ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for Portsmouth Gaseous Diffusion 

Plant – Occupational External Dose

Document Number:

ORAUT-TKBS-0015-6

Effective Date: 01/18/2005

Revision No.: 00

Controlled Copy No.: _______

Page 1 of 55

Subject Expert: Paul J. Demopoulos

Document Owner

Approval: Signature on File Date: 12/30/2004

Mark D. Notich, TBD Team Leader

Approval: Signature on File Date: 12/30/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 01/03/2005

Richard E. Toohey, Project Director

Approval: Signature on File Date: 01/18/2005

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

5

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

6.0 Occupational External Dosimetry ................................................................................................8

6.1 Introduction.................................................................................................................................8

6.2 Basis of Comparison...................................................................................................................9

6.3 Dose Reconstruction Parameters ............................................................................................. 10

6.3.1 PORTS Historic Administrative Practices ......................................................................... 10

6.3.2 PORTS Dosimetry Technology......................................................................................... 11

6.3.2.1 Beta/Photon Dosimeters.......................................................................................... 13

6.3.2.1.1 Film Dosimetry Two-Element, 1954-1980 ....................................................... 13

6.3.2.1.2 Thermoluminescent Dosimeters, 1981-present............................................... 15

6.3.2.2 Neutron Dosimeters................................................................................................. 16

6.3.3 Calibration........................................................................................................................18

6.3.3.1 Film Badges Beta/Photon........................................................................................ 18

6.3.3.2 TLD Badges Beta/Gamma....................................................................................... 20

6.3.3.3 TLD Badges Albedo Neutron................................................................................... 20

6.3.4 Workplace Radiation Fields.............................................................................................. 22

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

6.3.4.2 PORTS Workplace Neutron Response.................................................................... 24

6.3.4.2.1 Calibration Laboratory in Building 710 ............................................................ 25

6.3.4.2.2 Nondestructive Assay Laboratory in Building 710 ........................................... 28

6.3.4.2.3 HEU Storage Vault in Building 345 ................................................................. 29
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6.3.4.2.4 Neutron-to-Photon Ratio................................................................................. 30

6.3.4.3 PORTS Workplace Neutron Dosimeter Response................................................... 30

6.3.5 PORTS Workplace Dose Uncertainty............................................................................... 31

6.4 Adjustments to Recorded Dose................................................................................................. 31

6.4.1 Beta Dose Adjustments.................................................................................................... 31

6.4.2 Photon Dose Adjustments ................................................................................................ 31

6.4.3 Neutron Dose Adjustments............................................................................................... 32

6.5 Missed and unmonitored Dose.................................................................................................. 33

6.5.1 Missed and Unmonitored Shallow Dose ........................................................................... 34

6.5.2 Missed and Unmonitored Photon Dose ............................................................................ 40

6.5.3 Missed and Unmonitored Neutron Dose........................................................................... 43

6.6 Organ Dose ..............................................................................................................................44

6.6.1 Organ Dose Conversion Factors ...................................................................................... 44

6.7 Dose Reconstruction.................................................................................................................44

References ........................................................................................................................................45

Glossary ............................................................................................................................................49
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LIST OF TABLES

Table

6-1 Film badge meter calendar.....................................................................................................

Page

10

6-2 Security badge letter prefix..................................................................................................... 11

6-3 PORTS dosimeter type, period of use, exchange frequency, LOD, and 

potential annual dose equivalent missed................................................................................ 12

6-4 PORTS historical dosimetry events........................................................................................ 13

6-5 IARC testing results for US beta/photon dosimeters............................................................... 15

6-6 Testing results for Hanford two-element and multi-element film 

dosimeters for energy and angular response ......................................................................... 15

6-7 Harshaw Type L card element and filter description............................................................... 16

6-8 Personal neutron dosimetry results November 1996 – February 1997 ................................... 17

6-9 Area neutron dosimetry results November 1996 – February 1997.......................................... 18

6-10 Roentgen-to-rem conversion factors used for TLD dosimeter calibration ............................... 20

6-11 Laboratory sources of uncertainty for beta/photon dosimeter calibration 

parameters.............................................................................................................................21

6-12 Laboratory sources of uncertainty for neutron dosimeter calibration 

parameters.............................................................................................................................21

6-13 PORTS workplace beta/photon dosimeter response.............................................................. 23

6-14 Major facilities at PORTS where 99Tc might have accumulated.............................................. 24

6-15 Reactor returns fed to cascade .............................................................................................. 24

6-16 Properties of radionuclides that might be found at uranium facilities ...................................... 26

6-17 Common workplace beta/photon dosimeter Hp(10) performance ........................................... 27

6-18 Dose fractions for PORTS calibration facility.......................................................................... 28

6-19 Dose fractions for PORTS nondestructive assay laboratory facility in 

Building 710 ...........................................................................................................................29

6-20 Dose fractions for PORTS HEU storage vault in Building 345 ................................................ 29

6-21 PORTS neutron-to-photon dose ratios ................................................................................... 30

6-22 Typical workplace neutron dosimeter Hp(10) performance..................................................... 30

6-23 Estimates of uncertainty......................................................................................................... 31

6-24 Adjustments to reported PORTS deep photon dose............................................................... 31

6-25 Historical neutron quality or weighting factors ........................................................................ 32

6-26 PORTS facility neutron dose fractions and associated ICRP (1991) 

correction factors ...................................................................................................................33

6-27 Missed beta dose according to dosimeter type....................................................................... 34

6-28 Beta or SDE dose by year...................................................................................................... 35

6-29 Departmental SDE dose ratios............................................................................................... 37

6-30 Beta contamination plant limits for PORTS ............................................................................ 38

6-31 Shallow dose-equivalent rates for 99Tc ................................................................................... 39

6-32 Missed photon dose adjustments to recorded deep dose according to 

PORTS facility........................................................................................................................40

6-33 Photon or DDE dose by year.................................................................................................. 41

6-34 Departmental DDE dose ratios............................................................................................... 42

6-35 PORTS dosimeter type, period of use, exchange frequency, LOD, and 

potential annual dose missed................................................................................................. 43
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LIST OF FIGURES

Figure

6-1 Measured Hanford two-element dosimeter photon response 

characteristics........................................................................................................................

Page

19

6-2 Results of neutron spectrum measurements made at 1 meter from bare 

252Cf fission neutron source.................................................................................................... 28

6-3 Results of neutron spectrum measurements made in the front of scuffler 

unit29

6-4 Results of neutron spectrum measurements made about 24 in. in front of 

93 – 96% HEU cylinders ........................................................................................................ 29

6-5 PORTS SDE average and minimum dose by year ................................................................. 36

6-6 PORTS DDE average and minimum dose by year................................................................. 42
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 11/21/2003 00-A Technical Basis Document for Portsmouth Gaseous 

Diffusion Plant – Occupational External Dose. Initiated 

by Mark D. Notich

Draft 12/22/2003 00-B Incorporates OCAS and internal review comments. 

Initiated by Mark D. Notich.

Draft 5/10/2004 00-C Incorporates GDP task group comments. Initiated by 

Mark D. Notich.

Draft 12/07/04 00-D Incorporates changes for consistency with other 

gaseous diffusion plants and completes comment 

resolution. Incorporates comment resolutions based on 

Task 5 recommendations. Initiated by Mark D. Notich.

Draft 12/29/2004 00-E Incorporated NIOSH review comments. Initiated by 

Mark D. Notich.

01/18/2005 01/18/2005 00 First approved issue. Initiated by Mark D. Notich.
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 6 of 55

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ANL Argonne National Laboratory

AP anterior-posterior (X-ray view)

AWE Atomic Weapons Employee

BJC Bechtel Jacobs Company

CDC Centers for Disease Control and Prevention

CFR Code of Federal Regulations

cm2 square centimeter

cpm counts per minute

DDE deep dose equivalent

DE dose equivalent

DL detection limit

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm discintigrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ERP extended range product

EU enriched uranium

FBI Federal Bureau of Investigation

GAT Goodyear Atomic Corporation

GDP gaseous diffusion plant

GM Geiger mueller detector

HASA High-Assay Storage Area

HEU highly enriched uranium

HEW Hanford Engineering Works

HFM Hand and Foot monitor

HMPD Hanford Multipurpose TLD

HP health physics

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

hr hour

IARC International Agency for Research on Cancer

ICN International Chemical and Nuclear Corporation

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

LANL Los Alamos National Laboratory
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LAT lateral geometry

LAW low-assay withdrawal

LLDR lower limit of dose limit

LOD limit of detection

LODR limit of dose rate

MDL Minimum Detection Level

MED Manhattan Engineering District

mg milligram

mm millimeter

mR milliroentgen

mrem millirem

mSv milliseivert

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NRC Nuclear Regulatory Commission

NVLAP National Laboratory Accreditation Program

ORNL Oak Ridge National Laboratory

OW open window (i.e., no filter) nonpenetrating dose

PGDP Paducah Gaseous Diffusion Plant

PNL Pacific Northwest Laboratory

PORTS Portsmouth Gaseous Diffusion Plant

PW product withdrawal

QF quality factor

R Roentgen

rem radiation equivalent man

rep radiation equivalent physical

RN radionuclide

ROT rotational geometry

RU recycled uranium

SDE Shallow Dose equivalent

SNM special nuclear material

SRS Savannah River Site

TBD technical basis document

TED track-etch dosimetry 

TEPC Tissue Equivalent Proportional Counter

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

U.S.C. United States Code

USEC United States Enrichment Corporation

WB whole-body
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6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSIMETRY

Technical Basis Documents (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 Employees 

Occupational Illness Compensation Program Act of 2000 (EEOICPA; 42 U.S.C. § 7384l (5) and (12)). 

PORTS operations, which involved several processes of the nuclear enrichment cycle, played a 

significant role in the U.S. energy program and the U.S. Department of Defense (DOD) nuclear fuel 

program. These processes included nuclear fuel enrichment; radiochemical separations; refining, 

finishing, and storing uranium; and handling the associated radioactive waste. 

PORTS workers, especially those employed during the production decades of the 1950s and 1960s, 

have been exposed to radiation types and energies associated with nuclear energy development 

processes. PORTS used facility and individual worker monitoring methods to measure and control 

radiation exposures. Evaluations are difficult because the extensive scope of facility, process, and 

worker information relevant to an individual worker’s potential dose might involve many years or even 

decades after employment. 

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

coworkers are available for PORTS operations beginning in 1954. Doses received by these 

dosimeters were recorded at the time of measurement and routinely reviewed by the PORTS 

operations and radiation safety staff for compliance with radiation control limits. The External Dose 

Reconstruction Implementation Guide (NIOSH 2002) indicates that these records represent the 

highest quality records for retrospective dose assessments. 

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

radium and X-ray medical diagnostic and therapy applications. These methods were well advanced 

at the start of the Manhattan Engineering District (MED) program to develop nuclear weapons in 

about 1940. The primary new challenges encountered by MED, and later U.S. Atomic Energy 

Commission (AEC), operations to measure worker dose to external radiation involved:

• Comparatively large quantities of high-level radioactivity

• Mixed radiation fields involving beta, photon (gamma and X-ray), and neutron radiation with 

low, intermediate, and high energies

• Neutron radiation

From 1954 until 1986, Goodyear Atomic Corporation (GAT) operated the site. In 1986, Martin 

Marietta Energy Systems, Inc., assumed responsibility for PORTS operations. The Energy Policy Act 

of 1992 transferred responsibility for the PORTS site from the U.S. Department of Energy (DOE) to a 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 9 of 55

newly created entity, the United States Enrichment Corporation (USEC), which leased the 

nonoperational portion of PORTS to Bechtel Jacobs, the DOE primary contractor (USEC 2003). On 

July 1, 1993, the operational side of PORTS officially transferred to USEC (USEC 2003), which is 

regulated by the U.S. Nuclear Regulatory Commission (NRC). From 1993 to 1999, Martin Marietta 

Utility Services, which became Lockheed Martin Utility Services, operated the USEC portion of the 

site as the primary contractor. In May 1999, USEC assumed control without a primary contractor. 

The remaining portion of PORTS (including cylinder storage and legacy wastes) is under the 

operation of DOE; Bechtel Jacobs Company (BJC ) has been the primary contractor since April 1998.

PORTS employees have been exposed to gamma, beta, neutron, and X-ray radiation. Early 

employees were exposed to higher levels of radiation due to either (1) recycled uranium that might 

have included transuranics and 99Tc or (2) highly enriched uranium, or just to greater amounts of 

processed materials. Since the beginning of operations in 1954 through 1980, PORTS monitored its 

employees for external exposures beginning with film dosimetry and, from 1981 to the present, with 

various forms of thermoluminescent dosimetry. 

Records of surveys, investigations, procedures, and facility controls are available. The types of 

instrumentation and procedures used over the history of the PORTS site have varied. NIOSH (2002) 

recognizes that personnel external dosimetry issued to workers is the best way to determine external 

dose. Methods of calibration and limitations of dosimetry systems have been documented with onand offsite references, as indicated in this TBD. 

6.2 BASIS OF COMPARISON 

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

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

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

penetrating radiation of significance to whole-body dose, d = 10 mm and is noted as Hp(10). The

International Commission on Radiological Units and Measurements (ICRU) recommends the use of 

both Hp(0.07) and Hp(10) as the operational quantities to be recorded for radiological protection 

(ICRU 1993). In addition, Hp(0.07) and Hp(10) are the quantities used in the DOE Laboratory 

Accreditation Program (DOELAP), which the Department has used to accredit personnel dosimetry 

systems since the 1980s. The National Laboratory Accreditation Program (NVLAP), which is the NRC 

equivalent to DOELAP, uses the same operational quantities. 

PORTS has monitored photon (gamma) and beta radiation since 1954. Therefore, comparisons with 

similar dosimetry systems monitoring near-equivalent radiation work environments are possible. 

PORTS operated its’ own film dosimetry system from 1954 through 1980, and its’ own 

thermoluminescent dosimetry system from 1981 to January 1, 1999. The Plant did not monitor 

neutron dose with personal dosimetry until 1992, then operated an onsite neutron dosimetry system 

from 1992 through 1994. By 1995, vendors provided dosimetry for USEC [with the International 

Chemical and Nuclear Corporation (ICN) as the vendor] and Bechtel Jacobs Company (BJC) [Oak 

Ridge National Laboratory (ORNL) with Y-12 laboratories as the vendor]. As early as 1963, PORTS 

recognized that the health physics (HP) group did not have adequate instrumentation to monitor for 

neutrons. The policy was that neutron monitoring was not necessary (Cardarelli 1997). Pacific 

Northwest Laboratories conducted a neutron survey in 1992 (Soldat and Tanner 1992). PORTS did 

not routinely monitor for neutrons until 1997. 
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6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray and gamma ray), and neutron radiation type, energy, and 

geometry of exposure in the workplace, and the characteristics of the PORTS dosimeter responses 

are crucial to the assessment of bias and uncertainty of the original recorded dose in relation to the 

radiation quantity Hp(10). Earlier dosimetry systems can be compared to current systems to evaluate

their performance, based on the premise that current systems have more stringent criteria as 

indicated in DOELAP and NVLAP programs. 

Accuracy and precision of the recorded individual worker doses depend on (Fix et al. 1997):

• Administrative practices that facilities adopt to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations.

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

as the response to different types and energies of radiation, in particular in mixed radiation 

fields.

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

sources of exposure in the workplace.

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

exposure geometries, and environmental conditions.

An evaluation of the original recorded doses, as available, combined with detailed examinations of 

workplace radiation fields and dosimeter responses to those fields is the recommended option to 

provide the best estimate of Hp(d) for individual workers. 

6.3.1

Monitoring at PORTS included radiation level monitoring with portable and area instrumentation, use 

of pocket ionization chambers when necessary, establishment of radiation areas and high radiation 

zones, and criticality monitoring with personnel and area instrumentation and personnel dosimeters. 

PORTS Historic Administrative Practices

PORTS based the frequency of personnel dosimetry on safety policy. When operations began in 

1954, personnel received dosimetry based on job assignments. In 1960, personnel dosimetry was 

assigned to all employees, contractors, and visitors as a picture identification and personnel badge. 

PORTS analyzed all badges assigned to radiation workers, but not all visitor or contractor badges 

were analyzed. In 1981, personnel dosimetry changed from film dosimeters to thermoluminescent 

dosimeters (TLDs) with frequencies varying from monthly to quarterly based on work assignments. 

As indicated below, a number of frequency and monitoring selection changes have occurred over time 

at PORTS. (GAT 1964) 

One administrative tool used at PORTS to maintain film badge, and later TLD badge, control is color 

coding. Badges for selected departments used the color codes listed in Table 6-1. This was first 

listed in the film badge procedure dated May 3, 1963, and continued until PORTS used outside 

personal dosimetry vendors (January 1, 1999, for USEC employees and January 1, 1997, for BJC 

employees for neutron monitoring).

Table 6-1. Film badge meter 

calendar (GAT 1971).
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Insert colors Quarter

Black – Red 1

Blue – Green 2

Black – Red 3

Blue – Green 4

Table 6-2 lists letter prefixes used for badge inserts (GAT 1971); these codes might occur in 

dosimetry records or databases.

Table 6-2. Security badge letter prefix.

Badge HP insert Personnel

A A Akron GT&R officials

PM AO AEC personnel

SO SO FBI personnel

B B Nationwide and all food vendors

C C OVEC

CC CO Construction contractors

F F All temporary visitors and employee lost badges, etc.

J J Miscellaneous (IBM, General Telephone, etc.)

H H Hurst dosimeters (Criticality Dosimeters)

E E Equipment and emergency badges

Note: no prefix was used for PORTS employees.

In another administrative method, PORTS places badges that cannot be assigned to personnel, 

visitors, or contractors in the “bucket” file. In addition, since at least December 29, 1969, PORTS has 

processed badges at random for employees not among the “selected employees” at a rate of 100 per 

quarter for spot checks; the Plant records positive results only for readings greater than the limit of 

detection (LOD) of 30 mrem (GAT 1971). “Selected employees” were determined by department, job 

category, or job assignments. Anyone likely to exceed 10% of the contemporary regulatory limits 

probably received permanent badges. This policy continued from the film dosimetry program though 

the TLD program until it ended on January 1, 1999.

On occasion, PORTS has assigned extremity dosimeters to workers based on their job categories 

and potential to exceed 10% of the DOE/NRC regulatory limit. [The LOD is about 30 mrem. This is 

the same as the Minimum Detectable level (MDL).] Table 6-3 lists the dosimeters used at PORTS.

6.3.2

PORTS maintained onsite personnel dosimetry from 1954 through 1998. The dosimetry section 

followed operational and technical guidelines, as indicated in the available procedural information. As 

listed in Table 6-4, which summarizes major events in the PORTS personnel dosimetry program, the 

program was dynamic. Changes occurred due to changes in dosimetry technology, regulatory 

guidance and plant operations. 

PORTS Dosimetry Technology
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Table 6-3. PORTS dosimeter type, period of use, exchange frequency, LOD, and potential annual dose equivalent missed (rem).

Dosimeter Period of use Exchange frequency

Laboratory 

LOD

Maximum 

annual 

missed dosea

Beta/photon dosimeters

PORTS film 2-elementb 9/22/54 - 7/16/57 Weekly (n=52) {selected groups} 0.03c 0.78

7/17/57 - 9/30/58 Biweekly (n=26) {selected groups} 0.03c 0.39

10/01/58-4/8/59 Weekly (n=52) {chemical operators and material handlers}

Monthly (n=12) {remainder of selected groups}

0.03c

0.03c

0.78

0.18

4/9/59 - 7/31/60 Every 4 weeks (n=13) {all selected groups} 0.03c 0.195

8/1/60 - 7/5/64 Monthly (n=12) {all selected groups}

Quarterly (n=4) {all other employees}

0.03c

0.03c

0.18

0.06

7/6/64 - 12/28/69 Quarterly (n=4) {all employees} 0.03c 0.06

12/29/69 - 12/30/73 Quarterly (n=4) {selected employees} 0.03c 0.06

12/31/73 - 6/29/75 Quarterly (n=4) {selected employees}

Semiannual (n=2) {unselected employees}

0.03c

0.03c

0.06

0.06

6/30/75 - 12/31/80 Quarterly (n=4) {selected employees}

Monthly (n=12) {selected female employees only}

0.03c

0.03c

0.06

0.18

PORTS Harshaw 2276 4-element TLD 

without window

1/1/81 - 12/31/82 Monthly (n=12) {all monitored}

Quarterly (n=4) {all monitored}

0.015d

0.015

0.09

0.03

PORTS Harshaw 2276, 8000, 8800 

4-element TLD with window

1/1/83 - 12/31/98 {1/1/93-12/31/96 for BJC 

employees}

Quarterly (n=4) 0.010e

(0.04 SDE)

0.04

(0.08 SDE)

ICN TLD 760 1/1/99 – present {USEC employees} Quarterly (n=4) 0.01f

(0.03 SDE)f

0.02

(0.06 SDE)

ORNL Panasonic 8805/8806 

4-element TLD with window

1/1/99 – present {BJC employees} Quarterly (n=4) 0.01h

(.03 SDE) h

0.02

(0.06 SDE)

Neutron dosimeters

PORTS TLD albedo dosimeter {USEC 

and BJC}

1/1/1992 – 12/31/94 {unmoderated Cf-252 

calibrated}

Quarterly (n=4) 0.02i 0.04

1/1/95-12/31/96 {moderated Cf-252 calibrated} Quarterly (n=4) 0.02i 0.04

ICN TLD 760 {USEC} 1/1/97 - present {moderated Cf-252 calibrated} Quarterly (n=4) 0.01f 0.02

Y-12 Panasonic TLND {BJC 

employees}

1/1/97 - 12/31/98 Quarterly (n=4) 0.01g 0.02

ORNL Panasonic TLND 8806 

4-element TLD {BJC employees}

1/1/1999 - present Quarterly (n=4) 0.01h 0.02

a. Maximum annual missed dose (NIOSH 2002). [For photon/beta missed dose = LOD/2 × n(frequency, p. 18), for neutron missed dose = LOD/2 × n(frequency), p. 29]

b. Kodak personnel type 2 film with gold sandwiched with cadmium for high-energy gamma, OW with aluminum for low-energy gamma and beta. LOD for SDE and DDE are the same, 0.03 rem, and the 

reporting level.

c. GAT Film badge procedure, May 5, 1963, reporting level for gamma and beta.

d. Personal communication (Wagner 2003)

e. Bassett (1986, p. 3)

f. ICN (2003).

g. Personal communication (Souleyrette 9/12/2003)

h. Personal communication (McMahon 9/17/2003)

i. GAT (undated) 
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Table 6-4. PORTS historic dosimetry events.

Date Description

9/22/54 PORTS Film 2-element system started 

8/1/60 Film badge combined with security badge (picture)

7/6/64-12/28/69 All employees monitored

1/1/1981 Harshaw 2276 TLD system/4-element enacted (without window), Sr-90 calibration; 

depleted uranium beta, radium gamma (1985). In addition, company photo was placed 

over beta chip during 1981 preventing beta (skin) determinations for 1981.

1/1/1983 Window added to TLD badge

4/20/1986 DOELAP accreditation process begins

1/1/97 Neutron monitoring begins with ICN TLD 760; beta/gamma monitoring continues 

1/1/97 BJC splits with USEC, utilizes Y-12 Panasonic TLND dosimetry 

1/1/99 USEC ends PORTS TLD program, utilizes ICN TLD 760 or equivalent

1/1/99 BJC utilizes ORNL Panasonic 8805 (beta/gamma)/8806 (neutron) 

See Table 6.3 for references. Based on GAT procedures and USEC/BJC procedures.

In the mid-1990s, an internal investigation indicated that PORTS did not keep dosimetry records in 

accordance with procedures, resulting in improper assignment of doses. From 1993 to 1995, some 

employee exposures were recorded as zero for exposed damaged TLDs. The HP group 

reconstructed these records, resulting in minor adjustments. All doses were much less than DOE or 

NRC limits (DOE 2002a, p. 37).

6.3.2.1 Beta/Photon Dosimeters

6.3.2.1.1 Film Dosimetry Two-Element, 1954-1980

The film dosimetry program began in 1954. The dosimeter description from the documentation 

obtained is cryptic. A Description of Co-Operative Work Assignments in Industrial Hygiene and 

Health Physics (Wooldridge 1964) describes the dosimeter as a “film badge with Kodak Type-2 

personal monitoring film combined with aluminum, cadmium, and gold filters for beta-gamma, low 

energy gamma, and high energy gamma radiation. There are also sulfur and gold filters for neutron 

exposures.” This document indicates a detection range from 30 to 2,000 mrem. None of the 

documentation indicated any analysis for neutron or extremity exposure during the film badge era 

(Wooldridge 1964, p. 3).

The film badge had two elements with an open window and a shield of gold sandwiched between 

cadmium to admit mostly high-energy gamma. To measure high-energy gamma deep dose 

equivalent (DDE), the shield should have been at least 1 mm thick (Wooldridge 1964, p. 5).

There is some description of film processing. Kodak Type 2 film had two emulsions. The first was 

fast and, when developed and monitored for density, would yield results for an exposure range from 

30 to 2,000 mrem. The second, which was a fast emulsion that would measure gamma radiation from 

5 to 300 roentgen (R), was used for an extended exposure range (Wooldridge 1964, p. 3).

Series of films, including a control for every 150 film badges, were developed for each batch. The 

control film was taken from the same emulsion as the series. Controls were used to determine the 

amount of natural film darkening from the normal wear cycle and to enable differences in emulsions 

and developing. 

A densitometer was used to measure the density of the control badge and then zeroed. The zeroed 

densitometer was used to read the badges in that series. If the measurement in the open window 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 14 of 55

was less than or equal to zero, the beta, gamma, and combined beta/gamma measurements were 

recorded as zero. In addition, total exposures less than 30 mR were recorded as zero. 

If the shielded portion of the film read zero, the cause of open window darkening was believed to be 

beta exposure, so gamma exposure was recorded as zero. A calibration graph generated twice a 

year was used to determine the beta dose. 

If the shielded portion of the film read the same as the open window, or within the ratio of about 1.3 

(open window to shielded portion) on the graph of the open window reading, all darkening was caused 

by gamma exposure. Beta exposure was recorded as zero. The densitometer reading from the 

shielded portion was found on the vertical axis of the graph and the corresponding gamma exposure 

was derived from the gamma or shielded labeled curve.

If the shielded portion of the film read greater than zero, but the ratio of the open window reading to 

the shielded portion reading was greater than the ratio on the calibration graph, the formula:

OW – S(R) = beta

was used to determine the darkening due to the gamma exposure, 

where

OW = open window densitometer reading

S = shielded portion densitometer reading

R = ratio given on the graph

The corresponding beta exposure was derived from the graph and recorded. The gamma exposure 

was obtained by using the shielded portion densitometer reading and the curve labeled gamma. Total 

beta/gamma exposure was obtained by adding these exposure readings (GAT Film Badge Procedure 

1964, 1971, p. 2-6). 

PORTS used emergency and equipment badges in potentially high radiation areas to determine the 

cause of high exposure readings in personnel badges because those badges remained in the same 

locations and were exposed to the same conditions for known periods. 

One way to determine the effectiveness of an external dosimetry system is to compare similar 

systems used in comparable environments. The International Agency for Research on Cancer (IARC) 

evaluated a two-element film dosimeter used at the Hanford Site to monitor several photon energies 

in exposure orientations as a combination of anterior-posterior (AP), rotational, and isotropic 

geometries (Thierry-Chef et al. 2002). As indicated in Table 6-5, the two-element dosimetry system 

overestimated the Hp(10) for most exposure orientations, especially for lower energies. The listed 

118-keV energy is the most applicable energy for PORTS exposure potential. (Table 6-16 lists 

radionuclide information.) 

Wilson et al. (1990) conducted another study on dosimetry at the Hanford Site. That study used only 

the AP orientation for low energies for film dosimeters (see Table 6-6). This is the only orientation 

that PORTS used for film dosimeter calibration. The lower energies of 16 and 59 keV representing 

plutonium photons are close to the 13, 30, 53, 63 and 68 keV representative of low-energy uranium, 

daughter and contaminant photon energies present at PORTS. There are also many intermediate 

energy photons most of which fall in the 30 to 250 keV energy bin. (See Table 6-16.)
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 15 of 55

Table 6-5. IARC testing results for U.S. beta/photon dosimetersb,c.

Geometry Phantom

118 keV 208 keV 662 keV

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

US-2 (Hanford two-element film dosimeter)

AP Slab 3.0 2.1 1.3 1 1.0 0.8

AP Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8

Rotational Anthropomorphic 2.2 2 1.4 3 1.2 3.2

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7

US-8 (Hanford multielement film dosimeter)

AP Slab 1.0 1.5 1.0 0.8 0.8 1.7

AP Anthropomorphic 0.8 9.5 0.9 6 0.8 1.8

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8

Isotropic Anthropomorphic 1.0 3 1.2 9 1.0 2.3

US-22 (SRS multielement thermoluminescent dosimeter)

AP Slab 0.9 4.4 0.9 3.9 0.9 3.5

AP Anthropomorphic 0.8 3.1 0.9 2.1 0.9 3.9

Rotational Anthropomorphic 1.1 3.1 1.2 1.5 1.0 4.1

Isotropic Anthropomorphic 0.9 0.3 1.0 2.5 0.9 1.6

a. Ratio of recorded dose to HP(10)

b. (Thierry-Chef et al. 2002).

c. (ORAU 2003a or ORAU 2003b)

Table 6-6. Testing results for Hanford two-element and multielement film dosimeters for energy and 

angular response.

a,b 

Beam

(energy, keV)

AP exposure Rotational exposure

Film dosimeters

TLD

1972–present

Film dosimeters

TLD 

1972–93

Two-element

1944–56

Multielement

1957–71

Two-element 

1944–56

Multielement

1957–71

16 0.1 0.9

59 0.5 1.1

M150 (70) 0.7 0.70 0.95 1.31 1.31 1.77

H150 (120) 1.6 0.64 0.87 3.00 1.20 1.64 

137Cs (662) 1.0 1.0 1.0 1.46 1.46 1.46

a. Divide recorded dose by table value to estimate Hp(10).

b. Based on Wilson et al (1990).

6.3.2.1.2 Thermoluminescent Dosimeters, 1981-present 

PORTS Four-Element TLD, January 1, 1981, to December 31, 1998

The TLD program began in 1981 under Goodyear Atomic Corporation. Most of the following 

information is from Thermoluminescent Dosimeter Response and Calibration (Bassett 1986). 

PORTS used the Harshaw Type L card with three TLD-700 chips (7

Li) and one TLD 600 chip (6Li). 

Generally, Chip 1 and 2 were used for skin dose, chip 3 for deep dose, and chip 4 for lens dose. 

Although not indicated in the calibration procedure, the TLDs were probably irradiated in the AP 

geometry. Table 6-7 lists the TLD shielding configurations.

Manufacturer performance specifications for the Harshaw Type L card for response are as follows: 10 

mrem at 90% confidence ±15% for 60Co and 40 mrad at 90% confidence ±20% for natural uranium 

beta particles. 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 16 of 55

A TLD system similar to that used at the Hanford or Savannah River Site (SRS) (Panasonic 802D 

four-element) was evaluated. The TLD was irradiated by several photon energies in exposure 

orientations as a combination of AP, rotational, and isotropic geometries). As indicated in Table 6-5, 

Table 6-7. Harshaw Type L card element and filter description.

a

Chip 

position TLD type Shield density thickness (mg/cm2)

Total density 

thickness (mg/cm2)

1 TLD-700 Mylar/8, Teflon/8 16

2 TLD-700 Laminated Photo/75, Polyethylene/84, Teflon/8 167

3 TLD-600 Laminated Photo/75 Polyethylene/52 Cadmium/790, 

Gold/245, Tape and Teflon/15

1,177

4 TLD-700 Laminated Photo/75 Polyethylene/65 Aluminum/281, Tape 

and Teflon/15

436

a. (Bassett 1986, p. 3).

the TLD dosimetry system closely estimated Hp(10) for all exposure orientations within ± 20% of the 

expected exposure. The 118-keV listed energy (AP and rotational orientation) is the most applicable 

energy for PORTS exposure potential. A similar French Harshaw TLD with a plastic filter of 1,000 

mg/cm2 was within ± 10% of the U.S. Panasonic TLD. Although the Panasonic system might have 

different filters, the overall response was similar (Thierry-Chef et al. 2002, Table 2, p. 106).

ORNL (Y-12) Four-Element TLD, January 1999 to present – BJC Employees

PORTS used TLDs in a few configurations, including this configuration with a Panasonic 8805/8806 

four-element card. It was designed to monitor beta, photon, and neutron radiation (McMahon 2003).

Commercial ICN Four-Element TLD System, January 1999–Present - USEC Employees

PORTS implemented a commercial ICN TLD system on January 1, 1999. This system includes a

four-chip beta/photon dosimeter and a separate neutron dosimeter. Technical characteristics are 

described at the ICN Internet site (www.ICN.com). This dosimetry is NVLAP-accredited.

6.3.2.2 Neutron Dosimeters

PORTS Four-Element TLD, January 1, 1992, to December 31, 1998

PORTS used the Harshaw Type L card with three TLD-700 chips (7

Li) and one TLD-600 chip (6Li) 

(thermal neutron sensitive) with a 235-mg/cm2 density thickness. The Plant performed neutron 

monitoring with this dosimeter but only of selected groups. From January 1, 1992, to December 31, 

1994, the PORTS thermoluminescent neutron dosimeter (TLND) system was calibrated against an 

unmoderated 252Cf neutron source. From January 1, 1995, to December 31, 1996, a moderated 252Cf 

source was used. The change was in response to a perceived over-response of area neutron 

dosimeters. 

ORNL (Y-12) Four-Element TLD, January 1999 to present – BJC Employees

PORTS used TLDs in a few configurations. This configuration has a Panasonic 8805/8806 fourelement card. It was designed to monitor beta, photon, and neutron radiation (McMahon 2003).

Commercial ICN Four-Element TLD System, January 1999–Present - USEC Employees

PORTS implemented a commercial ICN TLD system on January 1, 1999. This system includes a 

four-chip beta/photon dosimeter and a separate neutron dosimeter. Technical characteristics are 

described at the ICN Internet site (www.ICN.com). This dosimetry is NVLAP-accredited.

Despite changes in neutron dosimetry systems at PORTS, neutron sensitivities are essentially the 

same. 
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Neutron Surveys

Neutron dosimetry was not used at PORTS until January 1, 1992. In 1996, at the request of several 

union representatives, NIOSH prepared a hazard evaluation report (Cardarelli 1997). This report 

referred to a survey of neutron radiation levels around 5-in. cylinders of highly enriched uranium 

(HEU) in storage in 1985. This survey resulted in measurements of 3 mrem/hr at the surface and 

0.5 mrem/hr at 1 meter. Lower enrichment 10-ton storage cylinders produced radiation levels of 

0.5 mrem/hr neutron dose equivalent (DE) rate at the surface. Based on this information, a worker 

who spent about 3 hours a week exposed to 0.5 mrem/hr would receive 75 mrem of neutron dose per 

year. Because this is 1.5% of the regulatory limit of 5,000 mrem, the HP group deemed neutron 

personnel monitoring to be unnecessary (Cardarelli 1997). [Cylinder lot surveys measured an 

average gamma DE rate of about 22 mrem/hr at the surface and 3 to 12 mrem/hr at 1 meter (Cylinder 

lot-745D and Cylinder lot-745G)] 

Actual personnel exposures monitored from November 1996 to February 1997 indicated no 

measurable neutron dose above the 20-mrem/quarter LOD or MDL for the dosimeter used (Table 6-

8). This indicates possibly 80 mrem of neutron dose missed per year, a value close to that estimated 

by surveys conducted about 10 years earlier, as mentioned above.

Table 6-8. Personal neutron dosimetry results November 1996 – February 1997a

.

Job title Department Buildings Comments Doseb (mrem)

Process operator 720, 730, 

740

X-333, X-330, 

X-326

Cold recovery, tails, product, low-assay 

withdrawal, unit operator, extended 

range product station

< 20 mrem

Security guard 151, 152 X-326, X-705, 

X-345 

Product withdrawal, rotation P-12 < 20 mrem

Chemical operator 721, 771, 

791

X-344, X-345, 

X-744G, X-326 

Recovery, cylinder lots, small parts, 

tunnel

< 20 mrem

Health physics technician 300 X-342, X-343, 

X-344, X-705 

HP coverage <20 mrem

Laborer 147 Cylinder yards Paint and scrap in yards (20 hr/wk) < 20 mrem

Uranium material handler 791 X-344, X-745C, 

X-745E, X744G, X-326

Cylinder lots, L-cage, warehouse 

storage, autoclave, shipping and 

receiving, vault

< 20 mrem

a. Cardarelli (1997)

b. Note: LOD or MDL = 20 mrem

The quality factors (QFs) used historically for neutrons have changed significantly. In current 

regulations, QFs that are used to convert radiation dose (mrad) to dose equivalent (mrem) are based 

on ICRP Publication 38 (ICRP 1983). The most current QFs from ICRP (1991) are about 2 times 

higher than the ICRP (1983) values. Because a QF of 10 was used for the referenced radiation 

measurements, the personnel dose results (Table 6-8), and PORTS personnel dosimetry, an 

adjustment to ICRP (1991) of at most a factor of 2 times higher would be necessary.

Average neutron energy is about < 1 MeV, 510 keV for 2% 235U, 770 keV for 5% 235U, and 860 keV for 

97% 235U (Cardarelli 1997, p. 9). QF equals 10 for ICRP (1983), or about 20 for the ICRP (1991) 

revision. The average neutrons from depleted and natural uranium cylinders ranged from 210 to 

360 keV (Cardarelli 1997, p. 9). Unmoderated and deuterium (water) 252Cf neutrons created were 

between 1,403 and 1,306 keV. This means the dose as monitored at PORTS since 1992 (for 

calibration facility personnel) and 1994 (others included) was overestimated and, therefore, claimantfriendly.

PORTS determined from a study of dosimeters in the eight work areas that neutron dose would be 

12.5% of the photon dose equivalent (Cardarelli 1997, p. 8). A Hanford study yielded a 26% average 

neutron-to-photon dose (reactor average – Fix, Wilson, and Baumgartner 1996). SRS experience 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 18 of 55

indicates somewhat less, a 10% neutron-to-photon dose (reactor average) (ORAU 2003a, Section 

5.3.4.2.3.2). From the Guide of Good Practices for Occupational Radiological Protection in Uranium 

Facilities (DOE 2000, p. 2-19), a 0.01- to 0.2-mrem/hr neutron DE rate for natural to 5% enriched 

cylinders in cold storage is likely. A neutron DE rate of 2 to 5 mrem/hr at contact and 1 to 2 mrem/hr 

at 3 feet is likely for 97% enriched cylinders. Table 6-9 lists neutron dosimetry results for the work 

areas.

Table 6-9. Area neutron dosimetry results November 1996 – February 1997a

.

Buildings Locations/comments Dose (mrem)

X-326 Extended range product (ERP) {Dynamics St. 2)

Product withdrawal (PW) {T-57-8-3 Bed 3}

80 mrem (3 mos)

60 mrem (3 mos)

X-330 Tails

Deposit;dd-4;29AB-1

G-33;29-3-7-7 {randomly selected}

Low-assay withdrawal (LAW) 

< 20 mremb

< 20 mremb

< 20 mremb

< 20 mremb

X-343 Typical movement (300 cylinders/month) < 20 mremb

X-345 Behind Phantom 5 on wall < 20 mremb

X-705 Portal 

DOE Lot 11,200;2-3%, row 22-23 Sec 44

< 20 mremb

420 mremb

210 mremb

710 mremb

X-745G DOE Lot; 11,200;2-3%, row 20-21 Sec 3, heel 210 mremb

510 mremb

320 mremb

a. Cardarelli (1997), pp. 21 and 22.

b. Measurement period of 1 month

From November 1996 through February 1997, specific process areas and personnel were monitored, 

using the Landauer/Neutrak ER badge (Landauer 2003). This badge combines an albedo dosimeter 

with a CR-39 chip, which monitors neutrons of > 30-50 keV to 35 meV. The TLD albedo chip monitors 

neutrons in the energy range of 0.5 to 100 keV. Thermal neutrons were not monitored due to the 

assumption that most are intermediate to fast neutrons (Cardarelli 1997) 

6.3.3

Potential error in recorded dose is dependent on dosimetry technology response characteristics to 

each radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; 

and the similarity between the radiation fields used for calibration and that in the workplace. 

Calibration

6.3.3.1 Film Badges Beta/Photon

For PORTS film dosimeter calibration, eight badges were exposed to a 22.5-mCi 226Ra source AP at 

various distances free in air to yield 0 (control), 30, 60, 100, 200, 500, 750, and 1,000 mR of gamma 

exposure (GAT 1964). To correct for this free-in-air exposure [a conversion from roentgen (87.6 

ergs/gram) to rad (95 ergs/gram)], ratios of the differences of rad to roentgen were taken, as follows: 

{87.6 – 95}/87.6 = 8.45 % (6-1)

In other words, if the badges were calibrated with an anthropomorphic-type phantom, they would have 

received 95 ergs/gram (plus scatter) rather than the 87 ergs/gram they received. Assuming a QF of 1 

for rad-to-rem conversion, the correction factor due to not using a phantom would be -8.45% relative 

to using a phantom. 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 19 of 55

Because 226Ra or 137Cs was used for calibration, a 8.0 % or 3.9% under-response, respectively , due 

to higher energy calibration energies would result, compared to the actual lower exposure energies 

encountered in the workplace (ORAU 2003a). A correction of 16.5% for film dosimetry (1954-1980), 

12.5 % for TLD dosimetry (1981-1986) and 4 % for TLD dosimetry (1987-present) is recommended.

Figure 6-1 shows a comparison of a two-element film dosimetry system to the other film dosimetry 

systems. Wilson et al. (1990) measured the AP photon energy response of the Hanford systems. As 

indicated in Figure 6-1, the dosimeter open-window response shows a significant over-response to 

lower energy photon radiation. Most of the photon energy spectrum and the dose equivalent will be 

from the 30 to 250 keV range at PORTS where there was an over-response from the 2 element film 

shielded portion of the dosimeter (silver shield). This indicates that recorded results for the PORTS 

two-element film would overall be claimant-favorable and no corrections are needed for the response 

of the dosimetry to the radiation work environment. The under-response of the 2 element film shielded 

portion to low energy of less than 50 kev should be of little consequence since the vast majority of the 

photon energy spectrum and dose equivalent will be from photons greater than 50 kev. This is due to 

the fact that most radiation work environments involve shielded or self shielded uranium sources 

which would allow little exposure from < 30 kev photons. In the case of open systems such as 

processing, recovery, and maintenance the exposure to low energy photons is more probable. In 

these situations, since the low energy photons will be monitored conservatively by the open window it 

may be claimant favorable to equate the shallow dose to the deep dose if the deep dose was found 

less than the shallow dose in the claimants’ records. 

Figure 6-1. Measured Hanford two-element dosimeter photon response 

characteristics (Wilson et al. 1990).

The film nonpenetrating (i.e., open window) and penetrating (i.e., gold-cadmium filter) response was 

used to estimate skin dose from beta and photon radiation.

For beta calibration, six badges were exposed to depleted uranium slugs (240 mR/hr) in an AP 

geometry at various times to yield 0, 60-, 120-, 240-, 480-, and 960-mR beta radiation exposure. The 

slugs were placed directly on the badges (GAT 1964). This meant that the calibration closely matched 

the work environment and no corrections for beta response is necessary. 
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6.3.3.2 TLD Badges Beta/Gamma

TLD badges were calibrated in free air with a 137Cs source, NBS filtered X-rays, and depleted uranium 

(slab geometry). Because different energies were used for calibration, exposure to DE factors was 

used in the calibration algorithms, as listed in Table 6-10. 

Table 6-10. Roentgen-to-rem conversion factors used for TLD 

dosimeter calibrationa

.

Conversion 

factors (rem/R)

shallow (0.007 cm)

Conversion 

factors (rem/R)

deep(1.0 cm)

NBS filtered X-rays:

M30 (20 keV) 1.08 0.45

S60 (36 keV) 1.15 1.07

M150 (70 keV) 1.41 1.47

H150 (120 keV) 1.41 1.41

Cs-137 662 ---- (b) 1.03

a. Bassett (1986, p. 5)

b. This value, which is dependent on source geometry, should be measured

Controls were placed at the plant entrance portals – drive gate, X-100 lobby, X-108B gate, H-lot gate, 

and C-lot gate. About 16 combinations of sources were used, such as 137Cs and 20-, 36-, 70-, and 

120-keV X-rays and U-slab; U-slab and 20-, 36-, 70-, 120-keV X-rays as representative of possible 

mixed fields encountered at PORTS. 

The general form of the DDE equations used in the TLD calibration system was:

DDE = Do + D1(CR3) + D2(CR4) (6-2)

SDE = S0 + S1(CR1) + S2(CR2) + S3(CR3) + S4(CR4) (6-3)

where:

Di and Si are the deep and shallow multiple linear regression coefficients for their respective chip 

positions, and CRi are the nanocoulomb TL chip responses for chips 1 through 4. 

TLD chips 1 and 2 (TLD 700, 16 and 167 mg/cm2

, respectively) are not used for DDE determination. 

The basis for dose algorithm calculations is chip proportionalities, which enables discrimination of 

radiation energy types to unknown fields (Basset 1986, pp. 22-23).

Table 6-11 summarizes laboratory sources of uncertainty parameters in beta and photon calibrations. 

If uncertainty or bias is positive, make no change. In the case of negative bias, make the appropriate 

corrections using the values in Table 6-11. Uncertainty was estimated using PORTS procedure 

information or dosimetry performance information from similar dosimetry systems. 

6.3.3.3 TLD Badges Albedo Neutron

In 1992, some workers in the PORTS Radiation Calibration Facility and the Applied Nuclear 

Technology Department were placed on a routine neutron dose monitoring program. In October 

1994, only workers entering a Controlled or Restricted Area were monitored. From 1992 through 

1994, the albedo dosimeter was calibrated with an unmoderated 252Cf source resulting in higher doses 

than expected. In 1995, the calibration procedure was modified to utilize a moderated 252Cf source 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 21 of 55

that would result in lower doses than expected due to the fast neutron environs of PORTS (210 – 860 

keV neutrons) (Cardarelli 1997).

The use of the ICN TLND, the Y-12 TLND, and the ORNL TLND from 1997 to the present should not 

present much variation because all are either NVLAP- or DOELAP-certified, and all use phantoms for 

calibration and similar geometries. ICN is a large commercial vendor 

Table 6-11. Laboratory sources of uncertainty for beta/photon dosimeter calibration parameters.

Parameter Historical description Uncertaintya Comment

In-air calibration Prior to 1/1/1987, calibrations were 

performed free in air. In 1997 for BJC 

and 1999 for USEC, use of commercial 

or outside vendors for dosimetry began. 

Phantom for calibration introducing 

increase of response from backscatter 

occurred relative to dosimetry used 

before 1997 and 1999. 

±10% After 1997, recorded dose of record too 

high. Backscatter radiation from worker’s 

body is highly dependent on dosimeter 

design (ORAU 2003b, Table 6-2). Before 

1987, recorded dose of record too low. 

(See section 6.3.3.1.)

Radiation 

quantity

Before 1981, PORTS used Cs-137 and 

Ra-226 for beta and photon beam 

calibration.

±5% For higher energy Cs-137 and Ra-226 for 

beta and photon beam calibration, this 

caused about 4% or 8% under-response 

in recorded dose (Basset 1986).

Tissue depth of 

dose

Historically, PORTS used specified 

depth of 10 mm to estimate deep dose. 

±5% Numerical effect of this for photon 

radiation is comparatively low. PORTS 

dosimeter designs had filtration density 

thickness of about 1,000 mg/cm2 that 

would relate closely to 1-cm depth in 

tissue. 

Angular 

response

PORTS dosimeter system is calibrated 

using AP laboratory irradiations.

~ -25%

100 keV

Recorded dose of record likely too low 

because dosimeter response is usually 

lower at non-AP angles. Effect is highly 

dependent on radiation type and energy 

(Table 6-13 and 6-16)

Environmental 

effects

Workplace heat, humidity such as 

dosimeter fading impact dosimeter 

results.

±5% Fading should have been less than 2% for 

TLDs. Heat effect should have been 

much less than 1%. Recorded dose due 

to these effects likely too low.

a. Uncertainty estimate in recorded dose compared to Hp(10) based on judgment from Hanford dosimeter laboratory studies. 

with references to its dosimetry program available at its Internet site (www.ICN.com). Table 6-12 lists 

possible dosimeter lab parameter uncertainty.

Table 6-12. Laboratory sources of uncertainty for neutron dosimeter calibration parameters.

Parameter Historical description Uncertaintya Comment

Source energy 

spectra

PORTS used unmoderated and 

moderated Cf-252 source for 

calibration. Unmoderated source led to 

increased bias.

±100% Delivered dose used in calibrating 

neutron dosimeters is uncertain. 

Uncertainty listed is claimantfavorable.

Radiation 

quantity

QF or spectrum used for PORTS by 

outside vendors. 

±50% This represents significant and 

complicated issue. 

Angular 

response

PORTS TLND dosimeters calibrated 

using AP laboratory irradiations.

±50% Recorded dose of record likely 

too high because dosimeter 
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response is often higher at angles 

other than AP. Effect is highly 

dependent on energy. 

Environmental 

stability

TLD systems are subject to signal fade 

with time, heat, humidity, light, etc.

±50% Recorded dose of record likely 

too low because of fading; 

however, this effect depends 

strongly on such routine 

dosimetry practices as when 

calibration dosimeters were 

irradiated. 

a. Uncertainty in recorded dose compared to Hp(10) based on judgment from laboratory studies based on ORAU (2003b, 

Table 6-3). 

6.3.4

The PORTS radiation work environment is comprised of a variety of combinations of complex beta, 

gamma (photon), X-ray, neutron, and bremsstrahlung (X-ray) radiations. The worker’s job, location, 

and occurrences of incidents would affect the external exposure acquired. Worker location in relation 

to radiation sources during the course of the work history is one of the most challenging parameters to 

establish, as is estimation of actual radiation source levels. Assumptions related to these parameter 

definitions need to be based on available worker information and the work environment. The basis of 

the assumptions will be given. For the PORTS areas listed in Table 6-13, claimant-favorable 

assumptions were used. As suggested in NIOSH (2002), 30 to 250 keV for photons is used if a 

specific gamma spectrum is not known. Several exceptions occur in Building X-330, where a large 

amount of tails that would contain 238U in higher concentrations might be present. Because the 

daughter products of 238U, 234mPa and 234Th in equilibrium with 238U, have photon energies greater 

than 250 keV, the energy bin of 30 to 250 keV was allocated 85% photon field percentage, and > 250 

keV was allocated 15% photon field percentage. Buildings X-710 and X-720 have calibration source 

usage, 137Cs and 226Ra, which yield more than 250 keV photons. Therefore, 25% of the photon field 

percentage was allocated to these areas. This is claimant-favorable because exposure from or usage 

in areas other than calibration facilities or structural analysis areas is limited. Bremsstrahlung could 

be present in the ambient work area but would be of a lower level than the ambient gamma radiation.

Workplace Radiation Fields

The energy bin selection for beta radiation for every area at PORTS is greater than 15 keV. Of 

special note for beta areas is X-330 or any area of tails with potential for exposure during processing. 

Technetium-99 exposure could have occurred during cascade maintenance, removal of the 

magnesium traps, and waste processing. 

PORTS processed recycled uranium (RU), which contained trace amounts of radioactive impurities 

not present in natural uranium feed material. Because these impurities were present at such minute 

concentrations, their radiological impact was usually negligible. However, some routine chemical 

processes would concentrate them. The most significant impurity found in RU is the pure beta 

emitter, 99Tc, which tends to deposit in enrichment equipment and “pocket” in the higher sections of 

the diffusion cascade (DOE 2000a). In addition, 99Tc was concentrated for recovery and removal. 

The relatively low-energy beta particles (maximum 294 keV) from 99Tc pose minimal external 

exposure potential because of their limited range. Neither film nor TLD efficiently detect them, 

particularly in the presence of uranium. Clothing and gloves provide adequate shielding. Skin 

contamination is the most credible scenario in which a significant shallow dose could occur from 99Tc. 

Table 6-14 lists the principal locations and periods for which recovery operations at PORTS are 

believed to have occurred (DOE 2000b).
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Table 6-15 summarizes the reactor returns (RU). (See also Section 2 and Table 2.4.2-1 of this TBD). 

RU comprised about 1,094 MTU of the 330,000 MTU fed to the PORTS cascade (BJC 2000, p. 22). 

6.3.4.1 Workplace Beta/Photon Dosimeter Response

All PORTS radiological work areas involve beta/photon radiation covering a wide range of energies, 

which are characterized by radionuclides present in the work environment. Table 6-16 lists properties 

of the radionuclides and machine sources at PORTS. Information on the actual radiation environment 

can be reviewed in the Site Description TBD (ORAU 2004) and Attachment A of this PORTS Site 

Profile. 

Table 6-13. PORTS workplace beta/photon dosimeter response.

Process/

buildings Description

Operations Radiation

type

Energy 

Begin End selection %

X-326 High assay withdrawal station (HASA) {HEU} 1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

Purge system (top and side) {EU – HEU} 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X – 330 Tails withdrawal station (DU) 1954 2001 Betaa

Photon

> 15 keV

30 – 250 keV

>250 keVb

100

85

15

X-333 Reactor grade LAW (EU) [when used as backup for 

tails withdrawal; see X-330]

1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-342, 

342A, & 

343

Fluorine generation (342) and fixed feed (EU) 1954 2001 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-344 Feed manufacturing plant 1958

1962

1962

2003

Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-345 Special Nuclear Material (SNM) storage and HASA 1978 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-700 Maintenance 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-705c,d & 

705Ec Decontamination, cleaning and recovery, oxide conversion plant

1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

100

100

X-710e Analytical labs 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

>250 keV

100

75

25

X-720d Compressor shop 1954 2003 Beta

Photon

> 15 keV

30 – 250 keV

> 250 keV

100

75

25

X-744 Smelter for aluminum recovery 1961 1983 Beta

Photon

> 15 keV

30 – 250 keV

100

100

a. Technetium-99 is expected, especially from 1977 – 1980s.

b. Uranium-238 in equilibrium with daughters 234mPa and 234Th.

c. Expect all RNs including TRU materials and 99Tc. 

d. Beta exposures are more probable with the treatment of wastes and opening of equipment.

e. Calibration sources such as 137Cs and 226Ra and X-ray equipment from 40 to 200 Kv have been used in parts of this facility.
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 24 of 55

Table 6-14. Major facilities at PORTS where 99Tc might have accumulated.

Building 

No. Name

Dates of 

operation Activities

X-326 Gaseous Diffusion Process Bldg. 1954–1991 High Assay Product 

X-330 Gaseous Diffusion Process Bldg. 1954–2001 Intermediate process & tails withdrawal

X-333 Gaseous Diffusion Process Bldg. 1954–2001 Initial enrichment & reactor product

X-344 UF6 Feed Manufacturing Plant 1958–1962 Conversion of UF4 to UF6

X-345 Special Nuclear Material Storage 1978–2003 HEU storage

X-700 Maintenance Building 1954–2003 Large component repairs

X-705 Decontamination & Cleaning Bldg. 1954–2003 Equipment wash & uranium recovery

X-705E Oxide Conversion Plant 1957–1978 Conversion of U3O8 to UF6

X-720 Compressor Shop 1954–2003 Disassembly & repair of compressors

X-744G Smelter & Aluminum Recovery 1954–1978 Recover aluminum from scrap

Source: DOE (2000b), p. 16.

Table 6-17 summarizes the common beta/photon personnel dosimeter parameters important to 

Hp(10) performance in the workplace. PORTS dosimetry has made reasonable measurements of 

workplace radiation fields. The two-element dosimeter would have over-responded to the average 

photon field of about 100 keV by about 50% or more. No corrections to over-response are 

recommended. As of 1987, PORTS estimated shallow dose by adding the gamma (deep) and beta 

(shallow) doses together. This policy would be an overestimate of shallow dose. Extremity

Table 6-15. Reactor returns fed to cascade.

Fiscal year

Amount 

fed (MTU)

Enrichment 

(% U-235) Source Remarks

1955 105.8 0.64 – 0.68 Paducah Fed May – Sept. 1955

1956 54.5 0.64 – 0.68 Paducah

1956 293.4 0.64 – 0.68 Oak Ridge

1957 6.2 0.64 – 0.68 Paducah

1958 64.2 0.64 – 0.68 Paducah

1970 168.1 0.64 – 0.68 Paducah Fed Oct. & Nov. 1969

1974 398.8 0.64 – 0.68 Paducah Fed Jan. 1974

1974 – 1978 1.86 2 – 50 PORTS Oxide Conversion

1968 – 1977 0.15 78 – 80 Division of International Affairs

1977 – 1998 0.15 78 – 97 Babcock & Wilcox

1969 – 1993 0.07 78 AEC Office of Safeguards & Materials 

Management

1997 – 1998 1.10 56 – 82 France

1997 – 1998 0.33 80 NUMEC

TOTAL 1,094.66

Source: Table 2.2.2.5-1, BJC (2000, p. 22).

exposures were monitored infrequently. BJC has not monitored for extremity exposure.

6.3.4.2 PORTS Workplace Neutron Response

In general, PORTS radiation workers were exposed to ambient neutron radiation produced primarily 

from three reactions – spontaneous fission of 235U, subcritical fission of 235U, an alpha reaction on 

fluorine from the decay of uranium [19F(α,n) 22Na] and an alpha reaction on oxygen from the decay of 

uranium [18O(α,n)21Ne]. The most likely places for neutron exposures are in storage areas or cylinder 

yards (X-345, cylinder lots 745), feed and withdrawal process areas (X-326, 330, and 333), calibration 

and laboratory assay areas where 252Cf sources were used, and areas where uranium deposits 

formed in the cascades. 
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Two of these areas have been characterized for neutron energy – the assay laboratory in Building 710 

and the storage vault for HEU in Building 345 (Soldat and Tanner 1992). The 252Cf calibration facility 

in Building 710 energy characterizations will be based on measurements of a similar facility.

One phenomenon that occurs at GDPs is the formation of “slow cookers,” which are uranium deposits 

that can accumulate in the cascade. At a fission rate just below critical, a slight increase of neutron 

production occurs. All of these processes produce fast neutrons with energies usually less than 

2 meV. 

6.3.4.2.1 Calibration Laboratory in Building 710

Measurements were made of the bare 252Cf calibration source in the low scatter room, where it was 

used with a Tissue Equivalent Proportional Counter (TEPC) at 1 meter. On the day of the 

measurement (March 12, 1992), the source was rated at 27.54 μgm, corresponding to a dose 

equivalent rate of 64.9 mrem/hr at 1 meter. The TEPC measurements ranged from 48 to 

57 mrem/hour (Soldat and Tanner 1992, p. 3.3). 

Pacific Northwest Laboratory (PNL) measured the ORNL bare 252Cf calibration source for a neutron 

energy spectrum. The results are indicated by the solid lines in Figure 6-2. Table 6-18 lists the dose 

fractions for the neutron energy groups indicated by the dashed lines in Figure 6-2. The dose 
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Table 6-16. Properties of radionuclides that might be found at uranium facilities.

a

Nuclide Half-life

Energies (MeV) and abundances 

of major radiations

Alpha Beta Gamma

Primary uranium isotopes

U-238 4.51 × 109 yr 4.15 (25%) -- 0.013 (8.8%)

4.20 (75%)

U-235 7.1 × 108 yr 4.37 (18%) -- 0.144 (11%)

4.40 (57%) 0.185 (54%)

0.013 (31%)

U-234 2.47 × 105 yr 4.58 ( 8%) -- 0.204 (5%)

4.72 (28%) 0.053 (0.20%)

4.77 (72%) 0.013 (10%)

U-236 2.34 × 107 yr 4.49 (76%) 0.013 (10%)

4.44 (24%)

U-232 72 yr 5.26 (31%) 0.013 (12%)

5.32 (69%)

Decay Products

Th-234 (U-238 parent) 24.1 d -- 0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.093 (4%)

Pa-234m (U-238 parent) 1.17 m 2.29 (98%) 0.765 (0.30%)

1.001 (0.60%)

Th-231 (U-235 parent) 25.5 hr 0.140 (45%) 0.026 (2%)

0.220 (15%) 0.084 (10%)

0.305 (40%)

Th-230 (U-234 parent) 77,000 yr 4.62 (23.4%) 0.068 (0.4%)

4.69 (76.3%) 0.012 (8.4%)

Th-228 (U-232 parent) 1.913 yr 5.34 (26.7%) 0.012 (9.6%)

5.42 (72.7%) 0.084 (1.2%)

Impurities

Tc-99 2.12 × 105 yr -- 0.294 --

Np-237 2.14 × 106 yr 4.78 (75%) 0.030 (14%)

4.65 (12%) 0.086 (14%)

0.145 ( 1%)

Pu-238 86.4 yr 5.50 (72%)

5.46 (28%)

Pu-239 24.4 × 104 yr 5.16 (73%)

5.14 (15%)

5.10 (12%)

Pu-240 6.6 × 103 yr 5.17 (76%)

5.12 (24%)

Pu-241 13.2 yr 0.021

Cs-137/calibration sources 30.6 yr 0.514 (95%) .662 (85%)

Ra-226 1,600 yr 4.60 (6 %) .186 (3.6 %)

4.78 (94%) 

Machine generated X-rays 0.07 – 0.2

a. In most part from Bassett (1986).

fractions for the lower (< 10 keV) and intermediate (10 to 100 keV) energy neutron groups were less 

than 1% of the total dose from the measurements. Thus, combining the lower and intermediate 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 27 of 55

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

calculation.

The use of the ORNL 252Cf calibration source is reasonable because the energy spectra would be the 

same. However, neutron distribution would differ from calibration facility to facility based on scatter of 

the neutrons or room rate return. This difference should be minimal because calibration facilities are 

Table 6-17. Common workplace beta/photon dosimeter Hp(10) performance.

a

 

Parameter Description Uncertaintyc Workplace biasb

Exposure 

geometry

PORTS dosimeter system 

calibrated AP lab irradiations. 

Workplace exposure 

geometries are highly variable.

> 100 keV: 

Two–element film 

dosimeter, ~ 

+200%

Others, ±25%

< 100 keV: 

Probably too low.

Recorded dose probably too low because 

dosimeter response can be much lower at ROT 

and LAT angles. This effect is highly energy 

dependent. Highest doses are probably 

associated with AP geometry where work is 

performed close to radiation source. Effect is 

highly dependent on radiation energy.

Energy 

response

PORTS film dosimeter 

response to photon radiation < 

100 keV too low and > 100 

keV was too high.

Depends on actual 

field

100 keV ~ - 25%

Bias in recorded dose depends on photon 

spectrum in workplace, especially for film 

dosimeter. Reasonable estimate of Hp(10) dose is 

likely (Figure 6-1) (Wilson et al. 1990). Estimate 

based on 100 keV and Table 6-6.

Mixed fields PORTS dosimeters responded 

to both beta and photon 

radiation. TLD PORTS system 

was calibrated against number 

of different possible mixed 

fields.

Depends on actual 

field

About 16 types of mixed fields were calibrated 

under PORTS TLD system. PORTS film system 

was not calibrated against mixed fields. 

Reasonable estimate of Hp(10) dose is likely.

Missed dose Doses less than MDL or LOD 

are recorded as zero dose.

Recorded dose of 

record probably too 

low.

PORTS recorded doses < MDL for all years. Issue 

is significant, primarily in earlier years with frequent 

dosimeter exchange and film dosimeters with 

higher MDLs. 

Environmental 

effects

Workplace heat, humidity such 

as dosimeter fading impact 

dosimeter results.

±5% Fading should have been less than 2% for TLDs. 

Heat effect should have been much less than 1%. 

Recorded dose due to these effects probably too 

low.

a. Judgments based on PORTS dosimeter response characteristics and workplace radiation fields.

b. Recorded dose compared to Hp(10) 

c. Uncertainty based on recorded dose compared to Hp(10) on judgment on similar dosimeter laboratory studies and onsite procedures.

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

Neutron Energy (MeV)

10|-7

10|-6

10|-5

10|-4

10|-3

10|-2

10|-1

Dose Equivlanet (mrem)

9.75e-5 mrem 1.21e-4 mrem

7.59e-3 mrem

2.61e-2 mrem
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Figure 6-2. Results of neutron spectrum measurements 

made at 1 meter from bare 252Cf fission neutron source 

(Soldat and Tanner 1992). (See Y-12 TBD, Fig. 6.3.4.2-3)

Table 6-18. Dose fractions for PORTS calibration facility.

Neutron energy group Near unshielded Cf-252 source

< 10 keV 0.003

10-100 keV 0.004

0.1-2 MeV 0.224

2-20 MeV 0.769

Claimant-favorable dose fractions

0.1-2 MeV 0.23

2-20 MeV 0.77

designed to minimize scatter and maintain consistency. This difference would probably not be more 

than 20% (NCRP 1991).

6.3.4.2.2 Nondestructive Assay Laboratory in Building 710

The assay laboratory contained a californium shuffler unit and a segmented gamma scanner 

separated by a concrete wall that extended halfway down the room. Measurements of the 252Cf 

shuffler unit in its fixed open position yielded a dose equivalent rate of 0.2 mrem/hr and an integrated 

dose on environmental dosimeters on a phantom of about 1 mrem. The multisphere measurement at 

the same location as the phantom resulted in an average neutron energy of 0.52 MeV and a dose 

equivalent rate of 0.15 mrem/hr. 

The solid lines in Figure 6-3 show the calculated energy spectrum from the multisphere detectors 

(Bonner Spheres). Table 6-19 lists the dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-3). The dose fractions for the lower (< 10 keV) and intermediate (10 – 100 

keV) energy neutron groups were about 20% of the total dose from the measurements. Combining 

the lower and intermediate energy groups into the 0.1 – 2 MeV group is a reasonable simplification of 

the neutron dose calculation.
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Figure 6-3. Results of neutron spectrum measurements 

made in the front of shuffler unit (Soldat and Tanner 1992). 

Table 6-19. Dose fractions for PORTS nondestructive 

assay laboratory facility in Building 710.

Neutron energy group Near unshielded Cf-252 source

< 10 keV 0.034

10-100 keV 0.153

0.1-2 MeV 0.598

2-20 MeV 0.215

Claimant-favorable dose fractions

0.01-2 MeV 0.785

2-20 MeV 0.215

6.3.4.2.3 HEU Storage Vault in Building 345

Cylinders of highly enriched (93% to 96%) uranium were measured with a TEPC mounted on a 

phantom about 24 in. from the cylinders. The dose equivalent from the cylinders was about 

0.8 mrem/hr with a total dose equivalent of 14 mrem. The multisphere measurement at the same 

location as the phantom resulted in an average neutron energy of 0.53 MeV and a dose equivalent 

rate of 0.5 mrem/hr. 

The solid lines in Figure 6-4 show the calculated energy spectrum from the multisphere detectors 

(Bonner Spheres). Table 6-20 lists the dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-4). The dose fractions for the lower (< 10 keV) and intermediate (10 –100 

keV) energy neutron groups were about 47 % of the total dose from the measurements. 

Figure 6-4. Results of neutron spectrum measurements 

made about 24 in. in front of 93 – 96% HEU cylinders

(Soldat and Tanner 1992).

Table 6-20. Dose fractions for PORTS HEU storage vault in 

Building 345.

Neutron energy group Near unshielded Cf-252 source
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< 10 keV 0.300

10-100 keV 0.172

0.1-2 MeV 0.447

2-14 MeV 0.081

Claimant-favorable dose fractions 

< 10 keV 0.300

0.01- 2 MeV 0.610

0.1-2 MeV 0.081

6.3.4.2.4 Neutron-to-Photon Ratio

A 0.125 neutron-to-photon ratio was calculated in a Centers for Disease Control and Prevention 

(CDC) study of PORTS from data obtained from an onsite study conducted in 1995. Eight locations at 

the site were monitored. The observed median quarterly neutron doses were at the Feed Plant 

(5 mrem), X-344 autoclave area (7 mrem), Shipping and Receiving (7 mrem), cylinder lots (7 mrem), 

X-345 (3 mrem), X-744G (2 mrem),X-326 L Cage (< LOD), and the burn area (4 mrem). The mean 

total beta and gamma dose was 40 mrem. The mean neutron dose was 5 mrem and, thus, the 

calculated neutron-to-photon ratio was 0.125 (Cardarelli 1997).

A neutron dose can be calculated at all PORTS facilities with a potential for such a dose using the 

neutron-to-photon ratio of 0.125 applied to a fraction of the recorded photon dose. For example, a 

worker in a process facility with a recorded photon dose of 500 mrem in 1992 would be assigned a 

62.5-mrem neutron dose (i.e., 500 × 0.125 = 62.5). The photon dose should be adjusted for missed 

dose before estimation of the neutron dose. Since the routine monitoring for neutron exposure began 

in 1997, this ratio method should be used before 1997. The LOD method can apply after 1997 for 

neutron exposures.

PORTS measurement of the process and storage areas was actually about 0.125:1. For the PORTS 

neutron-to-photon ratio, a default neutron-to-photon ratio of 0.20 obtained from a PGDP cylinder 

survey (Meiners 1999) would be reasonable. Table 6-21 lists these values for PORTS facilities.

Table 6-21. PORTS neutron-tophoton dose equivalent ratios.

Facilities Neutron-to-photon ratio

General areas 0.20

Calibration facility 0.20

Nondestructive lab 0.20

HEU/EU storage 0.20

Process facilities 0.20

a. Multiply adjusted (i.e. for any missed dose) annual 

photon dose by the tabulated value and use this 

fraction times the neutron-to-photon ratio to estimate 

neutron dose.

6.3.4.3 PORTS Workplace Neutron Dosimeter Response

Table 6-22 summarizes typical neutron personnel dosimeter parameters important to Hp(10) 

performance in the workplace. The most important parameter related to Hp(10) performance of 

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

Table 6-22. Typical workplace neutron dosimeter Hp(10) performance.

a

Parameter Description Potential workplace biasb
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 31 of 55

Workplace neutron 

energy spectra

TLND response increases with decreasing 

neutron energy. 

Depends on workplace neutron spectra. 

Exposure 

geometry

TLND response decreases with increasing 

exposure angle.

TLD recorded dose is lower at angles other than AP. 

Effect is highly dependent on neutron energy. 

Missed dose Doses less than MDL recorded as zero 

dose.

Recorded dose of record is probably too low. Impact of 

missed dose is greatest in earlier years because of 

higher MDLs of neutron dosimeters. 

Environmental 

effects

Workplace environment (heat, humidity, 

etc.) fades dosimeter signal.

Recorded dose of record is probably too low.

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

b. Recorded dose compared to Hp(10).

6.3.5

Uncertainty in the recorded dose emanates primarily from two sources, laboratory bias and workplace 

radiation field composition and geometry. Sections 6.3.3 and 6.3.4 discuss potential effects of these 

parameters on the recorded dose. Table 6-23 lists the judged estimates of uncertainty in the recorded 

dose at PORTS, based on the combination of parameters. 

PORTS Workplace Dose Uncertainty

Table 6-23. Estimates of uncertainty.

Dosimeter PORTS

Laboratory 

uncertaintya Workplace uncertaintyb

Beta/gamma dosimeters

Two-element film Used 1954-1980 ±20% ±50%

TLD Used 1981-2003 ±10% ±50%

Neutron dosimeters

TLND Used 1992-2003 ±25% ±50%

a. In relation to Hp(10) response of the dosimeter.

b. 95 % confidence level.

6.4 ADJUSTMENTS TO RECORDED DOSE

Adjustments to the PORTS reported dose are necessary considering the uncertainty associated 

primarily with the complex workplace fields and exposure orientations.

6.4.1

No adjustment in recorded nonpenetrating or skin dose is recommended. PORTS incident reports 

would typically address nonroutine worker exposure to significant beta or photon radiation. The 

assessed doses in the incident reports, based on investigations conducted at the time of the incident, 

probably provide the best estimate of dose received. 

Beta Dose Adjustments 

6.4.2

Utilizing only AP geometry, the only major dose adjustments necessary to use is a correction of 

16.5% for film dosimetry (1954-1980), 12.5 % for TLD dosimetry (1981-1986) and 4 % for TLD 

dosimetry (1987-present) as listed in Table 6-24. 

Photon Dose Adjustments

Table 6-24. Adjustments to reported PORTS deep photon dose.

Period Dosimeter Facility Adjustment to reported dose

Prior to 1/1/1981 All beta/photon dosimeters All facilities Multiply reported film/TLD deep dose by 

factor of 1.165 .
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1/1/1981 -1986 All beta/photon dosimeters All facilities Multiply reported film/TLD deep dose by 

factor of 1.125.

1987 to the 

present

All beta/photon dosimeters All facilities Multiply reported film/TLD deep dose by 

factor of 1.04.

6.4.3

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

between historic and current scientific guidance, as described in NIOSH (2002). The quality factor is 

incorporated in the calibration methodology, which used flux-to-dose-rate conversion factors for 

varying neutron energies for each calibration source. Flux-to-dose-rate conversion factors were 

based on National Council on Radiation Protection (NCRP) Report 38 (NCRP 1971). This report lists 

both flux-to-dose-rate conversion factors and associated quality factors that vary from 2 at energies 

less than 1 keV to 11 at 1 MeV. To convert from NCRP (1971) quality factors to ICRP (1991) 

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

neutron energy. The average quality factor for each neutron energy group was developed by 

integrating the area under the curve and dividing by the neutron energy range, as shown in equation 

6-4. 

Neutron Dose Adjustments

2.0

0.1

,0.1 2.0

( )

( ) (2.0 0.1)

f

n MeV

Q E dE

Q E

Range − = −

∫

(6-4)

Table 6-25 summarizes historic changes in the quality factors and the average NCRP (1971) quality 

factor for the neutron energy groups used in dose reconstruction. In addition, Table 6-25 lists the 

average quality factor for the four neutron energy groups that encompass PORTS neutron exposures. 

The neutron dose equivalent correction factor can be calculated by dividing the dose fractions from 

Section 6.3.4.2 for each neutron energy group (Df(En)) by the corresponding energy specific average 

(NCRP 1971) quality factor (Q(En)) and then multiplying by the ICRP (1991) radiation weighting factor 

(wR), as shown in equation 6-5. 

( ) ( ) ( )

f n

f n R

n

D E C E w

Q E

= × (6-5)

Table 6-26 summarizes the default neutron dose fractions by energy for PORTS work areas where 

field measurements of neutron spectra occurred, using the associated ICRP (1991) correction factors. 

The neutron dose equivalent is calculated by multiplying the recorded neutron dose by the areaspecific correction factors. For example, consider a 50-mrem recorded neutron dose to a worker at 

the calibration facility. The corrected neutron dose is 22 mrem from neutrons between 0.1 MeV and 

2.0 MeV, (i.e., 50 × 0.44 ) and 51 mrem from neutrons with energy between 2 MeV and 20 MeV, (i.e., 

50 × 1.02 ). These adjustments should be applied to measured dose, missed dose, and dose 

determined based on a neutron-to-photon ratio.

Table 6-25. Historical neutron quality or weighting factors.

Neutron 

energy 

(MeV)

Historical 

dosimetry 

guidelinea

NCRP 38

quality 

factorsb

Average quality factor used at 

PORTS ( ) Q En

ICRP 60 neutron weighting 

factor, wR

c

2.5 × 10-8 3 2 2.35 5

1 × 10-7 10 2

1 × 10-6 2
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1 × 10-5 2

1 × 10-4 2

1 × 10-3 2

1 × 10-2 2.5 5.38 10

1 × 10-1 7.5 10.49 20

5 × 10-1 11

1 11

2 10 7.56 10

2.5 9

5 8

7 7

10 6.5

14 7.5

20 8 Not applicable 5

40 7

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix, Wilson, and Baumgartner 1997). 

b. Recommendations of NCRP Report 38 (NCRP 1971).

c. ICRP Publication 60 (ICRP1991)

Table 6-26. PORTS facility neutron dose fractions and associated ICRP (1991) correction factors. 

Process Description/buildings

Operations Neutron 

energy

Default dose 

fraction

ICRP (1991) 

Begin End correction factor

Calibration Area X-710 (40 mCi Cf-252 source) 1954 Present < 0.01-2 MeV

2-20 MeV

0.23

0.77

0.44

1.02

Nondestructive 

Laboratory Area

X-710 (Cf-252 source) 1954 Present < 0.01-2 MeV

2-20 MeV

0.78

0.22

1.50

0.28

HEU/EU Storage 

Areas

X-345 and Cylinder Yards 1954 Present < 0.01-2 MeV

2 – 20 MeV

0.92

0.08

1.81

0.11

General Facilities 

including process 

areasa

Uranium enrichment, recovery, 

testing, maintenance, transport 

and storage areas

1954 Present < 0.01-2 MeV

2 – 20 MeV

0.92

0.08

1.81

0.11

a. Neutron dose fraction is assumed that of the HEU/EU storage areas.

6.5 MISSED AND UNMONITORED DOSE

There is undoubtedly missed recorded dose for PORTS workers. Missed dose applies to workers who 

were monitored but had results below the LOD of their personal radiation monitors. In the early years 

of radiation monitoring, when relatively high detection limits are combined with short monitoring 

durations, missed doses can be significant. Methodologies for estimating missed doses are 

discussed in this section.

Unmonitored dose may have occurred because workers may have the potential of receiving less than 

10 % of the radiation protection guidelines or that they worked in uncontrolled areas and not 

considered radiation workers and therefore were not assigned dosimetry. The following sections 

discuss beta, photon and neutron missed or unmonitored dose. Dose reconstructors should apply 

adjustments to beta, photon, and neutron dosimetry discussed in Section 6.4 to the missed dose 

calculations discussed below. For these cases, dose reconstructors must rely on coworker data 

and/or population data to estimate a worker’s potential unmonitored dose. These methods are also 

discussed in this section.

For the special case when a worker’s exposure potential has been determined to be low, the 

environmental dose should be assigned.
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 34 of 55

6.5.1 Missed and Unmonitored Shallow Dose

Missed shallow dose can occur if a zero dose was recorded for a dosimeter reading less than the 

LOD. Estimates of missed dose can be computed from personnel dose data from either before or 

after the missing dose period or from coworker data. NIOSH (2002) recommends calculating missed 

dose by multiplying LOD/2 by the number of zero dose results. The missed shallow dose for 

dosimeter results less than the LOD is particularly important for earlier years because LODs were 

higher and dosimeter exchange frequency was higher. The last column in Table 6-27 lists the 

resulting estimates of this annual missed dose for different years at PORTS. The LOD/2 method 

seems to be more conservative for most periods. Figure 6-5 shows the PORTS SDE average and the 

maximum dose by year.

Table 6-27. Missed beta dose according to dosimeter type.

Dosimeter

LOD 

(rem) Period of use

Exchange 

frequency

Max. annual 

missed dose 

(rem)

Two-element 

film

0.03 9/22/54-7/16/57 Weekly (n=50) 0.75

7/17/57-9/30/58 Biweekly (n=25) 0.38

10/01/58-4/8/59 Weekly (n=50) 0.75

10/01/58-4/8/59 Monthly (n=12) 0.18

4/9/59-7/31/60 Every 4 wks 

(n=13)

0.20

8/1/60-7/5/64 Monthly (n=12) 0.18

8/1/60-7/5/64 Quarterly (n=4) 0.06

7/6/64-12/28/69 Quarterly (n=4) 0.06

12/29/6912/30/73 Quarterly (n=4) 0.06

12/31/73-6/29/75 Quarterly (n=4) 0.06

12/31/73-6/29/75 Semiannual 

(n=2)

0.03

6/30/75-12/31/80 Quarterly (n=4) 0.06

6/30/75-12/31/80 Monthly (n=12) 0.18

TLD 0.04 1/1/81-12/31/82 Monthly (n=12) 0.24

1/1/81-12/31/82 Quarterly (n=4) 0.08

0.03 1/1/83-12/31/98 Quarterly (n=4) 0.06

1/1/93-12/31/96 (BJC ) Quarterly (n=4) 0.06

0.03 1/1/99-present 

(USEC)

Quarterly (n=4) 0.06

0.03 1/1/99-present Quarterly (n=4) 0.06

For calculating unmonitored dose, the dose reconstructor might choose to utilize the departmental 

ratios listed in Table 6-29 to approximate doses to unmonitored personnel or unprocessed badges 

more closely. Take the results of applying the LOD/2 method or the Table 6-28 values and multiply by 

the corresponding department ratio from Table 6-29. The result should not exceed the cumulative 

dose that can be obtained by adding the maximum SDE column results for the years in question from 

Table 6-28. If more than one department is involved, those department ratios can be included by 

weighting the fraction of time worked in the department. For departments not in the list, choose the 

department with the closest function. 

Table 6-29 is based on the average accumulated skin dose of PGDP employees from 1953 to 1988. 

The ratios were calculated from dividing the average cumulative dose column by the Mechanical 

Inspection group cumulative total. Since PORTS and PDGH had similar operations, the relative dose 

in each departmental area should be similar. 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 35 of 55

Another option for estimating unmonitored dose to personnel that should have been considered 

radiation workers is to use the average or maximum doses as listed in Table 6-28. In Table 6-28 the 

listed average dose is considered the geometric mean, the listed maximum dose the 99% confidence 

level dose with the assumed distribution log normal as indicated with the geometric standard deviation 

listed. The listed geometric mean and geometric standard deviation can be directly utilized in the 

IREP code. 

The beta dose reported from 1981 through the end of 1982 may not be as reliable because there was 

no open window for the PORTS TLD badge for this period. However, the average SDE for 1981-1982 

is close to those recorded from subsequent years as indicated in Table 6-28. Table 6-27 includes the 

potential missed beta dose for the respective periods of use, dosimeter types, LOD, and exchange 

frequency.

Individuals with no dose recorded and if it is definitely established that the individual was not a 

radiation worker, then the assigned dose is the environmental dose discussed in the occupational 

environmental dose portion of this PORTS site profile. 

Table 6-28. Beta or SDE dose by year (rem).a

Year

Number 

monitored Total SDE

Maximum 

SDE

Average 

SDE 

(GeoMean)

Geometric 

Standard 

Deviation 

1954 172 57.374 1.640 0.334 1.98

1955 903 128.187 0.867 0.142 2.17

1956 610 53.579 0.655 0.088 2.37

1957 482 77.845 2.095 0.162 3.00

1958 873 229.719 5.770 0.263 3.76

1959 866 442.237 10.075 0.511 3.60

1960 488 106.240 3.420 0.218 3.26

1961 549 113.291 4.890 0.206 3.89

1962 693 94.380 1.340 0.136 2.67

1963 640 78.012 1.510 0.122 2.94

1964 621 49.253 1.388 0.117 2.89

1965 356 37.018 1.150 0.104 2.80

1966 236 24.590 0.790 0.104 2.39

1967 141 18.606 1.135 0.132 2.52

1968 611 90.098 1.040 0.147 2.32

1969 901 101.743 1.890 0.113 3.35

1970 414 70.396 2.028 0.170 2.90

1971 167 30.015 2.385 0.180 3.03

1972 234 35.116 1.440 0.150 2.64

1973 207 36.700 2.140 0.177 2.91

1974 189 27.925 2.185 0.148 3.18

1975 285 59.540 2.060 0.209 2.67

1976 363 63.585 1.350 0.175 2.40

1977 523 71.895 1.410 0.137 2.72

1978 618 127.100 1.840 0.206 2.56

1979 393 58.970 1.120 0.150 2.37

1980 1,004 122.872 1.500 0.122 2.94

1981 587 26.203 0.535 0.045 2.89

1982 784 37.638 1.709 0.048 4.63

1983 1,092 61.048 1.749 0.056 4.38

1984 1,002 46.394 0.958 0.046 3.68

1985 1,428 45.647 0.452 0.032 3.12

1986 1,308 49.436 2.198 0.038 5.71

1987 1,147 51.602 2.160 0.045 5.27
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 36 of 55

1988 1,189 42.922 0.608 0.036 3.36

1989 1,491 53.851 0.593 0.036 3.33

1990 1,862 81.447 0.694 0.044 3.27

1991 1,509 61.066 0.517 0.040 3.00

1992 889 50.365 0.910 0.057 3.28

1993 667 41.737 1.551 0.063 3.95

1994 583 24.610 0.507 0.042 2.91

1995 2,112 67.011 1.996 0.032 5.89

1996 3,461 100.426 1.951 0.029 6.09

1997 3,394 116.483 1.281 0.034 4.75

1998 355 14.238 0.300 0.040 2.37

1999 189 9.423 0.247 0.050 1.98

2000 703 19.103 0.338 0.027 2.96

2001 182 5.479 0.199 0.030 2.25

2002 224 10.710 0.353 0.048 2.35

2003 385 23.498 0.569 0.061 2.61

a. Data provided by PORTS HP department. (Litton 2004 and PORTS and PDGH 

spread sheet 2004)

Figure 6-5. PORTS SDE average and maximum dose by year. Data provided by PORTS HP 

department (PORTS and PGDP spread sheet 2004).

Energy Range

Although the different elements in a multielement dosimetry system can give specific energy exposure 

characteristics, this analysis cannot consider shielding and radiation scattering. Attempts were made 

with the PORTS TLD system to calibrate dosimeters based on a number of mixed field combinations. 

If the badge read close to the regression results of a particular mixed field calibration and the person 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 37 of 55

probably worked in an area with that expected mixed field, the dose calculations would use the mixed 

field calibration factors. More detail of the procedure was not available. 

Skin Contamination

Skin 99Tc contaminations could have gone undetected because of the low-energy beta radiation. 

Although routine surveys involved monitoring for beta contamination, the uranium daughters 234Th and 

234mPa could have masked the presence of 99Tc. 

Potential doses from 99Tc skin contamination have been evaluated by using the VARSKIN computer 

code (PNL-7913 DP). The calculated shallow dose rate from uniform 99Tc skin contamination is 1.6 × 

10-3 mrem/hr per dpm/cm2 (Swinth 2004). Because the nuclide is difficult to remove from skin, the 

integrated shallow dose resulting from 99Tc skin contamination might be relatively large. For example, 

with a residence half-time of 1.5 days, the dose is 8.1 × 10-2 mrem per dpm/cm2

. 

Some skin contamination events involving 99Tc could have gone undetected. In some cases, 

therefore, it might be appropriate to consider an additional skin dose component for a reported 

shallow dose of a worker who might have had direct contact with 99Tc. To estimate an annual missed 

shallow dose in the absence of specific data, one must make assumptions regarding the number of 

times per year an affected skin region might have been contaminated and the extent of each 

contamination. For example, one might assume a monthly contamination event at a specific location 

on the skin with an average level of 25,000 dpm/100 cm2 (the action limit for 99Tc contamination on 

work surfaces and hand tools at PORTS, see Table 6-30). With the residence half-time of 1.5 days, 

assumed above, it follows that the annual shallow dose equivalent would be (12 × 250 dpm/cm2 × 

0.081 mrem per dpm/cm2 ) = 240 mrem. The direct external dose rate at a distance of 10 cm from a 

Table 6-29. Departmental SDE dose ratios.

Description

Average 

cumulative 

dose (mrem)a

Ratio (using mechanical 

inspection employees 

as the basis)

Feed Plant Operators 15,834 30.99

Decontamination 12,369 24.21

Feed Plant Mechanics 9,767 19.11

Chemical Processors 3,794 7.42

Feed Plant Mechanics 1,968 3.85

Process Maintenance 1,954 3.82

Cascade Operators 1,824 3.57

Instrument 1,407 2.75

PEMU Decontamination 1,223 2.39

Electrical 987 1.93

Convertor Shop 933 1.83

Material Term Serv. 931 1.82

Analytical Chemistry 877 1.72

Laundry 851 1.67

Converter Test 836 1.64

Equipment Maintenance 586 1.15

Equipment Control 572 1.12

Transportation Pool 557 1.09

Fabrication Shop 517 1.01

Mechanical Inspection 511 1.00

a. Adapted from PACE (2000, Table 7-3).
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 38 of 55

surface contaminated at this same level would be (250 dpm/cm2 × 10-4 mrem/hr per dpm/cm2 ) = 

0.025 mrem/hr. At 30 cm, the rate would be 0.00025 mrem/hr.

Significant nonroutine worker doses, as might occur from skin contamination events, might be 

addressed in specific incidence reports. In such cases, dose reconstructors should consider 

assessments based on investigations conducted at the time of the incident as the best resource.

The limit of dose rate (LODR) appears to be between 0.07 and 0.04 mrem/hr (see Table 6-30) based 

on the PORTS monitoring limits. One method of skin contamination monitoring was for workers to 

wrap their hands around a GM tube. A typical GM tube is about 2% efficient for 99Tc beta particles. 

(A contemporary example is a Ludlum model 44-7 – end window GM)

For the routine meter used from at least the 1960s through the 1980s, the Samson meter appears to 

have had a LOD of 60 cpm, because this was the lowest count rate action limit. This would translate 

to a 3,000-dpm activity level, which corresponds to a 0.07 mR/hr 99Tc beta radiation level. Note in 

Table 6-30 that the lowest radiation field level utilized for an action level is 0.04 mR/hr. This is 

probably the LODR obtainable from a Cutie Pie ion chamber (used for beta and photon radiation field 

monitoring from 1965 to 1985).

External Beta Dose Rates

As a first method, dose reconstructors might estimate dose from probable external beta dose source 

terms. PORTS personnel have been exposed to some beta radiation levels, typically during 

maintenance. For example, with 99Tc trap maintenance, a typical Mg trap for 99Tc used for liquid UF6

is a 10-in. diameter by 13.25-in. tall cylinder with about 5,650 grams of Mg trap material. About 1.7 

μCi of 99Tc/gram of absorbent material is trapped, leading to about 10 mCi of 99Tc in the trap (gaseous 

Table 6-30. Beta contamination plant limits for PORTS.

a

Area

Type of 

monitoring Action limit 1963

Action limit

1975

Action limit 1980 and 

1990 (Tc-beta)

Floor, work 

surfaces, etc.

Contamination

Radiation level

500 cpm (Samson 

meter)

0.15 mR/hr @ 1 in.

0.15 mR/hr @ 

1 in.

Surface-25,000 dpm 

500 cpm Samson

Removable-5000 dpm 

100 cpm

Hands Contamination

Radiation level

Hand and foot monitor 

as posted

Hands Contamination

Radiation level

(Hands wrapped around 

OW GM tube) 100 cpm

0.04 mR/hr 

0.04 mR/hr @ 

1 in.

Body Radiation level 0.08 mR/hr 3,000 dpm/100 cm2 

(Samson – 60 cpm)

Shoes Contamination 

or radiation 

level

Personal 500 cpm 

(HFM)

Issued 5,000 cpm

0.15 mR/hr

1.5 mR/hr

Issued 400,000 

dpm/100 cm2

Clothing, tables, 

equipment, tools, 

etc.

Radiation level 0.15 mR/hr Coveralls, gloves200,000 dpm/100 cm2

Personal clothing –

9,000 dpm/100 cm2

Airborne activity Contamination

Radiation level

20 cpm/ft3 (filter paper 

wrapped around GM 

tube)

2000 dpm/ft3

Hand tools and 

other equipment

Fixed

Removable

25,000 dpm/100 cm2

3,000 dpm/100 cm2

Process shop and Fixed 200,000 dpm/100 cm2
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 39 of 55

purge equipment Removable 3,000 dpm/100 cm2

a. Radiation Protection Manual, Portsmouth Facility

UF6

99Tc traps accumulate much less activity). Using the VARSKIN code, the beta dose rate result for 

10 mCi in a trap at 30 cm (Swinth 2004) is 0.16 mrem/hr at the face. Because there are three traps 

for every changeout (PORTS) and 50 hr/yr, the resultant dose is 0.16 mrem/hr × 3 traps/changeout × 

50 hours = 24 mrem/yr. This method assumes that the personnel were potentially exposed to these 

beta radiation fields even though most personnel did not perform 99Tc trap maintenance. In addition, 

this is a conservative estimate because it assumes there was no shielding of the Mg trap material, the 

accumulation of activity at the top of the cylinder occurred, and the personnel contact was very close.

In addition, there are alumina traps located throughout the cascade buildings. They usually 

accumulate uranium daughters and some transuranics (237Np), but 99Tc is possible. PORTS indicated 

that external beta fields as high as 2 R/hr have been measured around these alumina traps, and that 

uranium daughter buildup is probably responsible for these fields. However, personal dosimetry 

would have monitored uranium daughters adequately.

A second method can be based on the LODR for an end GM of 0.08 mrem/hr (Table 6-30) for the 

whole body. With the assumption of 2,000 maximum hours of exposure, the exposure calculation 

would be: 0.08 mrem/hr × 2,000 hr exposure/ year = 160 mrem. Another approximation can be made 

using the conservative assumption that the technetium: uranium daughter ratio of about 0.40 (Basset 

1986) along with the maximum exposure from ambient levels of about 0.2 mrem/hr (Basset 1986) 

would yield a result of: 0.20 × 0.40 × 2,000 hours/year = 160 mrem.

This method might be more plausible because it represents a chronic exposure situation that was 

more common among all PORTS workers. 

In general, direct external beta dose from 99Tc is minimal. The unshielded shallow dose rate to bare 

skin (no clothing) at a distance of 10 cm in air from a uniformly contaminated surface is about 10-4

mrem/hr per dpm/cm2

, as estimated with VARSKIN. The dose rate at 30 cm is only about 10-6

mrem/hr per dpm/cm2

. Table 6-31 summarizes these benchmark values for shallow dose equivalent 

rate as determined from VARSKIN for skin contamination and for external exposure with intervening 

air. 

Table 6-31. Shallow dose-equivalent rates for 99Tc.

Condition

Dose rate

(mrem/hr per 

dpm/cm2) Dose rate mrem/hr

Skin contamination 1.6 × 10-3 N/A

External, 10 cm air 1 × 10-4 0.08 mrem/hr LLDR whole body

External, 30 cm air 1 × 10-6 0.20 mrem/hr ambient

(X.40 99Tc/U daughter ratio)

For method 1, the total SDE would equal 240 mrem skin dose + 24 mrem external beta dose = 264 

mrem/yr.

For method 2, the total SDE would equal 240 mrem skin dose + 160 mrem external beta dose = 400 

mrem/yr.
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 40 of 55

6.5.2

Missed photon dose can occur if a zero dose was recorded for a dosimeter reading less than the 

LOD. Estimates of missed dose can be computed from dose before or after the missing dose period 

or from coworker data. NIOSH (2002) recommends calculating missed dose by multiplying LOD/2 by 

the number of zero dose results. The missed photon dose for dosimeter results less than the LOD is 

particularly important for earlier years because LODs and dosimeter exchange frequency were higher. 

. The last column in Table 6-32 lists the resulting estimates of this annual missed dose for different 

years at PORTS. The LOD/2 method seems to be more conservative for most periods. Figure 6-6 

shows PORTS DDE average and maximum dose by year.

Missed and Unmonitored Photon Dose

Individuals with no dose recorded and if it is definitely established that the individual was not a 

radiation worker, then the assigned dose is the environmental dose discussed in the occupational 

environmental dose portion of this PORTS site profile. 

For calculating unmonitored dose, the dose reconstructor might choose to utilize departmental ratios 

listed in Table 6-34 to approximate doses to unmonitored personnel or unprocessed badges. Take 

the results of applying the LOD/2 method or the Table 6-33 values and multiply by the corresponding 

department ratio from Table 6-34. The result should not exceed the cumulative dose that can be 

obtained by adding the maximum DDE column results for the years in question from Table 6-33. If 

more than one department is involved, include those department ratios by weighting the fraction of 

time worked in the department. For 

Table 6-32. Missed photon dose adjustments to recorded deep dose according to PORTS facility.

Facility 

type Dosimeter

LOD 

(rem) Period of use

Exchange 

frequency

Max. annual 

missed dose 

(rem)

All facilities PORTS two-element 

film

0.03 9/22/54-7/16/57 Weekly (n=50) 0.75

7/17/57-9/30/58 Biweekly (n=25) 0.38

10/01/58-4/8/59 Weekly (n=50) 0.75

10/01/58-4/8/59 Monthly (n=12) 0.18

4/9/59-7/31/60 Every 4 wks (n=13) 0.20

8/1/60-7/5/64 Monthly (n=12) 0.18

8/1/60-7/5/64 Quarterly (n=4) 0.06

7/6/64-12/28/69 Quarterly (n=4) 0.06

12/29/6912/30/73 Quarterly (n=4) 0.06

12/31/73-6/29/75 Quarterly (n=4) 0.06

12/31/73-6/29/75 Semiannual (n=2) 0.03

6/30/75-12/31/80 Quarterly (n=4) 0.06

6/30/75-12/31/80 Monthly (n=12) 0.18

PORTS Harshaw w/o 

window 4-element TLD

0.015 1/1/81-12/31/82 Monthly (n=12) 0.09

1/1/81-12/31/82 Quarterly (n=4) 0.03

All facilities PORTS Harshaw with 

window 4-element TLD

0.010 1/1/83-12/31/98 Quarterly (n=4) 0.02

1/1/93-12/31/96 (BJC ) Quarterly (n=4) 0.02

ICN TLD 760 0.010 1/1/99-present (USEC) Quarterly (n=4) 0.02

ORNL TLD Panasonic 0.010 1/1/99-present (BJC ) Quarterly (n=4) 0.02

departments not in the list, choose the department with the closest function. Table 6-33 is based on 

the average accumulated deep dose of PGDP employees from 1953 to 1988. The ratios were 

calculated from dividing the average cumulative dose column by the Steam Plant group cumulative 

total. Since PORTS and PDGH had similar operations, the relative dose in each departmental area 

should be similar. 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 41 of 55

Another option for estimating unmonitored dose to personnel that should have been considered 

radiation workers is to use the average or maximum doses as listed in Table 6-28. In Table 6-33 the 

listed average dose is considered the geometric mean, the listed maximum dose the 99% confidence 

level dose and the assumed distribution log normal as indicated with the geometric standard deviation 

listed. The listed geometric mean and geometric standard deviation can be directly utilized in the 

IREP code. 

Facility/Location

Table 6-32 lists the missed photon dose at PORTS facilities. The same information is listed in Table 

6-3, which contains the table references. The maximum annual missed dose is the LOD/2 multiplied 

by the exchange frequency. The LOD is based on laboratory irradiations or from citations from 

dosimetry vendors. 

Table 6-33. Photon or DDE dose by year (rem).a

Year

Number 

monitored Total DDE

Maximum 

DDE

Average 

DDE

(Geo Mean)

Geometric 

Standard 

Deviation

1954 172 51.945 1.500 0.302 1.99

1955 903 112.711 0.867 0.125 2.30

1956 610 48.002 0.590 0.079 2.37

1957 482 25.805 0.355 0.054 2.24

1958 873 130.699 1.090 0.150 2.34

1959 866 296.067 5.000 0.342 3.16

1960 488 35.840 0.900 0.073 2.94

1961 549 38.691 1.740 0.070 3.97

1962 693 65.015 0.765 0.094 2.46

1963 640 69.772 0.960 0.109 2.54

1964 621 45.244 1.388 0.073 3.00

1965 356 27.710 1.150 0.078 3.17

1966 236 21.884 0.790 0.093 2.50

1967 141 16.460 0.885 0.117 2.38

1968 611 86.630 1.040 0.142 2.35

1969 901 94.836 1.890 0.105 3.46

1970 414 64.633 1.145 0.156 2.35

1971 167 22.595 1.420 0.135 2.75

1972 234 26.931 1.440 0.115 2.96

1973 207 32.415 2.140 0.157 3.07

1974 189 24.395 1.720 0.129 3.04

1975 285 46.225 1.420 0.162 2.54

1976 363 55.424 1.280 0.153 2.49

1977 523 62.290 1.410 0.119 2.89

1978 618 75.170 1.210 0.122 2.68

1979 393 22.835 0.715 0.058 2.94

1980 1,004 117.394 1.500 0.117 2.99

1981 587 26.203 0.535 0.045 2.89

1982 784 18.467 0.864 0.024 4.66

1983 1,092 20.014 0.543 0.018 4.32

1984 1,002 17.855 0.529 0.018 4.27

1985 1,428 16.765 0.452 0.012 4.75

1986 1,308 20.277 1.973 0.016 7.90

1987 1,147 27.207 0.826 0.024 4.57

1988 1,189 21.957 0.395 0.018 3.76

1989 1,491 29.255 0.348 0.020 3.41

1990 1,862 39.745 0.430 0.021 3.65

1991 1,509 24.650 0.306 0.016 3.55
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 42 of 55

1992 889 19.173 0.446 0.022 3.64

1993 667 18.992 0.428 0.028 3.22

1994 583 8.852 0.259 0.015 3.40

1995 2,112 24.181 0.821 0.011 6.37

1996 3,461 35.346 0.317 0.010 4.41

1997 3,394 30.374 0.292 0.009 4.45

1998 355 10.182 0.285 0.029 2.67

1999 189 6.406 0.183 0.034 2.06

2000 703 15.404 0.173 0.022 2.42

2001 182 4.259 0.176 0.023 2.40

2002 224 7.168 0.316 0.032 2.67

2003 385 17.590 0.496 0.046 2.77

a. Data provided by PORTS HP department. (Litton 2004 and PORTS and PDGH 

spread sheet 2004)

Figure 6-6. PORTS DDE average and maximum dose by year. Data provided by PORTS HP 

department. (PORTS and PGDP spread sheet 2004)

Table 6-34. Departmental DDE dose ratios.

Description

Average cumulative 

dose in mrema

Ratio (using steam 

plant employees 

as the basis)

Plant feed operators 3,814 34.67

Decontamination 2,788 25.35
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 43 of 55

Feed Plant Mechanics 2,587 23.52

Cascade Operators 627 5.70

Chemical Operators 595 5.41

Instrument 538 4.89

Transportation Pool 371 3.37

Process Maintenance 364 3.31

Environ Control 338 3.07

Electricians 298 2.71

Mat. Term Mgr 295 2.68

PEMU Decontamination 253 2.30

Mechanical Inspection 170 1.55

Plant services 147 1.34

Converter Test 145 1.32

Feed Plant Mechanics 143 1.30

Nitrogen Plant 142 1.29

Metals Building 132 1.20

Fabrication Shops 127 1.15

Steam Plant 110 1.00

 a. Adapted from PACE (2000, Table 7-2).

6.5.3

Table 6-35 summarizes missed neutron dose for PORTS; see Table 6-3 for references. For the 

PORTS neutron-to-photon ratio is 0.20. Table 6-21 lists these values for different PORTS facilities. 

The photon dose should be adjusted for missed dose before estimation of the neutron dose. Since 

the routine monitoring for neutron exposure began in 1997, the neutron to photon ratio method should 

be used before 1997. The LOD method can apply after 1997 for neutron exposures. Multiply this dose 

equivalent by the neutron dose correction factors listed in Table 6-26 for different processes or 

buildings.

Missed and Unmonitored Neutron Dose

Unmonitored neutron dose may be assigned or be calculated with photon co-worker data or 

population data as presented in Tables 6-33 and 6-34. The same method as described in the above

paragraph can than be applied to the unmonitored neutron dose calculation. 

Table 6-35. PORTS dosimeter type, period of use, exchange frequency, LOD, and potential annual 

dose missed.

Dosimeter Period of use

Exchange 

frequency

Laboratory 

LOD (rem)

Maximum 

annual missed 

dose (rem)a

Neutron dosimeters

PORTS TLD albedo dosimeter 

{USEC and BJC }

1/1/1992 – 12/31/94

{Unmoderated Cf-252 

calibrated}

Quarterly (n=4) 0.02j 0.04

PORTS TLD albedo dosimeter 

{USEC and BJC }

1/1/95-12/31/96

{Moderated Cf-252 calibrated}

Quarterly (n=4) 0.02j 0.04

ICN TLD 760 {USEC} 1/1/97-present

{Moderated Cf-252 calibrated}

Quarterly (n=4) 0.01g 0.02

Y-12 Panasonic TLND {BJC 

employees}

1/1/97-12/31/98 Quarterly (n=4) 0.01h 0.02

ORNL Panasonic TLND 8806 fourelement TLD {BJC employees}1/1/1999 - present Quarterly (n=4) 0.01i 0.02
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 44 of 55

6.6 ORGAN DOSE

6.6.1

NIOSH (2002) describes the methodology to calculate the organ dose distribution for the respective 

radiation types using identified exposure geometries. The missed or unmonitored dose is to be 

estimated as described in Section 6.5. A correction of 16.5% for film dosimetry (1954-1980), 12.5 % 

for TLD dosimetry (1981-1986) and 4 % for TLD dosimetry (1987-present) is recommended. After the 

corrections are applied, the R to organ DCFs should be applied for the years 1954-1986 and the 

HP(10) to organ DCFs should be applied for the years from 1987 to the present. The DCFs can be 

found in appendix B of the External dose reconstruction implementation guideline, rev 1. (NIOSH 

2002) 

Organ Dose Conversion Factors

6.7 DOSE RECONSTRUCTION

As much as possible, dose to individuals should be based on dosimetry records. It is important to 

distinguish between the recorded non-penetrating and penetrating doses and the actual Hp(0.07) and 

Hp(10). The following list summarizes appropriate information:

• Dosimetry records that provide nonzero beta-photon values for Hp(10) and Hp(0.07) are 

considered adequate. A correction of 16.5% for film dosimetry (1954-1980), 12.5 % for TLD 

dosimetry (1981-1986) and 4 % for TLD dosimetry (1987-present) is recommended. Beta 

energies are greater than 15 keV and photon energies should be considered to be in the range 

30 keV to 250 keV.

• Workers for whom dosimetry records provide zero beta-photon values for Hp(10) and Hp(0.07) 

should have missed dose assigned on the basis of LOD/2 times the number of zeros, as 

described in Sections 6.5.1 and 6.5.2 (NIOSH (2002). 

• Individuals with no dose recorded and if it is definitely established that the individual was not a 

radiation worker, then the assigned dose is the environmental dose discussed in the 

Occupational Environmental Dose portion of this PORTS Site Profile. 

• The missed dose is to be estimated as described in Section 6.5. A correction of 16.5% for film 

dosimetry (1954-1980), 12.5 % for TLD dosimetry (1981-1986) and 4 % for TLD dosimetry 

(1987-present) is recommended. After the corrections are applied, the R to organ DCFs 

should be applied for the years 1954-1986 and the HP(10) to organ DCFs should be applied 

for the years from 1987 to the present. The DCFs can be found in appendix B of the External 

dose reconstruction implementation guideline, rev 1. (NIOSH 2002) 

• Reported and missed neutron dose equivalents should be adjusted according to 6.4.3 to 

adjust for ICRP (1990).

• Cylinder yard workers, security guards and general workers for whom no neutron dose is 

recorded should have missed neutron dose equivalent estimate assigned based on a neutronto-photon ratio of 0.20 for dose equivalent (Meiners 1999). 

• Special attention should be paid to the possibility of skin contamination incidents for workers 

involved with 99Tc recovery operations (Section 6.5.1).

• Uncertainty is discussed in Table 6.11, 6.12,6.17 and Table 6.23. 
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 45 of 55

REFERENCES

Bassett, A. C., 1986, Thermoluminescent Dosimeter Response and Calibration, GAT-S-56, Piketon, 

Ohio, April 20.

BJC (Bechtel Jacobs Company LLC), 2000, Recycled Uranium Mass Balance Project, Portsmouth, 

Ohio Site Report, BJC/PORTS-139/R1, U. S. Department of Energy, Office of Environmental 

Management, Washington, D.C.

Cardarelli, J. J., 1997, NIOSH Health Hazard Evaluation Report, HETA 96-1298-2651, prepared for 

the U.S. Department of Labor, Centers for Disease Control and Prevention, National Institute 

for Occupational Safety and Health, Cincinnati, Ohio.

Cylinder lot-745D, Special Purpose Radiological Survey, 6/23/1998, p 1

Cylinder lot-745G, Special Purpose Radiological Survey, 7/9/1998, p 8

DOE (U.S. Department of Energy), 1986, Department of Energy Standard for the Performance 

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Washington, D.C.

DOE (U.S. Department of Energy), 2000, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2000, Washington, D.C. 

DOE (U.S. Department of Energy), 2002a, Independent Investigation of the Portsmouth Gaseous 

Diffusion Plant, Volume 1: Past Environment, Safety, and Health Practices, Office of Oversight 

Environment, Safety, and Health, Washington, D.C.

DOE (U.S. Department of Energy), 2002b, Independent Investigation of the Portsmouth Gaseous 

Diffusion Plant, Volume 2: Current Environment, Safety and Health Practices, Office of 

Oversight Environment, Safety, and Health, Washington, D.C.

Fix, J. J., R. H. Wilson, and W. V. Baumgartner, 1996, Retrospective Assessment of Personnel 

Neutron Dosimetry for Workers at the Hanford Site, PNNL-11196, Pacific Northwest National 

Laboratory, Richland, Washington. 

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997. “A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiological Study,” Radiation Protection 

Dosimetry 74, 39-53.

GAT (Goodyear Atomic Corporation), 1963, Film Badge Procedure, Piketon, Ohio, (original dated 

May, 3, 1963).

GAT (Goodyear Atomic Corporation), 1964, GAT Film Badge Program, Piketon, Ohio, February 26.

GAT (Goodyear Atomic Corporation), 1971, Film Badge Program, Piketon, Ohio.

ICN (International Chemical and Nuclear Corporation), 2003, Internet website, http://www.ICN.com, 

Dosimetry section, accessed October 2003. 

ICRP (International Commission on Radiological Protection), 1983, “Radionuclide Transformations: 

Energy and Intensity of Emissions,” Publication 38, Annals of the ICRP, Vol. 21, Pergamon 

Press, Oxford, England.
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ICRP (International Commission on Radiological Protection), 1991, “1990 Recommendations of the 

International Commission on Radiological Protection,” Publication 60, Annals of the ICRP, Vol. 

11, Pergamon Press, Oxford, England.

Landauer (Landauer Inc.), 2003, Internet site, http://www.landauerinc.com/neutron.htm, accessed 

October 2.

Litton, R., personal communication. February 2004. “PORTS Historical Film/TLD Summary – DDE>0, 

SDE > 0 or SDEWB > 0”

Meiners, S., 1999, Paducah UF6 Cylinder Painting Project, Bechtel Jacobs Company LLC., Paducah, 

Kentucky.

McMahon, 9/17/2003 Personal communication, “ORNL Panasonic 8805/8806 TLD and TLND LOD”

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation, Report 38, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements), 1991, Calibration of Survey 

Instruments Used in Radiation Protection for the Assessment of Ionizing Radiation Fields and 

Radioactive Surface Contamination, Report 112, Bethesda, Maryland.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, Rev. 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

ORAU (Oak Ridge Associated Universities), 2003a, Technical Basis Document for the Savannah 

River Site, ORAUT-TKBS-0003, Oak Ridge, Tennessee.

ORAU (Oak Ridge Associated Universities), 2003b, Technical Basis Document for the Hanford Site, 

Draft, ORAUT-TKBS-0006, Oak Ridge, Tennessee.

ORAU (Oak Ridge Associated Universities), 2004, Technical Basis Document for the Portsmouth Site, 

Draft, ORAUT-TKBS-0015-2, Oak Ridge, Tennessee.

PACE (Paper, Allied Industrial, Chemical and Energy Workers International Union) and University of 

Utah, 2000, Exposure Assessment Project at the Paducah Gaseous Diffusion Plant, Paducah, 

Kentucky.

PNL-7913 DP, NUREG/CR-5873, “Varskin Mod 2 and SADDE MOD 2: Computer Codes for 

Assessing Skin Dose from Skin Contamination.” 

PORTS and PDGH spread sheet 2004. Developed by Demopoulos, P. J. May 2004

Radiation Protection Manual, Portsmouth Facility, “Acceptable Plant Limits for radiation and 

Radioactive Contamination(4/1/1963), Section 243.5 (7/1/1975) and GAT/GDP-1073, vol 2 pg 

4.3-35 (5/10/1990). 

Soldat, K. L., and J. E. Tanner, 1992, Neutron Dose Equivalent and Energy Spectra Measurements at 

the Portsmouth Gaseous Diffusion Plant, Pacific Northwest Laboratory, Richland, Washington.

Souleyrette, 9/12/2003 Personal communication, “Y-12 TLND LOD” 
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Stewart, D. N., 2004, A Standard Complex-wide Correction Factor for Over-estimating External Doses 

Measured with Film Badge Dosimeters, ORAUT-OTIB-0010.

Swinth, K. L., 200 4, E-mail correspondence dated March 5 and March 19, External Dosimetry Task 

Group. 

Thierry-Chef, I., F. Pernicka, M. Marshall, E. Cardis, and P. Andreo, 2002, “Study of a Selection of 10 

Historical Types of Dosimeter: Variation of the Response to Hp(10) with Photon Energy and 

Geometry of Exposure,” Radiat Prot Dos, 102(2): 101-113.

USEC (United States Enrichment Corporation), 2003, “The History of PORTS,” Internet site, 

http://www.usec.com/v2001_02/html/facilities_portsoverview.asp, accessed August 21.

Wagner, 9/2003 Personal communication, “LOD of PORTS Harshaw 2276 4-element TLD”

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program From 1944 Through 1989, PNL-7447, Pacific 

Northwest Laboratory, Richland, Washington.

Wooldridge, F., 1964, A Description of Co-Operative Work Assignments in Industrial Hygiene and 

Health Physics, Goodyear Atomic Corporation, Portsmouth, Ohio, June 1.

Bibliography

Bassett, A. C., T. H. Maggard, and W. S. Terry, 1985, TLD Program Processing Procedures Harshaw 

2276 Version BGXI, GAT-S-54, Goodyear Atomic Corporation, Piketon, Ohio.

Cochran, S., 2002, Radiation Calibration Facility Normal Operating Procedure, XP4-GP-RI401, p. 5, 

December 31.

Fix, J. J., L. Salmon, G. Cowper, and E. Cards, 1997, “A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiologic Study,” Radia. Prot. Dosi., Vol. 74, pp. 

39-53.

GAT (Goodyear Atomic Corporation), 1963, Transportation of RAM, pp. 2-6, Piketon, Ohio, May 14.

GAT (Goodyear Atomic Corporation), 1963, Film Calibration, pp. 1-4 and appendix, Piketon, Ohio, 

June.

GAT (Goodyear Atomic Corporation), 1964, Procedure for Extending Film Badge Range, pp. 1-3, 

Piketon, Ohio, May 12.

GAT (Goodyear Atomic Corporation), 1971, General Procedure for X-raying Film Badges, Piketon, 

Ohio, April.

Hill, L. R. L., and D. J. Strom, 1993, Internal Dosimetry Technical Basis Manual for Portsmouth and 

Paducah Gaseous Diffusion Plants, PNL-8723, Batelle Memorial Institute, Pacific Northwest 

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Against External Radiation,” Publication 51, Annals of the ICRP, Volume 17, Pergamon Press, 

Oxford, England.
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Ruggles, D. J., 1986, TLD Error Analysis, GAT-S-58, Goodyear Atomic Corporation, Piketon, Ohio, 

January 22.
Effective Date: 01/18/2005 Revision No. 00 Document No. ORAUT-TKBS-0015-6 Page 49 of 55

GLOSSARY

absorbed dose, D 

Amount of energy imparted by radiation to unit mass of absorbing material (100 ergs per 

gram), including tissue. The unit used prior to the use of the International System of metric 

units (SI) is the rad; the SI unit is the gray.

accreditation 

In relation to this document, recognition that a dosimeter system has passed the performance 

criteria of the DOE Laboratory Accreditation Program (DOELAP) standard (DOE 1986) in 

specified irradiation categories.

accuracy 

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

albedo dosimeter 

A TLD device that measures the thermal, intermediate, and fast neutrons that are scattered 

and moderated by the body from an incident fast neutron flux.

algorithm 

A computational procedure.

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a successor to the 

Manhattan Engineering District (MED) and a predecessor to the U.S. Department of Energy 

(DOE).

backscatter 

Deflection of radiation by scattering processes through angles greater than 90 degrees, with 

respect to the original direction of motion.

beta particle 

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) of the direct fission products emit (negative) beta 

particles. Physically, the beta particle is identical with an electron moving at high velocity.

Bonner Sphere

See Multisphere Neutron Spectrometer

bremsstrahlung 

Secondary photon or X-ray radiation produced by deceleration of charged particles passing 

through matter.

buildup 

In relation to this document, increase in flux or dose due to scattering in the medium.

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).
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control dosimeter

A dosimeter used to establish the dosimetry system response to radiation dose. The 

dosimeter is exposed to a known amount of radiation dose.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep absorbed dose (Dd)

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

deep dose equivalent (Hd)

The dose equivalent at the respective depth of 1.0 cm in tissue.

densitometer

Instrument that has a photocell to determine the degree of darkening of developed 

photographic film.

density reading

See optical density.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The special unit is the rem. When D is expressed in Gy, H is in Sieverts (Sv). 

(1 Sv = 100 rem.)

DOELAP

The DOE Laboratory Accreditation Program (DOELAP) accredits DOE site dosimetry 

programs based on performance testing and onsite reviews performed on a two year cycle.

dose equivalent index

For many years the dose equivalent used to calibrate neutron sources that were used to 

calibrate neutron dosimeters a concept of summing the maximum dose equivalent delivered in 

the ICRU sphere at any depth for the respective neutron energies even though the maximum 

dose occurred at different depths.

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See albedo dosimeter, film 

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, 

and/or extremities. This includes the fabrication, assignment, and processing of the 

dosimeters as well as interpretation and documentation of the results.

error

Term used to express the difference between the estimated and "true" value. Error can also 

be used to refer to the estimated uncertainty.
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exchange period (frequency)

Time period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

As used in the technical sense, a measure expressed in roentgens of the ionization produced 

by gamma (or X) rays in air.

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.

fast neutron

Neutron of energy between 10 keV and 10 MeV.

field calibration

Dosimeter calibration based on radiation types, intensity and energies present in the work 

environment.

film

Generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be measured 

using an optical densitometer. 

film density

See optical density.

film dosimeter

A small packet of film within a holder that attaches to a worker.

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus. 

gamma ray interactions

Interaction of gamma rays with matter occurs through three primary processes as follows:

Photoelectric absorption - Process whereby a gamma ray (or X-ray) photon, with energy 

somewhat greater than that of the binding energy of an electron in an atom, transfers all its 

energy to the electron, which is consequently removed from the atom.

Compton scattering - Attenuation process observed for X-ray or gamma radiation in which an 

incident photon interacts with an orbital electron of an atom to produce a recoil electron and a 

scattered photon of energy less that the incident photon.
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Pair production - Absorption process for X-ray and gamma radiation in which the incident 

photon is annihilated in the vicinity of the nucleus of the absorbing atom, with subsequent 

production of an electron and positron pair. This reaction only occurs for incident photon 

energies that exceed 1.02 MeV.

hurst dosimeter

Film-based criticality dosimeter.

intermediate energy neutron

Neutron of energy between 0.5 ev (assumed to be 0.4 ev because of cadmium cutoff in 

neutron response) and 10 keV.

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

isotopes

Forms of the same element having identical chemical properties but differing in their atomic 

masses. Isotopes of a given element all have the same number of protons in the nucleus but 

different numbers of neutrons. Some isotopes of an element may be radioactive.

kiloelectron-volt (keV)

An amount of energy equal to 1,000 electron-volts.

luminescence

The emission of light from a material as a result of some excitation.

Minimum Detection Level (MDL)

Often confused because the statistical parameters necessary to its calculation are not 

explicitly defined. Nonetheless, often assumed to be the level at which a dose is detected at 

the two-sigma level (i.e., 95% of the time). The MDL should not be confused with the 

minimum recorded dose. 

minimum recorded dose

Based on a policy decision, the minimum dose level that is routinely recorded. A closely 

related concept is the dose recording interval. PORTS has generally recorded minimum 

doses of 10 mrem and at intervals of 10 mrem (10, 20, 30, etc.).

million-electron volt (MeV)

An amount of energy equal to 1,000,000 electron-volts.

multisphere neutron spectrometer

A series of neutron moderating spheres of tissue-equivalent material with a neutron detector 

positioned at the middle of the respective spheres. Algorithms are used to unfold the data to 

calculate the neutron spectra.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.
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neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, intermediate

Neutrons with energy between 0.4 eV and 10 keV.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons with energy 

less than the cadmium cutoff at about 0.4 eV.

open window (OW)

Designation on TLD dosimeter reports is of little use because there was no OW throughout 

1981. Otherwise, an OW existed for the film and TLD badges used at PORTS.

optical density

The quantitative measurement of photographic blackening the density defined as D = Log10

(Io/I).

penetrating dose

Designation (i.e., P or Pen) on film dosimeter reports that implies a radiation dose, typically to 

the whole body, from higher energy photon radiation.

personal dose equivalent, Hp(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. Represented by Hp(d), where d identifies the depth (in mm) and represents 

the point of reference for dose in tissue. For weakly penetrating radiation of significance to 

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

whole-body dose, d = 10 mm and is noted as Hp(10). 

photon

A unit or "particle" of electromagnetic radiation consisting of X- and/or gamma rays. 

precision

If a series of measurements has small random errors, the measurements are said to have high 

precision. The precision is represented by the standard deviation.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation. 

radiation monitoring

Routine measurements and the estimation of the dose equivalent for determining and 

controlling the dose received by workers.
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radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

radionuclide

Unstable nuclides that emit radiation, generally alpha or beta particles, gamma rays, and 

neutrons to eventually form stable nuclei.

random errors

When a given measurement is repeated, the resulting values, in general, do not agree exactly. 

The causes of the disagreement between individual values must also be causes of their 

differing from the "true" value. Errors resulting from these causes are random errors.

rem

A unit of dose equivalent, which is equal to the product of the number of rads absorbed and 

the quality factor.

roentgen

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kg of dry air. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue.

scattering

The diversion of radiation from its original path as a result of interactions with atoms between 

the source of the radiation and a point at some distance away. Scattered radiations are 

typically changed in direction and of lower energy than the original radiation.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.07 mm in a material of specified geometry and 

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 mm in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2

.

systematic errors

When a given measurement is repeated and the resulting values all differ from the "true" value 

by the same amount, the errors are systematic.

thermal neutron

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons of energy less 

than the cadmium cutoff of about 0.4 ev.
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tissue equivalent

Used to imply that the radiation response characteristics of the material being irradiated are 

equivalent to tissue. Achieving a tissue equivalent response is typically an important 

consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

Tissue Equivalent Proportional Counter (TEPC)

Device used to measure the absorbed dose from neutron radiation in near tissue equivalent 

materials and, through analysis of the counter data, determination of the effective quality factor 

and the dose equivalent.

TLD chip

A small block or crystal made of LiF used in the TLD.

TLD-600 - A TLD chip made from 6

Li (>95%) used to detect neutrons.

TLD-700 - A TLD chip made from 7

Li (>99.9%) used to detect photon and beta radiation.

thermoluminescent

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated will release the stored energy as 

light. The measurement of this light provides a measurement of absorbed dose. The solid 

chips are sometimes called crystals.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1,000 mg/cm2

); also 

used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extranuclear origin.