 92

Table 6-4. Dose equivalent fractions for 

IREP neutron energy groups for 238Pu:Be 

and 252Cf sources in a water-shielded drum 

in Building 3038 [6]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 3.0

0.01–0.1 0.0

0.1–2.0 39.1

2.0–20.0 57.9

Favorable to claimant values

0.1–2.0 42

2.0–20.0 58

More than 92% of the neutron dose equivalent measured by PNL for the two sources in the 

water-shielded drum was from neutrons with energies above the 500-keV cutoff for NTA film. Thus, 

NTA film should have provided reliable results if used for personnel monitoring in similar spectra. 

TLD irradiations performed with the two sources in the water-shielded drum indicated a 

neutron-to-gamma dose ratio of approximately 3.7. In contrast, TLD irradiations performed with the 

238Pu:Be source in the Lucite glovebox showed a neutron-to-gamma dose ratio of approximately 32.2. 

For choosing appropriate TLD correction factors for assigning neutron dose to workers in this area, 

ORNL assumed 80% of the exposure came from sources in the shielded configuration and 20% from 

unshielded sources (MMES 1992). 

6.3.4.2.2.2 Building 3100: Source and Special Nuclear Materials Vault

At the time of the PNL measurements, the Building 3100 vault contained approximately 75 drums of 

241Am-oxide. The drums were stacked two high and two deep in two rows. Characterization 

measurements were performed at approximately the midpoint of the rows at a distance of 

approximately 84 cm. 

Figure 6-3 shows the neutron spectrum determined for the 241Am-oxide drums in terms of dose 

equivalent rate. Table 6-5 lists the contribution to the total dose equivalent from the four IREP 

neutron energy groups of interest for ORNL workers. 

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 30 of 92

Figure 6-3. Neutron energy spectrum measured by PNL for drums of 

241Am-oxide in Building 3100 [7]. 

Approximately 80% of the neutron dose equivalent measured by PNL for the drums of 241Am-oxide 

was from neutrons having energies above the 500-keV cutoff for NTA film. NTA film should, 

therefore, have provided reasonable results if used for personnel monitoring in similar spectra.

TLD irradiations performed at the same location as the spectral characterization indicated a 

neutron-to-gamma dose ratio of approximately 1.2. 

Table 6-5. Dose equivalent fractions for 

IREP neutron energy groups for drums of 

241Am-oxide in Building 3100 [8]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 5.4

0.01–0.1 2.4

0.1–2.0 70.4

2.0–20.0 21.9

Favorable to claimant values

0.1–2.0 78

2.0–20.0 22

6.3.4.2.2.3 Building 7910: High Flux Isotope Reactor

The only location at the HFIR where personnel could have been exposed to neutron radiation is the 

beam room. This room contains four stations (HB-1 through HB-4) where intense beams of nearly 

monoenergetic thermal neutrons can be directed. The thermal neutron beams are very well defined in 

terms of both energy and physical size. Neutron energies between approximately 0.02 and 0.08 eV 

can be selected. Personnel exposures to neutrons are the result of scatter from target materials 

being irradiated in the thermal neutron beams. 

Neutron spectral characterizations in the HFIR beam room are unnecessary given the incident 

neutron energies. In terms of the IREP neutron energy groups, neutron exposures at HFIR should be 

assigned to the less-than-0.01-MeV group. NOTE: NTA film would show no response to neutrons in 

this energy range. 

Two sets of TLD irradiations were performed with the HFIR operating at its maximum power level of 

85 MW. One irradiation was at a distance of 54 cm from a 7

Li target in a 0.020-eV beam, and the 

other was at a distance of 67 cm from a nickel target in a 0.050-eV beam. Both indicated 

neutron-to-gamma dose ratios of approximately 0.2 (MMES 1992). 

6.3.4.2.2.4 Building 6000: Holifield Heavy Ion Research Facility

The Holifield facility contains two accelerators. One of the accelerators is the Oak Ridge Isochronous 

Cyclotron (ORIC), which occupied Building 6000 before construction of the Holifield facility. The 

ORIC began operation in 1962. Following construction of the Holifield facility in the mid-1970s, the 

ORIC served as an ion source for the other accelerator, a 25-MV tandem Van de Graaff. The two 

accelerators can be operated in a number of different configurations that result in a variety of 

workplace exposure conditions. Nuclei are accelerated to very high energies and directed into target 

materials in any of several experiment rooms. However, personnel are not allowed into the 

experiment rooms unless dose rates (as determined by area monitors) are very low. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 31 of 92

The operating conditions at the Holifield facility at the time PNL performed its neutron dosimetry 

characterizations did not allow the performance of neutron spectral measurements. Dose 

reconstructors should, therefore, assign neutron exposures at the Holifield facility to the favorable to 

claimant default 0.10-to-2.0-MeV energy group. 

TLD exposure measurements were performed under conditions resulting in high-energy neutrons in 

one of the experiment rooms so that conservative estimates of TLD response per unit dose equivalent 

would be obtained. Specifically, exposures were made with the two accelerators running in tandem 

using a broad-range magnet and carbon target to produce 16O and 17O nucleons with energies of 375 

and 353 MeV, respectively (MMES 1992). These irradiations indicated a neutron-to-gamma dose 

ratio of approximately 19. 

6.3.4.2.2.5 Building 6010: Oak Ridge Electron Linear Accelerator

The ORELA is in the basement of Building 6010. It began operation in the late 1960s and is an 

intense, pulsed neutron source for time-of-flight experiments. Neutron energies produced by the 

ORELA range from thermal to beyond 50 MeV (MMES 1992). There is no personnel access to the 

target room or the accelerator room while the ORELA is in operation. Personnel neutron exposures 

result from neutron back-scattering from the target down the length of the beam line and out the two 

entry mazes. The PNL measurements were at the door to the west entry maze, which is the closest 

personnel can get to the accelerator during operation. 

The measurements were made with the accelerator operating at approximately 500 W and a rate of 

approximately 13.5 pulses/second. The measurement location was approximately 23 cm from the 

grated steel door at the west entrance to the accelerator room at a height of about 1 m. 

The neutron spectrum present at the measurement location was characterized using a 3

He 

proportional counter. The data from the proportional counter were subsequently rebinned to the same 

neutron energy group structure used in unfolding the Bonner sphere results for consistency. 

Figure 6-4 shows these data in terms of dose equivalent rate. Table 6-6 lists the contribution to the 

total dose equivalent from the four IREP neutron energy groups of interest for ORNL workers. 

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-4. Neutron energy spectrum at the west entrance to the 

ORELA room [9]. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 32 of 92

Approximately 65% of the neutron dose equivalent measured by PNL at the west entrance to the 

ORELA room was from neutrons with energies above the 500-keV cutoff for NTA film. NTA film 

would, therefore, have underestimated neutron dose if it was used for personnel monitoring in similar 

spectra. 

TLD irradiations performed at the same location as the spectral characterization indicated a 

neutron-to-gamma dose ratio of approximately 2.5. 

Table 6-6. Dose equivalent fractions for 

IREP neutron energy groups for the west 

entrance to the ORELA room [10]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 13.6

0.01–0.1 0.1

0.1–2.0 85.6

2.0–20.0 0.7

Favorable to claimant values

<0.01 14

0.1–2.0 86

6.3.4.2.2.6 Building 7735: Radiation Calibration Laboratory

The ORNL RADCAL includes a neutron irradiation room that shares walls with other irradiation rooms 

and storage areas. Characterization measurements were performed in areas accessible to personnel 

with the 252Cf neutron sources exposed. Personnel are not allowed in the neutron irradiation room 

when a source is in use. However, an additional characterization was performed with one of the 

neutron sources raised to near the surface of the source storage well. The sources are stored in a 

water-filled vessel 4 ft in diameter and 4 ft deep when not in use. There were two 252Cf sources in use 

at RADCAL when the characterization measurements were performed. One source, of approximately 

1,741 mg, was used primarily in a moderated configuration. The other, of approximately 573 mg, was 

used primarily in a bare configuration. 

Five characterizations of workplace neutron exposure conditions were performed at RADCAL. Three 

of these included both spectral characterizations and TLD irradiations, one consisted of TLD 

irradiations only, and another of spectral characterization only. 

Two sets of characterization measurements were made in the beta/X-ray irradiation room: One with 

the moderated 252Cf source exposed and the other with the bare source exposed. Both sets of 

measurements were performed at a distance of approximately 62 cm from the wall shared between 

the beta/X-ray room and the neutron room. Figures 6-5 and 6-6 show the neutron spectra measured 

at this location for the bare and moderated sources, respectively. 

Tables 6-7 and 6-8 list the contributions to the total dose equivalent from the four IREP neutron 

energy groups of interest for ORNL workers for the bare and moderated sources, respectively. 

A spectral characterization was performed in the hallway outside the neutron irradiation room at a 

distance of approximately 30 cm from the shielded entry door. These measurements were performed 

with the moderated 252Cf source only. Figure 6-7 and Table 6-9 indicate the spectrum and 

contribution to the total dose equivalent from the four IREP neutron energy groups of interest, 

respectively. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 33 of 92

A spectral characterization was performed in the RADCAL equipment room approximately 66 cm from 

the wall it shares with the neutron irradiation room. These measurements were performed with the 

moderated 252Cf source only. Figure 6-8 and Table 6-10 indicate the spectrum and contribution to the 

total dose equivalent from the four IREP neutron energy groups of interest, respectively. 

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-5. Neutron energy spectrum in the RADCAL beta/X-ray 

irradiation room with a bare 252Cf source exposed in the adjacent 

neutron irradiation room [11]. 

1.00E-15

1.00E-13

1.00E-11

1.00E-09

1.00E-07

1.00E-05

1.00E-03

1.00E-01

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-6. Neutron energy spectrum in the RADCAL beta/X-ray 

irradiation room with a moderated 252Cf source exposed in the 

adjacent neutron irradiation room [12]. 

TLD irradiations were performed in the beta/X-ray irradiation room in the same locations as the 

spectral characterizations discussed above using both the bare and moderated 252Cf sources. TLD 

irradiations were also performed in the hallway outside the neutron irradiation room with the 

moderated source exposed. TLD irradiations were not performed in the equipment room. A fourth set 

of TLD irradiations was performed in the neutron irradiation room itself at the edge of the neutron 

source storage well. TLDs were exposed at a height of approximately 56 cm above the surface of the 

water with one of the 252Cf sources raised to approximately 25 cm below the surface. Thus, four sets 

of TLD irradiation data were available from which approximate neutron-to-gamma dose ratios could 

be obtained. Table 6-11 summarizes these values. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 34 of 92

The data in Table 6-12 show that NTA film might have provided a reasonable estimation of personnel 

neutron exposure in fields similar to that near the wall of the beta/X-ray irradiation room with the bare 

Table 6-7. Dose equivalent fractions for 

IREP neutron energy groups for the 

RADCAL beta/X-ray irradiation room with a 

bare 252Cf source exposed in the adjacent 

neutron irradiation room [13]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 7.0

0.01–0.1 2.0

0.1–2.0 60.3

2.0–20.0 30.7

Favorable to claimant values

0.1–2.0 69

2.0–20.0 31

Table 6-8. Dose equivalent fractions for 

IREP neutron energy groups for the 

RADCAL beta/X-ray irradiation room with a 

moderated 252Cf source exposed in the 

adjacent neutron irradiation room [14]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 16.4

0.01–0.1 0.0

0.1–2.0 83.6

2.0–20.0 0.0

Favorable to claimant values

<0.01 16

0.1–2.0 84

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-7. Neutron energy spectrum in the hallway outside the 

RADCAL neutron irradiation room with a moderated 252Cf source 

exposed [15]. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 35 of 92

Table 6-9. Dose equivalent fractions for 

IREP neutron energy groups for the hallway 

outside the RADCAL neutron irradiation 

room with a moderated 252Cf source exposed 

[16]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 33.2

0.01–0.1 9.9

0.1–2.0 42.6

2.0–20.0 14.3

Favorable to claimant values

<0.01 33

0.1–2.0 53

2.0–20.0 14

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-8. Neutron energy spectrum in the RADCAL equipment 

room with a moderated 252Cf source exposed in the adjacent neutron 

irradiation room [17]. 

source in use, but would have underestimated exposures in fields similar to those resulting from the 

moderated source—substantially so in one case. 

Table 6-12 summarizes the fraction of the total dose equivalent measured at locations in RADCAL as 

an element of neutron spectral characterizations that was due to neutrons with energies above the 

500-keV cutoff for NTA film. This information provides a basis for estimating the performance of NTA 

film in similar neutron fields. NTA film was never used for personnel dosimetry at RADCAL. 

6.3.4.2.2.7 Building 2007: Radiation Standards and Calibration Laboratory

At the time of the PNL characterization measurements (October 1989), the RASCAL modern, 

low-scatter facility was not yet in use. Measurements were performed to characterize the neutron 

field experienced by instrument calibration technicians as they performed calibrations using 238Pu:Be 

neutron sources. Specifically, the measurements were in Room 102 at 130 cm from the source. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 36 of 92

Table 6-10. Dose equivalent fractions for 

IREP neutron energy groups for the 

RADCAL equipment room with a moderated 

252Cf source exposed in the adjacent neutron 

irradiation room [18]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 25.0

0.01–0.1 3.0

0.1–2.0 53.9

2.0–20.0 18.2

Favorable to claimant values

<0.01 25

0.1–2.0 57

2.0–20.0 18

Table 6-11. Neutron-to-gamma dose ratios for several locations in RADCAL (MMES 1992). 

Location 252Cf source configuration

Approx. 1n:γ 

dose ratio

Beta/X-ray irradiation room at wall shared with neutron room Bare 15.6

Beta/X-ray irradiation room at wall shared with neutron room Moderated 3.6

Hallway outside neutron room 30 cm from entry door Moderated 2.1

Edge of source storage well, 56 cm above water 25 cm below surface of storage well 1.2

Table 6-12. Fraction of dose equivalent above NTA cutoff for several locations in RADCAL [19]. 

Location

252Cf source 

configuration

Fraction of dose equivalent 

above 500 keV

Beta/X-ray irradiation room at wall shared with neutron room Bare 82%

Beta/X-ray irradiation room at wall shared with neutron room Moderated 76%

Hallway outside neutron room 30 cm from entry door Moderated 29%

Equipment room at wall shared with neutron room Moderated 61%

Figure 6-9 and Table 6-13 indicate the neutron spectrum and contribution to the total dose equivalent 

from the four IREP neutron energy groups of interest, respectively. These data were determined for 

the 238Pu:Be neutron source used for instrument calibrations at RASCAL before use of the new 

low-scatter facility about 1991. 

Approximately 94% of the neutron dose equivalent measured by PNL for the RASCAL Pu:Be source 

was due to neutrons with energies above the 500-keV cutoff for NTA film. NTA film should, therefore, 

have provided reliable results if used for personnel monitoring in similar spectra. 

TLD irradiations performed at the same location as the spectral characterization showed a 

neutron-to-gamma dose ratio of approximately 28.9. 

6.3.4.2.2.8 Buildings 7920 and 7930: Radiochemical Engineering Development Center

The mission of the REDC is to separate and purify transuranic isotopes (primarily 252Cf) from 244Cm 

target material that has been irradiated in the HFIR. The REDC consists of two facilities: 

Building 7920 and Building 7930. Both buildings contain hot cells and gloveboxes for working with 

irradiated targets and separated products. The separation and purification processes occur in 

Building 7920. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 37 of 92

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE Rate (mrem/hr)

Figure 6-9. Neutron energy spectrum in RASCAL Room 102 at 130 

cm from the 238Pu:Be neutron calibration source [20]. 

Table 6-13. Dose equivalent fractions for 

IREP neutron energy groups for RASCAL 

Room 102 at 130 cm from the 238Pu:Be 

neutron calibration source [21]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 1.3

0.01–0.1 0.8

0.1–2.0 26.4

2.0–20.0 71.5

Favorable to claimant values

0.1–2.0 29

2.0–20.0 71

Products are prepared to customer specifications and packaged in Building 7930. Elements handled 

in the REDC include californium, berkelium, einsteinium, and fermium. 

Given its importance as a source of personnel neutron exposures, a number of characterization 

measurements were performed in the REDC. ORNL subsequently defined different groups of neutron 

workers for assigning appropriate TLD correction factors based on their duties. Some worker groups 

were defined by determining occupancy factors for individuals for different areas in the REDC and 

using these in conjunction with characterization data to establish weighted average TLD correction 

factors. For other groups, conservative assumptions were made to essentially bound the radiation 

fields to which they could be exposed, and then a 100% occupancy factor was used to compute any 

neutron dose. 

This section presents neutron spectra determined for five locations in the REDC, followed by a 

summary of neutron-to-gamma dose ratios estimated from TLD irradiation data from 11 locations. 

The fraction of the total dose equivalent from neutrons having energies above the 500-keV cutoff for 

NTA film are given for the five spectral measurements. 

Neutron spectra were measured for five locations in REDC Building 7920: 

• At the face of a glovebox in Room 111
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 38 of 92

• At the side of a glovebox in Room 211

• At the viewing window of the Transuranium Decontamination Facility (TDF)

• Next to a waste cask in the Limited Access Area

• At the exit of the Waste Transfer Area (WTA) transfer tunnel

At the time of the PNL measurements, Room 111 (the Shielded Cave Area) contained two shielded 

hot cells and two gloveboxes. Most of the neutron dose to personnel in this area came from the 

gloveboxes. It was estimated that less than 10% of neutron dose to workers in this area came from 

hot cell operations (MMES 1992). The neutron spectrum was measured at about 37 cm from the face 

of a glovebox containing waste and general contamination consisting primarily of 252Cf. The spectrum 

was, therefore, expected to be representative of that when 252Cf sources were present. The 

measurement was performed at a height of 1.25 m above the floor. Figure 6-10 and Table 6-14 

indicate the neutron spectrum and contribution to the total dose equivalent from the four IREP neutron 

energy groups of interest for this location, respectively. 

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-10. Neutron energy spectrum measured by PNL in REDC 

Room 111 at 37 cm from a glovebox containing 252Cf [22]. 

Table 6-14. Dose equivalent fractions for 

IREP neutron energy groups for REDC 

Room 111 at 37 cm from a glovebox 

containing 252Cf [23]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 2.9

0.01–0.1 0.7

0.1–2.0 72.3

2.0–20.0 24.1

Favorable to claimant values

0.1–2.0 76

2.0–20.0 24

REDC Room 211 (the Alpha Laboratory) contained several banks of gloveboxes used for 

radiochemical procedures. The neutron spectrum was measured in a walkway at the end of the 

glovebox banks between the side of one of the boxes and the wall. The location was approximately 

32 cm from the side of the box. None of the gloveboxes was in use at the time, but all were 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 39 of 92

contaminated with transuranic material to some degree. The box nearest the measurement location 

held a small container of waste, thought to be primarily einsteinium. Figure 6-11 and Table 6-15 

indicate the neutron spectrum and contribution to the total dose equivalent from the four IREP neutron 

energy groups of interest for this location, respectively. 

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E+02

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-11. Neutron energy spectrum in REDC Room 211 at 32 cm 

from the end of a glovebox bank [24]. 

Table 6-15. Dose equivalent fractions for 

IREP neutron energy groups for REDC 

Room 211 at 32 cm from the end of a 

glovebox bank [25]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 1.6

0.01–0.1 0.3

0.1–2.0 73.2

2.0–20.0 24.8

Favorable to claimant values

0.1–2.0 75

2.0–20.0 25

Measurements were made to characterize the neutron field experienced by personnel using the TDF, 

which is a movable hot cell on top of the Building 7920 hot cell bank. The TDF is a means of moving 

materials out of hot cells and into other containers. Sources or waste can be transferred into the TDF, 

when it is in position, through its bottom access port. Items are decontaminated as necessary in the 

TDF before being moved to other containers, and measurements to verify source strength can be 

performed in the TDF. The inner cavity of the TDF is 3 ft by 4 ft by 4 ft high. It is shielded on all sides 

by 3 ft of water. One face of the TDF contains a viewing window, which is where the PNL 

characterization measurements occurred. The measurement location was 33 cm from the window. A 

12-mg 252Cf source was placed on a shelf in the center of the cavity to facilitate the measurement. 

Figure 6-12 and Table 6-16 indicate the neutron spectrum and contribution to the total dose 

equivalent from the four IREP neutron energy groups of interest for this location, respectively.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 40 of 92

1.00E-17

1.00E-15

1.00E-13

1.00E-11

1.00E-09

1.00E-07

1.00E-05

1.00E-03

1.00E-01

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-12. Neutron energy spectrum for the REDC TDF containing 

12 mg 252Cf [26]. 

Table 6-16. Dose equivalent fractions for 

IREP neutron energy groups for the REDC 

TDF containing 12 mg 252Cf [27]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 12.2

0.01–0.1 0.0

0.1–2.0 87.8

2.0–20.0 0.0

Favorable to claimant values

<0.01 12

0.1–2.0 88

A waste cask on a concrete pad in the Building 7920 Limited Access Area is a receptacle for waste 

generated in the process cubicles. Waste is placed in the cask as necessary by removing the cask 

cover using an overhead crane. The waste casks have walls of 24 cm of reinforced concrete plus 

10 cm of armor-plate steel. Characterization measurements were performed approximately 18 cm 

from the wall of the cask at a location where the neutron dose rate was fairly uniform. The location 

was approximately 1 m above the support pad. Figure 6-13 and Table 6-17 indicate the neutron 

spectrum and contribution to the total dose equivalent from the four IREP neutron energy groups of 

interest for this location, respectively.

Waste moves into the REDC WTA through a square tunnel that passes through the wall of process 

cubicle Number 10. The dimensions of the tunnel are 18 by 18 in. A metal table is attached to the 

outer wall of the cubicle just below the exit of the tunnel. Measurements were made directly in front of 

the tunnel exit at the edge of this table. This placed the measurement location approximately 58 cm 

from the tunnel exit and 83 cm above the floor. There were no sources present in the tunnel or in the 

WTA at the time of the measurements. The intent was to characterize the neutron field from 

contamination present on the tunnel walls. Figure 6-14 and Table 6-18 indicate the neutron spectrum 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 41 of 92

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-13. Neutron energy spectrum for a waste cask in the REDC 

Limited Access Area [28]. 

Table 6-17. Dose equivalent fractions for 

IREP neutron energy groups for a waste 

cask in the REDC Limited Access Area [29]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 13.8

0.01–0.1 2.7

0.1–2.0 56.6

2.0–20.0 26.9

Favorable to claimant values

<0.01 14

0.1–2.0 59

2.0–20.0 27

and contribution to the total dose equivalent from the four IREP neutron energy groups of interest for 

this location, respectively. 

TLD irradiations were performed at all five of the locations discussed above where neutron spectral 

characterizations were made and at an additional six locations in the REDC where spectral 

characterizations were not performed:

• The face of a glovebox in Building 7920, Room 211

• Approximately 4.6 m behind the glovebox bank in Building 7920, Room 211

• The face of the WTA glovebox

• The face of a glovebox in Building 7930 in a low-scatter geometry

• The face of a glovebox in Building 7920, Room 208 

• The low-dose waiting area in Building 7920, Room 208

TLD irradiations in Room 211 were performed roughly 2 years after the measurements to characterize 

the neutron spectrum in 1989. These follow-up characterizations were in July 1991. The gloveboxes 

still contained transcalifornium waste, however. One irradiation was performed approximately 22 cm 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 42 of 92

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-14. Neutron energy spectrum for contamination on the walls 

of the REDC WTA transfer tunnel [30]. 

Table 6-18. Dose equivalent fractions for 

IREP neutron energy groups for 

contamination on the walls of the REDC 

WTA transfer tunnel [31]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 2.5

0.01–0.1 1.0

0.1–2.0 90.9

2.0–20.0 5.6

Favorable to claimant values

0.1–2.0 100

from the face of the same glovebox where the neutron spectrum was measured from the side in 1989. 

This irradiation was performed at a height of approximately 1 m off the floor. A second irradiation was 

performed at a location approximately 4.6 m behind this glovebox, also at a height of approximately 

1 m. 

An irradiation was performed 28 cm from the face of the glovebox in the WTA. This glovebox 

contains a rotary table for moving items into it from a small hot cell. A 252Cf source was placed in the 

glovebox to facilitate the irradiations. 

TLDs were irradiated at a distance of 76 cm from a glovebox in an open floor area in Building 7930. 

The location was selected because of its low-scatter geometry. The glovebox was in an area with no 

structure or other objects in its vicinity. The nearest wall was 3 m away, the roof was 4.6 m above, 

and the bottom of the box was more than 1 m off the floor. The glovebox was internally contaminated 

and read about 4 mrem/hr with no sources present. The TLD irradiations were performed with two 

252Cf sources in the glovebox (in addition to the contamination mix). 

Room 208 of the 7920 building is used to prepare samples for analysis. At the time of the PNL 

measurements, it contained a row of gloveboxes, plus bench tops and a fume hood. Glovebox faces 

were shielded with a layer of Lucite in addition to their inherent shielding. TLDs were irradiated at 

10 cm from the Lucite shield at the face of a glovebox containing californium and curium. This 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 43 of 92

exposure was performed at a height of 1.1 m above the floor. TLDs were exposed in the same room 

in a low-dose area where workers were asked to stand during waiting periods. 

Table 6-19 summarizes neutron-to-gamma dose ratios indicated by the TLD irradiations performed in 

various workplace radiation fields in the REDC. 

Table 6-19. Neutron-to-gamma dose ratios for several locations in the 

REDC (MMES 1992). 

Location Approx. 1n:γ dose ratio

Face of glovebox in Building 7920, Room 111 5.8

Side of glovebox in Building 7920, Room 211 11.7

Face of glovebox in Building 7920, Room 211 5.3

4.6 m behind glovebox in Building 7920, Room 211 1.3

TDF viewing window (Building 7920) 0.2

Limited Access Area waste cask (Building 7920) 4.4

WTA transfer tunnel exit (Building 7920) 6.7

WTA glovebox (Building 7920) 1.2

Low-scatter glovebox in Building 7930 16.7

Face of glovebox in Building 7920, Room 208 1.3

Low-dose waiting area in Building 7920, Room 208 1.3

Table 6-20 summarizes the fraction of the total dose equivalent measured at locations in the REDC as 

an element of neutron spectral characterizations that was from neutrons having energies above the 

500-keV cutoff for NTA film. The data in Table 6-20 show that NTA film should have provided a 

reasonable estimate of personnel neutron exposure in fields similar to those for glovebox workers in 

Rooms 111 and 211 and for the exit of the WTA transfer tunnel. Exposures from fields similar to that 

near the wall of the waste cask would have been underestimated. NTA film would not have shown 

any response in fields comparable to that associated with the TDF. 

Table 6-20. Fraction of dose equivalent above NTA cutoff for several 

locations in the REDC [32]. 

Location

Fraction of dose equivalent 

above 500 keV (%)

Face of glovebox in Building 7920, Room 111 91

Side of glovebox in Building 7920, Room 211 94

TDF viewing window (Building 7920) 0

Limited Access Area waste cask (Building 7920) 72

WTA transfer tunnel exit (Building 7920) 87

6.3.4.2.2.9 Solid Waste Storage Area 5

In the late 1980s, there were two areas in SWSA 5 where personnel neutron exposures could occur. 

The first was the Waste Examination and Assay Facility (WEAF; Building 7824) where the waste 

assay equipment in use included a small deuterium-tritium (D-T) accelerator. The other was the TRU 

waste storage facility, which is a series of four concrete bunkers approximately 30 m behind the 

WEAF. At the time of the PNL measurements, three of the four bunkers had been filled and sealed. 

The fourth (northernmost) bunker was still in use and contained three casks of TRU waste. The casks 

contained californium and transcalifornium wastes, including berkelium and einsteinium, presumably 

from the REDC. 

Characterization measurements were not attempted for the WEAF D-T accelerator, because the 

operating time required to get a reliable measurement would have significantly depleted the tritium 

target. Instead, WEAF workers were assumed to spend 10% of their time in the neutron field from the 

D-T accelerator. These exposures were evaluated using the TLD correction factors determined for 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 44 of 92

the Holifield facility, a very conservative choice (MMES 1992). WEAF workers were assumed to 

spend the other 90% of their time in the field associated with the three waste casks in the storage 

bunker. 

Two neutron spectral characterization measurements were made in the SWSA 5 TRU storage bunker. 

One was at 82 cm from the center cask in a row of three at a height of 1 m. The other was at 

approximately 41 cm from the wall shared with the adjacent bunker. This measurement was 

performed 4 m along the wall from the bunker entrance at a height of just over 1 m. This location was 

5.7 m from the nearest of the three waste casks. Figure 6-15 and Table 6-21 indicate the neutron 

spectrum and contribution to the total dose equivalent from the four IREP neutron energy groups of 

interest for the waste casks, respectively. Figure 6-16 and Table 6-22 indicate spectrum and 

contribution results for the measurement along the bunker wall, respectively. 

TLD irradiations performed at the same location as the spectral characterization for the three waste 

casks indicated a neutron-to-gamma dose ratio of approximately 2.8. TLD irradiations were not 

performed at the location near the shared wall. 

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E+02

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-15. Neutron energy spectrum for waste casks in the SWSA 

5 TRU storage bunker [33]. 

The fraction of the total dose equivalents measured for the waste casks and along the shared wall in 

the SWSA 5 TRU storage bunker that were from neutrons with energies above the 500-keV cutoff for 

NTA film were 68% and 65%, respectively. Thus, NTA film would have underestimated neutron 

exposures if used for personnel monitoring in similar fields. 

6.3.4.2.3 Data from Neutron Dosimetry Records

The purpose of the characterization measurements performed by PNL at ORNL between 1989 and 

1991 was to establish correction factors needed to compute neutron dose equivalent from the 

response of the neutron-sensitive elements of the CEDS albedo neutron dosimeters. The individual 

characterization measurements resulted in individual values of these TLD correction factors. These 

Table 6-21. Dose equivalent fractions for 

IREP neutron energy groups for waste casks 

in the SWSA 5 TRU storage bunker [34]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 45 of 92

Measured values

<0.01 12.4

0.01–0.1 2.9

0.1–2.0 66.5

2.0–20.0 18.2

Favorable to claimant values

<0.01 12

0.1–2.0 70

2.0–20.0 18

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E-07 1.00E-05 1.00E-03 1.00E-01 1.00E+01

Energy (MeV)

DE rate (mrem/hr)

Figure 6-16. Neutron energy spectrum at a wall shared between two 

adjacent TRU storage bunkers at SWSA 5 [35]. 

Table 6-22. Dose equivalent fractions for 

IREP neutron energy groups at a wall shared 

between two adjacent TRU storage bunkers 

at SWSA 5 [36]. 

IREP neutron energy 

group (MeV)

Dose equivalent 

fraction (%)

Measured values

<0.01 13.2

0.01–0.1 2.3

0.1–2.0 48.6

2.0–20.0 35.9

Favorable to claimant values

<0.01 13

0.1–2.0 51

2.0–20.0 36

factors, and their associated neutron-to-photon dose ratios, are conservative and consider only 

photon exposure associated with the neutron-photon field. They are single measurements and do not 

provide information on the distribution of neutron dose or neutron-to-photon dose ratios. Therefore, 

neutron dosimetry results were obtained from ORNL for its various neutron workers groups to get 

additional information on neutron exposures and neutron-to-photon dose ratios for different facilities 

and job duties. 

ORNL defines a number of different neutron groups, so appropriate TLD response-to-dose-equivalent 

correction factors can be applied when evaluating an individual’s neutron dose. Correction factors for 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 46 of 92

each group are determined based on worker’s occupancy in different workplace neutron fields. There 

are currently 22 groups defined, 19 of which can be used to evaluate dose for ORNL workers. Some 

groups are used for ORNL facilities exclusively, and others are used for workers performing duties at 

ORNL or Y-12. There are also groups defined exclusively for other sites, such as Portsmouth and 

Y-12 (ORNL 1994). Some groups are defined based on an assumption of 100% occupancy in a 

specific neutron field. Others use weighted averages of occupancy in two or more fields. Occupancy 

in this context refers to time spent in the neutron field and not necessarily total time spent in a 

radiation area. 

The ORNL external dosimetry group queried its database of personnel dosimetry records to produce 

a set of quarterly neutron and photon dose results for all of the various neutron worker groups. The 

query returned 20,037 records covering the period from 1990 through the second quarter of 2004. 

Each record consisted of five fields: Calendar year, calendar quarter, neutron group, neutron dose, 

and gamma dose. ORNL determined the gamma dose values from the response of the TLD-700 chip 

below 300 mg/cm2 of plastic. The gamma dose values were determined from the neutron dosimeter’s 

response rather than the gamma dose result from the worker’s beta-gamma dosimetry because 

workers are sometimes issued more than one neutron dosimeter during a quarterly monitoring period. 

This occurs when a neutron worker’s duties involve neutron exposure corresponding to more than one 

of the defined neutron worker groups. It can also occur when workers have interim dose evaluations 

performed at the request of health physics staff because higher dose is anticipated. The photon dose 

results determined from the neutron dosimeters should still represent the individual’s total gamma 

dose (and not just that associated with neutron fields) because ORNL requires both dosimeters be 

worn regardless of the work being done. Therefore, the neutron-to-photon dose ratios derived from 

the dosimetry records (see Attachment D) are relative to total photon dose rather than just that 

received in mixed neutron-photon fields. These values differ from those derived from the PNL 

characterization data (see Attachment C) in that they inherently account for the workers’ occupancy 

and thus can be applied directly to an individual’s reported photon dose. 

For neutron groups for which there were an adequate number of positive results, the dosimetry data 

obtained from ORNL were used to establish distributions of neutron dose and neutron-to-photon dose 

ratios. Typical values (rather than distributions) were determined from the dosimetry data for groups 

for which there were only a few results. Summary statistics, such as the fraction of total results that 

showed a positive neutron dose, minimum and maximum dose values, and so forth were determined 

for each group for which data were available. This analysis used measured TLD results for dosimetry 

data. The results do not include doses assigned through such other means as investigations or 

estimations. The reporting limit for quarterly neutron dose is 10 mrem. Photon dose values down to 1 

mrem (net) were computed, but values below 10 mrem (above background) were not used. 

It should be noted that of the 20,037 neutron dosimetry records obtained for the period from 1990 

through the second quarter of 2004, 16,785 of them (84%) showed a reported neutron dose of zero, 

which means the individual’s neutron dose evaluated to less than 10 mrem. Researchers at HFIR 

account for 40% of the records. For these individuals, only 5 out of 7992 (less than 0.1%) had an 

assigned neutron dose in excess of the reporting limit (the highest of the five was 14 mrem). 

The sections that follow discuss the results obtained from the dosimetry data provided by ORNL for 

the various neutron groups. Comparisons between the neutron-to-photon dose ratios derived from 

the personnel dosimetry records and those from the PNL characterization measurements are 

discussed where applicable. For the ratios derived from the dosimetry records, in general the 

comparisons are made relative to the geometric mean of the distribution of these values. 

Attachment D summarizes the neutron-to-photon dose ratios for each group, giving the geometric 

mean and geometric standard deviation (GSD) for the groups for which there were a sufficient number 

of results to establish a distribution. Minimum and maximum values are given. Attachment E 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 47 of 92

summarizes neutron dose information, also giving the geometric mean and GSD for the groups for 

which there were a sufficient number of results to establish a distribution. The maximum neutron 

dose and the fraction of the total number of results that were greater than or equal to the 10-mrem 

reporting level are also given. In general, dose reconstructors should use the information provided in 

Attachment D instead of that in Attachment C, as practical, because the data in Attachment D are 

expressed as distributions and account for occupancy in all photon fields. In most cases, the 

neutron-to-photon dose ratios derived from the personnel dosimetry records are significantly lower 

than those implied by the characterization measurements performed by PNL. Exceptions are ratios 

for the TDF used by REDC workers (Group 17 below) and the HFIR (Group 5). The

neutron-to-photon dose ratios indicated by the dosimetry records for TDF users were significantly 

higher than that implied by the PNL characterization, and the values for the HFIR were largely 

comparable. However, there were very few positive dose results among the records for HFIR 

personnel to use to compute a neutron-to-photon dose ratio.

6.3.4.2.3.1 REDC Workers: Groups 2, 3, 4, and 17

ORNL defines the following neutron groups for REDC workers:

• Group 2: Analytical chemistry (Building 7920 glovebox workers)

• Group 3: Packaging and shipping workers in Building 7930

• Group 4: REDC general area

• Group 17: TDF workers

REDC analytical chemistry (Group 2) workers prepare and analyze samples of materials at various 

stages of the production process. Most of this work takes place in gloveboxes, so ORNL makes the 

assumption that 100% of any neutron exposure for these individuals is from glovebox work. To be 

conservative, the TLD correction factors used are those determined for the glovebox in Room 111 

rather than those for the box in Room 211. 

Group 2 workers accounted for 1,535 (8%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 1,535 records, 1,197 (78%) showed a neutron dose greater than or equal to the 10-

mrem reporting limit. There were 1,104 records for which both the neutron and photon dose were 

positive. Figure 6-17 is a normal probability plot showing the distribution of the neutron-to-photon 

dose ratios determined from these 1,104 results. The ratios are approximately lognormally distributed 

with a geometric mean of 1.6 and a GSD of 2.2. They range from 0.1 to 8.8. 

y = 1.591e0.765x

R2 = 0.9643

0.1

1.0

10.0

100.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Dose Ratio
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 48 of 92

Figure 6-17. Normal probability plot for Group 2 neutron-to-photon 

dose ratios [37]. 

Figure 6-18 is a normal probability plot of the 1,197 positive neutron dose values for Group 2 workers 

in the period from 1990 through the second quarter of 2004. These data are approximately 

lognormally distributed with a geometric mean of 50 mrem and a GSD of 2.2. They range from 10 to 

584 mrem. 

Workers in the general area of the REDC include control room staff and office and support personnel 

who normally do not work with product materials. These individuals are assigned to Neutron Group 4. 

In terms of occupancy in neutron fields, Group 4 workers are assumed to spend 60% of their time in 

the control room, 20% in the Limited Access Area behind the hot cells, and 20% in the WTA (MMES 

y = 50.433e0.7771x

R2

 = 0.9911

10

100

1000

-4 -2 0 2 4

Normal Quantiles

Neutron Dose (mrem)

Figure 6-18. Normal probability plot for Group 2 neutron doses [38].

1992). The neutron spectrum assumed for the control room was a 50:50 combination of that 

determined for the TDF (as a proxy for the shielded hot cells) and that for the glovebox in Room 111 

(as a proxy for the gloveboxes in Room 109). No spectral measurements were possible in the REDC 

control room because the neutron dose rate is so low (about 0.1 mrem/hr). 

Group 4 workers accounted for 2,730 (14%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 2,730 records, 1,070 (39%) showed a neutron dose greater than or equal to the 10-

mrem reporting limit. There were 982 records for which both the neutron and photon dose were 

positive. Figure 6-19 is a normal probability plot showing the distribution of the neutron-to-photon 

dose ratios determined from these 982 results. The ratios are approximately lognormally distributed 

with a geometric mean of 1.1 and a GSD of 1.7. They range from 0.1 to 3.7.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 49 of 92

y = 1.0742e0.4976x

R2

 = 0.9661

0.1

1.0

10.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratio

Figure 6-19. Normal probability plot for Group 4 neutron-to-photon 

dose ratios [39]. 

Figure 6-20 is a normal probability plot of the 1,070 positive neutron dose values for Group 4 workers 

in the period from 1990 through the second quarter of 2004. These data are approximately 

lognormally distributed with a geometric mean of 23 mrem and a GSD of 1.6. They range from 10 to 

91 mrem. 

All of the neutron dose for workers in Building 7930 (Group 3) is assumed to be from the field 

determined for the glovebox in the low-scatter geometry. Similarly, 100% of the neutron dose for 

Group 17 workers using the REDC TDF is assumed to be from the spectrum determined for that unit. 

Group 3 workers accounted for 81 (0.4%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 81 records, 79 (98%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 65 records for which both the neutron and photon dose were positive. 

Figure 6-21 is a normal probability plot showing the distribution of the neutron-to-photon dose ratios 

determined from these 65 results. The ratios are approximately lognormally distributed with a 

geometric mean of 2.5 and a GSD of 1.6. They range from 1.0 to 8.1.

Figure 6-22 is a normal probability plot of the 79 positive neutron dose values for Group 3 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 42 mrem and a GSD of 1.8. They range from 11 to 137 mrem. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 50 of 92

y = 22.839e0.4693x

R2 = 0.9784

10

100

1000

-4 -2 0 2 4

Normal Quantiles

Neutron Dose (mrem)

Figure 6-20. Normal probability plot for Group 4 neutron doses [40]. 

y = 2.5008e0.494x

R2 = 0.9835

0.1

1.0

10.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratio

Figure 6-21. Normal probability plot for Group 3 neutron-tophoton dose ratios [41]. 

Group 17 workers accounted for 264 (1%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 264 records, 111 (42%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 86 records for which both the neutron and photon dose were positive. 

Figure 6-23 is a normal probability plot showing the distribution of the neutron-to-photon dose ratios 

determined from these 86 results. The ratios are approximately lognormally distributed, having a 

geometric mean of 2.1 and a GSD of 2.3. They range from 0.4 to 8.0.

Figure 6-24 is a normal probability plot of the 111 positive neutron dose values for Group 17 workers 

in the period from 1990 through the second quarter of 2004. These data are approximately 

lognormally distributed with a geometric mean of 37 mrem and a GSD of 2.3. They range from 10 to 

230 mrem. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 51 of 92

y = 42.33e0.5853x

R2 = 0.9922

10

100

1000

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-22. Normal probability plot for Group 3 neutron doses [42]. 

y = 2.1112e0.83x

R2

 = 0.9612

0.1

1.0

10.0

100.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratios

Figure 6-23. Normal probability plot for Group 17 neutron-to-photon 

dose ratios [43]. 

Comparison of the neutron-to-photon dose ratios computed from the personnel dosimetry records with 

those derived from the PNL characterization measurements shows the values from the dosimetry data 

are significantly lower for Groups 2, 3, and 4 and significantly higher for Group 17. The lower values 

are expected because the PNL measurements were performed in conservatively chosen locations 

with high neutron dose rates. Under real work conditions, workers do not spend all of their time in 

these locations and the neutron-to-photon dose ratios derived from the dosimetry data inherently 

account for occupancy in all photon fields and not just those associated with neutron sources. 

However, the neutron dose rate at the location where PNL performed its characterization 

measurements for the TDF (Group 17) was much lower than the corresponding photon dose rate. 

That the neutron-to-photon ratio determined from the personnel dosimetry records is much higher 

than that implied by the PNL measurements suggests that workers receive neutron exposures from 

other sources or other geometries where the neutron and photon dose rates are more comparable. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 52 of 92

y = 42.33e0.5853x

R2 = 0.9922

10

100

1000

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-24. Normal probability plot for Group 17 neutron doses [44]. 

Workers assigned to Group 17 are assumed to spend 100% of their time in the neutron spectrum 

determined for the TDF for the purpose of neutron dose evaluation. 

6.3.4.2.3.2 Group 7

Group 7 is a generic group used for personnel working in neutron fields for which specific 

characterization information is not available and exposure from high-energy neutrons is not expected. 

The TLD calibration factors used for dose evaluation for personnel assigned to Group 7 are based on 

a D2O-moderated 252Cf spectrum. 

Group 7 workers accounted for 1,329 (7%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 1,329 records, 37 (3%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 24 records for which both the neutron and photon dose were positive. 

Figure 6-25 is a normal probability plot showing the distribution of the neutron-to-photon dose ratios 

determined from these 24 results. The ratios have a geometric mean of 0.3 and a GSD of 3.2. They 

range from 0.1 to 2.2. The distribution of the Group 7 neutron-to-photon dose ratios appears bimodal, 

but this could be an artifact of the small number of values. The neutron doses associated with the 24 

ratios range from 11 to 27 with the higher ratios associated with the lower doses. The lower ratios are 

associated with statistically better data—the photon doses in particular—so using the geometric mean 

of 0.3 with the large GSD should be valid. 

Figure 6-26 is a normal probability plot of the 37 positive neutron dose values for Group 7 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 14 mrem and a GSD of 1.3. They range from 10 to 27 mrem. 

6.3.4.2.3.3 Unshielded Source Handlers: Group 9

Group 9 is a generic group defined for workers whose duties involve transfers or other operations with 

unshielded isotopic neutron sources. The TLD response-to-neutron-dose conversion factors for 

Group 9 reflect a weighted average of bare Pu:Be and D2O-moderated 252Cf fission spectra in an 

80:20 ratio (ORNL 1994). 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 53 of 92

y = 0.3485e1.1072x

R2 = 0.8929

0.0

0.1

1.0

10.0

-2 -1 0 1 2

Normal Quantiles

Neutron:Photon Ratio

Figure 6-25. Normal probability plot for Group 7 neutron-to-photon 

dose ratios [45]. 

y = 14.236e0.245x

R2 = 0.9104

10

100

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-26. Normal probability plot for Group 7 neutron doses [46]. 

Group 9 workers accounted for 1,687 (8%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 1,687 records, 405 (24%) showed a neutron dose greater than or equal to the 

10-mrem reporting limit. There were 163 records for which both the neutron and photon dose were 

positive. Figure 6-27 is a normal probability plot showing the distribution of the neutron-to-photon 

dose ratios determined from these 163 results. The ratios are approximately lognormally distributed 

with a geometric mean of 1.0 and a GSD of 2.8. They range from 0.1 to 14.1.

Figure 6-28 is a normal probability plot of the 405 positive neutron dose values for Group 9 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 24 mrem and a GSD of 1.8. They range from 10 to 212 mrem. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 54 of 92

y = 1.0037e1.0107x

R2 = 0.9737

0.1

1.0

10.0

100.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratio

Figure 6-27. Normal probability plot for Group 9 neutron-to-photon 

dose ratios [47]. 

y = 23.643e0.5927x

R2

 = 0.9465

10

100

1000

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-28. Normal probability plot for Group 9 neutron doses [48]. 

6.3.4.2.3.4 SWSA 5 workers: Groups 11, 12, and 21

ORNL defines three groups for workers exposed to neutrons from TRU wastes handled at SWSA 5:

• Group 11: Operators and other SWSA 5 personnel whose duties do not include moving 

wastes into the area

• Group 12: Personnel who transfer TRU wastes from other areas to SWSA 5

• Group 21: Operators and health physics staff associated with the transfer of drums into

Building 7879

Personnel assigned to Group 11 for the purpose of neutron dose evaluation could be exposed to 

neutrons from the SWSA 5 TRU waste storage bunker or the intermittent neutron field associated with 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 55 of 92

the small D-T accelerator housed at the WEAF. The TLD response-to-neutron-dose conversion 

factors for Group 11 reflect a weighted average of the spectrum measured for the TRU storage 

bunker and that for the Holifield Heavy Ion Facility in a 90:10 ratio (ORNL 1994). The high-energy 

component associated with the Holifield facility was included for conservatism to account for potential 

exposure from the D-T accelerator. 

Personnel assigned to Group 12 are assumed to receive 100% of their neutron exposure from the 

spectrum measured for the TRU storage bunker, and those assigned to Group 21 are assumed to 

receive 100% of their exposure from the spectrum determined for waste drums handled in Building 

7879. 

Group 11 workers accounted for 448 (2%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 448 records, 38 (8%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were nine records for which both the neutron and photon dose were positive. 

The neutron-to-photon dose ratios associated with these nine records show a geometric mean of 1.1 

and a GSD of 1.9. They range from 0.3 to 2.4. A normal probability plot was not created showing the 

distribution of the neutron-to-photon dose ratios given the small number of data points. 

Figure 6-29 is a normal probability plot of the 38 positive neutron dose values for Group 11 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 14 mrem and a GSD of 1.3. They range from 10 to 26 mrem. 

y = 13.705e0.2563x

R2

 = 0.9283

10

100

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-29. Normal probability plot for Group 11 neutron doses [49]. 

Group 12 workers accounted for 501 (3%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 501 records, 3 (1%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 3 records for which both the neutron and photon dose were positive. The 

neutron-to-photon dose ratios associated with these 3 records are 0.2, 0.7, and 1.1. The measured 

neutron doses associated with these three ratios were 17, 30, and 16 mrem, respectively. 

Group 21 workers accounted for 1,148 (6%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 1,148 records, 160 (14%) showed a neutron dose greater than or equal to the 10-

mrem reporting limit. There were 148 records for which both the neutron and photon dose were 

positive. Figure 6-30 is a normal probability plot showing the distribution of the neutron-to-photon 

dose ratios determined from these 148 results. The ratios are approximately lognormally distributed 

with a geometric mean of 0.5 and a GSD of 1.9. They range from 0.1 to 2.8.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 56 of 92

y = 0.4645e0.6181x

R2

 = 0.9921

0.1

1.0

10.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratio

Figure 6-30. Normal probability plot for Group 21 neutron-to-photon 

dose ratios [50]. 

Figure 6-31 is a normal probability plot of the 160 positive neutron dose values for Group 21 workers 

in the period from 1990 through the second quarter of 2004. These data are approximately 

lognormally distributed with a geometric mean of 17 mrem and a GSD of 1.5. They range from 10 to 

58 mrem. 

y = 17.059e0.4019x

R2 = 0.9343

10

100

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-31. Normal probability plot for Group 21 neutron doses [51]. 

Comparison between the neutron-to-photon dose ratio determined from the personnel dosimetry 

records and that from the PNL characterization measurements for Group 11 show the value from the 

dosimetry data to be about a factor of 3 lower. However, the distribution of neutron-to-photon dose 

ratios determined for Group 11 is based on a limited number of results, as few people in this group 

receive neutron exposures in excess of the reporting limit. There were not enough records to 

establish a ratio for Group 12 workers, and workplace characterization data are not currently available 

for Building 7879 to allow a comparison for Group 21. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 57 of 92

6.3.4.2.3.5 Holifield and ORELA Facilities: Groups 6, 18, and 19

ORNL defines three groups for workers exposed to neutrons from the Holifield and ORELA 

accelerator facilities:

• Group 6: Personnel that work exclusively at the Holifield facility

• Group 18: Personnel that work exclusively at the ORELA facility

• Group 19: Personnel that work at both facilities

The TLD response-to-neutron-dose conversion factors for Groups 6 and 18 reflect the spectra 

determined for those the two respective facilities. The factors for Group 19 reflect a weighted average 

of the spectra measured for the ORELA and Holifield facilities in a 90:10 ratio (ORNL 1994).

Group 6 workers accounted for 407 (2%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 407 records, 8 (2%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were four records for which both the neutron and photon dose were positive. 

The neutron-to-photon dose ratios associated with these 4 records are 0.7, 1.1, 1.2, and 1.5. The 

measured neutron doses associated with these three ratios were 17, 18, 28, and 16 mrem, 

respectively. 

Group 18 workers accounted for 191 (1%) of the 20,037 neutron dosimetry records obtained from 

ORNL. None of the 191 records showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. 

Group 19 workers accounted for 290 (1%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 290 records, 4 (1%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were three records for which both the neutron and photon dose were positive. 

The neutron-to-photon dose ratios associated with these 3 records are 0.7, 0.9, and 1.9. The 

measured neutron doses associated with these three ratios were 20, 13, and 47 mrem, respectively. 

There were not enough positive results in the dosimetry records to establish distributions of 

neutron-to-photon dose ratios for any of the three accelerator groups, but the few records available 

suggest that the actual ratios for Holifield facility workers are much lower than the value indicated by 

the conservative characterization performed by PNL. The neutron-to-photon dose ratio derived from 

the PNL characterization data was 19, whereas the personnel dosimetry results indicate values 

between 1.0 and 2.0. 

6.3.4.2.3.6 Shielded Source Users: Group 13

Group 13 is used to evaluate neutron dose for personnel who work with shielded sources, such as 

well-loggers and some Instrumentation and Controls technicians. The TLD response-to-neutron-dose 

conversion factors for Group 13 reflect a weighted average of D2O-moderated 252Cf and bare Pu:Be 

spectra in a 95:5 ratio (ORNL 1994). 

Group 13 workers accounted for 675 (3%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 675 records, 44 (7%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 37 records for which both the neutron and photon dose were positive. 

Figure 6-32 is a normal probability plot showing the distribution of the neutron-to-photon dose ratios 

determined from these 37 results. The ratios are approximately lognormally distributed with a 

geometric mean of 0.5 and a GSD of 2.4. They range from 0.1 to 3.5.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 58 of 92

y = 0.5031e0.8549x

R2 = 0.9791

0.1

1.0

10.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratios

Figure 6-32. Normal probability plot for Group 13 neutron-to-photon 

dose ratios [52]. 

Figure 6-33 is a normal probability plot of the 44 positive neutron dose values for Group 13 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 18 mrem and a GSD of 1.6. They range from 10 to 53 mrem. 

y = 17.601e0.469x

R2

 = 0.9097

10

100

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-33. Normal probability plot for Group 13 neutron doses [53]. 

6.3.4.2.3.7 Neutron Source Make-up Workers: Group 14

Group 14 is used to evaluate neutron dose for personnel fabricating and handling isotopic neutron 

sources in ORNL’s isotope production (Isotope Circle) facilities, Building 3038. The TLD 

response-to-neutron-dose conversion factors for Group 14 reflect a weighted average of shielded 

Pu:Be spectra. 

Group 14 workers accounted for 222 (1%) of the 20,037 neutron dosimetry records obtained from 

ORNL. Of the 222 records, 72 (32%) showed a neutron dose greater than or equal to the 10-mrem 

reporting limit. There were 45 records for which both the neutron and photon dose were positive. 

Figure 6-34 is a normal probability plot showing the distribution of the neutron-to-photon dose ratios 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 59 of 92

determined from these 45 results. The ratios are approximately lognormally distributed with a 

geometric mean of 1.1 and a GSD of 2.2. They range from 0.2 to 9.1.

y = 1.1113e0.7715x

R2 = 0.9576

0.1

1.0

10.0

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron:Photon Ratios

Figure 6-34. Normal probability plot for Group 14 neutron-to-photon 

dose ratios [54]. 

Figure 6-35 is a normal probability plot of the 72 positive neutron dose values for Group 14 workers in 

the period from 1990 through the second quarter of 2004. These data are approximately lognormally 

distributed with a geometric mean of 25 mrem and a GSD of 1.8. They range from 10 to 182 mrem. 

y = 24.843e0.5529x

R2 = 0.9334

10

100

1000

-3 -2 -1 0 1 2 3

Normal Quantiles

Neutron Dose (mrem)

Figure 6-35. Normal probability plot for Group 14 neutron doses [55]. 

The neutron-to-photon dose ratio determined for Group 14 from the dosimetry records is significantly 

lower than that from the PNL characterization measurements. 

6.3.4.2.3.8 Other Workers: Groups 1, 5, 10, and 22

There were not enough positive results recorded in any of these groups between 1990 and the 

second quarter of 2004 to allow distributions of either neutron-to-photon dose ratios or neutron doses 

to be established.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 60 of 92

Group 1 is used to evaluate dose for workers at RADCAL. These individuals accounted for 406 (2%) 

of the 20,037 neutron dosimetry records obtained from ORNL. Of the 406, 5 (1%) showed a neutron 

dose greater than or equal to the 10-mrem reporting limit. There were four records for which both the 

neutron and photon dose were positive. The neutron-to-photon dose ratios associated with these four 

records are 0.6, 0.9, 1.3, and 1.6. The measured neutron doses associated with these four ratios 

were 14, 26, 14, and 28 mrem, respectively.

Group 5 is used to evaluate doses for worker and researchers at HFIR. This group represents the 

largest number of individuals monitored for neutron exposure at ORNL, accounting for 7992 (40%) of 

the 20,037 neutron dosimetry records obtained. However, only 5 (<0.1%) of these 7992 records 

showed a positive neutron dose from 1990 through the second quarter of 2004, and all 5 values were 

at or near the 10-mrem reporting limit. The largest of the five values was 14 mrem. In addition, there 

were five records for which both the neutron and photon dose were positive. Four of these records 

gave a neutron-to-photon dose ratio of 0.1, and the other gave a value of 0.2. 

Group 10 is used to evaluate neutron dose for individuals whose duties include work in the Building 

3100 storage vault. These individuals accounted for only 53 (0.3%) of the 20,037 records obtained 

from ORNL. Only 1 of the 53 results exceeded the 10-mrem reporting limit in the period from 1990 

through the second quarter of 2004 (the result was 11 mrem). There were no individuals for whom 

both the neutron and photon doses were positive. 

Group 22 was created primarily to support dose evaluations for Instrumentation and Control 

technicians exposed to isotopic neutron sources in both moderated and unmoderated configurations. 

The TLD response-to-neutron-dose conversion factors for Group 22 reflect a weighted average of 

bare Pu:Be and D2O-moderated 252Cf fission spectra in a 20:80 ratio (ORNL 1994). Workers assigned 

to Group 22 represented only 48 (0.2%) of the 20,037 records obtained from ORNL. Of the 48 

records, 11 (2%) showed a neutron dose greater than or equal to the 10-mrem reporting limit. There 

were only two records for which both the neutron and photon doses were positive. The 

neutron-to-photon dose ratios associated with these two records are 1.1 and 1.4, which correspond to 

neutron dose values of 25 and 17 mrem, respectively.

6.3.4.2.3.9 Summary for all groups

Of the 20,037 neutron dosimetry records obtained from ORNL, 3,252 (16%) showed a neutron dose 

greater than or equal to the 10-mrem reporting limit. There were 2,684 records for which both the 

neutron and photon dose were positive. Figure 6-36 is a normal probability plot showing the 

distribution of the neutron-to-photon dose ratios determined from these 2684 results. The ratios are 

approximately lognormally distributed with a geometric mean of 1.2 and a GSD of 2.2. They range 

from 0.1 to 14.1. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 61 of 92

y = 1.2219e0.7931x

R2 = 0.9887

0.01

0.10

1.00

10.00

100.00

-4 -3 -2 -1 0 1 2 3 4

Normal Quantiles

Neutron:Photon Ratio

Figure 6-36. Normal probability plot for neutron-to-photon dose ratios 

derived from ORNL personnel neutron monitoring records for 1990 – 

2004 [56]. 

Figure 6-37 is a normal probability plot of the 3,252 positive neutron dose values from the neutron 

dosimetry records obtained from ORNL for the period from 1990 through the second quarter of 2004. 

The data are approximately lognormally distributed with a geometric mean of 31 mrem and a GSD of 

2.1. They range from 10 to 584 mrem. 

y = 30.73e0.7402x

R2 = 0.9596

10

100

1000

-4 -3 -2 -1 0 1 2 3 4

Normal Quantiles

Neutron Dose (mrem)

Figure 6-37. Normal probability plot for neutron doses ≥10 mrem for 

ORNL workers in the period from 1990 – 2004 [57]. 

6.4 REPORTED QUANTITIES AND ADJUSTMENTS TO REPORTED DOSE

6.4.1

One version of the Individual Meter Record cards used to record film badge and pocket meter 

readings in the 1944 to 1946 period included columns labeled G and B. The G (gamma) value was 

the response of the film badge under the cadmium shield as determined from 226Ra calibration curves. 

Reported Quantities and Associated Algorithms: Beta-Gamma
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 62 of 92

This value is analogous to the deep dose. The B (beta) value was determined from 226Ra calibration 

curves; however, the meaning of this value is uncertain. Deal and Hart (1949) assert that the B value 

was simply the OW reading from the dosimeter, the same as the W or OW readings recorded in the 

years that followed. In addition, Deal and Hart state that it was an early practice not to record the B 

value unless the film density was at least twice that of the G value (recorded as S on other forms). 

However, Hart (1966) asserts that the B value was the computed difference between the dosimeter 

response from under the OW and that from under the cadmium shield (the G value). Review of claim 

files supports the case of the B value being the difference between the OW and S readings; therefore, 

dose reconstructors should interpret it in this fashion. Thus, G and B should be summed if the OW 

result from the reading of a particular dosimeter is desired. NOTE: (1) Treating the B value as the 

difference between the window and shield results is the favorable to claimant choice; (2) other 

versions of the Individual Meter Record cards from that period (1944 to 1946) had data columns for 

film badge results labeled W and S (corresponding to the OW and cadmium-filter responses, 

respectively) rather than G and B. 

After 1946, ORNL practice was apparently to record results from the two-element film badges as OW 

and S, referring to readings behind the OW and the cadmium filter, respectively. The OW readings 

were sometimes called W. Data columns on the cards were used to record results for the sensitive 

and insensitive films in the beta-gamma badges. These results were noted as FILM [results from 

neutron dosimeters (NTA film) were called NEUTRON]. Values of S and OW (or W) were recorded 

for the sensitive beta-gamma films and for the neutron dosimeters. Apparently, only S values were 

recorded for the insensitive film, which was used only if the sensitive film blacked out. In 1949,

threshold (maximum dose) values were administratively set to 5,000 mR(ep) and 20,000 mR(ep) for 

sensitive and insensitive films, respectively. These values, considered the upper range of 

measurement for the respective films, were entered on the data cards in the event a blacked-out film 

was encountered. Other notations indicated anomalies such as fogging of the film, evidence of X-ray 

(low-energy photon) exposure, contamination, etc. NOTE: By 1952, the threshold value for a 

blacked-out sensitive film appears to have been increased to 10,000 mR(ep) (Craft, Ledbetter, and 

Hart 1952). 

Beginning in the second half of 1951, ORNL introduced the terms probable total reading (PTR) and 

probable maximum exposure (PME) into its dosimetry practices (Hart 1966). The intent of the PTR 

was to account for over-response of the OW reading by subtracting the gamma response from behind 

the cadmium shield. Thus, PTR was defined as

PTR ≡ OW – S

where OW is the response in millirep determined from the film density behind the OW element (based 

on calibration to uranium) and S is the response in milliroentgen based on the film density behind the 

cadmium shield (based on calibration to radium). However, if the S (shield) reading was higher than 

the PTR as computed above, the PTR was set to the value of S rather than OW minus S (Craft, 

Ledbetter, and Hart 1952). 

The PME was the sum of the PTR values for a given monitoring period unless there were artifacts in 

any of the dosimeter readings that necessitated a field investigation (Craft, Ledbetter, and Hart 1952). 

If an investigation was required, the PME was assigned as an estimate of the probable maximum 

exposure (Hart 1966). 

PTR and PME, along with OW (or W) and S, were apparently reported until the second half of 1956, 

when ORNL adopted the depth dose quantities of NBS Handbook 59 (NBS 1954). These quantities 

included a skin dose evaluated at a density thickness of 7 mg/cm2

, which at the time could not be 

determined from film badge readings. The skin (superficial) dose was designated as DS. Other 

quantities adopted at this time were the moderately penetrating dose (DM or MOP), a dose to the lens 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 63 of 92

of the eye (DL or LEN), and a penetrating dose (DP or PEN). DP was determined from the reading 

under the cadmium filter, and thus was physically the same as S. NOTE: There were fields for 

values of DS on the data cards used to record personnel exposure results, even though the quantity 

was not actually computed until the second half of 1961. 

In the second half of 1961, after ORNL began using the Model II film badge, it began reporting only 

the quantities DS and DC on the data cards used to record personnel exposure results. DC referred 

to critical organ dose, and was physically the same as the previously used penetrating dose. Both of 

these quantities are analogous to deep dose. ORNL began computing values for the skin (superficial) 

dose at this time, using the expression:

DS = 2.5(W – P) + C

where W was the response under the OW, P was the response under the plastic element, and C was 

the response under the cadmium filter (Thornton, Davis, and Gupton 1961). NOTE: ORNL used the 

quantities DS and DC to refer to superficial and deep dose, respectively, through the late 1980s and 

the adoption of the DOELAP-accredited CEDS. 

The purpose of the term 2.5(W – P) in the expression for DS above was to represent the beta portion 

of the total dose, with the factor of 2.5 being a correction for the attenuation by the OW and plastic 

filters relative to a dose depth of 7 mg/cm2

. The factor of 2.5 was based on the beta spectrum from 

natural uranium, but different factors could be used if there was evidence of exposure to a different 

beta energy (Thornton, Davis, and Gupton 1961). 

The above expression was used to compute values of DS from the second half of 1961 until the first 

quarter of 1968. However, it was not always applied because other administrative practices were in 

place that governed how values of DS (and DC for that matter) were computed. For example, it can 

be seen in the claim files that if the value of P was blank, the expression was not used and apparently 

DS was assigned as the value of W (i.e., the OW reading). However, if the value of C was greater 

than W (and P was blank), DS was assigned equal to DC (where DC = C, the reading from under the 

cadmium filter). Review of the data implies that blank entries for P on the data cards mean the value 

was less than the LOD of the dosimeter. NOTE: Computed values were apparently rounded to the 

nearest 10 mrem. 

The above examples of cases where the DS equation was not used is by no means a complete 

treatment of how values of DS were computed. Indeed, there are many examples in the claim files 

where the means used to assign a value for DS cannot be determined from the information on the 

dosimetry record. Gupton (1968) reports that dose evaluation at ORNL could have involved:

• Calculations using data obtained from personnel dosimeters

• Knowledge of workplace radiation fields and exposure conditions

• Radiation survey data

• Investigatory actions

• Cohort dosimetry

• Reconstruction of exposure conditions, etc.

Thus, one should not assume that values of DS were computed using the above expression in cases 

where only summary-level information is available. However, it appears that in most cases dose 

reconstructors will have dosimetry record cards from which to work, which include dosimeter readings 

in addition to assigned doses and remarks. 

Beginning with the first quarter of 1968, the expression used for computing the superficial dose DS 

from individual dosimeter element readings was revised to: 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 64 of 92

( )

( ) ( ) CPW

CP

CW DS +− −− = 8.0

where W, P, and C are defined as above. As with the previous expression, this expression was not 

always applied to compute values of DS. Gupton (1968) states that this expression was used for 

routine exposure conditions, meaning:

• No individual dosimeter element readings exceeded 500 units.

• No ratio of adjacent elements exceeded a factor of 2.

• There was no unusual appearance of the film.

Apparently, other administrative practices were used as well, such as one involving the first term of 

the DS expression (i.e., 80% of the ratio of the net response under the window and plastic elements). 

From review of claim data, this term apparently was constrained such that it was never less than 

unity; that is, if a value less than unity were obtained, the calculation of DS reduced to: 

DS ( ) +−= CPW

As with the initial expression for DS above, other administrative practices were associated with the 

assignment of values of DS, and there are instances in the claim files where the bases for assigned 

values are not immediately obvious from the dosimeter data. Gupton (1968) states that nonroutine 

cases were forwarded to external dosimetry supervisors, who could then assign a dose based on the 

dosimeter reading or initiate an investigation by the ORNL radiation survey section. 

TLDs appear to have first been employed at ORNL to measure deep dose (DC) for workers in the first 

quarter of 1974. The advent of the use of TLDs for assigning values of DC meant a slight modification 

of the expression used to compute DS. Specifically, the C term representing the penetrating 

contribution to the superficial dose was replaced with DC, making the expression: 

( )

( ) ( ) PW DC

CP

CW DS +− −− = 8.0

NOTE: The values of C from the film badge were still used to compute the net response of W and P

in the first term. 

After 1975, when ORNL switched to TLDs exclusively for routine monitoring, the methods used to 

compute the quantities DS and DC become less clear, complicated by the fact that several different 

TLD designs were in use. Thus, the methods used to compute deep and shallow dose varied 

somewhat depending on the type of dosimeter assigned to a worker. NOTE: The dosimeter model a 

worker was wearing for a given period can be ascertained from the Beta-Gamma Dosimetry Record 

cards by looking at the TLD elements employed (see Table 6-1). 

For TLDs issued to nonradiation workers, shallow (skin) dose was determined from the OW, which 

was covered by an effective density thickness of 60 mg/cm2 (Parrish 1979). The basic assumption 

behind this practice was that individuals assigned this dosimeter would not be exposed to radiation 

fields where there would be an appreciable difference between dose at the effective depth of the OW 

and that at 7 mg/cm2 (Gupton 1978). 

According to Gupton (1978), the skin dose from the Class 2 (yellow or yellow dot) and Class 3 (red or 

red dot) dosimeters (both of which were issued to radiation workers) was determined for a dose depth 

of 7 mg/cm2 with the expression 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 65 of 92

( )

( ) AP

OW A DS A −− +=

2

where OW is the response from the element under the OW, A is the response from the element under 

the aluminum filter, and P is the response from the element under the plastic filter. However, a 

criterion was applied against skin dose evaluations made in this manner to identify cases where the 

calculated dose could be inappropriate due to artifacts such as the energy of the incident radiation, 

the angle of incidence, read errors, and so forth. Specifically, if the ratio of the calculated skin dose to 

the response of the OW TLD element was greater than 4, this was an indication that further action 

was required in either an evaluation of the Type 2 film packet in each dosimeter or an exposure 

investigation (Gupton 1978). Claims data have not been reviewed to confirm if this expression for DS 

was used, or under what conditions it might have been applied. 

After 1988, ORNL changed to the DOELAP-accredited CEDS dosimeter and associated dose 

algorithms, for which assigned shallow doses can be assumed equivalent to Hp(0.07). 

6.4.2

In the period when NTA film was used exclusively for neutron personnel monitoring, the film was 

processed for a given individual only if the field health physicist recommended it. Similarly, when 

neutron personnel monitoring was performed using neutron-sensitive TLDs, the TLDs were issued 

only to individuals for whom neutron exposure was anticipated. The other TLDs in use were capable

of indicating neutron exposure, however, and cases have occurred where neutron dose was added to 

a worker’s deep and shallow dose totals on the beta-gamma dosimetry record cards. 

Reported Quantities and Associated Algorithms: Neutrons

Data fields (i.e., entry blanks) for neutron exposure data first appeared on ORNL dosimetry data cards 

in early 1948. Fields were added for OW and S readings under the heading NEUTRON. Thus, at that 

time, beta-gamma and neutron data were recorded on the same card. In 1952, ORNL changed to the 

use of separate data cards for beta-gamma and neutron dosimetry data. Photographic Film 

Dosimetry Record cards (or similar) were used for beta-gamma data, and Nuclear Track Film 

Dosimetry Record cards (or similar) were used for neutron data. 

The original ORNL method for computing what it called thermal neutron dose was to subtract the NTA 

film reading under the cadmium shield from that under the OW, assuming the tracks observed under 

the window were the result of fast and thermal neutrons, and those under the shield were from fast 

neutrons alone. However, it is unclear if neutron dose was reported in terms of different energy 

components. The dosimetry data cards examined either had fields for recording the window and 

shield readings separately or had single values for neutron dose (representing the contribution from 

all energies).

Between the late 1940s and early 1950s, neutron dose data were reported in terms of a fraction of 

tolerance, where the neutron exposure and tolerance were both stated in terms of tracks per field or, 

more typically, tracks for a number of fields (Hart and Walters 1950). For example, for the first half of 

1950 tolerance values asserted on the Neutron Film Record cards were in terms of a number of tracks 

per 12 fields. The tolerance values differed for the OW and S readings, with that for the OW being 22 

tracks per 12 fields and that for the S reading being 18 tracks per 12 fields. To convert such values to 

dose equivalent, dose reconstructors should apply a factor of 10 tracks/cm2

/mrem (Thornton, Davis, 

and Gupton 1961). This factor is for fast neutrons and inherently includes a quality factor of 10. The 

value 2 × 10-4 cm2

/field (Thornton, Davis, and Gupton 1961) should be used to convert a number of 

fields to area. Thus, for a reported value of “3/22” for the first half of 1950, which implies 3 tracks per 

12 fields observed relative to a tolerance level of 22 tracks per 12 fields, the dose conversion is as 

follows:
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 66 of 92

125 mrem.

10 tracks 

 1mrem cm 

 2 10 cm 

 1field

12 fields

tracks 3 2

24- =













 ⋅ 



 





×





 





Dose reconstructors will need to consider other factors before assigning neutron dose in this manner, 

such as missed dose and the adequacy of NTA film for the neutron field in question. The dose 

equivalent indicated by the values reported on the Neutron Film Record cards should also be 

considered versus the LOD for NTA film dosimeters. For instance, in the example above, a value less 

than 3 tracks per 12 fields would equate to less than 100 mrem, suggesting that the tracks could be 

from photon events or effects other than a small neutron dose. 

In later years neutron exposure data were expressed in terms of tracks per scan rather than tracks 

per field, where a scan was defined as a unit of volume equal to the depth of the emulsion, the width 

of the field of view, and a length of 2 mm (Gupton 1968). NTA films were read using 900- to 950-

power magnification under dark field illumination. Gupton (1968) states that as many as five scans 

could have been observed for a dosimeter depending on the track density. 

Neutron dose was assumed to be penetrating and thus was added to both superficial and penetrating 

dose values from other radiations as necessary when compiling summary information. As of 1969, 

ORNL began doubling neutron dose values to obtain the recorded dose for unknown spectra (Gupton 

1969). 

Following the adoption of TLDs for routine monitoring, personnel neutron dosimetry at ORNL was 

effected through use of the ORNL neutron dosimeter, known as the Class 3, red, or red dot badge. 

This dosimeter used combinations of neutron-sensitive and -insensitive TLD chips in conjunction with 

NTA film. Empirical algorithms were developed based on knowledge of workplace radiation fields and 

the neutron energy response characteristics of the dosimeter. These algorithms enabled assignment 

of a total neutron dose that reflected contributions from four energy groups: Thermal, intermediate, 

0.05 MeV to 0.5 MeV, and greater than 0.5 MeV. Dose from the thermal group was measured by the 

TLDs (pairs of TLD-600 and TLD-700 chips) alone. Similarly, dose from the greater-than-0.5-MeV 

energy group was determined from the NTA film alone. Doses from the other groups were computed 

by empirical algorithms based on the response of the TLDs and the NTA film. Thus, neutron dose 

assigned for ORNL personnel from 1975 to the present represent the dose contribution from all 

energies based on the known energy response characteristics of the dosimeters and characterization 

of workplace spectra. 

The algorithms employed to compute neutron dose for the thermal, intermediate, and fast energy 

groups using the ORNL neutron dosimeter between 1975 and approximately 1985 were as follows. 

Determine the parameters T1 and T2, where T1 is the albedo response from under the 1-mm cadmium 

filter, and T2 is the incident plus albedo response from under the aluminum filter. T1 is computed as 

the response of the TLD-600 chip minus that of the TLD-700 chip (from the pair under the cadmium 

filter). T2 was the same difference computed for the pair of chips under the aluminum filter. The 

thermal neutron dose was then computed as:

( ) 02 12 NT .0 −= TT

where

NT = thermal neutron dose equivalent

0.02 = thermal neutron constant
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 67 of 92

The fast neutron dose equivalent was computed as:

( ) T NTA

T

T NF − 













 = 1

2

1 1.0

where 

NF = fast neutron dose equivalent

0.1 = fast neutron constant

NTA = NTA film reading

NOTE: The ratio of T1 to T2 was a function of the hardness of the neutron energy spectrum (Gupton

1978). 

The dose from the intermediate neutron energy group was computed as:

NI = [T ( ) +− NTANF ] 035 1 .0

where

NI = intermediate energy neutron dose equivalent;

0.1 = intermediate energy neutron constant. 

The total neutron dose equivalent was then computed as the sum of the four components – NT, NI, 

NF, and NTA. 

Between 1981 and 1985, the NTA film packet was eliminated from routine use. The algorithm used to 

determine neutron dose equivalent from the ORNL red dot TLD was then modified as follows (Berger 

and Lane 1985): 

Compute the parameter A, which indicated the response from incident, thermal neutrons as: 

( )

2

12

T

TT A − =

and the parameter R, which indicated the quality of the neutron spectrum where:

2

1

T

T R = . 

An empirical calibration factor CAL was then computed as:

( ) AR CAL e .7 64 .1 46 − = , 

and finally, the neutron dose equivalent was computed as the product of T1 and CAL. 

Sometime around the mid-1980s, ORNL began using four-element Panasonic TLDs for neutron 

monitoring (MMES 1992). In 1989, ORNL began implementing the CEDS, which used two 

dosimeters for whole-body monitoring: A four-element beta-gamma dosimeter (blue holder) and a 

four-element neutron-sensitive dosimeter (red holder). The CEDS beta-gamma dosimeter became 

the dosimeter of record as of January 1989. However, the use of the Panasonic neutron dosimeter 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 68 of 92

continued as the neutron dosimeter of record until January 1990. The transition to CEDS at ORNL 

included a comprehensive effort to further characterize the workplace neutron spectra. As before, this 

information was used in combination with the known neutron energy response characteristics of the 

CEDS neutron dosimeter to develop methods for estimating neutron dose for individual workers 

based on evaluations of time spent in different workplace neutron fields. 

6.4.3

6.4.3.1 Deep Dose

Adjustments to Recorded Dose: Beta-Gamma

All personnel dosimetry employed at ORNL over its history included suitably filtered elements for the 

determination of deep dose. The ORNL film dosimeters, which were calibrated in terms of exposure 

from 226Ra photons, show either good agreement with or an over-response to the quantity Hp(10) (see 

Figure 6-1). However, because dosimeters were calibrated in terms of exposure over most of the 

Laboratory’s history, dose reconstructors should apply exposure to organ dose conversion factors to 

deep dose results recorded for ORNL workers for the period before 1989 (i.e., for the period before 

DOELAP accreditation). After 1989, reported deep dose values are considered equivalent to Hp(10). 

6.4.3.2 Shallow Dose

Shallow dose was not explicitly reported at ORNL until 1961 when the Laboratory began reporting the 

quantity DS, referred to as the superficial, nonpenetrating, or skin dose. DS was defined at a dose 

depth of 7 mg/cm2 and thus was intended to represent a skin dose. ORNL adopted this quantity (from 

NBS Handbook 59) in the second half of 1956, but had no algorithm by which to assign it at that time 

(NBS 1954). Before 1961, the Laboratory reported film badge results for the individual dosimeter 

elements with the exception of the period from 1944 through 1946 when the quantities B and G were 

used. Dose reconstructors should sum the B and G values to obtain the corresponding OW reading, 

and should use the OW results in cases where shallow dose data are required before 1961 when 

ORNL began reporting the quantity DS. However, some adjustments to the OW readings are 

required for certain periods so they are consistent with one another. 

The window dose assigned in the period from late June 1944 through the eighth week of February 

1947 was based on calibration curves determined with the use of a radium photon source. After that 

time, the conversion of film density under the OW to dose was based on calibration to a uranium slab. 

The uranium slab came into use in the ninth week of February 1947; however, the surface dose rate 

of 270 mrep/hr originally assumed for the slab at that time was revised to 240 mrep/hr on March 1, 

1951 (Hart 1966). Therefore, corrections are necessary for OW results reported between 1944 and 

March 1, 1951, as follows.

• Multiply OW results recorded from 1944 through the eighth week of February 1947 by a factor 

of 2.2. This factor represents the ratio of the window reading based on calibration to uranium 

to that based on 226Ra, with the surface dose rate for the uranium source taken to be 

240 mrep/hr (Hart 1966). 

• Multiply OW results recorded from the ninth week of February 1947 through February 1951 by 

0.9 to adjust from the original dose rate assumed for the uranium source (270 mrep/hr) to the 

240 mrep/hr value adopted later (Hart 1966).

These corrections should enable comparison of all OW (or W) readings recorded by ORNL on a 

consistent basis. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 69 of 92

6.4.4

Adjustments to neutron dose data for ORNL workers are required to separate doses into appropriate 

energy groups for input into IREP and to account for differences between the neutron quality factors in 

use at the time dose assignments were made and the radiation weighting factors were promulgated in 

ICRP Publication 60 (ICRP 1990). Neutron dose data for ORNL workers should be assigned to IREP 

neutron energy groups using the information in Attachment B. Adjustments required to state the 

values in terms of the neutron radiation weighting factors of ICRP 60 are described below. 

Adjustments to Recorded Dose: Neutron

Neutron quality factors historically used at ORNL were

• 5 for fast neutrons and 2 for thermal for the period from 1943 until 1950

• 10 for fast neutrons and 2.5 for thermal for the period from 1950 until 1971

• Values equivalent to those reported in Table 2 of NCRP Report 38 (NCRP 1971) from 1971 to 

the present

Fast neutrons at ORNL consisted primarily of those associated with nuclear fission, and thus should 

correspond to the 0.10-to-2.00-MeV energy group defined in ICRP Publication 60 (ICRP 1990) and 

used in IREP. The ICRP 60 radiation weighting factor for the 0.10-to-2.00-MeV energy group is 20. 

Thus, if any reported neutron dose data are used, dose reconstructors should multiply fast neutron 

doses reported for ORNL workers between 1943 and 1950 by a factor of 4 to account for the 

difference in historical and current quality factors. Similarly, dose reconstructors should multiply fast 

neutron doses reported for the period from 1950 to 1968 by a factor of 2. After 1968, ORNL began 

applying a factor of 2 to neutron dose data, apparently to account for uncertainty associated with 

unknown neutron spectra (Gupton 1969). From review of claim files it should be apparent when this 

factor of 2 was applied, thus dose reconstructors should know if a given result needs to be doubled to 

account for differences in quality factor or if a factor of 2 has been applied. NOTE: The factor of 2 

appears to have been applied for conservatism. It does not appear to have been a quality factor. 

This practice appears to have continued until the adoption of TLDs for personnel neutron monitoring 

in 1975. 

ORNL eventually began reporting neutron dose in terms of the contribution from all energies based on 

characterizations of workplace spectra and calibration of dosimeters to well-characterized calibration 

spectra. Because assigned dose equivalents from this era were based on the neutron quality factors 

of NCRP Report 38 (NCRP 1971), a factor is needed to convert from these factors to those of ICRP 

Publication 60 (ICRP 1990). This conversion is complicated by the fact that the neutron quality 

factors in NCRP 38 are point values, while those in ICRP 60 are in terms of neutron energy groups. 

Therefore, establishing a factor to adjust from NCRP 38 values to those from ICRP 60 requires 

application of a group-averaging technique to the NCRP 38 data. This process has been detailed in 

other site profiles developed for the NIOSH Dose Reconstruction Project [for the Y-12 plant for 

instance – see ORAUT (2006)] and is not repeated here. However, Table 6-23 summarizes the 

results of these calculations for neutron energy groups of interest. 

Table 6-23. Group-weighted NCRP 38 neutron quality factors and ICRP 60 radiation weighting 

factors.

a

ICRP 60 (IREP) neutron 

energy group (MeV)b

NCRP 38 group-averaged 

quality factor

ICRP 60 radiation 

weighting factor

Ratio 

(ICRP 60-to-NCRP 38)

Thermal (<0.01) 2.35 5 2.1

0.01–0.10 5.38 10 1.9

0.10–2.00 10.49 20 1.9

2.0–14.0a 7.56 10 1.3
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 70 of 92

a. ICRP 60 = ICRP (1991); NCRP 38 = NCRP (1971).

b. The ICRP energy group is actually from 2.0 to 20.0 MeV; it was truncated to accommodate the group averaging 

process. 

Thus, to adjust more recent reported neutron dose data for ORNL workers (i.e., results for which a 

neutron energy group was not specified), one would multiply by a factor of 1.9 to account for the 

difference between the previously applied neutron quality factors and the ICRP 60 weighting factors 

(assuming an intermediate or fast neutron energy group) (ICRP 1990). Thermal results would be 

scaled up by a factor of 2.1 [58].

In light of the above discussion on historical neutron quality factors in relation to the neutron weighting 

factors of ICRP 60 (ICRP 1991), dose reconstructors should consider applying a single factor of 2 to 

all reported neutron dose data from 1950 to the present for simplicity, with the caveat that data 

reported from 1969 through 1974 could already reflect application of such a factor [59]. However, the 

data cards should show both the corrected and uncorrected dose values for the 1969 to 1974 period if 

the factor of 2 was used. 

6.5 LIMITS OF DETECTION AND MISSED DOSE

6.5.1

Table 6-24 lists LODs for deep dose and dosimeter exchange frequencies for ORNL radiation 

workers. Dosimeters for other individuals (i.e., nonradiation workers) were sometimes exchanged 

less frequently, and dosimeters for radiation workers are sometimes exchanged more frequently 

under circumstances such as potentially high doses or pregnant workers. 

Deep Dose

Table 6-24. LODs for deep dose and dosimeter exchange frequencies for ORNL radiation workers. 

Period of use Dosimeter LOD (rem) Exchange frequency

June 1944 through 2nd qtr. 1956 Two-element and multielement film 0.03 Weekly

3rd qtr. 1956 through 1974 Multi-element film 0.03 Quarterly

1975–1988 ORNL TLDs 0.01 Quarterly

1989–present CEDS TLD 0.01 Quarterly

NOTE: Deviations from the LOD values in Table 6-24 can be observed in the claim files. However, 

the deviations noted have always been cases where the LOD was greater than the assigned dose. 

For example, instances of assigned DC (deep dose) values as low as 10 mR are seen as early as the 

mid-1960s, and cases of assigned deep doses of 20 mR are seen as early as the mid-1950s. 

6.5.2

Errors associated with reading film badges for beta exposure are essentially the same as those for 

X-ray and gamma exposure, and it appears to have been ORNL practice to equate the LOD for OW 

with that for the cadmium element (Morgan 1961). This practice continued after the adoption of the 

quantity DS and its associated algorithms and administrative practices, because DS values were 

generally assigned the deep (DC) LOD value in the event of a dosimeter reading less than its 

sensitivity threshold. This practice appears to have continued until the adoption of the CEDS 

dosimetry system in January 1989. 

Shallow Dose

Similar to the Morgan (1961) assertion for film badges, the technical basis for use of CEDS 

dosimeters (ORNL 1994) implies that the LODs for beta and photon radiations are essentially 

equivalent in terms of shallow dose from 90Sr/Y beta particles and photons from 137Cs. ORNL (1994) 

shows that the shallow dose, lower LOD for the CEDS dosimeter calculated on the basis of 

background variability is approximately 40% higher for a 204Tl beta spectrum (average energy 

approximately 0.24 MeV) than for a filtered 90Sr/Y spectrum (average energy approximately 0.8 MeV). 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 71 of 92

Therefore, for assigning missed shallow dose, dose reconstructors should apply the deep dose LOD 

values from Table 6-24 unless there is reason to suspect exposure to a field consisting primarily of 

low-energy electrons [60]. In the latter case, assigning a missed shallow dose would be problematic 

except in cases where the worker was monitored with the CEDS TLD, for which the shallow LOD for 

(pure) low-energy beta exposure would be approximately 14 mrem/quarter. 

6.5.3

There is considerable uncertainty about neutron doses that ORNL workers could have received 

before the advent of TLD technology in 1975. The available information indicates the use of NTA film 

on an experimental basis to monitor personnel at the Graphite Reactor during the period before its 

formal introduction. In addition, neutron radiation surveys were used to control personnel neutron 

dose and to define the need for shielding. The available information is vague on documentation of 

these surveys or of any potential neutron dose determined from them in employee exposure records. 

Some references indicate that neutron doses were relatively low until the onset of production of 

isotopic neutron sources, while others suggest potentially significant exposures for certain operations 

in the 1944 to 1945 period (Wirth, Morgan, and Curtis 1945). These potentially significant exposure 

conditions could have been isolated events, or they could have been corrected by radiation protection 

measures such as shielding or limits on occupancy times. 

Neutron Dose

The LOD for personnel neutron dosimetry using NTA film was 100 mrem (based on a quality factor of 

10) for neutron energies greater than 0.5 MeV. The sensitivity of NTA film to neutrons with energies 

less than 0.5 MeV is essentially zero; that is, neutrons at this energy or lower cannot produce a 

discernible track in the emulsion. Therefore, two components of missed dose must be considered for 

the period when NTA film was used exclusively for personnel neutron monitoring: Dose that was 

below the LOD of the dosimeter and dose associated with neutron energies below 0.5 MeV. 

Because the leakage neutron spectrum that would have been present in general areas of the Graphite 

Reactor was highly thermalized, it is doubtful that NTA film dosimeters were capable of providing an 

accurate and reliable estimate of personnel neutron exposures for workers in these areas. This logic 

can be extended to other ORNL facilities where there was potential for neutron exposures from 

degraded spectra. The neutron spectral characterizations performed by PNL in 1989 through 1991 

showed that NTA film should have provided reasonable results in harder spectra such as those 

associated with 238Pu:Be sources or glovebox operations where there was relatively little shielding. 

The characterization data show that NTA film probably would have under-responded in softer spectra 

where there was significant neutron scatter with the magnitude of this under-response ranging from 

modest to substantial (see Attachment F). Therefore, unless it is known that an individual’s neutron 

exposure for a given period occurred in a field where NTA film should have provided reliable 

monitoring, dose reconstructors should assign neutron dose when necessary by applying 

neutron-to-photon dose ratios (see Attachments C and D). Appropriate ratios should be applied to 

both monitored and missed photon dose for individuals and periods for which neutron dose could be 

anticipated and NTA film was used exclusively for personnel neutron monitoring (i.e., before 1975). 

Missed photon dose is addressed in Section 6.5.1. However, because an individual’s neutron dose 

could have been assigned based on survey data, a dose investigation, or other judgment, dose 

reconstructors should compare neutron dose values computed using neutron-to-gamma dose ratios to 

any neutron dose recorded for the same period (using consistent neutron quality factors) and select 

the higher value. As an alternative, the neutron dose distributions given in Section 6.3.4.2.3 and 

Attachment E could be considered as a means for estimating an individual’s neutron dose. 

Facilities at ORNL where neutron dose could be anticipated include reactors and those listed in 

Attachment B. The neutron-to-photon dose ratios given in Attachment D and discussed in Section 

6.3.4.2.3 are derived from personnel neutron monitoring data provided by ORNL for the period from 

1990 through the second quarter of 2004. These values reflect actual personnel neutron and photon 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 72 of 92

exposures and should be used in favor of the values given in Attachment C as practicable. The

values in Attachment C are derived from characterization of workplace neutron spectra performed in 

conservatively chosen locations with high neutron dose rates. 

At present, measured values of neutron-to-photon dose ratios for ORNL reactor facilities are available 

only for the Graphite Reactor, the HPRR, and the HFIR. The neutron-to-photon dose ratio estimated 

for the Graphite Reactor compares well with values for the early Hanford production reactors in 

ORAUT (2007). The ratios for the HFIR from both the PNL characterization data and the ORNL 

neutron dosimetry records are essentially the same as that for the Graphite Reactor. This is 

consistent because both facilities are characterized by highly thermalized spectra. Similarly, the 

neutron-to-photon dose ratio information reported by Kerr and Johnson (1968) for the HPRR is 

comparable with the values observed for the spectra encountered at the REDC. The highest ratio 

derived from the ORNL personnel neutron monitoring records for any of the neutron worker groups 

was a geometric mean of 2.5 having a GSD of 1.6. This ratio, from the dosimetry data for Group 3 

(see Section 6.3.4.2.3.1), corresponds to individuals working in a fission neutron spectrum in a 

low-scatter environment. Lower values are seen for other facilities and job duties. Looking at 

personnel neutron monitoring results for ORNL workers collectively, the records for the period from 

1990 through the second quarter of 2004 yield a geometric mean neutron-to-photon dose ratio of 1.2 

with a GSD of 2.2 (see Section 6.3.4.2.3.9). This value is derived from 2,684 individual monitoring 

records where both the measured neutron and photon dose were greater than or equal to 10 mrem. 

These records are a subset of a total of 20,037 personnel neutron monitoring records obtained from 

ORNL for this period, representing monitoring in all workplace neutron spectra. Therefore, for 

reactors or other facilities not addressed in Attachments C or D, dose reconstructors can evaluate the 

applicability of the distribution of neutron-to-photon dose ratios derived from the ORNL personnel 

monitoring records for similar facilities or job duties, the distribution for all facilities (a geometric mean 

of 1.2 with a GSD of 2.2), or the distribution yielding the maximum ratios (a geometric mean of 2.5 

with a GSD of 1.6 for Building 7930 workers). Dose reconstructors should refer to the 

neutron-to-photon dose ratios in Attachments C and D to assign missed and unmonitored dose for 

individuals who received neutron exposures in comparable facilities or job duties. 

Distributions of neutron-to-photon dose ratios or dose values should be used to estimate missed and 

unmonitored neutron dose for ORNL workers with potential neutron exposure for the period before 

1975. Dose reconstructors should use these methods in cases where individuals were monitored via 

NTA film and it is known or suspected there was significant dose associated with neutrons having 

energies less than 500 keV. Photon missed dose values for this period should be in accordance with 

Table 6-24. 

For 1975 to the present, neutron monitoring for ORNL workers should have been adequate in terms of 

sensitivity of the methods used to the various workplace neutron energy spectra. Thus, missed 

neutron dose can be estimated on the basis of LODs and dosimeter exchange frequencies in the 

same manner used for deep dose. Dose reconstructors should note that they could encounter 

instances where an individual was issued more than one dosimeter for a given monitoring period. If 

so, missed dose should be addressed for each dosimeter worn during the period. Some workers are 

issued multiple dosimeters because they work in different neutron groups, and dosimeters are 

sometimes processed more frequently than quarterly at the request of field health physics staff. The 

latter case typically applies to REDC workers for whom higher doses are anticipated. In any event, 

applying the quarterly LOD values from the table below for more frequent exchanges should be 

conservative. 

Table 6-25 lists LODs for neutron dose and dosimeter exchange frequencies for ORNL neutron 

radiation workers. An LOD for the brief period in which the Panasonic neutron dosimeter was used is 

not given because the value for the ORNL neutron TLD should be sufficiently bounding. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 73 of 92

Table 6-25. Limits of detection for neutron dose and dosimeter exchange frequencies 

for radiation workers from 1975 to the present. 

Period of use Dosimeter LOD (rem) Exchange frequency

1975 through 1989 ORNL neutron TLDs (red dot) 0.02 Quarterly

1990–present CEDS neutron TLD 0.01 Quarterly

6.6 UNCERTAINTY IN PHOTON AND NEUTRON DOSE

For film badges, the LODs that are quoted in the literature range from about 30 to 50 mrem for betagamma irradiation (Morgan 1961; Parrish 1979; West 1992) and from 50 to 100 mrem for neutrons 

(Morgan 1961; Parrish 1979). These are not the expected uncertainties at larger photon and neutron 

dose readings. For example, it was possible to read a photon dose of 100 mrem to with ± 15 mrem if 

the exposure involved photons with energies between several hundred keV and several MeV (Morgan 

1961). If the exposure involved photons with energies less than several hundred keV, the uncertainty 

was at least twice that for the more energetic photons. Thus, the standard error in the recorded film 

badge doses from photons of any energy is estimated here to be ± 30%. The standard error for the 

recorded dose from beta irradiation was essentially the same as that for photon irradiation, but when 

an unknown mixture of beta and photon irradiation was involved the standard error for the dose from 

beta irradiation was somewhat larger than 30% (Morgan 1961). The situation for neutrons was not as 

favorable as that for photons. With NTA films, the estimated standard error was much larger and 

varied significantly with the energy of the neutrons. Thus, the standard error for a neutron dose 

reading of approximately 100 mrem is estimated here to be ± 50%. For the TLD beta-gamma and 

neutron dosimeters used at ORNL after the mid-1970’s, the standard errors for a recorded dose 

reading of 100 mrem or more are estimated here to be approximately ± 15% for photons, beta 

particles, and neutrons. The standard errors for TLD dose measurements less than 100 mrem and 

those in mixed radiation fields would be expected to be somewhat larger. 

6.7 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

Joseph Mohrbacher served as the initial Subject Expert for this document. Mr. Mohrbacher was 

previously employed at ORNL and his work involved management, direction or implementation of 

radiation protection and/or health physics program policies, procedures or practices related to atomic 

weapons activities at the site. This revision (and original revision) has been overseen by a Document 

Owner who is fully responsible for the content, including all findings and conclusions. In all cases 

where such information or prior studies or writings are included or relied upon by Mr. Mohrbacher, 

those materials are fully attributed to the source. 

[1] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The mixture of 75% >250-keV photons and 25% 30-to 250-keV photons was selected to reflect 

a field comprised of high-energy and Compton-scattered photons, as would be expected for 

these facilities. For high-energy photon facilities (accelerators), 100% photons >250 keV was 

selected. For facilities where low-energy photons were possible, the 30- to 250-keV group 

was specified to be most favorable to the claimant since the presence of fission products could 

not be ruled out.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 74 of 92

[2] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These recommendations follow the best current practices for interpreting shallow dose in 

relation to ensuring favorability to the claimant.

[3] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Selecting the 0.1- to 2.0-MeV neutron energy group maximizes dose equivalent and ensures 

favorability to the claimant. 

[4] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Selecting the 0.1- to 2.0-MeV neutron energy group maximizes dose equivalent and ensures 

favorability to the claimant.

[5] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[6] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[7] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[8] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[9] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 75 of 92

[10] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[11] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[12] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[13] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[14] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[15] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[16] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[17] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 76 of 92

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[18] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[19] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[20] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[21] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[22] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[23] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[24] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 77 of 92

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[25] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[26] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[27] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[28] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[29] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[30] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[31] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 78 of 92

[32] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[33] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[34] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[35] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The plot of the neutron energy spectrum in terms of dose equivalent rate was created by 

expanding (rebinning) the spectral data for this location given by MMES (1992). Expanding 

the data allows the contributions to dose equivalent to be established in relation to the IREP 

neutron energy groups and the energy response threshold for NTA film. It also allows the 

spectrum to be plotted in a meaningful manner. The spectra provided by MMES (1992), which 

are reported in terms of average neutron energy, were established by using field 

measurements as input to a spectral unfolding code. The reported spectral data were 

rebinned by Mr. Burns to reflect the range of each energy bin rather than its average. Mr. 

Burns has experience with the unfolding code he used and its neutron energy group structure. 

[36] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These fractions were derived from the expanded spectral data that were used to create the 

supporting graph.

[37] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[38] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[39] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[40] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[41] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[42] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 79 of 92

[43] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[44] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[45] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[46] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[47] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[48] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[49] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[50] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[51] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[52] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[53] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[54] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[55] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[56] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[57] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Data for this analysis came from worker neutron monitoring records.

[58] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These data are directly from Table 6-23.

[59] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

This recommendation is made to ensure favorability to the claimant regardless of when 

neutron exposures occurred.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 80 of 92

[60] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

Assigning the deep dose LOD could result in an underestimate in the case of low-energy 

electrons.

[61] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The mixture of 75% >250-keV photons and 25% 30-to 250-keV photons was selected to reflect 

a field comprised of high-energy and Compton-scattered photons, as would be expected for 

these facilities. For high-energy photon facilities (accelerators), 100% photons >250 keV was 

selected. For facilities where low-energy photons were possible, the 30- to 250-keV group 

was specified to be most favorable to the claimant since the presence of fission products could 

not be ruled out.

[62] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

The values in this table come from Section 6.3.4.2. The 0.1- to 2.0-MeV (default) energy 

group is specified for facilities for which no data were available because this is most favorable 

to the claimant.

[63] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

This is a compilation summary of data in Section 6.3.4.2.

[64] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These data were derived from analysis of worker dosimetry records.

[65] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

These data were derived from analysis of worker dosimetry records.

[66] Burns, Robert E. Shonka Research Associates. Senior Health Physicist. April 2007. 

This is a compilation summary of data in Section 6.3.4.2. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 81 of 92

REFERENCES

Beck, J., P. Stansbury, and J. Watson, 1979, Interview with Ed Gupton on Oak Ridge National 

Laboratory’s Personnel Monitoring Program, August 6, Oak Ridge National Laboratory, Oak 

Ridge, Tennessee. 

Berger, C. D., and B. H. Lane, 1985, Neutron Dosimetry with the ORNL Badge, ORNL/TM-9450, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee.

Cameron, J. R., N. Sunthanalingham, and G. N. Kennedy, 1968, Thermoluminescent Dosimetry, 

University of Wisconsin Press, Madison, Wisconsin.

Craft, H. R., J. C. Ledbetter, and J. C. Hart, 1952, Personnel Monitoring Operating Techniques, 

ORNL-1411, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Davis, D. M., E. D. Gupton, and J. C. Hart, 1954, Applied Health Physics Radiation Survey 

Instrumentation, ORNL-332, Revision 1, Oak Ridge National Laboratory, Oak Ridge, 

Tennessee. 

Deal, L. J., and J. C. Hart, 1949, An Explanation Concerning Personnel Monitoring Records and 

Methods of Filing Prior to January 1, 1947, Oak Ridge National Laboratory, Oak Ridge, 

Tennessee.

DOE (U.S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Office of Environment, Safety and 

Health, Washington, D.C.

Gupton, E. D., 1968, Methods and Procedures for External Radiation Dosimetry at ORNL January 1, 

1968, Central Files Number 68-1-73, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Gupton, E. D., 1969, Methods and Procedures for External Radiation Dosimetry at ORNL January 

1969, Central Files Number 69-1-59, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Gupton, E. D., 1974, “Badge Dosimetry Components for 1975,” Intra-Laboratory Correspondence, 

December 6, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Gupton, E. D., 1975, “Photobadges and Dosimetry Components”, Intra-Laboratory Correspondence, 

August 13, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Gupton, E. D., 1978, Dosimetry with the ORNL Badge, 1978, ORNL/TM-6357, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee.

Gupton, E. D., 1981, Methods and Procedures for External Radiation Dosimetry at ORNL, 

ORNL/TM-7984, Oak Ridge National Laboratory, Oak Ridge, Tennessee, September.

Hart, J. C., 1966, A Progress Report Dealing with the Derivation of Dose Data From ORNL Personnel 

Exposure Records Applicable to the Mancuso Study, Central Files Number 66-1-84, January 

14, Oak Ridge National Laboratory, Oak Ridge, Tennessee. 

Hart, J. C., and P. O. Martin, 1949, Personnel Monitoring Records and Filing at Oak Ridge National 

Laboratory, Health Physics Division Report Number SUM-535-49, May 27, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 82 of 92

Hart, J. C., and B. T. Walters, 1950, Personnel Monitoring Recording and Filing Methods of the Health 

Physics Division of the Oak Ridge National Laboratory as Effected July 2, 1950, Health 

Physics Division Report Number SM-188-50, November 7, Oak Ridge National Laboratory, 

Oak Ridge, Tennessee.

Horowitz, Y. S., 1984, Thermoluminescence and Thermoluminescent Dosimetry, CRC Press, Boca 

Raton, Florida.

IAEA (International Atomic Energy Agency), 1990, Compendium of Neutron Spectra and Detector 

Responses for Radiation Protection Purposes, Technical Report Series No. 318, Vienna, 

Austria.

ICRP (International Commission on Radiological Protection), 1971, Data for Protection Against 

Ionizing Radiation from External Sources: Supplement to ICRP Publication 15, ICRP 

Publication 21, Pergamon Press, Oxford, England. 

ICRP (International Commission on Radiological Protection), 1990, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England. [SRDB Ref ID: 11517]

ICRU (International Commission on Radiation Units and Measurements), 1993, Quantities and Units 

in Radiation Protection Dosimetry, Report No. 51, Bethesda, Maryland. 

Kerr, G. D., 2003, “Neutron Dosimetry and Neutron-to-Photon Ratios for Various ORNL Facilities”, 

letter to R. Burns, November 29, Kerr Consulting Company, Knoxville, Tennessee.

Kerr, G. D., and D. R. Johnson, 1968, Radiation Survey and Dosimeter Intercomparison Study at the 

Health Physics Research Reactor, ORNL-TM-2334, Oak Ridge National Laboratory, Oak 

Ridge, Tennessee.

McLendon, J. D., 1980, External Radiation Monitoring of Union Carbide Nuclear Division Personnel, 

Y/DD-268, Y-12 Plant, Oak Ridge, Tennessee.

Mitchell, T. J., G. Ostrouchov, E. L. Frome, and G. D. Kerr, 1993, A Method for Estimating 

Occupational Radiation Dose to Individuals, Using Weekly Dosimetry Data, ORNL-6778, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee. 

MMES (Martin Marietta Energy Systems, Inc.), 1992, Dosimeter Response Characterization for 

Personnel Neutron Dosimetry (Volumes 1-6), Oak Ridge National Laboratory, Oak Ridge, 

Tennessee. [SRDB Ref IDs: 11927, 11938, 11939, 11966, 11973, and 11976]

Morgan, K. Z., 1961, “Dosimetry Requirements for Protection from Ionizing Radiations,” in Selected 

Topics in Radiation Dosimetry, proceedings of the symposium on selected topics in radiation 

dosimetry, International Atomic Energy Agency, Vienna, Austria. 

NBS (National Bureau of Standards), 1954, Permissible Dose from External Sources of Ionizing 

Radiation, Handbook 59, Gaithersburg, Maryland. 

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation, Report 38, Bethesda, Maryland.
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 83 of 92

NIOSH (National Institute for Occupational Health and Safety), 2006, External Dose Reconstruction 

Implementation Guidelines, OCAS-IG-001, Revision 2, Office of Compensation Analysis and 

Support, Cincinnati, Ohio. 

ORAUT (Oak Ridge Associated Universities Team), 2006, Technical Basis Document for the Y-12 

National Security Complex – Occupational External Dosimetry, ORAUT-TKBS-0014-6, Rev. 

00 PC-3, Oak Ridge, Tennessee, May 11.

ORAUT (Oak Ridge Associated Universities Team), 2007, Hanford Site – Occupational External 

Dose, ORAUT-TKBS-0006-6, Rev. 03, Oak Ridge, Tennessee, November 11.

ORNL (Oak Ridge National Laboratory), 1958, A Study of the Nuclear and Physical Properties of the 

ORNL Graphite Reactor Shield, ORNL-2195, Oak Ridge, Tennessee. 

ORNL (Oak Ridge National Laboratory), 1994, Technical Basis for the Centralized External Dosimetry 

System, Revision 2, Oak Ridge, Tennessee.

Pardue, L. A., N. Goldstein, and R. O. Wollan, 1944, Photographic Film as a Pocket Radiation 

Dosimeter, CH-1553, University of Chicago, Chicago, Illinois.

Parrish, B. R., 1979, Analysis of External Personnel Monitoring Devices and Data for Oak Ridge 

National Laboratory Epidemiological Study, University of North Carolina at Chapel Hill, Chapel

Hill, North Carolina. 

Roberson, J. H., 1947, excerpted material dated December 10, Oak Ridge National Laboratory, Oak 

Ridge, Tennessee. 

Sims, C. S., and G. G. Killough, 1983, “Neutron Fluence-to-Dose Equivalent Conversion Factors: A 

Comparison of Data Sets and Interpolation Methods,” Radiation Protection Dosimetry, volume 

5, number 1, pp. 45–48. 

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, “Study of a Selection of 10 

Historical Types of Dosimeters: Variation of the Response of Hp(10) with Photon Energy and 

Geometry of Exposure,” Radiation Protection Dosimetry, volume 102, number 2, pp. 101–113. 

Thornton, W. T., D. M. Davis, and E. D. Gupton, 1961, The ORNL Badge Dosimeter and its Personnel 

Monitoring Applications, ORNL-3126, Oak Ridge National Laboratory, Oak Ridge, Tennessee. 

Vallario, E. J., D. E. Hankins, and C. M. Unruh, 1969, AEC Workshop on Personnel Dosimetry, 

BNWL-1340, Battelle Pacific Northwest Laboratory, Richland, Washington.

West, C. M., 1992, Description of the ORNL External Monitoring Program 1943 – 1961, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee.

Wirth, J. E., K. Z. Morgan, and H. J. Curtis, 1945, Health Division and Biological Section of the 

Research Division Report for Month Ending February 15,1945, CH-2587, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee. 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 84 of 92

ATTACHMENT A

Selection of IREP Energy Groups for Beta-Gamma Exposures for ORNL Facilities [61]

Page 1 of 3

Process Building/description

Approx. dates of 

operation IREP energy group 

Begin End selection for deep dose

Accelerators 6000: Holifield Ion Beam Facility August, 1975 Present Photons: >250 keV (100%) 

6010: ORELA 1969 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

6000: ORIC 1962 Present Photons: >250 keV (100%) 

Calibration facilities 2007: RASCAL Ca. 1951 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7735: RADCAL (aka CalLab) 1989 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

Isotope/separation 

facilities

3019: Hot Pilot Plant/Radiochemical Development Facility 1943 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3026-C:Radiochem. laboratory/85Kr facility 1944 Ca. 1988 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3026-D:RaLa Bldg./segmentation facility 1945 Ca. 1988 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3028: Radioisotope Processing Building A/Alpha Powder 

Facility

1950 1985 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3029: Source Development Laboratory 1952 Late 1980s Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3030: Radioisotope Processing Building C Ca. 1950 Late 1980s Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3031: Radioisotope Processing Building D Ca. 1950 Late 1980s Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3032: Radioisotope Processing Building E Ca. 1950 Late 1980s Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3033: Radioisotope Processing Building F/Radioactive Gas 

Processing Facility

Ca. 1950 1990 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3038: Isotope Research Materials Laboratory/Alpha 

Handling Facility

1949 1990 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3047: Radioisotopes Development Laboratory 1962 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3505: Metals Recovery Facility 1952 1960 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 85 of 92

ATTACHMENT A

Selection of IREP Energy Groups for Beta-Gamma Exposures for ORNL Facilities [61]

Page 2 of 3

Process Building/description

Approx. dates of 

operation IREP energy group 

Begin End selection for deep dose

3508: Alpha Isolation Laboratory 1952 Ca. 1976 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3515: Fission Product Pilot Plant (F3P) 1948 1958 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3517: Fission Product Development Laboratory 1958 1975 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3525: High Radiation Level Examination Laboratory 1963 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

4501: High-Level Radiochemistry Laboratory 1951 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

4507: High-Level Chemical Development Facility 1957 1980 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

4508: Interim Plutonium Laboratory/Fuel Cycle Alpha Facility 1962 Present Photons: 30–250 keV 

(100%) 

7025: Tritium Target Preparation Facility 1967 Ca. 1989 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7920: REDC 1966 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7930: REDC 1966 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

Reactors 3001: Graphite Reactor Nov. 1943 Nov. 1963 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3005: Low-Intensity Test Reactor (LITR) 1949 Oct. 1968 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3010: Bulk Shielding Reactor/Pool Critical Assembly 1951 1987 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3042: ORNL Research Reactor (ORR) March, 1958 Ca. 1987 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7500: Homogeneous Reactor Experiment (HRE)/ 

Homogeneous Reactor Test (HRT)

1952 1961 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 86 of 92

ATTACHMENT A

Selection of IREP Energy Groups for Beta-Gamma Exposures for ORNL Facilities [61]

Page 3 of 3

Process Building/description

Approx. dates of 

operation IREP energy group 

Begin End selection for deep dose

7503: Molten Salt Reactor Experiment (MSRE) 1965 1969 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7702: Tower Shielding Reactor (TSR) 1954 1992 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7710: Health Physics Research Reactor (HPRR) 1963 1987 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7900: High Flux Isotope Reactor (HFIR) August, 1965 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

9213: Critical Experiments Facility (at Y-12) August, 1950 1987 Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

Waste/storage 

facilities

3023: North Tank Farm 1943 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3039: Central Offgas Facility 1950 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3100: Source/SNM Storage Vault 1960 Present Photons: 30–250 keV 

(100%)

3507: South Tank Farm 1943 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

3513: Waste Holding Basin 1943 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 

7830: SWSA 5 (Melton Valley storage area) 1959 Present Photons: >250 keV (75%)

Photons: 30–250 keV (25%) 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 87 of 92

ATTACHMENT B

Selection of IREP Energy Groups for Neutron Exposures for ORNL Facilities [62]

Page 1 of 2

Process Building/description

Approx. dates of operation IREP energy group selection

Begin End Energy groups (MeV) Dose fractions (%)

Accelerators 6000: Holifield Ion Beam Facility August 1975 Present 0.1–2.0 100 (default)

6010: ORELA 1969 Present <0.01

0.1–2.0

14

86

6000: ORIC 1962 Present 0.1–2.0 100 (default)

Calibration 

facilities

2007: RASCAL (238Pu:Be only) Ca. 1951 Ca. 1990 0.1–2.0

2.0–20.0

29

71

7735: RADCAL: beta room with 

moderated Cf-252 source exposed

1989 Present <0.01

0.1–2.0

16

84

7735: RADCAL: beta room with bare 

Cf-252 source exposed

1989 Present 0.1–2.0

2.0–20.0

69

31

7735: RADCAL: hallway with moderated 

Cf-252 source exposed

1989 Present <0.01

0.1–2.0

2.0–20.0

33

53

14

7735: RADCAL: equipment room with 

moderated Cf-252 source exposed

1989 Present <0.01

0.1–2.0

2.0–20.0

25

57

18

Isotope/separation 

facilities

3038: Isotope Research Materials 

Laboratory (shielded 238Pu:Be 

and Cf-252 sources) 

1949 1990 0.1–2.0

2.0–20.0

42

58

7920: REDC: Room 111 gloveboxes 1966 Present 0.1–2.0

2.0–20.0

76

24

7920: REDC: Room 211 gloveboxes 1966 Present 0.1–2.0

2.0–20.0

75

25

7920: REDC: TDF 1966 Present <0.01

0.1–2.0

12

88

7920: REDC: waste cask 1966 Present <0.01

0.1–2.0

2.0–20.0

14

59

27

7920: REDC: WTA tunnel exit 1966 Present 0.1–2.0 100 
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 88 of 92

ATTACHMENT B

Selection of IREP Energy Groups for Neutron Exposures for ORNL Facilities [62]

Page 2 of 2

Process Building/description

Approx. dates of operation IREP energy group selection

Begin End Energy groups (MeV) Dose fractions (%)

Reactors 3001: Graphite Reactor (general area) Nov. 1943 Nov. 1963 <0.01 100

3005: Low-Intensity Test Reactor (LITR) 1949 Oct. 1968 0.1–2.0 100 (default)

3010: Bulk Shielding Reactor/Pool 

Critical Assembly

1951 1987 0.1–2.0 100 (default)

3042: ORNL Research Reactor (ORR) March 1958 Ca. 1987 0.1–2.0 100 (default)

7500: Homogeneous Reactor Experiment 

(HRE)/ Homogeneous Reactor 

Test (HRT)

1952 1961 0.1–2.0 100 (default)

7503: Molten Salt Reactor Experiment 

(MSRE)

1965 1969 0.1–2.0 100 (default)

7702: Tower Shielding Reactor (TSR) 1954 1992 0.1–2.0 100 (default)

7710: Health Physics Research Reactor 

(HPRR)

1963 1987 0.1–2.0 100

7900: High Flux Isotope Reactor (HFIR) 

(beam room)

August 1965 Present <0.01 100

9213: Critical Experiments Facility (at 

Y-12)

August 1950 1987 0.1–2.0 100 (default)

Waste/storage 

facilities

3100: Source/SNM Storage Vault 

(241Am-oxide drums)

1960 Present 0.1–2.0

2.0–20.0

78

22

SWSA 5 TRU storage bunker: waste 

casks

1959 Present <0.01

0.1–2.0

2.0–20.0

12

70

18

SWSA 5 TRU storage bunker: at wall 

shared with adjacent bunker

1959 Present <0.01

0.1–2.0

2.0–20.0

13

51

36
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 89 of 92

ATTACHMENT C

Neutron-to-Photon Dose Ratios for ORNL Facilities for 100% Occupancy [63]

Process Building/description

Approx. dates of operation Neutron-to-gamma 

Begin End dose ratios

Accelerators 6000: Holifield Ion Beam Facility August, 1975 Present 19

6010: ORELA 1969 Present 2.5

Calibration 

facilities

2007: RASCAL (238Pu:Be only) Ca. 1951 Ca. 1990 28.9

7735: RADCAL: beta room with moderated Cf-252 source 

exposed

1989 Present 3.6

7735: RADCAL: beta room with bare Cf-252 source exposed 1989 Present 15.6

7735: RADCAL: hallway with moderated Cf-252 source exposed 1989 Present 2.1

7735: RADCAL: edge of neutron source storage well 1989 Present 1.2

Isotope/separation 

facilities

3038: Isotope Research Materials Laboratory (shielded 238Pu:Be 

and Cf-252 sources)

1949 1990 3.7

3038: Isotope Research Materials Laboratory (238Pu:Be source in 

Lucite glovebox)

1949 1990 32.2

7920: REDC: Room 111 glovebox (face) 1966 Present 5.8

7920: REDC: Room 211 glovebox (side) 1966 Present 11.7

7920: REDC: Room 211 glovebox (face) 1966 Present 5.3

7920: REDC: 4.6 m behind Room 211 gloveboxes 1966 Present 1.3

7920: REDC: TDF 1966 Present 0.2

7920: REDC: waste cask 1966 Present 4.4

7920: REDC: WTA tunnel exit 1966 Present 6.7

7920: REDC: WTA glovebox 1966 Present 1.2

7930: REDC: low-scatter glovebox 1966 Present 16.7

7920: REDC: Room 208 glovebox (face) 1966 Present 1.3

7920: REDC: Room 208 low dose waiting area 1966 Present 1.3

Reactors 3001: Graphite Reactor (general area) Nov. 1943 Nov. 1963 0.1

7710: Health Physics Research Reactor (HPRR) 1963 1987 1.0

7900: High Flux Isotope Reactor (HFIR) (beam room) August, 1965 Present 0.2

Waste/storage 

facilities

3100: Source/SNM Storage Vault (241Am-oxide drums) 1960 Present 1.2

SWSA 5 TRU storage bunker: waste casks 1959 Present 2.8

NOTE: Geometric mean, minimum, and maximum should be doubled to account for ICRP 60 radiation weighting factors
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 90 of 92

ATTACHMENT D

Neutron-to-Photon Dose Ratios for ORNL Neutron Worker Groups [64]

Neutron

Group Group Description No. Records* Geo. Mean GSD Min. Max.

1 RADCAL workers 4 -- -- 0.6 1.6

2 REDC Analytical Chemistry: Bldg. 7920 glovebox workers 1104 1.6 2.2 0.1 8.8

3 REDC TURF glove box: Bldg. 7930 workers 65 2.5 1.6 1.0 8.1

4 REDC General Area: control room, clerical, and support personnel. 982 1.1 1.7 0.1 3.7

5 HFIR Beam Room 5 -- -- 0.1 0.2

6 Holifield workers (exclusively) 4 -- -- 0.7 1.5

7 D2O-moderated 252Cf 24 0.3 3.2 0.1 2.2

9 Unshielded source handlers 163 1.0 2.8 0.1 14.1

10 Building 3100 storage vault 0 -- -- -- --

11 SWSA 5: storage bunker/WEAF workers 9 1.1 1.9 0.3 2.4

12 General TRU waste handlers 3 -- -- 0.2 1.1

13 Shielded source users: well-loggers, I&C tasks 37 0.5 2.4 0.1 3.5

14 Neutron source make-up 45 1.1 2.2 0.2 9.1

17 REDC TRU decon. facility workers 86 2.1 2.3 0.4 8.0

18 ORELA workers (exclusively) 0 -- -- -- --

19 ORELA and Holifield: staff whom work at both facilities 3 -- -- 0.7 1.9

21 SWSA 5: Bldg. 7879 drum storage handlers 148 0.5 1.9 0.1 2.8

22 R&D (primarily I&C techs.) 2 -- -- 1.1 1.4

all All groups combined 2684 1.2 2.2 0.1 14.1

*neutron and photon dose both ≥ 10 mrem

Neutron:Photon Dose Ratio

NOTE: Geometric mean, minimum, and maximum should be doubled to account for ICRP 60 radiation weighting factors
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 91 of 92

ATTACHMENT E

Neutron Dose Data for ORNL Neutron Worker Groups 1990–2004 [65]

Neutron Fraction of results

Group Group Description ≥ 10 mrem Geo. Mean GSD Max.

1 RADCAL workers 1% (5 out of 406) -- -- 28

2 REDC Analytical Chemistry: Bldg. 7920 glovebox workers 78% (1197 out of 1535) 50 2.2 584

3 REDC TURF glove box: Bldg. 7930 workers 98% (79 out of 81) 42 1.8 137

4 REDC General Area: control room, clerical, and support personnel. 39% (1070 out of 2730) 23 1.6 91

5 HFIR Beam Room 0.1% (5 out of 7992) -- -- 14

6 Holifield workers (exclusively) 2% (8 out of 407) -- -- 28

7 D2O-moderated 252Cf 3% (37 out of 1329) 14 1.3 27

9 Unshielded source handlers 24% (405 out of 1687) 24 1.8 212

10 Building 3100 storage vault 2% (1 out of 53) -- -- 11

11 SWSA 5: storage bunker/WEAF workers 8% (38 out of 448) 14 1.3 26

12 General TRU waste handlers 1% (3 out of 501) -- -- 30

13 Shielded source users: well-loggers, I&C tasks 7% (44 out of 675) 18 1.6 53

14 Neutron source make-up 32% (72 out of 222) 25 1.8 182

17 REDC TRU decon. facility workers 42% (111 out of 264) 37 2.3 230

18 ORELA workers (exclusively) 0% (0 out of 191) -- -- --

19 ORELA and Holifield: staff whom work at both facilities 1% (4 out of 290) -- -- 47

21 SWSA 5: Bldg. 7879 drum storage handlers 14% (160 out of 1148) 17 1.5 58

22 R&D (primarily I&C techs.) 23% (11 out of 48) -- -- 17

all All groups combined 16% (3252 out of 20037) 31 2.1 584

Neutron Dose (mrem)

NOTE: Geometric mean and maximum should be doubled to account for ICRP 60 radiation weighting factors
Document No. ORAUT-TKBS-0012-6 Revision No. 01 Effective Date: 09/10/2007 Page 92 of 92

ATTACHMENT F

Fractions of Dose Equivalent above Energy Cutoff for NTA Film for ORNL Facilities [66]

Process Building/description

Approx. dates of 

operation

Fraction of dose equivalent 

above the energy cutoff for 

Begin End NTA film (500 keV) (%)

Accelerators 6010: ORELA 1969 Present 65

Calibration facilities 2007: RASCAL (Pu-238:Be only) Ca. 1951 Ca. 1990 94

7735: RADCAL: beta room with moderated Cf-252 source 

exposed

1989 Present 76

7735: RADCAL: beta room with bare Cf-252 source exposed 1989 Present 82

7735: RADCAL: hallway with moderated Cf-252 source 

exposed

1989 Present 29

7735: RADCAL: equipment room with moderated Cf-252 

source exposed

1989 Present 61

Isotope/separation 

facilities

3038: Isotope Research Materials Laboratory (shielded 

238Pu:Be and Cf-252 sources) 

1949 1990 92

7920: REDC: Room 111 glovebox (face) 1966 Present 91

7920: REDC: Room 211 glovebox (side) 1966 Present 94

7920: REDC: TDF 1966 Present 0

7920: REDC: waste cask 1966 Present 72

7920: REDC: WTA tunnel exit 1966 Present 87

Reactors 3001: Graphite Reactor (general area) Nov. 1943 Nov. 1963 0 (estimated)

7900: High Flux Isotope Reactor (HFIR) (beam room) August, 1965 Present 0

Waste/storage 

facilities

3100: Source/SNM Storage Vault (241Am-oxide drums) 1960 Present 80

SWSA 5 TRU storage bunker: waste casks 1959 Present 68

SWSA 5 TRU storage bunker: at wall shared with adjacent 

bunker

1959 Present 65