ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 56

DOE Review Release 06/23/2009

Document Title:

Y 12 National Security Complex – Occupational 

External Dosimetry

Document Number: ORAUT-TKBS-0014-6 

Revision: 01

Effective Date: 06/02/2009

Type of Document TBD

Supersedes: Revision 00 PC-3 

Subject Expert(s): George D. Kerr

Site Expert(s): N/A

Approval: John M. Byrne Signature on File for Approval Date: 05/15/2009

William E. Murray, Document Owner

Concurrence: Signature on File Concurrence Date: 05/15/2009

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 05/15/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 05/21/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 06/02/2009

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-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 2 of 56

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/19/2003 00 New document to establish occupational external dosimetry for the 

Y-12 National Security Complex. First approved issue. Initiated by 

William E. Murray.

10/11/2005 00 PC-1 Page change initiated to incorporate the definition of U.S.C. on 

page 7 and details for the definition of a DOE facility on page 8. No 

sections were deleted. First approved page change revision for 

Revision 00. Retraining is not required. Initiated by William E. 

Murray. 

Approval:

William E. Murray, TBD Team Leader

Signature on File 10/06/2005

Judson L. Kenoyer, Task 3 Manager

Signature on File 10/04/2005

Richard E. Toohey, Project Director

Signature on File 10/04/2005

James W. Neton, Associate Director for Science

Signature on File 10/11/2005

02/14/2006 00 PC-2 Approved page change revision. Page changes initiated to 

incorporate definitions and directions for dose reconstruction for nonpresumptive cancers that are excluded from the 1943 through 1947 

Special Exposure Cohort. Initiated by William E. Murray. Affected 

pages are: page 7, pages 8 and 9 in Section 6.1. No sections were 

deleted. Training required: As determined by the Task Manager. 

Initiated by William E. Murray. 

Approval:

William E. Murray, TBD Team Leader

Signature on File 02/09/2006

Judson L. Kenoyer, Task 3 Manager

Signature on File 02/09/2006

Kate Kimpan, Project Director

Signature on File 02/07/2006

James W. Neton, Associate Director for Science

Signature on File 02/14/2006

05/11/2006 00 PC-3 Approved page change revision. Updates required language in 

Introduction (Section 6.1) on page 9. Revised to incorporate 

additional directions for dose reconstruction for non-presumptive 

cancers that are excluded from the 1943 through 1947 Special 

Exposure Cohort on page 10 Section 6.1. Revised to delete text on 

page 10 in Section 6.1, specified in comments from OCAS review. 

No sections were deleted. Training required: As determined by the 

Task Manager. Initiated by William E. Murray. 

Approval: 

William E. Murray, TBD Team Leader

Signature on File 04/10/2006
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 3 of 56

Judson L. Kenoyer, Task 3 Manager 

John M. Byrne Signature on File for 04/10/2006

Edward F. Maher, Task 5 Manager

Signature on File 05/10/2006

Kate Kimpan, Project Director

Signature on File 05/02/2006

James W. Neton, Associate Director for Science

Signature on File 05/11/2006

06/02/2009 01 Revision to add Section 6.1.3.3 and Attachment B to address the

Special Exposure Cohort (SEC-00098). Updates required language 

and adds purpose and scope in Section 6.1. Revised to incorporate 

additional directions for dose reconstruction for nonpresumptive 

cancers that are excluded from the March 1, 1943, through 

December 31, 1947, Special Exposure Cohort in Section 6.1.3.3. 

Adds Attributions and Annotations Section. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by William E. Murray. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 4 of 56

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

8

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

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

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

6.1.3 Special Exposure Cohort Designations.......................................................................13

6.1.3.1 March 1943 to December 1947.................................................................... 13

6.1.3.2 January 1948 to December 1957 ................................................................. 13

6.1.3.3 March 1, 1943 to December 31, 1947 .......................................................... 14

6.2 Basis of Comparison..............................................................................................................14

6.3 Dose Reconstruction Parameters........................................................................................... 15

6.3.1 Y-12 Historical Administrative Practices .....................................................................15

6.3.2 Y-12 Dosimetry Technology .......................................................................................18

6.3.2.1 Beta-Photon Dosimeters .............................................................................. 18

6.3.2.2 Neutron Dosimeters ..................................................................................... 22

6.3.3 Y-12 Dosimeter Calibration Procedures......................................................................22

6.3.4 Y-12 Workplace Radiation Fields................................................................................23

6.3.4.1 Workplace Beta/Photon Dosimeter Response.............................................. 23

6.3.4.2 Workplace Neutron Dosimeter Response .................................................... 25

6.3.4.2.1 Calibration Laboratory in Building 9983 ...................................... 25

6.3.4.2.1.1 Neutron Energy Spectrum..................................... 25

6.3.4.2.1.2 Neutron-to-Photon Dose Ratio .............................. 25

6.3.4.2.2 Enriched Uranium Storage Area in Building 9212 ....................... 26

6.3.4.2.2.1 Neutron Energy Spectrum..................................... 27

6.3.4.2.2.2 Neutron-to-Photon Dose Ratio .............................. 27

6.3.4.2.3 Nondestructive Assay Laboratory in Building 9720-5 .................. 28

6.3.4.2.3.1 Neutron Energy Spectrum..................................... 28

6.3.4.2.3.2 Neutron-to-Photon Dose Ratio .............................. 28

6.4 Adjustments of Recorded Dose.............................................................................................. 30

6.4.1 Photon Dose Adjustments ..........................................................................................30

6.4.2 Neutron Dose Adjustments.........................................................................................31

6.4.2.1 Neutron Weighting Factors........................................................................... 31

6.4.2.2 Neutron Dose Correction Factors................................................................. 32

6.5 Missed Dose ..........................................................................................................................33

6.5.1 Photon Missed Dose ..................................................................................................33

6.5.2 Neutron Missed Dose ................................................................................................. 35

6.6 Uncertainty in Photon and Neutron Dose ............................................................................... 36

6.7 Organ Dose............................................................................................................................36

6.8 Attributions and Annotations .................................................................................................. 37

References ........................................................................................................................................38
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 5 of 56

ATTACHMENT A, OCCUPATIONAL EXTERNAL DOSE FOR MONITORED 

WORKERS .......................................................................................................... 46

ATTACHMENT B, OCCUPATIONAL EXTERNAL DOSE FOR UNMONITORED 

WORKERS .......................................................................................................... 54
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 6 of 56

LIST OF TABLES

TABLE TITLE

6-1 Monitoring technique and exchange frequency at Y-12 Plant for external WB 

exposures ..............................................................................................................................

PAGE

16

6-2 Dosimeter type, period of use, exchange frequency, laboratory MDL, and 

maximum annual missed dose............................................................................................... 17

6-3 Y-12 historical dosimetry events............................................................................................. 19

6-4 IARC study results for U.S. beta/photon dosimeters............................................................... 21

6-5 Selection of beta and photon radiation energies and percentages for Y-12 site 

processes ..............................................................................................................................24

6-6 Dose fractions for Y-12 calibration laboratory......................................................................... 26

6-7 Neutron dose fractions for Y-12 EU storage area................................................................... 27

6-8 Dose fractions for Y-12 Nondestructive Analysis Laboratory .................................................. 29

6-9 Typical workplace neutron dosimeter Hp(10) performance ..................................................... 29

6-10 Number of neutron monitored workers, cumulative neutron dose, and 

average neutron dose to Y-12 workers for 12-year period after introduction of 

the four-element TLND dosimeter in 1989.............................................................................. 30

6-11 Adjustments to reported Y-12 deep photon dose ................................................................... 31

6-12 Neutron quality factor, Q, or weighting factor, wr .................................................................... 32

6-13 Summary of neutron dose fractions and associated ICRP Publication 60 

correction factors for Y-12 facilities ........................................................................................ 33

6-14 Default exposure geometries favorable to the claimant for calculating organ 

doses .....................................................................................................................................37

A-1 IREP dose parameter input screen ........................................................................................ 47

A-2 Selection of beta and photon radiation energies and percentages for Y-12 site 

processes ..............................................................................................................................47

A-3 Selection of neutron energies and percentages for Y-12 site facilities.................................... 48

A-4 Adjustments to reported Y-12 deep photon dose ................................................................... 49

A-5 Neutron dose fractions and associated ICRP 60 correction factors for Y-12 

site facilities............................................................................................................................50

A-6 Dosimeter type, period of use, exchange frequency, laboratory MDL, and 

maximum annual missed dose............................................................................................... 51

A-7 Default exposure geometries favorable to the claimant for calculating organ 

dose.......................................................................................................................................52

B-1 Cubicle x-ray radiation measurements ................................................................................... 55

B-2 Bounded WB dose equivalent for calutron-related uranium enrichment 

workers, March 1, 1943, to December 31, 1947..................................................................... 56
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 7 of 56

LIST OF FIGURES

FIGURE TITLE

6-1 Percentage of Y-12 workers monitored for external radiation exposure, 1948 

through 1961..........................................................................................................................

PAGE

12

6-2 Comparison of Hp(10) from a broad beam of normally incident photons with 

the energy responses for a sensitive DuPont 502 emulsion in a MED twoelement film badge versus a sensitive DuPont 555 emulsion in an ORNL 

multielement film badge ......................................................................................................... 21

6-3 Comparison of Hp(10) from normally incident neutrons to the energy 

responses of an NTA film and a neutron albedo dosimeter with a neutron 

TLND chip made of 6

LiF shielded by cadmium....................................................................... 22

6-4 Results of PNL neutron spectrum measurements in the Calibration 

Laboratory of Building 9983 and the PNL measurement results............................................. 26

6-5 Results of PNL neutron spectrum measurements in the enriched uranium 

storage area of Building 9212 and the PNL measurement results divided into 

the four neutron energy groups used in the dose reconstruction for Y-12 

workers ..................................................................................................................................27

6-6 Results of PNL neutron spectrum measurements made at 39 in. from a bare 

252Cf fission neutron source and the PNL measurement results divided into 

the four neutron energy groups used in the dose reconstruction for Y-12 

workers at the Nondestructive Analysis Laboratory in Building 9720-5 ................................... 28

6-7 Maximum and average photon deep dose to Y-12 workers, 1978 to 1987 ............................. 31

6-8 Comparison of the neutron quality factors in the PNL neutron spectrum 

measurements and the neutron quality factors from NCRP Report 38 shown 

both as point-wise data and grouped averaged data over the four neutron 

energy groups for dose reconstruction for Y-12 workers ........................................................ 32
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 8 of 56

ACRONYMS AND ABBREVIATIONS

ABRWH Advisory Board on Radiation and Worker Health

AEC U.S. Atomic Energy Commission

ANSI American National Standards Institute

A-P anterior-posterior

CATI computer-assisted telephone interview

C.F.R. Code of Federal Regulations

Ci curie (a unit of radioactivity)

cm centimeter

d day 

DCF dose conversion factor

DHHS U.S. Department of Health and Human Services

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

ft foot

g gram

hr hour

HEU highly enriched uranium

HPRR Health Physics Research Reactor

IAEA International Atomic Energy Agency

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

ISO International Standards Organization

K-25 Oak Ridge Gaseous Diffusion Plant

keV kiloelectronvolt

kV kilovolt

MDL minimum detection limit 

MED Manhattan Engineer District

MeV megaelectronvolt

mg milligram

mm millimeter

min minute

MMES Martin Marietta Energy Systems

mrem millirem

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 9 of 56

NTA nuclear track emulsion, type A

OD optical density

ORAU Oak Ridge Associated Universities

ORGDP Oak Ridge Gaseous Diffusion Plant (K-25)

ORISE Oak Ridge Institute for Science and Education

ORNL Oak Ridge National Laboratory

PIC pocket ionization chamber

PNAD personal neutron accident dosimeter

PNL Pacific Northwest Laboratory

POC probability of causation

R roentgen

RADCAL Radiation Calibration Laboratory at ORNL

REF radiation effectiveness factor

RPG radiation protection guideline

SD standard deviation

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SSN Social Security Number

TBD technical basis document

TEC Tennessee Eastman Corporation

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

UCC Union Carbide Corporation

UCC-ND Union Carbide Corporation-Nuclear Division

U.S.C. United States Code

WB whole body

wk week

Y-12 Y-12 Plant or Y-12 National Security Complex

§ section or sections 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 10 of 56

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

The statute also 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 excludes Naval Nuclear Propulsion Facilities from being covered under the Act, 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 occupational radiation exposures are considered valid for inclusion in a 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 (NIOSH 2007): 

] 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, 42 C.F.R. pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

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

conventional structures

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 11 of 56

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

reasons

6.1.1

The purpose of this technical basis document (TBD) is to describe the external dosimetry practices 

and systems at the Y-12 Plant (now the Y-12 National Security Complex). This document discusses 

historical and current practices in relation to the evaluation of external exposure data for monitored 

and unmonitored workers.

Purpose

6.1.2

The Y-12 Plant was first conceived in the fall of 1942 by the Manhattan Engineer District (MED) of the 

U.S. Army Corps of Engineers. The construction of the first building was completed in 1943 (Wilcox 

2001, pp. 8–13

Scope

2

In May 1947, management of Y-12 was assigned to the Carbide and Chemicals Company, a division 

of the Union Carbide and Carbon Corporation, and emphasis was directed away from enrichment to 

the fabrication of nuclear weapon parts. Numerous changes have occurred over the years in the 

fabrication procedures, but the general features have remained the same. Typically, enriched 

uranium (EU) was received at Y-12 in the form of UF6, converted to UF4, reduced to a metal, and then 

fabricated into weapon parts. These fabrication processes involved metal casting, rolling and forming, 

machining, and recycling of the EU salvage. The fabrication of weapon parts was expanded over the 

years to include other radioactive and non-radioactive materials. In addition to facility monitoring to 

measure and control the radiation exposure to workers, an external dosimetry program was started in 

1948 to monitor individual personnel working in the Assay Laboratories, Radiographic Shop, 

Spectrographic Shop, and the “Metal” Machine Shops (Murray 1948a, p. 3; Murray 1948b, p. 3; 

Struxness 1948, p. 8; Struxness 1949a, p. 9; 1949b, p. 6). This program, which monitored less than 

25% of the total number of Y-12 employees, was continued through the time that the criticality 

accident occurred at Y-12 in 1958 (Watkins et al. 1993, p. 39). As a result of the 1958 criticality 

accident, a program was instituted in 1961 to individually monitor all Y-12 workers for external 

radiation exposure using a dosimeter system that was an integral part of the worker’s identification 

badge and contained components for both routine and accident dosimetry (West 1993, p. 7). 

Therefore, Y-12 has used both facility monitoring and individual worker monitoring to measure and 

control radiation exposures to radiation workers since 1948. The percentages of Y-12 workers who 

). The Tennessee Eastman Corporation (TEC) operated Y-12 between June 1943 

and May 1947. During this period, the operations at Y-12 primarily involved the use of the 

electromagnetic separation process to enrich uranium in 235U; Y-12 shipped the enriched product to 

Los Alamos for production of atomic weapons. Until the latter part of 1945, Y-12 converted UO3 to 

UCl4, which was subsequently enriched in 235U by the electromagnetic separation process using two 

calutron stages (termed “alpha” and “beta”). In late 1945, Y-12 discontinued the use of the alpha 

calutron stage, and processes at Y-12 were changed to receive UF6 from the Oak Ridge Gaseous 

Diffusion Plant (ORGDP) or K-25 Plant. The UF6 was then further enriched at Y-12 by the beta 

calutrons and shipped to Los Alamos. In these early days of Y-12, TEC relied entirely on facility 

monitoring to measure and control the radiation exposure to workers (Murray 1948a, p. 476; Murray 

1948b, p. 469). The nature of the work at Y-12 in these early years resulted in internal occupational 

exposure to uranium dust being more important than occupational external exposure (Souleyrette 

2003a, p. 1).

 

2 Page numbers provided in the text are those specific to the referenced report. Page numbers that may be useful in 

locating a specific report within a Site Research Database (SRDB) file are provided after the SRDB Ref ID number in the

reference list.
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 12 of 56

were monitored for external radiation exposure from the start of the external dosimetry program in 

1948 through the switch to monitoring nearly all workers in 1961 are shown in Figure 6-1. The

external monitoring data for Y-12 workers from 1948 to 1950 are not readily available by Social 

Security Number (SSN) and are not being supplied by Y-12 in response to EEOICPA requests 

(Souleyrette 2003a, p.1). These 1948 to 1949 external monitoring data were previously made 

available to Oak Ridge Associated Universities (ORAU), along with other data for use in 

epidemiological studies of Y-12 and other DOE sites in Oak Ridge, Tennessee (Watkins et al. 1993, 

p. 15). The data have recently been analyzed and made available for use in dose reconstruction for 

Y-12 workers (ORAUT 2005, p. 5).

1945 1950 1955 1960

Year

0

5

10

15

20

25

External Monitoring of Y-12 Workers (%)

Figure 6-1. Percentage of Y-12 workers monitored for external 

radiation exposure from start of external monitoring program in 

1948 through the switch to monitoring nearly all workers in 1961

(Watkins et al. 1993, p. 39).

There are numerous Y-12 records about facility monitoring, safety evaluations, investigations, and 

other radiation safety practices. It is time consuming to find and evaluate these records for all Y-12 

facilities and processes since 1943. Evaluations of the extensive scope of facility, process, and 

worker information relevant to an individual worker’s potential dose, many years or even decades 

after employment, are difficult or even impossible in some instances. 

Records of radiation dose to individual workers from personnel dosimeters worn by the worker and 

coworkers are available for some employees from 1948 to 1950, for employees with the highest 

potential for external radiation exposure from 1950 to 1961, for all workers from 1961 to 1996, and to 

workers entering radiological areas after 1996 (West 1993; Souleyrette 2003a, p. 2; ORAUT 2005). 

The information in this section pertains to the analysis of these records but does not address 

parameters for skin and testicular or breast radiation dose that could result from acute exposure to 

beta particles in generally nonroutine workplace exposure situations.

Radiation dosimetry practices were initially based on experience during several decades of radium 

and X-ray usage in medical diagnostic and therapeutic applications. These methods were generally 

well advanced at the start of the MED project to develop nuclear weapons in the 1940s (Morgan 1961; 

Taylor 1971). The primary difficulties MED encountered in its efforts to measure worker doses to 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 13 of 56

external radiation were (1) the large quantities of high-level radioactivity that were not encountered 

previously, (2) the mixed radiation fields of beta particles and photons (X-rays and gamma rays), and 

(3) neutrons with a broad spectrum of energies (Morgan 1961, p. 5). 

6.1.3

6.1.3.1 March 1943 to December 1947

Special Exposure Cohort Designations

Based on the findings and recommendations of NIOSH and the Advisory Board on Radiation and 

Worker Health (ABRWH), the Secretary of the U.S. Department of Health and Human Services 

(DHHS) has designated a class of Y-12 employees as an addition to the Special Exposure Cohort 

(SEC): 

DOE employees or contractor or subcontractor employees who worked in uranium enrichment 

operations or other radiological activities at the Y-12 facility in Oak Ridge, Tennessee, from March 

1943 through December 1947, and who were employed for a number of work days aggregating at 

least 250 work days either solely under this employment or in combination with work days within the 

parameters (excluding aggregate work day requirements) established for other classes of employees 

included in the SEC (DHHS 2005).

NIOSH, however, has determined that, for the period from March 1943 to December 1947, the only 

radiological exposures at Y-12 for which doses can be reconstructed are those that resulted from 

occupational medical X-rays.

From 1948 onward, the internal dose from uranium, whether occupational or environmental, can be 

reconstructed using the available information, which includes environmental data, bioassay results, 

site technical information, and coworker data.

6.1.3.2 January 1948 to December 1957

Based on the findings and recommendations of NIOSH and the ABRWH, the Secretary of DHHS

designated the following class of employees as an addition to the SEC: 

DOE employees or contractor or subcontractor employees who were monitored or should have been 

monitored for:

1. Thorium exposures while working in Buildings 9201-3, 9202, 9204-1, 9204-3, 9206, or 9212 at 

Y-12 for a number of work days aggregating at least 250 work days from January 1948 

through December 1957 or in combination with work days within the parameters (excluding 

aggregate work day requirements) established for one or more classes of employees in the 

SEC; or

2. Radionuclide exposures associated with cyclotron operations in Building 9201-2 at Y-12 for a 

number of work days aggregating at least 250 work days from January 1948 through 

December 1957 or in combination with work days within the parameters (excluding aggregate 

workday requirements) established for one or more classes of employees in the SEC (DHHS

2006).

NIOSH has determined that it is possible to reconstruct or bound occupationally required medical 

doses and external gamma, beta, and neutron doses. NIOSH also found that sufficient bioassay data 

from January 1948 to December 1957 are available for reconstruction of:
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 14 of 56

1. Internal doses for workers with potential for exposure to uranium or recycled uranium 

contaminants (238Pu (239Pu in lesser quantities), 237Np, and 99Tc); and

2. Internal doses for workers involved in plutonium operations when plutonium was enriched with 

the calutrons.

6.1.3.3 March 1, 1943 to December 31, 1947

On August 15, 2008, as provided for under 42 U.S.C. § 7384q(b), the Secretary of DHHS designated 

the following class of employees as an addition to the SEC: 

All employees of DOE, its predecessor agencies, and DOE contractors or subcontractors who worked 

at the Y-12 Plant in Oak Ridge, Tennessee, from March 1, 1943, through December 31, 1947, for a 

number of work days aggregating at least 250 work days occurring either solely under this 

employment or in combination with work days within the parameters established for one or more 

classes of employees in the SEC (DHHS 2008).

NIOSH found that it is not possible to reconstruct radiation doses for these employees for all potential 

radiation exposures. However, NIOSH has decided that the occupational medical dose and external 

exposures directly associated with the calutron uranium enrichment process can be estimated. 

NIOSH will use limited available external dose rate measurements that might be applicable to an 

individual claim (and can be interpreted using existing NIOSH dose reconstruction processes or 

procedures). Therefore, dose reconstructions for individuals with nonpresumptive cancers or 

employment of fewer than 250 work days in the class period may be performed using these data as 

appropriate. (Attachment B contains details on reconstructing the external photon dose.)

6.2 BASIS OF COMPARISON

Since the creation of the MED in the early 1940s, various radiation dose concepts and quantities have 

been used to measure and record occupational dose. A basis of comparison for dose reconstruction 

is the personal dose equivalent, Hp(d), where d identifies the depth in (mm) and represents the point 

of reference for dose in tissue. For weakly penetrating radiation of significance to skin dose, d = 0.07 

mm and is noted as Hp(0.07). For penetrating radiation of significance to whole body (WB) dose, d = 

10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities recommended 

for use as the operational quantity to be recorded for radiological protection purposes by the 

International Commission on Radiation Units and Measurements (ICRU) in Report 51 (ICRU 1993). 

In addition, Hp(0.07) and Hp(10) are the radiation quantities that are used in the DOE Laboratory 

Accreditation Program (DOELAP) to accredit DOE personnel dosimetry systems since the 1980s 

(DOE 1986). The International Agency for Research on Cancer (IARC) also selected Hp(10) as the 

quantity to assess error in historical, recorded WB dose for workers in the IARC nuclear worker 

epidemiologic studies (Thierry-Chef et al. 2002, p. 5). 

The basis for comparison for neutron radiation is more complicated because the calibration of 

dosimeters to measure neutron dose was historically based on different dose quantities such as first 

collision dose, multiple collision dose, and dose equivalent index. The numerical difference in using 

these dose quantities in comparison with the Hp(10) dose in current DOELAP performance testing 

could be evaluated by using the relative values of the dose conversion factors (DCFs) for the 

respective dose quantities in conjunction with characteristics of the respective Y-12 neutron dosimeter 

response characteristics and workplace radiation fields (DOE 1986).
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 15 of 56

6.3 DOSE RECONSTRUCTION PARAMETERS

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

exposure geometries, and the characteristics of the respective Y-12 dosimeter response, are crucial 

for the assessment of bias and uncertainty of 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) (Oxley 2001, pp. 37–49). The performance of current dosimeters can often be compared with 

performance characteristics of historical dosimetry systems in the same, or highly similar, facilities or 

workplaces. In addition, current performance testing techniques can be applied to earlier dosimetry 

systems to achieve a consistent evaluation of all dosimetry systems (Wilson et al. 1990). 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 (2002) guidelines depend on (Fix et al, 1997a): 

• Administrative practices to calculate and record personnel dose based on technical, 

administrative, and statutory compliance considerations.

• Dosimetry technology including the physical capabilities of the dosimetry system, such as 

the response to radiation type and energy, especially in mixed radiation fields.

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

of calibration to sources of exposure in the workplace.

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

geometries, and environmental conditions.

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

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

overall uncertainty. 

6.3.1

A dosimetry program was started in 1948 to monitor individual external exposures of personnel 

working in the Assay Laboratories, Radiographic Shop, Spectrographic Shop, and “Metal” Machine 

Shops where uranium metals were handled (Murray 1948a, p. 3; Murray 1948b, p. 3; Struxness 1948, 

p. 8; Struxness 1949a, p. 9). At first, the external radiation monitoring was performed using pocket 

ionization chambers (PICs), typically exchanged on a weekly basis (Souleyrette 2003a, p. 1). Early 

efforts were concerned with using a photographic film pad on the hands of the uranium metalworkers

and attempting to correlate the film pad reading with WB exposures, which were recorded first with

PICs and later with personnel WB film badge dosimeters (Souleyrette 2003a, p. 1). The Y-12 Plant 

used film badges for external monitoring of exposures to beta particles and photons (X-rays and 

gamma rays) (Souleyrette 2003a, p. 2; West 1993, p. 11). In 1950, they began to use nuclear track 

emulsion, Type A (NTA) film for monitoring worker exposures to fast neutrons (Struxness 1954, 

p. 31).

Y-12 Historical Administrative Practices

In 1980, the film badge dosimeters for beta particles and photons were replaced with 

thermoluminescent dosimeters (TLDs) (McLendon et al. 1980, pp. 5–7; Howell and Batte 1982, p. 6). 

Thermoluminescent neutron dosimeters (TLNDs) were added to NTA badge dosimeters for neutrons 

(Gupton 1978, p. 3; Berger and Lane 1985, p. 3). The TLND was used to determine the doses from
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 16 of 56

thermal and intermediate energy neutrons, and the NTA films were maintained to determine the doses 

from fast neutrons (Gupton 1978, pp. 8–9). 

These combination TLND and NTA dosimeters were replaced in 1990 with newer dosimeters 

containing TLDs for the measurement of dose from beta particles and photons and TLNDs for the 

measurement of doses from neutrons of all energies (Oxley 2001, pp. 3–6; McMahan 1991). The 

frequency of exchange of these personnel dosimeters is summarized in Table 6-1 (Souleyrette 2003a; 

Watkins et al. 1993; West 1993).

Table 6-1. Monitoring technique and exchange frequency at Y-12 Plant for external WB exposures.

a

Period Monitoring technique Exchange frequency Monitored personnel

Beta/photon dosimeters

1948–1950 PICs and two-element 

film dosimeters

Some daily, some weekly Personnel expected to receive 

over 10% of the radiation 

protection guideline (RPG) 

1950–04/07/58 Two-element film 

dosimeters

Weekly Personnel expected to receive 

over 10% of the RPG

04/08/58–06/30/61 Two-element film 

dosimeters

Monthly Personnel expected to receive 

over 10% of the RPG

06/30/61–10/01/80 Four-element film 

dosimeters

Quarterly Nearly all personnel monitored

10/01/80–01/03/89 Two-element TLDs Some quarterly, some 

annually, a very limited 

group on a monthly basis

Quarterly exchange for 

personnel expected to receive 

500 mrem or more, annual 

exchange for personnel 

expected to receive less than 

500 mrem

01/03/89–present Four-element TLDs Mostly quarterly, some 

monthly

Nearly all personnel monitored 

from 1989 to 1996. After 1996, 

only personnel entering 

radiological areas.

Neutron dosimeters

1950–10/01/80 NTA film Biweekly, monthly, and 

quarterly

Personnel exposed to neutron 

sources

10/01/80–01/03/89 NTA film for fast neutrons 

and TLND dosimeters for 

other energy neutrons

Quarterly Personnel exposed to neutron 

sources

01/03/89–present TLNDs for neutrons of all 

energies

Quarterly Personnel exposed to neutron 

sources

a. Souleyrette (2003a), Watkins et al. (1993), West (1993).

The minimum detection level (MDL) of the various dosimeters used at Y-12 to monitor for 

beta/gamma and neutron exposures of the whole body is summarized in Table 6-2. The first film 

dosimeter that Y-12 used was identical to the badge used at the Oak Ridge National Laboratory 

(ORNL) in 1949 (West 1993, p. 2), which is described by Thornton, Davis, and Gupton (1961, 

pp. 2-4). This film badge was a U.S. Atomic Energy Commission (AEC) PF-1B film badge 

manufactured by the A. M. Samples Machine Company in Knoxville, Tennessee (Patterson, West, 

and McLendon 1957, p.21). It had an open window over a portion of the film to measure both beta 

and photon radiation, and a 1-mm cadmium filter over a portion of the film to measure only higher 

energy photon radiation (Handloser 1959, p. 152). This film dosimeter was used at Y-12 until 1961, 

when a newer film dosimeter was developed for use at all Union Carbide Corporation-Nuclear Division 

(UCC-ND) facilities (Thornton, Davis, and Gupton 1961, pp. 2–8). This newer film dosimeter served 

as an identification badge and also provided for both personnel routine and accident monitoring. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 17 of 56

Some of the MDLs in Table 6-2 were difficult to estimate, particularly for the film dosimeters. For the 

current TLD dosimeters, the MDLs are precisely identified in Technical Basis for the External 

Dosimetry Program at the Y-12 National Security Complex (Oxley 2001, pp. 43–45), which is based 

on a DOELAP protocol (DOE 1986). For the film dosimeters, the MDLs are subject to additional 

uncertainty because factors that involve the radiation field, film type, processing, developing, and 

reading system cannot be tested. The MDLs for the film dosimeters in Table 6-2 were based on 

information from Souleyrette (2003a, pp. 5–6), Watkins et al. (1993, pp. 28–29), West (1993), and 

Wilson et al. (1990, p. 7.6).

Table 6-2. Dosimeter type, period of use, exchange frequency, laboratory MDL, and maximum 

annual missed dose.

a

Dosimeter Period

Exchange 

frequency

Laboratory MDL 

(mrem)

Maximum annual 

missed dose (mrem)

Beta/photon dosimeters

Pocket ionization chamber 1948–1950 Daily <5 1,300

Weekly <5 260

Two-element film badge 1948–1958 Weekly 40 2,080

1958–1961 Monthly 40 480

Four-element film badge 1961–1980 Quarterly 40 160

Two-element TLD dosimeter 1980–1989 Quarterly 20 80

Four-element TLD dosimeter 1989–Present Quarterly 10 40

Neutron dosimeters

NTA film 1948–1980 Biweekly <50 1,300b

Monthly <50 600b

Quarterly <50 200b

Combination NTA film and TLND 

dosimeter

1980–1989 Quarterly <50 200b

TLND dosimeter 1989–Present Quarterly 10 40

a. Souleyrette (2003a, pp. 5–6), Watkins et al. (1993. pp. 28–29), West (1993), and Wilson et al. (1990, p. 7.6).

b. Potential annual missed dose based on data from laboratory irradiations may not be directly applicable to workplace 

missed dose (Wilson et al. 1990, p. 7.6; IAEA 1985, pp. 63-71).

The Y-12 Plant administration practices that are important to dose reconstruction include the following 

policies for:

• Assigning dosimeters to workers, 

• Exchanging dosimeters, 

• Recording notional dose (i.e., some identified values for lower dose workers based on a small 

fraction of the regulatory limit), 

• Estimating dose from lost or damaged dosimeters, 

• Replacing destroyed or missing records, 

• Evaluating and recording doses for incidents or accidents, and

• Obtaining and recording occupational dose to workers for other employer exposures.

Policies appear to have been in place at Y-12 for all of these parameters (BWXT Y-12 2000, Section 

2.7, pp. 22–26; Patterson, West and McLendon 1957, pp. 7–8, 21–26; McLendon 1963, pp. 11–13, 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 18 of 56

23–25, 95–101; McRee, West, and McLendon 1965, pp. 10–11, 19–21, 87–93; McLendon et al. 1980;

West 1993; Oxley 2001; Souleyrette 2003a, pp. 1–6). Their routine practices appear to have required 

assigning dosimeters to personnel who might receive an external radiation dose that was greater than 

10% of the radiation protection guidelines (RPGs) in effect at that time. According to West (1993), all 

workers were monitored from 1961 to 1979, after which only those workers having an exposure 

potential were monitored quarterly and others annually. Souleyrette (2003a, p. 2) states that nearly all 

workers were monitored from 1989 to 1996 and all workers entering radiological areas were 

monitored after 1996. 

Dosimeters were exchanged on a routine schedule. All beta/photon dosimeters were processed and 

the measured results were recorded and used for dose estimation to the individual workers. Unless 

the worker was actually or could have been exposed to neutrons, a TLND or NTA film was not issued 

to the worker (West 1955) or the NTA emulsion in the worker’s film badge dosimeter was not 

processed (West 1993, p. 5). There appears to be no use of recorded notional doses, although there 

are issues of recording dose for low-dose exposures (see Table 6-2). There is also the problem of 

missing dose components from the total WB dose for a worker designated simply as “not available” or 

“damaged film” in the worker’s records (West 1993). These missing dose components for workers 

could be estimated using a method described by Watson et al. (1994) and Watkins et al. (1994, pp. 

22–26) based on examination of continuity in the worker’s job and work activities.

6.3.2

The Y-12 dosimetry methods evolved during the years as improved technology was developed and 

the complex radiation fields in the workplace were better understood. The adequacy of the respective 

dosimetry methods to accurately measure radiation dose as discussed in later sections depends on 

radiation type, energy, exposure geometry, etc. The dosimeter exchange frequency of the dosimeters 

was gradually lengthened and corresponded generally to downward reductions in the RPGs (Morgan 

1961; Taylor 1971). During the early stages of the Y-12 program to monitor individual workers, a 

weekly dose control of 0.3 rem was in effect. This was changed to an annual limit of 5 rem in the 

latter part of the 1950s. A summary of the major historical events in the Y-12 dosimetry program for 

external radiation is provided in Table 6-3. 

Y-12 Dosimetry Technology

6.3.2.1 Beta-Photon Dosimeters

In 1948, Y-12 started an external dosimetry badge service with the assistance of ORNL (Murray 

1948a,b). ORNL had earlier implemented the beta/photon film dosimeter design that was developed 

originally at the Metallurgical Laboratory of the University of Chicago (Pardue, Goldstein, and Wollan 

1944). Several minor modifications had been made to this original design as discussed by Patterson, 

West, and McLendon (1957) and by Thornton, Davis, and Gupton (1961). The film badge was a socalled two-element badge because a portion of the film was covered with a comparatively “open 

window” and a portion of the film was covered by a cadmium shield or filter. In 1961, the two-element 

film badge was replaced by a multielement film badge with an open window over a portion of the film 

and three filters of plastic, aluminum, and cadmium over other portions of the film (Thornton, Davis, 

and Gupton 1961; McLendon 1963; McRee, West, and McLendon 1965). Cadmium filters have been 

consistently incorporated in all film badge designs at Y-12 since 1948 and have an approximate 

thickness of 1 mm or mass density of 1000 mg cm-2. The external doses to Y-12 workers from 

photons were always determined from film readings behind the cadmium filter (Sherrill and Tucker 

1973). In addition, the Y-12 film badges have always included a comparatively open window to 

measure significant beta radiation and to distinguish film exposures due to beta and photon radiation. 

The film areas behind the plastic and aluminum filters of the multielement film dosimeters were read 

at Y-12, but they were not used in the normal evaluation of worker doses (Sherrill and Tucker 1973). 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 19 of 56

Table 6-3. Y-12 historical dosimetry events. 

Date Event Reference

1948 An external dosimetry program was started to monitor personnel in 

radiological areas using PICs and two-element film badges. Film pads 

were also used as an extremity monitor for beta particle exposure to the 

hands of uranium “metal” workers.

Murray 1948a,b; 

Struxness 1949a 

1949 NTA film used to monitor personnel exposed to neutron sources. Struxness 1949b

1951 Converted from film pads to film rings as an extremity monitor for beta

particle exposure to the hands of uranium “metal” workers.

Struxness 1951

1955 Strips of indium foil with a mass of approximately 1 g each were included 

in the security badges of all employees at Y-12. The foils provided a 

quick means of identifying employees who might have been exposed to 

elevated radiation levels during a nuclear criticality accident.

McLendon 1959

June 16, 

1958

A nuclear criticality accident occurred during recovery of enriched 

uranium in Building 9212 of the Y-12 Plant. The indium foils in the 

security badges of 31 employees indicated that they were exposed to 

elevated levels of thermal neutron radiation during the accident. 

Callihan and Thomas 

1959; McLendon 1959; 

Hurst, Ritchie and 

Emerson 1959

July 1, 

1961

All UCC-ND employees were issued a badge meter that served as both a 

security pass and a routine and accident dosimeter. The badge meter 

contained a four-element film dosimeter for routine monitoring of beta 

particle, photon, and neutron exposures. The badge meter also 

contained indium, other neutron activation foils, and special photon 

dosimeters for nuclear accident dosimetry. 

Thornton, Davis and 

Gupton 1961; 

McLendon 1963; 

McRee, West and 

McLendon 1965

October 1, 

1980

All UCC-ND employees were issued a badge meter with (1) a security 

badge for identification and (2) a two-element TLD badge for personnel 

radiation monitoring of beta particles and photons. A 1-g foil of indium 

was included in the security badge to provide a quick means of identifying 

personnel exposed to elevated levels of radiation during a nuclear 

accident. An additional ORNL neutron badge containing NTA film and 

TLNDs was issued to Y-12 personnel who were exposed to neutrons.

McLendon et al. 1980; 

Howell and Batte 

1982; Gupton 1978; 

Berger and Lane 1985

1980 The accident dosimetry components from the 1961 UCC-ND badge 

meter were consolidated and incorporated into Y-12 security badge 

covers. All persons were required to use these security badge covers 

when entering controlled areas that had an installed criticality alarm.

McLendon 1980; Y-12 

Plant 1982a

January 3, 

1989

All Martin Marietta Energy Systems (MMES) employees were issued a 

badge with (1) a security badge for identification and (2) a four-element 

TLD badge for personnel radiation monitoring of beta particles and 

photons. A 1-g foil of indium was included in the security badge for 

accident dosimetry purposes, and an additional four-element TLND 

badge was issued to Y-12 personnel potentially exposed to neutrons. 

Y-12 Plant 1988a; 

Oxley 2001

1991 The beta/gamma TLD that was issued to all MMES employees became 

the official personnel nuclear accident dosimetry (PNAD) in the event of a 

criticality accident. All persons were required to use a beta/gamma 

dosimeter attached to their security badge before entering controlled 

areas with an installed criticality alarm.

Y-12 Plant 1991; Kerr 

and Mei 1993

The film badges Y-12 used from 1948 to 1963 contained DuPont type 552 film packets (Souleyrette 

2003a). These packets consisted of (1) a sensitive 502 emulsion with an effective range from 

approximately 30 mrem to 10 rem and (2) an insensitive 510 emulsion with an effective range from 

approximately 500 mrem to 20 rem (Craft, Ledbetter, and Hart 1952; Thornton, Davis and Gupton 

1961; Parrish 1979). In 1963, Y-12 switched to the use of DuPont 554 film packets (McLendon 1963; 

Souleyrette 2003a). These packets consisted of (1) a sensitive 555 emulsion with an effective range 

from approximately 30 mrem to 5 rem (McLendon 1963) and (2) an insensitive 834 emulsion with an 

effective range from approximately 5 rem to 150 rem (Thornton, Davis, and Gupton 1961; Davis and 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 20 of 56

Gupton 1963; Parrish 1979). In 1971, DuPont stopped manufacturing the 554 film packets and Y-12 

switched to Eastman type 2 film (Jones 1971; Souleyrette 2003a). The Eastman type 2 packet 

contained one film with two emulsions bonded to opposite sides of the base film. The sensitive 

emulsion had an MDL of approximately 30 mrem (Jones 1971). During the switch to the Eastman 

type 2 film, some film to be evaluated was removed from cold storage, inserted in 16 pairs of badges, 

and the badges placed in racks with the office area to investigate onsite radiation background. A 

sample of film was developed on the day of the background study and used as a base point for other 

measurements. Every 2 weeks, film from a pair of badges was unloaded, developed, and read with a 

densitometer. On the first day of the study, the fresh film had an optical density (OD) of 0.205, and 

after 215 days, the OD had reached a level of 0.405. This increase in the OD of 0.2 represented an 

increase of only about 15 mrem or approximately 0.5 mrem/wk. It was noted that these OD readings 

did not have the OD readings of an unexposed blank subtracted from them and, therefore, represent 

the actual OD readings of the films (Jones 1971).

The response of the film badge to photon radiation of different energies is illustrated in Figure 6-2. 

This figure also shows the Hp(10) response. Two responses are shown: one response is for a 

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

and one response is for a sensitive DuPont 555 emulsion in the a multielement film badge (Thornton, 

Davis, and Gupton 1961). The response of the sensitive Eastman type 2 film in a multielement film 

badge should be quite similar to that of the sensitive DuPont 555 emulsion. The film badges show an 

under-response at the 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 much 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). The two-element TLD badges 

Y-12 used from 1980 to 1989 had LiF chips covered by an open window and an aluminum filter for 

beta/photon discrimination (McLendon et al. 1980); the multielement TLD badges Y-12 adopted for 

use in 1989 had four LiF chips covered by an open window, a plastic filter, a copper filter, and a 

hemispherical Teflon button (Y-12 Plant 1988b; Oxley 2001). The photon doses were determined 

primarily from the readings of the LiF chips that were covered by the aluminum filter in the twoelement TLD badge and the hemispherical Teflon button in the multielement TLD Badge. A value for 

average background radiation of 0.75 mrem/wk from photons was determined for the Oak Ridge area 

by storing a total of 1,680 TLDs in 70 houses for up to 1 year (Sonder and Ahmed 1991). The 

distribution of results indicated a rather large variation in background among houses, with a few 

locations exhibiting a background of double the average. It was suggested that the results from the 

high-background houses be ignored in determining values of the MDL to be used in routine personnel 

dosimetry monitoring.

IARC has conducted a recent dosimeter intercomparison study of ten commonly used historical 

dosimetry systems from around the world (Thierry-Chef et al. 2002). The results of this IARC study, 

for the U.S. dosimeters only, are presented in Table 6-4. 

Three of the designs were from the United States. These included a two-element film dosimeter that 

was previously used at Hanford (identified as US-2), a multielement film dosimeter that was previously 

used at Hanford (identified as US-8), and the Panasonic TLD that is currently used at the Savannah 

River Site (identified as US-22). The IARC study considered that exposure to workers could be 

characterized as a combination of anterior-posterior (A-P), rotational, and isotropic irradiation 

geometries. Dosimeter responses for these various irradiation geometries were investigated using 

two different phantoms to represent the torso of the body. The first phantom was a water-filled slab 

phantom with polymethyl methacrylate walls, an overall width of 30 cm, an overall height of 30 cm, 

and an overall depth of 15 cm. This phantom is widely used for dosimeter calibration and 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 21 of 56

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

Figure 6-2. Comparison of Hp(10) from a broad beam of 

normally incident photons (ICRP 1996) with the energy 

responses for a sensitive DuPont 502 emulsion in a MED twoelement film badge (Pardue, Goldstein, and Wollan 1944) 

versus a sensitive DuPont 555 emulsion in an ORNL 

multielement film badge (Thornton, Davis, and Gupton 1961). 

Table 6-4. IARC study results for U.S. beta/photon dosimeters (Thierry-Chef et al. 2002).

Geometry Phantom

118 keV 208 keV 662 keV

Meana SD/Mean Meana SD/Mean Meana SD/Mean

US-2 (two-element film dosimeter)

A-P Slab 3.0 2.1 1.3 1.0 1.0 0.8

A-P Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8

Rotational Anthropomorphic 2.2 2.0 1.4 3.0 1.2 3.2

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7

US-8 (multielement film dosimeter)

A-P Slab 1.0 1.5 1.0 0.8 0.8 1.7

A-P Anthropomorphic 0.8 9.5 0.9 6.0 0.8 1.8

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8

Isotropic Anthropomorphic 1.0 3.0 1.2 9.0 1.0 2.3

US-22 (multichip TLD dosimeter)

A-P Slab 0.9 4.4 0.9 3.9 0.9 3.5

A-P Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6

a. Ratio of recorded dose to Hp(10). 

performance testing by the International Standards Organization (ISO). The second phantom was an 

anthropomorphic Alderson Rando phantom. This realistic phantom is constructed using a natural 

human skeleton cast inside material that has a tissue equivalent composition.

As noted previously, the multielement film badge was used at Y-12 in essentially the same manner as 

the two-element film badge. It should also be noted that the two-element film dosimeter can 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 22 of 56

significantly overestimate Hp(10) at the lower photon energies of 118 keV and 208 keV as shown in 

Table 6-4. 

6.3.2.2 Neutron Dosimeters

Several general types of neutron dosimeters that have been used at Y-12 differ significantly in their 

response to neutrons of different energies as illustrated in Figure 6-3 (IAEA 1990). An NTA emulsion 

was included in the same holder used for the Y-12 beta-gamma dosimeter until 1980. Between 1980 

and 1989, Y-12 workers who were exposed to neutrons were provided with a separate neutron 

dosimeter. This neutron dosimeter contained both an NTA film for measurement of the fast neutron 

dose and a TLND for measurement of the neutron dose from lower energy neutrons (Gupton 1978). 

From 1980 to 1985, the neutron doses to Y-12 workers were determined at ORNL using both the NTA 

and TLND dosimeters, as discussed by Gupton (1978) and Berger and Lane (1985). Since 1989, the 

neutron doses to Y-12 workers have been measured using a separate albedo TLND that is worn on 

the belt to keep it in close contact with the worker’s body (Gunter 1994; Oxley 2001). In general, 

TLNDs have a response that increases with decreasing neutron energy, while NTA films have very 

little response to neutrons with energies less than its detection threshold of approximately 500 keV 

(Figure 6-3). Results reported at the first AEC neutron dosimetry workshop in 1969 indicated that 

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

those that were measured with other methods including the TLND (Vallario, Hankins, and Unruh 

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

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

spectra for reliable results.

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1

Neutron Energy (MeV)

10|0

10|1

10|2

10|3

10|4

Relative Response

Hp(10)

NTA Film

6

LiF with Cd Shield

Figure 6-3. Comparison of Hp(10) from normally incident 

neutrons (IAEA 2001) to the energy responses of an NTA film 

and a neutron albedo dosimeter with a neutron TLND chip 

made of 6

LiF shielded by cadmium (IAEA 1990, 2001).

6.3.3

Potential error in recorded dose is dependent on the calibration methodology and the extent of the 

similarity between the radiation fields for calibration and those in the workplace. The potential error is 

Y-12 Dosimeter Calibration Procedures
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 23 of 56

much greater for dosimeters with significant variations in response such as film dosimeters for lowenergy photon radiation and both the NTA and TLND dosimeters for neutron radiation.

Beta/Photon Dosimeters

The Y-12 dosimeters were originally calibrated primarily using beta particles from a natural uranium 

slab and photons from a 226Ra source (Souleyrette 2003a). The dosimeters were exposed face down 

on the uranium slab and free in air (i.e., no phantom) facing the 226Ra source for preselected times to 

reproduce the beta particle and photon doses that were normally encountered in the workplace. This 

practice was similar to that at other AEC sites.

Neutron Dosimeters

A good account of the historical aspects of the calibration of the Y-12 neutron dosimeters is not 

available. It is known, however, that the NTA films were originally calibrated using a Po-Be neutron 

source (Struxness 1949a, 1952) and later an Am-Be neutron source (McLendon 1963; McRee, West, 

and McLendon 1965). The dosimeters containing the NTA films were exposed free in air (i.e., no 

phantom) to neutrons from the Po-Be and Am-Be sources for preselected times to reproduce the 

neutron doses that were normally encountered in the workplace. Some information on the calibration 

of Y-12 TLNDs can be found in Berger and Lane (1985) and Oxley (2001). 

6.3.4

The main workplace radiation fields at Y-12 are due to processes that involve either EU (235U) or 

depleted uranium (DU) (238U). Some other workplace radiation fields involve industrial radiation 

generating equipment (X-rays and electron accelerators) and isotopic gamma-ray and neutron 

sources for testing purposes (60Co and 252Cf). The current Y-12 external dosimetry TBD provides a 

discussion of the radiation fields due to different processes and primary nuclides in the Y-12 

workplace (Oxley 2007, pp. 3–5).

Y-12 Workplace Radiation Fields

6.3.4.1 Workplace Beta/Photon Dosimeter Response

Beta/photon radiation fields characteristic of the Y-12 facilities can be generally defined based on 

historical information as presented in Table 6-5. Because Y-12 is a nuclear weapons fabrication and 

disassembly facility, the most common materials are EU (235U) and DU (238U). Both 235U and 238U are 

primarily alpha particle emitters. However, 235U does emit a 185-keV photon in 54% of its decays. 

Most of the external dose from 238U comes from its short-lived 234Th, 234mPa, and 234Pa decay 

products. From an external dose standpoint, the most significant radiations from these decay 

products of 238U are: (1) the 2.29-MeV beta particle from 234mPa, and (2) the photons emitted by 234Pa 

with energies as large as 1.962 MeV. The various Y-12 dosimeters have filtration of about 1,000 mg 

cm-2 (i.e., nearly equivalent to 1-cm depth in tissue) for those regions of the dosimeter that are used to 

measure the WB dose. The response to beta radiation in Y-12 workplaces is limited because beta 

radiation usually cannot penetrate this much filtration.

The largest workplace exposures at Y-12 have historically occurred in the DU process areas 

(Struxness 1954; Henderson 1991). During casting operations, the decay products of 238U float to the 

top surface of the molten metal and remain as surface residues. These surface residues result in an 

increased exposure potential because of the high beta and photon energies that are associated with 

the 234Pa nuclide. The 234Pa nuclide emits a number of high-energy photons and has a specific 

activity that is approximately 2 x 1015 times larger than the specific activity of its 238U parent 

(Henderson 1991). For 234Pa, the percentages of photons with energies of 30 to 250 keV and 250 

keV or more are about 7% and 93%, respectively. For 238U in equilibrium with its short-lived 234Th, 

234mPa, and 234Pa, the percentages of photons with energies of 30 to 250 keV and 250 keV or more 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 24 of 56

Table 6-5. Selection of beta and photon radiation energies and percentages for Y-12 site processes. 

Y-12 site processes Building

Operations Radiation 

type

Energy 

Begin End selection Percent

Enriched uranium product recovery 

and salvage operations

9203 1947 1951 Beta >15 keV 100

9206a 1947 1959 Photon 30–250 keV 100

9211 1947 1959

9201-1 1952 1963

Uranium chemical operations and 

weapon production operations

9202 1947 1995 Beta >15 keV 100

9206a 1947 1995 Photon 30–250 keV 100

9212b 1949 Ongoing

Special nuclear material receiving 

and storage

9720-5 1949 Ongoing Photon 30–250 keV 100

Uranium forming and machining for 

weapon component operations

9201-5 1949 Ongoing Beta >15 keV 100

9204-4 1949 Ongoing Photon 30–250 keV 100

9215 1950 Ongoing

Depleted uranium process 

operations

9201-5 1949 Ongoing Beta >15 keV 100

9204-4 1949 Ongoing Photon 30–250 keV 50

9766 1949 ? >250 keV 50

9998 1949 Ongoing

Final weapon component assembly 

operations

9204-2 1952 Ongoing Beta >15 keV 100

9204-2E 1952 Ongoing Photon 30–250 keV 100

ORNL 86-in cyclotron 9201-2 1950 1983 Photon 30–250 keV 50

>250 keV 50

Chemical assay and mass 

spectrometry laboratories

9203 1947 Ongoing Photon

Specific to radiation 

source. Photon default 

values:

Radiographic laboratory 9201-1 1947 Ongoing Photon

Calibration laboratory 9983 1949 Ongoing Photon

Weapon component assay 

laboratory 

9995 1952 Ongoing Photon

30–250 keV 50

Nondestructive assay laboratory 9720-5 1980 Ongoing Photon >250 keV 50

West End waste treatment facility 9616-7 1984 Ongoing Beta >15 keV 100

Photon 30–250 keV 50

>250 keV 50

a. Building 9206 Complex includes Buildings 9768, 9720-17, 9409-17, 9510-2, 9767-2, and the east and west tank farm 

pits.

b. Building 9212 Complex includes Buildings 9809, 9812, 9818, 9815, and 9980.

are about 82% and 18%, respectively. Therefore, an artificially high percentage of photons with 

energies greater than 250 keV was assumed in Table 6-5 for the normal and DU process areas. This 

produces doses that are favorable to the claimant because of the increased exposure potential to 

high-energy photons from the short-lived 234Pa decay product of 238U.

One serious problem with the workplace response of the Y-12 beta/gamma dosimeters involves 

workers who performed waist-level uranium handling jobs in the DU process areas (Henderson 1991). 

A personnel dosimeter worn at the collar might underestimate the Hp(10) dose at the waist by rather 

significant factors. Y-12 now instructs these workers to wear the dosimeters at the waist, but many 

workers might have worn them on the collar in the past. Therefore, for all workers who performed 

waist-level uranium handling jobs, the recorded dose before 1991 should be multiplied by 1.34 

(Henderson 1991). To determine when to make such adjustments, the dose reconstructor must 

depend on information about routine duties and work locations in the computer-assisted telephone 

interview (CATI) file for a claimant.
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 25 of 56

6.3.4.2 Workplace Neutron Dosimeter Response

Three main facilities at Y-12 with a potential for neutron exposure are: (1) the Calibration Laboratory 

in Building 9983, (2) the EU Storage Area in Building 9212, and (3) the Nondestructive Assay 

Laboratory in Building 9720-5. The following sections discuss the neutron exposure spectra and 

neutron-to-photon dose ratios in these three areas using data from recent measurements from the 

Pacific Northwest Laboratory (PNL) (Soldat et al. 1990; McMahan 1991; Oxley Y-12 2001). 

6.3.4.2.1 Calibration Laboratory in Building 9983

The Calibration Laboratory has a highly shielded room used for storage of both photon and neutron 

sources. The walls of the room are all 3-in. steel, with a high-density concrete floor into which several 

source storage pits are sunk. The types of neutron sources stored in this room include twelve 2 to 

4 Ci americium-boron (Am-B) sources, several americium-lithium (Am-Li) sources, and several 

americium-beryllium (Am-Be) sources. At the time of the PNL measurements, the neutron sources 

were stored in the room in shielded containers. The neutron shielding of the containers was either 

paraffin or high-density plastic depending on the container. Several sources were stored in containers 

inside a steel safe, others were in their containers on the floor of the room, and still others were in the 

storage pits below floor level. The neutron sources have not been used for routine calibration 

purposes since the early 1970s, when ORNL began calibrating all Y-12 neutron detection and survey 

instruments. Workers at Y-12 do access the source storage area for other purposes, and the PNL 

measurements were made outside the door to the source storage room to determine appropriate 

calibration factors for a worker’s TLND.

6.3.4.2.1.1 Neutron Energy Spectrum

The PNL measurements of the neutron energy spectrum at a distance of 18 in. from the door to the 

source storage room are shown by the solid line in Figure 6-4. It should be noted that the PNL 

measurement data were provided as dose equivalent rates (Soldat et al. 1990); however, a 1-hour 

exposure was assumed here to show the results of the PNL measurements as dose equivalent. The 

fluence-to-dose equivalent conversion factors and neutron quality factors used in the PNL 

measurements were 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 1973). 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 can 

be found in a report by Sims and Killough (1983). The dashed line in Figure 6-4 shows the dose 

equivalent from the PNL measurements divided into the four energy groups in NIOSH (2002), and the 

dose fractions in each of these four energy groups are provided in Table 6-6. Although PNL 

measured some dose from lower (<10 keV) and intermediate energy (10 to 100 keV) neutrons, the 

contribution to the total dose was only about 6%. The radiation effectiveness factor (REF) used in the 

Interactive RadioEpidemiological Program (IREP) to calculate the POC for these two neutron energy 

groups is less than the fast (or fission) neutron energy group from 0.1 to 2 MeV (Kocher, Apostoaei, 

and Hoffman 2002). As a result, combining the lower and intermediate energy groups into the fast 

neutron group from 0.1 to 2 MeV is a reasonable simplification of the neutron dose calculation that is 

favorable to the claimant.

6.3.4.2.1.2 Neutron-to-Photon Dose Ratio

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 8:1. For 

workers in the Calibration Laboratory, no other data on neutron-to-photon dose ratios have been 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 26 of 56

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-4

10|-3

10|-2

10|-1

10|0

10|1

Dose Equivlanet (mrem)

0.255 mrem 0.0304 mrem

2.38 mrem 2.01 mrem

Figure 6-4. Results of PNL neutron spectrum measurements in 

the Calibration Laboratory of Building 9983 (solid line) and the 

PNL measurement results (dashed line) divided into the four 

neutron energy groups in the dose reconstruction for Y-12 

workers (Soldat et al. 1990).

Table 6-6. Dose fractions for Y-12 calibration laboratory. 

Neutron energy group Near source storage safe

<10 keV 0.055

10–100 keV 0.007

0.1–2 MeV 0.509

2–14 MeV 0.429

Dose fractions favorable to the claimant

0.1–2 MeV 0.57

2–14 MeV 0.43

found to use to estimate the missing neutron dose from measurements made with film dosimeters. 

However, the recent PNL studies indicate that more than 90% of the neutron dose is above the 

500-keV threshold of the NTA films. In addition, neutron dose measurements for Calibration 

Laboratory workers with NTA dosimeters are expected to be reasonably accurate to within the 

parameters that are discussed in NIOSH (2002).

6.3.4.2.2 Enriched Uranium Storage Area in Building 9212

Building 9212 contains a secure storage area for containers of enriched uranium tetrafluoride (UF4) 

and uranium trioxide (UO3). Neutrons are produced by alpha particle reactions with the nucleus of the 

fluorine and oxygen atoms of the UF4 and UO3, respectively. Containers of these materials are 

placed on a rack of shelves and arranged in a matrix that is critically safe. The containers are spaced 

approximately 2 to 2.5 ft apart on a shelf, and can be placed one deep per shelf. There are four 

shelves per rack and 24 in. between shelves. The PNL measurements were made 39 in. above the 

floor and 24 in. from the shelf at a location near the center of a rack that was filled with 20 containers 

of UF4. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 27 of 56

6.3.4.2.2.1 Neutron Energy Spectrum

The PNL measurements of the neutron energy spectrum near the center of the UF4 storage rack are 

shown by the solid line in Figure 6-5, and the dose fractions for the neutron energy groups shown by 

the dashed line in this figure are provided in Table 6-7. The dose fraction for the lower (<10 keV) and 

intermediate (10 to 100 keV) energy neutron groups were less than 2% of the total dose from these 

PNL measurements. As before, combining the lower and intermediate energy groups into the fast 

neutron group from 0.1–2 MeV is a reasonable simplification of the neutron dose calculation that is 

favorable to the claimant.

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-4

10|-3

10|-2

10|-1

10|0

10|1

Dose Equivlanet (mrem)

0.0193 mrem

 1.29 mrem

0.00381 mrem

0.340 mrem

Figure 6-5. Results of PNL neutron spectrum measurements in 

the enriched uranium storage area of Building 9212 (solid line) 

and the PNL measurement results (dashed line) divided into 

the four neutron energy groups used in the dose reconstruction 

for Y-12 workers (Soldat et al. 1990).

Table 6-7. Neutron dose fractions for Y-12 EU storage area. 

Neutron energy group Near storage rack

<10 keV 0.012

10–100 keV 0.002

0.1–2 MeV 0.781

2–14 MeV 0.205

Dose fractions favorable to the claimant

0.1–2 MeV 0.79

2–14 MeV 0.21

6.3.4.2.2.2 Neutron-to-Photon Dose Ratio

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 1:1. For 

workers in the Enriched Uranium Storage Facility, no other data on neutron-to-photon dose ratios in 

Building 9212 have been found to use to estimate the missing dose from earlier measurements made 

with film dosimeters. However, the recent PNL studies indicate more than 95% of the neutron dose is 

above the 500-keV threshold of the NTA films. In addition, the neutron dose measurements for 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 28 of 56

workers in the Enriched Uranium Storage Area of Building 9212 with NTA film dosimeters are 

expected to be reasonably accurate to within the parameters in NIOSH (2002).

6.3.4.2.3 Nondestructive Assay Laboratory in Building 9720-5 

The Nondestructive Assay Laboratory in Building 9720-5 is used for recovery of highly enriched 

uranium (HEU) from manufacturing wastes (Hogue and Smith 1984). The laboratory contains 

instruments for gamma scanning and neutron interrogation of containers of solid wastes, gamma 

analysis of solution samples, and measurements of solution density. Because measurements of the 

neutron spectrum were made previously for a 252Cf fission-neutron source at ORNL’s Radiation 

Calibration Laboratory (RADCAL), it was not necessary to make additional neutron measurements to 

characterize the workplace radiation fields near the 252Cf neutron source at the Nondestructive Assay 

Laboratory at Y-12. The neutron measurements at the RADCAL facility were made at a distance of 

39 in. from the bare 252Cf source and a height of approximately 39 in. above the floor.

6.3.4.2.3.1 Neutron Energy Spectrum

The results of the PNL neutron spectrum measurements made at 39 in. from the 252Cf fission-neutron 

source at ORNL’s RADCAL facility are shown by the solid lines in Figure 6-6. The dose fractions for 

the neutron energy groups (dashed lines) in this figure are provided in Table 6-8. The dose fractions 

for the lower (<10 keV) and intermediate (10–100 keV) energy neutron groups were less than 1% of 

the total dose from these PNL measurements. Therefore, combining the lower and intermediate 

energy groups into the fast neutron group of 0.1 to 2 MeV is a reasonable simplification of the neutron 

dose calculation that is favorable to the claimant.

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1 10|2

Neutron Energy (MeV)

10|-7

10|-6

10|-5

10|-4

10|-3

10|-2

10|-1

Dose Equivlanet (mrem)

9.75e-5 mrem 1.21e-4 mrem

7.59e-3 mrem

2.61e-2 mrem

Figure 6-6. Results of PNL neutron spectrum measurements

made at 39 in. from a bare 252Cf fission neutron source (solid 

line) and the PNL measurement results (dashed lines) divided 

into the four neutron energy groups used in the dose 

reconstruction for Y-12 workers at the Nondestructive Analysis 

Laboratory in Building 9720-5 (Soldat et al. 1990). 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 29 of 56

Table 6-8. Dose fractions for Y-12 Nondestructive 

Analysis Laboratory. 

Neutron energy group Near unshielded Cf-252 source

<10 keV 0.003

10–100 keV 0.004

0.1–2 MeV 0.224

2–14 MeV 0.769

Dose fractions favorable to the claimant

0.1–2 MeV 0.23

2–14 MeV 0.77

6.3.4.2.3.2 Neutron-to-Photon Dose Ratio

The neutron-to-photon dose ratio from the recent PNL measurements was approximately 25:1. For 

workers in the Nondestructive Assay Laboratory in Building 9212, no other data on neutron-to-dose 

ratios have been found to use to estimate missing dose from earlier measurements made with film 

dosimeters. However, the recent PNL studies indicate that more than 97% of the neutron dose is 

above the 500-keV threshold of the NTA films. In addition, the neutron dose measurements for 

workers in the Nondestructive Assay Laboratory with NTA film dosimeters are expected to be 

reasonably accurate to within the parameters in NIOSH (2002).

Typical Workplace Neutron Dosimeter Hp(10) Performance

Typical neutron personnel dosimeter parameters important to Hp(10) performance in the workplace 

are summarized in Table 6-9. The most important parameter related to Hp(10) performance of the 

neutron dosimeters is the difference between calibration and workplace neutron energy spectra. 

Measurements made by PNL in the late 1980s and early 1990s could be used to correct the response 

of the Y-12 TLND dosimeters to workplace neutron energy spectra in several Y-12 areas. These 

measurements were discussed in the previous section of this report and the results of the corrected 

TLND workplace measurements over a 12-year period starting in 1990 are shown in Table 6-10. 

Table 6-9. Typical workplace neutron dosimeter Hp(10) performance.

a

Parameter Description Potential workplace biasb

Workplace neutron 

energy spectra 

NTA dosimeter response decreases and 

TLND response increases with 

decreasing neutron energy. 

Depends on workplace neutron spectra. 

NTA recorded dose of record likely too 

low because of high 500-keV threshold 

for detection of neutrons. 

Exposure geometry NTA dosimeter response increases with 

increasing exposure angle and TLND 

response decreases with increasing 

exposure angle.

NTA recorded dose likely too high since 

dosimeter response is higher at angles 

other than A-P. TLD recorded dose is 

lower at angles other than A-P. Effect is 

highly dependent on neutron energy. 

Missed dose Doses less than the MDL are recorded 

as zero dose.

Recorded dose of record is likely too low. 

The impact of missed dose is greatest in 

earlier years because of the higher MDLs 

of the neutron dosimeters. 

Environmental effects Workplace environment (heat, humidity, 

etc.) fades the dosimeter signal.

Recorded dose of record is likely too low.

a. Judgment based on Y-12 dosimeter response characteristics.

b. Recorded dose compared to Hp(10). 

These data illustrate the low potential for routine exposure to neutrons at Y-12 both now and in the 

past. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 30 of 56

Table 6-10. Number of neutron-monitored workers, cumulative annual neutron dose, and average 

annual neutron dose to Y-12 workers for 12-year period after introduction of the four-element TLND 

dosimeter in 1989 (Souleyrette 2003b). 

Year

Number of neutronmonitored workersCumulative annual 

neutron dose (mrem)

Average annual

neutron dose (mrem)

1990 82 1,085 13.2

1991 64 463 7.2

1992 86 200 2.3

1993 215 343 1.6

1994 301 1,289 4.3

1995 165 116 0.7

1996 203 470 2.3

1997 38 10 0.3

1998 47 57 1.2

1999 141 121 0.9

2000 49 35 0.7

2001 73 55 0.8

The recent PNL measurements also indicate that the past NTA film dosimeters worked reasonably 

well in the Y-12 workplace because the Ra-Be and Po-Be neutron spectra used to calibrate them 

were reasonably well matched to the workplace neutron spectra. These measurements suggest that 

the NTA film dosimeters missed less than 10% of the neutron dose equivalent at the Calibration 

Laboratory of Building 9983 and less than 5% of the neutron dose equivalent at the Enriched Uranium 

Storage Area of Building 9212 and the Nondestructive Assay Laboratory of Building 9720-5. There 

are many recorded zeros in the neutron dose data for Y-12 workers for two reasons: (1) a worker’s 

NTA film was not developed and read, or (2) a worker’s neutron dose equivalent was less than the 

MDL for the NTA film.

6.4 ADJUSTMENTS OF RECORDED DOSE 

Adjustments to the Y-12 recorded doses are necessary to arrive at a dose that is favorable to the 

claimant because of the uncertainty that is primarily due to the complex workplace radiation fields and 

exposure geometries.

6.4.1

The average and maximum deep photon dose to Y-12 workers for the 10-year period from 1978 to 

1987 is shown in Figure 6-7 (Y-12 Plant 1978, 1980, 1981, 1982b, 1983, 1984, 1985, 1986, 1987, 

1988b). The UCC-ND policy at that time was to limit the maximum deep photon dose to workers to 

500 mrem or less per quarter and 2,000 mrem or less per year. The average deep dose from photons 

to all Y-12 workers was approximately 20 mrem from 1978 to 1987. This period covers the change 

from film dosimeters to TLDs in 1980. No abrupt change occurred in the deep penetrating data for 

photon dose in 1980. Therefore, the recorded doses for the photon deep dose from both film and 

TLDs appear to be in very close agreement, and no adjustments are deemed necessary to the 

recorded deep photon doses for most Y-12 workers. 

Photon Dose Adjustments

There is one group of Y-12 workers for which an adjustment in the recorded photon dose is 

recommended (see Section 6.3.4.1). These workers performed waist-level handing jobs in DU 

process areas (see Table 6-5). Examples of waist-level handling jobs are the unloading and sorting of 

DU scrap materials, shearing of larger pieces of scrap materials, cleaning of the scrap materials, 

crucible loading during the melting and casting operations, and materials sampling (Henderson 1991). 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 31 of 56

1978 1979 1980 1981 1982 1983 1984 1985 1986 1987

Year

10|0

10|1

10|2

10|3

10|4

10|5

Annual Dose Equivalent (mrem)

Maximum photon deep dose

Average photon deep dose

Figure 6-7. Maximum and average photon deep dose to Y-12 

workers, 1978 to 1987. 

Y-12 now instructs workers who perform these operations to wear their beta/photon dosimeters at the 

waist, but many workers might have worn their dosimeters at the collar in the past. The photon dose 

correction that is summarized in Table 6-11 is necessary to calculate an adjusted photon dose that is 

favorable to the claimant before 1991. From 1991 to the present, no correction is needed because 

the recorded photon dose is Hp(10) equivalent.

Table 6-11. Adjustments to reported Y-12 deep photon dose. 

Parameter Description

Period Before 1/1/1991

Dosimeters All beta/photon dosimeters

Facilities Depleted uranium process operations

Workers Waist-level metal handling operators

Adjustment to recorded dose Multiply reported deep photon dose by a factor of 1.34 to estimate Hp(10)

6.4.2

The Y-12 facility incorporated the energy variation of the dose equivalent from neutrons into their 

calibration methodology. Therefore, the recorded dose equivalent is a combination of all neutron 

energies. To calculate the POC, the recorded neutron dose must be separated into neutron energy 

groups as discussed in Section 6.3.4.2 and then converted into ICRP Publication 60 (1991) 

methodology. 

Neutron Dose Adjustments

6.4.2.1 Neutron Weighting Factors

An adjustment to the neutron dose is necessary to account for the change in neutron quality factors 

between historical and current scientific guidance as is discussed by NIOSH (2002). At Y-12, the 

TLNDs were calibrated using PNL measurements based on fluence-to-dose conversion factors and 

quality factors similar to those from ICRP Publication 21 (1973) and NCRP Report 38 (NCRP 1971). 

These quality factors are point-wise data because they were calculated for a broad-parallel beam of 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 32 of 56

monoenergetic neutrons incident on a 30-cm-diameter cylinder of tissue that represents the torso. 

The NCRP Report 38 (1971) quality factors are compared in Figure 6-8 with those used in the PNL 

measurements at Y-12. To convert from NCRP Report 38 quality factors to ICRP Publication 60 

radiation weighting factors, a curve was fit that described the quality factors as a function of neutron 

energy. A group average quality factor was then calculated as shown in Figure 6-8 for each of the 

neutron energy groups to define the radiation weighting factors in ICRP Publication 60 (1991). A 

summary of the group-averaged NCRP Report 38 (1971) quality factors for dose reconstruction is 

provided in Table 6-12. This table also compares the group averaged NCRP Report 38 quality factors 

with historical dosimetry guidelines from the First Tripartite Conference at Chalk River in 1949 (Fix, 

Gilbert, and Baumgartner 1994).

10|-8 10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1

Neutron Energy (MeV)

0

5

10

15

Neutron Quality Factor

PNL Measurements

NCRP 38: Point-wise data

NCRP38: Group averaged data

Figure 6-8. Comparison of the neutron quality factors in the PNL 

neutron spectrum measurements and the neutron quality factors 

from NCRP Report 38 (NCRP 1971) shown both as point-wise data 

and grouped averaged data over the four neutron energy groups for 

dose reconstruction for Y-12 workers. 

Table 6-12. Neutron quality factor, Q, or weighting factor, wr. 

Neutron energy 

(MeV)

Historical dosimetry 

guidelinea NCRP Report 38 (1971) group averaged quality factorb (Q)

ICRP Publication 60 (1991) 

neutron weighting factor (wr)

Thermal 3 2.35 5

0.5 eV–10 keV 10

10 keV–100 keV 5.38 10

100 keV–2 MeV 10.49 20

2 MeV–14 MeV 7.56 10

14 MeV–60 MeV Not applicable 5

a. First Tripartite Conference at Chalk River in 1949 (Fix, Gilbert, and Baumgartner 1994). 

b. See Figure 4.6.2-1 in NCRP Report 38 (1971).
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 33 of 56

6.4.2.2 Neutron Dose Correction Factors

The average quality factor for the four energy groups that encompass the Y-12 neutron exposures are 

provided in Table 6-12. The neutron dose equivalent correction factor for each of these four energy 

groups Cf(En) can be calculated by the use of the following equation: 

( ) ( )

( ) C E ( ) D E

Q E f n w E f n

avg n

= × R n

where 

Df(En) is the dose fraction from Section 6.3.4.2 for the specific neutron energy group of interest

Qavg(En) is the group average NCRP 38 (1971) neutron quality factor for that specific energy group

Wr(En) is the ICRP 60 (1991) neutron weighting factor for that specific energy group

The neutron dose distributions by energy for the various neutron exposure areas at Y-12 are 

summarized in Table 6-13. By multiplying the recorded neutron dose by the area-specific correction 

factors, the neutron dose equivalent is calculated as follows. Consider security personnel who 

inventory fissile material in the Enriched Uranium Storage Area of Building 9212. Assume that the 

worker receives a recorded annual neutron dose of 100 mrem. The corrected neutron dose is 

151 mrem for neutrons with energies between 0.1 and 2 MeV, 28 mrem for neutron with energies 

between 2 and 14 MeV, and 179 mrem for neutrons of all energies. These corrections should be 

applied to both measured neutron dose and missed neutron dose. The dose fractions by energy and 

the associated ICRP Publication 60 (1991) correction factors for various neutron exposure areas at 

Y-12 are summarized in Table 6-13. 

Table 6-13. Summary of neutron dose fractions and associated ICRP Publication 60 (1991) 

correction factors for Y-12 facilities. 

Y-12 facilities Building

Operations Neutron 

energy

Neutron 

dose fraction

ICRP 60 

Begin End correction factors

Calibration Laboratory 9983 1949 Ongoing 0.1–2 MeV 0.57 1.09

2–14 MeV 0.43 0.57

Enriched Uranium 

Storage Area

9212 1949 Ongoing 0.1–2 MeV 0.79 1.51

2–14 MeV 0.21 0.28

Nondestructive Analysis 

Laboratory

9720-5 1980 Ongoing 0.1–2 MeV 0.23 0.44

2–14 MeV 0.77 1.02

6.5 MISSED DOSE 

There is undoubtedly missed dose for Y-12 workers. Analysis of the missed dose has been 

separated according to photon and neutron missed dose. The missed photon dose is discussed first 

and then the neutron missed dose.

6.5.1

Missed photon dose to Y-12 workers could have occurred for the following reasons: (1) the worker 

was not monitored before 1961, (2) there is no recorded dose for short periods of time after 1961, and 

(3) the worker’s dose during a monitoring period was recorded as zero because the dosimeter 

response was less than the MDL. Before 1961, the policy at Y-12 was to issue a film badge only to 

those workers who might exceed 10% of the RPGs in effect at that time. This practice resulted in 

Photon Missed Dose
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 34 of 56

large numbers of workers not being monitored for external radiation exposure before 1961 (see 

Figure 6-1). 

If a worker’s routine duties and work location remained essentially the same during the 1950s and 

early 1960s, it might be feasible to use the recorded annual doses in the early 1960s to estimate the 

missed dose before 1961. Methods are being investigated at present based on department numbers, 

job descriptions, and work locations that might be used to estimate annual doses for other workers 

who were not monitored for external radiation exposure before 1961 and did not remain in the same 

jobs during the 1950s and early 1960s. Methods to be considered when there is no recorded dose for 

short periods for normally monitored workers have been discussed by Watson et al. (1994). 

Estimates of the missed dose can be made using dose results for coworkers or using the nearby 

recorded dose for the specific worker of interest before and after a period of missed dose. 

Regardless of how the missing dose is estimated for unmonitored periods, these situations do require 

careful consideration. 

The missed dose for dosimeter results less than the MDL is particularly important for earlier years 

when MDLs were higher and dosimeter exchange was more frequent. NIOSH (2002) describes 

several different options to calculate the missed dose in these situations. One option to estimate a 

maximum potential dose that is favorable to the claimant is to multiply the MDL by the number of zero 

dose results. This will provide an estimate of the maximum missed dose to the worker. The following 

sections consider missed photon dose for dosimeter results less than the MDL according to facility or 

location, dosimeter type, year, and energy range.

Facility or Location

Information has not been found that is adequate to describe the potential missing photon dose by 

facility or location within the Y-12 Plant. This is particularly true during the early years when the 

missed photon dose was most significant due to the frequent exchange of film dosimeters and the 

higher MDLs. 

Dosimeter Type

The missed photon dose by dosimeter type is discussed in Section 6.3.1 and summarized in 

Table 6-2. The MDLs for the respective Y-12 beta/photon dosimeters are based on results of 

laboratory irradiations, and the actual MDLs for the workplace might be somewhat greater than these 

values because of additional uncertainty in actual field use. Nevertheless, the values in Table 6-2 are 

expected to provide reasonable estimates of the missed dose for Y-12 workers.

Year

Analysis of the missed photon dose by year and dosimeter exchange frequency is discussed in 

Section 6.3.1 and summarized in Table 6-2. The missed photon dose for workers who were 

unmonitored for external radiation exposures before 1961 is also discussed in Section 6.5.1.

Energy Range

An estimate of the missed dose by energy range might be possible based on the type of facility and 

predominant radiation sources or radionuclides at the facility. The recorded dose from the dosimeter 

does not typically provide information to estimate discrete energy ranges. It is possible to examine 

the energy response characteristics of the multielement film and TLD dosimeters, but this analysis 

does not recognize the substantial uncertainties present in the workplace associated with differing 

exposure geometries and mixed radiation fields. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 35 of 56

6.5.2

There can be significant missed neutron dose at Y-12 because of the very low potential for neutron 

exposure as illustrated by the data in Table 6-10. The neutron missed dose is divided into three 

periods in the following discussion. The first period is before 1980 when only NTA film dosimeters 

were used. The second period is from 1980 to 1989 when the switch from NTA film dosimeters to 

TLNDs was being completed. The third period is after 1989 when only TLNDs were used. The 

estimated MDLs for these neutron dosimeters are summarized in Table 6-2. It is possible to estimate 

the missed neutron dose using the MDLs because the neutron dosimeters were calibrated with 

neutron sources that had energies similar to those in the workplace and more than 90% of the 

neutrons to which workers were normally exposed had energies greater than the 500-keV threshold of 

the NTA film dosimeters. There was, of course, no threshold energy for the TLNDs as illustrated in 

Figure 6-3. 

Neutron Missed Dose

Before 1980

The use of NTA films for neutron dosimetry before 1980 is well documented in various Y-12 reports. 

As noted above, it is possible to estimate the missed dose using the MDLs. It was also noted 

previously that there are many recorded zeros in the neutron dose data for Y-12 workers for two 

reasons: (1) a worker’s NTA film was not developed and read, or (2) a worker’s NTA film indicated a 

neutron dose equivalent that was less than the film’s MDL (approximately 50 mrem). If the MDL for 

NTA film is used in estimating the missed neutron dose, it should be multiplied by 1.10 for workers in 

the Calibration Laboratory and by 1.05 for workers in the Enriched Uranium Storage Area of Building 

9212 and the Nondestructive Analysis Laboratory. It is also possible to estimate the missed neutron 

dose in some facilities by use of neutron-to-photon dose ratios (NIOSH 2002). However, the only 

Y-12 facility where this dose ratio is expected to provide a reasonably reliable estimate of the missed 

neutron dose is the Enriched Uranium Storage Area of Building 9212. For this area, the dose ratio 

was determined by recent PNL measurements to be approximately 1:1. The dose ratios for other 

neutron exposure areas at Y-12 that were determined from recent PNL measurements were quite 

large, and the use of these data could lead to gross overestimates of missed neutron dose if a worker 

was also exposed to pure photon sources at the Calibration Laboratory or the Nondestructive Analysis 

Laboratory.

From 1980 to 1989

Y-12 became increasingly dependent over the years on ORNL to process the NTA films and to 

determine neutron doses because of the small numbers of workers who were exposed to neutrons. 

All workers at both ORNL and Y-12 were provided with a two-element TLD dosimeter for beta-particle 

and photon dosimetry (McLendon et al. 1980), and workers who were exposed to neutrons were 

provided with a separate neutron dosimeter (Gupton 1978). This dosimeter was a modification of the 

film badge dosimeter that was previously used at both ORNL and Y-12. The MDL of this neutron 

dosimetry is assumed to be about the same as that of the NTA film alone because most of the 

neutron dose at Y-12 comes from neutrons above the 500-keV threshold of the NTA film. 

After 1989 

Since 1989, the neutron dose has been measured using a newly developed albedo-type TLND that is 

worn on the belt to keep it in close contact with the body. The characteristics of this dosimeter are 

well documented (Oxley 2001), and the MDL to be used in estimating missed dose is 10 mrem (see 

Table 6-2). 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 36 of 56

6.6 UNCERTAINTY IN PHOTON AND NEUTRON DOSE

For film badges, the MDLs that are quoted in the literature range from about 30 to 50 mrem for 

beta/photon irradiation (Morgan 1961; Parrish 1979; West 1993; Souleyrette 2003a) and from 50 to 

100 mrem for neutrons (Morgan 1961; Parrish 1979; Wilson et al. 1990). 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 within ±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. Therefore, 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. Therefore, the standard error for a neutron dose reading of approximately 100 mrem is 

estimated here to be ±50%. For the TLDs that were used at Y-12 after 1980 and the TLNDs after 

1985, 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 and 

TLND dose measurements less than 100 mrem and for TLD and TLND dose measurements in mixed 

radiation fields would be expected to be somewhat larger. 

6.7 ORGAN DOSE

Once the photon and neutron doses and their associated standard errors have been calculated for 

each year, the values are then used to calculate organ doses of interest using the NIOSH External 

Dose Reconstruction Implementation Guideline (NIOSH 2002). There are many complexities and 

uncertainties when applying organ DCFs to the adjusted doses of record. Many of the factors that 

affect the recorded dose have already been discussed in this section. Some factors, such as 

backscattering (phantom calibration) and the over-response to low-energy photons, would indicate the 

recorded dose was too high. Other factors, such as calibration methodology, angular response, low 

energy threshold, and film fading, would result in a recorded dose that was too low. As a result, 

differences in film badge design (filtration) and calibration can have both positive and negative effects 

on the overall dose comparison to Hp(10). ICRU (1988) indicates that film badge dosimeters, while 

not tissue equivalent, can be used for personnel dosimetry. The report emphasizes, however, that it 

is difficult to establish the variations in film response as functions of photon energy and angle of 

photon incidence on the film badge at very low photon energies. Given these broad uncertainties, 

especially with film badge dosimetry in the 1950 to 1980s, an approach is used to estimate organ 

dose that is favorable to the claimant. Because use of exposure-to-organ DCFs results in a higher 

organ dose and higher POC, and given the REFs of the intermediate-energy photons, these DCFs 

should be used to convert recorded film badge doses to organ dose. In the conversion of all recorded 

photon and neutron doses to organ doses, the exposure geometry must also be given careful 

consideration. Some of the more common exposure geometries in the workplace are: (1) an A-P 

exposure is typical for an individual who works in a directional radiation field and faces the radiation 

source, (2) rotational exposure is typical of an individual who is constantly turning in a directional 

radiation field, and (3) an isotropic exposure is typical of an individual who is working in a highly 

nondirectional or omnidirectional radiation field. The proposed default options based on exposure 

geometries for long-term workers that are favorable to the claimant are listed in Table 6-14. 

Attachment A discusses the use of the parameters in Section 6 to aid the dose reconstructor in 

preparing dose reconstructions for long-term Y-12 workers. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 37 of 56

Table 6-14. Default exposure geometries favorable to the claimant for

calculating organ doses. 

Claim status Job category Exposure geometry Percentage

Likely non-compensable All A-P 100

Compensable worker All A-P 50

Rotational 50

Compensable supervisor All A-P 50

Isotropic 50

6.8 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document. 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 38 of 56

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Hygiene Information Bulletin, Number 82-2, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref 

ID: 711, pp. 3-4]

Y-12 Plant, 1983, “Ionizing Radiation Exposures – CY 1982,” Radiation Safety and Industrial Hygiene 

Information Bulletin, Number 83-1, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref ID: 715, 

pp.1-2]

Y-12 Plant, 1984, “Ionizing Radiation Exposures – CY 1983,” Radiation Safety and Industrial Hygiene 

Information Bulletin, Number 84-1, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref ID: 717]

Y-12 Plant, 1985, “Ionizing Radiation Exposures – CY 1984,” Radiation Safety and Industrial Hygiene 

Information Bulletin, Number 85-2, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref ID: 718, 

pp. 3-4]

Y-12 Plant, 1986, “Ionizing Radiation Exposures – CY 1985,” Radiation Safety and Industrial Hygiene 

Information Bulletin, Number 86-2, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref ID: 720]

Y-12 Plant, 1987, “Ionizing Radiation Exposures – CY 1986,” Radiation Safety and Industrial Hygiene 

Information Bulletin, Number 87-2, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref ID: 8547, 

pp. 25-26] 

Y-12 Plant, 1988a, “Ionizing Radiation Exposures – CY 1987,” Radiation Safety and Industrial 

Hygiene Information Bulletin, Number 88-13, Y-12 Plant, Oak Ridge, Tennessee. [SRDB Ref 

ID: 724, pp. 17-18]

Y-12 Plant, 1988b, “New Radiation Dosimeter to be Issued,” Health, Safety, Environment, and 

Accountability Information Bulletin, Health Physics Number 88-14, Y-12 Plant, Oak Ridge, 

Tennessee. [SRDB Ref ID: 724, pp. 19-20]
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 45 of 56

Y-12 Plant, 1991, “Nuclear Accident Dosimetry,” Health, Safety, Environment, and Accountability 

Information Bulletin, Health Physics Number 91-5, Y-12 Plant, Oak Ridge, Tennessee. [SRDB 

Ref ID: 8547, p. 37] 
Document No. ORAUT-TKBS-0014-6 Revision No. 01 Effective Date: 06/02/2009 Page 46 of 56

TABLE OF CONTENTS

SECTION TITLE

A.1 Occupational External Dose...................................................................................................

PAGE

46

A.2 Years of Exposure..................................................................................................................47

A.3 Exposure Rate .......................................................................................................................47

A.4 Radiation Type.......................................................................................................................47

A.4.1 Beta and Photon Radiation.........................................................................................47

A.4.2 Neutron Radiation.......................................................................................................48

A.5 Adjustments to Recorded Dose.............................................................................................. 48

A.5.1 Parameter #1..............................................................................................................48

A.5.2 Adjustment to Recorded Photon Dose........................................................................49

A.5.3 Adjustments to Recorded Neutron Dose.....................................................................49

A.5.4 Unmonitored Photon and Neutron Dose .....................................................................49

A.5.5 Missing Photon Dose..................................................................................................50

A.5.5.1 Facility or Location ....................................................................................... 50

A.5.5.2 Dosimeter Type............................................................................................ 50

A.5.5.3 Year ............................................................................................................. 51

A.5.5.4 Energy Range.............................................................................................. 51

A.5.6 Neutron Missed Dose ................................................................................................. 51

A.5.7 Organ Dose Equivalent...............................................................................................52

A.5.8 Parameter #2..............................................................................................................52

A.5.9 Organ Dose Conversion Factors ................................................................................53

A.1 OCCUPATIONAL EXTERNAL DOSE

The information necessary to evaluate claims is directed to the technical parameters of the annual 

estimates of the primary organ dose that is calculated from the dosimeter and interpreted as personal 

dose equivalents Hp(10) for organ doses and Hp(0.07) in the case of skin, testicular, and breast 

cancer. These are used as a consistent basis of comparison for all years of Y-12 occupational 

external dose starting in 1950.

The primary IREP screen used to input dose parameters is illustrated in Table A-1. The input to these 

fields is obtained from the Y-12 dose of record. The claim provides the primary organ of interest and 

other worker information necessary to run the IREP computer program. Guidance to the dose 

reconstruction analysis in selecting the external dose parameters to complete the respective fields in 

Table A-1 is provided in the following sections.

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Table A-1. IREP dose parameter input screen. 

Exposure

Radiation type

Distribution parameters

# Year Rate Type 1 2 3

1 1960 Acute Photon, 30-250 keV Normal 2 2 0

2 1961

3 1962

A.2 YEARS OF EXPOSURE 

The years of exposure should be identified from the Y-12 radiation dose reports. Missed dose is 

calculated for each of the following reasons: (1) the worker was not monitored for external radiation 

exposure before 1961, (2) there is no recorded dose for short periods after 1961, and (3) the recorded 

dose during a monitoring period was zero because the dosimeter response was less than the MDL. 

The missed dose should be calculated for a claimant for each year of record as an employee unless 

there are valid reasons for years in which there are no records. 

A.3 EXPOSURE RATE

Acute is selected for all types of external beta and photon dose, and chronic is selected for neutron 

dose (NIOSH 2002). 

A.4 RADIATION TYPE 

A.4.1 Beta and Photon Radiation

Assumptions favorable to the claimant should be made using guidance in Table A-2 for beta particles 

and for photons (X-rays and gamma rays) to ensure that dose is not underestimated. The values in 

the table are intended to provide a reasonable estimate of parameters used to calculate the organ 

dose without significant numerical error for long-term Y-12 workers in the respective facilities. There 

is no direct guidance for selecting the specific values other than considerations of the radiation 

sources and usual work tasks. In those cases where there is some doubt in the values, a range of 

realistic values and the option that is favorable to the claimant should be selected.

Table A-2. Selection of beta and photon radiation energies and percentages for Y-12 site processes. 

Y-12 Site Processes Building

Operations Radiation 

type

Energy

Begin End selection (keV) Percent

Enriched uranium product 

recovery and salvage operations

9203 1947 1951

9206a 1947 1959 Beta >15 100

9211 1947 1959 Photon 30-250 100

9201-1 1952 1963

Uranium chemical operations and 

weapon production operations

9202 1947 1995

9206a 1947 1995 Beta >15 100

9212b 1949 Ongoing Photon 30-250 100

Special nuclear material receiving 

and storage

9720-5 1949 Ongoing Photon 30-250 100

Uranium forming and machining 

for weapon component 

operations

9201-5 1949 Ongoing Beta >15 100

9204-4 1949 Ongoing Photon 30-250 100

9215 1950 Ongoing

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Y-12 Site Processes Building

Operations Radiation 

type

Energy

Begin End selection (keV) Percent

Depleted uranium process 

operations

9201-5 1949 Ongoing Beta

9204-4 1949 Ongoing Photon >15 100

9766 1949 ? 30-250 50

9998 1949 Ongoing >250 50

Final weapon component 

assembly operations

9204-2 1952 Ongoing Beta >15 100

9204-2E 1952 Ongoing Photon 30-250 100

ORNL 86-in cyclotron 9201-2 1950 1986 Photon 30-250 50

>250 50

Chemical Assay and Mass 

Spectrometry Laboratories

9203 1947 Ongoing Photon Specific to 

radiation source 

Photon default 

values: 

Radiographic Laboratory 9201-1 1947 Ongoing Photon

Calibration Laboratory 9983 1949 Ongoing Photon

Weapon Component Assay 

Laboratory 

9995 1952 Ongoing Photon 20-250 50

50

Nondestructive Assay Laboratory 9720-5 1980 Ongoing Photon >250

West End Waste Treatment 

Facility

9616-7 1984 Ongoing Beta >15 100

Photon 30-250 50

>250 50

a. Building 9206 Complex includes Buildings 9768, 9720-17, 9409-17, 9510-2, 9767-2, and the east and west tank farm 

pits.

b. Building 9212 Complex includes Buildings 9809, 9812, 9818, 9815, and 9980. 

A.4.2 Neutron Radiation

The default neutron dose distributions by energy for each of the neutron exposure areas at Y-12 are 

summarized in Table A-3.

Table A-3. Selection of neutron energies and percentages for Y-12 site facilities. 

Y-12 site facility Building

Operations Neutron 

energy (MeV)

Default dose 

Begin End fraction (%)

Calibration Laboratory 9983 1949 Ongoing 0.1-2 57

2-14 43

Enriched Uranium Storage Area 9212 1949 Ongoing 0.1-2 79

2-14 21

Nondestructive Assay Laboratory 9720-5 1980 Ongoing 0.1-2 23

2-14 77

A.5 ADJUSTMENTS TO RECORDED DOSE 

A.5.1 Parameter #1

Selection of the distribution parameters in Table A-1 involves the adjustments to the dose of record for 

missed dose before entry into IREP. The selection of a normal distribution for the “Type” determines 

the definition of Parameters #1 and #2. For a normal distribution, Parameter #3 is not used and 

Parameter #1 is the mean of the distribution of recorded dose for each year of monitoring. 

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A.5.2 Adjustment to Recorded Photon Dose 

Adjustments to the Y-12 reported photon dose are necessary to arrive at a dose that is favorable to 

the claimant considering uncertainties that are associated primarily with the complex workplace 

radiation fields and exposure geometries. Henderson (1991) identified such problems for workers 

who perform waist-level handling jobs in the DU process areas of Y-12. Examples of waist-level 

handling jobs are unloading and sorting of DU scrap materials, shearing of larger pieces of scrap 

materials, cleaning of the scrap materials, crucible loading during the melting and casting operations, 

and materials sampling. Y-12 now instructs workers who perform these operations to wear 

beta/photon dosimeters at the waist, but many of these workers might have worn their dosimeters on 

the collar in the past. The photon dose correction summarized in Table A-4 is necessary to calculate 

an adjusted photon dose that is favorable to the claimant for years before 1991. From 1991 to the 

present, no correction is needed because the recorded dose is Hp(10) equivalent. To determine when 

to make such adjustments, the dose reconstructor must depend on information about routine duties 

and work locations in the CATI file for a claimant.

Table A-4. Adjustments to reported Y-12 deep photon dose. 

Parameter Description

Period Prior to 1/1/1991

Dosimeters All beta/photon dosimeters

Facilities Depleted uranium process operations

Workers Waist-level metal handling operators

Adjustment to recorded dose Multiply reported deep photon dose 

by a factor of 1.34 to estimate Hp(10)

A.5.3 Adjustments to Recorded Neutron Dose

The Y-12 facility incorporated the energy variation of the dose equivalent from neutrons into their 

calibration methodology. Therefore, the recorded dose equivalent is a combination of all neutron 

energies. To calculate the neutron dose input to IREP (see Table A-1), the recorded neutron dose 

must be separated into neutron energy groups as shown in Table A-3 and subsequently converted 

into ICRP Publication 60 methodology (ICRP 1991). The dose fractions by neutron energy group and 

the associated Publication 60 correction factors for the various neutron exposure areas at Y-12 are 

summarized in Table A-5. As an example, consider security personnel who inventory fissile material 

in the Enriched Uranium Storage Area of Building 9212 and assume that the worker receives a 

neutron dose of 100 mrem. The corrected neutron dose is 151 mrem for neutrons with energies 

between 0.1 and 2 MeV and 28 mrem for neutrons with energies between 2 and 14 MeV. Therefore, 

the total corrected neutron dose is 179 mrem. These corrections should be applied to both recorded 

and missed doses. 

A.5.4 Unmonitored Photon and Neutron Dose

Missed photon and neutron dose would occur where there is no recorded dose because the worker 

was not monitored or the dose is unavailable for a short period because a film was either lost or 

damaged while being processed. Before 1961, the policy at Y-12 was to issue film badges only to 

those workers who might exceed 10% of the RPGs in effect at that time. This practice resulted in 

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Table A-5. Neutron dose fractions and associated ICRP Publication 60 (1991) correction factors 

for Y-12 site facilities. 

Y-12 facilities Building

Operations Neutron 

energy

Neutron 

dose fraction

ICRP 60 

Begin End correction factors

Calibration 

Laboratory

9983 1949 Ongoing 0.1-2 MeV 0.57 1.09

2-14 MeV 0.43 0.57

Enriched Uranium 

Storage Area

9212 1949 Ongoing 0.1-2MeV 0.79 1.51

2-14 MeV 0.21 0.28

Nondestructive 

Analysis Laboratory

9720-5 1980 Ongoing 0.1-2 MeV 0.23 0.44

2-14 MeV 0.77 1.02

large numbers of workers not being monitored for external radiation exposure before 1961 (see 

Figure 6-1). 

If a worker’s routine duties and work location remained essentially the same during the 1950s and 

early 1960s, it might be feasible to use the recorded annual doses in the early 1960s to estimate the 

missed dose before 1961. Methods are being investigated at present using department numbers, job 

descriptions, and work locations that might be used to estimate annual doses for other workers who 

were not monitored for external radiation exposure before 1961 and did not remain in the same jobs 

during the 1950s and early 1960s.

Methods to be considered when there is no recorded dose for short periods for normally monitored 

workers have been discussed by Watson et al. (1994). Estimates of the missed dose can be made 

using dose results for coworkers or using nearby recorded dose for the specific worker of interest 

before and after a period of missed dose due to a film that was either lost or damaged while being 

processed. Regardless of how the missing dose is estimated for unmonitored periods of time, these 

situations require careful consideration. 

A.5.5 Missing Photon Dose

Missing photon dose also occurs when the recorded dose is zero because the dosimeter response 

was less than the MDL. This kind of missed dose is most important for earlier years when MDLs were 

higher and dosimeter exchange was more frequent. NIOSH (2002) guidance should be followed to 

calculate the missing photon dose by using a missed dose that is favorable to the claimant. This is 

calculated by multiplying the MDL by the number of zero dose results to estimate the maximum 

potential missed dose. The following sections discuss the missed photon dose corrections according 

to facility or location, dosimeter type, year, and energy range. 

A.5.5.1 Facility or Location

Information has not been found that is adequate to describe the potential missing photon dose by 

facility or location within Y-12. This is particularly true during the early years when the missed photon 

dose was most significant due to the frequent exchange of film dosimeters and their higher MDLs.

A.5.5.2 Dosimeter Type

The missed photon dose by dosimeter type is summarized in Table A-6. The MDLs for the respective 

Y-12 beta/photon dosimeters are based on results of laboratory irradiations. The actual MDLs for the 

workplace could be somewhat greater than these values because of additional uncertainty in actual 

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field use. Nevertheless, the values in Table A-6 are expected to provide reasonable estimates of the 

missed dose for Y-12 workers.

Table A-6. Dosimeter type, period of use, exchange frequency, laboratory MDL, and maximum 

annual missed dose. 

Dosimeter Period

Exchange 

frequency

Laboratory 

MDL (mrem)

Maximum annual 

missed dose (mrem)

Beta/photon dosimeters

Pocket ionization chamber 1948–1950 Daily <5 1,300

Weekly <5 260

Two-element film badge 1948–1958 Weekly 40 2,080

1958–1961 Monthly 40 480

Four-element film badge 1961–1980 Quarterly 40 160

Two-element TLD dosimeter 1980–1989 Quarterly 20 80

Four-element TLD dosimeter 1989–present Quarterly 10 40

Neutron dosimeters

NTA film 1948–1980 Biweekly <50 1,300

Monthly <50 600

Quarterly <50 200

Combination NTA film and 

TLND dosimeter

1980–1989 Quarterly <50 200

TLND dosimeter 1989–present Quarterly 10 40

A.5.5.3 Year

Analysis of the missed photon dose by year and dosimeter exchange frequency is summarized in 

Table A-6. The missed photon dose for workers who were unmonitored for external occupational 

radiation exposure before 1961 is also discussed in Section A.5.4.

A.5.5.4 Energy Range

An estimate of the missed dose by energy range might be possible based on the type of facility and 

predominant radiation sources or radionuclides at the facility. The recorded dose from the dosimeter 

response does not typically provide information to estimate discrete energy ranges. It is possible to 

examine the energy response characteristics of the multielement film and TLD dosimeters, but this 

analysis does not recognize the substantial uncertainties in the workplace that are associated with 

differing exposure geometries and mixed radiation fields. 

A.5.6 Neutron Missed Dose

The estimated MDLs for the neutron dosimeters that were used at Y-12 are summarized in Table A-6. 

It is possible to calculate the missed neutron dose at Y-12 using the MDLs because the neutron 

dosimeters were calibrated with neutron sources that had energies similar to those in the workplace 

and more than 90% of the neutrons to which workers were normally exposed had energies greater 

than the 500-keV threshold of the NTA film dosimeters. If the MDL for NTA film is used in estimating 

the missed neutron dose, it should be multiplied by 1.10 for workers in the Calibration Laboratory and 

by 1.05 for workers in the Enriched Uranium Storage Area of Building 9212 and the Nondestructive 

Analysis Laboratory. It is also possible to estimate missed neutron dose in some facilities by use of 

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photon dose ratio is expected to provide a reasonably reliable estimate of the missed dose is for 

workers in the Enriched Uranium Storage Area of Building 9212. For this area, the neutron-to-dose 

ratio was approximately 1:1. The neutron-to-dose ratios for other neutron exposure areas at Y-12 

were quite large and the use of these data could lead to gross overestimates of missed neutron dose 

if an individual was also exposed to pure photon sources during the course of work at either the 

Calibration Laboratory or the Nondestructive Analysis Laboratory.

A.5.7 Organ Dose Equivalent

Once the adjusted photon and neutron doses have been calculated for each year, the values are used 

to calculate organ doses of interest using the NIOSH External Dose Reconstruction Implementation 

Guideline (NIOSH 2002). There are many complexities and uncertainties when applying organ dose 

conversion factors to the adjusted doses of record. Many of the factors that affect the recorded dose 

have already been discussed in the various tables throughout this section. Some factors such as 

backscattering (phantom calibration) and the over-response of low-energy photons would indicate the 

recorded dose was too high. Other factors such as calibration methodology, angular response, lowenergy threshold, and film fading would result in a recorded dose that was too low. As a result, 

differences in film badge design (filtration) and calibration can have both positive and negative effects 

on the overall dose comparison to Hp(10). ICRU (1988) indicates that film badge dosimeters, while 

not tissue equivalent, can be used for personnel dosimetry. The report emphasizes, however, that it 

is difficult to establish the variations in film response as functions of photon energy and angle of 

photon incidence on the film badge at very low photon energies. Given these broad uncertainties, 

especially with film badge dosimetry in the 1950 to 1980s, an approach is used to estimate organ 

dose that is favorable to the claimant. Because use of exposure-to-organ DCFs results in a higher 

organ dose and higher probability of causation, given the REFs of the intermediate-energy photons, 

these DCFs should be used to convert recorded film badge doses to organ dose. In the conversion of 

all recorded photon and neutron doses to organ doses, the exposure geometry must also be given 

careful consideration. The proposed default options based on exposure geometries for long-term 

workers that are favorable to the claimant are listed in Table A-7.

Table A-7. Default exposure geometries favorable to the claimant 

for calculating organ dose. 

Claim status

Job 

category

Exposure 

geometry Percentagea

Non-compensable All A-P 100

Compensable worker All A-P 50

Rotational 50

Compensable supervisor All A-P 50

Isotropic 50

a. Apply this percentage to the dose conversion factor in Appendix B of NIOSH 

(2002) to arrive at the total organ dose equivalent from the adjusted recorded 

dose.

A.5.8 Parameter #2

Parameter #2 is the standard deviation of the normal distribution for the organ dose. The individual 

dose result for each dosimeter exchange period is normally available to calculate the mean and 

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standard deviation for each year. If it is not available, the adjusted organ dose can be used for each 

year, and a default standard deviation value can be used for parameter #2.

A.5.9 Organ Dose Conversion Factors

A detailed discussion of the conversion of measured dose to organ dose equivalent is provided in 

Appendix A of NIOSH (2002). Appendix B of NIOSH (2002) contains the appropriate DCFs for each 

organ, radiation type, and energy range based on the type of monitoring. In some cases, simplifying 

assumptions are appropriate. 

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TABLE OF CONTENTS

SECTION TITLE

B.1 Calutron Related Uranium Enrichment Activities from March 1, 1943 

through December 31, 1947.................................................................................................. 

PAGE

54

B.2 Operators (Control or Cubicle Operators)............................................................................... 55

B.3 Other Workers........................................................................................................................55

B.1 CALUTRON RELATED URANIUM ENRICHMENT ACTIVITIES FROM MARCH 1, 1943 

THROUGH DECEMBER 31, 1947

The Y-12 calutrons used an electromagnetic process to separate 235U from natural uranium. This 

enrichment process began in Building 9731 in November 1943 when the operations supervisors and 

the first crews were trained on the use of the alpha and beta calutrons. These calutrons were in 

operation until December 23, 1946 (Wilcox 2001). 

The first run of a production (alpha) calutron occurred in Building 9201-1 on January 27, 1944. There 

were 864 alpha calutrons operated in Buildings 9201-1, 2, 3, 4, and 5. These operations were shut 

down by September 22, 1945. The beta calutrons began operations in March 1944. There were 288 

beta calutrons operated in Buildings 9204-1, 2, 3 and 4. These operations ceased by December 23, 

1946, marking the end of the calutron enrichment process (Wilcox 2001).

Each calutron was connected to a separate control panel called a cubicle, located about 30 ft from the 

calutron track. There were many dials, gauges, switches, and other controls on the front of the 

cubicle. The operator (control or cubicle operator) had to sit or stand in front of the cubicle and watch 

instrument readouts, make adjustments to maximize calutron performance, and record the calutron’s 

performance. If the operator could not adjust the calutron to run properly, the operator would call a 

maintenance person to make additional adjustments. In addition to the cubicle operator, electricians 

and other maintenance employees worked in the calutron area (Tankersley 2008a, 2008b). 

Both the alpha and beta cubicles and calutrons had high-voltage rectifier tubes, known as kenetrons, 

which could generate appreciable X-ray exposure levels. These tubes usually operated in the 50 to 

100 kV range, but the voltage could go as high as 150 kV or more with a sudden drop in potential 

across the tube (Sterner and Riley date unknown, pp. 11–12). 

Smith (1944a) made measurements of X-rays around the cubicles in Buildings 9731 and 9201-1 and 

the calutrons in Building 9731 using Eastman Type K Industrial X-ray film in dental packages. The 

response of the film to photon energy was determined, and the film was calibrated by Eastman Kodak 

using an ionization chamber (Dahl 1944). The operating voltage of the rectifier tubes was assumed to 

be 150 kV. Unless a specific measurement distance is specified in the report, it is assumed that the 

film was fixed to the cubicle or calutron. 
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The measurements were made on all sides of the cubicle and in the immediate area around the 

cubicles and the calutrons. However only the measurements made on the front of the cubicle are 

used in this analysis because that was the position where the operator worked. The data shown in 

Table B-1 were used to bound the operators’ external radiation exposure.

Table B-1. Cubicle x-ray radiation measurements (Smith 1944a). 

Film position during test

Cubicle 

number

Exposure

duration (h)

Exposure 

(R/8-h day)

Annual dose 

equivalent (rem)

XAX Cubicle in Bldg. 9731

Front window over 

instrument panel

1 53 0.0003 0.08

2 42.3 0.0008 0.20

XBX Cubicle in Bldg. 9731

Front window over 

instrument panel in direct 

line with regulator tube

1 11.5 0.001 0.25

11.5 0.001 0.25

12.2 0.008 2.00

Front window over 

instrument panel in direct 

line with regulator tube

2 20.6 0.011

32 0.005 2.75

1.25

Complementary X-ray exposure measurements were made using a condenser R-meter chamber by 

the General Electric Company (Schmidt, Jr. 1944a, pp. 5–6; 1944b, pp. 13–15). These 

measurements showed that the two sets of measurements were comparable and represented the 

X-ray radiation levels that the control operators were exposed to. 

Unshielded X-ray radiation levels from some equipment and to certain workers could have exceeded

0.1 R/d. The potential for exposure to these levels was recognized by TEC management. Shielding 

in combination with limiting the time employees worked near the equipment was used to keep 

exposures below the limit of 0.1 R/d established by the National Bureau of Standards (Hull 1944, p. 

12; Smith 1944b, p. 7). The accepted value of 0.1 R/d was exceeded only on rare occasions (Sterner 

and Riley, date unknown, p. 11). 

B.2 OPERATORS (CONTROL OR CUBICLE OPERATORS)

The film data can be used for an overestimate of X-ray exposures to cubicle operators that is 

favorable to the claimant. The highest measured exposure rate at the front of a cubicle was 0.011 R 

per 8-hour day, assuming that the rectifier was operating at 150 kV (Table V, Smith 1944a), which 

best represents the high end of the voltage range typically observed at the cubicle (Dahl 1944). This 

exposure rate applies to all control or cubicle operators.

B.3 OTHER WORKERS

Maintenance and/or trades workers were in the calutron enrichment area to monitor, adjust, and 

repair the equipment. They routinely looked through the windows (viewing ports) in the rear of the 

cubicle to ensure that the rectifier tubes were working properly. 

Electricians likely spent more time in and around the cubicles and the calutrons because these were

electrical equipment. They also would have worked in the cable and tube test operation. Here, 

overvoltages were applied to this electrical equipment to test it before putting it into service. Sterner 

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and Riley (date unknown, p. 12) said that testing operators could have been exposed to levels of 1-2 

R/d before the equipment being tested was enclosed with lead shields and leaded glass. This action 

reduced the exposure levels to 0.1 R/d. This is the only reference found where this operation is 

mentioned. Based on this limited information, NIOSH assumes that electricians were exposed at 

levels equal to the daily exposure limit of 0.1 R. 

Other maintenance and trades workers presumably had other duties and did not spend as much time 

near the cubicles and calutrons as the electricians did. Therefore, NIOSH assumes that daily 

exposure is one-half of the daily limit which is 0.05 R.

For all of the exposures to control operators and other calutron related workers, it is assumed that the 

listed doses are WB rather than partial-body doses and that the worker was exposed at the location 

where the exposure was measured rather than the worker’s actual physical location. The resultant

measured exposure rates are used to bound the worker’s annual, WB dose equivalent. Table B-2 is a 

summary of the recommended maximum annual dose assignments for various occupations. If an 

Energy Employee worked for only part of a year, the dose reconstructor can adjust the annual doses 

accordingly. 

Table B-2. Bounded WB annual dose equivalent for calutron-related uranium enrichment 

workers, March 1, 1943 to December 31, 1947. 

Occupation

Bounded annual WB annual

dose equivalent (rem)

Basis for bounded 

dose equivalent Reference

All control (cubicle) 

operators

2.75 Film measurements Smith (1944a) 

Electricians 25 Exposure limit Sterner and Riley

(date unknown)

All other maintenance

and trades workers

12.5 Exposure limit Sterner and Riley

(date unknown)