ORAU Team 

NIOSH Dose Reconstruction Project 

Technical Basis Document for Atomic Energy Operations at the 

Iowa Army Ammunition Plant (IAAP) 

Document Number: 

ORAUT-TKBS-0018 

Effective Date: 04/16/2004 

Revision No.: 00 

Controlled Copy No.: ________ 

Page 1 of 58 

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

Shipler/Bruce Napier 

Document Owner 

Approval: Signature on File Date: 04/16/2004 

Donald Bihl, TBD Team Leader 

Approval: Signature on File Date: 04/19/2004 

Judson L. Kenoyer, Task 3 Manager 

Concurrence: Signature on File Date: 04/19/2004 

Richard E. Toohey, Project Director 

Approval: Signature on File Date: 04/16/2004 

James W. Neton, OCAS Health Science Administrator 

Supersedes: 

None 

TABLE OF CONTENTS 

Section Page 

Record of Issue/Revisions ......................................................................................................................5
 

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

1.0 Site Description, Operational History, and Process ....................................................................7
 

2.0 Estimation of Internal Exposure ..................................................................................................8
 

2.1 Tritium Exposure ...................................................................................................................9
 

2.2 Depleted Uranium ...............................................................................................................10
 

2.2.1 DU Intakes from Hydroshots .........................................................................................11
 

2.2.2 DU Intakes from Machining Baratols.............................................................................13
 

2.2.3 DU Intakes by Operators at Burning Yard.....................................................................14
 

2.2.4 DU Intakes from Disassembly of Weapons...................................................................15
 

2.3 Other Sources .....................................................................................................................16
 

3.0 Estimation of External Radiation Exposure...............................................................................16
 

3.1 Introduction .........................................................................................................................16
 

3.2 Basis of Comparison ...........................................................................................................16
 

3.3 Workplace External Radiation Fields ..................................................................................17
 

3.3.1 Photon Radiation...........................................................................................................17
 

3.3.2 Neutron Radiation .........................................................................................................19
 

3.3.3 Depleted Uranium .........................................................................................................20
 

3.4 External Radiation Dose Records .......................................................................................20
 

3.5 Dosimeter Technology ........................................................................................................23
 

3.6 Dosimeter Performance Studies .........................................................................................24
 

3.6.1 Photon Radiation...........................................................................................................24
 

3.6.2 Neutron Radiation .........................................................................................................25
 
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3.7 IAAP Worker External Dose Reconstruction.......................................................................25
 

3.7.1 Photon Dose .................................................................................................................26
 

3.7.2 Neutron Dose ................................................................................................................27
 

3.7.3 Skin and Extremity Dose (Reserved) ............................................................................29
 

3.7.4 IAAP Facilities ...............................................................................................................29
 

3.7.5 IAAP Nuclear Workers ..................................................................................................29
 

3.7.6 Summary of Dose Reconstruction Recommendations .................................................29
 

3.7.7 Radiation Dose Fraction................................................................................................32
 

3.8 Uncertainty in Photon and Neutron Dose ...........................................................................32
 

3.9 Organ Dose.........................................................................................................................32
 

4.0 Occupationally Related Medical X-Rays ...................................................................................34
 

5.0 Occupational Environmental Dose............................................................................................35
 

5.1 Introduction .........................................................................................................................35
 

5.2 Intakes from Onsite Atmospheric Radionuclide Concentrations .........................................35
 

5.2.1 Intakes from Tritium Releases ......................................................................................36
 

5.2.2 Intakes from Release of DU from Burning Sites ...........................................................37
 

5.3 External Dose .....................................................................................................................38
 

5.4 Uncertainty..........................................................................................................................38
 

6.0 Radon .......................................................................................................................................38
 

7.0 Yearly Intakes ...........................................................................................................................39
 

References ...........................................................................................................................................40
 

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

Attachment A Position Titles and Tasks............................................................................................ 49 

Attachment B Occupational External Dose for IAAP Workers .......................................................... 50 

B1.0 DOE IAAP Dose Records .........................................................................................................50
 

B2.0 Dose Parameters......................................................................................................................51
 

B2.1 Years of Exposure ..............................................................................................................52
 

B2.2 Rate ....................................................................................................................................52
 

B2.3 Radiation Type ....................................................................................................................52
 

B2.4 Radiation Energy and Dose Fraction ..................................................................................52
 

B3.0 Distribution Parameters ............................................................................................................53
 

B3.1 Type ....................................................................................................................................53
 

B3.2 Parameter 1 ........................................................................................................................53
 

B3.3 Parameter 2 ........................................................................................................................56
 

Attachment C Summary of Intakes and Details of Ingestion Calculations......................................... 60 

C1.0 Intakes ......................................................................................................................................57
 

C2.0 Ingestion Intake Calculation ......................................................................................................57
 
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LIST OF TABLES 

Table Page
 

1 Mass and radiological characteristics of depleted uranium .........................................................10
 

2 Measured DU air concentrations from hydroshots.......................................................................11
 

3 Chronic DU intakes from hydroshots for test fire operators and drivers ......................................13
 

4 Chronic DU intakes from hydroshots for other AEC personnel....................................................13
 

5 Chronic DU ingestion intakes from hydroshots for cleanup crew ................................................13
 

6 Summary of examination of IAAP worker dose records ..............................................................21
 

7 IAAP worker recorded dose data statistics ..................................................................................22
 

8 IAAP Mason and Hanger monitored workers and reported whole-body doses ...........................23
 

9 Summary of IAAP beta/photon dosimeters, MDL, exchange frequency and 

potential missed dose..................................................................................................................24
 

10 IARC testing results for U.S. beta/photon dosimeters .................................................................25
 

11 Testing results for Hanford two-element and multielement film dosimeters for 

energy and angular response ......................................................................................................25
 

12 Missed photon dose adjustments to recorded deep dose ...........................................................27
 

13 Statistical parameters for lognormal probability distribution for neutron-to-photon 

dose ratios ...................................................................................................................................29
 

14 IAAP neutron dose fractions and associated ICRP 60 correction factors....................................29
 

15 IAAP facilities handling radiation..................................................................................................30
 

16 IAAP total manpower and Division B workers, and AEC Form 190 reported nuclear 

and monitored workers ................................................................................................................31
 

17 Summary of IAAP worker claimant-favorable dose reconstruction options .................................31
 

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

radiation exposures .....................................................................................................................32
 

19 Annual photon exposure to organ dose conversion factors.........................................................33
 

20 Annual neutron deep dose equivalent to organ dose conversion factors ....................................33
 

21 Organ doses from lumbar spine X-rays at IAAP ..........................................................................35
 

22 Summary statistics of 1990 radon measurements at Pantex.......................................................39
 

B-1 IREP dose parameter input screen..............................................................................................51
 

B-2 Selection of radiation type, energies and percentages ................................................................52
 

B-3 Distribution parameters for IAAP worker measured photon dose ................................................53
 

B-4 Missed photon dose adjustments to DOE reported gamma dose ...............................................54
 

B-5 Statistical parameters for lognormal probability distribution for neutron-to-photon 

dose ratios ...................................................................................................................................54
 

B-6 IAAP neutron dose fractions, energies, percentages, and associated ICRP 60 

correction factors .........................................................................................................................54
 

B-7 Annual photon exposure to organ dose conversion factors.........................................................55
 

B-8 Annual neutron deep dose equivalent to organ dose conversion factors ....................................55
 

C-1 Summary of intakes by AEC workers at IAAP .............................................................................57
 
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LIST OF FIGURES 

Figure Page
 

1 Layout of the IAAP site with AEC facilities marked ........................................................................7
 

2 MCNP calculated photon spectra emitted from 234mPa beta in 238U spheres................................18
 

3 Bremsstrahlung component of the calculated spectrum on a linear vertical axis ........................19
 

4 Characteristic X-ray component of calculated spectrum on a linear vertical axis ........................19
 

5 Trends in dosimeter measured IAAP worker deep, skin, neutron and extremity 

collective dose .............................................................................................................................22
 

6 Lognormal probability plot of measured annual IAAP worker deep (photon) dose ......................26
 

7 Lognormal plot of Pantex worker positive neutron 809/812 dosimeter data ................................28
 

8 Lognormal probability plot of Hanford plutonium facility worker measured neutronto-photon dose ratio.....................................................................................................................28
 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 5 of 58 

RECORD OF ISSUE/REVISIONS 

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 

reserved. External whole body 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. 
CV 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 6 of 58 

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 

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 

OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 7 of 58 

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 

1.0 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 (U.S. Army 

1988). 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 (JAYCOR 1996). 

Since operations began in 1941, IAAP has used 

explosives and lead-based initiating compounds 

to produce a wide variety of ordnance items. In 

1947, 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. 

The main function of the site was nuclear 

weapon fabrication and final assembly/ 

disassembly. Mason & Hanger-Silas Mason 

Co. assumed the contract for AEC functions at 

Figure 1. Layout of the IAAP site with AEC 

facilities marked. 
2.0 

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BAECP in 1950. Mason & Hanger-Silas Mason Co. also operated the Pantex Plant, which had a 

similar mission, in Amarillo, Texas. Many of the health physics policies and practices were similar. 

For example, both IAAP and Pantex used film badges first provided by Tracerlab and then by 

Landauer, but Pantex had a short period when it procured its dosimetry from Eberline. AEC functions 

at BAECP ended in July 1975. Only AEC operations between 1947 and 1975 are applicable to the 

National Institute for Occupational Safety and Health (NIOSH) Dose Reconstruction Project. The 

exact start date for assembly of nuclear weapons is uncertain, but radiation exposures did not likely 

occur until after the modification of Line 1 in 1948. Production of explosives for use in nuclear 

weapons began in late 1948 (TN & Associated 2001). The date of first assembly of nuclear weapons 

has not been discovered except that it was after June 1949 (Poole and Harrison 1954). It is assumed 

that radiation exposures began in 1948. 

Radiation sources at IAAP included depleted uranium (DU), enriched uranium (EU), plutonium, tritium, 

210Po, and radium, 60Co, and x-ray generating devices for radiography. Additionally, work-required 

medical screening x-ray examinations were included in the IAAP program. 

IAAP worker titles and tasks are summarized in Attachment A. 

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. § 7384l (5) and (12)). 

ESTIMATION OF INTERNAL EXPOSURE 

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

routinely monitored removable and nonremovable 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, the radionuclides most likely to result in an intake at 

IAAP were DU and tritium (3

H). A radiological survey of the plant conducted in 2001, years after 

radiological work had ceased, found only DU and 137Cs. The document providing the results of the 

survey states, “Radioactive materials used at the line were received in a sealed configuration and 

were swipe tested before use. Known radioactive materials include DU, enriched uranium, plutonium, 

tritium gas, and 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 or swipe records were discovered during the records search. Based on experience at 
 

2.1 

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Pantex, there was no potential for intakes during assembly. The canisters of tritium gas arrived 

sealed and assembly did not involve breaking seals or release of the gas. The plutonium and EU 

were also sealed and not available for intakes, even during disassembly. Pantex did some plutonium 

bioassay, but no intakes were ever recorded, with the exception of one well-documented accident. 

No reports of similar accidents at IAAP were found in the literature. The 210Po arrived sealed and was 

not disassembled at IAAP. 

Note: Intakes discussed below and in Sections 5 and 6 are summarized by year and work task in 

Table C-1 in Attachment C. 

TRITIUM EXPOSURE 

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

disassembly procedures. 

Concerning estimation of intakes of tritium, the technology of tritium usage in nuclear weapons is 

classified. Therefore, information is not available on source terms for tritium at IAAP. No tritium 

bioassay results or tritium air sampling results for IAAP 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. Allowing for 

possible different circumstances between Pantex in the 1970-80s and IAAP from the early 1950s, a 

reasonable upper bound for tritium uptake would be 800 μCi per year of exposure. The uptake as 

calculated using ICRP 10 methodology accounted for tritium in body fluids from any and all intake 

modes. 

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 be found; historical reviews of detection capabilities at 

other AEC programs indicate the calculated upper bound for the IAAP annual uptake of 800 μCi 

should have been readily detectable, with the possible exception of the later 1940s and first few years 

of the 1950s. 

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

information regarding the release of tritium from the site. 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 
2.2 

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escaped containment. Most of the tritium released from the stack was probably tritium gas, whereas 

the worker dose comes from tritiated water. The 800-μCi annual uptake estimated in the paragraph 

above is about 3% of the total airborne effluent for a typical year in the 1960s and greater than 10% of 

the effluent for other years. It is improbable that any single worker had an annual uptake approaching 

3% of the effluent, especially not of tritiated water. 

As discussed in the preceding two paragraphs, the 800-μCi uptake is an overestimate, especially 

when applied to all years of employment, and, therefore, for IREP input the distribution type is 

constant. Because of the way IMBA handles tritium intakes, even inhalation intakes, uptake equals 

intake (personal communication Anthony James 2003), so the 800-μCi uptake can be modeled as an 

800-μCi injection or inhalation. The tritium reservoirs came from the Savannah River Site (SRS), so 

the earliest possible date for tritium exposure at IAAP would be 1954. Intakes of tritium would have 

occurred only during disassembly, but it is not known when the first disassembly occurred. It is 

claimant-favorable to assume tritium intakes occurred for each year a worker was at the 

assembly/disassembly line from 1954 to 1975. 

DEPLETED URANIUM 

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

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

nuclear weapons core). 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 by the NIOSH Office of Compensation Analysis and Support 

(OCAS), but within the variability of batches of DU. Either the values in Table 1 or the IMBA values 

can be used. 

Table 1. Mass and radiological characteristics of depleted 

uranium. 

Weight percentagea 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, limited air 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 11 of 58 

sampling data were taken during DU operations. Concerning the 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. 

2.2.1 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. 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 plastic-bonded explosives (PBX) produced at IAAP. Mr. Shannan described the 

hydroshot operation (Fix and Bihl 2003). Hydroshots were conducted outdoors. 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. 

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

Shahan 1972), summarized in Table 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 claimantfavorable, 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). 

Table 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 shotc 0.0 – 9.12 0.9 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 12 of 58 

1 mile from shotd 0.0 – 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 

No particle size distribution information for the hydroshots was available so, the default of 5-μm 

Activity Median Aerodynamic Diameter (AMAD) particle size should be assumed. The intakes are 

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 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. 
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Table 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 for conservativism); 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 4 and C-1. The intakes are 

chronic, lognormally distributed, with a GSD of 4.0, obtained from the probability plot. 

Table 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 exploded baratols lying around ground zero by hand without gloves to bag them 

as waste. If so, 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 

ingestion of about 7.6 mg of DU per cleanup task. (Details of the calculation are provided in 

Attachment C.) There is no record of how many times cleanup of baratol pieces occurred; a claimantfavorable assumption is after each hydroshot. Results of the calculations are provided in Tables 5 

and C-1. This should be modeled as chronic ingestion, insoluble material, constant upper bound. 

Table 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 

2.2.2 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. Both baratols and hydroshot explosive charges might have 

contained a thin shell of DU. Machining or grinding these components might have released small 

quantities of DU to the machining room environment. DU was not machined directly but unintentional 

“nicking” of the DU occasionally occurred during machining on the explosive charges. DUcontaminated 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 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 14 of 58 

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 might 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 workhours 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 claimantfavorable 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 × 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 × 10-3 μCi per year 

(6.6 pCi/day for input into IMBA). 

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, OCAS 

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

(0.2)(1 ×10-12 μCi/cm3

)(106 pCi/μCi)(106 cm3/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 contamination 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”). . 

2.2.3 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 nonradioactive 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, page 4-9). The scenario considered most appropriate for bagging ash 

was described as “free-fall spill of cohensionless powders: free-fall <3 m, air velocity normal to 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 15 of 58 

powder flow, general forced enclosure ventilation or low-wind outside conditions.” The median 

airborne release fraction was 3 × 10-4 and the median respirable fraction was 0.5; the upper bound 

values for the same parameters were 2 × 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 min. 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 × 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. 

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 1948 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. 

2.2.4 DU Intakes from Disassembly of Weapons 

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

Pantex indicates that the DU material in the disassembled weapons was clean metal with no 

reasonable chance of airborne contamination until about 1980 (oxidized parts were encountered after 

that date). 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 the situation was different from 1949 to 1957, prior to when Mr. 

Shannan was employed at IAAP. Experience at Pantex indicated that “about a half of a cup” of 

oxidized DU was available for resuspension. 

1⁄2 cup = 118 cm3 

The density of UO2 is 11 g/cm3

, so the mass of UO2 is 

(118cm3

)(11 g/cm3) = 1,300 g 

of which about (238/270)(1,300g) = 1,140 g is DU. 

The airborne release fraction and respirable fraction were obtained from the DOE handbook (1994) 

using the scenario described as release of pressurized gases over a solid surface contaminated with 

loose oxide where the gas does not significantly pressurize the confinement in which the 

contamination exists. Only upper bound values were presented in the handbook for this scenario; the 

airborne release fraction was 0.005, and respirable fraction was 0.3. The volume of air into which the 
 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 16 of 58 

contamination was suspended was assumed to be 27 m3

, and the exposure time was 1 hr. 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)(0.005)(.3)(1.2 m3

/hr)(1 hr)(106 μg/g)/27m3 = 7.63 × 104 μg or 2.84 × 104 pCi. 

The frequency of disassembly of old weapons is not known but should have been rare considering 

that the primary mission at BAECP was assembly. An assumption was made that no single worker 

was involved in more that two disassemblies per year. The intake should be modeled as an acute 

intake of DU at the start of each year of 5.7 × 104

 pCi total alpha activity, constant upper bound. 

Because assembly of weapons did not occur until some time after June 1949, disassembly of oxidized 

weapons was assumed to start in 1950. Hence, these intakes would apply to 1950 through 1957 

only. 

2.3 OTHER SOURCES 

In addition to DU, EU, plutonium, 210Po, and perhaps thorium were present during assembly or 

disassembly of nuclear weapons. These sources were sealed, and with the careful control of 

contamination before work on the pits was allowed, it is unlikely these radioelements would have been 

available for intake. This is consistent with the experience at Pantex. 

3.0 ESTIMATION OF EXTERNAL RADIATION EXPOSURE 

3.1 INTRODUCTION 

Information concerning the early history of IAAP nuclear weapons assembly activities still involves 

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

DOE (1997) states that the AEC built two nuclear weapons assembly plants to supplement Sandia 

nuclear weapons assembly activities that began in about 1945. One of these was at IAAP in 1947 

and the second was the Pantex Plant in 1951. As described in Section 1, construction of IAAP 

actually began in 1941 to produce conventional explosives for the U.S. Army. As of 2004, IAAP 

continues to operate as an ammunition manufacturing plant; however, nuclear weapon assembly 

activities ended in 1975 when the work was transferred to the Pantex Plant. Archival 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 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. 

3.2 BASIS OF COMPARISON 

Since the initiation of the Manhattan Engineer District (MED) project in the early 1940s, various 

concepts and quantities have been used to measure and record occupational radiation dose at the 

many MED/AEC/DOE facilities. A common basis of comparison has been selected to assess the 

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. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 17 of 58 

consistency of the available historical recorded dose at IAAP, which terminated AEC/ERDA 

operations in 1975, and at Pantex, which assumed the IAAP activities in 1975, compared to current 

Pantex dosimetry performance and field tested capabilities. Dates of changes in the IAAP and Pantex 

dosimetry systems are known and comparisons of recorded doses prior to and following these 

changes provides a capability to assess consistency. Similar radiation beams have been used 

historically to calibrate and conduct performance testing of dosimetry systems (AEC 1955, Unruh et 

al. 1967, McDonald et al. 1983). This basis, to be used in dose evaluation or reconstruction, is the 

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

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

mm and is noted as Hp(0.07). For penetrating radiation of significance to whole-body dose, d = 10 

mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are recommended for use as the operational 

quantity to be recorded for radiological protection by the International Commission on Radiological 

Units and Measurements (ICRU 1993). These are the radiation quantities used in the DOE 

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

systems in the DOE Complex (DOE 1986) including Pantex. 

3.3 WORKPLACE EXTERNAL RADIATION FIELDS 

Doses to IAAP workers conducting nuclear weapon assembly activities were considered to be low 

(Fix and Bihl 2003). 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 workplace radiation fields at IAAP were due to processes involving depleted 

uranium (DU), nuclear weapon components associated with plutonium and highly enriched uranium 

(HEU) and their radioactive progeny. Some early nuclear weapons contained 210PoBe initiators in the 

center of the core, what we would now call a pit, which is a solid, hermetically sealed object (DOE 

1997). Thus, people at the weapons assembly and disassembly plants (Pantex, Burlington, 

Clarksville, and Medina) never directly handled Po-Be initiators. A nuclear weapon must be initiated 

by neutrons at the proper time. Stray neutrons are avoided around a nuclear weapon. Significant 

aspects of the information are classified. However, some general information is available. 

3.3.1 Photon Radiation 

Photon (x-ray and gamma) radiation was associated with several IAAP work activities. Sources of 

ionizing radiation at IAAP included 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. No inventory of these small sources was found in the archival material 

reviewed. Doses associated with the proper, and widespread, use of small check sources is generally 

comparatively negligible. 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, had the potential for producing significant exposure to workers if not used properly. 

Gamma radiation of 2.2 MeV resulted from 1

H (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 nuclear components of purified metals. The 

purification process separates natural progeny radionuclides from their parent metals. 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 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 18 of 58 

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. Thus, for IAAP nuclear weapon assembly 

activities between approximately 1948 and 1975, this is likely to be much less significant than 

experience with weapons disassembled years after their assembly where 241Am activity is significant. 

An important progeny nuclide for potential worker exposure in the 238U decay is 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, IRPA 1983). An additional source of exposure in the IAPP workplace is 

from bremsstrahlung produced in high Z materials from interactions with higher-energy beta particles. 

Beta particles emitted by 234mPa excite both bremsstrahlung and characteristic x-rays in depleted 

uranium or 238U. 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 sphere. Figure 2 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. The bremsstrahlung 

and characteristic x-ray components of the calculated spectrum in Figure 2 has been presented, 

respectively, on a linear vertical axis in Figures 3 and 4. The average energy of the photon spectrum 

is 0.41 MeV. The spectrum below 30 keV is insignificant. The characteristic x-ray photons also 

produce their own Compton scattered photons, visible as elevated fluences underlying the 

characteristic x-rays. Importantly the assembled weapon components were encased in a metallic 

cladding that significantly attenuates the photon, particularly lower energy, radiation. Also, much of 

the interior surfaces of the buildings where nuclear components are handled or stored are concrete. 

Most elements comprising ordinary concrete have a low atomic number. Oxygen (Z=8) is 50% of 

concrete by weight and silicon (Z=14) is 32%. Higher-energy photons would scatter within this facility 

losing energy in each collision and resulting in photons of lower energy. No comprehensive data have 

been located describing measured photon energy spectra in IAAP workplaces. Photon radiation 

could have been readily measured at IAAP, with available dosimeter technology, during all years of 

operation. 

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 2. MCNP calculated photon spectra emitted from 234mPa beta in 238U 

spheres. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 19 of 58 

238U Metal Bremsstrahlung Photon Energy Spectrum 

0 

0.000005 

0.00001 

0.000015 

0.00002 

0.000025 

0.00003 

MeV/pe 

0 0.5 1 1.5 2 2.5 

Energy (MeV) 

Figure 3. Bremsstrahlung component of the calculated spectrum on a linear 

vertical axis. 

MCNP Modeled 1-cm dia. 238U Metal Sphere
 

Bremsstrahlung & Characteristic X-Ray Photons
 

MeV/pe 

0.001 

0.0009 

0.0008 

0.0007 

0.0006 

0.0005 

0.0004 

0.0003 

0.0002 

0.0001 

0 

0 0.05 0.1 0.15 0.2 

Energy (MeV) 

Figure 4. Characteristic X-ray component of calculated spectrum on a linear 

vertical axis. 

3.3.2 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. The significance of neutron radiation exposure to IAAP nuclear 

weapon component assembly workers cannot be directly assessed because there are no available 
 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 20 of 58 

measurements using adequate technology during the period of IAAP operation. Mr. Shannan 

considered the neutron dose to IAAP workers from pits to be very low (Fix and Bihl 2003). 

Experience and measured doses at Pantex is used to infer neutron doses to IAAP workers. 

3.3.3 Depleted Uranium 

As noted in Chapter 2, 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 (ORAUT-TKBS-0001 2003). 

3.4 EXTERNAL RADIATION DOSE RECORDS 

A considerable number of R. S. Landauer Company dose reports for IAAP were found in the archival 

records (Landauer 1955 – 1975). The earliest obtained dose records apparently begin with the 

availability of eight Landauer-provided film badges for the week beginning November 14, 1955. There 

is mention of TracerLab dosimeters being used prior to Landauer dosimeter use (Fix and Bihl 2003), 

but no records of TracerLab dosimeters have been located. Based on written notations in some 

Landauer dose reports, the film badges apparently were assigned to selected IAAP workers based on 

work activity in the early years and the reported dose data were reviewed (i.e., hand-written on the 

report by a person using one of the dosimeters, numbered 130-0001 to 130-0008). In general, 

reported doses were noted only in the “Gamma and X-Ray Exposure” column of the dose report. 

Based on the records available for inspection, IAAP 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). Table 6 summarizes examination of available 

IAAP dose records. For the report for the weekly period beginning January 16, 1956, an apparent test 

of the dosimetry service was done by exposing one of the film dosimeters to about 600 mR (probably 

from one of the radiography sources). The reported dose was 560 mR “gamma and x-ray exposure” 

and 80 mR (assumed unit to be mR) open window. The dose reports examined are not complete for 

all periods of IAAP operation, as described in Fix and Bihl (2003) and listed in Table 6. 

Individual IAAP worker dose records were received for analysis. Only records with realistic personal 

and dosimeter badge identifications and non-AEC (i.e., Mason and Hanger) contractor affiliation were 

used. Several thousand acceptable records were identified and the results are presented in Table 7. 

The dose data available for analysis ranged from 1955 (only four positive dose results) through 1974 

(no positive dose in 1975). The data included primarily recorded deep dose and in some cases skin, 

extremity and neutron doses. Table 7 presents the respective lognormal and normal distribution 

parameters for equivalent annual doses at IAAP during the different years. It should be noted that 

prior to 1963 it was necessary to convert a weekly or monthly partial year doses to an equivalent 

annual dose (i.e., multiply the dose for the monitored period to estimate an equivalent annual dose). 

No individual dosimeter dose data has been located for the years prior to 1955, 1956, and 1957. 

There is dose data for 1961 and for 1971 but it was not available for the analyses presented in Table 

7, however this is expected to have little impact on the analysis. The geometric mean and standard 

deviations for these years was calculated from the adjoining years (i.e., mean of previous and 

subsequent year). 
 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 21 of 58 

Table 6. Summary of examination of IAAP worker dose records. 

Period Description Comments 

11/14/55–2/6/56 Weekly availability of 8 film dosimeters 

from Landauer apparently assigned to 

workers, as needed, based on notes on 

dose reports. 

Maximum reported dose was 40 mR; most results were 

zero. Test of dosimetry service done for week beginning 

1/16/56 using exposure of 600 mR (reported dose was 

560 mR “gamma and x-ray” and 80 mR (assumed) “open 

window.” 

5/6/57–7/29/57 Weekly availability of 8 film dosimeters 

apparently assigned to workers based 

on notes on dose reports. 

Maximum reported dose was 30 mR “gamma and x-ray.” 

7/28/57–11/27/58 Routine weekly assignment of film 

dosimeters to 19 workers and one area 

(near hospital X-ray) 

Maximum reported dose was 60 mR. Same workers 

monitored for each weekly exchange. 

6/29/59–9/28/59 Routine weekly assignment of film 

dosimeters to 16 workers 

Maximum reported dose was 95 mR. Same workers 

monitored for each weekly exchange. Pattern in worker 

dose with repeated higher relative dose for certain 

workers for each weekly period. 

3/7/60–6/13/60 Routine weekly assignment of film 

dosimeters to 23 workers. Calendar 

year dose included on routine reports 

(i.e., 3/7/60 report was No. 1) as well as 

total number of dosimeters for year. 

Maximum reported dose was 55 mR. Beginning with the 

6/13/60 exchange period, approximately, the notation “M” 

was used to indicate dose results less than the detection 

level. 

8/62–12/63 Routine biweekly assignment of film 

dosimeters to workers to include some 

area/facility monitoring locations (i.e., 

crane area). 

Exchange cycle is noted on dose report. Start date for 

continuing weekly service is estimated based on 

examination of next available dose report (1/26/63) that 

includes columns for number of badge reports to date, 

missing badges to date, and inception date for 

permanent dose total. 

1/64 - 12/74 Routine 4-week assignment of film 

dosimeters to about 60 workers. 

As above but now includes some reported neutron dose 

and exchange periods of 1 quarter. 

Figure 5 presents trends in the measured collective annual dose for the respective dose components 

summarized in Table 7. The data show a significant peak in measured dose during the early 1970s 

likely associated with IAAP operational activities. 

Another source of IAAP dosimetry data is the annual whole-body external dose statistical reports 

reported to the AEC (using AEC Form 190) from 1962 to 1973 (AEC 1962-1973b). The form provides 

information on the numbers of personnel monitored or not monitored and a distribution of whole-body 

doses received. Information on these forms reported for Mason and Hanger IAAP workers only is 

summarized in Table 8. The data for the operating contractor were selected as being the most 

meaningful to the evaluation of dose to workers. Although the lowest reported dose category on AEC 

Form 190 was 0 - 1 rem, in reality the recorded annual dose for most workers was much less than 1 

rem, as noted in Table 7 and in Figure 5, with many workers with annual recorded doses of zero. 

From the foregoing information and consideration of potential sources of external radiation exposure, 

it appears that IAAP worker whole-body dose was: 

•	 Generally not measured prior to 1955, inferred because monitoring data have not been 

located. 

•	 Relatively low prior to about 1968, as listed in Tables 7 and 8. Routine film dosimeter 

monitoring was begun in 1962 and all doses prior to about 1968 are relatively low in 

comparison with the criteria of 10% of AEC-allowed dose limits of 300 mrem per week prior to 

1960 (i.e., as noted on Landauer dose reports) and 1,25 rem per quarter beginning in about 

1969. The criterion of 10% was commonly used as a decision level to require routine 

monitoring. 
 

 

 

 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 22 of 58 

Table 7. IAAP worker recorded dose data statistics. 

Year 

IAAP deep dose dataa

 Lognormal fit 

Workers with 

reported dose 

Dose, mrem Median 

Mean Maximum Collective dose GSD

1948-54 reserved 

1955 reserved 

1956 reserved 

1957 reserved 

1958 22 548 1,500 12,052 444 1.88 

1959 13 350 446 4,551 344 1.22 

1960 18 366 609 6,591 352 1.33 

1961 (b) 298 1.24 

1962 43 245 364 10,535 243 1.14 

1963 57 229 1,060 13,043 195 1.70 

1964 84 96 292 8,062 87 1.51 

1965 44 139 375 6,123 115 1.83 

1966 74 168 833 12,449 127 2.04 

1967 62 145 620 8,997 112 1.95 

1968 170 135 1,625 22,927 94 2.01 

1969 199 119 1,207 23,685 91 1.80 

1970 308 426 7,750 131,310 174 3.27 

1971 224 331 3,496 75,069 109 4.53 

1972 514 179 4,806 92,108 106 2.25 

1973 332 210 4,350 69,590 113 2.41 

1974 226 236 3,813 53,304 117 2.62 

1975 0 

n.a. data not available for analysis 

a. Doses are “annualized” by multiplying the sum of the measured doses by (50 weeks / 

actual number of weeks monitored). 

b. Statistical parameters calculated as linear interpolation from 1960 and 1962. 

c. Dose data analyzed for 1971 only were based on the 1975 IAAP AEC termination (1965­

74) report. 

0 

10 

20 

30 

40 

50 

60 

70 

80 

1950 1955 1960 1965 1970 1975 

Year 

Collective Dose, rem

Deep 

skin 

Neutron 

Extrem ity 

Figure 5. Trends in dosimeter measured IAAP worker deep, skin, neutron and 

extremity collective dose. 
 

 

 

 

3.5 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 23 of 58 

Table 8. IAAP Mason and Hanger monitored workers and reported whole-body doses (AEC 1962– 

1973b). 

Calendar 

year 

Number not 

monitored 

Total 

monitored 

Percent not 

monitored 0–1 rem 1–2 rem 2–3 rem 3–4 rem 4–5 rem 

1962 1,030 29 97 29 

1963 796 41 95 40 1 

1964 650 36 95 36 

1965 692 35 95 35 

1966 824 62 93 62 

1967 818 61 93 61 

1968 798 131 86 130 1 

1969 812 152 84 152 

1970 803 243 77 225 12 5 1 

1971 825 293 74 261 15 8 8 1 

1972 818 312 72 304 5 3 

1973 635 226 74 220 3 2 1 

•	 Not completely monitored with personnel dosimeters particularly prior to 1963, such that 

recorded IAAP worker career whole body doses are underestimated. 

•	 There was awareness of the potential for extremity doses (Fix and Bihl 2003) and some 

workers have recorded extremity doses in the worker dose data as shown in Figure 5. 

•	 Probably considered a relatively low hazard because of the relatively low photon (and neutron) 

radiation exposure levels to workers in routine nuclear weapon component assembly activities. 

In addition, Tables 7 and 8 indicate that the number and percentage of workers routinely monitored 

with dosimeters generally increased, particularly in the 1970s when higher doses were recorded. The 

1970s likely represented a time of increased work activity on assembly of new weapons and perhaps 

preparation for the transfer of the AEC IAAP activities to Pantex by July 1975. 

DOSIMETER TECHNOLOGY 

The 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., nonpenetrating 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 9 summarizes the 

monitoring technique and exchange frequency for the IAAP dosimeters contracted from a commercial 

service. The history of the type of dosimeter used at IAAP is very similar to Pantex, which assumed 

the IAAP work in 1975. The Minimum Detectable Limit (MDL) of these dosimeters is summarized in 

Table 9 along with the maximum potential missed dose (NIOSH 2002). The MDL is widely used in 

other NIOSH documents and can vary depending on dosimeter type, processing equipment, 

calibration techniques, and procedures. Because of these variations, a review of typical MDLs for 

photon dosimeters was conducted and is documented in (NIOSH 2004b). The typical MDL (maximum 

missed dose) per exchange cycle for photon dose for film dosimeters is 40 mrem. 
 

 

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Table 9. Summary of IAAP beta/photon dosimeters, MDL, exchange frequency and potential missed 

dose. 

Period 

Dosimeter 

MDLb 

(rem) 

Exchange 

frequency 

Max. annual 

missed dose (rem)c Year Period of usea 

1948-55 None None 

1955–1962 Occasional Landauer Film 0.040 Weekly (n=50) 1.0 

1962 Occasional prior to 8/1962 Landauer Film 0.040 2-weeks (n=25) 0.5 

1963–1975 Routinely Landauer Film 0.040 4-weeks (n=13) 0.26 

a. Prior to 1963, dosimeters were not routinely assigned to IAAP workers throughout an entire year. 

b. Estimated MDLs for Landauer film dosimeter in the workplace. 

c. Maximum annual missed dose calculated from OCAS-IG-001 (NIOSH 2002). 

3.6 DOSIMETER PERFORMANCE STUDIES 

3.6.1 Photon Radiation 

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. Certainly, the data illustrate capabilities, in spite of many differences in 

organizations, emulsion types, and dosimeter designs, to reasonably detect and measure the photon 

radiation levels and energies potentially received by IAAP workers. 

In recent years, further studies of early dosimeter performance compared to Hp(10) have been done. 

The IARC conducted a dosimeter intercomparison study to higher energy (i.e., >100 keV) photons of 

10 dosimetry systems commonly used throughout the world (Thierry-Chef et al. 2002). Two of the 

U.S. film dosimeter designs were from the Hanford Site – the two-element design (identified as US-2) 

used from 1944 through 1957 and the multielement design (identified as US-8) used from 1958 

through 1971. These dosimeter designs were commonly used at many AEC laboratories (Hanford, 

Oak Ridge National Laboratory, University of Chicago, etc.) (NIOSH 2004b). The IARC Study 

considered that exposure to dosimeters worn by workers could be characterized as anterior-posterior 

(AP), rotational and isotropic irradiation geometries, or a combination thereof. Dosimeter response to 

selected photon energies was measured using two phantoms to simulate the effect of the worker’s 

body on the measured dosimeter response. The first phantom was the International Standards 

Organization (ISO) water-filled slab phantom, which is used for dosimeter calibration and performance 

testing. The second was an anthropomorphic Alderson Rando Phantom, which is constructed from a 

natural human skeleton cast material that has a tissue equivalent response. The results of IARC 

testing, for U.S. dosimeters only, are listed in Table 10. 

The Hanford Site conducted intercomparison testing of all its historic film dosimeter designs using AP 

(Wilson et al. 1990) and angular (Fix, Gilbert, and Baumgartner 1994) irradiations on an Alderson 

Rando phantom essentially identical to the phantom used in the IARC studies. These studies 

included lower energy (i.e., < 100 keV) photons that are significant in plutonium facilities. Data from 

Wilson et al. (1990) are summarized in Table 11. The dosimeter results for energies greater than 100 

keV are consistent with the IARC results, showing an overestimate of Hp(10) for the two- element 

dosimeter. For energies much less than 100 keV, such as in unshielded plutonium glovebox 

operations, the Hanford two-element dosimeter can underestimate the photon dose. However, this 

situation would not occur at IAAP with sealed nuclear weapon components. 
 

 

 

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Table 10. IARC testing results for U.S. beta/photon dosimeters (Thierry-Chef et al 2002). 

Geometry Phantom 

118 keV 208 keV 662 keV 

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

US-2 (Hanford two-element film dosimeter) 

AP Slab 3.0 2.1 1.3 1 1.0 0.8 

AP Anthropomorphic 3.0 4.2 1.2 1.9 1.0 1.8 

Rotational Anthropomorphic 2.2 2 1.4 3 1.2 3.2 

Isotropic Anthropomorphic 1.5 4.4 1.1 1.6 1.0 2.7 

US-8 (Hanford multielement film dosimeter) 

AP Slab 1.0 1.5 1.0 0.8 0.8 1.7 

AP Anthropomorphic 0.8 9.5 0.9 6 0.8 1.8 

Rotational Anthropomorphic 1.2 1.9 1.2 17 1.1 1.8 

Isotropic Anthropomorphic 1.0 3 1.2 9 1.0 2.3 

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

b. Ratio of mean to standard deviation. 

Table 11. Testing results for Hanford two-element and multielement film dosimeters for energy and 

angular response.a,b 

Photon AP exposure Rotational exposure 

beam Two-element 1944–56 Multielement 1957–71 Two-element 1944–56 Multielement 1957–71 

16b

 0.1 0.9 

59b

 0.5 1.1 

M150(70) 0.7 0.70 1.31 1.31 

H150(120) 1.6 0.64 3.00 1.20 

137Cs(662) 1.0 1.0 1.46 1.46 

a. From Table 6-5 in the Hanford Technical Basis Document (ORAU 2003a.) 

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

c. Based on Wilson et al (1990). 

3.6.2 Neutron Radiation 

A few neutron doses were reported by Landauer for some IAAP workers on dose reports beginning in 

about January 1964. Eastman Kodak nuclear emulsion type A (NTA) film probably was used for 

these measurements, similar to the NTA dosimetry by Landauer used at Pantex beginning in 1958 

(ORAU 2004a). 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). It is likely that NTA neutron 

dose measurements at IAAP were highly uncertain for all years and certainly not complete prior to 

1964. One of the most important parameters related to performance of NTA is the difference between 

calibration and workplace neutron energy spectra. There are no measurements of neutron spectra at 

IAAP, and the method(s) used to calibrate the Landauer NTA film known is unknown. Neutron 

spectra have been measured 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. 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 and PICs. 

3.7 IAAP WORKER EXTERNAL DOSE RECONSTRUCTION 

The primary objective of dose reconstruction for IAAP workers is to utilize a claimant-favorable 

method to retrospectively calculate: 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 26 of 58 

•	 Dose to unmonitored workers prior to the routine use of personnel dosimeters. 

•	 Missed dose for monitored workers for low-dose results less-than minimum detection level 

(MDL) of the personnel dosimeter. 

•	 Unmeasured neutron dose to monitored workers. 

•	 Unmonitored and missed skin dose to monitored workers. 

3.7.1 Photon Dose 

The available IAAP worker dose data from 1955 through 1975 were analyzed as providing the best 

option to estimate photon dose for years for which routine personnel dosimetry data are not available. 

Figure 6 presents a lognormal distribution plot of the recorded annual deep (photon) dose for 

measured dose for IAAP workers presented in Table 7. The year to year variability is significant with 

much of the recorded dose occurring in the 1970s. As such, the analysis of photon dose is done for 

each year as described in the following sections for unmonitored and monitored workers. 

1965-1975 IAAP Deep Doses 

0.01 

0.1 

1 

10 

100 

1000 

10000 

100000 

-4 -3 -2 -1 0 1 2 3 4 

Standard Normal Deviate (z) 

Deep Dose (mrem) 

Figure 6. Lognormal probability plot of measured annual IAAP worker deep 

(photon) dose (update for more years of data). 

Unmonitored workers 

Estimates of the dose at IAAP to unmonitored workers prior to 1962 when routine monitoring was 

implemented is based on the measured doses received by the monitored workers summarized in 

Table 7. It is assumed that unmonitored (i.e., non-radiation) workers did not and would not receive a 

significant dose compared the monitored workers. Therefore, assigning a photon dose distribution for 
 

 

 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 27 of 58 

each year as noted in Table 7 based on the dose received by monitored workers would certainly 

assure a claimant-favorable estimate of any unmonitored worker dose. For years when there was no 

recorded dose, a value was estimated from the adjacent years. For years, prior to 1955, the 

recommended dose to be assigned is based on 1958 as representing the year when routine 

assignment of dosimeters to identified workers, although still not complete throughout the year, was 

begun. Unmonitored workers would not be expected to have received significant neutron dose as 

discussed in section 3.7.2. 

Monitored workers 

The DOE reported dose for monitored workers should be adjusted for any missed photon dose using 

estimates from other doses for the same person doing similar work for different time periods (Watson 

et al. 1994) or calculated by multiplying the MDL by the number of zero dose results and dividing by 2 

to estimate the mean potential missed dose (NIOSH 2002) as shown in Table 12. 

Table 12. Missed photon dose adjustments to recorded deep dose. 

Period 

Dosimeter 

MDLb 

(rem) 

Exchange 

frequency 

Mean annual 

missed dose (rem)c Year Period of usea 

1948-55 None (d) 

1955–1962 Occasional Landauer Film 0.040 Weekly (n=50) 1.0 

1962 Occasional prior to 8/1962 Landauer Film 0.040 2-weeks (n=25) 0.50 

1963–1975 Routinely Landauer Film 0.040 4-weeks (n=13) 0.26 

a. Prior to 1963, dosimeters were not routinely assigned to IAAP workers throughout an entire year. 

b. Estimated MDLs for Landauer film dosimeter in the workplace. 

c. Mean annual missed dose calculated from NIOSH (2002). 

d. Estimate of dose for unmonitored workers based on Table 7. 

3.7.2 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 for Pantex workers using measured 

Pantex worker doses during the period of 1993 through 2003. These measurements were made with 

the performance validated Pantex 809/812 dosimetry system. Dose records were analyzed for each 

Pantex worker with a positive neutron dose greater than 50 mrem for the period of 809/812 use. 

Analysis of this information is shown in Figure 7 as a lognormal probability plot of the ratio of neutronto-photon doses. A Finney (1971) analysis, used to reduce the effect of outliers at the extremes, 

yields a geometric mean of 0.81 and a geometric standard deviation (GSD) of 1.51. The measured 

Pantex neutron and photon doses from 1993 through 2003 are expected to be associated with 

plutonium based nuclear weapon components. The neutron-to-photon ratio for the older HEU based 

nuclear weapon components expected to have been handled at IAAP is expected to be similar. 

However, it should be noted that the photon dose from HEU based components is low and this may 

explain why the IAAP worker doses were considered to be low by IAAP safety professionals (Fix and 

Bihl 2003) and the occasional film badge results were also low. Assigning the distribution of 

measured dose from Table 7, corrected for the missed dose, 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 worker measurements is expected to be claimant-favorable. 

Neutron radiation can be created with (alpha, neutron) interactions; however, for several reasons this 

is not considered to be a significant source of exposure. A conservative analysis of the potential 

neutron dose relative to the photon dose, could be based on neutron fields measured at Hanford and 

Savannah River sites associated with workers who actually handled plutonium [plutonium fluoride 

(alpha,neutron)] interactions. The analysis of the lognormal distribution of Hanford plutonium worker 

neutron-to-photon ratio in Figure 8 showed a geometric mean of 0.73, a geometric standard deviation 

of 2.10, and an upper 95th percentile of 2.47. Considering the uncertainty in the measured photon 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 28 of 58 

Pantex Neutron and Gamma (Gamma >= 50 mrem) 

10 

Data 

Ratio (neutron/gamma)

Finney

1 

0.1 

-3 -2 -1 0 1 2 3 

Standard Normal Deviate, z 

Figure 7. Lognormal plot of Pantex worker positive neutron 809/812 

dosimeter data. 

10

Neutron / Photon Dose Ratio 

1 

0.1 

-3 -2 -1 0 1 2 3 

Standard Normal Deviate (z) 

Figure 8. Lognormal probability plot of Hanford plutonium facility worker 

measured neutron-to-photon dose ratio. 
 

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dose at IAAP and the retrospective analysis of early doses, it is recommended to use this distribution 

of neutron-to-photon ratios of 2.5 (i.e., 2.47 rounded to 2.5) to estimate neutron doses to IAAP 

workers prior to 1960. 

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 13. 

Table 13. Statistical parameters for lognormal probability 

distribution for neutron-to-photon dose ratios. 

Parameter 

Neutron-to-photon dose ratio 

Pre-1960 1960–1975 

Geometric Mean 0.7 0.8 

Geometric Standard Deviation 2.1 1.5 

Upper 95% percentile 2.5 1.6 

This estimate of the neutron dose must be adjusted to include the conversion to the ICRP Publication 

60 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 14 (ICRP 

1994). The approach can be simplified using the following expressions: 

Neutron dose = adjusted photon dose × upper 95% percentile neutron/photon dose ratio × ICRP 60 CF 

Table 14. 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 Neutron exposure associated with weapon assembly and disassembly activities. 

component assembly 0.1 - 2 MeV 100 1.91 

3.7.3 Skin and Extremity Dose (Reserved) 

3.7.4 IAAP Facilities 

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 15 lists such facilities. 

3.7.5 IAAP Nuclear Workers 

An historical summary of IAAP workers associated with nuclear weapons assembly activities or with 

industrial radiography activities has not been located. Some information from archival records from 

1949 through 1973 for workers associated with Division B nuclear weapon assembly activities is listed 

in Table 16 along with the total number of workers monitored based on IAAP reports to the AEC 

(Form 190) from 1962 to 1975 (AEC 1962-1973b). These sources of data are generally consistent, 

although the exact number of workers in Line 1 or Division B activities is not fully known and not all of 

these workers were monitored as shown in the AEC Form 190 information. 

3.7.6 Summary of Dose Reconstruction Recommendations 

Substantial information necessary to evaluate worker exposure primarily during the early years is not 

available. Therefore, the following claimant-favorable assumptions have been made: 
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Table 15. IAAP facilities handling radiation. 

Building 

number Description Use Comments 

1-11 Vault storage for 

components (pits) 

Received and unloaded pits, some 

assembly. 

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 midto-late 1960s. Radiation swipe sampling, 1972-1975. 

1-13 Assembly area, U-235 pits Assembly of larger nonplutonium weapons 

(U-235) detonators and covers. Radiation 

swipe data - 1974 for bays A-G. 

1-18 Research and Development References to disposing of radioactive waste 

(1971) 

1-19 Assembly bays “Experimental Building” Tritium monitors 

1-61 Assembly area, U-235 pit 

assembly 

Nonplutonium-bearing weapons 

(i.e., uranium) 1950-1953 (dates not 

confirmed) 

Tritium monitors 

1-63 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 after 1957 as area to bring together 

explosives and plutonium physics package 

(prior to 1957 fissionable material inserted in 

flight). Tritium monitors. 

1-63-7 Assembly cell Gravel Gerty 

1-64 Storage Including pit storage 

1-65 

1-66-1 

1-66-2 

Radiation swipe data, 1972-73 

1-67 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. 

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 

1-100-1 X-Ray X-ray facility and film development 

center 

1-100-2 X-Ray X-ray built for Linitron accelerators Built but not used 

•	 The period of IAAP worker handling of nuclear weapon components (baratols) started in 1948 

even though significant activity probably did not occur until the early 1950s. The actual date of 

the start of nuclear weapons assembly has not been discovered and interviews with several 

retired plant workers lead to disparate dates. 

•	 Utilize options listed in Table 17 to reconstruct photon, neutron, skin (Reserved) and 

skin+extremity doses (Reserved) using the DOE reported deep, skin, neutron and/or extremity 

dose. For all years, the highest value for each dose parameter and for each year should be 

used. 

•	 Because of the overall uncertainty, a claimant-favorable assumption is recommended to 

assume that 100% of the neutron dose to the worker results from 0.1 to 2 MeV neutrons based 

on measurements at Pantex. These assumptions will provide the most claimant-favorable 

estimates of organ dose and probability of causation for most types of cancer. 

•	 Because of the overall uncertainty, a claimant-favorable assumption is recommended to 

assume that 100% of the photon dose to the worker results from 30- to 250-keV photon 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

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Table 16. IAAP total manpower and Division B workers, 

and AEC Form 190 reported nuclear and monitored 

workers. 

Calendar 

year 

IAAP archive dataa IAAP AEC Form 190 

Manpower Division B Total Monitored 

1948 0 

1949 0 

1950 0 

1951 4,879 0 

1952 6,739 1,535 0 

1953 4,603 0 

1954 2,150 0 

1955 2,020 0 

1956 1,803 0 

1957 1,621 0 

1958 1,945 0 

1959 2,019 0 

1960 1,951 0 

1961 2,012 0 

1962 2,137 1,030 29 

1963 2,173 796 41 

1964 1,276 920 650 36 

1965 1,268 979 692 35 

1966 2,629 1,081 824 62 

1967 6,066 818 61 

1968 6,278 798 131 

1969 6,218 812 152 

1970 4,014 803 243 

1971 3,330 825 293 

1972 3,170 818 312 

1973 1,807 635 226 

1974 1,462 

1975 1,255 

IAAP nuclear assembly activities transferred to Pantex 

a. Information collected by the University of Iowa. 

Table 17. Summary of IAAP worker claimant-favorable dose reconstruction options. 

Dose Period Monitored worker Unmonitored worker 

Photon dose <1958 Reserved Reserved 

1958-75 Fig. 6, Tables 7, 12 Table 7 

Neutron dose <1958 Reserved Reserved 

<1961 Fig. 7, Table 13 n.a. 

1961-75 Fig. 8, Table 13 n.a. 

Skin dose (Reserved) <1961 Reserved n.a. 

Skin + extremity dose (Reserved) <1961 Reserved n.a. 

n.a. normally not applicable 

radiation. The exposure to dose conversion factors should be used. These assumptions will 

provide the most claimant-favorable estimates of organ dose and probability of causation for 

most types of cancers. 

The methods to reconstruct dose are presented in Attachment B. Using the above claimant-favorable 

assumptions for retrospective calculation of IAAP workers dose, there is no need to correct further the 
 

 

 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 32 of 58 

recorded dose for underestimates in film dosimeter response because of lower energy photon 

radiation, which probably was not significant based on the sealed (and therefore shielded) nuclear 

components and the available nonpenetrating dosimeter data that implies a relatively low lower 

energy component (i.e., many nonpenetrating doses shown as zero). 

3.7.7 Radiation Dose Fraction 

Table 18 summarizes the recommended fractions for IAAP dose according to the energy categories 

required by IREP. For the photon dose, 100% is assigned to the 30 to 250-keV category and for the 

neutron dose, 100% is attributed to the 0.1- to 2-MeV category. This will provide claimant-favorable 

analysis of the organ dose and probability of causation for most cancers. 

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

exposures. 

Process/ 

buildings Description 

Operations Radiation 

type 

Energy 

Begin End selection %

Line 1 or Division B 

(see Table 15 

and Att. A) 

Assembly and disassembly of nuclear weapon (i.e., 

depleted uranium, enriched uranium, plutonium) 

components. 

Parameters to estimate dose to “whole body” organs 1949 1975 Beta > 15 keV 100a 

Photon 30–250 keV 100b 

Neutron 0.1-2 MeV 100c 

Radiography 

(see Table 15 

and Att. A) 

Industrial radiography 

Parameters to estimate dose to whole-body organs 1948 1975 Beta > 15 100 

Photon 30–250 keV 25 

> 250 keV 75 

a. Beta particles from depleted or enriched uranium are greater than 15 keV. 

b. Most photons from depleted uranium are greater than 30 keV; some are greater than 250 keV unless shielding materials are present, 

assuming that 100% of the photons from depleted uranium are between 30 and 250 keV is claimant-favorable. 

c. The neutron energy region of 0.1 – 2 MeV was selected to be claimant-favorable for whole-body dose to provide the highest calculated 

organ dose. 

3.8 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 30 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). The estimated standard error in recorded film 

badge doses from photons of any energy is ±30%. Estimation of the measured neutron dose is not 

nearly as precise as for photons. With NTA films, the estimated standard error is much larger and 

varied significantly with the energy of the neutrons. 

For the calculated photon and neutron dose assigned to IAAP workers, the estimated standard error 

for the assigned photon and neutron dose is ±30% and ±50%, respectively. 

3.9 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. Given the many uncertainties, especially with 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 33 of 58 

film badge dosimetry in the 1950s to the 1970s, a claimant-favorable approach is used to estimate 

organ dose. The exposure to organ dose conversion factors shown in Table 19 results in a higher 

organ dose and higher probability of causation, given the Radiation Effectiveness Factors (REFs) of 

the intermediate energy photons. As such, these dose conversion factors (DCFs) are used to convert 

recorded film badge gamma (photon) doses to organ dose. For neutrons, the deep dose to organ 

dose conversion factors are shown in Table 20. In the conversion of all photon and neutron doses of 

record to organ doses, the exposure geometry must be given careful consideration. The anteriorposterior exposure geometry is used to estimate the organ dose to conduct an initial screening. 

Table 19. Annual photon exposure to organ dose conversion 

factors.a 

Organ 

Exposure to organ dose factors 

(A-P geometry) 

< 30 keV 30-250 keV >250 keV 

Bladder 0.175 1.244 0.883 

Bone (red marrow) 0.025 0.626 0.720 

Bone (surface) 0.209 1.229 0.764 

Breast (female) 0.561 1.266 0.930 

Colon 0.075 1.060 0.844 

Esophagus 0.014 0.688 0.745 

Eye 0.936 1.236 0.880 

Gonads (female-ovaries) 0.047 0.955 0.819 

Gonads (male-testes) 0.622 1.434 0.941 

Liver 0.106 1.064 0.845 

Lung 0.100 0.986 0.842 

Remainder organs 0.071 0.879 0.787 

Skin 0.504 0.892 0.835 

Stomach 0.182 1.251 0.885 

Thymus 0.288 1.408 0.892 

Thyroid 0.473 1.440 0.972 

Uterus 0.061 1.011 0.786 

a. OCAS-IG-001, Rev 1(NIOSH 2002). 

Table 20. Annual neutron deep dose equivalent to organ dose 

conversion factors.a 

Organ 

Deep dose equivalent to organ dose factors 

(A-P geometry) 

10-100 keV 0.1-2 MeV 2-20 MeV 

Bladder 1.268 0.796 1.105 

Bone (red marrow) 0.651 0.361 0.720 

Bone (surface) 0.656 0.436 0.675 

Breast (female) 1.111 1.145 1.121 

Colon 0.947 0.490 0.912 

Esophagus 0.775 0.412 0.869 

Gonads (female-ovaries) 0.935 0.424 0.903 

Gonads (male-testes) 1.466 1.307 1.222 

Liver 0.983 0.641 0.990 

Lung 0.737 0.557 0.950 

Remainder organs 0.819 0.525 0.889 

Skin 0.986 0.853 0.918 

Stomach 1.221 0.824 1.099 

Thyroid 1.066 1.086 1.123 

a. OCAS-IG-001, Rev 1(NIOSH 2002) 
 

4.0 

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Some of the more common exposure geometries encountered in the workplace are defined as 

follows: 

•	 An AP exposure is typical for an individual who works in a directional radiation field and faces 

the source of the radiation source, such as a nuclear weapon component, while working. 

•	 A rotational (ROT) exposure is typical of an individual who is constantly turning in a directional 

radiation field, such as when conducting inventories in the nuclear weapon storage vaults, 

while working. 

•	 An isotropic (ISO) exposure is typical of a worker involved in activities involving a highly 

nondirectional or omnidirectional radiation field. An example of a work facility with an 

omnidirectional radiation field that leads to ISO irradiation of a worker might be maintenance 

activities where scattered neutrons and photon radiation are incident on the worker from all 

directions. 

The proposed initial screening option to identify likely noncompensable cases based on claimantfavorable organ dose estimates for long-term workers is to use the organ dose factors for an anteriorposterior exposure geometry as shown in Tables 19 and 20. Claims that require a more realistic 

assessment to determine compensability should consider the geometries mentioned above. NIOSH 

(2002) appendix B provides dose conversion factors to convert IAAP worker photon and neutron dose 

to the primary organ dose for many selections of exposure geometry, target organ and radiation 

quantities. 

OCCUPATIONALLY RELATED MEDICAL X-RAYS 

Medical examinations at the IAAP were similar to those at Pantex. A chest X-ray was given to each 

IAAP employee annually. There is also evidence that male employees in certain job categories 

(heavy lifters, perhaps) received lumbar spine examinations, the frequency of which was not 

available. At Pantex lumbar spine examinations were given to men when they were hired. Because 

the operating contractor was the same at Pantex and the IAAP AEC facilities, the same protocol was 

assumed for lumbar spine X-rays at IAAP. 

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

Related Diagnostic X-ray Procedures (NIOSH 2003). No information concerning X-ray machine 

parameters pertinent to dose received by the workers was uncovered. Therefore, default values 

provided in the NIOSH document (2003) should be used for chest X-rays (Tables 3.3-1 and 4.0-1). 

Assume an annual posterior-anterior (PA) chest X-ray for all employees applicable from 1947 through 

1975. Do not apply X-ray dose for years other than 1947 through 1975, regardless if the worker was 

employed at IAAP during other years. 

NIOSH (2003) 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 PA 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 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 35 of 58 

organ doses and geometric standard deviations listed in Table 21, varying the HVL from 1.5 to 3.0 

mm AL. NIOSH (2003) 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 but arbitrarily added by proximity are listed in Table 21 after the ICRP 34 organ. For skin. 

Table 21. Organ doses from lumbar spine X-rays at IAAP. 

Organ View 

Organ dose 

geometric mean 

(GSD) (mrem) 

Thyroid, eye, brain PA 0.50 (3.0) 

LAT 0.10 (1.9) 

Ovaries, liver, gall bladder, stomach, intestines, 

colon, rectum, kidneys, adrenals, pancreas, spleen 

PA 330 (2.6) 

LAT 230 (2.1) 

Lungs, thymus, esophagus, bone surfaces PA 120 (2.5) 

LAT 64 (2.0) 

Active bone marrow PA 58 (2.6) 

LAT 110 (2.1) 

Testesa PA 26 (3.1) 

LAT 40 (1.5) 

Skina PA 1,800 (2.4) 

LAT 5,800 (1.8) 

a. Calculated from information in Lincoln and Gupton (1958) as presented in the Rocky 

Flats site TBD (ORAU 2004b). 

and testes, consistent with the approach taken in the Rocky Flats Occupational Medical Dose TBD, 

the values measured in the Lincoln and Gupton paper were used directly to account for the difference 

in collimation in the Lincoln and Gupton measurements versus ICRP 34 factors. 

The PA 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. 

5.0 OCCUPATIONAL ENVIRONMENTAL DOSE 

5.1 INTRODUCTION 

The occupational environmental dose refers to the dose received by workers on the site but outside 

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. 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. 

5.2 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. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 36 of 58 

5.2.1 Intakes from Tritium Releases 

As indicated in Section 2, there were no records of atmospheric releases at IAAP. 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 claimantfavorable, 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: 

X(pCi/m3

) = f * Q(pCi/sec) / V(m3/sec) 

where X is the annual average upper-bound air concentration, 

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

to be 0.25, 

Q 

V 

is the release rate of the radionuclide from the source, and 

is the volumetric flow rate of the vent. The default value is 0.3 m3

/sec, typical 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 Χ 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. 

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. 
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5.2.2 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 DU-contaminated 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 

C = f Q P/ U 

where C is the annual average air concentration, g/m3 

, 

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

0.25, 

Q is the release rate, g/sec, and 

U is the average wind speed, assumed to be 2 m/sec. 

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 × 10-3, 2.0 × 10-4, and 5 × 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 × 10-9 g/m3

; and 

at 1,000 m C is about 1.7 × 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. 

Assuming a breathing rate of 2400 m3

/yr, the amount inhaled at a location in or near the C Yard would 

be: 

6.9 × 10-9 g/m3

 * 2,400 m3/yr = 1.7 × 10-5 g/yr or 6.3 pCi/yr DU. 

Similarly, for locations at or beyond Line 1: 

1.7 × 10-9 g/m3

 * 2,400 m3/yr = 4.1 × 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. 
 

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5.3 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, the Explosive Disposal 

Area, and the Firing Site. It is likely that workers handling, transporting, or storing weapons and 

weapon parts were monitored. Radiation workers accounted for a small fraction of the workers on the 

site. Employees working in other areas of the site were not monitored and there was a 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 

at least 1955 through1975, film badges were used for radiation monitoring of personnel and fixed 

operating areas. Before 1966, surveys of ambient radiation levels were probably measured with 

pressurized ionization chambers and Geiger-Mueller detectors. However, because no records of 

these monitoring activities were found, direct determination of external dose is not possible. Data 

from the Pantex Plant (ORAU 2004c) indicate that, outside of the nuclear weapons areas, there was 

no external radiation dose measured by monitoring devices above background anywhere on the 

Pantex site. Because the same operations were performed at IAAP, it was concluded that no external 

environmental dose of significance to dose reconstruction would have been received. Therefore, no 

environmental external dose should be assigned to unmonitored workers outside facilities at IAAP. 

5.4 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. 

6.0 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 22. Pantex radon 

measurements and dose calculations were assumed to be the best indicators of radon exposure at 

IAAP, and were used as discussed below. 
7.0 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 39 of 58 

Table 22. 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 22, the geometric mean (median) for all buildings at Pantex was1.4 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.” 

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. 

YEARLY INTAKES 

Attachment C, Table C-1 summarizes intakes by workers from both workplace tasks (Section 2), 

environmental releases (Section 5) and radon (Section 6). 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 40 of 58 

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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 (β) 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. 

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 radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter). 

dosimetry 

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 45 of 58 

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. 

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 light-tight 

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 (γ) 

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays but with higher energy; the only essential difference is that X-rays 

do not originate in the nucleus. 

Gerty 

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 Gerty. 

Gray 

SI unit of absorbed dose. Unit symbol, Gy. 1 Gy = 100 rad. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 46 of 58 

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) 

Al 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. 

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. 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 47 of 58 

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. 

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. 

radioactivity 

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei. 

rem 

A special unit of dose equivalent, which is equal to the product of the number of rad absorbed 

and the "quality factor." 

roentgen (R) 

A unit of exposure to gamma (or x-ray) radiation. It is defined precisely as the quantity of 

gamma (or x) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 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. 
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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. 

X-ray 

Ionizing electromagnetic radiation that originates external to the nucleus of an atom. 
 

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ATTACHMENT A 

POSITION TITLES AND TASKS 

The IAAP was principally an Army ordnance facility with only a fraction of the activities being 

performed under the direction of or on behalf of the AEC. The EEOICPA applies to the AEC work 

only. However, at the time of this writing, the OCAS position is that a claim in NOCTS has passed 

screening and the dose reconstructor should consider the claimant eligible. 

However, not every worker doing AEC work at IAAP had the same potential for intakes. Table 17 

provides radiation energy assumptions based on tasks, and Table C-1 relates assumed intakes to 

tasks. The information below provides an attempt to correlate position titles in NOCTS claims folders 

to the job tasks provided in Tables 17 and C-1. The correlations below should be considered 

estimates only. If possible, additional information in a worker’s file should be used to help determine if 

an individual worker did the tasks listed in Tables 17 or C-1. Table 14 may also be helpful toward 

determining potential for exposure in the workplace. 

All personnel: any claimant in NOCTS 

Machinist: machinist; foreman, repairman, machinist; set-up machinist, maintenance; foreman 

machine shop; machinist helper; press operator 

Assembly/disassembly: production operator line 1; foreman line 1, assembler; assembly; 

assembly and inspection; assembly line; assembly worker; dismantled nuclear weapons; inspector 

line 1; line 1 assembly production worker; line 1 worked underground; warhead assembly plant; 

tore down hydrogen and atom bombs; laborer in production; laborer line 1; load lines line 1; top 

secret underground; safety engineer; safety inspector; safety technician or anything implying 

radiation safety worker 

Hydroshot operations: driver; firing sites; test fire technician; test site supervisor; test fire 

specialist, truck driver 

Hydroshot cleanup: firing sites; test fire technician; clean-up crew; test fire specialist 

Burn yard crew: burning field; supervisor assembly burn yard 

Radiography: x-ray technician; metallurgy 

Positions with no exposure in the categories above: carpenter, maintenance, component 

operator, component inspector, electrician, electrical worker, explosive operator, ironworker, insulator, 

matron (in charge of the clothes changing stations), stoker, melt operator, millwright, pipefitter, 

administrative positions. 

Many job descriptions are vague in relation to the specific tasks listed above; examples are given 

below. Hopefully, interview information or other information in the claimant’s folder can help with 

assigning workers to tasks. 

Position descriptions needing more information: Line 9; engineer, engineering aid, engineering 

assistant, engineering tech., equipment operator, foreman, IEP operator, inspector, line inspector, 

machine operator, operating engineer, pipefitter, production worker, production operator, production, 

production control, production foreman, PS-10 cleaning line. 
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ATTACHMENT B 

OCCUPATIONAL EXTERNAL DOSE FOR IAAP WORKERS 

TABLE OF CONTENTS
 

Section	 Page 

B1.0 DOE IAAP Dose Records .........................................................................................................50
 

B2.0 Dose Parameters......................................................................................................................51
 

B2.1 Years of Exposure ..............................................................................................................52
 

B2.2 Rate ....................................................................................................................................52
 

B2.3 Radiation Type ....................................................................................................................52
 

B2.4 Radiation Energy and Dose Fraction ..................................................................................52
 

B3.0 Distribution Parameters ............................................................................................................53
 

B3.1 Type ....................................................................................................................................53
 

B3.2 Parameter 1 ........................................................................................................................53
 

B3.3 Parameter 2 ........................................................................................................................56
 

LIST OF TABLES 

Table Page
 

B-1 IREP dose parameter input screen..............................................................................................51
 

B-2 Selection of radiation type, energies and percentages 

B-5 Statistical parameters for lognormal probability distribution for neutron-to-photon 

................................................................52
 

B-3 Distribution parameters for IAAP worker measured photon dose ................................................53
 

B-4 Missed photon dose adjustments to DOE reported gamma dose ...............................................54
 

dose ratios ...................................................................................................................................54
 

B-6 IAAP neutron dose fractions, energies, percentages, and associated ICRP 60 

correction factors .........................................................................................................................54
 

B-7 Annual photon exposure to organ dose conversion factors.........................................................55
 

B-8 Annual neutron deep dose equivalent to organ dose conversion factors ....................................55
 

B1.0 DOE IAAP DOSE RECORDS 

U.S. Department of Energy (DOE)-provided dose of record information for IAAP energy employees 

might be incomplete for the following reasons: 

No records of personnel dosimeter assignment have been located prior to 1955 

•	 Only samples of workers were assigned dosimeters, and then only temporarily, prior to August 

1962 when dosimeters were apparently first routinely assigned to workers for each succeeding 

year. 

•	 Neutron dosimeters were not generally assigned, and the NTA dosimeter used beginning in 

about 1964 likely did not provide a complete and accurate measurement of neutron dose to 

workers in IAAP facilities. 
 

 

 

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•	 The nonpenetrating dose is often recorded as a blank or zero on the observed dose records 

when the skin (i.e., nonpenetrating) dose should be equal to or greater than the penetrating 

dose. 

The DOE provided dose of record information generally assigns a skin (nonpenetrating, beta, shallow, 

etc.) and whole-body (penetrating, gamma, deep, etc.) dose for each year of record. The recorded 

penetrating photon doses for 1963 through 1975 are expected to reasonably estimate the actual dose 

at exposure levels significantly greater than the minimum detection level (~0.04 R) of the dosimeter. 

IAAP nuclear weapon assembly activities are assumed to have first occurred in 1948 and continued 

until July 1975 when this work was transferred to the Pantex facility. Estimates of IAAP neutron, beta 

(Reserved) and beta plus extremity (Reserved) doses are based on using lognormal probability 

distributions of the respective neutron-to-photon, beta-to-photon and beta plus extremity-to-photon 

dose ratios based on measured IAAP individual worker doses from 1965 through 1975 and measured 

Pantex individual worker dose during period of 1993 through 2003 using improved dosimeter 

technology and field validation. The Pantex information is expected to be applicable to IAAP. 

IAAP workers involved in nuclear weapons assembly and radiography activities were identified as 

Line 1 or Division B. The IAAP facilities with potential worker exposure are presented in Table 15. 

B2.0 DOSE PARAMETERS 

Several technical parameters are considered in the evaluation of the DOE-provided dose of record 

information for each claim. The focus of this attachment is to identify options for a claimant-favorable 

analysis of the dose to be assigned to IAAP workers for each year of employment. The recorded 

doses are often based on less capable technology than currently available and there is often 

uncertainty in historical radiation monitoring practices. A basis of comparison for a consistent 

evaluation of the DOE-provided recorded dose through time and among different facilities is the 

Personal Dose Equivalent, Hp(d), where d identifies the depth (in mm) and represents the point of 

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

mm and is noted as Hp(0.07). For penetrating radiation of significance to whole-body dose, d = 10 

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

the operational quantity to be recorded for radiological protection purposes by the International 

Commission on Radiological Units and Measurements (ICRU). These are the dose quantities used to 

accredit DOE dosimetry programs since the mid-1980s. 

The primary Interactive RadioEpidemiological Program (IREP) screen used to input dose parameters 

is presented in Table B-1. Input to these fields is obtained from the DOE-reported dose of record 

information. The claim provides the primary organ of interest and other worker information needed to 

run IREP. Guidance for the selection of the external dosimetry parameters in Table B-1 by the dose 

reconstruction analyst is presented in the following sections. 

Table B-1. IREP dose parameter input screen. 

Exposure 

Radiation type 

Distribution parameters 

# Year Rate Type 1 2 3 

1 1960 Acute Photon, 30-250 keV Normal 

2 1960 Chronic Neutron, 0.1-2 MeV Normal 

3 .. .. .. .. .. .. 
 

 

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B2.1 YEARS OF EXPOSURE 

The years of exposure during the period from 1948 through 1975 should be identified from 

employment information on the claim and from DOE radiation dose reports. For unmonitored workers 

at any time during IAAP employment, an assigned dose is calculated based on the distribution of 

measured doses for monitored workers. For monitored workers with years with no recorded radiation 

dose, a missed dose, as described later in this section, is calculated for all zero or missing records. 

The early period of IAAP operations which apparently involved assembly of nuclear weapon 

components is uncertain because relevant information such as the number of components handled 

per year is classified. Therefore, a claimant-favorable assumption is that IAAP Line 1 or Division B 

nuclear weapon activities began in 1948 and a deep (photon) dose is assigned based on the 

lognormal probability distribution of measured dose in later years. The photon dose is assigned to 

each worker for every year of exposure from 1948 through 1975 using the methods described in 

Section 3 of this TBD and abbreviated in this attachment. (Reserved) Estimates of skin (beta - 

cancer site on torso) and beta plus extremity (cancer site on extremities) are estimated using 

lognormal probability distributions of the skin-to-photon dose, and skin plus extremity-to-photon dose 

ratios obtained from analyses of measured doses for IAAP and Pantex workers. 

B2.2 RATE 

Acute is selected for all types of external beta and photon dose and chronic is selected for neutron 

dose (NIOSH 2002). 

B2.3 RADIATION TYPE 

IAAP Line 1 or Division B nuclear weapon assembly workers were exposed, potentially, to beta, 

photon (X-rays and gamma rays) and neutron radiation. Nuclear weapon components were sealed 

with a metallic covering that significantly reduces the potential beta and lower energy photon 

exposure to workers. IAAP workers were likely exposed to depleted uranium (DU) particularly in 

picking up DU fragments after conventional detonations. The precise circumstances of individual 

worker exposures to nuclear weapon components are not available. Therefore, claimant-favorable 

assumptions of the radiation type and energies are recommended using the guidance in Table B-2 

based on recent individual worker photon and neutron dose measurements at Pantex. The purpose 

of the values in this table is to identify options to estimate the parameters used in calculating organ 

doses for long-term IAAP workers for time spent in any of the facilities listed in Table 14 that are 

generally identified as Line 1 or Division B. 

Table B-2. Selection of radiation type, energies and percentages. 

Process/ 

buildings Description 

Operations Radiation 

type 

Energy 

Begin End selection %

Line 1 or Division B 

(see Table 15 or 

Att. A) 

Assembly and disassembly of nuclear weapon (i.e., depleted uranium, enriched uranium, plutonium) 

components. 

Parameters to estimate dose to “whole body” 

organs 

1948 1975 Beta > 15 keV 100 

Photon 30–250 keV 100 

Neutron 0.1-2 MeV 100 

Radiography 

(see Table 15 or 

Att. A) 

Industrial radiography 

Parameters to estimate dose to “whole body” 

organs 

1948 1975 Beta > 15 100 

Photon 30–250 keV 25 

> 250 keV 75 
 

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B2.4 RADIATION ENERGY AND DOSE FRACTION 

Table B-2 summarizes the recommended IAAP dose fractions according to the energy categories 

required by the IREP. This will provide claimant-favorable analysis of the organ dose and probability 

of causation for most cancers. 

B3.0 DISTRIBUTION PARAMETERS 

The selection of the distribution parameters in Table B-1 is discussed in the following sections. 

B3.1 TYPE 

The type of distribution is selected. This determines the definition of Parameters 1 and 2. For a 

normal and lognormal distribution, Parameter 3 is not used. 

B3.2 PARAMETER 1 

The assigned photon and neutron dose is determined using the following steps. 

Photon Dose, Unmonitored Workers 

The dose to unmonitored workers at any time during their Table B-3. Distribution 

employment at IAAP is calculated for each year from the parameters for IAAP worker distribution of dose measured for monitored workers. The 

Year 

Lognormal fit 

Median dose GSD 

1948-57 Reserved 

1958e

 444 1.88 

1959e

 344 1.22 

1960e

 352 1.33 

1961 298 1.24 

1962e

 243 1.14 

1963 195 1.70 

1964 87 1.51 

1965 115 1.83 

1966 127 2.04 

1967 112 1.95 

1968 94 2.01 

1969 91 1.80 

1970 174 3.27 

1971 1.09 4.53 

1972 106 2.25 

1973 113 2.41 

1974 117 2.62 

1975a 

measured photon dose. respective geometric mean and geometric standard deviation is 

selected for each year of employment from Table B-3. It is 

assumed that unmonitored (i.e., non-radiation) workers did not 

and would not receive a significant dose compared that measured 

for monitored workers. Assigning a photon dose distribution for 

each year as noted in Table B-3. A neutron dose is not 

calculated for unmonitored workers since it is not expected that 

significant neutron dose would be received. 

Monitored Workers 

Workers are considered to have been monitored if, at any time 

during IAAP employment, a dose is assigned as noted in the DOE 

dose report. The DOE reported photon dose (i.e., whole-body, 

penetrating, or gamma) for IAAP workers will provide a 

reasonable estimate of the Hp(10) dose used as parameter #1 for 

the periods of dosimeter assignment. Missed photon dose for 

IAAP workers will occur particularly prior to 1955 when no 

personnel dosimeters were used and during the period of 1955 

through 1962 when dosimeters were not used throughout the 

work year. As such, a missed dose distribution is estimated for 

each year with parameter #1 calculated from MDL/2 for each a. It is understood there is no dosimeter and exchange period with a zero or less-than (i.e., “m”) reported IAAP dose in 1975. 

dose result using information shown in Table B-4. Table B-4 However, if necessary, use 1974 

summarizes the potential missed photon dose adjustments parameters. 

according to exchange period if all dose results were zero. 
 

 

 

 

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Table B-4. Missed photon dose adjustments to DOE reported gamma dose. 

Period 

Dosimeter 

MDLb 

(rem) 

Exchange 

frequency 

Mean annual 

missed dose (rem)c Year Period of usea 

1948-55 None (d) 

1955–1962 Occasional Landauer Film 0.040 Weekly (n=50) 1.0 

1962 Occasional prior to 8/1962 Landauer Film 0.040 2-weeks (n=25) 0.50 

1963–1975 Routinely Landauer Film 0.040 4-weeks (n=13) 0.26 

a. Prior to 1963, dosimeters were not routinely assigned to IAAP workers throughout an entire year. 

b. Estimated MDLs for Landauer film dosimeter in the workplace. 

c. Mean annual missed dose calculated from NIOSH (2002). 

d. Base estimate for unmonitored workers if there is no recorded dose at any time. 

Neutron Dose 

The proposed claimant-favorable option, for monitored workers only, is separated into distinct time 

periods: 1) pre-1959 (reserved) 2) 1958-60 and 3) 1960-through 1975, respectively, is to utilize the 

distribution parameters shown in Table B-5. 

Table B-5. Statistical parameters for lognormal probability distribution for 

neutron-to-photon dose ratios. 

Neutron-to-photon dose ratio 

Parameter Prior to 1958 1958 - 1960 1960 - 1975 

Geometric Mean Reserved 0.7 0.8 

Geometric Standard Deviation 2.1 1.5 

Upper 95% percentile 2.5 1.6 

The neutron dose must be adjusted to include the conversion to the ICRP Publication 60 neutron 

weighting factor required for input of the dose into Table B-1 using factor identified in Table B-6. The 

approach can be simplified using the following expressions: 

Neutron dose = adjusted photon dose × geometric mean neutron/photon dose ratio × ICRP 60 CF 

Table B-6. IAAP neutron dose fractions, energies, percentages and associated ICRP 60 correction 

factors. 

Process Description/buildings 

Neutron energy 

(MeV) 

Default dose fractiona,b 

(%) 

ICRP 60 correction 

factor 

Nuclear weapon Neutron exposure associated with weapon assembly and disassembly activities. 

component assembly 0.1 1 2 MeV 100 1.91 

Reserved - Beta Dose (skin cancer site on body torso) 

Reserved - Beta Dose (skin cancer site on extremity) 

Organ Dose 

Once the adjusted photon and neutron doses have been calculated for each year with the associated 

standard deviation, the organ dose is calculated for the primary organ of interest identified in the 

claim. For photon radiation, the exposure to dose conversion factors should be used to provide a 

claimant favorable analysis. The proposed initial screening option to identify likely non-compensable 

cases based on claimant-favorable organ dose estimates for long-term workers is to utilize an 

anterior-posterior. 

NIOSH (2002) appendix B provides dose conversion factors to convert IAAP worker photon and 

neutron dose to the primary organ dose for many selections of exposure geometry, target organ and 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 55 of 58 

radiation quantities. Photon and neutron organ dose conversion factors for a claimant-favorable 

analysis of worker anterior-posterior (A-P) exposure are: 

• Photon exposure to organ dose conversion factors in Table B-7. 

• Neutron deep dose to organ dose conversion factors in Table B-8. 

Table B-7. Annual photon exposure to organ dose conversion 

factors.a 

Organ 

Exposure to organ dose factors 

(A-P Geometry) 

< 30 keV 30-250 keV >250 keV 

Bladder 0.175 1.244 0.883 

Bone (red marrow) 0.025 0.626 0.720 

Bone (surface) 0.209 1.229 0.764 

Breast (female) 0.561 1.266 0.930 

Colon 0.075 1.060 0.844 

Esophagus 0.014 0.688 0.745 

Eye 0.936 1.236 0.880 

Gonads (female-ovaries) 0.047 0.955 0.819 

Gonads (male-testes) 0.622 1.434 0.941 

Liver 0.106 1.064 0.845 

Lung 0.100 0.986 0.842 

Remainder organs 0.071 0.879 0.787 

Skin 0.504 0.892 0.835 

Stomach 0.182 1.251 0.885 

Thymus 0.288 1.408 0.892 

Thyroid 0.473 1.440 0.972 

Uterus 0.061 1.011 0.786 

a. OCAS-IG-001, Rev 1(NIOSH 2002) 

Table B-8. Annual neutron deep dose equivalent to organ dose 

conversion factors.a 

Organ 

Deep dose equivalent to organ dose factors 

(A-P Geometry) 

10-100 keV 0.1-2 MeV 2-20 MeV 

Bladder 1.268 0.796 1.105 

Bone (red marrow) 0.651 0.361 0.720 

Bone (surface) 0.656 0.436 0.675 

Breast (female) 1.111 1.145 1.121 

Colon 0.947 0.490 0.912 

Esophagus 0.775 0.412 0.869 

Gonads (female-ovaries) 0.935 0.424 0.903 

Gonads (male-testes) 1.466 1.307 1.222 

Liver 0.983 0.641 0.990 

Lung 0.737 0.557 0.950 

Remainder organs 0.819 0.525 0.889 

Skin 0.986 0.853 0.918 

Stomach 1.221 0.824 1.099 

Thyroid 1.066 1.086 1.123 

Claims that require a more realistic assessment to determine compensability should consider the 

geometries described as follows: 
 

Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 56 of 58 

•	 An AP exposure is typical for an individual who works in a directional radiation field and faces 

the source of the radiation source, such as a nuclear weapon component, while working. 

•	 A rotational (ROT) exposure is typical of an individual who is constantly turning in a directional 

radiation field, such as when conducting inventories in the nuclear weapon storage vaults, 

while working. 

•	 An isotropic (ISO) exposure is typical of a worker involved in activities involving a highly 

nondirectional or omnidirectional radiation field. An example of a work facility with an 

omnidirectional radiation field that leads to ISO irradiation of a worker might be maintenance 

activities where scattered neutrons and photon radiation are incident on the worker from all 

directions. 

B3.3 PARAMETER 2 

Parameter 2 is the standard deviation of the identified distribution for the respective variables involved 

in the calculation of the organ dose. The individual dose result for each dosimeter exchange period 

reported by DOE may be used to calculate the mean and standard deviation for each year. 
 

 

 

 

 

 

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ATTACHMENT C 

SUMMARY OF INTAKES AND DETAILS OF INGESTION CALCULATIONS 

TABLE OF CONTENTS
 

C1.0 Intakes ......................................................................................................................................57
 

C2.0 Ingestion Intake Calculation ......................................................................................................57
 

LIST OF TABLES 

C-1 Summary of intakes by AEC workers at IAAP .............................................................................57
 

C1.0 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 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 Radonb

 Chronic inhalation Lognormal 0.072 

WLM/1 

2 month 

period 

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 0.5 mean meanc 

Burning yard operations 1/1/1948-7/1/1975 DU Chronic inhalation M, S 38 0.10 Lognormal Dose 3 

Weapons disassembly 1/1/1954-7/1/1975 HTO Chronic inhalation/ 

absorption 

2.2E+06 Constant Dose 

Weapons disassembly 1/1/50 – 12/31/1957 DU Acute inhalation at 

start of each year 

M,S 5.7E+04 pCi 150 mg 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 

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 (NIOSH 2004a). Instead an experiment was 

conducted for this TBD, wherein a small amount of ordinary soil, judged by the author to be similar to 
Effective Date: 04/16/2004 Revision No. 00 Document No. ORAUT-TKBS-0018 Page 58 of 58 

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 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)