ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for Atomic Energy Operations at the 

Iowa Army Ammunition Plant (IAAP)

Document Number:

ORAUT-TKBS-0018

Effective Date: 03/14/2005

Revision No.: 01

Controlled Copy No.:

Page 1 of 79

Subject Experts: John A. Leonowich/Don Bihl/Jack Fix/Dillard 

Shipler/Bruce Napier/Mark Rolfes/Tim Taulbee

Document Owner

Approval: Signature on File Date: 03/14/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/14/2005

Richard E. Toohey, Project Director

Approval: Signature on File _ Date: 03/14/2005

James W. Neton, Associate Director for Science

Supersedes:

Revision No.: 00

TABLE OF CONTENTS

Section Page

LIST OF TABLES............................................................................................................................ 3

LIST OF FIGURES ......................................................................................................................... 4

ACRONYMS AND ABBREVIATIONS ............................................................................................ 6

1.0 INTRODUCTION................................................................................................................. 8

2.0 SITE DESCRIPTION, OPERATIONAL HISTORY, AND PROCESS................................ 8

2.1 OPERATIONAL HISTORY.............................................................................................. 9

2.2 FISSILE MATERIALS ONSITE ..................................................................................... 10

2.2.1 Early Mission and Responsibilities ....................................................................... 10

2.2.2 Early Weapons Design......................................................................................... 10

2.2.3 Radiation Safety Program..................................................................................... 11

2.2.4 Fissile Material Assumption.................................................................................. 11

3.0 OCCUPATIONALLY RELATED MEDICAL X-RAYS ....................................................... 13

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE.................................................................. 15

4.1 INTRODUCTION............................................................................................................ 15

4.2 INTAKES FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS . 15

4.2.1 Intakes from Tritium Releases.............................................................................. 16

4.2.2 Intakes from Release of DU from Burning Sites .................................................. 17

4.2.3 DU Intakes from Potentially Contaminated Drinking Water................................. 18

4.3 EXTERNAL DOSE......................................................................................................... 20

4.4 UNCERTAINTY.............................................................................................................. 22

5.0 ESTIMATION OF INTERNAL EXPOSURE...................................................................... 22

5.1 TRITIUM EXPOSURE ................................................................................................... 23

5.2 DEPLETED URANIUM.................................................................................................. 25

5.2.1 DU Intakes from Hydroshots................................................................................. 26
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5.2.2 DU Intakes from Machining Baratols.................................................................... 29

5.2.3 DU Intakes by Operators at Burning Yard............................................................ 30

5.2.4 DU Intakes from Disassembly of Weapons.......................................................... 31

5.3 RADON.......................................................................................................................... 32

5.4 OTHER SOURCES........................................................................................................ 33

6.0 ESTIMATION OF EXTERNAL RADIATION EXPOSURE ............................................... 34

6.1 INTRODUCTION............................................................................................................ 34

6.2 WORKPLACE EXTERNAL RADIATION FIELDS......................................................... 34

6.2.1 Photon Radiation .................................................................................................. 34

6.2.2 Neutron Radiation ................................................................................................. 37

6.2.3 Depleted Uranium ................................................................................................. 38

6.3 EXTERNAL RADIATION DOSE RECORDS................................................................ 39

6.3.1 IAAP Annual Dose Distributions ........................................................................... 41

6.4 DOSIMETER TECHNOLOGY AND PERFORMANCE STUDIES................................ 43

6.4.1 Photon Radiation .................................................................................................. 43

6.4.2 Neutron Radiation ................................................................................................. 45

6.5 IAAP WORKER EXTERNAL DOSE RECONSTRUCTION.......................................... 46

6.5.1 Unmonitored or Intermittently Monitored Workers............................................... 46

6.5.2 Monitored Workers................................................................................................ 51

6.5.3 Adjustments to Photon Dose................................................................................ 51

6.5.4 Estimation of Neutron Dose.................................................................................. 52

6.5.5 Skin and Extremity Dose (Reserved) ................................................................... 53

6.6 UNCERTAINTY IN PHOTON AND NEUTRON DOSE................................................. 53

6.7 ORGAN DOSE............................................................................................................... 54

REFERENCES.............................................................................................................................. 55

GLOSSARY .................................................................................................................................. 60

Appendix A – Facility Information................................................................................................. 65

Appendix B – List of Weapons Programs at IAAP ....................................................................... 67

Appendix C – Summary of Intakes and Details of Ingestion Calculations................................... 68

Appendix D – Generic Pit Specifications...................................................................................... 70

Appendix E – Special Nuclear Material (SNM) Encapsulation and Pit Cladding ........................ 73

Appendix F – IAAP Annual Dosimeter Badge Distributions......................................................... 75

Appendix G –Bladder Cancer Example Calculation for an Unmonitored Worker ....................... 79
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LIST OF TABLES

Table Page

Table 2.1 Significant Radiation Safety Program Events .......................................................... 12

Table 3.1 Results of X-ray study .............................................................................................. 14

Table 3.2 Organ doses from lumbar spine X-rays at IAAP...................................................... 15

Table 4.1 Total uranium results from water samples at IAAPa................................................ 18

Table 4.2 Upper bound equivalent doses (mrem) from ingestion of DU contaminated drinking 

water.......................................................................................................................... 20

Table 5.1 Mass and radiological characteristics of depleted uranium..................................... 25

Table 5.2 Measured DU air concentrations from hydroshots.................................................. 27

Table 5.3 Chronic DU intakes from hydroshots for test fire operators and drivers................. 27

Table 5.4 Chronic DU intakes from hydroshots for other AEC personnel............................... 28

Table 5.5 Chronic DU ingestion intakes from hydroshots for cleanup crew............................ 28

Table 5.6 Summary statistics of 1990 radon measurements at Pantex.................................. 32

Table 6.1 Energy Distribution from Generic Pit........................................................................ 37

Table 6.2 Summary of Dosimetry Data Available .................................................................... 40

Table 6.3 Personnel Dosimeter Monitoring Eras ..................................................................... 40

Table 6.4 Summary of Annual Dose Distributions ................................................................... 42

Table 6.5 Work factor calculation using the two era dose rates.............................................. 48

Table 6.6 Modeled Annual Dose Using Work Factor Methodology......................................... 49

Table 6.7 Am-241 Build-up from decay of Pu-241................................................................... 50

Table 6.8 IAAP Photon Adjustment Factors............................................................................. 51

Table 6.9 Statistical Parameters for IAAP neutron-to-photon dose ratios............................... 53

Table 6.10 IAAP neutron dose fractions and associated ICRP 60 correction factors............. 53
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LIST OF FIGURES

Figure Page

Figure 2.1 Layout of the IAAP site with AEC facilities marked. ................................................. 9

Figure 4.1 Probability plot of bi-weekly non-storage area dosimeters from 1969 - 1974........ 21

Figure 4.2 Plot of Area Monitoring data from 1962 – 1974...................................................... 22

Figure 6.1 MCNP calculated spectra from generic pit ............................................................. 35

Figure 6.2 Measured neutron to photon ratio in IAAP area badges and a function of gamma 

exposure and the resulting fitted lognormal distribution......................................... 37

Figure 6.3 MCNP calculated photon spectra emitted from 234mPa beta in 238U spheres......... 39

Figure 6.4 Comparison of analysis using all dosimeter data versus including only non-zero 

data. ........................................................................................................................ 41

Figure 6.5 Annual dose distributions with the associated uncertainty..................................... 43

Figure 6.6 Comparison of film dosimeter badges exposed to a known quantity of radiation and 

the reported readings by the commercial vendor................................................... 45

Figure 6.7 Work Factor Probability Distribution........................................................................ 48

Figure 6.8 Summary of annual dose distributions using the model between 1949 -1962 and 

actual dosimetry data from 1963 to 1974............................................................... 50

Figure 6.9 Probability plot of neutron to photon ratios calculated from Pantex 809/812 neutron 

and photon dosimeter data..................................................................................... 52
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ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV.

NO. DESCRIPTION

Draft 11/04/2003 00-A

New document to establish the technical basis 

for the development of a radiation exposure 

matrix for IAAP. Completed by Don Bihl, filling 

in for John A Leonowich while he was on 

travel.

Draft 11/12/2003 00-B

Resolved security issues and revised section 

on tritium and added an additional table to the 

external dosimetry section.

Draft 01/30/2004 00-C

Incorporated changes in response to OCAS 

review and comments on 00-B. Changes 

affected all sections.

Draft 03/03/2004 00-D

Incorporated changes in response to internal 

reviews. Changes affected all sections. 

Added Attachments A, B, and C.

Draft 03/19/2004 00-E

Incorporated changes in response to OCAS 

comments. New intake scenario in section 

2.6. Numerous changes in section 3 and 

Attachment B incorporating statistical 

analyses of external doses.

Draft 04/14/2004 00-F

Improvements to section 3 incorporating 

statistical analyses of all the years measured 

doses and minor change to section 6 to make 

radon instructions more explicit. External skin 

doses prior to 1958 reserved.

Draft 04/16/2004 00-G

Adjustments made to DU ingestion intakes to 

be consistent with OTIB-009 per OCAS 

comments.

04/16/2004 04/16/2004 00 First approved issue. Initiated by Donald Bihl.

Draft 01/20/2005 01-A

Revision to incorporate public comments, and 

to expand time period for dose reconstructions 

to include early unmonitored workers. 

Changed document section order and 

formatting for conformance with typical site 

profiles.

This was a major revision with significant 

changes throughout the document in 

methodology, assumptions based on new data 

and other information. 

Draft 03/14/2005 01-B

Incorporated comments from OCAS Associate 

Director for Science.

03/14/2005 03/14/2005 01 Approved issue of Revision 01.
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ACRONYMS AND ABBREVIATIONS

AEC US Atomic Energy Commission

AMAD Activity Median Aerodynamic Diameter

AP Anterior-posterior (or front-to-back) irradiation of the body

BAECP Burlington Atomic Energy Commission Plant

CF conversion factor

CV coefficient of variation

DCF dose conversion factor

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

EDA Explosives Disposal Area

EU enriched uranium

FS Firing Site

GSD geometric standard deviation

HVL half value layer

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

IAAP Iowa Army Ammunition Plant (also sometimes IAAAP)

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

ISO International Standards Organization

ISO isotropic

keV kilo (thousand) electron volts, a unit of energy

LAT lateral view X-ray

MED Manhattan Engineer District

MeV million electron volts, a unit of energy

MPC maximum permissible concentration

MDL minimum detectable level

NCRP National Commission on Radiological Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NTA Eastman Kodak Nuclear Track Emulsion type A
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OCAS Office of Compensation Analysis and Support

ORAU Oak Ridge Associated Universities

PBX plastic-bonded explosive

PNL Pacific Northwest Laboratory [also PNNL – Pacific Northwest National 

Laboratory]

PA posterior-anterior

PAEC potential alpha energy concentration

R roentgen, unit of exposure to ionizing photons in air

REF radiation effectiveness factor

RGD radiation generating device

ROT rotational

SRS Savannah River Site

TBD technical basis document

TLD thermoluminescent dosimeter

WLM working level month
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1.0 INTRODUCTION

Technical basis documents are general working documents that provide guidance

concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute of

Occupational Safety and Health (NIOSH) in the completion of the individual work required 

for each dose reconstruction.

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

of buildings that served a specific purpose at IAAP. It does not necessarily connote an 

“atomic weapons employer facility” or a “Department of Energy facility” as defined in the 

Energy Employee Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 

7384I (5) and (12)).

This document provides information on interpretation of dosimetry records and exposure 

matrices to estimate internal and external doses for workers at the Iowa Army Ammunition 

Plant in Burlington, Iowa. The Iowa Army Ammunition Plant was responsible for high 

explosive (HE) fabrication, assembly of non-nuclear and nuclear components, retrofits,

modifications, surveillance, and disassembly of nuclear weapons.

2.0 SITE DESCRIPTION, OPERATIONAL HISTORY, AND PROCESS

The Iowa Army Ammunition Plant (IAAP) is a load, assemble, and pack munitions facility 

that began production in 1941 and continues to operate as a Government-owned, 

contractor-operated installation. IAAP is in the southeastern part of Iowa, near the town of 

Middletown in Des Moines County. It is about 10 miles west of the Mississippi River and the 

town of Burlington). Less than a third of the IAAP’s 19,015-acre (30-square-mile) property is 

occupied by active or formerly active production or storage facilities. The remaining land is 

evenly divided between leased agricultural acreage and woodlands.

Since operations begin in 1941, IAAP has used explosives and lead-based initiating

compounds to produce a wide variety of ordnance items. The Line 1 area, portions of the 

Firing Site (FS) area, the Explosive Disposal Area (EDA) sites, and Yards C, G, and L came 

under the jurisdiction of the Atomic Energy Commission [AEC; now the U. S. Department of 

Energy (DOE)]. In addition, the Security Command Center, the Emergency Response

Command Post, the Deactivation furnace, Line 3 Warehouse 301, and the North Burn Pads 

Landfill might have been utilized. This area, totaling around 1,630 acres, became known as 

the Burlington Atomic Energy Commission Plant (BAECP). The site was officially renamed 

the Iowa Army Ammunition Plant in 1965. The site has also been referred to as the Iowa 

Ordnance Plant. Specific buildings and the functions of various radiological buildings can be 

found in Appendix A of this report.
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Figure 2.1. Layout of the IAAP site with AEC facilities marked.

2.1 OPERATIONAL HISTORY

During the summer of 1947, Silas Mason Company entered into a contract with the

Ordnance Department to assist in the design and engineering to perform the construction 

and to operate a facility for the purpose of supplying the Atomic Energy Commission (AEC) 

with explosive components nuclear weapons. In May 1948, Silas and Mason Company 

supervisor personnel entered a training program at the Naval Ordnance Test Station (China 

Lake, California). Upon returning from training, these supervisors trained other IAAP

personnel. By the spring of 1949, IAAP was at full production for HE fabrication (Mitchell 

2003). In March 1949, it was decided that certain weapon assembly operations (nonnuclear components) would also be conducted at IAAP (Poole and Harrison, 1954, Mitchell, 

2003). Until March 1949, all of the initial work at IAAP focused solely on HE explosive 

manufacturing.

The initial Iowa Army Ammunition Plant (IAAP) site profile reserved the time period from 

1947-1958 due to a lack of information about the early time period operations. Based on a 

review of IAAP Project History reports (Poole and Harrison, 1954), NIOSH has concluded 

that the first nuclear weapon assembly operations began in 1949 with the Mark IV.

Throughout the remaining years until plant closure, IAAP workers tested, assembled,

conducted surveillance and disassembled a wide variety of nuclear weapons. Through a 

review of the IAAP project history reports, records sent to Pantex from the Burlington Plant, 

and other records stored at the Pantex plant, NIOSH has compiled a list of the weapons 

programs worked on at IAAP (Appendix B). With the assistance of the Department of 

Energy (DOE), the first assembly and disassembly years are also provided.
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2.2 FISSILE MATERIALS ONSITE

The early time period in the initial site profile (ORAU, 2004a) was reserved due to 

uncertainty as to when fissile materials (plutonium or uranium pits) were onsite. Based on 

the review conducted during the development of the initial site profile, NIOSH felt certain 

that fissile materials were onsite at least in 1958 forward. The concern for these materials is 

that they are generally the most radioactive component of a nuclear weapon and result in 

the largest external dose and if unsealed represent the greatest potential for internal dose. 

NIOSH has uncovered considerable evidence that indicated fissile materials were not onsite 

at IAAP until 1955. This evidence considers IAAP’s mission, early weapon design and the 

development of a radiation safety program.

2.2.1 Early Mission and Responsibilities

According to documentation reviewed, the Burlington Plant was responsible for High

Explosive (HE) fabrication, assembly of non-nuclear components, retrofits, modifications, 

and disassembly for weapon retirements (Mitchell, 2003). 

Further documentation indicates that 1956 was the first year that Mason & Hanger company 

handled fissionable material (Lemert, 1979).

Pantex opened in 1953 with the Procter & Gamble Company as operating Contractor. 

Three years later Mason & Hanger – first to manufacture the explosive components –

recorded two more firsts at the Burlington AEC Plant. The company began to 

assemble nuclear weapons for guided missiles, and in doing so it handled the

fissionable material for the first time. Beginning with the introduction of plutonium 

and Uranium 235 into the configuration of the Genie air-to-air missile, the list of 

nuclear weapons assembled at Burlington lengthened to include artillery fired

projectiles of various sizes and warheads for other air-to-air, air-to-ground, and 

ground-to-ground missiles (Lemert, 1979). 

2.2.2 Early Weapons Design

Early weapons (Mark 4, 5, 6, 7, and 12) assembled at IAAP were called In-Flight-Insertable 

(IFI) weapons. These weapons were of ball (tamper) and capsule (pit or special nuclear 

material (SNM)) design (Mitchell 2003, Loeber 2002). The main purpose of this design was 

to prevent an accidental nuclear detonation during transport and delivery. In the IFI weapon 

design, the nuclear capsules (pits) were kept physically separated (outside) the tamper and 

the surrounding high explosives. Prior to use of the weapon, the nuclear capsule (pit) would 

be inserted into the ball (tamper). From a review of the IAAP History of Line 1 Operations 

(Ahlstrand, 1955), non-radioactive “mock-up” pits were used during assembly. This was 

likely done to ensure an exact fit (i.e., tolerances were met) with the tamper once a real 

nuclear capsule (pit) was combined with the completed (assembled) explosives packages of 

the weapons. 

In the mid 1950s, a new “sealed pit” was designed, tested and produced. The following 

excerpt was found in DOE documentation (Mitchell 2003):

In this design, the non-nuclear mechanical assembly was built around the nuclear 

capsule, or pit; and the weapon, in the case of a bomb, was ready for use when it left 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 11 of 79

the assembly plant. In the case of a warhead, it was the military’s responsibility to 

mate the AEC-delivered warhead to the military’s delivery system, usually a missile. 

Working two 12-hour shifts, (seven days a week) at the Burlington Plant, this new 

design entered the stockpile in December 1956, in an “Emergency Capabilities” 

status as the W-25 warhead.

This encapsulated pit encompassed a “fail-safe” design such that accidental detonation was 

highly improbable. With this new design, nuclear components (fissile pits) were assembled 

with the explosive package such that a complete weapon was the final product from IAAP. 

As a result, the first IAAP assembly workers would have been exposed to fissile materials in 

late 1955 or early 1956 with the receipt of the first sealed pits. This is also evident in other 

information obtained from Mitchell (2003).

The sealed pit design weapons precipitated several fundamental changes in the 

nuclear weapons complex. New facility designs were required and constructed at 

Burlington and Pantex to accommodate production work involving encapsulated SNM 

for the first time at either site.

According to the History of Line 1 Operations (Ahlstrand, 1957b), the first “Gravel Gertie” 

assembly buildings at IAAP were placed in operation in October 1957.

2.2.3 Radiation Safety Program

According to the IAAP History of Line 1 Operations (Ahlstrand, 1956a, 1956b, 1957b), with

the planned assembly of these new weapons designs, IAAP instituted a radiological

monitoring program. This program primarily consisted of routine external dose monitoring 

for some workers using film badges, but also included a radiological survey program,

continuous air monitoring, and training of the workforce during safety meetings. Some of 

the training included specialized training for urinalysis. A chronology of significant events 

can be found in Table 2.1.

2.2.4 Fissile Material Assumption

Although the records indicate that fissile materials were not routinely handled at IAAP prior 

to the W-25 assembly in 1956, the possibility that nuclear weapons taken out of the 

stockpile for retrofit or disassembly were returned to IAAP with the fissile material capsules 

cannot be ruled out.

The first assembly operations began with the Mark IV in March 1949. Following a two year 

production schedule, the plant was shutdown in October 1951 for construction to enlarge the 

assembly operation for the next generation of weapon assembly (Poole and Harrision 1954, 

Lemert 1979). Operations were restarted approximately 8 months later in May 1952. As a 

result the first logical point when retrofits could have been conducted on the Mark IV would 

likely have been after May 1952. 

Since NIOSH cannot definitively state when fissile materials first appeared onsite, to error in 

favor of the claimant, NIOSH assumes that fissile materials were onsite since the beginning 

of assembly operations in 1949. Since workers were not routinely monitored onsite prior to 

1962, NIOSH has scaled the external dose distributions based on differences in dose rate 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 12 of 79

from later pits designs to a “generic” pit. This methodology is discussed in section 6.5 of this 

site profile.

Table 2.1 Significant Radiation Safety Program Events

Time Event Reference

November 1955 Selected workers issued film dosimetry on an intermittent 

basis. Badge exchange frequency was weekly.

TracerLab Reports 

(1955)

March 1956 Safety Manager attends Rad-Safe training at LASL. Ahlstrand (1956a)

July 1956

Formal establishment of a Rad-Safe program

Training of all Division B Personnel

Acquirement of Rad-Safe Equipment

Established Continuous Air Monitoring

Ahlstrand (1956b)

July 1957

The following excerpt is from the IAAP Project History Report 

in 1957.

Continued emphasis on Rad-Safe Training for the safe handling of 

radioactive materials in order to prepare all workers for the new 

phases of work in division “B”. Even though this material is received 

at this installation in sealed containers and is low energy radiation, 

employees are trained in all phases of safe handling of these 

materials.

Ahlstrand (1957b)

September 1957 Establishment of a team for Off-site Emergency Radiological 

Monitoring Ahlstrand (1957b)

January – June 

1958

Extensive Rad-Safe training conducted including training on 

urinalysis.

72 Employees attended Primer Course 2T

7 employees spent a week in April at the Nevada Test Site for 

special Rad-safe training

The Plant’s Registered nurses and certain members of the 

Safety, Mechanical and Chemical Laboratory departments 

spent approximately 200 hours special training on urinalysis 

procedures and the operation and maintenance of monitoring 

equipment.

Ahlstrand (1958a)

July 1959 Radiation Safety inspector visits Pantex and LASL to discuss 

radiological and health problems. Ahlstrand (1959b)

1962 First instance of AEC Form 191 documenting no internal 

exposures at IAAP during the year.

AEC Report 191

(1962-1974)

September 1962

Routine issuance of film dosimeter badges to personnel. 

Badge exchange frequency was weekly. First routine area 

dosimeters in buildings 1-11 and 1-77.

Landauer Reports 

(1962)

January 1963 Badge exchange frequency changed to bi-weekly. Landauer Reports 

(1963)

January 1964

Badge exchange frequency changed to 4 weeks. A total of 

13 cycles per year. Some highly exposed workers continued 

on bi-weekly basis.

1968 Significant increase (doubling) in the issuance of dosimeters 

to personnel.

Landauer Reports 

(1968)

October 1969

IAAP establishes hand monitoring program (extremity 

dosimeters) following an AEC recommendation in 1969:

Hand exposure studies have been made in the past, with results that 

indicate only nominal exposure. However, no hand monitoring has 

been done within the past year. With new items that are coming into 

production, and the strong possibility of a reduced exposure guide for 

extremities,a program of hand monitoring should be provided for 

those employees who work directly with the radioactive units that 

have significant surface dose rates.

Davis (1969)
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In summary, the assumption that fissile materials were onsite since March 1949 is

considered claimant favorable for the following reasons.

• Evidence that prior to the assembly of the W-25 in 1957, non-radioactive model pits 

(mock-ups) were used for assembly (Ahlstrand 1955).

• From NIOSH’s review of the semi-annual reports, the only potential contact with 

fissile materials would be during retrofit or during disassembly operations, whereas 

the dominant process was assembly.

• Other documentation reviewed (Mitchell 2003) indicates that the nation’s nuclear 

capsules (pits) were initially stored at Sandia and Los Alamos. The list of storage 

areas greatly expanded through 1957 to encompass military bases in Kentucky, New 

Mexico, Louisiana, Maine, South Dakota, Washington, California, Virginia, Nevada, 

Massachusetts, Texas, and New York. From the documentation it is apparent that,

during assembly and transport, nuclear capsules were initially kept and stored

separately from the high explosives manufactured by IAAP.

• IAAP History of Line 1 Operations Reports on the production of the W-25 by the end 

of 1956 correlates with initial personnel monitoring beginning in November 1955 and 

the institution of a radiation safety program.

3.0 OCCUPATIONALLY RELATED MEDICAL X-RAYS

Medical examinations at the IAAP were required as a condition of employment. Generally,

an annual chest X-ray was given to each IAAP employee. During the worker outreach 

meeting in July 2004, some workers indicated that they received chest X-rays on a semiannual basis, while others indicated that it was more frequent on a quarterly basis. 

Assuming medical monitoring practices were similar between IAAP and Pantex, there is 

evidence that male employees in certain job categories (i.e., heavy lifters) received lumbar 

spine examinations, the frequency of which was not available. At Pantex, lumbar spine 

examinations were given to men when they were hired to check for pre-existing back 

conditions. As a result a single lumbar spine x-ray should be assumed for occupations that 

involve heavy lifting such as (assembly workers, transportation personnel, crafts, etc...)

Background information on X-ray doses can be found in Dose Reconstruction from

Occupationally Related Diagnostic X-ray Procedures (ORAU 2003a). Only limited data was 

found within the 1973 IAAP dosimetry records concerning occupational medical X-ray 

exposures at the hospital. These records indicate an exposure study was conducted in 

which several Landauer film badges were exposed to select X-ray machine settings. Table 

3.1 below summarizes the X-ray settings and measured film badge dose for this special 

study.
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Table 3.1 Results of X-ray study

# Amperage (mA) Voltage (kVp) Duration (sec) Dosimeter

Reading (mR)

1 200 mA 85 kVp 0.1 s 20

2 100 mA 95 kVp 0.6 s 690

3 100 mA 70 kVp 0.4 s 300

4 100 mA 45 kVp 0.05 s 0

A chest exposure of 20 mAs (200mA x 0.1s) is fairly common for typical chest exposures as 

indicated in exposure #1. Since this exposure was first in the series, it is expected that 

these were the default values for the common chest X-ray. As indicated in ORAUT-OTIB0006, X-ray measurements with films results in large uncertainty, and great care is needed 

to properly evaluate the dose. This study likely did not consider the calibration of the 

commercial film badge nor the effects that filtration has on the resulting spectral energy. As 

a result, the 20 mR reading could be either an underestimate or an overestimate depending 

on a multitude of variables. Because of this, the claimant favorable default values provided 

in ORAUT-OTIB-0006 should be used for chest X-rays (Tables 3.3-1 and 4.0-1).

The dose reconstructor should assume an annual posterior-anterior (PA) chest X-ray for all 

employees applicable from 1947 through 1975. Semiannual chest X-rays should be 

assumed for all assembly workers, and quarterly chest X-rays should be assumed for

workers involved in radiography. If the job description is unknown, the claimant favorable 

default assumption should be semi-annual examinations. Do not apply X-ray dose for years 

other than 1947 through 1975, regardless if the worker was employed at IAAP during other 

years.

ORAU (2003a) does not provide default values for lumbar spine examinations. The

Occupational Medical Dose TBD for the Rocky Flats site (ORAU 2004b) provides a method 

for calculating organ doses from lumbar spine examinations. Estimated median entrance 

skin exposures were 1.79 R for the AP view and 5.79 R for the lateral view based on 

information in Lincoln and Gupton (1958). Distributions of entrance skin exposures were 

created using the Crystal Ball® computer program (Decisioneering Inc. 2000) for different 

filtration half-value layers (HVL) as presented in Lincoln and Gupton. Tables A2 through A8 

in ICRP Publication 34 provide organ doses in units of mGy per Gray entrance skin

exposure for the thyroid, ovaries, testes, lungs, female breast, uterus, and active bone

marrow (ICRP 1982). For practical purposes, the units can be considered mrem per R. 

Multiplying the entrance skin exposure to the kerma-to-organ dose factors in ICRP

Publication 34 results in the organ doses and geometric standard deviations listed in Table 

3.2, varying the HVL from 1.5 to 3.0 mm AL. ORAU (2003a) provides guidance for dose to 

organs not provided in ICRP 34 by relating their proximity to organs that are listed in ICRP 

34. To account for a field of direct exposure larger than that used by ICRP 34 (as might 

have happened in the 1940s and 50s), additional organs were added to the ovary category, 

including stomach, kidneys, adrenals, and pancreas. Organs not listed by ICRP 34 added 

to the modeled organ based on proximity to the ICRP 34 organ (Table 3.2).

For skin and testes, consistent with the approach taken in the Rocky Flats Occupational 

Medical Dose TBD (ORAU 2003b), the values measured in the Lincoln and Gupton (1958) 

paper were used directly to account for the difference in collimation in the Lincoln and 

Gupton (1958) measurements versus ICRP 34 factors.
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The AP view should be used unless person-specific information is found indicating

otherwise. The lumbar spine organ dose should be assigned to males in the year of hire 

from 1947 through July 1975. The dose is assigned in the year of hire and is assumed to be 

lognormally distributed.

Table 3.2 Organ doses from lumbar spine X-rays at IAAP.

Organ View

Organ Dose

Geometric Mean mrem 

Geo. Std. Dev. (GSD)

Thyroid, eye, brain AP

LAT

0.50 (3.0)

0.10 (1.9)

Ovaries, liver, gall bladder, stomach, intestines, colon,

rectum, kidneys, adrenals, pancreas, spleen

AP

LAT

330 (2.6)

230 (2.1)

Lungs, thymus, esophagus, bone surfaces AP

LAT

120 (2.5)

64 (2.0)

Active bone marrow AP

LAT

58 (2.6)

110 (2.1)

TestesaAP

LAT

26 (3.1)

40 (1.5)

SkinaAP

LAT

1,800 (2.4)

5,800 (1.8)

a. Calculated from information in Lincoln and Gupton (1958) as presented in the Rocky Flats site TBD (ORAU 2004b).

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

4.1 INTRODUCTION

The occupational environmental dose refers to the dose received by workers outside of the 

normal production facilities. These doses can be internal and external depending on the 

characteristics of the individual radionuclides. Radionuclides present at IAAP include

tritium, uranium, plutonium, and thorium. Although there are some enhanced exposures to 

naturally occurring radon from certain structures, no noble gases are used or released at the 

IAAP site. While most radionuclides when inhaled would give a dose to particular organs in 

the body, tritium gas would give a dose to the whole body. These radionuclides are

addressed in the following sections.

4.2 INTAKES FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

Intakes to workers outside facilities are determined from air concentrations that resulted 

from individual facility releases and ground-level releases (e.g. burning activities). 

Unmonitored workers could have received internal or external occupational doses (or both) 

from any or all of these sources.
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4.2.1 Intakes from Tritium Releases

The University of Iowa Needs Assessment for IAAP (Fuortes 2001) quotes an annual 

release of 0.006 Ci (6,000 μCi) of tritium from the site; however, an effluent summary

document lists a total of 0.13 Ci for the period December 1965 through December 1970, for 

an average of 26,000 μCi per year. To be claimant-favorable, the latter value was used. In 

comparison to other sites that handled tritium, this release level is very small and provides 

some indication that not much tritium escaped containment. It is likely that most of the 

tritium released from stacks or vents was tritium gas, whereas the worker dose would come 

from tritiated water.

To estimate the intakes to workers outside facilities when little or no atmospheric information 

is available, the NCRP has suggested screening techniques. These techniques were

originally published as NCRP Commentary No. 3 (NCRP 1989) and updated in NCRP 

Report 123 (NCRP 1996). The NCRP recommends a graded approach, with three

screening levels. Level 1 is the most conservative and requires the least amount of input 

information. The Level 1 method for determining a conservative upper bound air

concentration can be written as:

( / sec)

 x ( / sec) ( / )

3

3

V m

f Q pCi X pCi m = Equation 4.1

Where: X = annual average upper-bound air concentration,

f = assumed fraction of time the wind blows in the direction of the subject, 

assumed to be 0.25,

Q = release rate of the radionuclide from the source, and

V = volumetric flow rate of the vent. The default value is 0.3 m3/sec, type of hood 

ventilation rates.

The model essentially assumes that the subject breathes undiluted effluent from the vent or 

stack, slightly modified by the fraction of time the wind blows in his/her direction (the factor 

of 0.25).

Supporting documentation for Federal Guidance Report 13 (Eckerman et al 1999) shows 

that dose per unit intake of tritium in the form of water vapor (HTO) is 10,000 times larger 

than the dose per unit intake for elemental tritium gas (HT). Because tritium (as water) can 

be absorbed through the skin, the tritium inhalation intake is multiplied by a factor of 1.5 to 

obtain the total intake of tritium water.

With the NCRP assumptions, the annual average air concentration near the release point 

should be less than:

26,000 μCi/yr * 1 x 106 pCi/μCi * 0.25 */ (3.15 X 107 sec/yr * 0.3 m3 /sec) = 688 pCi/m3

Assuming a breathing rate of 2400 m3/yr and assuming all the tritium is in the form of water 

results in an estimated annual intake of

688 pCi/m3 * 1.5 * 2,400 m3 = 2.5 x 106 pCi
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This intake is claimant-favorable because it uses the highest value for tritium releases, 

assumes minimal dilution between the release point and the worker, and much, perhaps 

most, of the inhaled tritium would have been tritium gas. The dose reconstructor should 

assign a 2,500,000-pCi intake per year as HTO to all workers (chronic, 6,800 pCi/D). The 

annual doses from these intakes are constant upper bound distribution types for input into 

IREP.

4.2.2 Intakes from Release of DU from Burning Sites

High explosives contaminated with DU were routinely burned in the Explosive Disposal 

Area, which is an irregularly shaped region of slightly less than 1 square mile just north of 

the “C” Yard and about 1 km south southwest of Line 1.

The source term has been determined to be about 2,000 g/year of DU (TN & Associates 

2001). This is an estimate of the material handled in the burn yard. Burning of DUcontaminated high explosives can be assumed to create aerosolized particles of DU. The 

most likely form of uranium released in the air from the burning would be in the form of an 

oxide, although the TN & Associates report suggested that much of the metal might not 

have been oxidized because the temperature was too low and the burning of explosives too 

rapid. Because uranium metal and some uranium oxides can exist in a chemical form 

associated with absorption type M, dose reconstructors should assume exposure to either 

type S or type M to maximize the dose to the organ/tissue of concern.

The burning was sufficiently frequent that modeling might consider it a continuous source 

during normal working hours. Thus, the 2,000 g/yr can be estimated as 1 g/work-hr.

The NCRP screening models for atmospheric releases provide a generic and conservative 

approach for estimating atmospheric dispersion (NCRP 1996). This approach is depicted in 

equation 4.2.

u

f Q P

C = Equation 4. 2

Where: C = the annual average air concentration, g/m3,

f = the fraction of time the wind blows in the direction of the subject, assumed to be 0.25

Q = the release rate, g/sec

u = the average wind speed, assumed to be 2 m/sec.

P = the Gaussian diffusion factor appropriate for down wind distances x

Values of P are provided in NCRP (1996, Figure 2.2). For ground-level releases for 

distances of 100 m, 500 m, and 1000 m, the values are 3.5 x 10-3, 2.0 x 10-4, and 5 x 10-5, 

respectively.

Individuals not directly involved in the burning operations could have been anywhere on the 

IAAP Site. The nearest portions of the C Yard are about 500 m, and the nearest portions of 

Line 1 are about 1,000 m away. The average calculated air concentration at 500 m is about 

6.9 x 10-9 g/m3; and at 1,000 m C is about 1.7 x 10-9 g/m3, assuming that all the DU becomes 

airborne, certainly a conservative upper bound. This approach neglects lofting of the plume 

caused by heat from the fire, which would reduce the calculated concentrations.
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Assuming a breathing rate of 2400 m3/yr, the amount inhaled at a location in or near the C 

Yard would be:

6.9 x 10-9 g/m3 * 2,400 m3/yr = 1.7 x 10-5 g/yr of 6.3 pCi/yr DU

Similarly, for locations at or beyond Line 1:

1.7 x 10-9 g/m3 * 2,400 m3/yr = 4.1 x 10-6 g/yr or 1.5 pCi/yr DU

If there is no information regarding location of worker activities, use the claimant-favorable C 

yard value.

4.2.3 DU Intakes from Potentially Contaminated Drinking Water

Drinking water for the Iowa Army Ammunition Plant (IAAP) during the period of operation by 

the Atomic Energy Commission was obtained from Mathes Lake (also referred to as Long 

Creek Lake) which is within the IAAP site perimeter. Mathes Lake was fed in part by Long 

Creek which drained the hydroshot firing site designated as FS-12. FS-12 was the principal 

firing site for hydroshots containing depleted uranium (DU) from 1965 through 1973. Hence, 

there was potential for DU contamination in Long Creek, Mathes Lake, and drinking water 

for the site.

Results of uranium measurements from the water sampling program that have been 

discovered to date are shown in Table 4.1. The reported results are for total uranium with 

no distinction made between DU and natural uranium. Fluorimetry was the analysis

technique used for the uranium measurement. The apparent detection limit for this

technique was 1.0 μg/L

Table 4.1 Total uranium results from water samples at IAAPa

Sample date Location

Measured U

concentration

(μg/L)

U concentration in

pCi/L assuming

100% DU

12/1968 Long Creek between FS-12 Trace <0.37

12/1968 SW of FS-12 1.2 0.44

12/1968 Reservoir 0 <0.37

12/1968 Drinking water I-04 Bldg. 0 <0.37

11/1969 Long Creek between FS-12 and lake Trace <0.37

11/1969 SW of FS-12 1.2 0.44

11/1969 Reservoir 0 <0.37

9/1970 Long Creek between FS-12 and lake <1.0 <0.37

9/1970 SW of FS-12 1.0 0.37

9/1970 Boat dock in lake <1.0 <0.37

9/1970 Reservoir <1.0 <0.37

3/1971 Long Creek between FS-12 and lake 2.0 0.74

3/1971 SW of FS-12 <0.4 <0.15

3/1972 Long Creek influent 2.20 0.82

3/1972 Long Creek influent 1.40 0.52

4/1973 Long Creek at FS Bridge 0 <0.37

4/1973 Drinking water in I-04 Bldg. 0 <0.37

a. From Shahan 1970, Shahan 1971, Meek and Shahan 1972, Holmberg 1974.
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The Technical Report No. 181, Part 5 (Meek and Shahan 1972), from which the 1972 data 

were obtained, stated, “Experience at this installation indicates that the typical or

background concentration of uranium in this area is as follows: Small Streams 1-3 μg/L...” 

The report also concludes, “Only background amounts of uranium (D-38) concentrations 

were detected in the various plant streams...” It is hard to judge for sure whether the 

concentrations of uranium in Long Creek were due to DU transported from the firing site or 

from natural uranium, although the site safety staff were of the opinion that the material was 

natural background uranium.

Even if there was DU in the intake to the water treatment plant from Mathes Lake, some of it 

would have been removed in the treatment plant. Water treatment was described as, 

“Water is treated to control odors and tastes by pre-exposure to activated carbon and 

chlorine. It is then subject to coagulation with ferric sulfate, lime, and soda ash. 

Following suspended solids removal and the softening process, it is clarified through 

rapid anthrafilt filters” (Shaykin 1969).

The two measured concentrations in the drinking water (1968 and 1969) were less than

detectable, as were all lake samples. With additional concentration reduction in the water 

treatment system, it is reasonable to assume that earlier and subsequent concentrations 

were also less than detectable.

Assuming that the upper bound drinking water concentration was just at the detection level, 

that all the activity was DU, that a worker might consume 2 L/day of water onsite, and 

adjusting for 250 working days per year to convert to intake per calendar day, the upper 

bound daily intake might have been:

intake (pCi/d) = (0.37 pCi/L) (2 L/work day) (250 work day/yr) / (365 cal. day per yr)

intake = 0.51 pCi/cal. day.

This ingestion intake rate was assigned to the entire period of the hydroshots, 1965 through 

1973, and organ doses were determined using IMBA assuming all the activity was 234U (f = 

0.02) which maximizes the dose. The organs with the highest annual doses were the bone 

surface and kidneys. The doses to those organs are shown in Table 4.2. The highest 

annual organ dose is 0.3 mrem to the bone surface.

Maximizing assumptions included in the calculations were

• all of the uranium was DU from site operations

• the concentration in Mathes Lake was just at the detection limit for the analysis

• this concentration existed for the entire period 1965 through 1974

• no credit was taken for removal of uranium by the water treatment system

• all of the DU activity was in the soluble form, despite the source term originating from 

explosions

• workers consumed 2 L of the site drinking water each work day, year round

•

234U was used for calculating organ doses.
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Based on the maximizing assumptions and the calculations shown above, it can be

concluded that the upper bound annual dose to any organ was less than 0.3 mrem. Annual 

doses to most organs were less than 0.1 mrem.

Table 4.2 Upper bound equivalent doses (mrem) from ingestion of DU contaminated drinking water 

Calendar

Year

Bone Surface

Dose (mrem)

Kidney Dose

(mrem)

Calendar

Year

Bone Surface

Dose (mrem)

Kidney Dose

(mrem)

1965 9.80E-02 9.27E-02 1986 7.20E-02 1.40E-02

1966 1.46E-01 1.10E-01 1987 6.80E-02 1.22E-02

1967 1.78E-01 1.21E-01 1988 6.46E-02 1.08E-02

1968 2.06E-01 1.31E-01 1989 6.12E-02 9.44E-03

1969 2.29E-01 1.39E-01 1990 5.83E-02 8.30E-03

1970 2.51E-01 1.46E-01 1991 5.57E-02 7.30E-03

1971 2.70E-01 1.52E-01 1992 5.35E-02 6.44E-03

1972 2.88E-01 1.58E-01 1993 5.12E-02 5.66E-03

1973 3.03E-01 1.63E-01 1994 4.92E-02 5.00E-03

1974 2.19E-01 7.40E-02 1995 4.73E-02 4.41E-03

1975 1.83E-01 6.05E-02 1996 4.58E-02 3.91E-03

1976 1.64E-01 5.30E-02 1997 4.41E-02 3.46E-03

1977 1.48E-01 4.62E-02 1998 4.26E-02 3.07E-03

1978 1.33E-01 4.04E-02 1999 4.12E-02 2.72E-03

1979 1.21E-01 3.53E-02 2000 4.00E-02 2.43E-03

1980 1.11E-01 3.10E-02 2001 3.86E-02 2.16E-03

1981 1.02E-01 2.70E-02 2002 3.74E-02 1.93E-03

1982 9.43E-02 2.37E-02 2003 3.63E-02 1.73E-03

1983 8.76E-02 2.07E-02 2004 3.52E-02 1.55E-03

1984 8.19E-02 1.82E-02

1985 7.65E-02 1.59E-02

4.3 EXTERNAL DOSE

Dosimetry records for IAAP indicate that radiation workers were the only employees 

monitored for radiation exposure. These personnel worked primarily in facilities in Line 1, 

Yard C storage area, the Explosive Disposal Area, and the Firing Site. Radiation workers 

accounted for a small fraction of the workers on the site. Employees working in other areas 

of the site were not monitored; however, there was a small potential for external dose from 

occupational environmental sources.

Workers at IAAP were subjected to external doses from the ambient radiation levels on the 

site. From 1962 through 1974, bi-weekly film badges were used for area radiation 

monitoring. These area badges were placed in various assembly buildings, Gravel Gerties, 

and Storage Igloos. Most of this monitoring data ( ̃ 70%) was below the detection limit of 10 

mR (Figure 4.1). Assuming a 2080 hour work year and using the dose distribution data in 

figure 4.1, an onsite ambient dose is estimated to be 37 mR per year with a geometric 

standard deviation of 4.6.
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Probability (%)

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Bi-weekly Area Dose Measurement (mR)

0.1

1

10

100

1000

Coefficients:

GM = 6.0 mR

GSD = 4.6

r 2 = 0.94

n = 5164

Figure 4.1 Probability plot of bi-weekly non-storage area dosimeters from 1969 - 1974

Considering the placement of the area dosimeters (i.e. near fissile material handling

operations), the onsite ambient dose throughout the remainder of the plant was near 

background levels. Instead of using the distribution illustrated in Figure 4.1, the dose

reconstructor can use an upper bound of 260 mrem/year to estimate the onsite ambient 

dose for the unexposed (non-line 1) workers. This estimate assumes a non-line 1 worker 

was exposed at the detection level of the area dosimeters for an entire year.

Although area dosimetry data in process areas was not abundantly available prior to 1969, 

extrapolation to the early time periods is considered claimant favorable. Area dosimeter 

data is available for the storage areas as shown in figure 4.2 from 1962 -1974. Also

depicted in this figure is the upper 95th% of the non storage area distributions which are 

observed to be relatively stable and follow the general dose rates in the storage areas. As a 

result, the onsite dose distribution for non-line 1 workers is expected to be relatively stable 

throughout the operations. 
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Year

60 65 70 75 80 

Area Bi-weekly Dose Measurements (mR)

10

100

1000

10000

95th % in Assembly areas (non-storage areas)

Fissile Material Storage Areas

95th Percentile of Bi-weekly 

dose distribution in 

non-storage areas

Figure 4.2 Plot of Area Monitoring data from 1962 – 1974

4.4 UNCERTAINTY

As discussed above, estimates of annual occupational environmental doses were based on 

accepted screening techniques (NCRP 1989, 1996). These techniques in themselves rely 

on considerable conservatism based on nominal values and uncertainties of known

parametric values. In addition, the analyses made additional claimant-favorable 

assumptions, as stated. Because of the scarcity of available environmental data and the 

use of multiple claimant-favorable assumptions, the environmental intakes should be

considered a constant upper bound.

5.0 ESTIMATION OF INTERNAL EXPOSURE

A 1969 Health Protection Appraisal (Shaykin 1969) indicated that swipe counting and direct 

surveys routinely monitored removable and non-removable radioactive contamination at 

IAAP. A personal interview with Joe Shannan, long-time (1958-1985) employee in the 

Safety Department and eventually Radiation Safety Manager, confirmed that shipments of 

radiological materials were swipe tested on entry and before being sent to the assembly 

facilities (Fix and Bihl 2003). The contents of the containers were swipe tested when the 

containers were opened. Mr. Shannan said that during his employment at the BAECP, 

contamination outside or inside the incoming containers was rare.

Based on materials used to assemble weapons and considering that the fissile material was 

encapsulated (sealed), the radionuclides most likely to result in an intake at IAAP were 

depleted uranium (DU, D38, or tuballoy) and tritium (3H). A radiological survey of the plant 
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conducted in 2001, years after radiological work had ceased, found only DU and 137Cs. The 

document providing the results of the survey also states, “Radioactive materials used at the 

line were received in a sealed configuration and were swipe tested before use. Known 

radioactive materials include depleted uranium (DU), enriched uranium (EU), plutonium,

tritium gas, and polonium (210Po).” The source of the 137Cs on the swipes is unknown but 

seems inconsistent with the nature of the work. When questioned specifically on the source 

of the 137Cs, Mr. Shannan said he was not aware of any 137Cs used at IAAP other than small 

sealed sources at μCi levels as instrument check sources.

Selected employees were given bioassays to detect intakes of radioactive material. 

However, no bioassay records were discovered during the records search. Tritium gas 

reservoirs, which arrived sealed in metal containers, were purged under a ventilation hood 

prior to assembly into weapons. The actual emplacement of the reservoirs into the weapons 

did not involve breaking seals or the release of tritium gas. The plutonium and EU were also 

encapsulated and not available for intakes, even during disassembly. Pantex did some 

plutonium bioassay, but no intakes were ever recorded, with the exception of one welldocumented accident. No reports of similar accidents at IAAP were found in the literature. 

If early IFI style weapons were returned from the field for retrofit and disassembly with the 

nuclear capsule, then 210Po might be present. Polonium-210 was also encapsulated and 

was not further disassembled at IAAP.

Note: Intakes discussed below and in Sections 5 are summarized by year and work task in 

Table C-1 in Attachment C.

5.1 TRITIUM EXPOSURE

Tritium intakes could have occurred and probably did occur to a certain extent during 

weapons assembly and disassembly procedures. The tritium reservoirs came from the

Savannah River Site (SRS), so the earliest possible date for tritium exposure at IAAP would 

be 1954. Based on a review of classified documentation, first weapon assembled that used 

tritium was after 1954, thus it is a claimant-favorable overestimate to assume that tritium 

intakes occurred each year that a worker was involved in assembly, disassembly,

inspection, stores, shipping and receiving (and other duties with elevated tritium exposure 

potential) from 1954 to 1975.

The technology of tritium usage in nuclear weapons is classified. No tritium bioassay results 

for IAAP workers have been located. However, material and procedures at IAAP were 

almost certainly the same as those at Pantex because the same company operated both 

plants, and the materials and tasks were the same. This assumption was confirmed in a 

telephone interview with Mr. Herman Phillips, who was a safety engineer at Pantex but also 

worked for some time at IAAP. Hundreds of tritium bioassay results were obtained at 

Pantex in the 1970s and 1980s. The largest internal dose at Pantex from tritium recorded 

during any year in this period, with the exception of a major accident, was 122 mrem. Using 

the standard calculation for tritium in the 1970s, which used a quality factor of 1.7 [based on 

International Commission on Radiological Protection (ICRP) Publication 10 (ICRP 1968) and 

explained in NUREG-0938 (NRC 1983)][1.5 mCi = 425 mrem], 122 mrem was indicative of a 

chronic annual uptake of 430 μCi of tritium. The uptake as calculated using ICRP 10 

methodology accounts for tritium in body fluids from any and all intake modes.
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Routine IAAP tritium air sampling results from 1959 through 1964 were located and used to 

estimate the potential upper bound intakes for workers involved in tritium operations. Tritium 

reservoirs were shipped to the Iowa Ordnance Plant in pressurized metal vessels known as 

JP containers. Prior to the opening of the JP container, which contained the tritium

reservoir, the pressurized air was vented in a controlled manner under a ventilation hood. 

The air purged from the container was surveyed with a T-289 tritium monitor to ensure that 

the tritium air concentrations were less than 90 μCi/m3 before the container was released to 

production (Shaykin 1969).

Print out charts from the tritium monitors were used to estimate upper bound tritium intakes. 

It was assumed that 2 JP containers containing tritium air concentrations at the release level 

of 90 μCi/m3 were opened each day in the relatively small enclosed space of a gravel gertie. 

The JP container volume was estimated to be 0.0136 m3, based on an inner radius of 15.5 

cm and a container height of 18 cm (DOE EIS-0225), thus the tritium activity present in the 

headspace would be about 1.22 μCi per container. Gravel gertie dimensions (DOE NV-710, 

2001) were used to estimate a volume of 437 m3 into which the tritium was dispersed. Thus, 

for intake and dose calculation purposes, it was assumed that 2.445 μCi of tritium was 

vented into the cell volume of 437 cubic meters for 365 days each year. The resulting upper 

bound air concentrations to which workers were potentially exposed was estimated to be 

2.043 μCi/m3. For intake calculations, it was assumed that an assembly/disassembly worker 

was involved in tritium operations for 8 hours per day of a 2000 hour work year, with a 

breathing rate of 1.2 m

3

/hour. The resulting upper bound chronic tritium intake was

estimated to be 4902 μCi/year, or 13.4 μCi/day, which corresponds to an upper bound 

tritium dose of 0.331 rem/year in IMBA Expert OCAS Edition.

In all AEC Form 191s generated at IAAP from 1962 to 1973 (AEC 1962-1973a), the site 

reported “no internal deposition during the period.” In a health protection appraisal and 

report in September 1969, C. E. Davis, from the AEC Albuquerque Office, stated:

“Routine internal exposure monitoring is provided only for tritium. Two urine samples

are analyzed biweekly for various selected individuals who work in areas where there

is potential for exposure. To date, there has never been a positive result. If air

samples, radiation survey results, or unusual conditions should indicate the possibility 

of internal exposure to any radioisotope, special bioassays would be necessary. 

There has been no occasion for such tests to date.” (Davis 1969).

Although the criteria for IAAP tritium monitoring, reporting levels and positive results have 

not been found, historical reviews of detection capabilities at other AEC programs indicate 

that the IAAP annual uptake of 4902 μCi would have been readily detectable and, therefore, 

this value is considered an upper estimate.

Another approach for estimating intakes of tritium uses tritium effluent values. There was 

some information regarding the release of tritium from the site in the 1970s. The University 

of Iowa Needs Assessment for IAAP (Fuortes 2001) quotes an annual release of 0.006 curie 

(6,000 μCi) of tritium from the site. But another 2001 report (TN & Associates 2001) quotes 

an effluent summary that indicates annual releases of about 0.026 Ci in the latter half of the 

1960s. Either way, in comparison to other sites that handled tritium, this release level is 

very small and provides some indication that not much tritium ever escaped containment. 

Most of the tritium released from the stack was probably tritium gas, whereas the worker 
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dose comes from tritiated water. The 4902 μCi annual intakes estimated in the paragraphs 

above are almost 19% of the total airborne effluent for a typical year in the 1960s and are 

approximately 80% of the reported effluent for other years. It is improbable that any single 

worker had a chronic annual intake of the magnitude assumed for dose calculation

purposes, especially not of tritiated water.

As discussed in the preceding paragraphs, the 4902 μCi intake is an overestimate,

especially when applied for all years of employment; therefore, for IREP input the

distribution type is constant. Because of the way IMBA handles tritium intakes, uptake 

equals intake (personal communication Anthony James 2003), so the 4902 μCi should be 

modeled as an 4902 μCi injection. 

5.2 DEPLETED URANIUM

Intakes of DU might have occurred during disassembly of old, oxidized DU bomb parts, 

during hydrotesting, or during machining of baratols (explosive charges that surround the 

ball). DU emits less radiation per gram than natural uranium. By weight DU is essentially 

pure 238U. Isotopic abundances of 234U and 235U in DU can vary, but those isotopes 

generally contribute less than 10% of the alpha radioactivity. Typical weight percents and 

activity fractions of uranium isotopes are listed in Table 1. These activity fractions are 

slightly different from the default values in IMBA, compiled for the NIOSH’s Office of

Compensation Analysis and Support (OCAS), but are within the variability of batches of DU. 

Either the values in Table 5.1 or the default IMBA values can be used.

Table 5.1 Mass and radiological characteristics of depleted uranium.

Weight percentage a DU

234U 0.0005

235U 0.2500

236U Negligible

238U 99.7500

Specific constituent activity in mixture (μCi/g, nCi/mg, or pCi/μgb

234U 0.0313

235U 0.0054

236U Negligible

238U 0.3352

Total 0.3718

Specific constituent activity in mixture (dpm/μg)b

234U 0.0694

235U 0.0120

236U Negligible

238U 0.7441

Total 0.8254

a. From Carbaugh 2003.

b. Can be used to represent specific alpha activity as well.

The chemical and radiological risks of DU were acknowledged at IAAP, but they were

generally considered insignificant in relation to other chemical hazards (such as beryllium); 

therefore, only limited air sampling data were taken during DU operations. Concerning the 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 26 of 79

inhalation absorption type, the Pantex Internal Dosimetry Technical Basis and Quality

Assurance Document states that “the compounds of uranium at Pantex are pure metal or 

air-oxides; it is assumed that all forms encountered will exhibit class Y aerosol behavior” 

(Pantex 2001, Section 13.2.1). The same document lists in Table 7.3 that 238U should be 

considered 20% class D and 80% class Y. Uranium contamination at IAAP would be similar 

to that at Pantex with the exception of material detonated in hydroshots. The most likely 

form of uranium at IAAP would be very insoluble, associated with lung absorption type S 

and a gastrointestinal-tract-to-blood uptake factor, f1, of 0.002. However, uranium oxides 

can exist in many states, and it might be too simplistic to assume a pure absorption type 

when the chemical form is not known for certain. The dose reconstructor should assume 

either type M or type S to maximize the dose to the organ of concern. Exposure to type F 

uranium at IAAP is not considered credible.

5.2.1 DU Intakes from Hydroshots

From 1965 to 1973, there was a potential for workers to be exposed to DU oxide-bearing 

dust in proximity to the North Firing Site 12 (FS-12) immediately following the detonation of a 

hydroshot (hydrodynamic test)). In addition, there might have been a few DU tests at the 

South Firing Site 6 (FS-6), though the number there was small in comparison to FS-12. A 

hydroshot was a diagnostic operation that used DU as a surrogate for weapons-grade 

material, and was a quality control technique for measuring the performance of plasticbonded explosives (PBX) produced at IAAP. In a personal interview, Mr. Shannan described 

the manner in which the outdoor hydroshot operations were conducted (Fix and Bihl 2003). 

One or two persons occupied the test fire control bunker, which was next to ground zero. All 

other site employees were kept outside a fenced area with the closest proximity about 1 mile 

from ground zero. A cable tunnel ran underground from the test fire control bunker to 

ground zero. A driver was at the fence gate. Within minutes of the explosion, the driver 

would enter the restricted area, pick up the workers in the bunker, and drive to the blast area 

to retrieve instruments. Then the workers would leave the fenced area. Neither the driver 

nor the control bunker operators wore respirators. According to Mr. Shannan, exposure to a 

plume would have been for a few minutes at most. Records indicate that 701 hydroshots 

occurred between 1965 and 1973 at FS-12, reportedly involving approximately 4,000 kg of 

DU (ATSDR 2003). Mr. Shannan said that shots were infrequent but could have been 

bunched, including more than one on a given day. Records list 530 hydroshots between 

December 2, 1965, and March 3, 1969, 3 hydroshots under a different program presumably 

between March 4 and July 14, 1969, and 168 hydroshots between July 15, 1969, and 

December 31, 1973.

The limited amount of air sampling performed after hydroshots in 1971 and 1972 (Meek and 

Shahan 1972) is summarized in Table 5.2. The data were plotted on log-probability paper to 

determine the geometric mean (or median) and geometric standard deviation (GSD) air 

concentration values. Only a few samples were taken at the control bunker, and the results 

were less than the 100-yard concentrations. Nevertheless, the data from the FS-12 tunnel 

were more robust, were more claimant-favorable, and compensate for the trip to the blast 

area after the dust had settled. Assuming an exposure of about 30 minutes for the 

operators and driver for each shot, the intake of DU for each of the periods listed in the 

above paragraph would be:

(air concentration μg/m3) (1.2 m3/hr breathing rate)(0.5 hr) (no. of shots in the period)
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 27 of 79

Table 5.2. Measured DU air concentrations from hydroshots.

Location DU concentrations (μg/m3)

a Geometric mean air 

concentration (μg/m3)

b

FS-12 tunnel 0.0 – 21.82 2.8

100 yards from shotc0.0 – 9.12 0.9

1 mile from shotd0.0 – 2.47 0.24

a. Air sampling data from Meek and Shahan (1972).

b. Taken directly from log-probability plot of air concentrations

c. Includes three data at 150 yards.

d. Includes one datum at 0.75 mile.

Based on a field investigation at Los Alamos Scientific Laboratories in November 1974, it 

was determined that approximately 10% of the total mass of uranium was aerosolized in a 

hydroshot. Data show that uranium particle sizes were lognormally distributed with an 

aerodynamic mass median diameter of 0.1 to 1 micron and a standard deviation of 8 for 

three separate experiments (Dahl and Johnson 1977). For FS-12 workers, an activity 

median aerodynamic diameter (AMAD) of 1 micron for uranium particles will be used in lieu 

of the IMBA default value of 5 micron AMAD. The DU intakes from hydroshots were

assumed to be chronic, and lognormally distributed. For the drivers and test fire operators, 

the geometric mean of 2.8 μg/m3 was used and the GSD was 4.6. Intakes for the three 

periods are provided in Tables 5.3 and C-1 (Table C-1 also has results in pCi/d). For input 

into IMBA, the intake per calendar day is needed, which was determined by dividing the total 

intake for the period by the calendar days in the period.

Table 5.3. Chronic DU intakes from hydroshots for test fire operators and drivers.

Period of exposure Total intake 

(mg) Calendar days Daily chronic intake (mg)

December 2, 1965 through March 3, 1969 0.890 1187 7.5E-4

March 4, 1969 through July 14, 1969 0.00504 132 3.8E-5

July 15, 1969 through December 31, 1973 0.282 1630 1.7E-4

Everyone at the site might have been exposed to a small degree to the plumes from

hydroshots. It is unlikely that the plumes always drifted in the same direction, and the 

decrease in air concentration as the plume moved across the site is not known. But

because the nearest AEC facilities were about equally distant from the fence as was ground 

zero to the fence, an assumption of a factor-of-4 decrease from the mean of the 1-mile air 

concentration is claimant-favorable for an annual intake, especially considering the high 

density of DU in relation to dust or nearly all other types of airborne effluents and variable 

plume directions when averaged over a year. The assumed period of exposure at this 

distance was 2 hr (assumes turbulence type A at 4000 m and 1 m/s drift speed then doubled 

to ensure conservativism (claimant favorability)); other assumptions are the same as for the 

operators.

Intake = (0.24 μg/m3)(1.2 m3/hr breathing rate)(2.0 hr)(no. of shots in the period)/4

Intakes for all other personnel from the hydroshots are provided in Table 5.4 and C-1. The 

intakes are chronic, lognormally distributed, with a GSD of 4.0, obtained from the probability 

plot.
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Table 5.4 Chronic DU intakes from hydroshots for other AEC personnel.

Period of exposure

Total intake 

(mg) Calendar days Daily chronic intake (mg)

December 2, 1965 through March 3, 1969 7.63E-2 1187 6.4E-5

March 4, 1969 through July 14, 1969 4.32E-4 132 3.3E-6

July 15, 1969 through December 31, 1973 2.42E-2 1630 1.5E-5

An unsigned, undated record in the IAAP files (file locator 000916) indicates that FS-12 

employees picked up pieces of D-38 metal lying around ground zero by hand without gloves 

to bag them as waste. Therefore, it is possible that a potential for ingestion of DU existed. 

Assuming 76 mg of DU contamination on hands, of which about 10% is ingested (assuming 

hands are not washed before eating), results in an ingestion of about 7.6 mg of DU per 

cleanup task. (Details of the calculation are provided in Appendix C.) There is no record of 

how many times cleanup of uranium pieces occurred; a claimant-favorable assumption is 

after each hydroshot. Results of the calculations are provided in Tables 5.5 and C-1. This 

should be modeled as chronic ingestion, insoluble material, constant upper bound.

Table 5.5 Chronic DU ingestion intakes from hydroshots for cleanup crew.

Period of Exposure Total intake 

(mg) Calendar days Daily chronic intake (mg)

December 2, 1965 through March 3, 1969 4.0E3 1187 3.3

March 4, 1969 through July 14, 1969 2.3E1 132 0.17

July 15, 1969 through December 31, 1973 1.3E3 1630 0.78

In addition to the potential for intakes of DU during the hydroshots, there was also

determined to be a potential for DU intake from the resuspension of contaminated soil

following routine high explosives detonations at FS-12. Air monitoring data were located in 

a document titled: “Health Protection Survey of BAECP – 1974,” (Holmberg, 1974) which 

indicated that the DU deposited in the soil immediately surrounding ground zero of the firing 

site was being resuspended during routine detonation of high explosives (non-hydroshots).

Several weekly continuous air monitoring results were reported in 1974 for the resuspension

of DU during non-hydroshot detonations at FS-12. The highest air sample result located 

was reported to be 101.4 E-13 μCi/cm3 for an 8 minute sample time. As the air sampling 

volume and time after the HE detonation increased, the airborne DU activity quickly

dropped. The potential for DU resuspension was linked to the number of high explosive 

tests at FS-12 in any given day. Given that the exact number of high explosive tests could 

not be located, for the purpose of intake estimation and dose calculations, the second

highest of the reported DU air concentrations (1.96 E-12 microcuries/cm3) was assumed to 

have been continuous at FS-12 for the period of 1965 through 1975. Using the air

monitoring data from 1974 to estimate intakes from 1965 – 1974 is claimant-favorable, since 

no additional depleted uranium was being deposited at FS-12 following the last hydroshot in 

1973. It was assumed that an FS-12 employee would have been exposed to this level of 

airborne DU for an entire 40 hours per week, even though additional air monitoring data 

indicates that this continuous exposure potential did not exist. In addition, it was assumed 

that employees were exposed for 10 hours per week to the highest recorded DU air 

concentration of 101.3 E-13 μCi/cm3. This approach overestimates the actual intake
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 29 of 79

potential and the subsequent dose resulting from the intake of DU at FS-12. The 

calculations for this exposure potential are provided below.

Constant DU air concentration = (1.96 E-12 μCi/cm3) (1 E +6 pCi/μCi) (1 E + 6 cm

3

/m3)/(DU specific 

activity of 0.402 pci/μg) = 4.87 μg/m3

Intermittent DU air concentration = (101.4 E – 13 μCi/cm3)(1 E + 6 pCi/μCi)(1 E +6 cm3/m3)/(DU 

specific activity of 0.402 pCi/μg) = 25.22 μg/m3

Intake = (4.87 μg/m3)(1.2 m3/hr)(40 hrs/week)(52 weeks/year) = 12156 μg/year

 = (25.22 μg/m3)(1.2 m3/hr)(10 hrs/week)(52 weeks/year) = 15737 μg/year

 Total Annual DU Intake from resuspension = 27893 μg/year or 11217 pCi DU / year

For input into IMBA Expert, a chronic daily upper bound inhalation intake of 30.7 pCi/day for 

depleted uranium should be assigned as either absorption Type M or S.

5.2.2 DU Intakes from Machining Baratols

On Line 1, from 1948 through about 1962 (TN & Associates 2001), the first step of the 

production process was the casting of baratols. Machining or grinding these components 

might have released small quantities of DU if conducted during assembly. Although DU was 

not machined directly, unintentional “nicking” of the DU occasionally occurred during

machining on the explosive charges. DU-contaminated explosive waste was reportedly 

taken to the Explosives Disposal Area burn pads for burning. Beginning in about 1962, the 

process of casting baratols was replaced by a new process that involved pressing

explosives in a plastic state into molds. Thus, the need for machining was eliminated 

(ATSDR 2003).

Mr. Shannan confirmed that machining directly on DU was not done because it produces 

hot, smoldering filings that would have been extremely hazardous because of the intimate 

proximity to explosives (Fix and Bihl 2003). Machining on contaminated metals might have 

produced some airborne contamination and cleanup operations around the machines and

have created low, temporary airborne concentrations of DU. An assumption of some intake 

of DU by the machinists is reasonable, although not comparable to sites where actual

machining on uranium occurred. Airborne DU contamination was probably intermittent and 

did not exist for the full 40 work hours every week, and it is unlikely that any worker was 

exposed for the full 40 hours each week. An exposure at 2% of the maximum permissible 

air concentration (MPC) for 20 hours per week is assumed as an upper bound for machining 

or cleaning around the machines. This assumes concentration of airborne contamination is 

consistent with values measured at the Hanford Site from machining of uranium and

cleanup of machinery at a fuel fabrication plant (Wilson 1958), and is believed to be 

claimant-favorable because the source term at IAAP was much smaller. The MPC for 

insoluble 238U (which would apply to DU as well) established by the National Commission on 

Radiological Protection and Measurements (NCRP) in 1959 was 1 x 10-10 μCi/cm3 (NBS 

1959). Assuming a breathing rate of 9.6 m3 per workday (light work) results in a chronic 

intake of DU of about 2.4 x 10-3 μCi per year (6.6 pCi/day for input into IMBA).
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 30 of 79

In addition, the machinists might have ingested DU by transfer from contaminated hands to 

food or cigarettes. For estimating ingestion resulting from contamination inside buildings, 

ORAU recommends a daily ingestion of 0.2 times the air activity per m3 (ORAU 2004c). 

This approach includes ingestion from transfer from hands and settling of contamination 

onto open sources of drink, such as a coffee mug. The approach assumes continuous 

settling of material from the air onto surfaces for 24 hours per day, 7 days per week. The air 

concentration discussed in the preceding paragraph (2% MPC) was assumed to apply only 

about half the time, so the average continuous air concentration would have been 1% MPC 

or 1 x 10-12 μCi/cm3. The daily ingestion intake would have been

Ingestion Intake = (0.2)(1 x 10-12 μCi/cm3)(106 pCi/μCi)(106 cm

3

/m3) = 0.2 pCi/d.

0.2 pCi/d should be the mode of a triangular distribution with the minimum at 0.1 pCi/d (no 

open drinks) and the maximum at 0.4 pCi/d (to account for the possibility that some

contaminants on the hands may have come from handling a baratol that was contaminated 

on the surface as opposed to touching general work surfaces (infrequent but possible when 

the DU was “nicked”).

5.2.3 DU Intakes by Operators at Burning Yard

About 2,000 g/yr of DU as contamination on scrap explosive components was burned at the 

Explosive Disposal Area (TN & Associates 2001). Section 5.2.2 addresses inhalation by 

general plant workers from the effluent of the burning. The ash was bagged and shipped off 

the site. Intakes might have occurred during the bagging of the ash. Probably more than 

99% of the DU remained in the ash, so 100% was assumed (airborne release fractions from 

burning DU are generally 10-3 or 10-4 [DOE 1994]). This means about 10 g/workday was 

bagged. Even when mixed with non radioactive ash from the explosives, the total amount of 

ash bagged per day was small and should have taken only a few minutes to sweep up and 

dispose of in a bag or drum. Airborne release fractions and respirable fractions of

radioactive materials, including uranium, under many different scenarios have been

compiled by Doe (DOE 1994, pages 4-9). The scenario considered most appropriate for 

bagging ash was described as “free-fall spill of cohesionless powders: free-fall <3 m, air 

velocity normal to powder flow, general forced enclosure ventilation or low-wind outside 

conditions.” The median airborne release fraction was 3 x 10-4 and the median respirable 

fraction was 0.5; the upper bound values for the same parameters were 2 x 10-3 and 0.3. 

The upper bound values might apply to an acute event, but for daily intakes the median 

values were considered more appropriate. It was assumed that the dust produced from this 

process was dispersed in 1 m3 of air. Because of the small amount of ash, 5 minutes was 

assumed as the time for gathering the ash and disposing of it in a bag or drum. The DU air 

concentration from this activity was:

(source μg/d)(airborne release fraction)(respirable fraction)/(air vol. m3)

or (10,000,000 μg/d)(3 x 10-4)(0.5)/1 m3 = 1,500 μg/m3

The DU inhaled was: (airborne concentration)(breathing rate)(time)

(1,500 μg/m3)(1.2 m3/hr)(0.0833 hr) = 150 μg/workday.
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The inhalation per workday is equal to 100 μg/calendar day or 38 pCi/calendar day. This is 

assumed to be the median value of a lognormal distribution with a GSD of 3.

This intake would apply to the period from 1949 through 1975. Exposure to the plume from 

burning was considered for burning yard workers using the same approach applied to 

workers outside the burning yard, except 100 m was assumed for the source-to-worker 

distance. Section 5.2.2 describes the calculation of intakes from the plume. This source of 

intake was calculated to be < 1 μg/work day, and was considered negligible compared to the 

intake from the cleanup of the ash, which would apply to the same workers.

5.2.4 DU Intakes from Disassembly of Weapons

Disassembly of nuclear weapons might be another source of intake of DU. Evidence

gathered at Pantex indicated that the DU material in the disassembled weapons was

generally clean metal with minimal potential of airborne contamination. Mr. Shannan 

indicated that contaminated internal parts were rare, and, when necessary, decontamination 

was performed before work on the weapons was started (Fix and Bihl 2003). However, 

there is a possibility that disassembly activities were different in the early years at IAAP prior 

to when Mr. Shannan was employed and routine low level DU contamination was more 

common. Although the primary mission of IAAP was assembly of new weapons, there was 

some early disassembly conducted in the 1950s and surveillance activities through the

1960s and early 1970s.

Experience at Pantex indicated that “about a half of a cup” (118 cm3) of oxidized DU was 

available for resuspension. The density of UO2 is 11 g/cm3, thus the mass of UO2 is 

estimated to be approximately 1300 g, of which about 1140 g is DU [(238/270)(1,300g) = 

1,140g.]

The airborne release fraction and respirable fraction were obtained from the DOE handbook 

(1994) using the scenario described as the free fall spill of UO2 powder from a height of 1 

meter. The median value for the airborne release fraction was 0.00008, and the respirable 

fraction was 0.5. The volume of air into which the contamination was suspended was 

assumed to be 27 m3, and the exposure time was 1 hour/day. These latter two assumptions 

assume quick work and no dilution by ventilation. Conversely, it could have been assumed 

that the work pace was slower but the concentration was decreased by ventilation, which 

would have produced about the same result. The estimated intake of DU per disassembly is 

then:

(1,140g)(0.00008)(.5)(1.2 m3/hr)(1 hr)(106 μg/g)/27m3 = 2027 μg or 815 pCi

Based on conversations and interviews with former employees, it was determined that

surveillance, testing, and disassembly occurred more frequently than initially credited, thus 

providing the potential for exposures to DU. Although it is unlikely that the amount of 

oxidized DU estimated above was present in all cases, the claimant-favorable assumption 

was made that the potential for intake of this quantity did exist throughout the covered time 

period at IAAP (1949 – 1974). For the purpose of dose reconstruction, it was assumed that 

on average 100 surveillance/disassembly operations might be conducted by a single worker 

per year with the potential for internal exposures to the quantity of DU estimated above. 
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The intake should be modeled as a chronic annual intake of 81500 pCi of DU per year, as a 

constant upper bound.

It is important to note that not all weapons programs exhibited oxidized DU components. 

There is evidence from sampling records that some systems used a stainless form of

uranium in some components called Mulberry. As a result, the oxidation was greatly 

reduced such that the above calculation would overestimate any potential uptake. Since 

NIOSH cannot publicly identify the programs that had significant DU oxidation, the claimant 

favorable assumptions are applied to all disassembly workers over time.

5.3 RADON

Weapons assembly/disassembly was conducted in bays and special cells called Gravel 

Gerties that were at ground level but had an overlay of earth on the roof and part-way up the 

sides. Three hundred forty-two radon measurements were taken in various buildings at 

IAAP by the Army from December 1989 through January 1991 (not including a few outlier 

values that had been scratched off the dataset) (Tec/Ops Landauer, Inc. 1991). To date, 

the authors have gained access to the results of the measurements, but not the link 

between results and specific buildings; hence, the information is of marginal value. 

Nevertheless, the average, standard deviation and geometric mean of the data are less than 

the corresponding values from the Pantex data discussed below. Without additional 

information, it was deemed claimant-favorable to use the Pantex data.

A DOE complex-wide survey of radon levels was performed in 1990 (UNC Geotech 1990). 

Most of the Pantex measurements were made over a 2-month period during the winter, 

normally expected to be the time with the highest radon concentrations because buildings 

are closed and heated most of the time. There were 137 locations sampled at Pantex 

including in bays and Gravel Gerties of similar design to those at IAAP. The data for the 

Pantex Plant were listed in their entirety in Table 5-11 in the Pantex Internal Dosimetry TBD 

(ORAU 2004d) and are summarized in Table 5.6. Pantex radon measurements and dose 

calculations were assumed to be the best indicators of radon exposure at IAAP, and were 

used as discussed below.

Table 5.6. Summary statistics of 1990 radon measurements at Pantex.

Parameter All buildings Underground 

buildings

Above-ground 

buildings

GeoMean (pCi/L) 1.37 1.51 1.33

GSD 1.68 1.75 1.66

Min (pCi/L) 0.8 0.8 0.8

Max (pCi/L) 8.1 7.1 8.1

Max/Min 10.1 8.9 10.1

Count 137 31 106

As listed in Table 5.6, the geometric mean (median) for all buildings at Pantex was 1.4 pCi/L 

with a GSD of 1.7. Values ranged from 0.8 to 8.1 pCi/L. Underground buildings had a 

slightly higher median concentration than above-ground buildings. Gravel Gerties and bays 

were considered “underground,” albeit not below “grade.”
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The Pantex measured radon concentrations were converted to equilibrium equivalent 

concentrations by multiplying the radon concentration by the equilibrium factor F using an 

assumed F of 0.4, as recommended by the ICRP (1981) and the United Nations Scientific 

Committee on the Effects of Atomic Radiation (UNSCEAR 1993). The equilibrium

equivalent concentration was divided by 100 pCi/L per working level (WL) to arrive at a 

potential alpha energy concentration (PAEC). These operations were combined to create

PAEC = CxF/100 pCi/L/WL

where C is the radon concentration in pCi/L and PAEC is in working levels. The PAEC is 

multiplied by the months per year the worker is exposed to determine the exposure in 

working level months (WLM) for input into IREP.

Because knowledge of whether a worker spent most of his/her time in a facility with an 

earthen cover will probably not be obtainable, dose reconstructors should use the Pantex 

median value for underground buildings, 1.5 pCi/L, for C and 12 months for the period 

(unless the person only worked for part of a year.) This results in an annual average 

exposure of

(1.5 pCi/L)(0.4)(12 M)/100 [pCi/L]/WL = 0.072 WLM per year

Radon exposure is only assigned when lung is selected as the cancer model in IREP. The 

exposure distribution is lognormal. Parameter 1 is the median value in working level 

months. Parameter 2 is the GSD. Use a GSD of 3 to allow for uncertainties in the

application of the 1990 radon measurements to the full-time period 1948-1975 and possible 

differences between Pantex and IAAP.

5.4 OTHER SOURCES

Enriched uranium (EU), plutonium, thorium, and perhaps Po210 were present at various 

times during assembly or disassembly of nuclear weapons. All of these sources were

encapsulated (sealed), and with the careful control of contamination before release of 

components to production was allowed, it is unlikely these radioelements would have been 

available for intake.
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6.0 ESTIMATION OF EXTERNAL RADIATION EXPOSURE

6.1 INTRODUCTION

Archived external dosimetry information for IAAP workers has been collected by University 

of Iowa College of Public Health researchers under funding provided by DOE to support

medical screening of former IAAP workers.1 Additional records were obtained from the 

Pantex site who currently is the custodian for most of historical IAAP records. This 

information has been examined as part of the effort to develop this technical basis document 

(TBD). Workers involved in nuclear weapon assembly activities for the AEC from about 

1949 through 1975 at the IAAP were associated with a facility known as Line 1 or Division B. 

The primary work activity involving external radiation exposure involved testing nuclear 

components using DU, handling sealed nuclear components called pits containing enriched 

uranium or plutonium (Brinck and Jacobson 1977), and industrial radiography operations. 

6.2 WORKPLACE EXTERNAL RADIATION FIELDS

The nuclides in the sealed nuclear weapon component pits emit beta, X- and gamma rays, 

and neutron radiation. However, radiation exposure to the workers depended significantly 

on processes used in the preparation, design and construction of the respective weapons. 

The main contribution to external radiation at IAAP was due to processes involving the 

handling or working around the nuclear components (pits) which contained plutonium and/or 

highly enriched uranium (HEU). In addition, radiography of explosive components at various 

stages of assembly also results in some potential for external exposure. Some early nuclear 

weapons may also have contained 210PoBe initiators with the capsule or what we would now 

call a pit, which is a solid, hermetically sealed object (DOE 1997). 

6.2.1 Photon Radiation

Photon (x-ray and gamma) radiation was associated with several IAAP work activities. 

Sources of external ionizing radiation at IAAP are dominated by fissile materials used in pits; 

however, there is also some low level exposure to low-activity radioactive sources, such as 

those used to check or calibrate radiation detectors, as well as analytical devices employing 

X-rays produced by a radiation generating device (RGD). These sources could have 

included alpha, beta, photon, and neutron emitters and were of the types and source

strengths typically used by mainstream industrial or process-related users. Although some

inventory records exist, a complete historical inventory of small sources has not been found 

in the archival material reviewed. Doses associated with the proper and widespread use of 

small check sources is generally negligible compared to the fissile materials. In addition to 

the small sources, there were at least two larger 60Co sources with original activities of 5 and 

50 curies, respectively (Shaykin 1969). These larger 60Co sources, as well as the RGDs, 

 

1

In 1993, Congress passed Public Law 102-484. Section 3162 of this law required DOE to screen for

occupational health conditions among former employees who might be at risk.
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had the potential for producing significant exposure to workers if not used properly. Gamma 

radiation of 2.2 MeV resulted from 1H (neutron, gamma) 2H interactions caused by neutron 

radiation scattering (i.e., moderation) and absorption in the hydrogen-rich materials in the 

nuclear components and building materials (concrete) (Shleien, Slayback, and Birky 1998).

Weapons assembly at IAAP was performed with sealed nuclear components of purified 

radioactive metals. The purification process separates natural progeny radionuclides from 

their parent radionuclide. This process provides some insight into potential sources of

radiation. Plutonium is purged of progeny radionuclides when it is purified. However, 241Am 

starts growing in as its parent radionuclide 241Pu decays with a half-life of 14.4 years. The 

241Am reaches a maximum activity after about 80 years, but it reaches about 85% of this 

maximum in 40 years. Figure 6.1 depicts the MCNP spectra from a generic pit of weapons 

grade plutonium aged 15 years (Traub et al 2005)). The details and specifications for the 

generic pit can be found in Appendix D of this report.

Photon Energy (keV)

250 500 750 1000 1250 1500 1750 2000

Dose Rate (mrem/hr)

0.00001

0.0001

0.001

0.01

0.1

1

10

100

1000

Photon Energy (keV)

0 50 100 150 200 250

Dose Rate (mrem/hr)

0.001

0.01

0.1

1

10

100

1000

Figure 6.1 MCNP calculated spectra from generic pit

(Bare unshielded mass of weapons grade plutonium)

The generic pit spectrum clearly indicates that the external dose contribution is

predominately from 30-250 keV photons, with approximately 70% of the total dose rate 

coming from the 60 keV photopeak from Am-241. This high magnitude is primarily due to 

the claimant favorable assumption of 15 year aged weapons grade plutonium. During 

assembly operations, the plutonium is not expected to be more than 3-4 years old; however, 

considering disassembly operations, an exposure to 15-year-old plutonium is credible. The 

average age of the weapon’s components at IAAP was less than 10 years. Considering that 

the plutonium was a few years old during assembly, an average of 15 years at disassembly 

is reasonable and claimant favorable.
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Although the age of the plutonium greatly affects the magnitude of the low energy photon 

dose, this component is also dependent upon the shielding or cladding materials

surrounding the fissile materials. The type of cladding material is particularly significant for 

lower energy photons (<70 keV). High-Z materials such as uranium greatly reduce the 

photon dose rate and almost completely shield the worker from low energy photon

emissions. Conversely, low-Z materials such as beryllium, do not provide much shielding 

and allow a relatively large quantity of lower energy photons to pass. Through discussions 

with DOE, NIOSH has learned that exact information on the cladding material and thickness 

for each weapon design assembled and disassembled at IAAP remains classified.

The initial IAAP site profile (ORAU, 2004a) assumed a high Z material of sufficient thickness 

such that the low energy photon dose was negligible. Through further research, NIOSH has 

learned that while this may be correct for some components, it is not globally accurate, and 

in some instances could result in a claimant unfavorable dose estimate. As a result of this 

finding, NIOSH has significantly revised this site profile to include a low energy photon dose 

for all workers handling pits.

It is important to note that not all components had a significant low energy photon dose. 

There are three basic types of pits used in assembly and disassembly at IAAP: 1) enriched 

uranium pits; 2) plutonium pits; and 3) composite pits (combination plutonium and enriched 

uranium). In the composite pits, the plutonium always had an outer shell of enriched 

uranium. Since the low energy photon dose from enriched uranium is negligible, only the 

plutonium pits had the potential for significant low energy photon dose.

However, due to security considerations, NIOSH cannot publicly identify which programs 

contained plutonium only pits versus those for which low energy photon dose was negligible. 

As a result, a claimant favorable approach is adopted in which NIOSH assumes, for low 

energy photons dose calculations only, that plutonium pits were exclusively handled and 

that none of the pits were clad. The low energy photon dose is, therefore, calculated based 

on the ratio between the low energy photons and higher energy photons.

This assumption maximizes the possible low energy photon dose component. If actual data 

were used in the dose reconstruction, the low energy photon dose would be reduced. As a 

result, does reconstructions using this methodology report doses that will exceed the actual 

dose that the workers received. While these assumptions will overestimate the external 

dose, they are considered reasonable and necessary in order to meet the intent of the 

Energy Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA), 

and still preserve and protect national security interests. (See Appendix E for additional 

discussion about pit cladding).

Using the generic pit described in appendix D, the photon dose breakdown for each IREP 

photon energy interval is provided in Table 6.1.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 37 of 79

Table 6.1 Energy Distribution from Generic Pit

Photon Energy Interval Dose Rate

(mrem/hr) Percentage of Dose

< 30 keV 1.18 4%

30-250 keV 27.76 83%

> 250 keV 4.34 13%

Total 33.28 mrem/hr 100%

6.2.2 Neutron Radiation

Uranium (234U, 235U, 236U, and 238U) and plutonium (primarily 239Pu, but also 240Pu and 242Pu) 

are alpha-emitting nuclides with the expectation of (alpha, neutron) interactions with light 

elements in addition to spontaneous fission. Only a limited number of dosimeter badges (27 

of 215 per exchange cycle) at IAAP contained NTA film for neutron monitoring (Davis,

1969). The significance of neutron radiation exposure to IAAP nuclear weapon component 

assembly workers cannot be directly assessed due to limitations in the monitoring

technology. Although Mr. Shannan considered the neutron dose to IAAP workers from pits 

to be very low (Fix and Bihl 2003), a review of the area badge (building) data indicates that a 

few areas had some significant neutron exposures. In these areas when the neutron dose 

was detectable, the photon doses were also fairly high, thus indicating a neutron to photon 

ratio is feasible and appropriate. A neutron to photon ratio was calculated using paired nonzero area monitoring data (Figure 6.2). To ensure that the ratio was not dependent on the 

magnitude of the photon exposure, a plot was developed of the measured neutron—photon 

ratio as a function of the measured photon exposure (Figure 6.2).

Gamma exposure (mR)

0 200 400 600 800 1000 1200 1400

Measured Neutron to Photon Ratio

0.01

0.1

1

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Measured Neutron - Photon Ratio

0.01

0.1

1

Coefficients:

GM = 0.135

GSD = 2.02

r 2 = 0.98

n = 129

Figure 6.2 Measured neutron to photon ratio in IAAP area badges and a function of gamma exposure 

and the resulting fitted lognormal distribution.

It is important to note that the measured neutron to photon ratio in Figure 6.2 is an 

underestimate of the true neutron to photon ratio. This underestimate results due to the 

inability of the NTA film to accurately measure neutrons below about 800 keV. This 

underestimate can be corrected using MCNP to calculate the portion of the neutron dose 

that would not have been detected by the NTA film. Although the neutron to photon ratio 

varies significantly depending on pit design, MCNP calculations indicate that no more than 

40% of the typical neutron spectra was below 800 keV. As a result, the measured ratio was 

adjusted in a claimant favorable nature (doubled) to correct this technological shortfall. 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 38 of 79

Thus for IAAP, a claimant favorable neutron to photon ratio of 0.27 with a GSD of 2.0 was 

calculated.

Although an evaluation of the NTA film was conducted and indications are that the neutron 

to photon ratio was typically less than 30%, The recommended approach to estimate 

potential neutron dose for IAAP monitored workers is to utilize the distribution of neutron-tophoton dose ratio from measured Pantex dosimeters during the period of 1993 to 2003. 

This data resulted in a higher neutron to photon ratio and is thus more claimant favorable. 

This higher ratio is mostly due to the practice of wearing lead aprons at Pantex. This 

practice results in a lower photon dose than would be experienced by Iowa workers;

however, the neutron dose is not greatly affected by the lead apron. Since the ratio is 

developed by dividing the neutron dose by the photon dose, a low biased photon dose 

would result in a higher neutron to photon dose ratio and is therefore more claimant

favorable. Although it is usually better to use site specific information over surrogate facility 

data, the uncertainty in neutron monitoring and calculated correction factors justify using the 

more accurate measurements from surrogate source term data.

It should also be noted, however, that a detectable neutron dose was not always found with 

high photon doses. The presence and magnitude of the neutron dose associated with 

nuclear weapons components greatly depends on the weapon design. As a result,

estimating the neutron to photon ratio based solely on corrected paired (neutron-photon) 

measurements will overestimate the total neutron dose for the work force when applied to all 

photon doses.

6.2.3 Depleted Uranium

Another component of photon dose comes from depleted uranium. The important progeny 

nuclide for this external exposure is the decay of 238U decay to 234mPa with a half-life of 24 

days. Thus in a matter of a few months, DU components have 234mPa activities nearly equal 

to that of 238U. The radionuclide 234mPa emits beta radiation 98.6% of the time when it 

transitions to its ground state with a maximum energy of 2.28 MeV and an average energy 

of 0.825 MeV (Shleien, Slayback and Birky 1998). While this beta particle does not result in 

a significant deep dose to most organs, the bremsstrahlung radiation produced can be

significant for some workers who did not handle pits, but worked with depleted uranium. 

MCNP was used to model spectral characteristics of bremsstrahlung photons from 1-cm and 

30-cm diameter 238U spheres. The results were similar for both spheres. Figure 6.3 shows 

the MCNP calculated photon spectrum emitted from 238U as excited by the 234mPa beta 

spectrum shown on a logarithmic vertical axis. Note the smooth bremsstrahlung spectrum 

and the uranium characteristic K X-rays at 90-109 keV and the L X-rays in the range of 13-

19 keV. As can be observed from the figure, the dose contribution from low energy photons 

(<30 keV) is relatively insignificant.
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1E-8

1E-7

1E-6

1E-5

1E-4

1E-3

0 0.5 1 1.5 2 2.5

Energy (MeV)

MeV/pe

Figure 6.3 MCNP calculated photon spectra emitted from 234mPa beta in 238U spheres.

As noted in section 5, workers handled DU during disassembly of bomb components

containing DU, during and following hydrotesting, and during machining. DU fragments 

were collected by workers from the test area after each non-nuclear detonation (Archive 

010000914 “BAECP Former Worker Program Needs Assessment”). The DU could

contribute a significant extremity and skin dose to workers unless precautions were taken to 

protect workers from the beta radiation. A bare slab source of DU contributes an Hp(0.07) 

dose of approximately 230 mrad/h compared to an Hp(10) dose of approximately 2 mrad/h 

(ORAU 2003b).

6.3 EXTERNAL RADIATION DOSE RECORDS

A review of IAAP dosimetry data was conducted and only intermittent dosimetry data has 

been located between 1955 through 1962. The majority of the dosimetry data is from R. S. 

Landauer Company (1955-1974), and only a few results from TracerLab have been located 

to date. Table 6.2 indicates the time periods where dosimetry records have been located. 

As indicated in the table, starting in 1963 the full 52 weeks/year of monitoring data is 

available. Since prior to 1963, only partial year data is available; the dose reconstructor 

should not assume the absence of dosimetry data indicates an unmonitored worker. During 

these time periods, the dose should be assigned either based on an individual worker’s 

available data multiplied by the appropriate number of weeks exposed or based on the 

methodology described in section 6.5.

Based on these records and the number of dosimeters issued, the monitoring practices at 

IAAP can be subdivided into four monitoring eras. These eras are described in Table 6.3.

Additional dosimetry records indicate that the IAAP Rad-Safe department routinely assigned 

and evaluated pocket ionization chamber (PIC) measured doses from at least June 18, 

1965, through November 7, 1974. The content of these records is consistent with

information from Shannan that monitoring was done sporadically depending on need (Fix 

and Bihl 2003). Most of the pocket dosimeter results appear to be related to using high level 

radioactive source for radiography.
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Table 6.2 Summary of Dosimetry Data Available

Year

Exchange

Frequency

# Weeks

of Data First Badge Date

End Date for last

cycle

Minimum

Reported 

Dose

1955 Weekly 7 November 14, 1955 January 1, 1956 20

1956 Weekly 6 January 2, 1956 February 13, 1956 20

1957 Weekly 13 May 6, 1957 August 4, 1957 10a

1958 Weekly 14 July 28, 1958 November 2, 1958 5

a

1959 Weekly 14 June 29, 1958 October 4 1958 5

a

1960 Weekly 15 March 7, 1960 June 19, 1960 5

a

1961

1962 Weekly 12 September 3, 1962 December 3, 1962 10

1963 Bi-Weekly 52 January 7, 1963 January 5, 1964 10

1964 4 Week 52 January 6, 1964 January 3, 1965 10

1965 4 Week 52 January 4, 1965 January 2, 1966 10

1966 4 Week 52 January 3, 1966 January 1, 1967 10

1967 4 Week 52 January 2, 1967 December 31, 1967 10

1968 4 Week 52 January 1, 1968 December 29, 1968 10

1969 4 Week 52 December 30, 1968 December 28, 1969 10

1970 4 Week 52 December 29, 1969 December 27, 1970 10

1971 4 Week 56 December 28, 1970 January 23, 1972 10

1972 4 Week 52 January 24, 1972 January 21, 1973 10

1973 4 Week 52 January 22, 1973 January 20, 1974 10

1974 4 Week 52 January 21, 1974 January 19, 1975 10

aGiven the dosimeter designs of this time period, it is highly unlikely that the film dosimeter could measure

this level. For this early time period, a detection level of 20 mR is more credible.

Table 6.3 Personnel Dosimeter Monitoring Eras

Era Time Period Description Dose Reconstruction Method

1 1949-1954 No personnel monitoring 

conducted. Exposure Matrix (Section 6.5)

2 1955-1962

Limited personnel monitoring data 

available; however, time periods 

within the year are intermittent.

Exposure Matrix (Section 6.5)

3 1963-1967

Routine personnel monitoring and

continuous monitoring records are 

available.

Annual Dose Distributions or

personal monitoring data, 

whichever is greatest.

4 1968-1974

Extensive personnel monitoring, 

most radiological workers should 

have some dosimetry records.

Annual Dose Distributions or

personal monitoring data,

whichever is greatest.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 41 of 79

6.3.1 IAAP Annual Dose Distributions

The IAAP annual dose distributions were developed by evaluating individual dosimeter

results from 1955 through 1974. This analysis also considered missed dose that was not 

reported on the dosimetry reports. Typically there are two scenarios that result in a zero 

dosimeter reading. The first considers that a worker was exposed to low level radiation that 

resulted in a dose that was below the limit of detection or an administrative reporting limit. 

The second is that a worker wore a dosimeter badge but was not exposed to radiation.

During worker outreach meetings in July 2004, a new issue with missed dose was identified. 

The issue concerned the radiological monitoring practices at the site. Through discussions 

with former IAAP workers who conducted both assembly and disassembly, NIOSH

discovered that film badge dosimeters may not have been worn all of the time. One worker 

indicated that he always wore his film badge, while another indicated that he would only 

wear it when one was given to him. Through this discussion, it became apparent that in 

general workers were supposed to wear their film badges, but strict adherence was not 

necessarily enforced. As a result, a third scenario occurred in which a worker was issued a 

dosimeter badge but did not wear it during exposure to radiation. When this badge would 

be processed, this could also result in a zero reading. The effect of these three scenarios is 

that there is likely some missed or unrecorded dose (Figure 6.4). As shown in Figure 6.4, in 

1966 approximately half of the dosimeter readings were below the detectable or reporting 

level.

1966 - All Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Readings (mR)

1

10

100

1000

Coefficients:

GM = 8.9 mR

GSD = 2.33

r 2 = 0.88

n = 391

 

1966 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 19.0 mR

GSD = 1.62

r 2 = 0.86

n = 172

Figure 6.4 Comparison of analysis using all dosimeter data versus including only non-zero data.

In Rev 0 of this site profile (ORAU, 2004a), only the first two scenarios were considered in 

the development of Table 7. The same data were analyzed using standard regression 

methodology to account for readings that were below the limit of detection or reporting 

criteria. The effect of the third scenario is that too many zero readings are considered in the 

development of the initial annual dose distributions. To assess the true dose, those badges 

that were issued but not worn must be removed from the dataset. When these zero

measurements are removed, the regression line shifts to the left, thus the geometric mean 

of the distribution is effectively increased. NIOSH considered estimating a percentage of the 

zero readings that resulted from workers not wearing their badges; however, any estimate of 

the percentage would be somewhat arbitrary and subjective. Since it is not possible to 

either remove the erroneous zero data or to estimate the percentage of zero readings at this 

time, NIOSH reanalyzed the entire dataset discarding all zero readings. This analysis

assumes that all of the non-zero dosimeter readings are a representative sample of all the 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 42 of 79

dosimeters in a given year. This type of analysis introduces significant bias in that it is 

known that some radiation exposures are below a limit of detection and that in some

instance workers would have worn their dosimeters during non-radiological activities. The 

true minimum dose is not 10 mR per cycle, but zero for workers who were not exposed 

during the monitoring period.

In summary, the annual dose distributions provided in Table 6.4 were developed by fitting a 

lognormal distribution to all of the non-zero dosimetry data for a given year. This analysis 

overestimates the true dose; however, given the data limitations, is considered necessary in 

order to give the benefit of the uncertainty to the claimant. The geometric mean and 

geometric standard deviations listed in Table 6.4 were calculated based on the fitted 

distributions depicted in the figures of Appendix F. A summary plot of the annual dosimetry 

data is depicted in Figure 6.5.

Table 6.4 Summary of Annual Dose Distributions

Year

Total # of

Dosimeter

Readings

# Dosimeter

Readings > 10 mR

used in the analysis

Geometric

Mean (mR)

Geometric

Standard

Deviation (GSD)

Upper 95th

Percentile

(mR)

1955 64 7 1664 1.26 2434

1956 48 10 1347 1.30 2074

1957 104 33 910 1.36 1509

1958 275 99 988 1.85 2718

1959 182 59 520 1.54 1058

1960 311 92 582 1.52 1160

1961

1962 450 33 619 1.32 975

1963 783 295 468 1.67 1088

1964 693 156 176 1.44 320

1965 334 133 226 1.70 541

1966 391 172 247 1.62 546

1967 444 140 261 1.78 675

1968 900 222 359 2.03 1150

1969 2107 292 296 1.88 837

1970 2682 848 514 2.93 3009

1971 3890 1527 586 2.75 3094

1972 4236 1086 400 2.47 1772

1973 3108 627 553 2.36 2268

1974 1490 342 527 2.40 2222

Total 22492 6173
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Year

1950 1955 1960 1965 1970 1975

Annual Dose (mR)

0

500

1000

1500

2000

2500

3000

3500

4000

No Monitoring

Data

Intermittent 

Monitoring Data

Continuous Monitoring Data

Figure 6.5 Annual dose distributions with the associated uncertainty. 

The error bars are not one standard deviation, but represent the range 

from the 5th percentile to the 95th percentile of the log-normal distribution. 

6.4 DOSIMETER TECHNOLOGY AND PERFORMANCE STUDIES

6.4.1 Photon Radiation

Specific designs of the Landauer film dosimeters used at IAAP have not been located, and 

no records of the mentioned earlier use of TracerLab dosimeters have been found. 

However, from the content of IAAP-submitted AEC termination reports and personal

testimony (Fix and Bihl 2003), it is likely that the film dosimeter was, at least, a two-region 

design (i.e. non-penetrating dose calculated from film response to open window or generally 

unfiltered region of the film and penetrating dose calculated from film response under a

selected, usually metallic, filter). Table 6.2 and 6.3 have summarized the monitoring

technique and exchange frequency for the IAAP dosimeters contracted from a commercial 

service and the minimum reported dose. An important addition to the exchange frequencies 

noted in Table 6.2 is that some of the more highly exposed workers continued on a biweekly exchange frequency through 1974. As a result, the dose reconstructors should 

consider this additional frequency when calculating missed dose. A quick review of the 

individual’s dosimetry records identify the higher exchange frequency.

The AEC conducted performance testing of several commercial and in-house film dosimeter 

services during 1954 with exposures provided by the National Bureau of Standards (AEC 

1955). Specific dosimeter design specifications are included in the documentation. The 

testing included 40-, 70-, and 210-keV narrow spectral beam X-ray techniques, 60Co gamma 

radiation, and selected mixtures of these beams. Measured response data are provided in 

the report for each of the respective dosimeter open-window and filtered regions of the film. 

This information exhibits the significant overresponse of the open-window and lightly filtered 

regions of the film at lower (i.e. 40 and 70 keV) photon energies and an underresponse of 

the heavily filtered portions of the film dosimeters to photon energies less than 70 keV.
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There is considerable uncertainty as to whether the IAAP dosimeter badge could reliably 

measure a penetrating dose from the 60 keV photons from Am-241. As illustrated in figure 

6.1, the Am-241 60 keV photopeak contributes approximately 70% of the total photon dose. 

According to the R. S. Landauer Company, the IAAP film badge had a lower energy 

response threshold of approximately 30 keV (Landauer, 1965). However, dosimetry

measurements conducted at Hanford (Larson and Roesch, 1955) using a two element film 

badge similar to the early IAAP dosimeter, photons less than 70 keV were severely 

attenuated. Based on the Hanford report, at 60 keV, only 37% of the penetrating dose 

would be measured. At IAAP, several film badges were exposed to low energy photons at 

the medical facility. In an X-ray exposure at 70 kVp read less than half a similar exposure to 

95 kVp X-rays although the intensity of the beam was only decreased by 67%. 

Furthermore, the exposure at 45 kVp was not detectable. Since the level of filtration greatly 

affects the average photon energy, exact energy threshold information cannot be estimated. 

If the beam was relatively heavily filtered, the kVp would correlate to near the peak energy; 

however, with little filtration, the 95 kVp would result in an average photon energy of 

approximately 32 keV.

Based on the Hanford information and professional judgment, a claimant favorable fraction 

of 30% of the Am-241 photopeak is assumed to have been measured by the IAAP film 

badge. This value is likely to be slightly lower than reality, and since multiple filters were 

used throughout the DOE complex from the early 1960s, this is likely a considerable 

overestimate for the later years (1970s).

6.4.1.1 Photon Radiation Performance Studies

Within the IAAP dosimetry records, NIOSH has found that the IAAP Rad-Safe group 

intermittently exposed spare badges to known quantities of radiation and submitted them 

with other dosimeters as a quality assurance check. NIOSH’s review of this information 

indicates a slight overall bias (-7%) and a standard deviation of ±20%, with an increase to -

60% and +40% observed for two of the three low level exposures at 50mR. This greater 

uncertainty is common with film dosimeter as the exposure approaches the limit of detection. 

This effect is discussed in NIOSH’s External Dose Reconstruction Implementation Guideline 

(OCAS-IG-001, 2002) which provides guidance on how to incorporate this uncertainty in 

dose reconstructions. Figure 6.6 illustrates the reasonable agreement between the

irradiations and the reported readings.
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Exposure (mR)

0 500 1000 1500 2000 2500

Dosimeter Reading (mR)

0

500

1000

1500

2000

2500

Figure 6.6 Comparison of film dosimeter badges exposed to a known quantity

of radiation and the reported readings by the commercial vendor.

6.4.2 Neutron Radiation

Some neutron doses were reported by Tracerlab and Landauer on dose reports beginning 

about 1962. Eastman Kodak nuclear emulsion type A (NTA) film was used for these

measurements. NTA was basically the only common dosimeter method available to 

measure neutron dose in AEC facilities at that time. Results reported at the first AEC 

Neutron Dosimetry Workshop in 1969 indicated that SRS calibration laboratory dose

measurements made with NTA film were about one-half to one-fourth of those measured 

with other methods, including the neutron thermoluminescent dosimeter (TLD)(Vallario, 

Hankins, and Unruh 1969). This underestimation is due to the NTA film’s inability to 

accurately measure lower energy neutrons (<800 keV). As a result, IAAP’s neutron

dosimeters underestimated the total neutron dose.

One of the most important parameters related to performance of NTA is the difference 

between calibration and workplace neutron energy spectra. There are no known

measurements of neutron spectra at IAAP, and the method(s) used to calibrate the

Landauer NTA film known is currently unknown. Neutron dose calculations have been 

conducted using MCNP for various IAAP pits in a configuration that would incorporate 

albedo effects of surrounding materials to increase the neutron scatter. As noted in section 

6.2, these MCNP calculations indicate that no more than 40% of the neutron dose was less 

than 800 keV, thus indicating that IAAP’s neutron readings would be of similar quality to 

those reported by other DOE facilities such as the Savannah River Site.

Although neutron spectra were not measured at IAAP, neutron spectra measurements have 

been conducted at various Pantex Plant facilities (classified), and the performance of the 

Pantex 809/812 thermoluminescent dosimeter used beginning in 1993 has been validated 

for weapon and workplace exposures. As a result, this data is considered to be of higher 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 46 of 79

quality than is calculated from MCNP. Significant neutron exposure from nuclear weapons 

components at IAAP was typically associated with a photon dose that would be readily and 

reliably measured with film dosimeters.

6.5 IAAP WORKER EXTERNAL DOSE RECONSTRUCTION

The primary objective of dose reconstruction for IAAP workers is to utilize a claimantfavorable method to retrospectively:

• Calculate the dose to unmonitored or intermittently monitored workers prior to the 

routine use of personnel dosimeters.

• Interpret the measured dose for monitored workers for input to IREP for probability of 

causation calculations.

• Calculate the unmeasured or poorly measured neutron dose to both unmonitored 

and monitored workers.

6.5.1 Unmonitored or Intermittently Monitored Workers

At IAAP, there are basically three categories of unmonitored or intermittently monitored 

workers; 1) unexposed workers who did not need to be monitored, 2) intermittently or 

routinely exposed workers who did not directly handle radioactive materials, and 3) workers 

who directly handled radioactive materials.

Unexposed workers who did not need to be monitored

This category includes office and administrative personnel who may have been occasionally 

exposed to radioactive materials as they walked throughout the plant. Since their main 

responsibilities would put them in an office setting, assigning onsite ambient dose levels 

based on area monitoring data located throughout the plant in production areas would

constitute a reasonable upper bound. The dose estimate for these workers should be based 

on onsite ambient dose levels as described in section 4.2.

Intermittently or routinely exposed workers who did not directly handle radioactive materials

These workers would likely have been monitored by modern standards but may not have 

been due to higher regulatory dose limits or past radiological control practices. Most IAAP 

unmonitored workers fall into this category. This category includes workers who did not 

directly handle radioactive materials but may have conducted some work in proximity to 

sources of radiation. Their work around radioactive materials may have been intermittent or 

routine. This category generally includes trade workers, and engineers whose time was 

spent between plant operations and an office setting. The dose estimate for these workers 

should be based on the exposure distributions in Table 6.4 and Table 6.6.

Workers who directly handled radioactive materials

In general, since 1963, workers who directly handled radioactive materials at IAAP are 

expected to have been routinely monitored and dosimetry data should be available. 

However, some may have only been partially monitored. This category included workers 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 47 of 79

involved directly in the assembly, disassembly, inspection, or transport of radioactive 

materials.

6.5.1.1 Dose estimation prior to 1963

Prior to 1963, only a relatively small number of workers were monitored for exposure to 

external radiation and only intermittent monitoring data is available. As a result, the external 

dose is estimated based on a ratio methodology using dose rate information and annual 

doses. Since the magnitude of the radiation dose could have changed over time due to the 

handling of various pits, a generic or bare unshielded pit2 is assumed in order to estimate a 

claimant favorable annual dose prior to 1963. Details and specifications of the generic pit 

can be found in Appendix D. Based on these specifications, the dose rate for the generic pit 

was calculated using MCNP (Traub et al, 2005). With this dose rate, the only other factor 

necessary to estimate annual dose is the relative work factor or number of hours in which a 

worker would have handled the pit or been in close proximity to the pit. To estimate the 

relative time spent in proximity to a pit (work factor), dosimetry data post 1962 was

compared with an era dose rate. The methodology and assumptions used to develop the 

era dose rates remain classified as described in Traub et al (2005). Using the era dose 

rates of 1.08 mrem/hr and 1.48 mrem/hr for era #3 and #4 respectively, the work factor was 

calculated as follows:

D x 2000 hours

D x C

W

ERA

Annual f

f &

= Equation 6.1

Where: Wf = Work Factor

DAnnual = Annual Dose (Lognormal Distribution)

DERA = Era Dose Rate

Cf = Conversion factor Roentgen to rem (1.05)

The correction factor was obtained by integrating the area under the curve from 70 keV to 2 

MeV which represents a combination of the measured dose (energy threshold) and that 

most of the external dose photon spectra is less than 2 MeV.

 

2 The terminology for pit is used as a general term to indicate the fissile materials which are the dominant contributor to 

the external dose.
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Table 6.5 Work factor calculation using the two era dose rates.

Year Annual

Dose

Modified

for Hp(10)

Era Dose

Rate

Estimated

Annual

Dose

Work

Factor

Mean Work

Factor

1962 619 650 1.08 2160 0.301

1963 468 491 1.08 2160 0.228

1964 176 185 1.08 2160 0.086 ERA #3

1965 226 237 1.08 2160 0.110

1966 247 259 1.08 2160 0.120

1967 261 274 1.08 2160 0.127 0.162

1968 354 372 1.48 2960 0.126

1969 296 311 1.48 2960 0.105

1970 514 540 1.48 2960 0.182 ERA #4

1971 586 615 1.48 2960 0.208

1972 400 420 1.48 2960 0.142

1973 553 581 1.48 2960 0.196

1974 527 553 1.48 2960 0.187 0.164

 Geometric Mean 0.153 

 Geometric Standard Deviation 1.43

The average annual dose changed on an annual basis. This introduces uncertainty into the 

calculation. As a result, the work factor is treated as a distribution. Figure 6.7 depicts the 

work factor distribution, which has geometric mean of 0.153 and a geometric standard

deviation of 1.43.

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Work Factor - 

Wf

0.01

0.1

1

Coefficients:

GM = 0.153

GSD = 1.43

r 2 = 0.97

n = 13

Figure 6.7 Work Factor Probability Distribution

The annual dose is then calculated by multiplying the dose rate from the generic pit by a 

2000 hr. work year and the work factor (Equation 6.2). Since in the equation the generic pit 

is treated as a constant, and the work year is also a constant, the uncertainty is simply the 

geometric standard deviation of the work factor.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 49 of 79

DAnnual = D&Generic x 2000 hours x Wf x RAm-241 Equation 6.2

Where DAnnual = Estimated annual dose (mrem)

DGeneric = Estimated Measured dose equivalent from the generic pit

Wf = Work Factor

RAm-241 = Ratio of Am-241 in growth

Equation 6.2 is used to reconstruct doses prior to 1963 and the annual dose information 

from Table 6.4 is used to reconstruct doses from 1963 through 1974. Using these two 

methodologies, Table 6.6 and Figure 6.8 provides the total annual dose distribution

parameters for IAAP workers from 1949 through 1974.

Table 6.6 Modeled Annual Dose Using Work Factor Methodology

Annual

Dosimeter

Data ERA Year 1

Generic Pit

Dose Rate

(DGeneric)

Work

Factor

(Wf)

Am241

Ingrowth

(RAm-241)

Modeled

Annual

Dose GSD

Median GSD

1949 14.8 0.153 0.67 3034 1.43

1950 14.8 0.153 0.71 3215 1.43

1951 14.8 0.153 0.75 3397 1.43

1952 14.8 0.153 0.78 3532 1.43

1953 14.8 0.153 0.81 3668 1.43

#1

1954 14.8 0.153 0.84 3804 1.43

1955 14.8 0.153 0.87 3940 1.43 1664 1.26

1956 14.8 0.153 0.90 4076 1.43 1347 1.30

1957 14.8 0.153 0.93 4212 1.43 910 1.36

1958 14.8 0.153 0.95 4302 1.43 988 1.85

1959 14.8 0.153 0.98 4438 1.43 520 1.54

1960 14.8 0.153 1.00 4529 1.43 582 1.52

1961 14.8 0.153 1.00 4529 1.43

#2

1962 14.8 0.153 1.00 4529 1.43 619 1.32

1963 468 1.67

1964 176 1.44

1965 226 1.70

1966 247 1.62

#3

1967 261 1.78

1968 354 2.03

1969 296 1.88

1970 514 2.93

1971 586 2.75

1972 400 2.47

1973 553 2.36

#4

1974 527 2.40

a. The data has been converted to mrem by multiplying the dosimeter dose by a correction factor of 1.05

Table 6.6 and Figure 6.8 indicate a continuous increase of the annual dose between 1949 

and then constant for 1960 through 1962. This ramp effect is due to the estimate in growth 

of Am-241 from the decay of Pu-241 in weapons grade plutonium. The generic pit assumes 

15 year aged plutonium, whereas in 1949, the maximum age plutonium in the United States 

could have been 4 years old. By 1960, however, there could have been 15 year age 

plutonium in the stockpile. 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 50 of 79

Year

1950 1955 1960 1965 1970 1975

Annual Dose (mrem)

0

1000

2000

3000

4000

5000

6000

7000

8000

9000

10000

Dosimetry Data

Modeled Dose

No Monitoring

Data

Intermittent 

Monitoring Data Continuous Monitoring Data

Figure 6.8 Summary of annual dose distributions using the model between 

1949 -1962 and actual dosimetry data from 1963 to 1974

Table 6.7 provides the change in dose rate of the generic pit from Am-241 in-growth. It is 

also important to note that the main mission of IAAP and Pantex was to assemble nuclear 

weapons. The Clarksville and Medina facilities were primarily responsible for retrofit and 

disassembly during this early time period (Mitchell, 2003). As a result, the ramp effect and 

assumption of 15 years is considered claimant favorable.

Table 6.7 Am-241 Build-up from decay of Pu-241

Photon Dose Rate (mrem/hr)

30-250 keV Decay 

Year Year

< 30 keV Pu 

Component

Am 

Component Total

> 250 keV

Simulated 

Measureda

>70 keV + 

0.3*Am 

Ratio of 

Am241 ingrowth

1 1.18 3.14 2.28 5.42 4.49 8.1 0.55

2 1.18 3.14 4.46 7.60 4.49 8.8 0.59

3 1.18 3.14 6.52 9.66 4.49 9.4 0.63

4 1949 1.18 3.14 8.49 11.63 4.49 10.0 0.67

5 1950 1.18 3.14 10.36 13.50 4.49 10.5 0.71

6 1951 1.18 3.14 12.14 15.28 4.49 11.1 0.75

7 1952 1.18 3.14 13.84 16.98 4.49 11.6 0.78

8 1953 1.18 3.14 15.44 18.58 4.49 12.1 0.81

9 1954 1.18 3.14 16.97 20.11 4.49 12.5 0.84

10 1955 1.18 3.14 18.43 21.57 4.49 13.0 0.87

11 1956 1.18 3.14 19.81 22.95 4.49 13.4 0.90

12 1957 1.18 3.14 21.12 24.26 4.49 13.8 0.93

13 1958 1.18 3.14 22.37 25.51 4.49 14.2 0.95

14 1959 1.18 3.14 23.56 26.70 4.49 14.5 0.98

15 1960 1.18 3.14 24.68 27.82 4.49 14.8 1.00

a. Simulated Measured Dose Rate is the C7 value as indicated in Appendix D. Also note values may 

not sum due to rounding.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 51 of 79

In reality, between the years from 1949 to 1961, there would also be some additional 

variance if the actual pit data could be presented. Since the generic pit was used, the data 

in Table 6.6 should be considered a reasonable upper bound of what the measured dose 

would have been if pits were onsite, and they were assembled into the weapons.

6.5.1.2 Dose estimation between 1963-1974

Starting in late 1962, a moderate number of workers were monitored for radiation exposure 

at IAAP. As a result, the annual dosimeter data in Table 6.6 can be used as a reasonable 

surrogate of co-worker data for unmonitored or intermittently monitored workers. For partial 

year exposures, the dose should be prorated by dividing the annual dose by 52 weeks to 

obtain a weekly rate and then multiplied by the number of weeks in the partial year.

6.5.2 Monitored Workers

The IAAP reported dose for monitored workers should be adjusted for any missed photon 

dose in accordance with the NIOSH External Dose Reconstruction Implementation

Guidelines (NIOSH 2002). This adjusted dose should then be compared against the annual 

dosimeter doses in Table 6.6. Since the worker may not have always worn their dosimeter, 

the higher of the two values should be assigned unless there is an indication in the 

dosimetry records which would warrant a lower dose assignment.

6.5.3 Adjustments to Photon Dose

The first correction is to adjust the measured or estimated doses by the Roentgen to rem 

conversion factor of 1.05. This factor was derived by integrating the area under the curve 

from 70 keV to 2 MeV which represents a combination of the measured dose (energy 

threshold) and that most of the external dose photon spectra is less than 2 MeV.

As noted earlier in this report, the film badge dosimeter used at IAAP may or may not have 

been capable of measuring all of the 60 keV photons from Am-241. The claimant favorable 

assumption is made that the dosimeter could only measure about 30% of this photon. As a 

result, some adjustment to the measured photon dose is necessary for IREP inputs. Using 

the generic pit (i.e. no cladding and minimal attenuation), dosimeter adjustment factors were 

developed using MCNP to translate the measured annual doses (> 70 keV) into IREP

energy intervals. Table 6.8 provides the claimant favorable adjustment factors. These 

factors incorporate both the ratio to account for the lower (< 70 keV) photons that were not 

measured as well as the energy fractions from Table 6.1. The ratios were developed using 

the dose rate data found in Table D.1 of Appendix D and by adding 30% of the Am-241 60 

keV photopeak into the measured dose (P4). To obtain the IREP input, multiply the values 

in Table 6.8 by the measured or estimated annual dose after conversion to Hp(10).

Table 6.8 IAAP Photon Adjustment Factors.

 Energy

 Interval

 Adjustment

 Factor

Low Energy Photon Dose < 30 keV 0.080

Intermediate Energy Photon Dose 30-250 keV 1.878

High Energy Photon Dose > 250 keV 0.303
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 52 of 79

6.5.3.1 Bladder Cancer Example

In order to summarize the ratios and adjustments, a bladder cancer example is provided in 

Appendix G.

6.5.4 Estimation of Neutron Dose

The recommended approach to estimate potential neutron dose for IAAP monitored workers 

is to utilize the distribution of neutron-to-photon dose ratio calculated from Pantex 

dosimeters during the period of 1993 through 2003. These measurements were made with 

the performance validated Pantex 809-812 dosimetry system. These dosimeters effectively 

evaluate the neutron to photon doses from a given source term. Dosimetry records were 

analyzed for each dosimeter with a positive neutron and photon dose greater than 50 mrem 

for the period of 809/812 use. Analysis of this information is shown in Figure 6.9 as a 

lognormal probability plot of the ratio of neutron-to-photon doses. A regression analysis 

yields a geometric mean of 0.79 and a geometric standard deviation (GSD) of 1.57. The 

measured Pantex neutron to photon ratio from 1993 through 2003 is expected to be greater 

than the actual IAAP neutron to photon ratio due to the use of lead aprons at Pantex which 

reduce the total photon dose and therefore increase the overall neutron to photon ratio. As 

a result, assigning the distribution of modeled and measured dose from Table 6.6 to IAAP 

workers prior to the routine use of personnel dosimeters and using the distribution of the 

neutron-to-photon dose ratio based on the Pantex dosimeter measurements is claimantfavorable. 

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Neutron to Photon Ratio

0.1

1

10

Coefficients:

GM = 0.79

GSD = 1.57

r 2 = 0.91

n = 43

Figure 6.9 Probability plot of neutron to photon ratios calculated from Pantex 

809/812 neutron and photon dosimeter data.

The respective statistical parameters for the lognormal distribution of neutron-to-photon 

dose ratios for application to dose reconstruction for IAAP monitored workers are presented 

in Table 6.9.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 53 of 79

Table 6.9 Statistical Parameters for IAAP neutron-to-photon dose ratios

Parameter Neutron to Photon

Dose Ratio

Geometric Mean 0.79

Geometric Standard Deviation 1.57

Upper 95% percentile 1.66

The estimate of the neutron dose must be adjusted to include the conversion to the ICRP 

Publication 60 (1990) 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 6.10 (ICRP 1990). Fission spectrum neutrons were selected for 

several reasons. As noted in Table D.1 of Appendix D, the neutron energy distribution for 

the generic pit was relatively evenly split between the fission and fast neutrons. The generic 

pit, however, does not have any cladding material surrounding it to slow the neutron spectra 

like a real pit would have. In order to ensure claimant favorability, 100% of fission spectra 

neutrons is assumed.

Table 6.10. IAAP neutron dose fractions and associated ICRP 60 correction factors.

Process Description/

buildings

Neutron

Energy (MeV)

Default dose

Fraction (%)

ICRP 60 correction

Factor (CF)

Nuclear 

weapon

component

assembly

Neutron exposure

associated with weapon 

assembly and

disassembly activities.

0.1 – 2 MeV 100 1.91

6.5.5 Skin and Extremity Dose (Reserved)

This section is currently reserved and will be updated in a revision to the TBD. The major 

difficulties with estimating the skin dose at this point evolve around beta exposures to the 

extremities (hands and forearms). The ORAU team is currently working on this issue as part 

of the skin dose Technical Information Bulletin (TIB). 

Although this section is reserved, this reservation does not pre-empt dose reconstruction for 

likely compensable cases. The dose reconstructor should not that the deep photon and 

neutron dose may be sufficient for the dose reconstruction to be completed. 

6.6 UNCERTAINTY IN PHOTON AND NEUTRON DOSE

For the usual analysis of measured film badge doses, the minimum detection levels (MDLs) 

quoted in the literature range from about 10 to 50 mrem for beta/photon irradiation; it is 

possible to read a photon dose of 100 mrem to within ±15 mrem if the exposure involved 

photons with energies between several hundred keV and several MeV (Morgan 1961). For 

dose reconstruction, the uncertainty in photon doses is estimated to be ±20%. This value is 

based on the IAAP Quality Assurance checks that were conducted periodically by IAAP’s

Rad-Safe Department. The uncertainty in the neutron dose is incorporated in a claimant 

favorable manner in the neutron to photon ratio.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 54 of 79

6.7 ORGAN DOSE

Once the photon and neutron doses and their associated standard errors have been

calculated for each year, the values are used to calculate organ doses of interest using 

NIOSH (2002). There are many complexities and uncertainties when applying organ dose 

conversion factors to adjusted doses of record. Many of the factors that affect the dose of 

record have been discussed in tables in this TBD. ICRU (1988) indicated that film badge 

dosimeters, while not tissue-equivalent, can be used for personnel dosimetry. It also

indicated that it is more difficult to ensure that the variation in response with energy and

angle of incidence with low energy. Considering the IAAP film badge response and the

typical photon energy spectra, the measured dose can be reasonably converted to Hp(10) 

using a correction factor of 1.05. Since the generic pit calculations were conducted using 

ICRP 74 (1996) conversion factors from kerma to Hp(10), Hp(10) is the recommended organ 

dose conversion factors for IAAP dose reconstructions.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 55 of 79

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Ahlstrand, Franz A., (1956a) “Project History of Line 1 Operations at Iowa Ordnance Plant, 

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Company, Inc., Burlington, Iowa.

Ahlstrand, Franz A., (1956b) “Project History of Line 1 Operations at Iowa Ordnance Plant, 

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Company, Inc., Burlington, Iowa.

Ahlstrand, Franz A., (1957b) “Project History of Line 1 Operations at Iowa Ordnance Plant, 

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GLOSSARY

Atomic Energy Commission

Original agency established for nuclear weapons and power production; a 

predecessor to the U. S. Department of Energy.

baratol

A castable mixture of explosives used in nuclear weapons.

beta (ß) dose

A designation (i.e.beta) on some Pantex external dose records referring to the dose 

from less-energetic 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.

cladding

Material surrounding or encasing another material. In this TBD, cladding generally 

refers to the material or materials surrounding the fissile material of the pit. 

curie

A special unit of activity. One curie (1 Ci) exactly equals 3.7 x 1010 nuclear 

transitions per second.

deep absorbed dose (Dd)

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

deep dose equivalent (Hd)

The dose equivalent at the respective depth of 1.0 cm in tissue.

detection limit (lower)

The minimum quantifiable exposure or neutron flux that can be detected.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying 

factors. The special unit is the rem. When D is expressed in Gy, H is in sieverts (Sv).

(1 Sv = 100 rem).

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 radiationabsorbing element (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).
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 61 of 79

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.

Depleted Uranium (DU)

Depleted uranium; 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.

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 (R) 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, intensity and energies present in the 

work environment.

film

Generally means a “film packet” that contains one or more pieces of film in a lighttight wrapping. The film when developed has an image caused by radiation that can 

be measured using an optical densitometer. (See Dupont 552, Dupont 558, 

Eastman Kodak, Nuclear Emulsions).

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 differences is that X-rays do not originate in the nucleus.
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Gertie

A facility covered with crushed gravel used to suppress the potential radioactive 

contamination from the accidental explosion of a nuclear weapon during assembly. 

Also referred to as a Gravel Gertie.

Gray

SI unit of absorbed dose. Unit symbol, Gy. 1 Gy = 100 rad.

hydroshot

Detonation of a mixture of explosives and DU used as a quality control technique for 

measuring the performance of plastic-bonded explosives.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles 

through interactions with matter.

Line 1

Facilities and operations taken over by the AEC in 1947 for casting of baratols and 

processes related to the assembly of nuclear weapons.

Minimum Detectable Level (MDL)

A term used in this document and other NIOSH documents to refer to a statistically 

determined minimum detection level, Lower Limit of Detectability (LD), and related 

quantities.

Minimum Reportable Dose (MRD)

A general term used to identify the minimum dose recorded and reported, normally 

based on site-specific policy.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen 

atom.

neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, intermediate

Neutrons with energy between 0.5 eV and 10 keV.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. Generally, neutrons with 

energy less than about 0.5 eV.

neutron film dosimeter

A film dosimeter that contains a Neutron Track Emulsion, type A, film packet.

nuclear emulsion

Often referred to as “NTA” film and used to measure personnel dose from neutron 

radiation.
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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 1000X 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).

Parameter 1

The column in the IREP template where the dose reconstructor will enter the 

calculated dose. Multiple entries based on year of employment, type of radiation, 

and appropriate energy ranges; internal and external exposures are possible.

Parameter 2

The column in the IREP template where the dose reconstructor will enter the lower 

limit of the dose distribution based on the radiation type and the dose distribution 

type.

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 mm 

and 10 mm, respectively, for the skin and body. 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 keV and 100 keV whose source 

can be an X-ray machine or radioisotope.

pit

Term used to describe the nuclear physics package of a nuclear weapon. This 

component contains the fissile material that begins the fission chain reaction. 

Sometimes also referred to as the primary.

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 also ionizing radiation.
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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 x 10-4 coulomb 

in 1 kg of dry air. An exposure of 1 R is approximately equivalent to an absorbed 

dose of 1 rad in soft tissue for higher (>100 keV) energy photons.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.007 cm in a material of specified geometry and 

composition. 

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.007 cm 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 mg/cm2.

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 cm (1000 mg/cm2); 

however, this term is also used to refer to the recorded dose.

Work factor (Wf)

Used in this TBD to describe the relative portion of time that a worker would be in 

close proximity to a pit.

X-ray

Ionizing electromagnetic radiation that originates external to the nucleus of an atom.
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Appendix A – Facility Information

The IAAP archival information associated with the DOE program to screen former workers 

for potential occupational health conditions provides a summary of IAAP facilities with some 

potential for radiation exposure to workers. Table A1 lists such facilities.

Table A1. IAAP Line 1 facilities handling radiation.

Building

Number Description Use

Area

Dosimeter

Badges

Comments

1-11

Vault storage for

components 

(pits)

Received and unloaded pits,

some assembly.

1962 -

1974

Stationary air monitors in building. Some 

contamination measured at squash press 

area and squash removal area.

1-12

High explosives,

fabrication,

pressing and

machining

Explosives pressed 

beginning mid-to-late 1960s. None Radiation swipe sampling,1972-1975.

1-13 Assembly area, 

U-235 pits

Assembly of larger nonplutonium

weapons (U-235) and covers.

Radiation swipe data – 1974 for bays AG.

1-18 Research and

development

References to disposing of

Radioactive waste (1971)

1-19 Assembly bays “Experimental Building” Tritium monitors

1-61 Assembly area –

Operating Bays

Nonplutonium -bearing 

weapons

(i.e., uranium)

Tritium monitors. Tritium shipping 

canisters were

sampled and purged in this building.

1-63

Series

Operating bays

and assembly

cells, plutonium

pits post 1956

Cells with Gravel Gerties 

(i.e. used to contain 

distribution of radioactive 

material in event of

accident).

Built in 1957 as area to bring

together explosives and

plutonium physics package

1-64

Series Storage

Cells with Gravel Gerties 

(i.e. used to contain 

distribution of radioactive 

material in event of

accident).

1-65

Series

1-66

Series Radiation swipe data, 1972-73

1-67

Series

Radiation swipe data, 1972-73

1-73 Storage and

Receiving

Storage and receiving in 

early years until about 1957, 

then used as pit storage 

1957 to late 1960s.

1962-

Pits were received in 1958,

nuclear weapons built until 1-63

building was built, then was used

as storage area for pits.

1-77 Pit storage and

Inspection

Pit receiving and inert 

assembly, Built in mid-1960s

Radiation swipes and air sampling data, 

tritium bottles were charged here.

1-80 Radiation swipes data 1974

1-85-2

For subcomponents of depleted center 

items. Building built but not used by AEC.

1-100 X-Ray Linitron X-ray Radiation – explosives X-rayed

For air cavities
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Table A2 Yard C Facilities – Storage of Fissile Materials

Building

number

 

 Description

 

 Use

 Area

 Dosimeter

 Badges Comments

23-15-7 Storage Igloos Used for storage of pits 1965-1974

23-39-15 Storage Igloos Used for storage of pits 1965-1974

23-39-20 Storage Igloos Used for storage of pits 1965-1974

23-39-21 Storage Igloos Used for storage of pits 1965-1974
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Appendix B – List of Weapons Programs at IAAP

Designator Delivery System Assembly Disassembly

Mk IV Bomb 1949 1953

Mk 5 Bomb 1952 1963

Mk 6 Bomb 1951 1962

Mk 7 Bomb 1952 1967

B12 Bomb 1954 1963

B15 Bomb 1955 1965

B17 Bomb 1954 1957

W21 Bomb 1955 1957

W25 Genie 1957 1984

B27

W27

Bomb

Regulus II

1958

1958

1965

1965

B28

W28

Bomb

Hound Dog, Mace

1958

1958

1991

1991

W30 Talos, TADM 1959 1978

W31 Honest John, Nike Hercules, ADM 1958 1989

W34 Lulu, Hotpoint, Astor 1958 1964

B36 Bomb 1956 1962

B39

W39

Bomb

Snark, Redstome

1958

1958

1966

1965

W40 Bomarc, La Crosse 1959 1972

B41 Bomb 1960 1976

B43 Bomb 1961 1972

W44 ASROC 1961 1989

W47 Polaris A1/A2 1960 1975

W49 Thor, Jupitor, Atlas Titan I 1958 1975

W50 Pershing I 1963 1973

W52 Sergeant 1962 1977

B53

W53

Bomb

Titan II

1962

1962

Retired ?

1987

W54 Falcon, Davy Crokett, Special ADM 1964 1978

W56 Minuteman II 1964 1993

B57 Depth/Tactical Bomb 1963 1975

W58 Polaris A3 1964 1968

W59 Minuteman I 1962 1970

B61 Bomb 1967 Enduring

B66 Sprint 1974 1979

W68 Poseidon 1970 1977

W69 SRAM 1972 1992

W71 Spartan 1974 1993

W72 Walleye 1970 1979

*Note that the Designators Mk (Mark), B, W, XW are interchangeable and varied over time.
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Appendix C – Summary of Intakes and Details of Ingestion Calculations

C-1.0 INTAKES

Table C-1 lists intakes by work tasks for time periods applicable for input into IMBA. The 

table includes intakes from environmental releases and exposure from radon. The worker 

category, “all personnel,” applies to all AEC workers even if they are also included in the 

other categories.

Table C-1. Summary of intakes by AEC workers at IAAP.

Dose calculation parameters IREP input Parameters Work or worker 

categorya

Period Material Mode Absorption 

type pCi/d mg/d Distribution

Type 1 2

All personnel 1/1/1948-7/1/1975 DU Chronic inhalation M, S 0.017 4.5E-05 Constant Dose

All personnel 12/2/1965-3/3/1969 DU Chronic inhalation M, S 0.024 6.4E-05 Lognormal Dose 4

All personnel 3/4/1969-7/14/1969 DU Chronic inhalation M, S 0.0012 3.3E-06 Lognormal Dose 4

All personnel 7/15/1969-12/31/1973 DU Chronic Inhalation M, S 0.0055 1.5E-05 Lognormal Dose 4

All personnel 1/1/1954-7/1/1975 HTO Chronic inhalation/

absorption 6.8E+03 Constant Dose

All personnel 1/1/1948-7/1/1975 RadonbChronic inhalation Lognormal

0.072

WLM/

12 mo. 

3

Machinist 1/1/1948-12/31/1962 DU Chronic inhalation M, S 6.6 1.8E-02 Constant Dose

Machinist 1/1/1948-12/31/1962 DU Chronic ingestion Insoluble 0.2 5.4E-4

(mean) Triangular 0.5mean

me

an

c

Burning yard 

operations 1/1/1948-7/1/1975 DU Chronic inhalation M, S 38 0.10 Lognormal Dose 3

Weapons assembly,

disassembly,

surveillance,

inspection

1/1/1954-7/1/1975 HTO Chronic inhalation/

absorption 1.34E+7 Constant Dose

Weapons assembly,

disassembly,

surveillance,

inspection

1/1/1949-12/31/1974 DU Chronic inhalation M, S 223 0.555 Constant Dose

Hydroshot operations 12/2/1965-3/3/1969 DU Chronic inhalation M, S 0.28 7.5E-04 Lognormal Dose 4.6

Hydroshot operations 3/4/1969-7/14/1969 DU Chronic inhalation M, S 0.014 3.8E-05 Lognormal Dose 4.6

Hydroshot operations 7/15/1969-12/31/1973 DU Chronic inhalation M, S 0.064 1.7E-04 Lognormal Dose 4.6

Hydroshot cleanup 12/2/1965-3/3/1969 DU Chronic ingestion Insoluble 1.2E+03 3.3 Constant Dose

Hydroshot cleanup 3/4/1969-7/14/1969 DU Chronic ingestion Insoluble 6.4E+01 0.17 Constant Dose

Hydroshot cleanup 7/15/1969-12/31/1973 DU Chronic ingestion Insoluble 2.9E+02 0.78 Constant Dose

FS-12 Operations 12/2/1965-12/31/1974 DU Chronic inhalation M, S 30.7 0.076 Constant Dose

All personnel 12/2/1965-12/31/1974 DU Chronic ingestion Soluble 0.51 0.001 Constant Dose

a. Workers performing specific listed tasks, e.g. burning yard operations, are also assigned the intakes for “all

personnel.”

b. Applies to workers on Line 1. However, unless it is clear that the worker did not work inside the Line 1 buildings, 

apply the radon intake to everyone.

c. Parameter 3 is 2 x mean.

C2.0 INGESTION INTAKE CALCULATION

Because the ingestion intake calculation for the hydroshot cleanup crew involved picking up 

source material, not touching surfaces contaminated by settling of particles from the air in a 

room, the intake calculation was not based on the draft OTIB on ingestion (ORAU 2004c). 

Instead, an experiment was conducted for this TBD, wherein a small amount of ordinary soil, 

judged by the author to be similar to what a worker would get on his/her hands by handling a 

dirty object, was acquired on a small piece of paper with double-sided sticky tape. The net 

weight of the dirt was measured to be 83 mg with an uncertainty of about 2 mg. The largest 
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uncertainty was in the judgment of the volume of dirt that would be on someone’s hand. Soil 

has a density of about 2 g/cm3, so the volume of dirt was about 0.042 cm3,

For the hydroshot cleanup crew, it was assumed that surfaces of the DU chards were 

partially covered with unexploded HE and dirt from the impact with the ground. In addition, 

the action of picking up chards from the ground would have introduced more dirt to the 

hands. So it was assumed that 10% of the volume of material on the hands would have 

been DU. Assuming the form of the DU was mostly metal with some oxide, the density of 

the DU would have been about 18 g/cm3. So the DU contamination on the hands would 

have weighed about 76 mg [(18,000 mg/cm3) (0.0042 cm3)]. It was then assumed that 10% 

of the DU on the hands was ingested (as suggested in the draft OTIB on ingestion). This is 

an upper bound assumption which includes the possibility that the worker ate or drank 

without washing his/her hands. It was further assumed that this activity occurred after each 

hydroshot, so the ingestion rate in mg/calendar day is simply 

(7.6 mg)(no. shots in period)/(calendar days in period)
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Appendix D – Generic Pit Specifications

Specifications for a “bare” (unshielded) unclassified pit used to 

develop dose ratio for EEOICPA Dose Reconstructions

TD Taulbee

The following discussion is intended to 1) describe the assumptions and 2) provide 

specifications for a “bare” (unshielded) model pit for dose reconstructions conducted under 

the Energy Employees Occupational Illness Compensation Program Act (EEOICPA) of

2000. The “bare” (unshielded) pit specifications in conjunction with MCNP calculations will 

be used to estimate the photon and neutron dose rates and ratios. The dose ratios will 

subsequently be used in conjunction with co-worker data to estimate external radiation 

doses during time periods when individual radiation dosimetry (film badges) were not worn 

and/or were not capable of measuring the low energy photon dose.

Currently, actual pit specifications assembled and disassembled at IAAP remain classified. 

In order to develop unclassified “bare” pit specifications for the dose reconstruction

calculations, classified information was reviewed and discussed with the Department of

Energy (DOE). Under NIOSH procedures, dose reconstructions are to give the benefit of 

the doubt or uncertainty to the claimant. NIOSH considers the classified information as 

uncertainty and, therefore, has given the benefit of the doubt towards overestimating the 

dose.

In the development of these specifications, NIOSH has estimated the mass and the surface 

area of a typical pit handled at IAAP. The effect that each assumption has on the dose rate 

is discussed in more detail below. The tendency of these two estimations increases the 

calculated photon dose rate resulting in an overall overestimate of the photon dose rate. 

Thus when used in conjunction with coworker data from an earlier time period as a ratio, the 

result will overestimate the true dose experienced by IAAP workers. This overestimate is 

considered reasonable and necessary to meet the intent of the EEOICPA and to preserve 

and protect national security interests. If NIOSH used actual pit specifications to calculate 

the photon dose, the resulting dose will be lower than the estimates derived here.

Assumed Mass for Bare Pit

For the “bare” unshielded pit, NIOSH assumes a mass of 6 kg of 15 year aged weaponsgrade plutonium in delta phase (density: ? = 16.00 g/cm3

). This mass was selected since it 

is the approximate mass of plutonium used in the Trinity and Nagasaki nuclear devices.(1) 

This assumed mass is greater than the assumed plutonium mass released in accident 

scenario #1 of the Environmental Impact Statement for the Pantex Plant.(2) In scenario #1, 

an activity of 300 Ci (5 kg) of plutonium is assumed released following an accidental 

detonation in a cell. Based on these considerations, a mass of 6 kg is considered claimant 

favorable. For the bare unshielded pit, weapons-grade plutonium was assumed as opposed 

to enriched uranium since plutonium has a higher photon dose rate.

Assumed Diameter for Nominal Bare Pit

For this nominal “bare” unshielded pit, NIOSH assumes an outer diameter of one foot (12 

inches). The purpose of this assumption is to result in a relatively thin shell such that 

photons (especially low energy photons) are less self-shielded (attenuated) by the assumed 

mass. Using the two criteria (mass and assumed outer diameter), the metal shell thickness 

was calculated to be approximately 1.3 mm (i.e. thin shell). This thin shell will have minimal 
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 71 of 79

photon attenuation and will, therefore, overestimate the true photon dose rate from pits 

handled at IAAP.

Nominal Bare (Unshielded) Pit Specifications

Assumed Mass of Weapons-grade Plutonium: 6.0 kg

Assumed Outer Diameter: 12 inches

Calculated Shell Thickness: 1.3 mm

Assumed Cladding: None

Using these specifications, dose rates were calculated at 100 cm using MCNP. Table D.1 

provides the neutron and photon energy distribution results and dosimetry correction factors.

Table D.1 Generic Pit Dose Rate Information

Description Designator Energy Interval Dose Rate 

(mrem/h)

Correction 

Factors

(P*/C7)a

Total Photon PT 33.3

Low Energy P1 < 30 keV 1.18 0.080

Intermediate Energy P2 30 – 250 keV 27.8 1.878

High Energy P3 > 250 keV 4.49 0.303

Measured Energy P4 > 70 keV 7.4

Film Badge Energy C7 > 70 keV + 0.3*(50-70 keV) 14.8

Total Neutron NT 0.582

Thermal N1 < 10 keV 0.004

Intermediate Energy N2 10 – 100 keV 0.008

Fission N3 0.1 – 2 MeV 0.334

Fast N4 2 - 20 MeV 0.242

Ultra Fast N5 > 20 MeV 0.000

a. Correction factor for dosimeter response according to IREP input energy ranges.

In summary, while the assumptions listed above will result in a dose rate that exceeds the 

true dose rate, the assumptions are considered reasonable and necessary in order to meet 

the intent of EEOICPA and preserve and protect national security interest.

References

1. Groves, Leslie (1945) Memorandum concerning the Trinity test dated July 16, 

1945.

2. DOE (2003) Supplemental Information to the Final Environmental Impact

Statement for the Continued Operation of the Pantex Plant and Associated 

Storage of Nuclear Weapon Components. DOE/EIS-0225
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Appendix D cont. – Generic Pit Calculated Spectral Data (Figure 6.1)

Energy 

(keV)

Dose 

(mrem/h)

Energy 

(keV)

Dose 

(mrem/h)

Energy 

(keV)

Dose 

(mrem/h)

Energy 

(keV)

Dose 

(mrem/h)

10 3.948E-04 510 2.227E-03 1010 9.090E-06 1510 0.000E+00

20 1.127E+00 520 9.017E-03 1020 1.265E-05 1520 1.490E-05

30 5.683E-02 530 2.658E-05 1030 8.016E-03 1530 0.000E+00

40 3.118E-02 540 9.524E-03 1040 3.068E-05 1540 4.505E-03

50 1.141E+00 550 2.414E-03 1050 8.160E-06 1550 2.158E-05

60 2.354E+01 560 2.427E-03 1060 8.222E-06 1560 2.198E-05

70 4.134E-02 570 1.476E-02 1070 8.306E-06 1570 0.000E+00

80 1.158E-01 580 1.001E-02 1080 0.000E+00 1580 0.000E+00

90 8.219E-02 590 5.096E-03 1090 2.318E-05 1590 0.000E+00

100 4.204E-01 600 5.152E-03 1100 1.079E-05 1600 1.015E-05

110 5.301E-01 610 1.843E-02 1110 1.379E-05 1610 9.436E-06

120 4.922E-01 620 3.186E-02 1120 1.886E-05 1620 1.906E-05

130 2.082E-01 630 1.892E-02 1130 1.687E-05 1630 7.594E-06

140 1.422E-01 640 1.094E-02 1140 0.000E+00 1640 0.000E+00

150 1.637E-01 650 2.216E-02 1150 4.179E-03 1650 2.692E-05

160 1.092E-02 660 2.535E-02 1160 0.000E+00 1660 0.000E+00

170 1.300E-01 670 1.419E-02 1170 0.000E+00 1670 4.093E-06

180 4.614E-02 680 1.152E-02 1180 4.181E-03 1680 2.597E-05

190 4.075E-02 690 3.507E-02 1190 8.598E-06 1690 2.930E-05

200 3.985E-02 700 3.246E-02 1200 6.808E-06 1700 0.000E+00

210 7.157E-02 710 2.999E-03 1210 2.883E-05 1710 9.785E-06

220 1.024E-01 720 1.516E-02 1220 1.354E-05 1720 4.583E-03

230 8.090E-02 730 2.145E-02 1230 4.223E-03 1730 2.377E-05

240 9.768E-02 740 1.243E-02 1240 1.782E-05 1740 0.000E+00

250 1.369E-01 750 9.370E-03 1250 2.065E-05 1750 1.549E-05

260 2.929E-02 760 1.588E-02 1260 1.494E-05 1760 2.206E-05

270 2.033E-02 770 1.606E-02 1270 0.000E+00 1770 3.645E-05

280 2.476E-02 780 9.725E-03 1280 0.000E+00 1780 5.140E-05

290 2.611E-02 790 1.964E-02 1290 1.835E-05 1790 2.049E-05

300 3.307E-02 800 6.663E-02 1300 3.486E-06 1800 1.363E-05

310 1.239E-01 810 3.386E-03 1310 4.980E-05 1810 4.578E-06

320 1.331E-01 820 3.432E-03 1320 0.000E+00 1820 0.000E+00

330 1.427E-01 830 3.455E-03 1330 1.484E-05 1830 0.000E+00

340 1.737E-01 840 1.114E-05 1340 0.000E+00 1840 9.838E-06

350 1.584E-01 850 2.665E-05 1350 1.300E-05 1850 0.000E+00

360 3.086E-01 860 1.161E-05 1360 1.594E-05 1860 9.859E-06

370 3.272E-01 870 3.557E-03 1370 6.449E-06 1870 0.000E+00

380 3.153E-01 880 4.957E-06 1380 0.000E+00 1880 9.638E-06

390 3.451E-01 890 5.044E-06 1390 2.707E-05 1890 4.609E-06

400 3.527E-01 900 3.617E-03 1400 0.000E+00 1900 0.000E+00

410 2.086E-01 910 7.790E-06 1410 1.044E-05 1910 0.000E+00

420 2.304E-01 920 4.994E-06 1420 1.059E-05 1920 0.000E+00

430 2.450E-01 930 2.835E-05 1430 2.975E-05 1930 6.915E-06

440 2.758E-01 940 1.189E-05 1440 8.050E-06 1940 0.000E+00

450 2.612E-01 950 3.073E-05 1450 6.417E-06 1950 2.934E-05

460 1.350E-02 960 0.000E+00 1460 0.000E+00 1960 2.450E-05

470 1.390E-02 970 5.169E-05 1470 0.000E+00 1970 0.000E+00

480 1.636E-02 980 0.000E+00 1480 0.000E+00 1980 1.892E-05

490 1.471E-02 990 8.862E-06 1490 0.000E+00 1990 7.605E-06

500 2.584E-02 1000 2.087E-05 5000 2.190E-05 2000 0.000E+00
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 73 of 79

Appendix E – Special Nuclear Material (SNM) Encapsulation and Pit Cladding

During the public meeting in June 2004, there was mention of “bare” pits, and the pits were 

thermally warm to the touch. In a follow-up worker meeting in July 2004, there was some 

discussion about this thermal conductivity and that some types of gloves were better than 

others to protect energy employees’ hands during direct handling of pits. The radiological 

concerns with this issue have both internal and external dose implications. To address this 

issue, NIOSH has reviewed classified documentation and interviewed Department of Energy 

(DOE) personnel knowledgeable of pit designs and assembly and disassembly operations.

Internal Dose

Based on NIOSH’s review, all nuclear capsules and pits were clad with some type of 

material. From the energy employee viewpoint, the pits could have appeared and felt 

through thermal radiance (heat) as if they were bare plutonium or uranium metal. According 

to DOE, all the pits manufactured since the late 1940s are known to be clad. In some 

instances, this cladding could be a very thin layer of material, thus the pit could be thermally 

warm and could also look like bare metal. As a result of our analysis, NIOSH has

determined that there is no significant potential for internal exposure from pits.

It should be noted that radiological incident records at Pantex have identified a few

accidents where the cladding was compromised resulting in contamination and airborne 

radioactivity. These incidents were not common, and the specific instances have been well 

documented. NIOSH has reviewed over 200 incident reports from IAAP between 1959 and 

1974. From these reports, NIOSH has identified 15 incidents that involved radioactive 

materials in which special radiological surveys or precautionary building evacuations were 

conducted. To date, there is no evidence that a radiological incident resulted in the release 

of special nuclear materials at IAAP.

External Dose

While there is no significant potential for internal exposure from nuclear capsules (pits), the 

cladding material and thickness can affect the external dose rate. This is particularly 

problematic for lower energy photons (< 70 keV). High-Z materials such as uranium greatly 

reduce the photon dose rate and almost completely shield the worker from low energy 

photon emissions (< 30 keV). Conversely, low-Z materials such as beryllium, do not provide 

much shielding and allow a relatively large quantity of low energy photons (< 30 keV) to 

pass. Through discussions with DOE, NIOSH has learned that exact information on the 

cladding material and thickness for each weapon design assembled and disassembled at 

IAAP remains classified.

It is important to note that not all components had a significant low energy photon dose. 

There are three basic types of pits used in assembly at IAAP; 1) enriched uranium pits, 2) 

plutonium pits, and 3) composite pits (combination plutonium and enriched uranium). In the 

composite pits, the plutonium always had an outer shell of enriched uranium which would 

effectively shield worker exposure to lower energy photons. Since the low energy photon

dose from enriched uranium is negligible, only the plutonium pits had the potential for 

significant low energy photon dose.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 74 of 79

Although the cladding information necessary to precisely estimate the external dose remains 

classified, an effective maximum photon emission rate (radiation dose) can be determined 

by assuming that there was no cladding or shielding between the fissile materials (pits) and 

the energy employee. For external dose reconstruction only, NIOSH assumes that: 1) all 

pits were plutonium, and 2) there is no cladding of the fissile materials. This assumption will 

result in an overestimate of the true external dose for both low energy and intermediate 

energy photons. Furthermore, since low energy photons could not be accurately measured 

with the film badge dosimeter used at IAAP (i.e. the shielded window blocked low energy 

photons and attenuated photons less than about 70 keV), no cladding is assumed for low 

energy photon dose over the entire operations history at IAAP (1949-1974).

Since most of the fissile materials in weapons assembled and disassembled at IAAP were of 

a design that greatly diminished the low energy photon dose, the assumptions stated above 

will overestimate the total external photon dose. The assumption is, however, considered 

reasonable and necessary in order to meet the intent of the Energy Employees

Occupational Illness Compensation Program Act of 2000 (EEOICPA) and still preserve and 

protect national security interests.
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 75 of 79

Appendix F – IAAP Annual Dosimeter Badge Distributions

1955 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 32.0 mR

GSD = 1.26

r 2 = 0.82

n = 7

1956 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 25.9 mR

GSD = 1.30

r 2 = 0.80

n = 10

1957 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 17.48 mR

GSD = 1.36

r 2 = 0.90

n = 33

1958 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 19.02 mR

GSD = 1.85

r 2 = 0.93

n = 99

1959 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 9.96 mR

GSD = 1.54

r 2 = 0.85

n = 59

1960 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 11.21 mR

GSD = 1.52

r 2 = 0.88

n = 92
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 76 of 79

1962 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 11.9 mR

GSD = 1.32

r 2 = 0.52

n = 33

1963 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

Bi-Weekly Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 18.0 mR

GSD = 1.67

r 2 = 0.85

n = 295

1964 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 13.5 mR

GSD = 1.44

r 2 = 0.68

n = 156

1965 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 17.4 mR

GSD = 1.70

r 2 = 0.82

n = 133

1966 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 19.0 mR

GSD = 1.62

r 2 = 0.86

n = 172
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 77 of 79

1967 - Non-zero Data 

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 20.1 mR

GSD = 1.78

r 2 = 0.88

n = 140

1968 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 27.6 mR

GSD = 2.03

r 2 = 0.92

n = 222

1969 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 22.8 mR

GSD = 1.88

r 2 = 0.88

n = 292

1970 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Reading (mR)

1

10

100

1000

Coefficients:

GM = 39.5 mR

GSD = 2.93

r2 = 0.92

n = 848

1971 - Non Zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 45.1 mR

GSD = 2.75

r 2 = 0.95

n = 1527

1972 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 30.8 mR

GSD = 2.47

r 2 = 0.91

n = 1086
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 78 of 79

1973 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 42.5 mR

GSD = 2.36

r 2 = 0.96

n = 627

1974 - Non-zero Data

Probability

0.5 1 2 5 10 20 30 50 70 80 90 95 98 99

4 Week Dosimeter Badge Readings (mR)

1

10

100

1000

Coefficients:

GM = 40.5 mR

GSD = 2.40

r 2 = 0.96

n = 342
Effective Date: 03/14/2005 Revision No. 01 Document No. ORAUT-TKBS-0018 Page 79 of 79

Appendix G –Bladder Cancer Example Calculation for an Unmonitored Worker

Annual Dose (mrem) Bladder DCFs HP(10)

Year

Initial 

Dose 

(mrem)

Am241

ingrowth

R to 

HP(10) 

Correction

Estimated 

Annual 

Dose

< 30 

keV

30-250 

keV

>250 

keV

< 30 

keV

30-

250 

keV

>250 

keV

Bladder 

Total 

(mrem)

0.08 1.88 0.303 0.17 0.873 0.913

1949 4529 0.67 1 3034 243 5699 919 41 4975 839 5856

1950 4529 0.71 1 3216 257 6039 974 44 5272 890 6205

1951 4529 0.75 1 3397 272 6379 1029 46 5569 940 6555

1952 4529 0.78 1 3533 283 6634 1070 48 5792 977 6817

1953 4529 0.81 1 3668 293 6889 1112 50 6014 1015 7079

1954 4529 0.84 1 3804 304 7145 1153 52 6237 1052 7341

1955 4529 0.87 1 3940 315 7400 1194 54 6460 1090 7604

1956 4529 0.9 1 4076 326 7655 1235 55 6683 1128 7866

1957 4529 0.93 1 4212 337 7910 1276 57 6905 1165 8128

1958 4529 0.95 1 4303 344 8080 1304 59 7054 1190 8303

1959 4529 0.98 1 4438 355 8335 1345 60 7277 1228 8565

1960 4529 1 1 4529 362 8505 1372 62 7425 1253 8740

1961 4529 1 1 4529 362 8505 1372 62 7425 1253 8740

1962 4529 1 1 4529 362 8505 1372 62 7425 1253 8740

1963 468 1 1.05 491 39 923 149 7 806 136 948

1964 176 1 1.05 185 15 347 56 3 303 51 357

1965 226 1 1.05 237 19 446 72 3 389 66 458

1966 247 1 1.05 259 21 487 79 4 425 72 500

1967 261 1 1.05 274 22 515 83 4 449 76 529

1968 354 1 1.05 372 30 698 113 5 609 103 717

1969 296 1 1.05 311 25 584 94 4 510 86 600

1970 514 1 1.05 540 43 1014 164 7 885 149 1041

1971 586 1 1.05 615 49 1156 186 8 1009 170 1187

1972 400 1 1.05 420 34 789 127 6 689 116 810

1973 553 1 1.05 581 46 1090 176 8 952 161 1120

1974 527 1 1.05 553 44 1039 168 8 907 153 1068