ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 46

Document Title:

Site Profile for the Kansas City Plant

Document Number: ORAUT-TKBS-0031

Revision: 00 PC-1

Effective Date: 01/06/2006

Type of Document: TBD

Supersedes: None 

Subject Experts: Jack J. Fix and Richard J. Traub

Document Owner

Approval: Signature on File Approval Date: 05/24/2005

Jack J. Fix, TBD Team Leader

Approval: Signature on File Approval Date: 05/26/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 05/26/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 05/31/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 2 of 46

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

05/31/2005 00 New document to establish the technical basis for the development 

of a radiation exposure profile for the Kansas City Plant. First 

approved issue. No training required. Initiated by Donald E. Bihl.

01/06/2006 00 PC-1 Approved page change revision. Incorporates changes for 

environmental ambient external and internal doses for unmonitored, 

non-radiological workers as a result of formal internal review and 

informal OCAS comments. Page change includes new content 

added to pages 7 and 22 (Sections 1.0 and 5.1.4). Removed 

definition of unmonitored worker on page 26 (Section 6.4.1). Minor 

revisions were made to pages 17, 18, 19, and 23 (Sections 4.0, 5.0, 

5.1.2, 5.2 and 6.2.1). As a result of NIOSH formal review, 

incorporates Attachment A on pages 43 – 46, into Site Profile. This 

attachment describes technical basis of ambient environmental dose 

estimate for unmonitored workers. No sections were omitted. 

Retraining is not required. Initiated by Jack J. Fix.

Approval:

Document Owner:

Jack J. Fix, TBD Team Leader

Signature on File______________________________12/19/2005 

Judson L. Kenoyer, Task 3 Manager

Signature on File______________________________12/20/2005 

Kate Kimpan, Project Director

Signature on File________ ________________ 12/28/2005 

James W. Neton, Associate Director for Science

Signature on File_________ _ ________________ 01/06/2006 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 3 of 46

TABLE OF CONTENTS

Section

Publication Record ..............................................................................................................................

Page

2

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

1.0 Introduction.................................................................................................................................7

2.0 Site Description...........................................................................................................................8

2.1 Facilities................................................................................................................................8

2.2 Occupational Safety Program..............................................................................................10

2.3 Radiation Monitoring Program.............................................................................................10

2.4 Radiation Sources...............................................................................................................11

2.5 Radiation Detection Instrumentation....................................................................................12

2.6 Radiological Records ..........................................................................................................12

3.0 Occupational Medical Dose ...................................................................................................... 14

3.1 Dose Reconstruction...........................................................................................................14

3.2 Uncertainty Analysis............................................................................................................15

3.3 Instruction Guide for Dose Reconstructors..........................................................................15

4.0 Occupational Environmental Dose ............................................................................................ 17

5.0 Occupational Internal Dose....................................................................................................... 17

5.1 Depleted Uranium ...............................................................................................................18

5.1.1 Physical and Radiological Characteristics................................................................... 18

5.1.2 Workplace Monitoring................................................................................................. 18

5.1.3 Bioassay.....................................................................................................................19

5.1.4 Unmonitored Worker .................................................................................................. 22

5.2 Other Nuclides ....................................................................................................................22

6.0 Occupational External dose ...................................................................................................... 23

6.1 Basis of Comparison ...........................................................................................................23

6.2 Workplace External Radiation Fields...................................................................................23

6.2.1 Beta Radiation............................................................................................................ 23

6.2.2 Photon Radiation........................................................................................................ 24

6.2.3 Neutron Radiation....................................................................................................... 24

6.3 Dosimeter Technology ........................................................................................................24

6.3.1 Beta/Photon Dosimeters............................................................................................. 24

6.3.2 Neutron Dosimeters.................................................................................................... 25

6.4 Dose Reconstruction...........................................................................................................25

6.4.1 Potential Unmonitored Dose....................................................................................... 26

6.4.2 Potential Missed Dose for Monitored Workers ............................................................ 26

6.4.3 Neutron Dose Adjustments for Monitored Workers ..................................................... 27

6.4.4 Adjustments to Recorded Deep Dose......................................................................... 28

6.4.5 Radiation Dose Fraction ............................................................................................. 29

6.5 Uncertainty in Photon and Neutron Dose ............................................................................29

6.5.1 Deep Photon Dose for Film Dosimeters...................................................................... 30

6.5.2 Deep Photon Dose for TLDs....................................................................................... 30
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6.5.3 Alternative Approach .................................................................................................. 31

6.5.4 Shallow Dose ............................................................................................................. 31

6.5.5 Neutron Dose Uncertainty .......................................................................................... 32

6.6 Organ Dose.........................................................................................................................32

References ........................................................................................................................................34

Glossary ............................................................................................................................................39

ATTACHMENT A, EXTERNAL AND INTERNAL AMBIENT ENVIRONMENTAL DOSE ASSIGNMENT 

GUIDELINE.......................................................................................................................................43

A1.0 External Radiation.....................................................................................................................43

A2.0 Internal Radiation......................................................................................................................43

 A2.1 KCP DU Air Concentrations Outside the Plant From Activities Inside the Plant ................ 45

 A2.1.1 Receptor on the Roof........................................................................................... 45

 A2.1.2 Receptor on the Ground Downwind From the Source on the Roof ....................... 46
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LIST OF TABLES

Table

1 History of facility operations ........................................................................................................

Page

9

2 Categorization of KCP workers in special physical examination program 

working with radiation ...............................................................................................................10

3 Types of radiation sources ........................................................................................................ 11

4 Typical RGDs............................................................................................................................12

5 Radiation detection equipment list ............................................................................................ 13

6 Organ dose estimates for PA chest radiographs before 1964 assuming minimal 

collimation.................................................................................................................................16

7 Organ dose estimates for PA chest radiographs from 1964 to 2004.......................................... 16

8 PA chest occupational X-ray frequency..................................................................................... 16

9 Mass and radiological characteristics of DU.............................................................................. 18

10 Alpha contamination levels, 1962 to 1969................................................................................. 18

11 Statistical parameters of measured DU in workplace air ........................................................... 19

12 Statistical parameters of recorded DU in urine .......................................................................... 20

13 Number of recorded bioassay measurements and average of measurements 

for 1959 through 1971...............................................................................................................21

14 Personnel dosimetry systems ................................................................................................... 25

15 Statistical parameters of recorded penetrating annual doses .................................................... 28

16 Potential missed dose for monitored workers............................................................................ 29

17 Neutron dose fractions and associated ICRP 60 correction factors........................................... 29

18 Beta, photon, and neutron radiation energies and percentages for external 

radiation exposures...................................................................................................................29

19 Uncertainty for KCP film photon deep dose............................................................................... 30

20 Uncertainty in TLD photon deep dose....................................................................................... 31

21 AP photon exposure-to-organ DCFs ......................................................................................... 32

22 AP neutron deep-dose-equivalent-to-organ DCFs..................................................................... 33

A-1 IREP input for ambient external dose calculation ...................................................................... 43

A-2 KCP personnel dosimeter control dosimeter data ..................................................................... 44

A-3 IMBA input for ambient internal dose calculation....................................................................... 45

LIST OF FIGURES

Figure

1 KCP site map..............................................................................................................................

Page

9

2 Lognormal probability plot of KCP recorded annual penetrating dose ....................................... 27
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

AP anterior-posterior

DCF dose conversion factor

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DU depleted uranium 

DXT density times thickness

ESE entrance skin exposure

GSD geometric standard deviation

HVL half-value layer

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

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

KCP Kansas City Plant

MDL minimum detectable level

MED Manhattan Engineer District

NCRP National Commission on Radiological Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NNSA National Nuclear Security Administration

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

OSL optical stimulated luminescence

PA posterior-anterior

PIC pocket ionization chamber

RGD radiation generating device

TLD thermoluminescent dosimeter

U.S.C. United States Code
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1.0

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 NIOSH in the completion of the individual 

work required for each dose reconstruction.

INTRODUCTION

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

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

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

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 73841(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

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

73841(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the [probability of causation] guidelines established under 

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

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

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

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

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

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

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

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

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

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

Section 2.0 of this Site Profile describes the Kansas City Plant (KCP) in Kansas City, Missouri. 

Section 3.0 discusses occupational medical doses from X-ray examinations, and Section 4.0 

discusses onsite environmental doses. Sections 5.0 and 6.0 describe internal and external dosimetry, 

respectively, in relation to KCP.
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2.0

KCP is on approximately 141 acres of the 300-acre Bannister Federal Complex 12 miles south of 

downtown Kansas City, Missouri, within the city limits. The Plant (Figure 1) currently shares the site

with the Federal Aviation Administration, Defense Finance and Accounting Service, U.S. Marine 

Corps, General Services Administration, Internal Revenue Service, National Oceanic and 

Atmospheric Administration, and National Logistics Support Center. Built by the Navy during World 

War II to assemble engines for Navy fighter planes, the facility was operated by Pratt-Whitney from 

early 1943 until September 2, 1945. In 1947, Westinghouse began leasing the facility and the Fairfax 

Storage Company used part of the building as a warehouse for tires, raw rubber, sugar, and lumber. 

In February 1949, the U.S. Atomic Energy Commission (AEC; a DOE predecessor agency) asked the 

Bendix Corporation to manage KCP. Table 1 summarizes the history of the facility. 

SITE DESCRIPTION

At present, KCP is a major operational facility administered by Honeywell Federal Manufacturing and 

Technologies/Kansas City. KCP produces non-nuclear weapons components for the nuclear 

weapons program of the National Nuclear Security Administration (NNSA). The principal products of 

KCP, since 1949, have included arming systems, fusing and firing systems, radars, power supplies, 

rubber, plastic and foam parts, and outer casings associated with non-nuclear components of nuclear 

weapons. KCP is currently the only NNSA facility for manufacturing non-nuclear components for 

nuclear weapons (DOE 1997). 

The KCP Office reports directly to DOE Headquarters and has line management responsibility for 

manufacturing non-nuclear components for the NNSA. In recent years, DOE has relocated additional 

project roles from the Mound, Pinellas, and Rocky Flats Plant facilities to KCP (Reis 1998).

2.1 FACILITIES

The Main Manufacturing Building, constructed in 1942, is the largest facility on the KCP site with 

approximately 2.6 million square feet of contiguous space that houses the key manufacturing 

operations. The following buildings support the Main Manufacturing Building: Polymer Building, High 

Power Laboratory, Mold Heating and Cooling Building, Plating Building and Spray Mask Facility, 

Technology Transfer Center, Special Processes Building, and Manufacturing Support Building. 
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Figure 1. KCP site map. 

Table 1. History of facility operations. 

Year Event

1943 Pratt-Whitney operated facility from early 1943 until September 2, 1945, to assemble engines for 

Navy fighter planes.

1947 Westinghouse began leasing facility, and Fairfax Storage Company used part of building as 

warehouse for tires, raw rubber, sugar, and lumber.

1949 KCP opens - AEC asks Bendix Corporation to manage facility and begin building non-nuclear 

components for nuclear weapons.

1982 Allied merger - Bendix merges with Allied Corporation.

1984 AlliedSignal merger - Allied merges with Signal companies to form AlliedSignal, Inc.

1993 Site consolidation - KCP designated as consolidated site for all non-nuclear components for nuclear 

weapons.

1999 Honeywell merger - AlliedSignal merges with Honeywell.

These facilities support the use of technologies for testing and evaluating engineering characteristics 

of materials and structures. Typical examinations include location of flaws, cracks, contaminations 

(chemical), and internal features; measurement of thickness, density, internal dimensions, and 

percentage contamination; sorting and identification of alloys; evaluation of heat treatment conditions; 
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and image enhancement. These capabilities are applied to a variety of objects including welds and 

forgings, foams, plastics, adhesives, composites, ceramics, and coatings as well as electrical 

component and mechanical assemblies.

Waste management operations at KCP consist mainly of hazardous nonradiological waste storage in 

preparation for offsite treatment or disposal. Operations can generate small quantities of low-level 

radioactive waste. Some waste is classified for national security reasons due to the nature of the 

generating process or constituents. Classified hazardous wastes are shipped off the site for 

sanitization and reclamation. KCP performs no onsite waste disposal. Treatment operations are 

limited to industrial wastewater pretreatment and selective recycling. 

2.2 OCCUPATIONAL SAFETY PROGRAM

The Plant has maintained a formal procedures manual (KCD 1958); Procedure S-306 of this manual, 

dated November 24, 1958, defines health and safety measures for receiving, handling and storing the 

radioactive materials. Sections in the procedure address area access control, physical requirements, 

protective apparel, personal hygiene, material handling, production area, scrap and waste disposal, 

and injuries. This procedure was revised in 1960 (Foster 1960) and in 1962 (Foster 1962). 

In 1965 an evaluation was conducted of 23 categories of workers for potential exposure to solvents, 

alcohols, etc., and to radiation. Table 2 summarizes this analysis for categories of workers potentially 

exposed to radiation (Schiltz 1965). There was a routine program of measuring uranium in urine (and 

lead in blood) for selected workers identified on a permanent access list to specified work areas (KCP 

undated). This list was used by the security, medical, and safety departments. Those on the list were 

authorized to work in areas in the Plant where radioactive material was handled (Nasca 2004a). This 

list is no longer in use. 

Table 2. Categorization of KCP workers in special physical examination program working with 

radiation (Schiltz 1965). 

Group Description Occupational exposuresa

I Radiation Workers - 

Departments 22, 34C, and 

217-22

Could be exposed to uranium oxide and litharge (lead) powders 0-20 

microns, small), and fumes from chlorinated hydrocarbon solvents.

IV X-ray Workers, Department 

213C

Could be exposed to X-ray radiation or gamma radiation from Cs-137 source 

(7 Ci). Some handle cured epoxy containing uranium dioxide and litharge.

VI Radiation Workers, 

Departments 267, 268, and 

280

Could be exposed to radiation in that they handle electronic tubes containing 

very small amount of radioactive material; they also handle drums that 

contain radioactive waste material (uranium dioxide) from Departments 22 

and 34C. 

X Electron Beam Welders, 

Departments 45, 585, 201, 

851, and Maintenance

Could be exposed to radiation from electron beam welders, which operate 

much like X-ray machines.

XV Personnel Handling Neutron 

Sources

There is none of this work at Plant at present; however, Test Laboratory 

could be working with neutron sources in near future.

XX Personnel Using Micro-Derm 

Thickness Gauge

Could be exposed to small amount of beta radiation.

a. Note that physicals normally include chest x-rays. Firemen and patrolmen receive physicals every 2 yr and have exit physicals. 

Cafeteria workers receive physicals every 3 yr but no exit physicals. All other groups receive an initial physical, annual periodic 

physical, and exit physical.

2.3 RADIATION MONITORING PROGRAM

The KCP Safety Department had primary responsibility for industrial hygiene and health physics. 

Potential exposure from X-rays and isotopic beta, photon, and neutron radiation sources represented 

a significant concern; however, the historical emphasis appears to have involved higher potential 
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nonradiological exposures to solvents such as acetone, methyl ethyl ketone, trichloroethylene, 

alcohol, toluene, and xylene; solvent fumes; epoxy resins and amine catalysts; polyester resins and 

peroxide catalysts, silicones, polysulfides, polyurethane, and tolylene isocyanate catalysts; and to 

fumes from acid, base, and salt chemicals of a wide variety. The focus appears to have involved 

ongoing evaluations of industrial, chemical, and radiological workplace hazards. There are records of 

frequent reviews of ionizing radiation equipment to evaluate the adequacy of shielding and interlocks. 

The earliest survey of radiation-generating devices (RGDs) apparently occurred in 1953 through the 

contracted services of a certified radiological physicist to evaluate the 1-MeV and 250-keV industrial 

X-ray units (Hoecker 1953). There are records of contamination surveys of the workplace using forms 

that imply a routine evaluation (Baldwin 1966). 

2.4 RADIATION SOURCES

KCP operations utilize radiation as one of the analytical tools to accurately manufacture, fabricate, 

and inspect non-nuclear components of nuclear weapons. Table 3 summarizes the types of 

radioactive sources at KCP during 1964 and 1987. Significant differences involve the extensive 

earlier use of radium sources and the presence of the 230-Ci 137Cs instrument calibration source in 

1987. The primary radiation sources at KCP involve analytical laboratory technologies for the 

manufacturing and testing of electronic and mechanical devices. Fabricated materials, parts, and 

assemblies are examined for internal flaws and defects to ensure compliance with engineering 

specifications and requirements. 

Table 3. Types of radiation sources.

Nuclide Use

Lists Predominant radiation

1964a 1987b Type Major energies, keV

C-14 Calibration source, beta scope, thickness 

gauges

x x Beta 45-156

Fe-55 Testing x Gamma bremsstrahlung to 230

Co-60 Calibration sources x Gamma 1,170, 1,330

Ni-63 Gas chromatograph x Beta 67

Sr-90 Calibration, thickness gauges x Beta 546 

Tc-99 Calibrations x Beta 292

Tl-204 Beta scope, thickness gauges x Beta 766

Cd-109 Plating thickness gauge x Gamma 88

Ba-133 Neutron generator x Gamma 30-382

Cs-137 Calibration sources x x Beta 514-1,176

Gamma 662

Pm-147 Calibration sources, thickness gauges x Beta 70-256

Tl-204 Beta scope, thickness gauges x Beta 766

Radium Calibration sources, thickness gauges x x Alpha 4,600-4,780

Gamma 186-610

Uranium Calibration sources Alpha/gamma progeny radiations 

from Th-230, Ra-226, etc.

Pu-239 Calibration sources, boron analysis x x Alpha 5,110-5,160

Gamma: 39-770

a. Source: Schiltz (1964). 

b. Source: KCP (1987). 

In addition to the radiation sources, radiographic inspections involve the use of numerous RGDs. 

Table 4 lists typical RGDs. The following list summarizes radiation-generating examinations.
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Table 4. Typical RGDs (Nasca 2004b).

Devices Energy (keV) Types Typical use Period of use

Industrial X-ray units 50 to 2,000 (X-ray) Rooms & Cabinets Radiography of parts 1950s to 2004

DXT Devicea 12 to 200 (X-ray) Cabinets Density thickness 1960s to 1980s

DXT Device Cs-137 (1 Ci) Cabinets Density thickness 1960s to 1980s

Electron Beam Welders 35 to 150 (X-ray) Cabinets Welding small parts 1960s to 2004

Electron Microscopes 30 to 200 (X-ray) Cabinets Analysis 1960s to 2004

Electron Beam Vacuum 

Deposition Systems

10 (X-ray) Cabinets Plating metals 1960s to 2004

Neutron generators 14.7 MeV (neutron) Open & Cabinets Generate microsec pulsed 

radiation

1960s to 2004

Neutron source Pu-239/Be (73 mCi) Cabinets Boron-10 analysis 1966 to 2004

Gamma camera Co-60 (19 Ci) Exposure Room Test electronic products 1950s to 1960s

Febetron Accelerator 2,300 (X-ray) pulser Cabinets Irradiation of electronic 

components

1970s to 1980s

Cesium Irradiators Cs-137 (230 Ci) Exposure Room Calibration of radiation 

detection instruments

1950s to 1980s

Medical X-ray 125 (X-ray) Exposure Room Patient diagnostic tests 1960s to 1990s

Electro Curtain 175 (X-ray) Cabinet Radiation curing of adhesives 1980s to 1990s

a. DXT refers to density times thickness measurement devices.

• X-Ray Radiography. KCP uses X-ray machines of different sizes and power to examine 

fabricated materials for structural defects such as voids or inclusions in weld or braze joints. 

These capabilities are also useful for failure analysis of electrical and mechanical assemblies. 

• Neutron Radiography. This is similar to X-ray radiography except it uses neutrons rather 

than X-rays. The primary advantage of neutron radiography is that high-density materials 

such as iron, lead, and uranium are nearly transparent to neutrons while low-density materials 

such as organic compounds are highly absorptive to neutrons. This technique is particularly 

well suited to examining foams, encapsulants, or seals through relatively thick metal cases.

• Gauging. Beta and X-ray techniques are used under a variety of applications to measure the 

thickness of materials. Utilizing the backscatter of beta radiation, thicknesses of platings and 

coatings can be measured.

KCP operations involve numerous analytical capabilities. Those involving ionizing radiation include 

nuclear magnetic resonance spectrometry, electron probe microanalysis, and scanning electron 

microscopy. Radiation protection surveys of equipment installations appear to have been an integral 

component of the safety program. For example, workplace surveys during 1960 and 1964 contain 

detailed evaluations of the installations, normal operating conditions, routine survey instrumentation, 

enclosures, safety features, and electrical safeguards (Schiltz 1960; Harrison, Meunier, and Schiltz 

1964a). 

2.5 RADIATION DETECTION INSTRUMENTATION

The KCP Industrial Hygiene Department appears to have had extensive radiation detection instrument 

capabilities. Table 5 reproduces the equipment listed in a memorandum that was apparently first 

written on May 19, 1964, and updated in August 1965 and March 1966 (Schiltz 1966). 

2.6 RADIOLOGICAL RECORDS

KCP has developed and maintained a radiological records database that contains records for all 

monitored worker exposures at KCP for all years of record. The database contains exposure data for 

about 4,400 workers. General trends in the recorded dose are:
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Table 5. Radiation detection equipment list.

a

Description No.

Victoreen Low Range Beta-Gamma Survey Meter, Model #592B, Range 0 - 1,000 mR/hr 2

Gas Proportional Alpha Survey Meters, Eberline Model #PAC-3G, Range 0 - 100,000 cpm 4

Beta/Gamma Survey Meter, Nuclear Chicago Corporation, Model 2586, Range 0 – 25, 0 – 250, and 0 – 2,500 

mR/h 

1 

Low-Range Beta-Gamma Survey Meters, Victoreen Model #389C, Range 0 - 20 mR/hr 4

Fast-Slow Neutron Survey Meter, Nuclear Chicago Model #2715, Range 10 - 104 neutrons/cm2 2

Air Proportional Alpha Survey Meter, Eberline Instrument Corp., Model #PAC-1A, Range 0 - 100,000 cpm 1

Tritium Monitors, Atomic Accessories, Model #TSM91, Range 0 - 100,000 mCi/cm3 of air of tritium 2

Gamma Survey Meters, Victoreen Model 61720, Range 0 - 500 R/hr 2

High Range Gamma Survey Meters, Eberline Instrument Corp., Model Gadora 1-B, Range 0 - 5,000 R/hr 2

Victoreen Condenser R-Meter Model #570 with assortment of probes for energy and dose 2

Victoreen Model 510 Roentgen rate meter with assortment of probes 1

Gamma Radiation Monitor, Eberline Instrument Corp., Model RM-2, Range 0 - 0.2 mR/hr 1

Beta-Gamma Count Rate Meters, Victoreen Model 743, Range 0 - 60,000 cpm. 2

Eberline Alpha Gas Proportional Floor Monitors, Model FM-2G, Range 0 - 100,000 cpm 2

Pocket Ionization Dosimeters, charge readers, and approximately 170 dosimeters with ranges from 0-100 R

Tritium Monitor, Atomic Accessories Corp., Model #TSM-91-C, BKC #29549 1

Alpha, Beta, Gamma Proportional Counter, Nuclear Measurements, Model PC3, Serial No. 275, BKC #14346 1

Proportional Counter Converter, Nuclear Measurements Corp., Model #PCC12A, Serial No. 145, BKC #17098 1

Cutie Pie Survey Meter, Model 2510, Nuclear Chicago, 0 - 2500 mR/hr 1

a. Source: Schiltz (1966) copied to Emergency Radiation Monitoring Team File.

• Positive deep, shallow, and extremity doses were first recorded in 1950.

• Positive neutron doses were first recorded in 1966. 

• Before about 1959, the recorded deep and shallow doses were essentially equal.

• Recorded extremity dose was higher in 1951 and 1952 than in any other year.

• Relatively high shallow dose (in comparison with deep dose) was recorded from 1959 to 1964 

and during 1973. 

• Recorded neutron dose is typically equal to recorded deep dose and to recorded shallow 

dose. The shallow dose is typically equal to the deep dose.

• There are comparatively few records with positive neutron dose.

• The only years with recorded positive uranium bioassay results are 1959 to 1971.

It appears that some dose has been recorded for nearly all occupational categories. The higher dose 

records were examined. Based on information from Nasca (2004d), some doses in the KCP 

radiological records system have been assigned to workers even though investigations at the time of 

measurement have shown unexplainable causes. Given the information available, these doses 

should be accepted as actual. 

Dose reconstructors should use recorded doses from the KCP database to supplement the hard-copy 

original dosimeter processing information submitted by DOE because the respective claims are often 

difficult to read. 
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3.0

KCP has a history of required routine medical examinations of workers that include routine chest 

X-rays. The medical X-ray units belonged to and were operated by the KCP Medical Department. 

There are handwritten records of chest X-rays being taken beginning on October 7, 1949 (Todd 

2004a). KCP conducted routine examinations that included medical laboratory screening, pulmonary 

function testing, vision screening, audiometric screening, vital signs, health history updates, and 

physical examinations (Todd 2004b). Before 1993, a chest X-ray was offered with every physical 

examination. Beginning in 1993, chest X-rays were offered every 5 yr or more frequently if worker 

history or physical circumstances indicated a need. In 1997, medical X-ray services were outsourced; 

thereafter, X-rays were not routinely offered as part of any physical examination with the exception of 

beryllium surveillance. The beryllium surveillance includes a chest X-ray every 5 yr (Todd 2004c). 

There is no evidence that photofluorographic techniques were used based on a review of X-ray films 

from the 1950s and 1960s. The first date of a chest X-ray being examined was April 12, 1950.

OCCUPATIONAL MEDICAL DOSE

Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAU 2003a) 

contains background information on X-ray doses. There is limited information about X-ray machine 

parameters pertinent to doses received by KCP workers. A radiation protection survey of the main 

medical X-ray unit on January 16, 1964 (Harris, Meunier, and Schiltz 1964b) identified the following:

• General Electric, Model 11 DA1, Serial No. 620871LDXC, 80 kVp maximum, 100 mA 

maximum, 0.5-mm Al inherent and 2.0-mm Al external filters with a cone.

• For chest exposures the beam is normally directed toward the patient at a distance of 72 in. 

For extremities, the beam is directed downward approximately 28 to 36 in. from the tabletop. 

• The normal operating settings (assumed for chest X-rays) were 70 to 870 kVp and 100 mA for 

0.1 s.

• Measured exposure to evaluate collimation of the beam for the Series II (chest X-ray) 

examination (75 kVp, 100 mA, 0.2 s) showed a well-collimated beam with a maximum 

exposure of 26.3 mR.

• There were instructions on focusing the light field on the patient or extremities so the minimum 

body area would be exposed to the X-ray beam. The useful X-ray beam is further collimated 

by adjusting the size of the light field. 

This implies an entrance skin exposure (ESE) of approximately 13 mR measured in 1964 for the usual 

chest posterior-anterior (PA) examination (i.e., 70 to 80 kVp, 100 mA, 0.1 s). Measurements of the 

ESE before 1964 were not found. 

3.1 DOSE RECONSTRUCTION

The DOE dose reports for each claim should provide the schedule of actual X-ray examinations. KCP 

routine practices were apparently to conduct periodic physical examinations depending on the 

worker’s job classification as follows:

• Firemen and patrolmen receive physicals every 2 yr and have exit physicals.

• Cafeteria workers receive physicals every 3 yr and have no exit physicals.

• All other groups receive an initial physical, annual periodic physicals, and an exit physical.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 15 of 46

Dose reconstructors should use the following approach based on current information to assess dose 

from medical chest X-rays: 

1. Before 1964, use the default values in Tables 3.3-1 and 4.0-1 of ORAU (2003b) for chest Xrays. 

2. Beginning in 1964, use an ESE of 13 mR.

3. If there is no reasonable information in the claim documentation, assume an annual PA 

radiographic chest X-ray for each claimant’s employment period from 1949 through June 

1993. 

4. If there is no reasonable information in the claim documentation, assume a PA radiographic 

chest X-ray for each 5-year period beginning in July 1993 for each claimant’s employment 

period.

Organ Dose

Organ dose calculations for workers at KCP from 1949 to the present involve only the 14- by 17-in. 

PA chest film. The analysis in this Site Profile evaluated only doses from this technique. Other 

radiographic examinations of KCP employees that could have occurred are likely to be 

nonoccupational in the sense that they were associated with illness or injury and were not part of a 

routine examination process. There is no indication in the examined records that other diagnostic 

radiographic examinations routinely occurred as part of the occupational medical program. 

Before 1964, the default organ dose recommendations for PA chest examinations in ORAU (2003a) 

were used (i.e., kerma of 2 rem) with minimal collimation due to the lack of detailed information 

concerning measurements of the ESE and collimation for KCP X-ray equipment. Historic 

documentation of measurements on January 16, 1964 does show a well-collimated system with an

ESE of 13 mR. Tables 6 and 7 summarize organ doses for the pre-1964 and 1964-2004 periods, 

respectively, for PA chest examinations. This analysis assumed that an exposure of 1 R is equivalent 

to a kerma of 1 rad, 10 mGy, or 1 rem (ORAU 2003a).

3.2 UNCERTAINTY ANALYSIS

The description of error and uncertainty in ORAU (2003a) is directly applicable to evaluation of 

medical X-ray dose to KCP workers. Given the information available, and the apparent unavailability 

of measurements of the ESE for the specific X-ray equipment and diagnostic technics used before 

1964 at KCP, the default options from ORAU (2003a) should be used. ORAU (2003a) recommends 

that dose reconstructors assume that errors are all positive (i.e., use only +30%). 

3.3 INSTRUCTION GUIDE FOR DOSE RECONSTRUCTORS

KCP practice involved maintaining a log of the date and type of medical examination and a record of 

the associated radiography films. DOE-provided medical X-ray dose information should include a 

copy or transcription of information from this log. Dose reconstructors should compare the frequency 

of examinations in this log with the default assumptions listed in Table 8. 

Dose reconstructors should use the greater number of examinations from the log or from Table 8 in 

preparing their dose calculations. A normal distribution should be assumed with an uncertainty of 

±30% at the one-sigma confidence interval. Dose reconstructors may multiply the doses listed in 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 16 of 46

Table 6. Organ dose estimates for PA chest radiographs before 1964 assuming 

minimal collimation. 

Organ

Dose conversion factor (mGy/Gy air kerma)

(beam quality for 2.5 mm Al HVL)a Organ dose (rem)

Thyroid 174 3.48E-02

Eye/brain 32 6.40E-03

Ovaries 168 2.5E-02

Urinary bladder 168 2.5E-02

Colon/rectum 168 2.5E-02

Testes 9.1 5.0E-03

Lungs 451 9.02E-02

Thymus 451 9.02E-02

Esophagus 451 9.02E-02

Stomach 451 9.02E-02

Bone surfaces 451 9.02E-02

Liver/gall bladder/spleen 451 9.02E-02

Remainder 451 9.02E-02

Female breast 49 9.80E-03

Uterus 149 2.5E-02

Bone marrow 92 1.84E-02

Skinb 2.72E-01

a. HVL = half-value layer; dose conversion factors (DCFs) from ORAU (2003a) and ICRP Publication 

34 (ICRP 1982, Tables A.2 to A.8)

b. Skin dose was determined by multiplying ESE by a backscatter of 1.36 for an HVL of 2.5 mm Al 

from NCRP Report 102 (NCRP 1989, Table B-8).

Table 7. Organ dose estimates for PA chest radiographs from 1964 to 2004. 

Organ

Dose conversion factor (mGy/Gy air kerma)

(beam quality for 2.5 mm Al HVL)a Organ dose (rem)

Thyroid 32 4.16E-04

Eye/brain 32 4.16E-04

Ovaries 1 1.30E-05

Urinary bladder 1 1.30E-05

Colon/rectum 1 1.30E-05

Testes 0.01 1.30E-07

Lungs 451 5.86E-03

Thymus 451 5.86E-03

Esophagus 451 5.86E-03

Stomach 451 5.86E-03

Bone surfaces 451 5.86E-03

Liver/gall bladder/spleen 451 5.86E-03

Remainder 451 5.86E-03

Female breast 49 6.37E-04

Uterus 1.3 1.69E-05

Bone marrow 92 1.20E-03

Skinb 1.77E-02

a. Dose conversion factors (DCFs) from ORAU (2003a) and ICRP Publication 34 (ICRP 1982, 

Tables A.2 to A.8)

b. Skin dose was determined by multiplying ESE by a backscatter of 1.36 for an HVL of 2.5 mm Al. 

From NCRP Report 102 (NCRP 1989, Table B-8).

Table 8. PA chest occupational X-ray frequency. 

Examination Frequency

Before 1993 Chest X-ray offered with annual physical examinations.

Beginning in 1993a Chest X-rays offered every 5 yr.

a. X-rays could be more frequent if a worker’s history or physical circumstances 

indicated.
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 17 of 46

Tables 21 and 22 in Section 6.0 by a factor of 1.3 (i.e., 30% increase) to include uncertainty (ORAU 

2003a) in the analysis. 

4.0

KCP prepared a site safety assessment report in 1995 in compliance with DOE requirements (AlliedSignal 1995); the report concluded that Plant operations had produced no undue hazard to the 

general public and no significant effect on the environment. It appears that there are minimal, if any, 

historical radiological effects. Air and water effluents have been monitored routinely to assess 

compliance with relevant criteria. KCP routinely handles hazardous chemicals, but there is limited 

handling of radioactive materials and thus essentially little likelihood of a significant occupational 

environmental exposure associated with releases. The Plant does generate low-level radioactive 

waste that includes equipment radiation sources, tritium exit signs, irradiated components, gap tubes, 

smoke detectors, and small amounts of cleanup materials and personal protective equipment. Mixed 

wastes consist of encapsulated electronic assemblies and spent solvents used in cleaning and 

decontaminating radioactive materials (primarily defective sources). There is no offsite environmental 

exposure. As such, variability calculated from background control personnel dosimeters (i.e., 2-sigma 

= 25 mrem) is recommended as a claimant-favorable option to assign external dose to unmonitored 

workers, as described in Attachment A.

OCCUPATIONAL ENVIRONMENTAL DOSE

The primary and apparently only substance handled in large quantities with a potential for significant 

environmental release involved depleted uranium (DU) from 1958 through 1970. As described in 

Attachment A, the recommended approach to assign an internal environmental dose to unmonitored 

nonradiological workers is to assume an annual intake fraction of 0.01 times the median annual 

measured air concentrations in Table 11 (i.e., multiply annual concentrations in Table 11 by 

6.58 × 1010 to obtained pCi/day). The 0.01 factor is a “claimant favorable” estimate of the decrease in 

concentration based on dispersion between inside and outside air concentrations. For the period 

following 1970, there are no workplace air sampling data available for analysis. Since there is the 

potential for unknown small releases or residual contamination, such as the incident with 147Pm in 

1989, the same 0.01 factor is recommended using the average median annual measured air 

concentration (for all years) from Table 11. Beginning in 1990, there is assurance that no significant 

environmental releases occurred based on the multi-agency response findings to the 147Pm incident. 

5.0

KCP has handled numerous types of radioactive sources (Table 3). Most of these sources are sealed 

or contained as components of various types of equipment. No incidents involving confirmed internal 

intakes have been reported (Lund 2004). The primary workplace exposure was associated with 

machining items containing DU oxide from 1958 to about 1971; the program that used DU oxide 

ended in 1972 (Nasca 2005a). No uranium was processed until 1997 when a new program was 

initiated. From 1997 onward, DU metal was reduced in size and shape by an electrochemical process

that involves the placement of DU metal in an acid bath. The parts are rinsed with water and dried 

before handling. Because the uranium does not become volatile during the electrochemical process, 

remaining in the acid solution, there is minimal personnel internal dose hazard with this process. 

There is no removable contamination with this process (Nasca 2005a). Bioassay data (urine assays) 

and air monitoring data were found only for 1959 through 1971. Intakes of DU after 1972 are not 

likely. 

OCCUPATIONAL INTERNAL DOSE
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 18 of 46

5.1 DEPLETED URANIUM 

The uranium at KCP was DU. For internal dose calculations, use the NIOSH accepted default

isotopic ratios for DU that are found in the Interactive Modules for Bioassay Analysis (IMBA) computer 

program. Table 9 lists typical weight percents and activity fractions of uranium isotopes. When 

estimating intakes, dose reconstructors should assume that exposures to uranium dust were chronic.

5.1.1

KCP had substantial quantities of UO2 on the site at various times. Order number ICO-020757 

(Bendix 1962) shows that UO2 was ordered in 10,000-pound lots. The relevant specifications from 

Physical and Radiological Characteristics 

Table 9. Mass and radiological characteristics of DU. 

Isotope

Weight 

percentagea

Specific constituent activity in mixtureb

Bq/mg pCi/mg dpm/mg

U-234 0.0010 2.3116 62.4757 138.696

U-235 0.1991 0.1592 4.3028 9.5523

U-236 0.0003 0.0072 0.1941 0.4308

U-238 99.7996 12.4111 335.4345 744.6647

Totals 14.889 402.4071 893.3437

a. From IMBA computer program (Birchall et al. 2003).

b. Could vary from IMBA values due to rounding.

Specification Control No. 4542260-00 (KCP 1998) were that the minimum density should be no less 

than 10.8 g/cm3

, the surface area of the powder should be no greater than 1.1 m2/g, at least 97% by 

weight of the material should be less than 10 μm in diameter, and 100% by weight should be less than 

15 μm in diameter. These specifications are consistent with a powder having an activity median 

aerodynamic diameter (AMAD) of 1.175 μm and a geometric standard deviation (GSD) of 2.48. 

In fitting bioassay data, dose reconstructors may wish to start with a 1-μm AMAD, a GSD of 2.5, a 

density of 10.97 g/cm3

, a lung solubility Type S, and f1 of 0.002 (ICRP 1995). However, it is not 

known what impact processes at KCP had on the particle size of uranium. Use of the default 5-μm 

AMAD particle size is also acceptable unless it is known that the intake was of unaltered UO2 powder. 

5.1.2

Table 10 summarizes alpha radiation contamination results from 1962 to 1969 in DU work areas 

considered to be most significant to potential worker exposure (Nasca 2004b). In addition, from 1958 

to 1970 KCP workplaces were routinely monitored with air samplers for DU concentrations (Nasca 

2004e). Table 11 lists the maximum measured workplace concentrations used to calculate the 

median and 95th-percentile statistical parameters.

Workplace Monitoring 

Table 10. Alpha contamination levels, 1962 to 1969 (Nasca 2004b).

Facility Work area

Measured levels (dpm/100 cm2)

Average Maximum

D/34C (D/27C) Air lock 226 20,000

Locker room 570 7,000

General area 2,564 45,000

D/220-22 (D/443-20D

D/216-22, D/217-20D)

Air lock 190 800

Wash-up 180 650

General area 425 1,000

D/22, D/20D Air lock 206 350

Clean area 468 2,000

General area 892 16,000
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 19 of 46

Eating and smoking were prohibited in DU processing areas. However, at least one survey report 

indicated the presence of cigarette butts, candy wrappers, and coffee cups in the exclusion area 

(Baldwin 1966). This indicates that ingestion of DU was a possible route of occasional intake. Dose 

reconstructors may assume inhalation or ingestion intakes to be claimant-favorable. 

The parameters in Table 11 were based on maximum measured air concentrations at several 

locations in the plant. Some measurement locations were identified only by a number. When 

locations were indicated, most were near the walls of the work areas. Other locations were labeled as 

“Mixing Rm, West,” “Air Lock,” “Over Shower,” “Rubber Mill Rm,” “Mill Stack West,” and “Dispersion 

Roll.” 

Table 11. Statistical parameters of measured DU in workplace air.

a,b

Year

KCP measured resultsa Lognormal fit

No. of 

measurements

(μCi/cm3) Air concentration (μCi/cm3)

Mean Maximum Median 95% GSD

1958 22 7.18E-12 4.90E-11 4.01E-13 1.74E-10 4.02E+01

1959 27 8.82E-13 1.22E-11 2.89E-13 2.53E-12 3.74E+00

1960 33 1.32E-12 1.50E-11 3.41E-13 3.94E-12 4.43E+00

1961 31 1.00E-12 2.04E-11 1.97E-13 1.52E-12 3.46E+00

1962 31 7.73E-13 1.13E-11 2.50E-13 2.03E-12 3.58E+00

1963 31 1.25E-12 1.63E-11 2.47E-13 1.90E-12 3.46E+00

1964 31 2.21E-12 3.90E-11 3.91E-13 2.98E-12 3.44E+00

1965 31 1.99E-13 8.70E-13 1.05E-13 8.02E-13 3.45E+00

1966 23 7.01E-13 6.24E-12 2.00E-13 2.00E-12 4.06E+00

1967 22 1.40E-12 1.30E-11 5.70E-13 3.12E-12 2.81E+00

1968 19 1.21E-12 9.88E-12 2.31E-13 3.47E-12 5.19E+00

1969 19 1.88E-11 8.55E-11 3.88E-12 1.42E-10 8.92E+00

1970 19 7.32E-14 5.91E-13 4.02E-14 1.98E-13 2.64E+00

Average 1958-70 2.85E-12 2.15E-11 5.49E-13 2.62E-11

a. All departments.

b. Based on maximum measured KCP workplace airborne uranium concentrations at several 

monitoring locations.

5.1.3

KCP workers were individually monitored for DU intake from 1959 to 1971 using a fluorophotometric 

method to measure the level of uranium in urine. KCP (1962) states that the method is sensitive to 

concentrations of uranium from 1 × 10-10 to 5 × 10-11 g per 0.25 g of sodium fluoride with a precision of 

±10% (KCP 1962). This sensitivity equates to 0.5 to 1 μg U/L of urine. However, this sensitivity may

be the theoretical best based on ultra pure water blanks, as opposed to urine blanks. The urine 

volume used for bioassay analysis was 0.1 ml and the lowest quantity of uranium used to determine a 

standard curve was 1 × 10-9 g. The urine concentration that equates to the lowest uranium quantity of 

the standard curve is 10 μg/L. Bioassay data for four individuals (Nasca 2005b) shows that 

concentrations as low as 1 μg U/L were recorded. However, most sites using fluorophotometry at this 

time were claiming more modest detection levels; for instance, Hanford - 4 μg/L (ORAU 2004a), 

Paducah – 10 μg/L (ORAU 2004b), University of Rochester (used by many AWEs) 5 – 10 μg/L 

(ORAU 2005). Because a definitive statement of the detection limit achieved by the Kansas City Plant 

was not found, an MDA of 10 μg/L is recommended.

Bioassay

The frequency of bioassay analysis for KCP personnel who worked with DU powders is not known. 

The available data are shown on an annual basis and might be the sum of one or more bioassay 

measurements. If the individual case information does not yield additional information, dose 

reconstructors should make the claimant favorable assumption that the recorded bioassay quantities 

represent a single bioassay measurement taken at the end of the calendar year.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 20 of 46

By procedure (KCP undated), the workers in radiation areas 20D, 34C, and 443E-20 were to receive 

uranium-in-urine bioassays semiannually (May and November). Bioassay data were recorded on 

either the individual’s film badge envelopes or the annual 3-in × 5.5-in. radiation exposure record 

(Nasca 2005b). The actual frequency of bioassay analysis for KCP personnel who worked with DU 

powders varied from person to person and from year to year (Nasca 2005b). The date and results of 

individual bioassay results are in each individual’s dosimetry file. 

The only bioassay data that are available electronically (Nasca 2004c) are the numeric sums of all 

bioassay measurements taken during the year for an individual and could be the sum of one or more 

bioassay measurements. The number of bioassay measurements that comprise the annual sum is 

not recorded in the electronic database. Bioassay data from four individuals were studied (Nasca 

2005b) and the number of bioassay samples taken per year ranged from 0 to 6. Zero samples per 

year means that there are gaps in the bioassay record where one or more years of no bioassay data 

are bracketed by years for which there are bioassay data. In 1960 and 1961, there appear to have 

been more urine samples collected and the results of the samples were greater than for other years. 

No information was found to explain why the bioassay data for 1960 and 1961 were greater than for 

other years. There were no records of an incident involving DU powder in those years (Nasca 

2005a).

Table 12 summarizes an analysis of the electronic bioassay records (Nasca 2004c). These data 

show a peak in 1960 and 1961. The peak is apparently not the result of an incident (Nasca 2005c) 

but could be due to a large number of bioassay samples being collected in those years for each 

worker. The bioassay data for 1971 are very low, indeed, they are less than the sensitivity level. The 

low bioassay levels may indicate that no intakes of uranium occurred during that year. As noted 

above, the excreta data in Table 12 represent the sum of an unstated number of bioassay 

measurements. An improved co-worker analysis on bioassay data is scheduled as part of the overall 

plan for co-worker analyses of DOE sites.

Table 12. Statistical parameters of recorded DU in urine.

a

Year

Recorded annual urine concentrationb Lognormal fit

Chronic intakes (pCi/d)c No. of workers 

reported 

Concentration (μg/L) Concentration (μg/L)

Mean Maximum Median GSD 5th Median 95th

1959 214 4.125 52.60 2.642 2.675 1.05E+02 6.42E+02 3.92E+03

1960 281 36.58 140. 19.53 3.813 7.79E+02 4.75E+03 2.89E+04

1961 123 51.40 192.1 37.44 2.402 1.49E+03 9.10E+03 5.55E+04

1962 148 4.327 15.75 3.162 2.508 1.26E+02 7.69E+02 4.69E+03

1963 211 10.96 72.00 7.564 2.532 3.02E+02 1.84E+03 1.12E+04

1964 219 5.627 78.38 3.888 2.431 1.55E+02 9.46E+02 5.76E+03

1965 175 9.572 38.00 5.583 3.422 2.23E+02 1.36E+03 8.27E+03

1966 223 6.432 45.05 4.214 2.640 1.68E+02 1.02E+03 6.24E+03

1967 159 5.438 21.50 3.574 2.713 1.43E+02 8.69E+02 5.30E+03

1968 11 6.055 6.600 6.052 1.029 2.42E+02 1.47E+03 8.97E+03

1969 1 0.15 0.150 0.150 1.000 5.99E+00 3.65E+01 2.22E+02

1970 59 11.64 45.00 7.576 2.686 3.02E+02 1.84E+03 1.12E+04

1971 47 0.03596 0.1000 0.02993 1.903 1.19E+00 7.28E+00 4.44E+01

ALL 1,871 14.1 192.1 5.5 4.7

a. All bioassay measurements.

b. The recorded annual sum of urine concentration is the sum of all bioassay results for the year. There is one sum for 

each person-year record. The listed statistics are based on the analysis of the data, which are the sums of all bioassay 

data for every person for that year.

c. Chronic intakes that produce the urinary excretion per day on the 365th day of intakes corresponding to the median 

excretion from the lognormal fit and 5th and 95th percentile intakes using a GSD of 3. Assumes 5-μm AMAD particle 

size; intakes for 1-μm AMAD particle size, 10.97 g/cm3 density, and absorption type S are smaller.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 21 of 46

Because these data are the sum of all bioassay measurements for each person in each year, they 

represent the maximum bioassay result possible. Also the default 5-μm AMAD particle size was used 

for the intake estimates; however, using the more likely 1-μm AMAD particle size and density of 

uranium oxide results in smaller intakes. Hence, the intakes in Table 12 are overestimates so long as 

a whole year’s exposure is used. The different percentile intakes are used with types of exposures 

explained in Section 5.1.4. 

Due to the nature of the work performed at KCP, and because no accident reports have been found, it 

is reasonable to assume that intakes of DU from 1959 through 1971 were chronic unless the 

individual’s dosimetry records indicate otherwise.

The persons who received bioassay for uranium had Organization Codes 530001 and 531002. Table 

13 lists the number of bioassay results for the two organizations and the occupations that had 

bioassay results. The Organization Code does not provide much information about what groups 

might have been exposed to airborne uranium because Code 531002 refers to the DOE contractor 

that operated KCP (e.g., Bendix, Allied-Signal), and Code 530001 refers to DOE workers. The data in 

Table 13 indicate that nearly all types of workers might have been exposed to uranium at KCP. The 

information in Table 13 may be of some benefit to dose reconstructors if a worker’s job description 

and organization code are known. However, see Section 5.1.4 for more general instructions for 

unmonitored workers. 

Table 13. Number of recorded bioassay measurements and average of measurements for 1959 

through 1971. 

Occupation description

Occ 

Code

Number of individual measurements Bioassay 

Organization code measurements (μg/L)

530001 531002 Total Totala Ave.b

Managers and administrators 110 10 342 352 3,210.44 9.12

Engineers 160 228 228 2,017.84 8.85

Scientists 170 15 15 35.16 2.34

Health physicists 184 3 3 12.90 4.30

Miscellaneous professionals 200 17 17 149.64 8.80

Repair technician 350 44 44 580.58 13.19

Health technician 360 3 3 16.17 5.39

Technologist, engineering 370 38 38 519.03 13.66

Miscellaneous technicians 390 47 47 418.52 8.90

Administrative support/clerical/work planners 450 60 60 717.74 11.96

Fire fighter 512 58 58 228.33 3.94

Security inspector & guard 513 3 4 7 11.25 1.61

Food service employees 521 1 1 0.00 0.00

Custodian/janitor 524 11 11 31.33 2.85

Mechanics/repairers 610 350 350 2,687.45 7.68

Electrician 643 153 153 1,371.99 8.97

Pipe fitter 645 177 177 1,566.47 8.85

Machinist 681 152 152 2,140.77 14.08

Sheet metal worker 682 2 2 0.00 0.00

Operators, plant/system/utility 690 2 2 2.45 1.23

Machine setup/operators 710 5 28 33 262.95 7.97

Welders/solderers 771 11 11 77.10 7.01

Miscellaneous precision/production workers 780 14 640 654 6,894.03 10.54

NAc 781 1 1 14.05 14.05

Drivers 840 28 28 217.13 7.75

Handlers/laborers/helpers 850 145 145 1,279.55 8.82

All groups 2,592 24,472.87 9.44

a. Grand total of all bioassay measurements for that occupational code for all years (1959-1971).

b. Average of recorded bioassay measurements.

c. NA – Information not available.
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 22 of 46

Table 13 also lists the sum of all bioassay results, for all years, for each occupational code and the 

average of the bioassay measurements for each occupational code.

The electronic record of uranium bioassay contains the sum of all measurements performed during 

the year for the particular individual. The electronic record does not indicate the number of samples 

that comprise the recorded measurement. The results of individual measurements were written on 

cards that are nearly illegible. For these reasons, KCP should be a candidate site for a co-workers 

study. The Managers and Administrators occupation category included individuals who performed job 

estimates and who were commonly in the work areas.

5.1.4

Table 12 lists the statistical parameters of measured DU concentrations in the urine of KCP workers

for the years for which urine data were found and associated intakes. Table 13 provides some 

insights into job categories. The uncertainties in the intakes in Table 12 do not warrant assigning 

specific intakes to each job category; however, the data in Table 12 do lend themselves to grouping 

workers into 4 exposure categories: (1) workers routinely exposed to airborne or loose material, (2) 

workers occasionally exposed, (3) workers rarely exposed or exposed only to very low workplace 

airborne or contamination levels, and (4) workers with little or no potential for radiological exposure. 

Unmonitored workers in category 1 should be assigned the 95th percentile intakes; category 2 to the 

median intakes; category 3 to the 5th percentile intakes; and category 4 workers should be assigned 

internal exposures per the environmental section of this document. The dose reconstructor should 

use the information in Table 13, the worker’s radiation exposure file, and the computer assisted 

interviews, and Technical Information Bulletin 014 (ORAU 2005c) to assign unmonitored workers to 

exposure categories. Generally, the occupations for which bioassay data have not been found are 

nurses, miscellaneous repairers/ construction workers, and equipment operators. 

Unmonitored Worker

Due to how Table 12 intakes were calculated, they must be applied to the whole year to ensure 

claimant-favorability, even if the energy employee worked a partial year. 

Dose reconstructors will likely not find bioassay results in 1958 yet the air sample data indicate that

airborne contamination existed. For 1958 for category 1 workers assume chronic intakes of 5,430

pCi/d; for category 2 workers assume chronic intakes of 891 pCi/d; for category 3 workers, 146 pCi/d. 

The median intake for 1958 was determined by ratio of the median air concentrations for 1958 / 1959 

from the lognormal distributions in Table 11 times the median intake for 1959 from Table 12. The 

default GSD of 3 was applied to the median intake for 1958 to determine the 5th and 95th percentile 

intakes for 1958. This is consistent with the GSD for the intakes for the other years in Table 12. 

Because several maximizing assumptions were used to determine the intakes in Table 12, the 

uncertainty distribution of doses determined from these intakes should be entered into IREP as 

constants. 

5.2 OTHER NUCLIDES

There is apparently no indication or expectation of significant worker intakes of any other nuclide. The 

response to an incident on February 10, 1989 involving 147Pm initially concluded, apparently in error, 

that an intake had occurred according to a report of what appears to be an extensive investigation (KCP 

1989). According to this report, the Department of Energy, Environmental Protection Agency and the 

Missouri Department of Health, Radiological Health Division representatives were notified at the time 

of this incident. A DOE team of investigators arrived at KCP on February 14 to assume technical 

management of the situation. Based on what eventually were determined to be false-positive 
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 23 of 46

bioassay results, the homes of 4 KCP workers were inspected and some contamination was found. 

There were undoubtedly many activities to identify the cause and extent of this contamination. Pm147 is a relatively low-energy 100% beta emitting nuclide (e.g., maximum energy of 224.7 keV, 

average energy of 62 keV) with a half-life of 2.6 years. The primary concern with this type of nuclide 

would be direct skin contamination and intake. The report addresses the chronology of steps, and the 

results, to examine potential exposure and contamination. KCP operations do include the use of 

several radioactive sources that can be fragile and relatively unsealed, as occurred with the 147Pm

source incident. However, any intakes from KCP use of various small sources would probably be

comparatively insignificant.

6.0

Information concerning the early history of KCP nuclear weapons assembly activities involves 

classified information; therefore, a clear description of events at that time is not publicly available. As 

described in Section 2.0, AEC operations at KCP began in 1949. The primary work activity involving 

external radiation exposure was fabrication and quality control testing of non-nuclear components of 

nuclear weapons. 

OCCUPATIONAL EXTERNAL DOSE

6.1 BASIS OF COMPARISON

Since the initiation of the Manhattan Engineer District (MED; a DOE predecessor agency) project in 

the early 1940s, various concepts and quantities have been used to measure and record occupational 

radiation dose at the many MED/AEC/DOE facilities. A common basis of comparison has been 

selected to assess the consistency of the available historical recorded dose at KCP with current KCP 

dosimetry performance. With the known dates of changes in KCP dosimetry systems, comparisons of 

recorded doses before and after these changes provide an ability to assess consistency. Similar 

sources have been used to calibrate and conduct performance testing of dosimetry systems (AEC 

1955; Unruh et al. 1967; McDonald et al. 1983). This basis, to be used in dose evaluation or 

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

whole-body dose, d is 10 mm and Hp(d) is noted as Hp(10). The International Commission on 

Radiological Units and Measurements has recommended both Hp(0.07) and Hp(10) for use as the 

operational quantities to be recorded for radiological protection (ICRU 1993). These are the radiation 

quantities used in the DOE Laboratory Accreditation Program (DOELAP) since the 1980s to accredit 

personnel dosimetry systems in the DOE complex, including KCP (DOE 1986). 

6.2 WORKPLACE EXTERNAL RADIATION FIELDS

Section 2.4 describes the primary sources of workplace radiation fields at KCP. These sources are 

historically associated with processes involving industrial RGDs (X-rays and electron accelerators); 

isotopic beta, gamma-ray, and neutron radiation-emitting sources; and DU. The isotopic sources are 

typically used in the manufacturing or quality control processes to monitor fabrication of non-nuclear 

weapons components. 

6.2.1

Beta radiation fields in KCP workplaces are associated with the sources of beta-emitting nuclides 

listed in Table 3, uranium handled from about 1958 to 1971, and RGD-produced electrons identified in 

Tables 2 and 4. Potential exposures to beta radiation sources are typically limited to maintenance 

activities or failures. For example, a 147Pm source was used to measure the thickness of a film, and 

Beta Radiation
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failure of the integrity of this source did result in some worker exposure and was the subject of an 

official investigation (KCP 1989). Worker exposure to electrons would typically be associated with an 

error or equipment failure. 

6.2.2

Photon (X-ray and gamma) radiation associated with KCP work activities covers a broad energy 

range. Sources of ionizing radiation at KCP listed in Table 3 include several radioactive sources that 

emit photon radiation. These were typically used to check or calibrate processes to gauge 

thicknesses, perform instrument calibrations, and so forth. Many sources were beta and photon 

emitters and are of the types and source strengths typically used by mainstream industrial or processrelated users. In general, doses associated with the proper, and widespread, use of these small 

check sources are comparatively negligible. In addition, KCP used a variety of larger photon-emitting 

radiation sources and X-ray RGDs, as listed in Table 4. These were typically used to radiograph parts 

or perform operational tasks. There were numerous X-ray machines in locations around the Plant.

Photon Radiation

The photon energy spectra in KCP workplaces have not been measured. The spectra are related to 

the configuration of the X-ray machines, the process, and the extent of shielding. However, 

regardless of the precise spectra, significant photon radiation would have been readily measured at 

KCP by the available dosimeter technology during all years of operation. 

6.2.3

Sources of neutron radiation at KCP involve neutron generators and nuclide (alpha, neutron) 

interactions, such as 239PuBe sources, as listed in Table 4. The first presence of neutron-emitting 

nuclides apparently occurred after 1965 based on a review of workplace hazards (Schiltz 1965).

Neutron Radiation 

6.3 DOSIMETER TECHNOLOGY

KCP has historically used beta/photon and neutron dosimeters to measure potential radiation 

exposure of personnel. 

6.3.1

Beta and photon (X-ray and gamma) dosimeters used at KCP include:

Beta/Photon Dosimeters

• Pocket ionization chamber (PIC) 

• KCP Film Dosimeter (Bendix 1964) – a two-piece, stainless-steel, film holder with front and 

rear matching rectangular windows. The front and rear faces have a 1-mm-thick cadmium 

filter in the area of the open window. Two different types of personal dosimetry film packets 

are used – Kodak Type 2 for mixed beta and gamma radiation and DuPont Type 558 for X-ray 

and gamma radiation. The DuPont 558 film packet contains two films, Types 508 and 1290. 

The type 508 film has a range of 0.003 to 30 R, and Type 1290 has a range of 0.68 to 

3,000 R. KCP calibration data using uranium and 60Co sources are available (Bendix 1964). 

• KCP (two-chip, TLD-100) thermoluminescent dosimeter (TLD) – used from 1973 to 1982.

• Eberline standard (three-chip, TLD-100) TLD – used from 1983 through 1990. This dosimeter

employs one chip under a 10-mg/cm2 filter to measure the shallow or skin dose and one or two 

chips under a 285-mg/cm2 filter to measure the deep or whole-body dose (TMA 1990). The 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 25 of 46

use of this dosimeter was terminated due to its inability to pass DOELAP lower energy photon 

performance testing categories although it had passed National Voluntary Laboratory 

Accreditation Program (NVLAP) testing (Allied-Signal 1991).

• The Landauer K1 (three-chip, TLD-700) TLD – use began April 1, 1991, to measure beta, Xray, and gamma radiation exposure to KCP workers.

• Landauer optical stimulated luminescent (OSL) aluminum-oxide dosimeters – use began in 

2000.

6.3.2

KCP used the Landauer Neutrack I dosimeter to measure neutron doses. This dosimeter is a 

polycarbonate (Lexan) neutron recoil track registration device used to monitor fast neutron 

interactions. The Lexan responds to neutrons by recording ionization damage caused by neutron 

interactions with carbon and oxygen atoms, which leaves a track. It has a uniform energy response 

from 3 to over 14 MeV with a threshold of 1 MeV.

Neutron Dosimeters

The dosimeters were used to record the official dose of record and the PIC was used to provide 

administrative control until dosimeter results were available. Table 14 summarizes the use of 

personnel dosimetry techniques and exchange frequencies. 

Table 14. Personnel dosimetry systems. 

Period Description Dose measured

Routine exchange 

period

Beta/photon dosimeters

1950 - 1954 KCP in-house film badge 

system

Beta/nonpenetrating

X-ray/gamma

Weekly

1955 Biweekly

1956 - 1964 Monthly

1965 -1972 Bimonthly

1973 - 1982 KCP two-chip TLD Bimonthly

1983 - 1990 Eberline three-chip TLDs Quarterly

1991 - 2000 Landauer three-chip TLDs Quarterly

2001 - 2003 Landauer OSL Semiannual

Neutron dosimeters

1961 - 1967 Controls for Radiation, Inc. Neutron Biweekly

1967 - 1973 Landauer, film

1974 - 1982 Landauer NTA film Quarterly

1983 - 1990 Landauer Lexan track-etch

1991 - 2000 Landauer Neutrak I Poly carb

2001 - 2003 Landauer Neutrak 144 Semiannual

KCP participated in DOELAP performance testing using Landauer-provided services beginning in 

October 1992, and most recently passed DOELAP performance testing in May 1995. As of May 

1998, KCP has been exempt from DOELAP accreditation based on the low-level potential for worker 

dose and accreditation by NVLAP.

6.4 DOSE RECONSTRUCTION

Evaluation of KCP worker dose to ensure that the occupational dose for each worker claim is not 

underestimated involves assessment of:
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• Potential unmonitored dose for workers who were not monitored for occupational radiation 

exposure throughout their employment at KCP. 

• Potential missed dose for monitored workers because of missing dosimeter results, actual 

dose that was less than the detection capabilities of the dosimeters, or unrecorded doses in a 

worker’s exposure history.

• Potential adjustments to the recorded dose because of considerations of the dosimetry 

technology, calibration methods, and workplace radiation fields that could have resulted in 

error in the recorded dose.

6.4.1

Based on KCP safety policies and the recorded dose to categories of workers, it appears that 

monitoring occurred for most categories of workers. However, for workers without a recorded dose, it 

is reasonable to assume that any unmonitored dose would be less than the dose received by 

monitored workers. Using the measured doses for these workers is more reasonable than an 

analysis based on the AEC radiation protection guidelines to assign dosimeters to workers who

potentially exceed 10% of the dose limit. Figure 2 shows a statistical analysis of recorded penetrating 

(i.e., photon) doses in a lognormal probability plot for 1950 through 2003. Table 15 summarizes for 

KCP recorded annual penetrating dose the arithmetic average and maximum value for all recorded 

penetrating doses and the lognormal probability statistical parameters for all positive recorded doses 

(i.e., dose > 0). Dose reconstructors should assign the ambient environmental dose to an 

unmonitored worker with minimal potential for radiation exposure from KCP operations, median 

coworker dose to an unmonitored worker with minimal likelihood of actual workplace exposure and 

the 95th percentile coworker dose to workers with a potential for workplace radiation exposure for 

each year of employment without a recorded dose. There should not, typically, be a significant 

neutron exposure of unmonitored workers because sources of neutron radiation were very limited.

Potential Unmonitored Dose

6.4.2

Missed dose occurs when the dose of record is zero because the interpreted dose had a negative 

bias, dosimeter response was less than the minimum detectable level (MDL), or there was no dose of 

record for an assigned badge for a period. There is no evidence of a systematic negative bias, as 

could occur from background control dosimeters in locations of elevated ambient dose (ORAU 

2003b). Missed dose because the response is less than the MDL is typically the most important 

consideration and is important for earlier years when MDLs were higher and dosimeter exchange was 

more frequent. Dose reconstructors should follow NIOSH (2002) guidance to calculate the missed 

photon dose. Methods for estimating the potential missed photon dose include: 

Potential Missed Dose for Monitored Workers

• Method 1: Estimate the missed dose reading from other readings for the same person doing 

similar work for different periods (Watson et al. 1994). 

• Method 2: Assign a maximum missed photon dose based on the MDL and the number of 

routine exchange periods (NIOSH 2002) for the respective dosimetry systems in Table 16. 

Dose reconstructors should use Method 1 if there is sufficient information and only occasional missed 

doses, and Method 2 if significant doses are zero or missing. Use Method 2 for workers who were 

monitored with missing or zero recorded dose or who worked in an area or an occupation where 

positive dose would be expected but was not recorded. 
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Annual Penetrating Dose: KCP

Annual Penetrating Dose (mrem)

Lognormal Probabilities

1 10 100 1000 10000

.0001

.0005

.002

.005

.02

.05

.1

.2

.3

.5

.7

.8

.9

.95

.98

.995

.998

.9995

.9999

Figure 2. Lognormal probability plot of KCP recorded annual penetrating dose. 

6.4.3

KCP worker recorded positive neutron dose was recorded a total of 35 times over a period from 1966 

through 1996, with most instances occurring in the 1980s with good dosimetry capabilities, from a 

total of 14,758 annual dose records. In all cases, except for 2, the recorded annual deep dose is 

equal to the recorded annual neutron implying all recorded deep dose resulted from the neutron dose. 

The two exceptions occurred in 1976 and 1983 with recorded deep and neutron annual doses of 64 

mrem and 26 mrem, and 3 mrem and 1 mrem, respectively. As such, the recommended approach is 

to use a neutron to photon dose ratio of 1:1 as a conservative estimate of neutron dose to the few 

workers with any potential of neutron exposure. Overall, the recorded neutron dose is a relatively 

insignificant component of the worker dose. The recorded neutron dose must be adjusted to include 

conversion to the International Commission on Radiological Protection (ICRP) Publication 60 (1991) 

neutron weighting factor required for input of the dose into the Interactive RadioEpidemiological 

Program (IREP) by using the assumed neutron energy and dose fraction listed in Table 17. Because 

there are no workplace measurements of neutron spectra, this analysis assumed that 100% of the 

workplace neutron radiation is within the range of 0.1 to 2 MeV because this option provides the 

highest organ dose estimate. The approach can be simplified using the following expression:

Neutron Dose Adjustments for Monitored Workers

Neutron dose = adjusted photon dose × upper 95th-percentile neutron/photon dose ratio × ICRP 60 DCF
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 28 of 46

Table 15. Statistical parameters of recorded penetrating annual doses. 

Year

Arithmetic, all recorded dose Lognormal, dose>0

No. of 

workers

Dose, rem No. of 

workers

Dose, rem

Mean Maximum Median 95% GSD

1950 46 7.09E-02 6.47E-01 29 5.01E-02 5.54E-01 4.31E+00

1951 227 2.78E-01 5.32E+00 158 4.97E-02 2.02E+00 9.52E+00

1952 233 1.91E-01 5.90E+00 220 2.45E-02 7.42E-01 7.96E+00

1953 103 1.37E-01 5.32E+00 72 2.13E-02 4.50E-01 6.38E+00

1954 65 1.61E-01 2.99E+00 21 1.19E-01 4.74E+00 9.39E+00

1955 41 9.18E-02 4.18E-01 25 1.10E-01 4.59E-01 2.39E+00

1956 26 2.89E-01 1.75E+00 26 1.35E-01 1.07E+00 3.52E+00

1957 65 5.24E-01 7.89E+00 36 1.77E-01 6.33E+00 8.80E+00

1958 301 3.90E-02 8.43E-01 89 5.44E-02 4.84E-01 3.78E+00

1959 464 8.97E-03 2.30E-01 72 3.57E-02 1.78E-01 2.66E+00

1960 1,043 1.21E-02 1.35E+00 165 4.53E-02 2.31E-01 2.70E+00

1961 948 2.39E-02 5.90E-01 400 3.17E-02 1.64E-01 2.72E+00

1962 700 1.25E-02 1.65E+00 59 4.25E-02 4.53E-01 4.22E+00

1963 597 1.18E-02 1.63E+00 100 3.65E-02 2.02E-01 2.83E+00

1964 530 6.85E-03 6.90E-01 59 3.09E-02 1.97E-01 3.08E+00

1965 436 1.35E-03 8.00E-02 26 1.66E-02 5.59E-02 2.09E+00

1966 415 2.32E-03 2.00E-01 27 2.54E-02 8.95E-02 2.15E+00

1967 370 1.71E-03 2.50E-01 20 2.06E-02 7.14E-02 2.13E+00

1968 469 1.75E-03 3.50E-01 5 1.10E-01 6.84E-01 3.04E+00

1969 577 (a) 0.00E+00 0

1970 580 1.60E-03 3.20E-01 29 1.61E-02 7.47E-02 2.54E+00

1971 575 9.39E-04 2.00E-01 16 2.02E-02 9.09E-02 2.50E+00

1972 195 1.19E-02 4.26E-01 68 2.04E-02 8.61E-02 2.40E+00

1973 199 9.86E-02 1.81E+01 69 1.87E-02 1.09E-01 2.91E+00

1974 169 1.30E-02 8.80E-02 67 2.86E-02 6.89E-02 1.71E+00

1975 150 5.72E-03 8.60E-02 44 1.25E-02 5.96E-02 2.58E+00

1976 126 9.65E-03 2.48E-01 53 1.37E-02 7.34E-02 2.77E+00

1977 123 5.45E-03 3.00E-01 12 2.76E-02 1.72E-01 3.04E+00

1978 152 7.53E-03 1.25E-01 18 4.55E-02 2.25E-01 2.64E+00

1979 162 2.69E-03 8.30E-02 17 2.12E-02 5.64E-02 1.81E+00

1980 185 4.19E-03 1.33E-01 22 2.74E-02 8.49E-02 1.99E+00

1981 210 4.12E-03 1.20E-01 24 2.72E-02 9.27E-02 2.11E+00

1982 209 2.00E-03 5.20E-02 22 1.67E-02 3.72E-02 1.63E+00

1983 226 2.98E-03 3.14E-01 12 2.76E-02 2.06E-01 3.39E+00

1984 216 1.89E-02 3.57E+00 18 2.44E-02 2.80E-01 4.41E+00

1985 201 4.67E-02 8.66E+00 49 1.41E-02 8.34E-02 2.95E+00

1986 194 3.22E-03 5.50E-02 27 2.05E-02 4.71E-02 1.66E+00

1987 196 1.65E-03 4.00E-02 20 1.18E-02 4.82E-02 2.35E+00

1988 188 3.52E-03 1.80E-01 11 4.62E-02 1.81E-01 2.29E+00

1989 233 8.03E-04 1.60E-02 17 1.09E-02 1.37E-02 1.15E+00

1990b 217 1.16E-03 4.10E-02 17 1.37E-02 2.50E-02 1.44E+00

a. Recommend using values for 1968. All 1969 recorded doses = zero.

b. Recommend using values for 1990 for all subsequent years. 

6.4.4

The DOE-reported photon deep dose for KCP workers is likely to provide a reasonable estimate of the 

actual photon dose. The dosimetry technology used is capable of detecting and measuring the 

photon dose based on studies of historical performance (AEC 1955; Unruh et al. 1967). The energy 

of prevalent photons based on the radiation sources is readily measured, so no adjustment of 

recorded dose would be likely based on the response characteristics of the different dosimeters. 

Adjustments to Recorded Deep Dose
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Table 16. Potential missed dose for monitored workers. 

Dosimeter Period Exchange

MDL (mrem)

Maximum missed 

annual dose (rem)

Shallow Deep Shallow Deep

KCP in-house film badge system 1950 - 1954 Weeklya 40b 40b 1.100 1.100

KCP in-house film badge system 1955 Biweekly 40 40 0.960 0.960

KCP in-house film badge system 1956 - 1964 Monthly 40 40 0.480 0.480

KCP in-house film badge system 1965 -1972 Bimonthly 40 40 0.240 0.240

KCP 2-chip TLD 1973 - 1982 Bimonthly 30 30 0.180 0.180

Eberline 3-chip TLDs 1983 - 1990 Quarterly 30 30 0.120 0.120

Landauer 3-chip TLDs 1991 - 2000 Quarterly 30 30 0.120 0.120

Landauer OSL 2001 - 2003 Semiannual 30 30 0.060 0.060

a. Earliest estimated exchange for radiation workers is weekly.

b. Estimated MDL typical of film dosimeter capabilities.

Table 17. Neutron dose fractions and associated ICRP 60 correction factors (ICRP 

1991).

Description

IREP neutron 

energy (MeV)

Default dose 

fraction (%)

ICRP 60 correction 

factor (CF)

KCP workplace exposures 0.1 - 2 MeV 100 1.91

6.4.5

Table 18 summarizes the recommended fractions for KCP worker deep dose according to the energy 

categories required by IREP. For the reported deep dose, 50% is assigned to the respective 

categories of photon energy from 30 to 250 keV, and greater-than-250-keV. For the reported neutron 

dose, 100% is attributed to the 0.1- to 2-MeV category. For the reported shallow dose, the relative 

contribution from beta and photon radiation cannot be distinguished. As such, it is recommended that 

the dose reconstructor calculate the maximum dose to the organ of interest assigning 100% to the 

beta radiation > 15 keV category or assigning 100% to the photon radiation < 30-keV category. This 

provides claimant-favorable analysis of the dose to the organ of interest. 

Radiation Dose Fraction

Table 18. Beta, photon, and neutron radiation energies and percentages for external radiation 

exposures. 

Description

Operations Radiation 

type

Energy 

Begin End selection %

Parameters to estimate dose to whole-body organs 1949 2004 Beta > 15 keV 100a

Photon 30–250 keV 50b

> 250 keV 50c

Neutron 0.1-2 MeV 100d

a. Beta particles (electrons) from DU, sealed sources, and electron accelerators are greater than 15 keV.

b. Workplace photon energies from scattered X-rays and DU are greater than 30 keV. Primary photon energies 

for DU are within the IREP category of 30 to 250 keV.

c. Workplace photon energies cover a broad range. Photons from sealed sources and RGDs are typically of

higher energies but are also heavily shielded. 

d. The neutron energy region of 0.1 – 2 MeV was selected to be claimant-favorable for whole-body dose to 

provide the highest calculated organ dose.

6.5 UNCERTAINTY IN PHOTON AND NEUTRON DOSE

The analysis for this Site Profile evaluated bias and uncertainty associated with the KCP external 

dosimetry systems.
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6.5.1

The uncertainty in the KCP recorded photon deep dose results from various parameters related to

dosimeter response, calibration, and workplace radiation fields can only be estimated. This was 

calculated using guidance in NIOSH (2002) and the National Research Council analysis of dose from 

atmospheric nuclear tests (NRC 1989). Table 19 lists estimates of uncertainty for film dosimeters 

used at KCP from 1950 to 1973.

Deep Photon Dose for Film Dosimeters 

Table 19. Uncertainty for KCP film photon deep dose. 

Recorded 

dose (mrem)

Estimated IREP 

Parameter 2a

Estimated 95% 

uncertainty factor

Upper 95% confidence 

photon deep dose (mrem)

10 2.32 5.21 52

20 1.78 3.11 62

50 1.37 1.85 93

100 1.20 1.43 143

200 1.10 1.22 244

500 1.10 1.20 600

1,000 1.10 1.20 1,200

2,000 1.10 1.20 2,400

3,000 1.10 1.20 3,600

4,000 1.10 1.20 4,800

5,000 1.10 1.20 6,000

a. Recommended input value for IREP Parameter 2 is the highest value that bounds the annual 

dose value (i.e., use 1.36 for doses between 50 and 100 mrem). 

The values in Table 19 were calculated from guidance in NIOSH (2002) as follows:

where

Upper 95% confidence photon dose = E × K(E) (roentgens)

E = dose (roentgens)

σ* = 0.015

D∞ = 2.8

γ = 0.25

and K(E) is limited to a minimum of 1.20 (NRC 1989).

6.5.2

KCP has used TLDs since 1974 to measure the deep photon dose. The uncertainty in the recorded 

deep photon dose TLD results was calculated using guidance in NIOSH (2002) and the National 

Research Council analysis of dose from atmospheric nuclear tests (NRC 1989). Table 20 lists

estimates of uncertainty for the TLDs used at KCP beginning in 1974.

Deep Photon Dose for TLDs 

The respective values in Table 20 were calculated from guidance in NIOSH (2002) as follows:
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 31 of 46

where

Upper 95% Confidence Dose = E × K(E) (millirem)

E = Dose (millirem)

K(E) = 1 + 1.96 * (σ(E)/E)

Lc = 30 mrem

σ* = 10%

Table 20. Uncertainty in TLD deep photon dose. 

Recorded dose 

(mrem)

Estimated IREP 

Parameter 2a Estimated 95% uncertainty factor

Upper 95% confidence 

photon deep dose (mrem)

10 2.00 4.01 40.1

20 1.59 2.51 50.2

50 1.28 1.63 81.5

100 1.17 1.36 136

200 1.12 1.25 250

500 1.10 1.20 600

1,000 1.09 1.20 1200

2,000 1.09 1.20 2400

3,000 1.09 1.20 3600

4,000 1.09 1.20 4800

5,000 1.09 1.20 6000

a. Recommended input value for IREP Parameter 2 is the highest value that bounds the annual dose 

value (i.e., use 1.36 for doses between 50 and 100 mrem). 

6.5.3

The approach in sections 6.5.1 and 6.5.2 based on the recorded dose may not be feasible for many 

workers. Dose reconstructors may incorporate consideration of uncertainty in the dose calculation for 

measured and missed doses as follows:

Alternative Approach

• The technology used to measure worker dose at KCP is similar to the technology used by 

commercial and AEC laboratory facilities. The errors in the penetrating dose are anticipated to 

be approximately ±30% and normally distributed. Dose reconstructors can assume that errors 

are all positive (i.e., use only +30%) and multiply the measured dose by a factor of 1.3 (i.e., 

increase of 30%) to be used for IREP Parameter 1 and to set Parameter 2 to zero.

• For missed dose, a lognormal distribution is assumed. Dose reconstructors should calculate 

the unmonitored dose using Section 6.4.1 and the missed dose using Section 6.4.2 to arrive at 

Parameter 1 input and to set Parameter 2 equal to 1.52.

6.5.4

The primary workplace radiation that contributes to KCP worker exposure is photon radiation. Dose 

reconstructors should apply the uncertainties in Tables 19 and 20 for film and TLD measured deep 

dose, respectively, as needed to the shallow dose component as well.

Shallow Dose
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 32 of 46

6.5.5

Few if any KCP workers have received a significant neutron dose. The recommendation to estimate 

the neutron dose for workers with potential for neutron exposure based on their work activities with 

neutron emitting equipment using a 1:1 ratio between the measured photon deep dose, and 

uncertainty in Tables 19 and 20, is sufficient to address uncertainty in the neutron dose considering 

the low probability of significant neutron dose. 

Neutron Dose Uncertainty

6.6 ORGAN DOSE

After calculating the photon and neutron doses and their associated standard errors for each year, 

dose reconstructors can use the values to calculate organ doses of interest using NIOSH (2002). 

There are many complexities and uncertainties when applying organ DCFs to adjusted doses of 

record. Many of the factors that affect the dose of record have been summarized in tables in this Site 

Profile. ICRU (1988) indicated that film badge dosimeters, while not tissue-equivalent, can be used 

for personnel dosimetry though it is more difficult to ensure that the variation in response with energy 

and angle of incidence is correct for lower energy photon radiation. Given the many uncertainties, 

especially with film badge dosimetry in the 1950s through 1970s, a claimant-favorable approach is 

used to estimate organ dose using an anterior-posterior (AP) exposure geometry. The exposure-toorgan DCFs for an AP orientation listed in Table 21 result in a higher organ dose given the radiation 

effectiveness factors of the intermediate energy photons. As such, these DCFs are used to convert 

Table 21. AP photon exposure-to-organ DCFs.

a

Organ

Exposure-to-organ DCFs

(AP geometry)

< 30 keV 30-250 keV >250 keV

Bladder 0.175 1.244 0.883

Bone (red marrow) 0.025 0.626 0.720

Bone (surface) 0.209 1.229 0.764

Breast (female) 0.561 1.266 0.930

Colon 0.075 1.060 0.844

Esophagus 0.014 0.688 0.745

Eye 0.936 1.236 0.880

Gonads (female-ovaries) 0.047 0.955 0.819

Gonads (male-testes) 0.622 1.434 0.941

Liver 0.106 1.064 0.845

Lung 0.100 0.986 0.842

Remainder organs 0.071 0.879 0.787

Skin 0.504 0.892 0.835

Stomach 0.182 1.251 0.885

Thymus 0.288 1.408 0.892

Thyroid 0.473 1.440 0.972

Uterus 0.061 1.011 0.786

a. Source: NIOSH (2002).

recorded film badge gamma (photon) doses to organ dose. Table 22 lists the deep-dose-to-organ 

DCFs for an AP orientation for neutrons. In the conversion of photon and neutron doses of record to

organ doses, the exposure geometry is an important consideration. The AP exposure geometry is 

used to estimate the organ dose to conduct an initial screening.

Some of the more common exposure geometries encountered in the workplace are defined as 

follows: 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 33 of 46

Table 22. AP neutron deep-dose-equivalent-to-organ DCFs.

a

Organ

Deep-dose-equivalent-to-organ DCFs

(AP geometry)

10-100 keV 0.1-2 MeV 2-20 MeV

Bladder 1.268 0.796 1.105

Bone (red marrow) 0.651 0.361 0.720

Bone (surface) 0.656 0.436 0.675

Breast (female) 1.111 1.145 1.121

Colon 0.947 0.490 0.912

Esophagus 0.775 0.412 0.869

Gonads (female-ovaries) 0.935 0.424 0.903

Gonads (male-testes) 1.466 1.307 1.222

Liver 0.983 0.641 0.990

Lung 0.737 0.557 0.950

Remainder organs 0.819 0.525 0.889

Skin 0.986 0.853 0.918

Stomach 1.221 0.824 1.099

Thyroid 1.066 1.086 1.123

a. Source: NIOSH (2002).

• An AP exposure is typical for an individual who works in a directional radiation field and faces 

the source of the radiation source, such as a nuclear weapon component, while working.

• A rotational exposure is typical of an individual who is constantly turning in a directional 

radiation field while working, such as when conducting inventories in the nuclear weapons 

storage vaults. 

• An isotropic exposure is typical of a worker involved in activities involving a highly 

nondirectional or omnidirectional radiation field. An example of work in an omnidirectional 

radiation field that leads to isotropic irradiation of a worker would be maintenance activities 

where scattered neutrons and photon radiation are incident on the worker from all directions. 

The proposed initial screening option to identify likely noncompensable cases based on claimantfavorable organ dose estimates for long-term workers is to use the organ DCFs for an AP exposure 

geometry as indicated in Tables 6 and 7. Claims that require a more realistic assessment to 

determine compensability should consider the geometries mentioned above. Appendix B of NIOSH 

(2002) provides DCFs to convert KCP worker photon and neutron doses to the primary organ doses 

for many selections of exposure geometry, target organ, and radiation quantity.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 34 of 46

REFERENCES

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Bendix (Bendix Corporation), 1962, Purchase Order ICO-020757 to Union Carbide and Nuclear 

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F. T. Tate, Bendix Corporation, Kansas City, Missouri, May 7. 

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E. G. Applegate (Bendix Aviation Corporation), Honeywell Federal Manufacturing & 

Technologies, LLC, Kansas City, Missouri, September 24. 

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Diagnostic Radiology, Publication 34, Pergamon Press, Oxford, England. 

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England

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for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Age-Dependent Doses to 

Members of the Public from Intake of Radionuclides: Part 4 Inhalation Dose Coefficients,

ICRP Publication 71, Pergamon Press, Oxford, England.

ICRU (International Commission on Radiation Units and Measurements), 1988, Determination of 

Dose Equivalents from External Radiation Sources - Part 2, Report 43, Bethesda, Maryland.

ICRU (International Commission on Radiation Units and Measurements), 1993, Quantities and Units 

in Radiation Protection Dosimetry, Report 51, Bethesda, Maryland.

KCD (Kansas City Division), 1958, Organization, Policies and Procedures Manual, Health and Safety 

Measures, Procedure No. S-306, Kansas City, Missouri. 

KCP (Kansas City Plant), 1962, The Fluorophotometric Determination of Uranium in Urine, Kansas 

City, Missouri.

KCP (Kansas City Plant), 1987, List of Radioactive Sources, Major Use and Major Energies, Kansas 

City, Missouri.

KCP (Kansas City Plant), 1989, Report of Investigation of Pm 147 Contamination February 10, 1989, 

Kansas City, Missouri. 

KCP (Kansas City Plant), 1998, Uranium Dioxide Powder (U), control number 4562260-00, Issue H 

release/change 982076KC, Kansas City, Missouri, October. 

KCP (Kansas City Plant), undated, Uranium in Urine and Lead List, Kansas City, Missouri.

Lund, A., 2004, “Confirmed Intakes,” electronic mail to J. Fix (U.S. Department of Energy, Pacific 

Northwest Laboratory), Honeywell Federal Manufacturing & Technologies, LLC, Kansas City, 

Missouri, May 10. 

McDonald, J. C., P. L. Roberson, J. J. Fix, G. W. R. Endres, R. T. Hadley, L. L. Nichols, K. L. 

Holbrook, R. B. Schwartz, and R. C. Yoder, 1983, Response Characteristics of Selected 

Personnel Neutron Dosimeters, PNL-3982, Pacific Northwest Laboratory, Richland, 

Washington.
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Nasca, B. M., 2004a, “Summary of KCP Response to List of Information Needs, permanent access 

list,” electronic mail to J. Fix (U.S. Department of Energy, Pacific Northwest Laboratory), 

August 5, Honeywell Federal Manufacturing & Technologies, LLC, Kansas City, Missouri.

Nasca, B. M., 2004b, “Summary of KCP Response to List of Information Needs,” attachment to letter 

from A. Lund to J. Fix (U.S. Department of Energy, Pacific Northwest Laboratory), Honeywell 

Federal Manufacturing & Technologies, LLC, Kansas City, Missouri, June 23.

Nasca, B. M., 2004c, “KCP Radiation Exposure History,” electronic mail to J. Fix (U.S. Department of 

Energy, Pacific Northwest Laboratory), Honeywell Federal Manufacturing & Technologies, 

LLC, Kansas City, Missouri, March 31. 

Nasca, B. M., 2004d, “unexplainable dose,” electronic mail to J. Fix (U.S. Department of Energy, 

Pacific Northwest Laboratory), Honeywell Federal Manufacturing & Technologies, LLC, 

Kansas City, Missouri, April 2. 

Nasca, B. M., 2004e, “air sampling data,” electronic mail to J. Fix (U.S. Department of Energy, Pacific 

Northwest Laboratory), Honeywell Federal Manufacturing & Technologies, LLC, Kansas City, 

Missouri, September 8. 

Nasca, B. M., 2005a, “subject: Re; Email dated 2/8/05, Uranium Bioassay,” letter to R. J. Traub, 

February 9; dosimetry records of four individuals were attached to this letter.

Nasca, B. M., 2005b, “Subject: Re: Uranium at KCP,” electronic mail to R. J. Traub, February 24. 

Nasca, B. M., 2005c, “Subject: Re: Uranium Bioassays,” electronic mail to R. J. Traub, February 3.

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Releases of Radionuclides to Atmosphere, Surface Water, and Ground, Report 123 I, 1-316, 

Bethesda, Maryland. 

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Beam and Gamma-Ray Protection for Energies Up to 50 MeV (Equipment Design, 

Performance and Use), Report 102, Bethesda, Maryland.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Revision 1, Office of Compensation Analysis and

Support, Cincinnati, Ohio.

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy Press, Washington, D.C.

ORAU (Oak Ridge Associated Universities), 2003a, Dose Reconstruction from Occupationally 

Related Diagnostic X-ray Procedures, ORAUT-OTIB-0006, Revision 02, Oak Ridge, 

Tennessee.

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of External Radiation, ORAUT-OTIB-0007, Revision 00, Oak Ridge, Tennessee.

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Diffusion Plant - Occupational Internal Dose, ORAUT-TKBS-0019-5, Revision 00, Oak Ridge, 

Tennessee.

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Environmental Internal Doses for Employees Not Exposed to Airborne Radionuclides in the 

Workplace, ORAUT-OTIB-0014, Revision 00, Oak Ridge, Tennessee.

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the Mallinckrodt Chemical Company St. Louis Downtown Site and St. Louis Airport Site, St. 

Louis, Missouri, Period of Operation: 1942-58, ORAUT-TKBS-0005, Revision 01, Oak Ridge, 

Tennessee.

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Department of Energy, before the Subcommittee on Military Procurement Committee on 

National Security, US House of Representatives, 1998, U.S. Department of Energy, 

Washington, D.C., March 19. 

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Corporation, Kansas City, Missouri.

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Bendix Corporation, Kansas City, Missouri, May 19.

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E. L. Nelson, and D. C. Whitman, Bendix Corporation, Kansas City, Missouri, March 30. 

Schiltz, J. O., 1966, “Memorandum to Emergency Radiation Monitoring Team File,” Bendix 

Corporation, Kansas City, Missouri, May 19, 1964, with updates in August 1965 and April 

1966. 

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Environmental Monitoring, TMA Eberline, Albuquerque, New Mexico.

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xrays,” electronic mail to J. Fix (U.S. Department of Energy, Pacific Northwest Laboratory), 

Honeywell Federal Manufacturing & Technologies, LLC, Kansas City, Missouri, April 8. 

Todd, M., 2004b, “Response to historical X-ray practice questions, examination frequency,” electronic 

mail to J. Fix (U.S. Department of Energy, Pacific Northwest Laboratory), Honeywell Federal 

Manufacturing & Technologies, LLC, Kansas City, Missouri, June 2. 

Todd, M., 2004c, “Response to historical X-ray practice questions, routine or special physical exams,” 

electronic mail to J. Fix (U.S. Department of Energy, Pacific Northwest Laboratory), Honeywell 

Federal Manufacturing & Technologies, LLC, Kansas City, Missouri, August 19. 

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Film Dosimeter Performance Criteria, BNWL-542, Battelle Northwest Laboratory, Richland, 

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Watson, J. E., Jr., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, “Estimation of Radiation 

Dose for Workers Without Monitoring Data for Retrospective Epidemiologic Studies,” Health 

Physics, volume 67, number 4, pp. 402-405. 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 39 of 46

GLOSSARY

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy.

beta dose

A designation (i.e., beta) on some external dose records referring to the dose from lessenergetic beta, X-ray, or gamma radiation.

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Physically, the beta particle is 

identical to an electron moving at high velocity.

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 centimeter in a material of specified geometry and 

composition.

deep dose equivalent (Hd) 

The dose equivalent at the depth of 1.0 centimeter in tissue.

depleted uranium (DU)

Uranium having less than the natural mass of 235U; used as components in nuclear weapons 

or as a surrogate for enriched uranium or plutonium in testing. 

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 grays, H is in sieverts

(1 sievert = 100 rem).

dose of record

The dose files provided by DOE to NIOSH as part of the individual worker files. 

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 film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter. 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, and 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 40 of 46

DXT Units

DXT stands for density times thickness. When either the density or thickness is known and 

can be assumed constant, DXT units are used to measure the alternate parameter to a 

precision of about 0.5%. DXT units use either X-ray generators or 137Cs radioactive sources.

exchange period (frequency)

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 photons (i.e., gamma and 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.

field calibration

Dosimeter calibration based on radiation types, intensities, 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. See nuclear emulsion. 

film density

See optical density. 

film dosimeter

A small packet of film in a holder that attaches to a wearer.

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 but with higher energy; the only essential difference is that X-rays 

do not originate in the nucleus. 

gray

International System unit of absorbed dose (1 gray = 100 rad). 

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

minimum detectable level (MDL)

A term used in this and other NIOSH documents to refer to a statistically determined minimum 

detection level, Lower Limit of Detectability (LD), and related quantities.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 41 of 46

neutron film dosimeter

A film dosimeter that contains an NTA film packet.

nuclear emulsion

Often referred to as NTA film and used to measure personnel dose from neutron radiation.

nuclear track emulsion, type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons

that can be seen by using an appropriate imaging capability such as oil immersion and a 

1,000-power microscope or a projection capability.

open window

Designation on film dosimeter reports that implies the use of little shielding. It commonly is 

used to label the film response corresponding to the open window area. 

optical density

The quantitative measurement of photographic blackening with the density defined as 

D = Log10 (Io/I).

personal dose equivalent, Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 millimeter and 

10 millimeter for the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), 

respectively. 

photon

A unit or particle of electromagnetic radiation consisting of X- or gamma rays. 

photon – X-ray

Electromagnetic radiation of energies between 10 and 100 kilovolts-electron whose source 

can be an X-ray machine or radioisotope.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Alpha, beta, neutron, and photon radiation with sufficient energy to ionize atoms. See ionizing 

radiation. 

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

A special unit of dose equivalent, which is equal to the product of the number of rad absorbed 

and the quality factor.

roentgen (R)

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 kilogram of dry 
Document No. ORAUT-TKBS-0031 Revision No. 00 Effective Date: 05/31/2005 Page 42 of 46

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue for higher (above 100 kilovolts-electron) energy photons.

shallow absorbed dose (Ds) 

The absorbed dose at a depth of 0.007 centimeter in a material of specified geometry and 

composition.

shallow dose equivalent (Hs) 

Dose equivalent at a depth of 0.007 centimeter in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation. 

skin dose

Absorbed dose at a tissue depth of 7 milligrams per square centimeter. 

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. 

Whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams 

per square centimeter); however, this term is also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation that originates external to the nucleus of an atom. 
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 43 of 46

ATTACHMENT A

EXTERNAL AND INTERNAL AMBIENT ENVIRONMENTAL DOSE ASSIGNMENT GUIDELINE

Potential external and internal radiation ambient environmental exposures for unmonitored workers 

were based on KCP measured workplace data. 

A1.0

No measurements of environmental external dose from KCP operations have been located. The 

expectation from site documents is little if any radiological ambient environmental impact. However, 

in spite of no evidence of an impact, it is recommended as a claimant-favorable approach that dose 

reconstructors assign an external dose of 25 mrem per year to an unmonitored, nonradiological 

worker for each year of employment at KCP. The IREP input to calculate the organ dose is shown in 

Table A-1. This potential external radiation dose was determined based on the 95% variability 

calculated from background control personnel dosimeters (i.e., 2-sigma = 25 mrem), as shown in 

Table A-2. 

EXTERNAL RADIATION

Table A-1. IREP input for ambient external dose calculation. 

Exposure 

rate Radiation type

Dose 

distribution type Parameter 1 Parameter 2 Parameter 3

Chronic Photons E=30-250keV Constant 0.025 0.000 0.000

A2.0

The primary and apparently only radionuclide handled in large quantities with a potential for significant 

environmental release was depleted uranium (DU) from 1959 through 1971. The recommended 

approach to assign an ambient environmental internal dose to unmonitored nonradiological workers is 

based on analysis of dispersion to the outside of measured laboratory uranium concentrations. There 

are no windows and all potential releases were from stacks. The status of filtration for these stacks is 

not known. KCP representatives identified two primary laboratories used for handling depleted 

uranium as follows:

INTERNAL RADIATION

Department

Stack velocity 

(ft3/min)

Work room 

area (ft2)

Machining area (20D) 2,400 8,800

Mixing/curing (27C) 1,987 7,000

Based on this information, a “claimant favorable” estimate of dispersion between outside and inside 

air concentrations of 0.01 is estimated as described in Section A2.1 based on National Council on 

Radiation Protection and Measurements Report 123, Screening Models for Releases of 

Radionuclides to Atmosphere, Surface Water, and Ground” (NCRP 1996). An annual intake fraction 

of 0.01 times the median annual measured laboratory air concentrations in Table 11 (i.e., multiply 

annual concentrations in Table 11 by 6.58 × 1010 to obtain pCi/day intake) is used to estimate a 

potential daily intake. For the period following 1970, there are no workplace air sampling data 

available for analysis. Since there is the potential for unknown small releases or residual 

contamination, such as the incident with 147Pm in 1989, it is recommended that the average median 

concentration for all years from Table 11 be used with the factor of 0.01 for each year of employment 

following 1970 until 1990 when there is assurance that no significant environmental releases occurred 

based on the multi-agency response findings to the 147Pm incident. The resulting input for IMBA is 

shown in Table A-3.
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 44 of 46

Table A-2. KCP personnel dosimeter control dosimeter data. 

Account 

series

Photon deep 

dose DDE (mrem)

Photon shallow 

dose SDE (mrem) Days mrem/day mrem/hr

DUO 37 37 206 0.179612 0.007

DUO 34 34 210 0.163726 0.007

DUO 41 42 216 0.188441 0.008

DUO 40 39 223 0.179878 0.007

ENG 35 34 206 0.169734 0.007

ENG 35 37 210 0.16878 0.007

ENG 36 38 216 0.167176 0.007

ENG 39 38 223 0.174036 0.007

FMT 30 29 206 0.145881 0.006

FMT 29 28 210 0.135928 0.006

FMT 33 33 216 0.152259 0.006

FMT 33 32 223 0.148526 0.006

GEN 30 30 206 0.146249 0.006

GEN 32 33 210 0.151753 0.006

GEN 35 36 216 0.161556 0.007

GEN 34 34 223 0.152212 0.006

MET 53 53 206 0.259466 0.011

MET 52 53 210 0.247593 0.010

MET 59 62 216 0.271173 0.011

MET 58 57 223 0.260149 0.011

NDE 45 45 206 0.220097 0.009

NDE 48 50 210 0.226797 0.009

NDE 49 52 216 0.228593 0.010

NDE 52 52 223 0.235022 0.010

NGA 41 41 206 0.200631 0.008

NGA 37 36 210 0.174383 0.007

NGA 42 44 216 0.196286 0.008

NGA 42 41 223 0.189626 0.008

RAD 29 29 206 0.14282 0.006

RAD 32 32 210 0.153984 0.006

RAD 34 34 216 0.155472 0.006

RAD 35 35 223 0.156726 0.007

ULT 41 40 206 0.200863 0.008

ULT 41 40 210 0.194969 0.008

ULT 45 46 216 0.208939 0.009

ULT 47 46 223 0.209522 0.009

WMT 39 39 206 0.188718 0.008

WMT 35 35 210 0.168635 0.007

WMT 44 47 216 0.201806 0.008

WMT 44 44 223 0.197593 0.008

XRY 40 39 206 0.194161 0.008

XRY 43 43 210 0.203215 0.008

XRY 49 51 216 0.224537 0.009

XRY 48 47 223 0.213842 0.009

Average = 0.1888947

Stdev = 0.0345245

year, 1-sigma 12.60143 mrem

2-sigma 25.20286 mrem
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 45 of 46

Table A-3. IMBA input for ambient internal 

dose calculation. 

Year

Median KCP measured 

air concentration

(μCi/cm GSD 3) pCi/day

1958 4.01E-13 2.64E-02 3.0

1959 2.89E-13 1.90E-02 3.0

1960 3.41E-13 2.24E-02 3.0

1961 1.97E-13 1.30E-02 3.0

1962 2.50E-13 1.65E-02 3.0

1963 2.47E-13 1.63E-02 3.0

1964 3.91E-13 2.57E-02 3.0

1965 1.05E-13 6.91E-03 3.0

1966 2.00E-13 1.32E-02 3.0

1967 5.70E-13 3.75E-02 3.0

1968 2.31E-13 1.52E-02 3.0

1969 3.88E-12 2.55E-01 3.0

1970 4.02E-14 2.65E-03 3.0

1971-89 5.49E-13 3.62E-02 3.0

A2.1 KCP DU AIR CONCENTRATIONS OUTSIDE THE PLANT FROM ACTIVITIES INSIDE 

THE PLANT

Calculations of the possible air concentration of DU outside the KCP facility due to activities inside the 

building were done based on the measured inside air concentration and engineering details of the 

primary laboratories. KCP work with DU was performed in two laboratories described as follows: 

Department

Stack velocity 

(ft3/min)

Work room 

area (ft2)

Machining area (20D) 2,400 8,800

Mixing/curing (27C) 1,987 7,000

Ref: Email Nasca to Fix, 21-Oct-2005. Subject Re: Uranium 

handling at KCP. 

For purposes of calculating dilution factors, the measured room air concentration was assumed to be 

1 Bq/m3 (a unit air concentration). The results of the calculated outside air concentration was then 

used to calculate the potential occupational dose for selected receptor locations. 

A2.1.1

Using the Machining Area (20D) as an example, the quantity of DU exhausted from the room is: 

1.133 Bq/s1

. 

Receptor on the Roof

Equation 1 (Equation 2.8 from NCRP 1996) was used to calculate the air concentration downwind 

from the release point but on the same surface as the release point (i.e. both the source and receptor 

are on the roof).

2 μ 0 x

Q

h

= BC (Eq. 1)
Document No. ORAUT-TKBS-0031 Revision No. 00 PC-1 Effective Date: 01/06/2006 Page 46 of 46

Where B0 = 30, Q = the release rate (1.133 Bq/s), uh = mean wind speed (10.6 mph = 4.74 m/s), and 

x is the distance downwind (m). To obtain the distance where C = 0.01, Equation 1 is rewritten as 

Equation 2 and solved for x. 

Cuh

QB

x 0 = (Eq. 2)

Substitution of the numerical values given above, C = 0.01 when x exceeds 26.7 m (87.9 ft) and the 

wind always blows towards the recipient.

A2.1.2

For this case the calculations use Equation 3 (Equation 2.6 from NCRP 1996) 

Receptor on the Ground Downwind From the Source on the Roof

u

fQP C = (Eq. 3)

Where f is the fraction of the time that the wind blows from the source to the receptor and P is the 

dilution factor from figure 2.2 of the NCRP report. The greatest air concentrations result when the 

height of the building is assumed to be zero, in which case the value of P is 3.6 × 10-2. Substituting 

the numeric values given above, the value of C is 8.6 × 10-3 Bq/m3

. The dilution factor of 0.0086 is for 

a distance of 100 m (the closest distance plotted in the reference figure) from the source point. This 

analysis is based on the claimant-favorable assumption that the wind is always blowing toward the 

receptor.