ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Mound Site – Occupational 

External Dosimetry

Document Number:

ORAUT-TKBS-0016-6 

Effective Date: 08/11/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 51

Subject Experts: C. Wesley Proctor, Jr. and Faisal Alguitifan

Document Owner

Approval: Signature on File Date: 08/04/2004

Jeff Vollmer, TBD Team Leader

Approval: Signature on File Date: 08/10/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 08/08/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 08/11/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS 

Section

Record of Issue/Revisions ...................................................................................................................

Page

4

Acronyms and Abbreviations ...............................................................................................................5

6.1 Introduction and Brief History of Dayton Project and Mound Facility ........................................... 6

6.2 Basis of Comparison...................................................................................................................7

6.3 Dose Reconstruction Parameters ...............................................................................................8

6.3.1 Mound Administrative Practices .........................................................................................9

6.3.1.1 Dosimeter Assignment Policies and Dosimetry........................................................ 10

6.3.1.2 Historic Dose Limit Standards.................................................................................. 10

6.3.1.3 Dosimeter Exchange Policies .................................................................................. 10

6.3.1.4 Policies To Record Notational Dose ........................................................................ 11

6.3.1.5 Policies To Estimate Dose from Lost or Damaged Dosimeters ................................ 12

6.3.1.6 Unmonitored Doses at Mound ................................................................................. 13

6.3.2 Dosimeter Types and Badge Distribution ......................................................................... 15

6.3.3 Energy Spectra ................................................................................................................24

6.3.3.1 Beta/Photon Calibration........................................................................................... 24

6.3.3.2 Neutron Energy Calibration...................................................................................... 24

6.3.3.3 Neutron Energy Spectra .......................................................................................... 24

6.3.3.4 Photon Energy Spectra............................................................................................ 25

6.3.3.5 Dose Calculation for Film Badges............................................................................ 25

6.3.3.6 Dose Calculation for Thermoluminescent Dosimeters.............................................. 26

6.4 Missed or Underreported Photon or Beta Dose......................................................................... 28
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6.5 Missed or UnderReported Neutron Dose at Mound................................................................... 29

6.6 Dosimeter Uncertainty Analyses ............................................................................................... 31

6.6.1 Photon Dosimeters........................................................................................................... 31

6.6.2 Neutron Dosimeters ......................................................................................................... 32

6.6.3 General Field Characterization......................................................................................... 34

6.6.3.1 Photon Field ............................................................................................................ 35

6.6.3.2 Beta Field ................................................................................................................ 36

6.6.3.3 Neutron Field........................................................................................................... 37

6.6.3.4 Mixed Field ..............................................................................................................37

6.7 Workday Hours and Exposure Geometry.................................................................................. 37

6.8 Guide to Important Tables......................................................................................................... 38

References ........................................................................................................................................39

Glossary ............................................................................................................................................44

Attachment 6A Summary Instructions for Mound Dose Reconstructors ........................................... 50
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LIST OF TABLES

Table

6-1 Synopsis of major programs with external dose implications....................................................

Page

8

6-2 Dosimeter assignment policies............................................................................................... 10

6-3 Historic dose limit standards .................................................................................................. 11

6-4 Dosimeter exchange frequency.............................................................................................. 11

6-5 Dose distribution from 1948 to 1959....................................................................................... 14

6-6 Mound neutron dosimeters..................................................................................................... 15

6-7 Neutron flux values for 300 mrem/40-hr wk dose ................................................................... 16

6-8 Radiation weighting factors for neutrons ................................................................................ 17

6-9 Radiation energy intervals and ICRP 60 correction factors by facility and 

process ..................................................................................................................................18

6-10 Neutron TLD card components .............................................................................................. 19

6-11 Neutron track fading in NTA film versus neutron energy ........................................................ 20

6-12 Neutron track and signal fading study and corrective action history ....................................... 20

6-13 Beta-photon dosimeter characteristics ................................................................................... 23

6-14 Unmonitored photon dose at Mound ...................................................................................... 23

6-15 Beta/photon TLD card components........................................................................................ 26

6-16 Missed photon dose...............................................................................................................28

6-17 Missed or underreported neutron dose................................................................................... 30

6-18 Photon dosimeter limits of detection....................................................................................... 31

6-19 Neutron dosimeter limits of detection ..................................................................................... 31

6-20 Photon dosimeter relative uncertainty .................................................................................... 33

6-21 IARC testing for three photon qualities performed on US beta/photon 

dosimeter...............................................................................................................................33

6-22 Neutron dosimeter relative uncertainties ................................................................................ 34

6-23 Externally hazardous radionuclide matrix ............................................................................... 35

6-24 Externally hazardous beta emitters from polonium production at Mound................................ 36

6A-1 Corrected neutron doses........................................................................................................ 50

6A-2 Corrected photon doses......................................................................................................... 51

LIST OF FIGURES

Figure

6-1 Neutron flux producing 300 mrem/40-hr exposure rate versus neutron 

energy....................................................................................................................................

Page

17

6-2 Cross-section versus neutron energy for several common 

neutron-detection processes .................................................................................................. 21
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 01/12/2004 00-A New Technical Basis Document for the Mound 

Site – Occupational External Dose. Initiated by 

Jeff Vollmer.

Draft 04/14/2004 00-B Incorporates NIOSH and internal review 

comments. Initiated by Jeff Vollmer.

Draft 05/28/2004 00-C Incorporates additional comments. Initiated by 

Jeff Vollmer.

Draft 07/16/2004 00-D Incorporates additional NIOSH comments. 

Initiated by Jeff Vollmer.

08/11/2004 08/11/2004 00 First approved issue. Initiated by Jeff Vollmer.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 5 of 51

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission (a DOE predecessor agency)

AP anterior–posterior

C Celsius (with degree symbol)

cm centimeter

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

ERAD External Radiation Analysis Data

EXAS External Exposure Analysis System

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiological Units and Measurements

keV kilovolt-electron, 1,000 electron volts

LOD limit of detection

MDL minimum detectable level

MESH Mound Environmental Safety and Health

MeV megavolt-electron, 1 million electron volts

mm millimeter

mo month

MPL maximum permissible limit

mR milliroentgen

mrad millirad

mrem millirem

NBS National Bureau of Standards

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

qtr quarter, 3 months

R roentgen

RBE relative biological effectiveness

RTG radioisotope thermoelectric generator

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wk week

yr year
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6.1 INTRODUCTION AND BRIEF HISTORY OF DAYTON PROJECT AND MOUND 

FACILITY

This technical basis document (TBD) contains an analysis for external exposure received by Mound 

Laboratory workers while inside site facilities. Discussions include analysis for missed doses for 

monitored workers and estimated doses for unmonitored workers. The discussion covers dosimetry 

practices at Mound from its establishment as part of the Manhattan Engineering District project until 

the planned closure of the facility in 2006. 

TBDs and Site Profile Documents are general working documents that provide guidance concerning 

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 the National Institute for Occupational Safety and Health (NIOSH) 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 facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 [42 U.S.C. Sections 7384l(5) and (12)].

Beginning in September 1943, the Monsanto Chemical Company investigated polonium and its 

production at the Scioto research laboratory in Dayton, Ohio. This was the Dayton Project, a part of 

the Manhattan Project. The project moved to Miamisburg, Ohio, in 1948 and became the Mound 

Laboratory in 1949. Mound was the first U.S. Atomic Energy Commission (AEC) facility. The Mound 

facility supported many activities from 1946 to 1992. The name derives from the large prehistoric 

Adena Indian burial mound across the road from the facility.

In comparison with other AEC facilities, the Mound site was never very large, about 2 square miles in 

area. Of that, about 306 acres were actively used. The main use area is on a steep-sided glacial 

moraine jutting about 300 feet above the valley of the Great Miami River. The significance of this 

topography is that the wind blows over the Mound facility almost constantly. There is very little 

stagnant air on the Mound site, so airborne radiological releases disperse quickly. This TBD does not 

address airborne releases. 

During the initial part of the Manhattan Project the Dayton facility separated 210Po from naturally 

occurring materials for PoBe source neutron generators for atomic bomb initiators. By 1944 the 

decision was made to transmute 209Bi into 210Po via neutron bombardment (Warren 1944). The 

reaction is: 83 209Bi + η→ 83 210Bi → 84 210Po + β-. Initially, 58-pound bismuth bricks were irradiated in 

the Clinton reactor at Oak Ridge. After January 1945 irradiated bismuth was the sole source of 

polonium (Wolf 1945). The beta radiation from the irradiated slugs was intense enough that lead 

gloves and tongs were necessary to handle them (Ferry 1944a). When the Mound facility opened in 

1948, the bismuth was cast into slugs that were then inserted into aluminum cans, which were welded 

shut. The canned bismuth was then irradiated at the Hanford facility and shipped to Mound in lead 

casks. The casks were emptied into a pool on the second floor of the T Building. After testing 

mechanical cutting, chemical dissolution, and melting, it was decided that the bismuth could best be 

separated from the aluminum can by dissolution in a bath of 17% HCl. The bismuth was then 

dissolved in a bath of HCl and HNO3. This separation occurred on the second floor of the T Building. 

It is Important that both the aluminum and the bismuth contained impurities including iron, silicon, 

cobalt, lead, tin, zinc, silver, chromium, vanadium, and gallium. During irradiation these impurities 

were neutron activated and produced gamma-emitting isotopes that created an external radiation 

hazard during the bismuth processing.
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The second phase of the polonium project consisted of manufacturing polonium heat sources for 

radioisotope thermoelectric generators (RTGs). The SNAP 3A polonium-powered RTG was built in 

1958. However, because of the short half-life of 210Po, this program was halted after 1969 and 

cancelled in 1971 when cleanup began. 

Because of the short half-life of 210Po, a program was begun in the late 1940s to find other 

longer-lived alpha emitters that could function in bomb initiators, which began the radium and actinium 

programs at Mound. Actinium-227 could be made from 226R via neutron irradiation. The reaction is: 

226Ra + η → 227Ra → 227Ac + β-. All the associated research projects were carried out at Mound in 

the late 1940s and early 1950s. In July 1953 the program was de-emphasized.

The plutonium project at Mound began with 238Pu research in the mid-1950s. Research on 238Pu was 

transferred from the Lawrence Livermore National Laboratory to Mound in 1959. Mound produced 

over 500 RTGs from 1961 to 1992. 

In the mid-1950s Mound was involved in a thorium refinery program. This was designed to develop a 

system for refining thorium for the breeder reactor program. The Monex process pilot plant was 

assembled in Room 1B of the SW Building in the spring of 1955. Construction on a refinery on the 

west side of the SW Building was begun on March 11, 1955, but was cancelled on May 9, 1955.

In November 1955 Mound received a directive from AEC to construct a facility to research the process 

chemistry of 230Th, also called ionium. This was to be used as a tracer element in the Redwing 

nuclear weapons tests. This program was terminated late in 1956. There was also a brief 231Pa 

program begun in 1954 in Room 167 of the R Building when 231Pa was selected as a research 

surrogate for 233Pa. A pilot plant was installed in the HH Building in July 1955, but after March 1956 

work was moved to the SW Building and used equipment from the Monex process.

From the early 1960s through the mid 1970s Mound was involved again in the ionium and 

protactinium programs. 

The last program at Mound was the program to produce and separate rare radioactive isotopes. It 

began at Mound in the mid 1950s and continued until 1985. The processes were carried out in the 

SW, SM, and R Buildings (DOE 1993). 

Mound continued its role in nuclear weapons development and production until 1992. The U.S. 

Department of Energy (DOE) has transferred all nuclear work from Mound to other DOE facilities. 

As an aid to understanding the health physics program at Mound, Table 6-1 lists the eight major 

programs, with external dose implications, that were conducted there over five decades. 

6.2 BASIS OF COMPARISON

Since the Manhattan Engineering District project began in the early 1940s, a great many radiation 

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

of the comparison for dose reconstruction, as described in External Dose Reconstruction 

Implementation Guidelines (NIOSH 2002), is the personal dose equivalent, Hp(d), where d identifies 

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

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

the lens of the eye, d is 0.3 mm and is noted as Hp(0.3). For penetrating radiation of significance to 

whole-body dose, d is 10 mm and is noted as Hp(10). The International Commission on Radiological 
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Table 6-1. Synopsis of major programs with external dose implications. 

Program Time frame Products Major isotopes of interest Building

Polonium A 1943-1959 Po-210 in PoBe & PoX 

neutron generators. 

Bi-210, Fe-59, Sc-46, Co-60, 

Pb, Sn, Zn-65, Ag-110m, Cr, 

Va, Ga, Hg-203

T, HH 

Polonium B 1953-1969 Po-210 as a heat source. 

SNAP 3A and Poodle RTGs. 

Po-210 T, HH 

Radium-actinium 1949?- 6/1953; 

4/1954-12/1954 

Ac-227Be neutron sources Mallinckrodt K-65 residues, 

Ra-226, Ra-227, Ac-227, 

Th-228 

R, SW including 

Old Cave 

Plutonium 1959-1995 Pu-238 heat sources. 500+ 

Pu RTGs 

Pu-238, Pu-238O-18 PP, WD SW 

Thorium-232 Refinery 

(Monex process) 

1954?- 5/1955; 

sludges stored 

until 7/1975 

Th-232 for breeder reactor 

program 

Th-232 containing sludges 

from uranium extraction and 

scrap 

SW 1-B, stored 

in warehouse 15, 

W, G, Quonset 

hut, bldg. 21

Thorium-230 (Ionium) 11/1955 -1956 Tracer element for Redwing 

weapons test 

Th-230 and Pa-231 containing 

sludges from uranium 

extraction (Cotter concentrate 

from St. Louis) 

R 

Protactinium-231 & 

233 

1954 – 9/1979 Pa-231 was a surrogate for 

study of the properties of 

Pa-233 

Th-230 and Pa-231 containing 

sludges from uranium 

extraction (Cotter concentrate 

from St. Louis); products of 

irradiation of Th-230 & Th-232 

R 167 & 145; HH 

’7/55-3/’56; SW 

4/’56 onward 

Rare isotopes Mid 1950’s1985 Pu-238; Pa-231; Th-228,229, 

230; U-233, 234; Ra-226; 

Ac-227; Po-209 

Uranium and thorium sludges, 

Pu-238, U-233, target 

irradiation 

SW 22, 132, 

140; SM 1; R; 

bldg. 21 

Units and Measurements (ICRU) has recommended Hp(0.07), Hp(0.3), and Hp(10) for use as the 

operational quantities to be recorded for radiological protection (ICRU 1993). In addition, DOE has 

used these quantities in its Laboratory Accreditation Program (DOELAP) since the 1980s for 

personnel dosimetry systems. In the body of this TBD, the doses have not been converted to Hp(10)

values.

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

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

collision dose, multiple collision dose, and dose equivalent index. The starting point for neutron dose 

reconstruction is the recorded neutron dose values. Additions to the neutron dose would only involve 

adjustments for missed dose due to the limits of detection (LODs) of the dosimeters. The distribution 

of missed dose is log normal. Measured dose uncertainty includes fading, missed low-energy dose, 

and so forth, and its distribution is Gaussian. Table 6-9 later in this document sets forth the 

International Commission on Radiological Protection (ICRP) Publication 60 correction factors (ICRP 

1991) that dose reconstructors should apply to the adjusted dose of record. Attachment 6A shows all 

of the doses used for reconstruction converted to Hp(10) values. 

6.3 DOSE RECONSTRUCTION PARAMETERS

Examination of the beta, photon, and neutron radiation types, energies, and geometries of exposure, 

and the characteristics of the respective Mound dosimeter response are crucial to the assessment of 

bias and uncertainty of the original recorded dose in relation to the Hp(10) radiation quantity. The 

bias and uncertainty for current dosimetry systems is typically well documented for Hp(10). At some 

facilities the performance of current dosimetry can be compared with performance characteristics of 

historic dosimetry systems in similar facilities or workplaces. In addition, current performance-testing 
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techniques can be applied to earlier dosimetry systems to achieve a consistent evaluation of all 

dosimetry systems.

Mound operations included many different tasks in many different facilities, but almost all the work 

involved either experimenting with or building neutron or heat sources. Therefore, the geometry of the 

workers in relation to radiation sources is assumed to be anterior-posterior (AP). Unlike some other 

facilities, in particular Hanford and Savannah River, Mound never had a well-documented quantitative 

study of its external dosimetry programs. The Mound dosimetry program is clearly derivative. It 

began in 1948 by emulating the program at Oak Ridge National Laboratory (Meyer 1994, Volume I, 

1948 Appendix III, pp. 31-34).1

Under NIOSH guidelines, the overall accuracy and precision of the original recorded and individual 

worker doses and their comparability depend on four factors:

 Twenty years later Mound was emulating the Nevada Test Site 

thermoluminescent dosimeter (TLD) program (Meyer 1994, Volume II, p. 104). By 1980, Battelle 

Columbus Laboratory was implementing the Mound TLD system (Meyer 1994, Volume V, p. 286). In 

between, there were comparisons with the Hanford and Savannah River systems.

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

technical, administrative, and statutory compliance considerations

• Dosimetry technology, which includes the such physical capabilities of the dosimetry system 

as the response to different types and energies of radiation (mixed radiation fields in particular) 

• Calibration of the respective monitoring systems and similarity of the methods of calibration to 

sources of exposure in the workplace

• Workplace radiation fields that can include mixed types of radiation as well as variations in 

exposure geometries and environmental conditions

6.3.1

Administrative practices germane to dose reconstruction include policies to:

Mound Administrative Practices

• Assign dosimeters to workers

• Exchange dosimeters

• Record notational dose 

• Estimate dose for missing or damaged dosimeters

• Replace destroyed or missing records

• Evaluate and record dose for incidents

• Obtain and record dose from exposures while working for another employer

Mound had policies for all these administrative practices, although the policies changed over time.

 

1. History of Personal External Dosimetry Program at the Dayton Project and Mound Laboratory (Meyer 1994) is a 

2,614-page compilation covering the 46 years from 1946 to 1993. The volume numbers and appendixes are from the Meyer 

report. Appendix numbers begin anew for each year. There are no page numbers in the Meyer report itself; the cited page 

numbers are from the Adobe Acrobat Reader electronic copy of the report furnished to the TBD authors. 
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6.3.1.1 Dosimeter Assignment Policies and Dosimetry 

Beginning as early as February 1944, the Dayton Project undertook a personnel monitoring program 

to measure penetrating gamma radiation received by employees in the process of extracting 210Po 

from irradiated bismuth slugs from the Clinton reactor at Oak Ridge. Table 6-2 lists the history of 

Mound dosimetry. In Table 6-2 and the remainder of this document 1 wk means a 40-hr workweek in 

a 7-day period ending Friday unless otherwise noted.

Table 6-2. Dosimeter assignment policies. 

Date Photon dosimetry

February 1944 Finger ring (beta) and whole-body film badges provided to some workers. Weekly limits & reports. 

August 1946 Wrist film badges provided to production employees. No individual employee data recoverable until 

January 17, 1947. 

February 21, 1949 Operations began in T Building. Film badges, one sensitive and one insensitive film, with 1-mm 

cadmium filter read weekly and two pocket ion chambers read daily. Beta window not read. 

December 1951 Film badges read every 2 wk. 

May 1966 Use of pocket ion chambers discontinued. 

September 1968 New dosimeters with security credential. Based on Nevada Test Site badges but with different 

filters. Modified later with TLD photon dosimeters plus Nevada Test Site neutron dose film. In 1977, 

questions were raised about low-energy photon dose underestimates with tantalum shields. 

1972 Routine extremity dosimetry began. 

December 1977 All-TLD dosimeters introduced. 

October 1986 New soft-pack dosimeters. 

June 1991 New hard-pack dosimeters with neutron track etch capability. 

June 1993 Added neutron track etch capability for future use. 

Date Neutron dosimetry

Before September 

1949

No neutron dosimetry data found. Doses estimated at 50 mrem/40-hr wk based on historic modal 

dose. 

September 1949 NTA track etch film for fast neutrons first used in T and R Buildings. Read 10 fields at 980 power 

magnification and averaged readings. 

September 1956 Began reading 10 fields at 430-power magnification with projection microscope and averaged 

values. 

December 2, 1957 Returned to using 980-power projection microscope to read NTA film. 

January 21, 1963 Began routine monitoring of night maintenance staff in Building P. 

March 1963 Began using 400-power projection microscope. Reading 64 fields and averaging. 

October 1964 Began using new projection microscope with better light. 

March 1966 Began badging all hourly Engineering Department staff members who spent a portion of their time in 

radiation control areas of plant. 

April 1967 Began using another new projection microscope, first at 500 power, then at 1,000 power. 

August 1, 1968 Exchange frequency for Buildings R and SW personnel changed to 4 wk.

Nov./Dec. 1977 All-TLD dosimeters introduced. Began using TLDs for neutron exposure monitoring. 

October 1986 New soft-pack dosimeters. 

June 1991 New hard-pack dosimeters with neutron track etch capability. 

June 1993 Added neutron track etch capability for future use. 

6.3.1.2 Historic Dose Limit Standards

The record on historic dose limit standards is ambiguous. As a consequence, a great deal of the 

record must be inferred. Table 6-3 lists the best set of dose limit standards that can be derived from 

the record. 

6.3.1.3 Dosimeter Exchange Policies

By February 1944 the rings and film badges were exchanged weekly (Ferry 1944b). Between 

February 1949 and May 1966, pocket ion chambers were exchanged and read daily. After November 
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1951, film badges were exchanged biweekly. Later, film badges were exchanged monthly or quarterly 

for those who worked in lower radiation areas. Table 6-4 lists the details. 

Table 6-3. Historic dose limit standards. 

Period Radiation type Dose limits

12/1943-1946(?) Photon 100 mrem/d, 26rem/yr γ; 500 mrem/d-3.5/wk β, 182 

rem/yr β a 

August 1946–November 1970 Photon 300 mrem/40-hr wk; 15 rem/yr 

December 1970–December 1988 Photon 3 rem/qtr; 12 rem/yr not to exceed an accumulated value 

of 5(N-18), N = age in years; 200 mrem/wk

January 1, 1989–1997 Photon + neutron 5 rem/yr 

August 1, 1949–January 1, 1985 Neutron 300 mrem/40-hr wk; 15 rem/yr 

January 1985–December 31, 1988 Neutron 200 mrem/40-hr wk; 12 rem/yr; 3 rem/qtr 

January 1, 1989–1997 Neutron + photon 5 rem/yr 

1997–2006 Neutron + photon + 

internal dose 

5 rem/yr 

a. Source: Ferry (1944b). The doses are from finger ring (beta) and body badges (gamma); 26 rem/yr gamma and 182 

rem/yr extremity beta are imputed (52 × weekly limits). The records appear to indicate that only weekly dose limits were 

imposed. 

Table 6-4. Dosimeter exchange frequency. 

Period Dosimeter type

Dosimeter 

holder

Exchange 

frequency Comments

1943-1946(?) Dental film Ring and brass 

badge

Weekly Not all workers were 

monitored 

8/1946–1/1949 Kodak dental films Unknown Weekly Films worn on wrist only

2/1949–11/1951 DuPont D552 Oak Ridge/steel Weekly All radiation workers

12/1951–1957 DuPont D552 Oak Ridge/steel Biweekly All radiation workers

1957–1960 DuPont (model unknown) Oak Ridge/steel Biweekly All radiation workers

1961–1962 DuPont 558 Oak Ridge/steel Biweekly All radiation workers

1963–1968 DuPont D556 Oak Ridge/steel Weekly Some radiation workers

Biweekly Most radiation workers

Monthly General area workers

1968–1969 DuPont D556 Oak Ridge/steel Biweekly Dependent on work area

Monthly

Quarterly

1970–1977 Kodak Type 3 Mound holder Biweekly Dependent on work area

Monthly

Quarterly

1977–1986 Harshaw 8810 TLD-700 Cycolac Biweekly Dependent on work area

Monthly

Quarterly

1986–1987 Harshaw 8810 TLD-700 Cycolac Monthly Dependent on work area

Quarterly

1987–1991 Harshaw 8810 TLD-700 Cycolac Monthly All employees monitored

Quarterly

1991–present Harshaw 8801 TLD-700 Cycolac Monthly Dependent on work area

Quarterly

6.3.1.4 Policies To Record Notational Dose

By February 22, 1944, extremity beta doses using finger ring film badges and whole-body photon 

doses using film badges were reported by worker name (Ferry 1944b). The dose limits were per 

workday and week, so there is no indication that a continuous record of the doses was kept, only that 

weekly reports were issued. Beginning in 1946, all doses were recorded. The 1946 results were 
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entered by code rather than by worker name. With the exception of two names entered for the week 

of July 19, 1946, the code-name listing is no longer available (Author unknown 1947). The records 

that begin with the workweek ending Friday, January 17, 1947, have names associated with the 

recorded doses. The historical threshold of film response to photons was 40 mrad. Several entries 

greater than 40 mrad in the 1946 data cannot be assigned to anyone because of the lost coding. 

After polonium operations began in February 1949, the results of the pocket dosimeter measurements 

were entered in each employee’s radiation exposure record on Form 1015-X, which remained in use 

until December 1959. Beginning on May 4, 1959, neutron exposure data were entered on punch 

cards and processed by the External Exposure Analysis System (EXAS) program. Beginning on 

January 2, 1960, results of both photon and neutron monitoring film were entered on punch cards 

(Meyer 1994, Volume I, 1960 Appendix I, pp. 133–138). After this time, no film exposure or film dose 

data were entered on Form 1015-X.

Beginning on March 17, 1963, pocket dosimeter results were entered into a file-card system. The use 

of punch cards for recording pocket dosimeters ended, and a Kardex file cabinet was purchased for 

the file cards. The file cards were used from March 25, 1963, until December 27, 1963 (Meyer 1994, 

Volume I, 1963 Appendix III, p. 263). Beginning on December 30, 1963, pocket dosimeter data were 

entered on Form 1333, which carried a revision date of March 19, 1963. This form, which permitted 

the use of a Rolodex V-File for more efficient handling of daily records, remained in use through May 

1966.

The External Radiation Analysis Data (ERAD) program replaced EXAS in 1978, and the Mound 

Environmental Safety and Health (MESH) program replaced ERAD in September 1989. The MESH 

database contains all radiation doses since 1947, although only the annual summary dose is available 

for years before 1974. 

6.3.1.5 Policies To Estimate Dose from Lost or Damaged Dosimeters

Beginning in 1949, each person carried a film dosimeter and two pocket ion chambers every day. 

The ion chambers were read the next day and their readings recorded. The program included two 

pocket ion chambers because they were subject to false readings, usually high, from mechanical 

stress and heat. The system was, therefore, redundant in relation to damaged or lost pocket 

dosimeters.

The available records do not indicate any policies for recording dose due to lost or damaged film or 

TLD photon or neutron dosimeters.

Some records were missing or destroyed. For example, at the end of 1977, the Dosimetry group was 

unable to recover any individual monitoring data for the previous years from the mainframe computer. 

Therefore, the system generated a summary report of the annual exposures of all monitored 

employees from 1947 through 1977 (Guido 2003; Meyer 1994, Volume III, 1973 Appendix 3, 

pp. 207-211).

Reports written for each incident at the plant contain descriptions of the incident, health physics data 

for the individuals affected, and recommendations for improved safety for the operations involved.

There were provisions for reporting radiation doses received by individuals who worked at other sites 

before working at Mound (Meyer 1994, Volume III, 1973 Appendix 3, p. 210).
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6.3.1.6 Unmonitored Doses at Mound 

Until February 1949, Mound monitored the production workers but not researchers or administrative 

personnel. Production workers wore the wrist film badges. Therefore, the dose of record before

February 1949 is extremity dose only rather than whole-body dose. Because the body is usually 

farther from the radiation source than the wrist, this analysis assumed that the unmonitored 

whole-body dose would be less than the monitored extremity dose. 

The earliest documented reference to neutron dose monitoring is from a March 26, 1946 letter from J. 

Russell Hayes to Dr. B.S. Wolf (Hayes 1946). Hayes stated that neutron monitoring experiments had 

been carried out during the two previous months but recommended that Hayes’ lab continue to read 

the Dayton neutron monitoring film. Writing on August 24, 1948, John E. Bradley stated that the 

tolerance level for neutrons was 0.1 R/8 hr. He states that 150 n/cm2

/8-hr was considered to equal 

0.1 R, which is lower than the 200 n/cm2

/8-hr in effect until the year before (Bradley 1948). However, 

no neutron dosimetry records have as yet been found for periods before September 1949, so neutron 

dose before that date is considered to be unmonitored. (Neutron dose monitoring began on August 

17, 1949, but was not reported until the Monthly Information Health Report of September 1949.) For 

the period from 1943 through August 1949, dose reconstructors should assign a neutron dose of 50 

mrem/wk. This estimate is based on the modal value of the historic dose distribution recorded after 

September 1949 (see Table 6-5). 

Neutron Dose

The analysis for neutron dose incorporated certain assumptions. First, the analysis assumed a 

40-hr wk in a radiation environment. If the workweek was longer, the dose reconstructor should 

adjust the neutron doses appropriately. Second, the analysis assumed that worker posture during 

exposure was AP, so the recorded exposure would result in the highest dose for a specific workweek. 

After September 1949, when both photon and neutron doses were recorded, the basic dose is the 

dose of record. Additions should be from missed dose, lost or damaged dosimeters or records, and 

other such unrecorded or flawed results.

There is a discrepancy in the records about when Mound began to manufacture neutron sources. 

Meyer (1994) states that the Mound Laboratory was commissioned in August 1948 and that polonium 

operations did not actually begin at the T Building until February 21, 1949. Meyer further states that 

monitoring for neutron radiation did not begin until August 1949 when the fabrication of PoBe sources 

began (Meyer 1994, Volume I, pp. 2, 28, and 84). However, a progress report covering October 1 to 

31, 1948, states, “Three sources (WS-9, WS-10, and WS-11) have been prepared and shipped on 

October 25, 1948 to Unit IV” (Richmond 1948). The neutron emissions of sources WS-9 and WS-10 

were counted on October 15, 1948. The emissions of all three sources were counted on October 20 

and 25, 1948. These records indicate that Mound Laboratory was manufacturing PoBe sources at 

least as early as October 15, 1948. 

Richmond (1948, pp. 5–6) indicates that Mound shipped a 210PoBe neutron source to Brookhaven 

National Laboratory on October 25, 1948. The opening paragraph of Richmond (1948, p. 4) 

references a report (John H. Birden, Work Done in Preparing Neutron Sources from Polonium and 

Beryllium between September 5, 1945 and May 25, 1947, NIM-187) and implies that PoBe sources 

were manufactured at one or more of the Mound precursor sites between September 5, 1945, and 

May 25, 1947. Whenever the manufacture of the first PoBe source occurred, Meyer (1994) is clear 

that monitoring for neutron dose began in August 1949. However, there are no records of the August

1949 neutron monitoring doses. The first report of neutron doses did not occur until the September 

1949 Monthly Health Information Report, dated October 18, 1949.
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The significance of the discrepancy for dose reconstructors is that they must estimate neutron doses 

incurred before September 1949. There are records neither for individuals nor for worker cohorts, and 

Table 6-5. Dose distribution from 1948 to 1959.

a

Year

Photon dose, mrem/exchange Neutron dose, mrem 

Data completenessb

0-50

(0-100)

50-150

(>100) 150-300

300-600

(>500, >300) 0-100 100-300 >300

1948 Note: 1948 is extremity dose, not whole-body dose.

Workers 95.39% 3.08% NA 0.85% 11 mo

1949 Note: Whole-body dose measured after January 1949; neutron data began 9/49. 12 mo

Workers 98.06%

Visitors 100% (0-120) (120-300)

Neutrons Note: increments 0-0.4 MPL = 0-120 mrem, etc. 99.23% 0.55% 0.22%

1950 12 mo

Workers 99.53%

Visitors 99.78% (0-120) (120-300)

Neutrons Note: Increments 0-0.4 MPL = 0-120 mrem, etc. 99.23% 0.61% 0.17% 1.6X max

1951 (0-120) (120-300) <thru 8/51 12 mo

Workers 99.01% 0.72% 0.19% 0.08%

Visitors 99.72% 0.13% 0.01% 0.13%

Neutrons Note: Increments changed to 10-100 mrem 9/51 99.22% 0.78% 0%

1952 Biweekly badge changeouts began 12/51. 11.5 mo

Workers* 94.23% 2.28% 1.98% 0.15% *13 received >600 in 1 mo

Visitors 99.86% 0.14% 0% 0% *Nov low-dose data missing.

Neutrons Note: Increments are 0-100 and >100 mrem 99.60% 0.07%

1953 11.5 mo

Workers 92.80% 6.39% 1.41 % 0.61 % Low-dose data unavailable for 

Visitors 99.26% 0.5 7% 0.12% 0.04% February and June.

Neutrons 99.51% 0.49% 0%

1954 3 mo, 3 qtr

Workers 91.74% 5.65% 1.75% 0.94%

Visitors 98.53% 1.47% 0% 0%

Neutrons 98.45% 1.55% 0%

1955 2.5 qtr

Workers*** 88.81% 6.10% 2.49% 2.04% ***5 received >600 mrem in 1 qtr

Visitors 93.29% 6.71% 0 0

Neutrons 97.55% 2.26% 0.19%

1956 2 qtr

Workers 92.86% 6.84% 0.30% 0%

Visitors 93.27% 6.73% 0% 0%

Neutrons 98.57% 1.43% 0%

1957 3.5 qtr

Workers 91.23% 8.72% 0.05% 0%

Visitors 98.96% 1.04% 0% 0%

Neutrons 93.72% 4.19% 2.09%

1958 4 qtr

Workers 98.20% 1.75% 0% 0%

Visitors 97.90% 1.95% 0% 0%

Neutrons 87.82% 7.44% 4.75%

1959 2 qtr

Workers 94.80% 5.20% 0% 0%

Visitors 97.54% 2.46% 0% 0%

Neutrons 92.86% 5.29% 1.85%

a. Historic neutron doses at Mound were calculated using the quality factor of 7. See Attachment 6A for ICRP 60 (1991) converted doses.

b. Months and/or quarters of available health physics reports per year 

source terms are not available. Therefore, in the absence of measured doses, dose reconstructors 

should apply the historic modal dose of 50 mrem/wk for all employees for periods before 1949.

Between the beginning of neutron dose monitoring in August 1949 and some indeterminate date 

(probably the end of 1969), Mound used three terms to set worker exposure limits: maximum 
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permissible exposure, maximum permissible rate of exposure, and maximum permissible level (MPL) 

of exposure. The value of this measure was set consistently at 300 mrem/40-hr wk (3.75 rem/qtr or 

15 rem/yr) (Meyer 1959; Bigler 1959). Hogerton (1963, Section 7.10) reaffirms this standard.

By November 1970, the maximum allowable external exposure was 3 rem/qt or 12 rem/yr (Author 

unknown 1970, p. 1). A notation for when this standard changed to 100 mrem/wk or 5 rem/yr was not 

identified. The 3-rem/qtr standard was still in use on July 15, 1985 (Daily and Wallace 1985), but it 

was superseded by March 1989. Dose reconstructors should assume that it changed on January 1, 

1989.

Table 6-6 lists the types of neutron dosimeters used at Mound.

Table 6-6. Mound neutron dosimeters. 

Period Dosimeter type

Dosimeter 

holder

Exchange 

frequency Comment

1946–1949 None None None No neutron monitoring

1949–8/1951 Kodak NTA Oak Ridge/steel Weekly All radiation workers

8/1951–1957 Kodak NTA Oak Ridge/steel Biweekly All radiation workers

1957–1960 Kodak NTA Oak Ridge/steel Biweekly All radiation workers

1961–1962 Kodak NTA Oak Ridge/steel Biweekly All radiation workers

1963–1968 Kodak NTA Oak Ridge/steel Weekly Some radiation workers

Biweekly Most radiation workers

Monthly General area workers

1968–1977 Kodak NTA Mound holder Weekly Dependent on work area

Biweekly

Quarterly

1977–1986 Harshaw TLD-600 TLD-700 Cycolac Weekly Dependent on work area

Biweekly

Quarterly

1986–1987 Harshaw TLD-600 TLD-700 Cycolac Monthly Dependent on work area

Quarterly

1987–1991 Harshaw TLD-600 TLD-700 Cycolac Monthly Dependent on work area

Quarterly

1991–2003 Harshaw TLD-600 TLD-700 Cycolac Monthly Dependent on work area

Quarterly

6.3.2

Mound used a film badge external dosimetry monitoring system from 1946 to 1977. From 1977 to the 

present, Mound has used a TLD system.

Dosimeter Types and Badge Distribution

Before use, dosimeters were placed in low-risk areas of men’s and women’s exchange rooms in the 

different buildings. The dosimeters were on wooden racks with name identification with a section for 

the film dosimeter and a section for two pocket ionization chambers. The ionization chambers were 

exchanged daily and read the next morning. The use of pocket ionization chambers stopped in 1966. 

Film badges were collected on Friday and developed on Monday of the following week (see Table 6-4 

for exchange frequency).

Every film packet consisted of two films, one sensitive and one insensitive. If evaluation of the 

sensitive film was not possible, the insensitive film was used to evaluate exposure (Meyer 1994, 

Volume I, p. 121). The batch number, week number, and an “HP” number were marked on each film 

using a 1- or 2-keV X-ray machine.
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The neutron flux, which constituted 300 mrem/40-hr wk, underwent a series of changes over this 

same period. Table 6-7 lists the chronology of how many neutrons per second per centimeter 

squared per 40 hours constitutes 300 mrem of dose. The term millirem came into use at Mound on 

May 4, 1959. This TBD uses this term for convenience in relation to earlier neutron dose quantities.

Table 6-7. Neutron flux values for 300 mrem/40-hr wk dose. 

Date of flux 

standard Flux/300-mrem neutron dose

1946?-

7/1947

200 n0/s/cm2/40 hr = 2.873E+07 neutrons/cm2 = 500

8/1947

mrem/40-hr workweek through 7/1947 

150 n0/s/cm2/40 hr = 2.160E+07 neutrons/cm2 PoBe sources.

10/1951 75 n0/s/cm2/40 hr = 1.080E+07 neutrons/cm2 PoBe and Po-n sources.

1/1955 35 n0/s/cm2/40 hr = 5.040E+06 neutrons/cm2.

1/1956 30 n0/s/cm2/40 hr = 4.320E+06 neutrons/cm2 PoBe, PoCaF, and Po-n sources. 

4/13/1959 55 n0/s/cm2/40 hr = 6.480E+06 neutrons/cm2 PoBe, PoCl, and Po-n sources. 

8/1963 70 n0/s/cm2/40 hr = 1.008E+07 neutrons/cm2 PuF4 sources measured at 0.75-MeV workplace 

average energy. 

7/15/1968 Track fading correction applied to all neutron film read after this date. Some film (1 wk?) 

corrected for 36% fading. Two-wk film corrected for 56% fading. 

1/1/1969 Exposed neutron film for 4 wk for calibration to compensate for effects of track fading. 

Percentage corrections for track fading were discontinued after this date. 

1/12/1969 Effective date on which 55 n0/s/cm2/40 hr = 7.92E+06 neutrons/cm2. 1.3-MeV neutron energy 

of bare PuF4 source.

11/12/1997 TLDs used for dosimetry, assume 100 mrem/wk.

Meyer (1994) claims that each neutron flux value from 150 n0/s/cm2/40 hr to 30 n0/s/cm2/40 hr to 

55 n0

/s/cm2/40 hr equals 300 mrem/wk. Meyer states these changes without explaining why the 

changes occurred. It might have been that the true workweek was longer than 40 hr assumed by the 

National Bureau of Standards (NBS 1957), so the 300-mrem/wk flux was reduced to reflect the longer 

period. The changes might reflect the manufacture of neutron sources that used polonium in 

combination with elements other than beryllium that had lower neutron energies. For example, on 

November 26, 1956, there was an incident during the manufacture of a PoCaF source (Richmond 

1956). All that is known is that these changes occurred. 

According to Meyer (1959), the energy of a neutron from a PoBe source is about 5 MeV. [The 

4.2-MeV value is from Shleien, Slaback, and Birky (1998, Table 7.5).] The neutron energy of a PoF 

neutron is 1.4 MeV, and that of a PoLi source is 0.42 MeV (Shleien, Slaback, and Birky 1998, Table 

7.2). The neutron energy from a PoCaF source is lower than the energy from a PoBe source, 

probably close to the 1.4-MeV value for PoF. 

All other values in Meyer (1994) through the beginning of the PuF4 neutron calibrations in December 

1963 have been accepted as stated. The PuF4 calibrations for 1.3-MeV neutrons have been 

normalized to 55 n0

/s/cm2/40 hr being equal to 300 mrem/wk. 

It is difficult to document the neutron energy assumptions that might have driven the early standard 

values set forth above. Meyer (1959, footnote 2) stated that the energy of PoBe neutrons was about 

5 MeV and their relative biological effectiveness (RBE) was 7, for which he cited four references to 

NBS Handbook 63 (NBS 1957; see Figure 6-1). However, differing neutron energy assumptions 

drove the values for 1963 and following years. Quality factor and weighting factor are more recent 

terms than RBE for the same concept. Table 6-8 lists contemporary quality factors, which neutron 

dose reconstructors should consider. Table 6-9 lists energy types and groups for the Dayton Project 

and Mound Laboratory.
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Figure 6-1. Neutron flux producing 300 mrem/40-hr 

exposure rate versus neutron energy (NBS 1957). 

Table 6-8. Radiation weighting factors for neutrons. 

Energy

Historic mean 

quality factor

Mound quality 

factor

ICRP weighting 

factor

<10 keV 2,10 7 5

10 keV – <100 keV 10 7 10

100 keV – <2 MeV 10 7 20

2 MeV – <20 MeV 10 7 10

≥ 20 MeV 10 7 5

For example, in August 1963 health physicist John B. Kahle measured neutron levels in the SM 

Building (from a PuF4 source) using a DePangher-type instrument he had assembled. He found the 

average neutron energy in SM Building to be 0.75 MeV after accounting for energy lost to the 

laboratory environment. Bigler (1963) cited NBS Handbook 63 (NBS 1957; see Figure 6-1), which 

indicated that a 0.75-MeV neutron flux of 70 n0

/s/cm2 would result in a dose rate of 300 mrem/40 hr. 

This is the first citation in Mound data of a reference for the conversion of a neutron flux to a dose. 

The corresponding number of neutrons equal to 300 mrem/40 hr is 1.008 × 107 neutrons/cm2 or a flux 

of 70 n0

/s/cm2/40 hr (Bigler 1963; Meyer 1994, 1963 Appendix I, Neutron Film Calibration, pp. I, 255). 

In August 1969 (effective December 1, 1969), the unmoderated neutrons from a PuF4 source were 

measured with an energy of 1.3 MeV. For 1.3 MeV, the number of neutrons equivalent to 300 mrem 

was reduced to 7.92 × 106 neutrons/cm2 or a flux of 55 n0

/s/cm2/40 hr. Even though the average 

energy of the dosing neutrons might actually have been closer to 0.75 MeV than to 1.3 MeV, taking 

into account collisions that would occur before the neutrons hit a person, Mound dosimetrists thought 

it conservative to assume the maximum neutron energy and reduce the number of neutrons 

equivalent to 300 mrem.

The original (August 1949) neutron source for calibration of films was a PoBe source. Therefore, the 

energy of the neutrons used for calibration and the energy of the neutrons to which workers were 

exposed were the same. It is unlikely that lowering of the neutron flux per 300 mrem in 1951 reflected
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Table 6-9. Radiation energy intervals and ICRP 60a correction factors by facility and process. 

Facility Process type Period

Radiation 

type

Energy 

interval Percentage

ICRP 60

correction 

factor

Unit 3 Made PoBe and Po-n sources Oct. 1943–1948 Photon 30–250 keV 25

>250 75

Neutron 100–2,000 keV 25 0.72b

2–20 MeV 75 1.1

Unit 4 Calibrated PoBe sources Feb, 1944–late 

1948 

Photon 30–250 keV 25

>250 75

Neutron 100–2,000 keV 25 0.72

2–20 MeV 75 1.1

Mound Facility August 1949–

2000 

Buildings T, HH, WD, R, B Po operations 

PoBe sources 

Feb/May 1949–

1963 

HH Building Beta >15 keV 100

HH 3, 4, 5, 8 Separation of He-3; Kr-85 1960–1995 Photon 30–250 keV 50

HH 5, 6 Pa-231/sludge separation 1955–1956 >250 keV 50

HH 8,10-12 Hydrolysis of Al and Bi 

chlorides

1949–1958

PP Building Plutonium processing 12/1967–1995 Beta >15 keV 100

Waste Handling Area Th-232 storage & redrumming 1955–1975 Photon 30–250 keV 75

>250 keV 25

Neutron 100–2,000 keV 75 2.2

2–20 MeV 25 0.36

R Building R&D for Po program 1948–1951 Beta >15 keV 100

R 117 Storage vault 1949–1990 Photon 30–250 keV 50

>250 keV 50

Neutron 100–2,000 keV 50 1.43

2–20 MeV 50 0.72

SW Building Beta >15 keV 100

SW 8 Separation of xenon isotopes 1957–1960 Photon 30–250 keV 50

SW 11-14 & 16 (area 

1B) 

Decon of BiSO4

Pa-231 purification

1951–1953 

1956

>250 keV 50

Neutron 100–2,000 keV 50 1.43

SW 19 (Old Cave) Radium-actinium separation 1951–1953 2–20 MeV 50 0.72

SW 140 New Cave, 

SW-132, 136, 137 

Processing of residue from 

pitchblende processing 

1970–1979 

SW 146-147 Metrology laboratory 1967–1985 

SW 219 Neutron source production 1962–1965 

T Building Beta >15 keV 100

All but T43,44,48, 

50-51, 57-59, & 348-359 

Po neutron source program 1949–1973 Photon 30–250 keV 50

>250 keV 50

Neutron 100–2,000 keV 25 0.72

2–20 MeV 75 1.1

H Building Hot laundry Photon 30–250 keV 25

>250 keV 75

WD Building Waste handling Beta >15 keV 100

WD 1, 8, 101,104 Wastewater treatment 1949–1990 Photon 30–250 keV 50

WD 107 Ultra filtration 1976–1981 >250 keV 50

WD 112, 113 Analytical work 1966–1979 

WD 113, 118 A&B Storage, waste incineration 1979–1994 

WD Penthouse Filter banks 1949–1993 

B Building Internal dose research 1950–1955 Photon 30–250 keV 25

>250 keV 75

a. ICRP (1991).

b. Sample calculation: Let fraction df = 0.25 for neutron energy between 0.1 and 2.0 MeV. ICRP weighting factor wf60 = 20 sieverts/gray. Mound historic 

weighting factor mwf = 7 sieverts/gray. Correction factor = (df × wf60)/mwf = (0.25 × 20)/7 = 0.72. 

a change in the neutron energy to which workers were exposed, because PoBe sources were still 

being made. 

The PoBe source was last used for calibration before August 9, 1965, for T and SW Buildings (Meyer 

1994, Volume I, p. 308). After December 1 and December 19, 1963, respectively, a plutonium 

fluoride source was used for calibration of film badges worn in the R and SM Buildings (Meyer 1994, 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 19 of 51

Volume I, p. 274). This analysis assumed that this was because polonium sources were no longer 

processed in those buildings (Meyer 1994, 1963 Appendix VI, Volume 1, p. 282). However, there was 

a significant polonium inhalation incident on July 22, 1968 (Storey 1969). Although no new polonium 

neutron sources were made after early 1963, a 1969 summary report indicates that research and 

development work was provided to the AEC. Among other things this work consisted of “... the 

isotopic separation and purification and sales of highly purified stable noble gases, carbon-13, 

uranium-234, and polonium-210” (Author unknown 1969, p. 7). However, lowering the 300-mrem/wk 

neutron flux in years after 1959 probably represents the manufacture of lower energy neutron 

sources.

Plutonium operations began in the SM Building early in 1962. On March 8, 1963, a PuF4 source was 

used for a study of neutron calibration. This was the first neutron calibration using a plutonium 

source. The effective date of this calibration was retroactive to January 1, 1963. On March 12, 1963, 

a weekly report noted that, for an equivalent exposure, 2.35 times as many tracks were observed from 

a PoBe source as from a PuF4 source (Meyer 1994, 1963 Appendix II, Volume 1, p. 258). This meant 

that dosimetrists had been underestimating neutron exposures received by personnel in the SM 

Building. Bigler reported that underestimates would go back to January 1, 1963, which they did. 

Meyer (1994, Volume 1, pp. 257–258) believed that underestimates should go back to the beginning 

of plutonium work in SM Building, or early 1962. On August 9, 1965, Bigler wrote, “As of August 9, 

1965, all neutron exposures received in the plant were to be estimated on the basis of PuF (sic) 

neutron calibrations. Before this time we have used PoBe calibrations for exposure estimates in the 

‘T’ and ‘SW’ Buildings” (Meyer 1994, Volume 1, p. 308).

On December 2, 1969, the value of 55 n0

/s/cm2/40 hr or 7.920 × 106 neutrons/cm2/40-hr wk was 

chosen as the Mound standard for a 300-mrem/wk neutron dose. In promulgating this standard 

Sheehan (1969a) commented, “Although the value is arbitrary, we now feel that it is more appropriate 

to use the value for 1.3-MeV neutrons since this energy is closer to that emitted from PuO2, which is 

the major source of our neutron emissions in the radiation areas.” This is the first available record 

that describes the rationale for using a particular value of neutron flux as a dosimetry standard. The 

arbitrary nature of this value is confirmed in Phillabaum (1973), which reported that the average 

neutron energy at Mound was less than 1 MeV based on experience at a Minnesota dosimetry school.

An outline of a publication about a Mound dosimeter provided perspective on neutron energies at 

Mound: “The neutron spectra in working areas at Mound Facility vary from heavily moderated 

neutrons averaging 0.5-MeV in production hoods to >2-MeV neutrons from transferring bare sources. 

NTA [neutron track emulsion, type A] film was not adequate to measure the lower-energy neutrons” 

(Crain 1978).

Table 6-10 lists the components of neutron TLDs at Mound.

Table 6-10. Neutron TLD card components. 

Period Back filter TLD Chips Front filters

1977–1991 Cd None TLD-600 .Absent Sn None

Sn Sn TLD-700 TLD-600 Sn Cd

1991–present Cd/ABS ABS TLD-600 Absent Sn/ABS ABS

Sn/ABS Sn/ABS TLD-700 TLD-600 Cd/ABS Sn/ABS

The energy spectrum of the neutrons has a direct bearing on another persistent problem at Mound, 

that of track fading in the NTA film. Table 6-11 lists track fading versus neutron energy. While the 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 20 of 51

relationship is not linear, it is clear that tracks from higher energy neutrons fade less than tracks from 

lower energy ones.

Table 6-11. Neutron track fading in NTA film versus neutron 

energy. 

Source Neutron energy Percent fading/time

PoBe 4.2 MeV 9%/wk (?) 

Moderated 

238PuO2 

≈1.3 MeV 16%/1 wk 

30%/2 wk 

PuF4 1.3 MeV (bare source) 33%/1 wk

56%/2 wk 

a. Meyer (1994, Volume III, p. 23) indicates that the neutron emission 

energy of this 238PuO2 source and the PuF4 source were about the 

same.

b. Source: Meyer (1994, E. N. Arnett, Volume II, pp. 107–113).

c. Source: Meyer (1994, C. T. Meyer, Volume II, pp. 96–102).

Table 6-12 lists the 40-yr history of neutron track and TLD signal fading at Mound.

Table 6-12. Neutron track and signal fading study and corrective action history. 

Period Dosimeter type Study/corrective action taken

1951 NTA film Meyer (1951) PoBe source track-fading study. No action taken. 

Summer 1967 NTA film Meyer (1968) PuF4 source track fading study. 

July 15, 1967 NTA film Films corrected for track fading: 33% for 1 wk, 56% for 2 wk.

January 13, 1969 NTA film Films calibrated over 4 wk to reduce track fading. 

March 15, 1976 NTA film All neutron exposures of record 1970 through 1976 doubled. 

Nov/Dec 1977 TLD Better low neutron energy response than NTA film. 

July 16, 1981 TLD Annealing begun to eliminate 5%/wk for 6 wk signal fading.

January 1, 1983 TLD Ca-252 used for neutron calibrations.

1991 TLD DOELAP-approved dosimeters. 

The first mention of fading of neutron tracks in NTA film at Mound is in Experiment for Fading of 

Latent Image on Type NTA Film Emulsions (Meyer 1951). The primary purpose of the experiment 

was “to determine the percentage of proton tracks [proton recoil tracks] which will fade out after a 

lapse of varying amounts of time between exposure and developing.” The secondary purpose was “to 

[find] out if increasing the developing time for ‘old’ films will bring out any of the faded tracks.” The 

details of the experiment are set forth carefully, but the available records do not contain the results.

The amount of neutron track fading measured must have been accepted in 1951, or Meyer would 

have made some changes to the calibration procedure to minimize it. The important point is the 

recognition of the issue of neutron track fading as early as September 1951. Nothing was done to 

address the issue until July 15, 1968, which turned out to be significant.

Meyer (1994, 1967 Appendix VII, p. 82, through 1967 Appendix X, p. 119) discusses neutron track 

fading studies for NTA film in detail, including studies of PoBe, PuF4, and 238PuO2 sources. A 

conclusion noted that, for exposure to plutonium-fluoride neutrons, an evaluation error of 

approximately 9% results for each week the worker wore the film due to latent image fading (Meyer 

1994, Volume II, pp. 67, 104).

On July 15, 1968, management directed correction for track fading of results of all neutron film 

processed on and after that date (Meyer 1994, Volume II, p. 119). This included correction of 1-wk 

film badges for 33% fading and 2-wk badges for 56% fading based on C. T. Meyer (1968). Beginning 

on January 13, 1969, neutron calibration films were exposed progressively over 4 wk to match the 

times that different groups of workers wore their badges. This compensated for track fading.
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During late 1969 and early 1970, there were experiments on signal fading with TLD-200 phosphors 

(CaF2:DY) exposed to 137Cs gamma photons. By February 1970, it had become clear that TLD-200 

was not suitable for routine long-term photon monitoring because of signal fading (Meyer 1994, 

“TLD-200 Signal Fading Studies,” Volume III, pp. 86–91). Thereafter, only TLD-600 (6

LiF:Mg Ti for 

neutron detection) and TLD-700 phosphors (7

LiF:Mg Ti for photon detection) were investigated for 

use. Lithium-6 is sensitive to neutrons via the alpha,n reaction. Lithium-7 is sensitive only to thermal 

and epithermal neutrons but not to higher energy neutrons. Lithium-7 neutron sensitivity was not an 

issue at Mound because there were no thermal or epithermal neutron sources. The important 

conclusion is that the neutron energy spectrum must be known to calibrate dosimeters or assign dose 

accurately.

Figure 6-2 shows the cross-section versus neutron energy for several common neutron-detecting 

processes. The neutron cross-section of 6

Li is linear over almost 6 orders of magnitude of change in 

neutron energy. However, it has a very pronounced resonance peak at about 200 keV and another 

smaller peak at about 2,500 keV. In other words, the most nonlinear portion of the 6

Li neutron 

response lies in those portions of the neutron energy spectrum that were important at Mound after 

1962 when plutonium sources were manufactured. 

Figure 6-2. Cross-section versus neutron energy for several common 

neutron-detection processes (Price 1964, p. 312).

In January 1973, Thermoluminescent Dosimeter (TLD) Fade Study of 7

LiF Phosphors (Bigler 1973), a 

study of gamma photons and TLD-700 phosphors, indicated that track fading amounted to only about 

6%/qtr, which was insignificant (Meyer 1994, Volume III, pp. 200–201). Gamma signal fading on 

TLD-700 phosphors might have been insignificant but neutron track fading on NTA film was not. On 

March 15, 1976, Mound management concluded that neutron exposures for 1970 through 1976 were 

low by a factor of 2 and required adjustment to include a doubling of recorded neutron dose 

assessments for those years (Meyer 1994, Volume IV, p. 170). 

Documentation of Variables in the Assessment of Neutron Doses at Mound Laboratory (Bigler 1978) 

is an overview of the history of neutron dosimetry at Mound and as such is invaluable in fitting the 
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documented pieces into the whole picture. One of the many important comments is that Mound 

began working with 238Pu in 1959 and began manufacturing RTGs in 1960. The PoBe sources 

manufactured previously had average energies of about 4.5 MeV, whereas the PuO2 RTG sources 

had an average energy of about 1.0 MeV. This work occurred in either the R or SM Building until the 

PP Building became operational in 1970. RTG manufacture in the PP Building in 1970 allowed the 

stripping of 16O and 17O from 18O before the manufacture of PuO2 RTG sources, which lowered the 

neutron dose from the RTG sources significantly (Meyer 1994, Volume IV, pp. 169–185). 

From 1970 to 1977, workers wore a TLD photon dosimeter and a separate NTA film neutron 

dosimeter. November and December 1977 saw the introduction of the new all-TLD dosimeter. 

Because the TLDs responded well to the lower energy neutrons that characterized the Mound working 

environment, it seemed that doses were finally being measured accurately. The signal fading 

problem now affected both neutron and photon dose measurements. Specifically, there was an 

observed 25% to 30% stored signal fading on both photon and neutron (TLD-600) chips. On June 3, 

1981, fading was about 5%/wk for 6 wk (Meyer 1994, Volume V, pp. 310–320), after which it stopped, 

which meant that some people would get slightly larger doses of record after the TLDs were annealed 

to stop the signal fading. In addition, Meyer stated that the workers receiving the highest neutron 

doses were in decontamination and decommissioning in Building PP and that their neutron doses 

were about 30 mrem/2 wk. This means that they would receive an extra 30 mrem over the span of 

the second half of 1981, which is not significant. 

The penultimate change in neutron dosimetry at Mound was on July 16, 1981. At that time, 

management decided to anneal the TLDs for 6 hr at 80 ̊C to eliminate signal fading (Meyer 1994, 

Volume V, p. 344). As of January 1, 1983, 252Cf was used as the neutron calibration source (Meyer 

1994, Technical Manual MD-21766, Issue 1, Volume IV, p. 43). In 1991, Mound adopted an all-TLD 

hard pack DOELAP-approved dosimeter.

Photon Dosimetry

This analysis assumed that workers spent 40 hr/wk in a radiation environment and that their posture 

was AP, so they received the maximum positional dose for any specific workweek. Dose 

reconstructors should reconstruct unrecorded dose from coworker doses, area surveys, source terms, 

or historic summary data. Corrections to the dose of record would be additions to the dose of record.

Table 6-13 lists beta photon dosimeter characteristics. Table 6-5 lists historic dose distribution. Dose 

reconstructors should use the modal values of this dose distribution to assign unmonitored photon 

and neutron doses.

Incomplete individual dosimetry has been found from Febuary through June, 1944, but none after that 

until May, 1946. In the 1946 dosimetry, worker names were numerically coded in the record and the 

code key has been lost. As a consequence, individual dose records before January 17, 1947, when 

exposures were again logged by worker name, are not recoverable. No worker cohort doses, records 

of area monitoring, or source terms are available for this period. Therefore, based on data from 1948 

to 1959, this analysis assigned a modal dose of 50 mrem/wk (see Table 6-14).

Recovery of 227Ac from neutron-irradiated 226Ra began in SW Building in the early 1950s. In the range 

of 500 mR, the film overresponded to the actinium gamma by a factor of 1.7. However, because the 

gammas from 226Ra, 227Ac and 228Th could not be separated on the dosimeter film, the radium 

calibration graph was used and all increased film density was treated as radium equivalent. This was 

a conservative dosimetric assumption. By 1952, the minimum detection for film badges was 34 mrem 

(Author unknown 1952), perhaps because of the improved film and the low dose calibration (Meyer 
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1994, Volume I, p. 28). The low-energy window was not read and its dose was not recorded until 

January 1960 (Meyer 1994, 1960 Appendix III, Volume I, p. 250). After that, the window was read 

Table 6-13. Beta-photon dosimeter characteristics. 

Period Packet type

Film 

component

Exposure 

range(R)

Filter 

type Comment

1944-1946? Brass, coin envelopes Dental Unknown Brass; 

none

Too many uncertainties to 

quantify 

1946–1948 Dental size film packet Unknown Unknown Unknown Possibly Kodak Type K film

1949–1957 DuPont D552 502 0.02-10 Cd 1-mm Cd filter, one on front 

510 5-800 of film packet

1957–1960 DuPont D/Unknown 

model

508 0.03-5 Cd 1-mm Cd filter, one on front 

510 5-50 of film packet

1961–1962 DuPont D558 502 0.02-10 Cd 1-mm Cd filter, one on front 

1290 20-3000 of film packet

1963–1968 DuPont D556 508 0.03-5 Cd 1-mm Cd filter, one on front 

834 5-1000 of film packet

1968–1969 DuPont D556 508 0.03-5 Cd --

Te

834 5-1000 ABS

Ta

July 1970–

1977

Kodak Type 3 Single-coated 0.030-10 Cd --

Te

Double-coated 10-800 ABS

Ta

Table 6-14. Unmonitored photon dose at Mound. 

Period Dose reconstruction action

Any facility before January 17, 1947 Assign photon dose of 50 mrem/wk, based on modal dose from Table 6-5. 

January 17, 1947, through June 8, 1963 Use average of individual known doses or worker cohort data or modal dose. 

June 8, 1963 onward Use average of individual known doses, worker cohort data, or area 

monitoring data in R, SM, SW, and T Buildings. 

and its dose recorded, so the doses in the records were correct. Table 6-5 lists dose distribution at 

Mound from 1948 to 1959. 

Based on whole-body exposure data from 1965 through July 1969 (Sheehan 1969b), dose 

reconstructors should retroactively assign maintenance staff 38 mrem/2-wk badge exchange period or 

1 rem/yr for the unmonitored period before March 22, 1966 (Meyer 1994, Volume I, 1966 Appendix II, 

pp. 320-325). 

Area monitoring began in the T, R, SM, and SW Buildings between May 25 and June 8, 1963 (Meyer 

1994, Volume I, pp. 264-270). If a worker was in these buildings after these dates, dose 

reconstructors should substitute area-monitoring results for missing doses of record. 

Mound introduced new dosimetry on September 1, 1968. Meyer (1994, Volume II, pp. 104–133) 

asserted that the new dosimeter represented a technical advance in that it incorporated filters to 

differentiate among the various radiation fields present in the work environment. There are no 

specifics about how different types of radiation were measured for the record or what the radiation 

standard was at that time. The training program for November 1970 stated that the radiation dose 

limits were 3 rem/qtr and 12 rem/yr, with a cumulative dose not to exceed 5 (age 18 yr) rem.
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At some point after July 1985, dose standards came into conformance with the International 

Commission on Radiological Protection recommendation of a maximum of 5 rem/yr (ICRP 1977). The 

effective date of this change is uncertain, but it was after July 15, 1985, and before March 1989. DOE 

issued Order 5480.11, “Radiation Protection for Occupational Workers” (updated as DOE 1992) to 

implement the recommendations in 1987. This analysis assumed that use of the 5-rem/yr standard 

began at Mound on January 1, 1989.

6.3.3

Dosimeter calibration performed at Mound was in-air calibration. Mound did not start to use phantom 

calibration until the late 1970s. Since the Mound dosimeter database contains only annual summary 

doses for the years through the end of 1977, dose reconstructors should use the Exposure to Organ 

Dose DCFs for dose conversions through 1977. Photon calibration was performed by exposing a set 

of films for the same period but at different distances from the source. Neutron calibrations were 

performed for the same distance but for different times. This section explains calibration from the 

start of the Dayton Project in 1946 until the present.

Energy Spectra

6.3.3.1 Beta/Photon Calibration

A photon calibration strip consisted of eight films. One film was left unexposed and seven were 

exposed to 226Ra and its decay products at equilibrium to identify a calibration curve of net optical 

density and exposure. The calibration was performed in an 26Al calibration well with a diameter of 

65 cm. Seven film holders were affixed to the calibration well. Film packets were inserted in steel 

holders distributed about the well at different distances from and angle to the source and exposed for 

7 hr. A calibration strip was created with every batch of film processed. The exposure was performed 

on Friday of every week or the last day of the exchange period. No beta calibration was performed 

before 1979.

6.3.3.2 Neutron Energy Calibration

Fast neutron calibration was based on exposing a set of five NTA films at a fixed distance from the 

source but at different exposure times. Calculation of the neutron fluence rate was based on a source 

strength that yielded 300 mrem/40-hr wk. A set of 10 fields of vision for each film was counted and 

averaged to derive the number of tracks per field. The number of tracks was plotted against 

exposure, and the slope of the line was used to calculate the exposure by multiplying the number of 

tracks per field times the slope of the line.

6.3.3.3 Neutron Energy Spectra

Neutron energy spectra depend on the neutron source. Mound manufactured 210PoBe neutron 

sources and measured the neutron energy at about 5 MeV. Mound calculations assumed this energy 

and an RBE of 7. In addition, Mound experimented with other polonium compounds for creating 

neutrons (for example, a PoCaF source in 1956; Richmond 1956). According to the Handbook of 

Health Physics and Radiological Health, 3rd Edition (Shleien, Slaback, and Birky 1998, Table 7.2), the 

energy of a PoF source is 1.4 MeV. In the late 1950s, Mound was making PoCl sources (Incident 

Report 22 1961).

When Mound switched from polonium to plutonium neutron source manufacturing, neutron energies 

dropped. This was reflected in the lowering of the neutron flux, which constituted 300 mrem/40-hr wk. 

In 1963, the average energy to which workers on the PuF4 manufacturing line were exposed was 

0.9 MeV (Meyer 1994, Volume VIII, p. 107). In December 1969, management decided that the proper 
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average neutron energy to use for dosimetry of PuF4 sources was 1.3 MeV (Meyer 1994, 1969 

Appendix VIII, Volume III, pp. 9-10). The calculations and energy spectrum for a 238PuF4 source are in 

Meyer (1994, Volume VIII, pp. 71–72). On February 18, 1974, a health physics technician reported a 

neutron energy survey in R and PP Buildings. The average neutron energy was 1.01 MeV (Meyer 

1994, Volume IV, p. 888). On March 6, 1989, Fix (1989, p. 2) recommended the use of a bare 252Cf 

spontaneous fission source for neutron dose calibrations.

6.3.3.4 Photon Energy Spectra

That photographic film overestimates the dose from low-energy photons (50-175 keV) has been 

known for decades. A Manhattan Project report entitled Photographic Film as a Pocket Radiation 

Dosimeter (Perdue, Goldstein, and Wollan 1944) contains a graph of the response of DuPont D502 

film to photons in the energy range of about 50 to 1,000 keV. The graph shows an overresponse of 

about 2 at 50 keV decreasing to 1 at about 175 keV. A recent study by the International Agency for 

Research on Cancer (IARC; Thierry-Chef et. al. 2002) determined that the overestimation was by a 

factor of 3 at 118 keV and 1.2 at 208 keV. 

There is a family of photon energy-versus-dosimeter response curves from May 25, 1977. A 

memorandum and accompanying graph show the photon energy and responses of the TLD photon 

dosimeter (Anderson 1977). Anderson stated that the tantalum shield in the TLD caused an 

underestimate of the 60-keV photon from 241Am, which he noted was a dose contributor in Building 

PP. In addition, he noted that some low-energy photon dose from plutonium would be 

underestimated. 

6.3.3.5 Dose Calculation for Film Badges

Photon dose for the two-element dosimeter was calculated as high- and low-gamma dose. The dose 

under the cadmium filter was the high-gamma dose. The low-gamma dose was calculated by 

subtracting the high dose from the open-window dose. No beta dose was calculated before 1979. 

The following have been used to calculate whole-body and skin doses in the periods specified:

• 1946-1952: High-energy gamma dose was the slope of the line of calibrated photon film times 

the reading of the field film under the cadmium shield. Low-energy gamma dose was not 

calculated. 

• 1953-1959: High-energy gamma dose was the slope of the line of calibrated photon film times 

the reading of the field film under the cadmium shield. 

Low-energy gamma dose was the slope of the line of calibrated film under the open window 

times the reading of the field film under the open window. 

• 1960-1967: Whole-body gamma dose was the slope of the line of calibrated photon film times 

the reading of the field film. 

Skin gamma dose was the slope of the line of the calibrated film under the open window times 

the reading of the field under the open window. 

• 1968-1977: Use of a multielement film badge began during this period. Whole-body gamma 

dose was calculated as follows:
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Whole-body gamma dose was the slope of line under the tantalum filter of calibrated film times 

the reading of the field film under the tantalum filter.

Skin gamma dose was the open window reading times the calibration factor. 

Neutron doses were calculated as follows:

• 1949-1977: The neutron dose was calculated by the product of the slope of the tracks per 

millirem line from calibrated neutron film and the number of tracks per field of the dosimeter 

film.

Skin dose was calculated by adding skin gamma dose and neutron dose. 

6.3.3.6 Dose Calculation for Thermoluminescent Dosimeters

Table 6-15 lists the history of TLD card components. 

Table 6-15. Beta/photon TLD card components. 

Period Back filter TLD chips Front filters

1977–1991 Brass None TLD-700 TLD-700 Brass Open 

window

None None Absent Absent None None

1991–2006 Cd/ABS ABS TLD-700 TLD-700 Cu/ABS Teflon/ABS

ABS ABS TLD-700 TLD-700 Mylar ABS

Three photon doses were calculated using TLDs. These were deep dose, shallow dose, and dose to 

the lens of the eye. Beta dose was calculated in certain parts of the laboratory. The dose evaluations 

were performed as follows:

• 1977 – 1991: The whole-body gamma dose or penetrating dose was calculated using the 

following equation:

Whole-body gamma dose (mrem) = PG7-BGR

CF

where

PG7 = the penetrating gamma dose measured under the brass filter; which is the value 

of the TLD-700 chip in position 1 on the gamma card

BGR = the background of the gamma cards

CF = the correction factor of the TLD reader

The skin dose is calculated as follows: 

Skin dose = Skin gamma dose + neutron dose + beta dose (if measured).

The skin gamma dose due to low-energy gamma is calculated as follows:

Skin gamma dose = OW7 BGR

CF

1.20x ( - )
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where

OW7 = the value of TLD-700 chip in position 2 on the gamma card

1.20 = the correction factor for overresponse of the chip at low energies. 

If the calculated skin dose is less than whole-body gamma dose, skin dose is recorded as 

equal to whole-body gamma dose.

To calculate neutron dose the following equation is used:

Neutron dose = 0.4 × (FS6 – BGR) × R .1 87

where 

R = ( - ) - ( - )

( - ) - ( - )

FS6 BGR NG7 BGR

BS6 BGR NG7 BGR

and

BS6 = TLD -600 in position 1 (Cd shielded on the back of neutron card close to worker’s 

body)

NG7 = TLD 700 in position 4 in the neutron card (Sn shielded on both sides)

FS6 = TLD-600 in position 3 in the neutron card (Cd shielded on the front)

• 1991-2006: Calculation of the whole-body gamma dose and skin dose uses the same 

equations used for the period from 1977 to 1991. The neutron dose is calculated by the 

NDOSE computer program, which implements the following calculation (FS6, BS6, and NG7 

are as before):

Neutron dose = N × 0.4 × (FS6-NG7) × ( ( - )

(- )

FS6 NG7

BS6 NG7 ) .1 87

where

N = 0.739

If (FS6-NG7) is less than zero or (BS6 - NG7) is less than zero, the neutron dose is equal to 

zero. 

If ( - )

(- )

FS6 NG7

BS6 NG7 is less than 0.5, then ( - )(- )

FS6 NG7

BS6 NG7

= 0.5. 

If ( - )

(- )

FS6 NG7

BS6 NG7 is more than 2.3, then \ = 2.3. 
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6.4 MISSED OR UNDERREPORTED PHOTON OR BETA DOSE 

Table 6-16 lists missed photon doses at Mound. There are four possible sources of missed or 

underreported photon or beta dose:

1. Unread beta or low-energy photon doses. The narrative is not certain, but it seems that the 

clear portion of the film badges, which would record beta and low-energy photon dose, was 

not read until at least January 1953 (Meyer 1994, Volume I, p. 90). For example, in a 

memorandum dated September 1, 1948, Meyer describes the film badge system he had gone 

to Oak Ridge to study. His report states in part, “Film exposed to gamma rays are developed 

and read on a densitometer” (Meyer 1994, Volume I, 1948 Appendix III, p. 33). There is no 

mention of exposure to anything other than gamma rays. On the other hand, the next 

document (Meyer 1994, Volume I, 1948 Appendix IV, pp. 35-40), by a health physicist from 

Oak Ridge, states that Oak Ridge used uranium metal to calibrate film badges for beta 

exposure (Meyer 1994, Volume I, 1948 Appendix IV, pp. 39). 

Table 6-16. Missed photon dose. 

Type of dosimeter Period of use

Mode of 

missed dose

Annual exchange 

frequency

Mean annual 

missed dose 

(mrem)

Two-element film: D502/ D510 Jan 17, 1947–Nov 1951 MDL 40 52 1,040

Two-element film: D502/ D510 Dec 1951–1957 MDL 34 26 442

Two-element film: D508/ D510 1957–1963 MDL 10 26 130

Two-element film: D508/ D834 1963–Dec 1, 1965 MDL 30 26 390

Two-element film: D508/ D834 Dec 1, 1965–1968 MDL 30 26 390

Two-element film: D508/ D834 1968–Nov 13, 1970 MDL 30 12 180

26 390

Kodak type 3 film only Nov 13, 1970–Apr 21, 1976 MDL 13 12 (R & SW Bldgs.) 78

4 26

TLD with Ta shield Apr 21, 1976–Jul 6, 1981 MDL 20 12 120

Signal fadea 2, 6 4 40

Tantalum shield 

lossb

26 Bldg. PP 260

12 24

4 24

26 NA

26, 12, 4 312

Total 26 572

Total 12 456

Total 4 376

TLD without Ta shield July 6, 1981–1991 MDL 20 Monthly 120

Quarterly 40

Biweekly 260 (Bldg PP)

a. Source: Bigler (1973). 

b. Anderson (1977) contains graphs showing that the dose recorded by the TLD with the tantalum shield in place underestimates 

photons below 500 keV. At 200 keV the underestimation is 35%; at 110 keV it is 90%. The 61-keV 241Am photons would have 

been completely missed. Taylor et al. (1995) identified calibration corrections of 1.119 for Hp(10) for Savannah River Site TLD 

doses before 1985. Because the energy spectrum of photons is unknown, this analysis assumed 11.9% of the permissible 

dose of 100 mrem/wk, 12 mrem/wk, or 24 mrem/ 2 wk for biweekly badge holders in Building PP. For biweekly badge holders, 

the geometric mean dose is 24 mrem ÷ 2 wk × 26 biweekly badges = 312 mrem. For monthly badge holders, the geometric 

mean is 12 mrem/wk × 4.33 wk/mo = 52 mrem/mo ÷ 2 × 12 mo = 312 mrem/yr. 

It is clear that it was possible to use the open window of the film badge system to measure 

beta exposure. It is also clear that when Mound was dealing only with polonium there was no 

beta radiation. Because there were no beta emissions, there was no beta monitoring. Later, a 

few areas of the plant had significant beta fields; these were areas in R and T Buildings that 
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worked with materials containing 90Sr90Y. On December 1, 1965, technicians began to read 

the open window as well as the shielded area, and insensitive film as well as sensitive film. 

The purpose of the additional measurements was to permit better evaluation of low-energy 

gamma photons (Meyer 1994, Volume I, p. 310). This analysis assumes that Mound 

evaluated all gamma doses after this date.

2. The 40-mR (40-mrem) threshold for the film dosimeter response. The 40-mR (40-mrem) 

threshold for the photon film dosimeter response is significant. However, a November 1952 

memorandum on film darkening measurement comparisons between 226Ra and 227Ac uses 

40 mR as the minimum measure for film darkening by 226Ra (Meyer 1952). Film dosimeters 

were exchanged and read on a weekly basis. Therefore, in a 52-wk yr, 2,080 mrem could 

have been missed. Because the distribution is lognormal, its geometric mean is [(52 × 40) ÷ 2] 

1,040 mrem annual missed dose.

When operations began at Mound in 1949, the combination of films could cover a gamma 

radiation range of 20 to 100 R (Meyer 1994, Volume I, p. 90). However, a March 1952 

memorandum identified the lowest calibration mark as 40 mR (Meyer 1952). 

3. The pocket ion chambers were read only to the nearest 5 mrem. This analysis assumed 

that this means the nearest 5 mrem higher or lower than the actual reading. For example, a 

value of 22 mrem would be read as 20 mrem but a value of 23 mrem would be read as 25 

mrem. The greatest dose that could be underreported would be 2 mrem/dosimeter reading or 

10 mrem/wk. If the values were read on the high side of a measurement that was 3 or 4 mrem 

greater than the nearest lower 5-mrem mark, the low and high readings would result in a net of 

zero over time. Therefore, this analysis neglected this source of error.

4. The cadmium or tantalum shield blocked photons below 200 keV from the TLD chip. 

The cadmium or tantalum shield blocked the recording of low-energy photons. The original 

purpose of the cadmium shield was twofold: First, to allow imprinting of identifying codes on 

the monitoring films using a 1- or 2-keV X-ray machine (Meyer 1994, Volume I, p. 90); and 

second, “to avoid any contribution by low energy photons” (Meyer 1994, Volume I, p. 20).

As stated in Section 6.3.3.3, when Mound switched to TLDs, the tantalum shield caused a 

significant underrecording of gamma photons below about 200 keV. Anderson (1977) 

contains graphs of TLD response versus gamma photon energy from 10 to 1,000 keV. The 

graphs make clear that the tantalum shield would have completely attenuated any gamma 

photons below about 110 keV, and any photons below about 200 to 300 keV would be 

significantly unrecorded. In particular, this would include the 60-keV gamma from 241Am in 

area PP. Anderson’s recommendation was to read photon doses exclusively from the 

open-window TLD. P. C. Adams wrote a note on his copy of Anderson (1977) indicating a 

request for management opinion on the recommendation. There is no record of what action, if 

any, management took in response to this recommendation.

6.5 MISSED OR UNDERREPORTED NEUTRON DOSE AT MOUND

There are six possible sources of missed or underreported neutron dose at Mound:

• Track fading in NTA film

• Minimum detectable level (MDL) of 50 mrem for NTA film
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• Underresponse of NTA film to lower energy neutrons, particularly those less than 500 keV, 

including albedo neutrons

• Signal fading in TLD neutron dosimeters

• MDL of 10 mrem for TLDs

• Underresponse of TLDs to high-energy neutrons

Table 6-17 lists specific values for missed or unreported neutron doses. Tables 6-18 and 6-19 list 

MDL values for photon and neutron dosimeters. 

Table 6-17. Missed or underreported neutron dose. 

Dosimeter Period of use

Mode of 

missed 

dosea

Annual badge 

exchange 

frequency

Missed dose per 

badge exchange 

(mrem)

Maximum annual 

missed dose (mrem)

NTA film 9/1949–11/1951 1a Weekly 16.5 858b/yr

2 Weekly 50 2,600c/yr 

3 Weekly 7 364d/yr 

73.5 total/wk 3,822 max total/yr

NTA film 12/1951–1967 1b Biweekly 28 728/yr

2 Biweekly 50 1,300/yr 

3 Biweekly 14 364/yr 

92 biweekly total 2,392 max total/yr

NTA film 1967–1976 1b Biweekly 28 728/yr 

2 Biweekly 50 1,300/yr 

3 Biweekly 14 364/yr 

92 biweekly total 2,392 max total/yr 

1c Monthly 19.5 e 234/yr 

2 Monthly 50 600/yr 

3 Monthly 100 1,200/yr 

169.5 monthly total 2,034 max total/yr 

TLDs 1976–1981 4 Monthly 11 132/yr 

5 Monthly 10 120/yr

6 Monthly Can’t quantify

21 monthly total 252 max total/yr 

4 Quarterly 15 60/yrf

5 Quarterly 10 40/yr

6 Quarterly Can’t quantify 

25 quarterly total 100 max total/yr

TLDs 1981–1991 5 Monthly 10 120 max/yr

5 Quarterly 10 40 max/yr

6 Any Can’t quantify

TLDs 1991–present 5 Quarterly 10 40 max/yr

a. Modes of missed dose: 1a, weekly track fading; 1b, biweekly track fading; 1c, monthly track fading; 2, MDL 50 mrem; 

3, poor low-energy response; 4, signal fading; 5, minimum detection limit 10 mrem; 6, poor high-energy response.

b. 33% track fading per week calculated as 50 mrem/wk × 0.33 = 16.5 mrem/wk × 52 wk/yr = 858 mrem/yr. For 2-wk 

badge exchanges, the fading was 56%/2-wk period (50 mrem/wk × 0.56 × 26 badge changes) = 28 mrem/2 wk or 728 

mrem/yr. 

c. Unrecorded dose due to 50-mrem minimum detection limit calculated as 50 mrem/wk × 52 wk/yr = 2,600 mrem/yr for 

weekly badge exchanges, 50 × 26 = 1,300 mrem/yr for biweekly badge exchanges, and 50 × 12 = 600 mrem/yr for 

monthly badge exchanges. 

d. Poor low-energy response calculated from the Savannah River Site correction factor of 1.14 neutrons in the energy 

range of 0.1 to 2 MeV. 50 mrem/wk × 0.14 = 7 mrem/wk or 364 mrem/yr. 
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e. Calculated as 50 mrem/wk × 52 wk/yr × 4.33 ÷ 12 × 0.09/wk = 19.5 mrem/mo or 234 mrem/yr. 

f. Signal fade calculated as 50 mrem/wk × 0.05/wk × 4.33 wk/mo = 11 mrem/mo or × 6-wk fade/qtr = 15 mrem for the first 

6 wk of each quarter × 4 qtr = 60 mrem/yr. 

6.6 DOSIMETER UNCERTAINTY ANALYSES

6.6.1

This section discusses laboratory uncertainty. Dosimeter response uncertainty has a normal 

distribution. Laboratory uncertainty includes the exposure of film to laboratory sources, the 

developing process, the densitometer reading process, and the calibration values used to calculate 

film exposure to film net optical density. Those uncertainties are calculated and combined using the 

propagation of error and assuming that all variables are independent. Dosimeter uncertainty is a 

Photon Dosimeters

Table 6-18. Photon dosimeter limits of detection. 

Period Dosimeter type LOD (mR)a

1946–1949 Dental Film 40

1949–1957 DuPont D552 34 

1957–1960 DuPont unknown model 10

1961–1962 DuPont 558 20

1963–1968 DuPont D556 30

1968–1969 DuPont D556 30

1970–1977 Kodak Type 3 13

1977–1991 Harshaw TLD-700 20

1991–2006 Harshaw TLD-700 20

a. Photon LODs based on Mound historical record. 

Table 6-19. Neutron dosimeter limits of detection. 

Period Dosimeter type LOD (mrem)a

1949-1957 Kodak NTA 50 

1957-1960 Kodak NTA 50

1961-1962 Kodak NTA 50

1963-1968 Kodak NTA 50

1968-1977 Kodak NTA 50

1977-2006 Harshaw TLD-600 10

a. NTA film LOD value based on Wilson et al. (1990).

function of exposure: As the exposure gets closer to the dosimeter LOD, the uncertainty becomes 

±100%. Table 6-20 lists relative uncertainty at the 95% confidence level for a study performed at the 

Nevada Test Site, which used the same type of film as Mound (NTS 2003). Similar values were 

assumed for films that were different.

An IARC study simulated different exposure geometries and three photon energies (Thierry-Chef et 

al. 2000). Irradiation was performed on a slab phantom and the anthropomorphic Alderson RANDO 

phantom. The amount of air kerma used in this study was 5.0 milligray. Table 6-21 lists the IARC 

comparison between measured dose and Hp(10). The mean ratio reflects the ratio between reported 

dose and the Hp(10) delivered. For the two-element film dosimeter at 118 keV, the measured dose 

overestimates the Hp(10) delivered by a factor of 3.0 for the AP geometry. At higher photon energies 

the measured dose and delivered dose are the same for the AP geometry.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 32 of 51

6.6.2

Calculation of neutron uncertainty is based on the simplified method in External Dose Reconstruction 

Implementation Guidelines (NIOSH 2002, p.28), which also considers laboratory uncertainty. The 

major uncertainty source in neutron film is the reading of tracks per field. At Mound, the number of 

fields and field size changed over time. At the beginning of the neutron-monitoring program in 1949, 

the number of fields read was 10 or 11; later it was 40, then 64. In addition, the area changed from 

1.22 mm2 to 1.88 mm2 (from Meyer 1994, Volume VIII, Neutron Film Calibrations, pp. 5-9, 11-12). 

The relative uncertainty was calculated and tabulated based on the dose range and the number of 

films at each dose level. When information for a year was unavailable, uncertainty was assumed from 

the years for which information was available. Table 6-22 lists the 95% confidence level uncertainty. 

Uncertainty becomes large, almost 100%, as dose level gets closer to the dosimeter LOD. 

Neutron Dosimeters
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Table 6-20. Photon dosimeter relative uncertainty. 

Period Dosimeter type Exposure range (mR) 95% confidence level uncertainty

1946–1948a Dental films 10-30 ±100%

30-1,000 ±28%

Above 1,000 ±10-15%

1949–1956 DuPont D552 10-20 ±100%

20-1,000 ±28%

Above 1,000 ±10-15%

1957–1960 DuPont/unknown model, may be D552 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1961–1962 DuPont D558 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1963–1969 DuPont D556 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1970–1977 Kodak Type 3 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1978–1991 Harshaw TLD-700 10-30 mrem ± 100%

30-100 mrem ±30%

1992–2003 Harshaw TLD-700 10-30 mrem ± 100%

30-100 mrem ±30%

a. Dental film was used until January 1949. 

Table 6-21. IARC testing for three photon qualities performed on US beta/photon dosimeter. 

Geometry Phantom

118 keV 208 keV 662 keV

Mean ratio* SD/mean % Mean ratio* SD/mean % Mean ratio* 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

* Column title altered by C. Wesley Proctor.

Uncertainty in neutron TLD card response depends on several sources, including TLD reader 

sensitivity, variability of LiF crystal responses, energy of the neutron spectrum, interference of photon 

radiation on neutron dose, and level of exposure to be detected by the TLD. 

Reader sensitivity and LiF crystal response can be estimated at about 10%. At low neutron energies 

(around 200 keV), the estimated overall uncertainty due to overresponse is about 120% unless 

corrected for in calibration (see Figure 6-2). At very low exposure levels close to background, the 

overall uncertainty approaches 100%. 

Missed Dose Uncertainty Analysis

Doses reported as zero or less than the LOD can be calculated using the LOD/2 method suggested in 

NIOSH (2002). The missed dose is a function of dosimeter exchange frequency, the dose of record 

reporting format, and whether the number of zeros is known. Table 6-18 lists photon dosimeter LOD 

values. Table 6-19 lists LOD values for neutron dosimeters. The modeling of uncertainties in annual 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 34 of 51

Table 6-22. Neutron dosimeter relative uncertainties. 

Period Dosimeter type Dose range (mrem) 95% confidence level uncertainty

9/1949-1957 Kodak NTA 10-20 ±100%

20-1,000 ±28%

Above 1,000 ±10-15%

1957-1960 Kodak NTA 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1961 -1962 Kodak NTA 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1963-1969 Kodak NTA 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

1970-11/1977 Kodak NTA 10-30 ±100%

30-1,000 ±20%

Above 1,000 ±10-15%

12/1977-1991 Harshaw TLD 10-30 mrem ± 100%

30-100 mrem ±30%

1992-2003 Harshaw TLD 10-30 mrem ± 100%

30-100 mrem ±30%

missed photon dose uses a lognormal distribution with a median dose equal to the product of LOD/2 

and the number of zero measurements. The upper 95% dose is calculated by the product of LOD and 

number of zero measurements. When applying this method, the dose reconstructor must keep in 

mind that the LOD/2 methodology does not apply to neutron doses when they are large in relation to 

photon doses, as is the case at Mound (NIOSH 2002, p. A38).

6.6.3

Two sources of radiation in the facilities are considered:

General Field Characterization

• Process Radiation Field: This source is dependent on the process and time. 

• Cumulative radiation source due to accidents and engineering control failures: This source is 

deposited on walls and floors in the facility where radiation work occurs, is different for each 

facility, and accumulates with time. In accordance with NIOSH dose reconstruction project 

direction, only the AP body geometry is to be considered in assigning dose, so this analysis 

does not address these sources. 

Uniform radiation fields were assumed for this analysis and were considered routine fields. Accidental 

fields and nonuniform fields were not considered. All site facilities were characterized by assumptions 

or from energy measurements performed at the sites (Meyer 1994, Vol. III, 1974 Appendix I, 

pp. 217-222). Energy measurements of neutrons were performed in certain years; a grid of 

measurement would be performed and average neutron energy would be calculated for the entire 

building. Assumptions for other energy groups are based on the type of isotope and possible 

energies emitted per decay. Table 6-23 lists the facilities and the corresponding energy ranges. 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 35 of 51

Table 6-23. Externally hazardous radionuclide matrix a Facility Process type Period Externally hazardous radionuclides

Unit 1 Po separation research 1943–1945 Ra-226, Sc-46, Zn-65, Hg-203, Ag-110m, 

Fe-59, Co-60, Bi-210

Unit 2 Chemical 1943–1945 No radioisotopes 

Unit 3 Po production; made PoBe 

and Po-n sources 

Oct. 1943–1948 Sc-46, Zn-65, Hg-203, Ag-110m, Fe-59, 

Co-60, Bi-210; PoBe sources and other 

Po-n sources 

Unit 4 Po production, calibrated 

PoBe sources 

Feb, 1944–late 1948 Sc-46, Zn-65, Hg-203, Ag-110m, Fe-59, 

Co-60, Bi-210; PoBe sources 

Warehouse Bioassay 1946–1949 None 

Mound Facility August, 1949–2000 

Buildings T, HH, WD, 

R, B

Po operations 

PoBe sources 

Feb/May 1949 – 1963 PoBe sources, Ra calibration sources, 

Co-60, Cr-55, Sr-90, Pb-210, Cs-137 

HH Building 

HH 3, 4, 5, 8 Separation of He-3; Kr-85 1960–1995 Xe-131 from reactor fission, Kr-85

HH 5, 6 Pa-231/sludge separation 1955–1956 Co-60

HH 8,10-12 Hydrolysis of Al & Bi chlorides 1949–1958 Co-60, Fe-55, Cr-55, Sr-90, 100-mrem 

gamma at 1 foot

PP Building Plutonium processing 12/1967–1995 Pu-238,239 plus trace amounts of Am-241 

Waste Handling Area Th-232 storage & redrumming 1955–1975 Th-232 and daughters, Pu-238, 239 

R Building R&D for Po & Pu programs 1948–2002 

R 117 Storage vault 1949–1990 PoBe, 238PuBe and 239PuBe neutrons 

SW Building 

SW 8 Separation of xenon Isotopes 1957–1960 Xe-131 from reactor fission (process 

moved to HH Building later)

SW 11-14 & 16 (area 

1B) 

Decon of BiSO4

Pa-231 purification

1951–1953 

1956 

Implementation of HH Building pilot 

project. Sr-Y-90, Cs-137, High gamma 

activity. Fe-59, Co-60 

SW 19 (Old Cave) Radium-actinium separation 1951–1953 Ra-226, Ac-227. By 1957, area was filled 

with concrete for external radiation 

protection. High gamma. 

SW 140 New Cave, 

SW-132, 136, 137 

Processing of residue from 

pitchblende processing 

1970–1979 Ra-223, -224, -226

SW 146-147 Metrology laboratory 1967–1985 Co-60 source 

SW 219 Neutron source production 1962–1965 PoBe, 238PuBe, 239PuBe neutron sources 

T Building 

All but T43,44,48, 

50-51, 57-59, & 

348-359 

Po neutron source program 1949–1973 Co-60, Cs-137, Bi-210, Sr-90, Y-90, Sc46, Zn-65, Hg-203, Ag-110m, Fe-59, Bi210 

H Building Hot laundry Possibly trace Ra-226 

WD Building Waste handling 

WD 1, 8, 101,104 Wastewater treatment 1949–1990 PoCl2; Fe-55; Co-60; Cs-137; Sr-90; Y-90; 

Am-241; Ra-224, -226, -228; K-40, Pu238, 239

WD 107 Ultrafiltration 1976–1981 Cs-137, Am-241 

WD 112, 113 Analytical work 1966–1979 Cs-137, Co-60 

WD 113, 118 A&B Storage, waste incineration 1979–1994 Cs-137, Co-60, Sr-90 

WD Penthouse Filter banks 1949–1993 Cs-137, Co-60, Sr-90, Am-241 

B Building Internal dose research 1950–1955 Minor Ra-226 contamination 

a. Plutonium emits both spontaneous fission neutrons and gamma photons. 

6.6.3.1 Photon Field

Sources of photons exist in almost all Mound facilities (see Table 6-1). The assumption of the two 

different groups was based on the processes and the surrounding shielding of each component of the 

facility and work areas. When photons are created, they are at high energies. As they slow down due 

to their interaction with air and shielding materials, the energies decrease due to Compton scattering. 

K and L fluorescent X-ray photons are produced.
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6.6.3.2 Beta Field

Based on site isotopic inventory, most betas are created with energies above 14 keV. Polonium-210 

does not emit betas. Strontium-90 was present in some places at Mound (see Table 6-22) during this 

period (King 1995). 

In addition, there were several beta-emitting contaminants of the irradiation process that produced 

210Bi slugs in production reactors, including 210Bi itself. The slugs were shipped to Mound for the 

extraction of polonium. For example, a shipment of irradiated slugs was received from Hanford on 

June 19, 1956. After the slugs were removed from the shipping casks, the water in the casks was 

drained and analyzed. The beta activity of 1 lambda of water from each cask was too high to count 

(Meyer 1956, p. 5). Another bismuth slug shipment from Hanford, received on April 12, 1957, had 

beta-contaminated water (Meyer 1957, p.4). 

According to Wilcox (1965) the polonium production slurry, which had been aged from 6 months to a 

year, contained the following beta emitters: Sc-46, Zn-65, Se-75, Hg-203, Ag-110m, Fe-59, and Co60. Significantly, Wilcox stated that there was no Sr-90 or Cs-137 in the slurry. The slug shipments 

themselves almost certainly contained radiologically significant quantities of 210Bi, but because 210Bi 

has a half-life of 5 days it was gone by the time the waste slurry had been aged for disposal. The 

electron emission from Se-75 has an energy of 12 keV. It is not externally hazardous and will not be 

considered further. Wilcox wrote that between October 1961 and July 1964, 7,599 30-gallon drums of 

210Po slurry had been shipped to Oak Ridge for burial and that burials before that time were 

“comparable in quantity and content.” It was anticipated that 2,000 more barrels would be shipped in 

1965, each containing less than two curies (Wilcox 1965). In 1971, a pilot unit was constructed to 

explore improvements in the processing of gross alpha and beta materials passing through the Mound 

Waste Disposal Facility (Flitcraft 1971, p. 15). Table 6-24 lists the beta emmiters of concern from 

polonium production.

Table 6-24. Externally hazardous beta emitters from 

polonium production at Mound. 

Isotope

Maximum energy

(MeV)

Average energy

(MeV) Half-life

Sc-46 0.357 0.112 84 days

Zn-65 0.329 (β+) 0.143 244 days

Hg-203 0.213 0.580 47 days

Ag-110m 0.531 0.165 250 days

Fe-59 1.565 0.615 45 days

Co-60 0.317 0.096 5.27 yr

Bi-210 1.161 0.389 5 days

Because of the quantity of bismuth processed, the contaminating isotopes listed above are the source 

of most of the potential beta dose at Mound through the decommissioning and cleanup of polonium 

operations in 1971. As early as October 1944, Captain Ferry reported that lead-lined gloves and 

tongs were needed to handle the irradiated bismuth slugs because of the beta radiation they emitted 

(Ferry 1944a) In 1944 the tolerance level for beta dose to the hands was 500 mrem/8-hr day or 3.5 

rem/40-hour week (Ferry 1944c). Therefore, the fact that lead lined gloves and tongs were necessary 

for handling bismuth slugs must mean that there was such a magnitude of beta radiation that even 

these lax extremity dose limits could not be met without them. A 1995 study performed at Mound 

(EG&G 1995) shows a lower LOD for beta particles to be about 53 mrem. This MDL for beta particles 

is based on a quarterly monitoring interval (Mound 2002). 
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6.6.3.3 Neutron Field

Mound has dealt with polonium-based neutron sources of all kinds, principally PoBe sources, as well 

as plutonium-based sources, principally PuF4 and PuO2. In the early years, Mound experimented with 

liquid and gaseous polonium neutron sources, and finally settled on solid-state sources. These 

products were used for initiators for nuclear weapons, calibration, and electric power for spacecraft. 

Neutrons were generated by the reaction of an alpha particle with an atomic nucleus, the (alpha,n) 

reaction. Polonium and plutonium produce alpha particles. In the early years, Mound settled on the 

use of beryllium as a target for alpha particles. With its low atomic mass number, beryllium produces 

neutrons with very high energies, between 4.0 and 4.5 MeV. After about 1959, Mound made sources 

with higher atomic mass numbers, such as oxygen and fluorine, which produced neutrons with lower 

energies, around 1.3 MeV. For the manufacture of RTGs for spacecraft, Mound selectively used 

plutonium and 18O, the highest common atomic mass of oxygen, to minimize neutron emissions from 

the RTGs. The use of plutonium as a component of neutron and electricity sources added a minor 

component of spontaneous fission neutrons to the neutron energy mix. 

Neutron weighting factor adjustments

At Mound a quality factor of 7.0 is used with all neutron energies (Meyer 1959). To obtain the 

correction factors, the dose fraction per neutron energy is multiplied by the ICRP Publication 60 

weighting factor (ICRP 1991) and the product is divided by the Mound quality factor. For example, a 

dose fraction of 25% for neutrons with energies between 0.1 and 2.0 MeV with a Publication 60 

weighting factor of 20 and a Mound quality factor of 7.0 results in a correction factor of 0.72. Neutron 

doses must be adjusted accordingly. Table 6-23 lists the correction factors. 

6.6.3.4 Mixed Field

Most Mound facilities have or had mixed fields of neutrons and photons. The beta field probably is 

not a concern for external radiation protection because the shielding around the glovebox or process 

shielding can stop the range of the maximum beta. For researchers who deal with samples of 

beta-emitting isotopes, gloves and a laboratory coat would stop most beta radiation. In general, the 

ratio of neutrons to photons produced at Mound was 2 to 1 (Anderson and Crain 1981; Anderson 

1985). 

6.7 WORKDAY HOURS AND EXPOSURE GEOMETRY

The amount of time spent at each task is based on the lowest amount of exposure. A health physicist 

determines the amount of time required for the task and assigns the number of workers for the task 

that will yield the minimum amount of exposure. To be claimant-favorable, an 8- to 10-hr workday is 

assumed.

All worker geometry is assumed to be AP for 100% of worked hours for all workers.
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6.8 GUIDE TO IMPORTANT TABLES 

To calculate the following parameters: Use table

Monitoring intervals or exchange frequency 6-4

Facility and radiation field characterizations 6-9

Neutron dosimeter uncertainty 6-19

Neutron dosimeter lower LOD 6-21 

Neutron dosimeter correction factors 6-17

Neutron missed dose 6-17

Neutron missed dose uncertainty 6-19

Dosimeter used or dosimeter characteristics 6-5

Facility and radioisotopes of external hazard 6-22 

Photon dosimeter uncertainty 6-18

Photon dosimeter lower LOD 6-20

Photon dosimeter correction factors 6-16

Photon missed dose 6-16

Photon missed dose uncertainty 6-18
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 39 of 51

REFERENCES

Anderson, M. E., 1977, memorandum to W. A. Bigler on the accuracy of Mound TLD dosimetry for 

low-energy photons, May 25, Mound Laboratory, Miamisburg, Ohio.

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Handbook 57, U.S. Government Printing Office, Washington D.C.

NBS (National Bureau of Standards), 1960, Measurement of Neutron Flux and Spectra for Physical 

and Biological Applications, Handbook 72, U.S. Government Printing Office, Washington D.C.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0016-6 Page 43 of 51

NBS (National Bureau of Standards), 1961, Measurement of Absorbed Dose of Neutrons and of 

Mixtures of Neutrons and Gamma Rays, Handbook 75, U.S. Government Printing Office, 

Washington D.C.

NBS (National Bureau of Standards), 1964, Physical Aspects of Irradiation, Recommendations of the 

International Commission on Radiological Units and Measurements, Handbook 85, U.S. 

Government Printing Office, Washington, D.C.
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GLOSSARY

absorbed dose

The energy absorbed per unit mass at a specific place in a material. The historical unit of 

absorbed dose is the rad; the International System unit of absorbed dose is the gray, which is 

joules per kilogram; 1 gray equals 100 rad.

accident dosimetry

Determination of high levels of deep absorbed dose resulting from uncontrolled conditions.

albedo effect

As used in this document, the neutron dosimeter response caused by the moderating and 

backscattering properties of a phantom or the human thorax for neutron radiation.

algorithm

A computational process for calculating dose equivalent from the response of individual 

elements in a dosimeter.

angular dependence

The response of a dosimeter as a function of the angle of incidence of the radiation detected 

compared with its response at normal incidence.

backscatter

Scattered radiation with an angle more than 90 degrees from the incident direction.

beta particle

An electron or positron emitted from a nucleus during beta decay.

beta radiation

Radiation consisting of beta particles.

blackness

The effect of radiation on a film dosimeter.

calibration

Adjustment or determination of the response or reading of an instrument relative to a standard 

or a series of conventionally true values.

deep dose

Deep dose equivalent at a tissue depth of 1,000 milligrams per square centimeter.

densitometer

A device used to measure the amount of blackness on a film dosimeter. 

density reading

The reading obtained from the densitometer when a film badge is read. It is the log of the ratio 

of reference light intensity to the transmitted light intensity of the film. 

DOE Laboratory Accreditation Program (DOELAP)

Laboratory accreditation program administered by the U.S. Department of Energy.
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dose equivalent

The product of absorbed dose and radiation quality factor at the point of interest in tissue. The 

historic unit of dose equivalent is the rem. When absorbed dose is expressed in rad, dose 

equivalent is expressed in rem. When absorbed dose is expressed in gray, dose equivalent is 

expressed in sieverts, where 1 sievert equals 100 rem.

dosimeter

Complete assembly consisting of dosimeter card containing radiation sensitive phosphors or a 

film packet inserted into a dosimeter holder.

dosimeter card

An aluminum card containing one or more radiation responsive phosphors.

dosimeter holder

A plastic holder used to contain the dosimeter card. The holder typically has one or more 

metallic filters used to modify the response of the phosphor to radiation.

dosimetry system

A system used to assess dose equivalent from external radiation. This system includes the 

selection, placement, and processing of the dosimeters; interpretation and recording of results; 

and the means by which the quality of results is assured.

External Radiation Analysis Database (ERAD)

Contains exposure history of a radiation worker. Used at Mound from 1978 to 1989.

External Exposure Analysis System (EXAS)

Contains exposure history of a radiation worker. Used at Mound from 1958 to 1978.

external dosimetry

Theory and application of the principles and techniques involved in the measurement and 

recording of radiation absorbed dose, dose equivalent, and effective dose equivalent from

external sources of radiation.

extremities

The hand and arm below the elbow, or the feet and legs below the knee

eye dose

Refers to dose at a tissue depth of 300 milligrams per square centimeter.

facility-specific calibration factor

The dosimeter calibration factor applicable to a particular occupational environment. These 

factors are determined by comparison of reference instrument with dosimeter response 

measurements. Both measurements are performed in the workplace.

fast neutrons

Neutrons with energies greater than 10 keV and less than 10 MeV.

field calibration

The calibration of dosimeters based on radiation types, intensities and energies of the work 

environment.
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film

A radiation-sensitive medium used to measure radiation absorbed by radiation worker. 

film density

See optical density.

Form 1015-X 

The form on which the worker dose record is coded. It was used at Mound before 1958.

free field dose equivalent

The dose equivalent assigned for neutron irradiation as if it were performed in free space with 

no background due to air and room scattering and no source asymmetry.

high gamma dose

Penetrating gamma dose measured under the shielded portion of film badge.

in-air exposure

As used in this document, exposure of a dosimeter without any phantom.

ionizing radiation

Any radiation capable of displacing electrons from atoms or molecules, thereby producing ions

isotopes

Different forms of the same element that contain the same atomic number.

lower limit of detection

The minimum evaluated dose equivalent for which the readout value of a dosimeter is 

significantly different at the 95% confidence level from the readout value at the detection 

threshold.

low gamma dose

Nonpenetrating gamma dose.

Manhattan Engineering District (MED)

A government-based project to develop a nuclear bomb and study the characteristics of all 

elements of interest associated with all elements of the program.

MESH database

A Mound environmental, safety, and health database that includes radiation exposure history 

of an individual radiation worker. It has been used at Mound from 1989 to the present.

radiation monitoring

Actions intended to detect and quantify radiological conditions.

neutron activation

The process in which atomic nuclei become radioactive by absorption of neutrons.

neutron radiation

One of the fundamental particles of the atomic nucleus with a neutral charge.
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nonuniform fields

The condition when a portion of the body is expected to receive a radiation dose equivalent 

that varies by more than 50% from dose equivalent expected at a reference location.

nonpenetrating dose

As used in this document, the photon dose of long wavelength measured under the open 

window.

on-phantom

As used in this document, exposure of dosimeters affixed to a phantom to simulate the 

dosimeter response while the dosimeter is being worn by a person.

open window

The part of a dosimeter that contains no metallic filter and no holder mass. The only material 

between radiation and film is the wrapping paper of the film.

optical density

A measure of a degree of blackness of film dosimeter. It is the log of the ratio of reference 

light intensity to the transmitted light intensity of the film. 

penetrating dose

Deep photon dose measured at 1,000 milligrams per square centimeter.

phantom

A slab of plastic, typically measuring either 30 by 30 centimeters square by 15 centimeters 

deep or 40 by 40 centimeters square by 15 centimeters deep, used to simulate the effect of 

the body on dosimeter response. Can also refer to an anthropomorphic phantom used for the 

same purpose.

photon radiation

Either X- or gamma radiation.

protection dosimetry

Routine estimation of the shallow and deep dose.

quality factor

A modification factor used to obtain dose equivalent form absorbed dose.

radiation

As used in this document, ionizing radiation.

radiation worker

An occupational worker whose job assignment requires work on, with, or in the proximity of 

radiation–producing machines or radioactive materials and/or who has the potential of being 

routinely exposed above 0.1 rem per year, where the total annual dose is the sum of the 

annual effective dose equivalent from external irradiation and the committed effective dose 

equivalent from internal irradiation.

radioactivity

Unstable isotopes that release energy in the form of particles and/or electromagnetic radiation 

by a process of disintegration.
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rad

The historical unit of absorbed dose. One rad is equal to 100 ergs per gram. The word 

derives from radiation absorbed dose. 

relative biological effectiveness (RBE)

The ratio of dose of X-rays required for a biological effect to dose of other radiation that 

produces the same effect.

roentgen

A special unit of radiation used to quantify ionization in air from photon radiation. One 

roentgen is equivalent to 2.58 × 10-4 coulomb per kilogram.

rep

The amount of energy absorbed per unit mass of tissue; 1 rep is equal to 93 ergs per gram of 

absorbed in tissue. The term derives from roentgen-equivalent-physical. 

rem

The historical unit of dose equivalent. The word derives from roentgen equivalent [in] man. 

scattering

A process by which particles can change direction and have a momentum and energy change.

sievert

The special name for the International System unit of dose equivalent; 1 sievert equals 1 joule 

per kilogram, which equals 100 rem.

shallow dose equivalent 

Dose equivalent measured at 7 milligrams per centimeter squared.

skin dose

Shallow dose equivalent at a tissue depth of 7 milligrams per centimeter squared.

superficial dose

Another term for skin dose.

thermal neutrons

Neutrons in thermal equilibrium in the medium. The mean energy of thermal neutrons is 

0.025 electron-volts.

TLD chip

A compound made of lithium fluoride and activated by manganese and thallium that stores 

energy in crystal imperfections.

thermoluminescent dosimeter (TLD)

A type of dosimeter that relies on excitation of the crystalline lattice by radiation of certain 

fluorescent materials that emit light when heated.

whole-body dose equivalent

The dose equivalent that results when the whole body is irradiated. If the irradiation is 

uniform, whole-body dose equivalent is the same as effective dose equivalent. Whole-body 

dose equivalent is expressed in the same units as dose equivalent.
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whole-body irradiation

Uniform radiation exposure of the gonads, active blood-forming organs, head, trunk, lens of 

the eye, arms above and including the elbow, and legs above and including the knee.
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ATTACHMENT 6A

SUMMARY INSTRUCTIONS FOR MOUND DOSE RECONSTRUCTORS

The entire record of weekly, biweekly, or monthly doses to individuals between 1947 and the end of 

1977 has been lost (Guido 2003). As a consequence, unless other records are found or the claimant 

can provide detailed records, all dose records for 1947 through 1977 consist only of annual dose 

summaries for each monitored worker. Therefore, missed dose increments through the end of 1977 

are given in annual units so the dose reconstructor can add them directly to the annual dose 

summaries that comprise the extant dose of record. These missed dose increments are the 

maximum possible missed doses calculated from the most frequent badge exchanges for each 

period. If there is no annual dose of record, dose reconstructors should assume that the base annual 

dose for this period is the cohort dose or the modal neutron and photon dose for 1949 through 1959 

(i.e., 50 mrem/wk for 52 wk/yr is 2,600 mrem/yr for neutrons or 1,040 mrem/yr for photons).

Corrected neutron doses: The base and missed neutron doses are multiplied by the ICRP 60 

correction factors (ICRP 1991) to obtain the corrected neutron doses. Table 6A-1 lists the corrected 

doses.

Table 6A-1. Corrected neutron doses. 

Period Base dosea

Missed neutron 

dose/unit timea

ICRP 60 

correction factors 

Corrected base neutron 

dose mrem/unit time 

Corrected missed neutron 

dose mrem/unit time 

1943-8/1949 2,600 mrem/yrb

imputed 

3,822 mrem/yrc 25% 0.l-2.0 MeV = 0.72, 

75% 2.0-20 MeV = 1.1

1,872+2,860 = 4732/yr 2,751+4,204 = 6,955/yr

9/1949–11/1951 Dose of record 

or cohort dose

3,822 mrem/yrd 25%: 0.72; 75%: 1.1 (DORf× CF) 2,751+4,204/yr = 6,956/yr

12/1951–1960 Dose of record 

or cohort dose

2,392 mrem/yre 25%: 0.1-2.0 MeV = 0.72; 

75%: 2.0-20 MeV = 1.1 

(DOR × CF) 1,722+2,631/yr = 4,353/yr 

1960-12/1973 Dose of record 

or cohort dose

2,392 mrem/yre 50% 0.1-2.0 MeV = 1.43; 

50% 2.0-20 MeV = 0.72

(DOR × CF) 3,420+1,722/yr = 5,142/yr 

1/1974–12/1976 Dose of record 

or cohort dose

2,392 mrem/yre 75% 0.1-2.0 MeV = 2.2;

25% 2.0 - 20 MeV = 0.36 

(DOR × CF) (2,392 × 2.2) +(2,392 × 0.36) 

= 6,123/yr 

1977 Dose of record 

or cohort dose 

21/mo = 252 

mrem/yre

2.2 (DOR × CF) (252 × 2.2) = 554/yr 

1978–1981 Dose of record 

or cohort dose

21/mo or 25/qtr 2.2 (DOR × CF) 46/mo or 55/qtr

1981–1991 Dose of record 

or cohort dose

10/mo or 10/qtr 2.2 (DOR × CF) 22/mo or qtr

1991–2002 Dose of record 

or cohort dose

10/qtr 2.2 (DOR × CF) 22/mo or qtr

Sample calculation, 1943-8/1949: 2,600 mrem/yr imputed dose and 3,822 mrem/yr missed dose. ICRP 60 corrected dose assumes that 25% of the dose lies in 

the energy range 0.1 to 2.0 MeV whose integrated correction factor is 0.72 and 75% of the dose lies in the energy range 2.0 to 20 MeV whose integrated 

correction factor is 1.1. The calculation is: [(2,600 × 0.72) + (2,600 × 1.1)] = 1,872 + 2,860 = 4,732 mrem/yr. The correction factors change over time as the 

neutron energy mix changes. 

a. Based upon the historic Mound neutron RBE of 7. 

b. Dose imputed from 50 mrem/wk historical modal neutron dose, 9/1949-1959 calculated as follows: (50 mrem/wk × 52 wk/yr) = 2,600 mrem/yr. 

c. Assumes a weekly badge exchange frequency and 52 exchanges. 

d. Assumes a biweekly badge exchange frequency and 26 exchanges. 

e. Assumes a biweekly badge exchange frequency and 26 exchanges. 

f. DOR = Dose of Record or Cohort Dose; 7 CDOR = Corrected Dose of Record.
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Corrected photon doses: The base and missed photon doses are multiplied by the photon correction 

factors to obtain the corrected photon doses. Table 6A-2 lists the corrected doses.

Table 6A-2. Corrected photon doses. 

Period Base dose Missed photon dose

Photon energy

interval in keV

1943–1/1949 2,600 mrem/yr imputeda 1,040 mrem/yr 30–250:

>250:

25%

75%

2/1949–11/1951 Dose of record or cohort dose 442 mrem/yr 30–250:

>250:

50%c

50%

12/1951–1957 Dose or record or cohort dose 442 mrem/yr 30–250:

>250:

50%c,d

50%

1957–1963 Dose of record or cohort dose 130 mrem/yr 30–250:

>250:

50%d

50%

1963–1968 Dose of record or cohort dose 390 mrem/yr 30–250:

>250:

50%d

50%

1968–11/12/1970 Dose of record or cohort dose 390 mrem/yr 30–250:

>250:

50%d

50%

11/13/1970–12/31/75 Dose of record or cohort dose 78 mrem/yr 30–250:

>250:

50%d

50%

1976-1977 Dose of record or cohort dose 572/yre 30–250:

>250:

50%

50%

01/01/1978–7/6/1981 Dose of record or cohort dose 572/2 wk, 456/mo or 376/qtr 30–250:

>250:

50%

50%

7/7/1981–1991 Dose of record or cohort dose 260/2 wk (bldg PP)d, 120 

monthly, 40/qtr

30–250:

>250:

50%

50%

1992-2006 Dose of record or cohort dose Not applicable 30–250:

>250:

50%

50%

a. Dose imputed from 50 mrem/wk historical modal photon dose, 9/1949-1959 calculated as follows: (50 mrem/wk × 52 

wk/yr) = 2,600 mrem/yr. 

b. For years before 1978, only annual dose summaries are available from the record. Therefore, all the corrections 

through the end of 1977 are in terms of annual additions to the dose of record.

c. The internal dose research section in B Building from 1950 to 1955 had a photon energy interval of (30 to 250 keV: 

75%, >250: 25%). 

d. The waste handling in Building PP from 1955 to 1975 had a photon energy interval of (30 to 250: 75%, >250: 25%). 

e. This is the maximum missed dose for 1976 to 1977 assuming tantalum-shielded badges and 26 badge exchanges 

throughout the entire period. Even though tantalum-shielded badges came into use on April 21, 1976, this analysis 

assumed their use for the entirety of both 1976 and 1977 for ease of computation.