ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 82

DOE Review Release 05/03/2010

Document Title:

Area IV of the Santa Susana Field Laboratory, the 

Canoga Avenue Facility, the Downey Facility, and 

the De Soto Avenue Facility (sometimes referred 

to as Energy Technology Engineering Center 

[ETEC] or Atomics International) – Occupational 

Internal Dose

Document Number: ORAUT-TKBS-0038-5

Revision: 01

Effective Date: 04/26/2010

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Eugene W. Potter and Thomas R. LaBone

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/16/2010

Melton H. Chew, Document Owner

Concurrence: Signature on File Concurrence Date: 04/16/2010

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 04/16/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 04/19/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/26/2010

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 2 of 82

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/22/2006 00 Technical basis document for Energy Technology Engineering 

Center – Occupational Internal Dose. First approved issue. Initiated 

by Melton H. Chew.

04/26/2010 01 Revision to change the site name to Area IV of the Santa Susana 

Field Laboratory, the Canoga Avenue Facility, the Downey Facility, 

and the De Soto Avenue Facility (sometimes referred to as Energy 

Technology Engineering Center [ETEC] or Atomics International. 

Updated required language in Section 5.1. Additional information 

added on the solubility of uranium aluminide and sensitivity of 

bioassay analyses. Added Attachment G, Sensitivity of Bioassay 

Analyses by Bioassay Code U. S. Testing and Other Laboratories. 

Acronyms added for ICRP, kCi, kg, mo, POC, TTF, wk, μCi, μm, and 

§. Changes are being made to Section 5.1.3, Scope, as requested 

by DOL to clarify which organizations performed work eligible for 

compensation. Reporting levels and MDAs were adjusted in Table 5-

3 for NSEC. A paragraph was deleted from Section 5.2, Internal 

Dosimetry Overview. The abbreviation “UR” for uranium radiometric 

was added to Section 5.3.1.3. Last paragraph of Section 5.3.1.4 was 

modified to clarify the sensitivity (reporting level) vs. MDA. The 

numbers and order in the documents for tables 5-7 and 5-8 were 

switched. An additional footnote was added to Table 5-9. Minor 

changes were made to the references and glossary. A list of figures 

was added to Attachment B. The figures and descriptions in 

Attachment A were reordered, the descriptions of two forms were 

combined, and in some cases descriptions were added to the same 

page as the figures. Attachment D was split into two attachments (D 

and E) and the original attachments E and F were re-lettered to F 

and G. The Table in Attachment G was split into in-vitro and in-vivo 

tables. Numerous minor edits were also made to clarify or improve 

the readability of the document without changing the technical 

content. Added Attributions and Annotations section. Activity 

fractions for reactors and related facilities were eliminated and 

references to OTIB-0054 were added. Added SEC information and 

added information in Section 5.3.1.5 pertaining to the use of CEP 

data. Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of document. Training required: As 

determined by the Objective Manager. Initiated by Melton H. Chew.

Trademark Information

Mathsoft® and Mathcad® are registered trademarks of Mathsoft Engineering & Education, Inc. in the 

United States and/or other countries.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 3 of 82

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

6

5.1 Introduction ..............................................................................................................................9

5.1.1 Purpose......................................................................................................................10

5.1.2 Special Exposure Cohort Petition Information............................................................. 10

5.1.3 Scope.........................................................................................................................11

5.2 Internal Dosimetry Overview ..................................................................................................12

5.3 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Protocols ............. 16

5.3.1 In Vitro Urine Analysis ................................................................................................ 16

5.3.1.1 1948 to 1957................................................................................................ 16

5.3.1.2 1958 to 1966................................................................................................ 16

5.3.1.3 1967 to 1974................................................................................................ 18

5.3.1.4 1975 to 1988................................................................................................ 21

5.3.1.5 1989 to 2005................................................................................................ 22

5.3.2 In Vitro Methods for Individual Radionuclides ............................................................. 23

5.3.3 Fecal Sample Analysis ............................................................................................... 23

5.4 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices.............. 24

5.4.1 Whole-Body Counting................................................................................................. 24

5.4.2 Chest Counting........................................................................................................... 24

5.5 Uncertainty.............................................................................................................................25

5.6 Detection Limits .....................................................................................................................26

5.7 Sample Kit Information...........................................................................................................28

5.8 Solubility Type, Activity Fraction, and Particle Size by Facility................................................29

5.9 Facility-Specific Radionuclide Conversions ............................................................................36

5.10 Radon ....................................................................................................................................36

5.11 Attributions and Annotations .................................................................................................. 36

References ........................................................................................................................................38

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

ATTACHMENT A, SUMMARY OF URANIUM FUEL FABRICATION INTERNAL DOSIMETRY 

ISSUES................................................................................................................46

ATTACHMENT B, EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS............................ 48

ATTACHMENT C, ADJUSTMENT TO HELGESON NUCLEAR SERVICES 235U LUNG 

COUNTS, 1981 to 1983 .......................................................................................60

ATTACHMENT D, ISOTOPIC COMPOSITION OF ATR-ETR FUEL USED IN THE POWDER 

ROOM, 1966 to 1967 ...........................................................................................61
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ATTACHMENT E, ISOTOPIC COMPOSITION OF URANIUM USED FOR ATR FUEL 

FABRICATION, 1979 to 1982 ..............................................................................62

ATTACHMENT F, DETECTION LIMITS FOR GAMMA SCANS OF BIOASSAY SAMPLES, 

1967 to 1975 ........................................................................................................63

ATTACHMENT G, SENSITIVITY OF BIOASSAY ANALYSES BY BIOASSAY CODE, UST AND 

OTHER LABORATORIES....................................................................................64

ATTACHMENT H, CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE.......... 65

LIST OF TABLES

TABLE TITLE

5-1 Internal dose control program (in vitro)...................................................................................

PAGE

24

5-2 Internal dose control program (in vivo) ................................................................................... 25

5-3 Detection limits, 1958 to 1966 ................................................................................................ 26

5-4 Detection limits, 1967 to 1974 ................................................................................................ 27

5-5 Detection limits, 1975 to 1988 ................................................................................................ 28

5-6 Detection limits, 1989 to 2005 ................................................................................................ 28

5-7 Excreta sample media and types ........................................................................................... 28

5-8 Solubility type, activity fraction, and particle size by facility..................................................... 30

5-9 Facility-specific radionuclide conversions............................................................................... 36

G-1 In vitro analyses .....................................................................................................................64

G-2 In vivo analyses .....................................................................................................................64

2-1 Parameters for Type S and Type K uranium .......................................................................... 66

4-1 Limiting uranium solubility class for chronic intakes................................................................ 69

4-2 Limiting uranium solubility class for an acute intake ............................................................... 70

A-1 Acute intake urine bioassay at 2 days after intake.................................................................. 71

A-2 Acute intake urine bioassay at 5 days after intake.................................................................. 72

A-3 Acute intake urine bioassay at 10 days after intake................................................................ 73

A-4 Acute intake urine bioassay at 40 days after intake................................................................ 74

A-5 Acute intake urine bioassay at 100 days after intake.............................................................. 75

A-6 Acute intake urine bioassay at 400 days after intake.............................................................. 76

A-7 Acute intake urine bioassay at 1000 days after intake............................................................ 77

A-8 Acute intake urine bioassay at 10000 days after intake.......................................................... 78

A-9 Chronic intake (10 days) with urine bioassay on day after end of intake................................. 79

A-10 Chronic intake (100 days) with urine bioassay on day after end of intake............................... 80

A-11 Chronic intake (1000 days) with urine bioassay on day after end of intake............................. 81

A-12 Chronic intake (10000 days) with urine bioassay on day after end of intake........................... 82

H-1 Parameters for Type S and Type K uranium .......................................................................... 66

H-2 Limiting uranium solubility class for chronic intakes................................................................ 69

H-3 Limiting uranium solubility class for an acute intake ............................................................... 70

H-4 Acute intake urine bioassay at 2 days after intake.................................................................. 71

H-5 Acute intake urine bioassay at 5 days after intake.................................................................. 72

H-6 Acute intake urine bioassay at 10 days after intake................................................................ 73

H-7 Acute intake urine bioassay at 40 days after intake................................................................ 74

H-8 Acute intake urine bioassay at 100 days after intake.............................................................. 75

H-9 Acute intake urine bioassay at 400 days after intake.............................................................. 76

H-10 Acute intake urine bioassay at 1000 days after intake............................................................ 77

H-11 Acute intake urine bioassay at 10000 days after intake.......................................................... 78

H-12 Chronic intake (10 days) with urine bioassay on day after end of intake................................. 79

H-13 Chronic intake (100 days) with urine bioassay on day after end of intake............................... 80
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H-14 Chronic intake (1000 days) with urine bioassay on day after end of intake............................. 81

H-15 Chronic intake (10000 days) with urine bioassay on day after end of intake........................... 82

LIST OF FIGURES

FIGURE TITLE

B-1 Example of an AI Request for Film Badge and Bioassay Services form, 1963 .......................

PAGE

49

B-2 Example individual personnel McBee Keysort card, 1966 ...................................................... 50

B-3 Example Bioassay Data Sheet, front...................................................................................... 52

B-4 Example Bioassay Data Sheet, back ..................................................................................... 53

B-5 Example Bio-Science Laboratories UF bioassay data sheet .................................................. 54

B-6 Example Bio-Science Laboratories UR bioassay data sheet .................................................. 55

B-7 Example Bio-Science Laboratories UR bioassay data sheets showing bottle tags ................. 56

B-8 Example UST UF bioassay data sheet................................................................................... 57

B-9 Example UST UR bioassay data sheet .................................................................................. 57

B-10 Example UST MFP bioassay data sheet................................................................................ 58

B-11 Example AI Wound Monitoring Report ................................................................................... 59

H-1 Comparison of urinary excretion patterns following inhalation intakes of 5 μm

AMAD Type M and Type S 234U and 1 μm AMAD Type K 234U............................................... 67
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AERD Atomic Energy Research Department

AETR Advanced Epithermal Thorium Reactor

AI Atomics International

AMAD activity median aerodynamic diameter

ATR Advanced Test Reactor

Bq becquerel

CF critical or criticality facility

cm centimeter

CMD count median diameter

CTF critical or criticality test facility

D&D decontamination and decommissioning

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ETEC Energy Technology Engineering Center

ETR Engineering Test Reactor

EU enriched uranium

F fast (solubility rate)

FSCF (SNAP) Flight System Critical Facility

g gram

GIF Gamma Irradiation Facility

HDEHP bis-(2-ethylhexyl) phosphoric acid 

HEU highly enriched uranium

HMRFSR Heavy Metal Reflected Fast Spectrum Reactor

hr hour

HRTM Human Respiratory Tract Model

in. inch

ICRP International Commission on Radiological Protection

IREP Interactive RadioEpidemiological Program

IVLC in vivo lung count

JAERI Japan Atomic Energy Research Institute

kCi kilocurie

keV kiloelectron-volt, 1,000 electron-volts

KEWB Kinetics Experiment Water Boiler Reactor

kg kilogram

L liter

L-85 alternate name for the WBNS after 1972

LLD lower limit of detection
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M moderate (solubility rate)

MDA minimum detectable amount

MFP mixed fission product

mg milligram

min minute

mL milliliter

mo month

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPLB maximum permissible lung burden

NAA North American Aviation

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NMDF Nuclear Materials Development Facility

NRTS National Reactor Testing Station

NSEC Nuclear Science and Engineering Corporation

OCY Old Conservation Yard

OMR organic moderated reactor

OMRE Organic Moderated Reactor Experiment

ORNL Oak Ridge National Laboratory

oz ounce

pCi picocurie

POC probability of causation

PUA code for plutonium activity determined by autoradiography

PUB code for plutonium activity determined using a proportional counter to count alpha 

particles

R&D research and development

RMHF Radiation Materials Handling Facility

RSRMS Radiation Safety Records Management System

S slow (solubility rate)

S10FS SNAP 10 Flight Simulation Reactor

S2DR SNAP 2 Development Reactor

S8DR SNAP 8 Development Reactor

S8ER SNAP 8 Experimental Reactor

SER SNAP Experimental Reactor

SGR Sodium Graphite Reactor

SNAP Systems for Nuclear Auxiliary Power

SRE Sodium Reactor Experiment

SSFL Santa Susana Field Laboratory

STIR Shield Test and Irradiation Reactor (Modified STR)

STR Shield Test Reactor

TBC total-body count

TBD technical basis document

TTA thenoyltrifluoroacetone

TTF transient test facility

UAlx uranium aluminide

UCLA University of California, Los Angeles
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UF uranium fluorometric

UR uranium radiometric

U.S.C. United States Code

UST United States Testing (Company)

W watt

WBC whole-body count

WBNS Water Boiler Neutron Source

wk week

ZnS(Ag) zinc sulfide scintillation crystal activated with silver

μCi microcurie

μg microgram

μL microliter

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 9 of 82

5.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

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

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

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

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

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

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2007):

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 10 of 82

5.1.1

The purpose of this TBD is to describe internal dosimetry systems and practices at Area IV of the 

Santa Susana Field Laboratory, Canoga Avenue Facility, Downey Facility, and De Soto Avenue 

Facility (sometimes referred to as Energy Technology Engineering Center [ETEC] or Atomics 

International). This document discusses historical and current practices in relation to the evaluation of 

internal exposure data for monitored and unmonitored workers.

Purpose

5.1.2

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the SEC 

classes below, NIOSH intends to use any internal and external monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed at 

the Area IV, Canoga Avenue, De Soto Avenue, or Downey facilities during the SEC periods stated 

below, but who do not qualify for inclusion in the SEC, may be performed using these data as 

appropriate.

Special Exposure Cohort Petition Information

Class Added to the SEC

• All employees of the Department of Energy (DOE), its predecessor agencies, and DOE 

contractors and subcontractors who worked in any area of Area IV of the Santa Susana Field 

Laboratory for a number of work days aggregating at least 250 work days from January 1, 

1955 through December 31, 1958, or in combination with work days within the parameters 

established for one or more other classes of employees in the SEC (Sebelius 2009).

Area IV of SSFL (1955-1958)

NIOSH determined in 2009 (NIOSH 2009) that it cannot estimate internal exposures with sufficient 

accuracy during the period from 1955 through 1958. This includes the time from the beginning of 

Area IV radiological activities in 1955, to late 1958 (selected as December 31, 1958 for SEC 

evaluation purposes). NIOSH intended the end of the SEC period to correspond with the date after 

which a routine bioassay program existed at SSFL and after which sufficient internal monitoring was 

thought to have been identified. Later in 2009 after further data review, NIOSH determined that 

sufficient data are not available for adequate dose reconstruction until January 1, 1965; NIOSH 

subsequently initiated a second SEC class for Area IV for the years 1959-1964, as described below

Classes Recommended by NIOSH for addition to the SEC

• All employees of the Department of Energy, its predecessor agencies, and their contractors 

and subcontractors who worked in any area of Area IV of the Santa Susana Field Laboratory 

from January 1, 1959 through December 31, 1964, for a number of work days aggregating at 

least 250 work days, occurring either solely under this employment or in combination with work 

days within the parameters established for one or more other classes of employees included in 

the Special Exposure Cohort (NIOSH 2010a).

Area IV of SSFL (1959-1964)

Through the course of ongoing dose reconstruction, continued data capture efforts, and investigations 

associated with SEC-00093, NIOSH determined in 2010 (NIOSH 2010a) that it cannot estimate with 

sufficient accuracy the potential internal exposures to various radionuclides to which the proposed 

class may have been subjected during the time period from January 1, 1959 through December 31, 
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 11 of 82

1964. NIOSH found that it is likely feasible to reconstruct external dose, including occupational 

medical dose, for Area IV of the Santa Susana Field Laboratory workers with sufficient accuracy.

• All employees of the Department of Energy, its predecessor agencies, and its contractors and 

subcontractors who worked at the Canoga Avenue Facility, Los Angeles, California, from 

January 1, 1955 through December 31, 1960 for a number of working days aggregating at 

least 250 work days, occurring either solely under this employment or in combination with work 

days within the parameters established for one or more other classes of employees in the SEC 

(NIOSH 2010b).

Canoga Avenue Facility

NIOSH determined in 2010 (NIOSH 2010b) that it cannot estimate internal exposures with sufficient 

accuracy during the period from 1955 through 1960. This includes the entire covered period for the 

Canoga Avenue Facility. NIOSH found that it is likely feasible to reconstruct external dose, including 

occupational medical dose, for Canoga Avenue workers with sufficient accuracy. 

• All employees of the Department of Energy, its predecessor agencies, and their contractors 

and subcontractors who worked at the Downey Facility in Los Angeles County, California, from 

January 1, 1948 through December 31, 1955, for a number of work days aggregating at least 

250 work days, occurring either solely under this employment or in combination with work days 

within the parameters established for one or more other classes of employees included in the 

Special Exposure Cohort (NIOSH 2010c).

Downey Facility

NIOSH determined in 2010 (NIOSH 2010c) that it cannot estimate internal exposures with sufficient 

accuracy during the period from 1948 through 1955. This includes the entire covered period for the 

Downey Facility so only partial dose reconstructions will be performed for individuals who do not 

qualify for inclusion in the SEC.

De Soto Avenue Facility

• All employees of the Department of Energy, its predecessor agencies, and their contractors 

and subcontractors who worked at the De Soto Avenue Facility in Los Angeles County, 

California, from January 1, 1959 through December 31, 1964, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other classes of 

employees included in the Special Exposure Cohort (NIOSH 2010d).

NIOSH determined in 2010 (NIOSH 2010d) that it cannot estimate internal exposures with sufficient 

accuracy during the period from 1959 through 1964. The De Soto Avenue Facility has a covered 

period of DOE 1959-1995 and DOE Remediation 1998. Therefore, dose reconstructions will be 

performed for 1965 through 1995 and partial dose reconstructions will be performed for individuals 

who do not qualify for inclusion in the SEC for 1959 through 1964.

5.1.3

The SSFL facility, which includes four locations, has been identified in various ways over time. This 

TBD uses SSFL to refer to all locations unless more specific location information is warranted. In that 

context, SSFL includes Area IV of the Santa Susana Field Laboratory [SSFL, which has also been 

Scope
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 12 of 82

known as Nuclear Development Field Laboratory, Liquid Metal Engineering Center, and Energy 

Technology Engineering Center (ETEC)], portions of the Downey facility, the Canoga Avenue Facility, 

and the De Soto facility.

North American Aviation (NAA) entered into a contract with AEC to conduct nuclear research 

operations at Area IV and the Downey, Canoga and De Soto sites. At that time, Atomics 

International, an internal division of NAA, was the company’s designated nuclear research and 

development division. In addition to the employees of the AI division, other employees of NAA who 

worked at any of the above sites during the AEC contract period are potentially eligible for EEOICPA 

benefits.

SSFL facilities, operated by The Boeing Company and its predecessors (North American Aviation), 

have played an important role in the development of the U.S. nuclear program. SSFL operations 

have involved research and development (R&D) in the areas of development and testing of nuclear 

reactors, nuclear support operations, and non-nuclear energy (Sapere and Boeing 2005, p. 2-2). This 

TBD is part of the overall SSFL Site Profile, which describes plant facilities and processes, historical 

information, and environmental data in relation to dose reconstruction for SSFL workers. It contains 

supporting documentation to assist in the evaluation of occupational internal doses from these 

processes in accordance with the Internal Dose Reconstruction Implementation Guideline (NIOSH 

2002).

The methods and concepts of measuring occupational internal doses to workers have evolved since 

the beginning of SSFL operations. An objective of this document is to provide supporting technical 

data to evaluate internal SSFL occupational doses that can reasonably be associated with worker 

radiation exposures covered by the EEOICPA legislation.

In addition, this document presents the technical basis of methods used to prepare SSFL worker dose 

information for input to the NIOSH Interactive RadioEpidemiological Program (IREP) software. 

Information on measurement uncertainties is an integral component of the approach. This document 

describes the evaluation of uncertainty for SSFL exposure and dose records.

5.2 INTERNAL DOSIMETRY OVERVIEW

Section 2 of this Site Profile (ORAUT 2006) contains some of the following discussion of historical 

activities. However, this TBD discusses the information in the context of internal dosimetry.

From 1948 to 1955, the Atomic Energy Research Department (AERD) of North American Aviation 

(NAA) occupied a portion of Building 001 at the Downey Plant on Lakewood Boulevard in the City of 

Downey, California. On April 21, 1952, AERD constructed a small aqueous homogeneous reactor in 

an area that is now a loading dock. The reactor and its associated exponential pile were apparently 

used as a source of neutrons for reactor physics experiments. The Water Boiler Neutron Source 

(WBNS) used a 93%-enriched uranyl sulfate solution and operated at power levels up to 4 W. The 

WBNS operated at Downey until December 1955. In 1956, it was dismantled and moved to a facility 

at the SSFL. Little information is available about the specific uses of the area after the WBNS 

relocation. All available records indicate that the Plant was not left in a contaminated condition, and 

this was confirmed by a survey in 2000 (Liddy and Rutherford 2001). Other radioisotopes at the 

Downey Plant were apparently used for industrial radiography and were not a likely source of internal 

exposure [1].

In 1956, NAA formed SSFL as one of its divisions to replace AERD. Between 1956 and 1960, SSFL 

performed nuclear R&D at a facility known as the Canoga Avenue Facility (Vanowen Building, 

Building 038) in Canoga Park, California. SSFL designed, developed, and operated two small, 

aqueous, 93%-enriched uranyl sulfate research reactors at the Canoga Avenue Facility; the reactors 
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 13 of 82

were designated L-47 and L-77. The maximum power ratings were 5 W and 10 W, respectively. 

Other operations included reactor design, fuel development, and radiochemistry. L-47 operated 

between August 1957 and June 1958, and L-77 operated between May 1958 and February 1960. 

The reactors were in the southeast corner of the building which is now a loading dock. Historical 

information indicates that the areas that supported nuclear operations were routinely surveyed until 

and after the removal of all radioactive material. No contamination above the current limits for 

unrestricted use was found. This was confirmed by a 1995 survey of the facility by the U.S. Nuclear 

Regulatory Commission (Rutherford 2002). Exposure of subsequent occupants to contamination 

levels of importance for internal dose reconstruction of the building would be unlikely [2].

In 1959, SSFL moved to a new facility on De Soto Avenue in Canoga Park and transferred all 

radiological activities from the Canoga Avenue Facility, including the L-77 reactor, to the new facility. 

Until 1983 nuclear operations were conducted at the De Soto facility in Buildings 101 and 104 (until 

1984 these buildings were known as 001 and 004, respectively). Work continued at a “much-reduced 

level” in Building 104 until the mid-1990s. The L-77 reactor in Room 416-61 of Building 104 was 

decontaminated and decommissioned in the late 1970s. The facility was released for unrestricted use 

in February 1982. Many nuclear fuel manufacturing operations were conducted in the 1960s and 

1970s in Buildings 101 and 104. These operations used 2%- to 93%-enriched uranium metal and 

composites. Most notable of these was fuel for the Advanced Test Reactor (ATR), which used 

uranium-aluminum alloy plates. This operation was the source of internal exposures that were difficult 

to evaluate because of the unanticipated insolubility and small particle size (when airborne) of the 

alloy. Attachment A summarizes these problems. Attachment H evaluates the impact of the uranium 

aluminide-specific model on intake and dose calculations.

Two other operations in Building 104 at the De Soto facility involved radioactive material. The 

Gamma Irradiation Facility (GIF) used sealed 137Cs and 60Co sources for hardness testing and food 

irradiation research. Biannual leak tests of these sources confirmed that internal exposures from 

these devices were unlikely. Operations ended in the late 1980s and, after the removal of all 

equipment, the State of California released the GIF for unrestricted use in July 1999. The other 

Building 104 operation, the Mass Spectroscopy Laboratory or Helium Laboratory, analyzed 

radioactive samples for helium content. Operations ended in 1995; the mass spectroscopy equipment 

was sent to Richland, Washington. In 1998, all remaining support equipment was removed and the 

facility was decontaminated. In October 1998, the State released the Helium Laboratory for 

unrestricted use. Thirteen separate radiation surveys in Buildings 101 and 104 demonstrated that no 

residual contamination existed that would be of interest for dose reconstruction (Boeing 2003).

SSFL consisted of four administrative areas and a buffer zone. NAA established Area IV in 1953 as a 

nuclear R&D facility. Starting in 1954, several nuclear reactors and critical assemblies were built in 

Area IV; but no nuclear operations took place until 1955. In 1959, a significant incident occurred 

when a sodium-cooled, graphite-moderated reactor in Building 4143 [the Sodium Reactor Experiment 

(SRE)] had a loss of coolant, which resulted in damage to 13 fuel assemblies (Sapere and Boeing 

2005, pp. 2-1, 2-3, 2-5). A government-owned, contractor-operated organization was formed to 

conduct non-nuclear research at the site, which was renamed the Energy Technology Engineering 

Center (ETEC) in 1978. Most nuclear research programs and operations ended in 1988. All research 

ended in 1998. ETEC was given the job of decontaminating and decommissioning (D&D) at the 

former nuclear facilities. D&D is complete at many of the facilities in Area IV, but some of this work is 

still ongoing.

SSFL documents from the early 1960s describe all the elements of a comprehensive radiation safety 

program including a laboratory with bioassay capability. The program was under central direction and 

covered the facilities operated by SSFL (Lang undated, Lang 1960). There is one possible exception. 

The Organic Moderated Reactor Experiment (OMRE) was a SSFL facility at the National Reactor 

Testing Station (NRTS) in Arco, Idaho. NRTS apparently provided dosimetry services to SSFL 
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personnel (Lang 1960). It is uncertain, based on current information if any bioassay results were sent 

to SSFL.

SSFL established health and safety files on each employee that contained radiation exposure 

records, injury records, and other “pertinent” data. Even though records were established for each 

employee, this does not imply that all personnel were monitored. Only persons working in “radiation 

areas” were required to wear film badges and, in some cases, pocket dosimeters and finger rings. Invitro bioassay samples were only obtained from workers whose assignments involved “potential 

internal exposure to radioactive materials” (Lang undated, Lang 1960). Between 1948 and 1999, the 

entire workforce of Rocketdyne/Atomics International employees numbered 46,970, most of whom 

were engaged in non-nuclear activities. A 2006 study determined 5,801 workers were monitored for 

external or internal radiation: 3,569 external radiation only, 58 internal radiation only, and 2,174 both 

internal and external radiation (Boice 2006). All personnel exposure records were retained at the site 

and are accessible and available (Sapere and Boeing 2005, p. 3-12).

The environmental monitoring program at Area IV was established in May 1954 before construction of 

the first radiological facility. Stack air emissions were measured in all facilities with radiological work 

areas or where unencapsulated or unpackaged radioactive material was handled. Gross alpha and 

beta activity was monitored on a weekly basis (Sapere and Boeing 2005, p. 3-13). From 1959 to the 

present, ambient gross beta activity in air has been measured continuously in five locations. From 

1963 on, gross alpha activity was also measured. At present, ambient air samples are analyzed for 

isotopic content (Sapere and Boeing 2005, p. 3-15).

By 1959, routine urine samples were requested on Fridays, and each monitored employee was 

required to submit the first voiding on Monday morning (after an absence from work of 48 hr or more). 

The time of the previous voiding was recorded to determine the excretion rate. If the Monday morning 

sample was verified as positive, a series of 24-hr samples was collected to determine the body 

burden. Employees were requested to fill these samples on Sunday (Kellehar 1959). Appropriate 

adjustments to this schedule were made for weekend work, etc. Uranium urinalyses were performed 

for fuel fabrication workers, and gross beta or mixed fission product (MFP) urinalyses were performed 

for many workers at SSFL and the Canoga Avenue Facility. Commercial laboratories were available 

for analysis of alpha emitters other than uranium and for soft beta emitters (Alexander 1959). And 

eventually, most routine samples were analyzed by a vendor. Special samples were requested when 

an intake was suspected or if a routine sample exceeded 10% of the urinary excretion expected from 

1 maximum permissible body burden (MPBB). Special samples were analyzed by the onsite 

laboratory, the commercial lab, or both (Kellehar 1966a).

Entry into the bioassay program at SSFL was apparently based on job assignment [3]. By the early 

1960s, the bioassay program “normally” consisted of urinalysis for personnel whose work 

assignments involved “potential exposure to radioactive materials.” The frequency of sampling varied 

from one to four per year, depending on the nature of the employee’s work, past exposure history, etc. 

(Lang undated). A 1963 procedure called for completion of a “Request for Film Badge and Bioassay 

Services” form for new employees who required such services (Garcia 1963). Figure B-1 in 

Attachment B is an example. The Health and Safety Operation Unit completed the bottom portion, 

which included a place to specify bioassay for the employee.

In 1970, standards for bioassay sampling were published (Staszesky 1970). Work in areas where 

unencapsulated radioactive material was present required baseline and termination urine samples. A 

new baseline could be required for a change in job assignment. For new operations, a “pilot” 

bioassay program consisting of weekly urine samples could be required until a pattern was 

established. Regular work in these areas required a routine quarterly urine sample, but monthly 

samples could be required in a case of high exposure potential. Periodic fecal samples and in vivo

counts could also be required. Employees who performed work in these areas only periodically were 
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subject to semiannual urine samples. Personnel such as project engineers, industrial engineers, etc., 

who frequently entered these areas but did not perform hands-on work, provided routine annual urine 

samples.

By the mid-1970s, the definition of who was included in the routine bioassay monitoring program had 

changed to “personnel whose work assignments potentially expose them to respirable-sized 

radioactive aerosols” (Hart 1979). By the late 1980s, the criterion was “personnel whose work 

assignments potentially expose them to radioactive aerosols” (Tuttle 1989). Quarterly urine sampling 

was standard through the 1980s (Hart 1979; 1980a,b,c; Eggleston 1983, 1984; Tuttle 1985, 

1986a,b,c, 1988a,b, 1989).

Special bioassay that consisted of more frequent urine sampling was in place very early (1960), and 

in the mid-1970s fecal sampling was also used but “only when gross internal contamination” was 

suspected (Hart 1979). Using the concept of an MPBB, an excretion rate was determined for each 

radionuclide that would indicate that 1 MPBB had been received. For several years before 1968, the 

policy was to restrict employees from work in potential airborne areas until their body burdens were 

less than 25% of the MPBB (Alexander 1968a). Starting in January 1968, SSFL imposed a restriction 

from work in areas with potential airborne exposure (or in some cases from all radiation areas) if the 

bioassay results indicated the receipt of 50% or more of the MPBB. The restriction remained in place 

until two consecutive bioassay samples indicated that the remaining deposition was less than 25% of 

the MPBB (Staszesky 1970).

It appears that the first in vivo counting occurred in 1967 at a U.S. Atomic Energy Commission (AEC)-

funded facility at the University of California, Los Angeles (UCLA). These counts were performed in 

conjunction with the insoluble enriched uranium (EU) intakes described in Attachment A. Although 

these counts were frequently referred to as whole-body counts (WBCs) or total-body counts (TBCs) in 

site documents, they were really attempts to quantify the amount of 235U activity in the lung. At that 

time chest (lung) counting was in its infancy, and standard calibration methods using phantoms had 

not been developed. SSFL ultimately determined that the results of these counts were significant 

overestimates. From 1968 to 1983, in vivo counting was done on the site using portable counters 

operated by Helgeson Nuclear Services. Most of these counts were also chest counts for EU. Work 

with unirradiated highly enriched uranium (HEU) ended in 1983, and no more chest counts occurred 

(Tuttle 1986b). WBCs for fission and activation products apparently occurred during and after this 

period, probably in conjunction with suspected exposures or new projects [4].

Even though SSFL ended research activities with radioactive materials in 1988, the internal dosimetry 

program continued; most of the features of the program are listed above [5].

The bioassay records in the individual files generally consist of:

• Individual personnel McBee Keysort cards (Figure B-2, Attachment B), which were used to 

track the type, frequency, and week of sample collection. In addition, the card summarized 

individual results. The cards can be difficult to read due to the quality of the copies, and dose 

reconstructors should refer to the forms listed below for urine and fecal data. This card might 

be the only place that in vivo data are listed. Two versions were produced and used from 

1963 to 1966 and from 1967 to 1970, respectively. The latter continued in use at least until 

1983. This card was the successor to one that recorded only results before 1963 (example 

not included).

• Bioassay Data Sheets (Figures B-3 and B-4, Attachment B), which were two-page forms that 

contained all the information on analyses conducted by the onsite laboratory. Some, if not 

most, offsite analysis results were apparently transferred to this form.
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• Individual analysis results sheets (Figures B-5 to B-10, Attachment B), which contained all the 

information on analyses conducted by offsite laboratories. This form might show the 

calculated result (even if it was below the limit of detection), the sensitivity of the analysis, and 

the uncertainty.

• Wound Monitoring Reports (Figure B-11, Attachment B). The entries on this form are selfexplanatory.

• Individual hand-drawn or computer-generated plots of bioassay data, apparently done as 

follow-up to high results. Some can be difficult to read due to the quality of the copies.

• Incident workups after high bioassay results. These might contain data summaries in a more 

legible and understandable form than information in other locations because they were 

produced by health physicists familiar with the data.

5.3 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.3.1

5.3.1.1 1948 to 1957

In Vitro Urine Analysis

It appears that no bioassay program existed before August 1958. Useful source term data or facilityspecific ambient airborne concentrations are probably not available. This unmonitored period 

includes the operation of the WBNS from 1948 to 1955 at the Downey Plant, the operation of the L-47 

reactor at the Canoga Avenue Facility, and operations at SSFL from 1954 to 1957 (including, for 

example, the operation of the WBNS in 1956 and 1957). Documents that might indicate 

environmental releases or incidents from these facilities have not been found. Bioassay data after 

1957 in similar facilities, such as the WBNS or L-77 reactor, might be the only data available to the 

dose reconstructor as an indicator of missed dose. Because these data would be from a different 

period, dose reconstructors should be cautious in applying them.

5.3.1.2 1958 to 1966

According to Kellehar (1966a) the bioassay program was initiated in August 1958. As mentioned 

above, routine bioassay was normally by urinalysis at a frequency of 3 months. At first, gross alpha or 

gross beta measurements were made of the samples. Specific radionuclides could be determined 

“where required” (Lang 1960). Some detail has been found on early urinalysis methods. In addition 

to the in-house laboratory capability, bioassay services were contracted to the following vendors:

• U. S. Nuclear, Burbank, California

• Controls for Radiation, Cambridge, Massachusetts

• Biomedical Procedures, North Hollywood, California

• Bio-Science Laboratories, Santa Monica, California

• Atomic Corporation of America, Panorama City, California

• United States Testing Company (UST), Richland, Washington

Information on what vendors were used by the site and when they were used has been captured in a 

bioassay data base. Due to various problems with the other laboratories (Fisher 1963), it appears 

that Bio-Science Laboratories analyzed the most samples early in this period. In 1964, SSFL initiated 

a contract with UST, which became International Technology Analytical Services, Quanterra 

Environmental Services, and finally Severn Trent Laboratories. UST appears to have been the main 

laboratory vendor between 1965 and 1968. After that time various vendors were used.
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SSFL Uranium Methods [A and B (fluorometric) and EU (radiometric)]

The use of Method A and Method B to designate uranium bioassay methods apparently changed over 

the years. In general, it should be easy to distinguish fluorometric methods if the units were reported; 

fluorometric results were reported in mass (micrograms or milligrams) per unit volume rather than in 

radioactivity per unit volume. The fluorometric methods fused uranium from raw urine with sodium 

fluoride and measured the fluorescence when the compound was exposed to ultraviolet light. The 

uranium fluorescence can be quenched by certain heavy metal ions, and the primary difference 

between the two methods was that Method A considered these quenching effects negligible and that 

Method B took into account possible quenching effects (Mason and Burr 1958).

Due to its higher specific activity, EU activity could be determined by counting. The procedure 

involved digesting the urine, removing some of the interfering cations, and electroplating the uranium 

onto nickel disks. The disks were counted in a proportional counter (Mason and Burr 1958).

No specific information on sensitivities for the in-house laboratory was obtained. The following values 

for the contractor laboratories should be used for dose reconstruction because they should be 

comparable to the in-house laboratory [6].

Vendor Uranium Methods [codes: 1A (fluorometric) and 1B (radiometric)]

The fluorometric method was implemented during the first half of 1948 at AEC sites (ORAUT 2010, 

p. 30). In May 1959, U. S. Nuclear quoted a price for fluorometric analysis with a minimum 

measurable concentration of 0.002 μg/mL based on a minimum volume of 1 L (Shepard 1959). 

Controls for Radiation quoted 0.0001 μg/mL based on a minimum volume of 150 mL (O’Brien 1959). 

NSEC gave its minimum measurable concentration as 0.0002 μg/mL (Edelmann 1959).

The early radiometric methods generally used separation chemistry followed by counting on a gasflow proportional counter or a ZnS(Ag) scintillation counter (ORAUT 2010, p. 30). Shepard (1959) 

quoted a minimum measurable concentration of 7.5 dpm/L for radiometric determination of EU. 

NSEC gave its minimum measurable concentration for EU as 20 dpm/24-hr sample (NSEC 1957).

Gross Alpha

Shepard (1959) gave a minimum measurable concentration of 7.5 dpm/L for gross alpha counting. 

NSEC gave its minimum measurable concentration as 2.0 dpm/mL (NSEC 1957).

Gross Beta

Shepard (1959) gave a minimum measurable concentration of 75 dpm/L for gross beta counting. 

NSEC gave its minimum measurable concentration as 4.0 dpm/mL (NSEC 1957).

Mixed Fission Products

The early procedure for MFPs was generally to add strontium carrier to the aluminum oxide solution 

for the plutonium procedure, then to precipitate lanthanum hydroxide. This procedure extracted rare 

earths and strontium with yields ranging from 90% for cerium to 23% for strontium. The dried 

planchet was counted for beta activity with an approximate minimum detectable amount (MDA) of 

60 dpm/sample (ORAUT 2010, p. 34). This procedure separated and counted radionuclides of 

alkaline earths and rare earths such as strontium, yttrium, barium, lanthanum, cerium, europium, and 

promethium. It might not have accounted for radionuclides of ruthenium, cesium, zinc, cobalt, 

manganese, niobium, or zirconium (ORAUT 2010, p. 34). NSEC gave its minimum measurable 

concentration as 10.0 dpm/mL (NSEC 1957).
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Polonium-210

Bio-Science used electrodeposition on a nickel disk and counting in a low-background proportional 

counter. The sensitivity quoted was 0.001 dpm/mL (Lee 1963).

Plutonium

One of the first plutonium bioassay analyses consisted of lanthanum fluoride precipitation, 

thenoyltrifluoroacetone (TTA) extraction, and gross alpha counting. Electrodeposition on a stainlesssteel disk was followed by counting with nuclear track emulsion (autoradiography). Plutonium 

bioassay results are available as early as 1963. The urine maximum permissible concentration 

(MPC) that indicated 1 MPBB was listed as 4 dpm/24 hr. Therefore, in keeping with normal practice 

at the time, 0.4 dpm/24 hr would be the sensitivity required [7]. Bio-Science quoted a sensitivity for 

plutonium bioassay of 0.00006 dpm/mL, but stated that it would have to subcontract the analyses 

(Lee 1963).

Strontium-90

Bio-Science coprecipitated 90Sr and 90Y with calcium as oxalates. Yttrium-90 was isolated and 

purified using a milking procedure with tributylphosphate, assayed, and reported as 90Sr equivalent. 

The method assumed that 90Sr and 90Y were in secular equilibrium. If not, counting could be delayed 

for ingrowth. The sensitivity quoted was 0.1 dpm/mL (Lee 1963).

Tritium

Tritium was not a significant personnel exposure hazard from research reactor operations, primarily 

due to the low power levels. However, 15 kCi of tritium in a triple-walled container was in 

Room 416-51 in Building 004 at the De Soto Facility until November 1967 (Alexander 1967a). The 

earliest viable method for tritium analysis seems to have been liquid scintillation counting of raw urine 

in a scintillation cocktail. NSEC gave its minimum measurable concentration as 2,220 dpm/mL (0.001 

μCi/mL) (NSEC 1957).

Thorium

No details of early thorium analyses were found. Tracerlab analyzed three urine samples for SSFL in 

1959. The detection limit was 0.2 μg/125-mL sample (Tracerlab 1959).

5.3.1.3 1967 to 1974

Partial documentation on bioassay methods from 1967 to 1974 was found. The sources of these data 

were mainly statements of work for bioassay services. These documents are believed to refer to 

services offered by UST (Kellehar 1966b, 1967; Spielman 1968; Bales 1969; Staszesky 1971). A 

review of the bioassay data captured confirmed that UST was the primary vendor during this period. 

Many of the methods listed in this TBD were undoubtedly in place before 1967.

Uranium Fluorometric (Procedure A)

The sample was acidified. A 100-μL aliquot was analyzed directly. Recovery was 93% ±18%. The 

detection limit was 0.5 μg/L. A fluorophotometer was used to measure the uranium.
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Uranium Fluorometric (Procedure B)

Uranium was extracted from the ashed residual salts with methyl isobutyl ketone using a salting 

solution of acid lammonium hydroxide. A fluorophotometer was used to measure the uranium 

present. Recovery was 83% ±8%. The detection limit was 0.05 μg/L.

Uranium Radiometric (UR)

Uranium was isolated as in Procedure B above. It was measured by a gas-flow proportional counter 

or a ZnS(Ag) scintillation counter. Recovery was 83% ±8%. The detection limit was 0.5 dpm/sample. 

At SSFL the standard sample volume per day was 1,500 mL. The result was the total alpha activity.

Plutonium (Procedure A)

Plutonium was isolated by coprecipitation as the fluoride, extraction with TTA, and electrodeposition. 

Autoradiography was used to detect the plutonium. Recovery was 82% ±14%. The detection limit 

was 0.05 dpm/sample. At SSFL the standard sample volume per day was 1,500 mL. Plutonium 

results would have included activity from 238Pu, 239Pu, and 240Pu, but not 241Pu or 241Am.

Plutonium (Procedure B)

Plutonium was isolated as in Procedure A and counted in a gas-flow proportional counter. Recovery 

was 85% ±10%. The detection limit was 0.5 dpm/sample. At SSFL the standard sample volume per 

day was 1,500 mL. Plutonium results would have included activity from 238Pu, 239Pu, and 240Pu, but 

not 241Pu or 241Am.

Tritium

Tritium was determined by liquid scintillation counting of an aliquot of the sample. Recovery was 

100% ±10%. The detection limit was 5,000 dpm/mL.

Polonium

Polonium was spontaneously deposited on a silver disk from a dilute hydrochloric acid solution of the 

residual salts. The disk was alpha counted. Recovery was 93% ±7%. The detection limit was 

0.5 dpm/sample.

Strontium

Strontium was isolated by precipitation as the oxalate and then as the nitrate. Yttrium was removed 

by a nitric acid wash. Barium was removed as the chromate. Strontium was precipitated as the 

carbonate and counted with a gas-flow proportional counter. This provided a gross strontium result. 

Recovery was 85% ±9%. The detection limit was 4 dpm/sample. After the ingrowth of 90Y (in 2 wk), 

the 90Y was isolated as the hydroxide, and then as the oxalate, which was burned to the oxide and 

counted in a gas-flow proportional counter. The 90Sr was computed from the 90Y, which was in secular 

equilibrium at separation time. Recovery was 78% ±11%. The detection limit was 4 dpm/sample.

Thorium

Thorium was isolated by a double fluoride precipitation and extraction with TTA followed by 

spectrophotometric determination with morin. Recovery was 78% ±12%. The detection limit was 

1.0 μg for a 1,000-mL sample. Thorium-232 was determined by planchetting the thorium fraction and 

alpha counting. The detection limit was 0.5 dpm/sample.
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Phosphorus-32

Phosphorus-32 was separated first as the phosphomolybdate and then as the magnesium ammonium 

phosphate, which was planchetted and beta counted in a gas-flow proportional counter. Recovery 

was 86% ±7%. The detection limit was 4 dpm/sample.

Sulfur-35

The sample was ashed to drive off tritium and carbon and counted by liquid scintillation techniques. 

Recovery was 95% ± 10%. The detection limit was 10 dpm per sample for a 1-mL sample aliquot.

Carbon-14

Carbon-14 was determined by liquid scintillation techniques. Recovery was 65% ±15%. The 

detection limit was 10 dpm for a 1-mL sample aliquot.

Gross Alpha

Gross alpha was determined by extracting most actinides from a 9N nitric acid solution into diethyl 

ether. This provided recoveries from 80% to 99% of most actinides. The detection limit was 

1 dpm/sample.

Gross Beta

Gross beta was determined from a beta count of the ashed residual salts. A 40K correction could be 

made. Recovery was 95% ±5%. The detection limit was 2 dpm for a 50-mL sample.

Promethium-147

Promethium-147 was chemically isolated by precipitation as the fluoride and extraction into TTA then 

counted by liquid scintillation techniques. Recovery was 83% ±10%. The detection limit was 

5 dpm/sample. In 1967, for at least one case of suspected promethium exposure (147Pm-oxide), 

bioassay analyses were performed on urine and fecal samples (Alexander 1967a).

Americium and Curium

Americium and curium were isolated from contaminating actinides with bis-2-(ethylhexyl) phosphoric 

acid (HDEHP) in toluene extraction out of 4N HNO3. Americium and curium were extracted into the 

HDEHP toluene solution at a pH of 4.5, back-extracted using 4N HNO3, electrodeposited, and 

counted using alpha spectroscopy. The recovery was 80% ±15% with a detection limit of 0.5 

dpm/sample.

Gross Fission Products

Gross fission products were precipitated as the oxalate from a basic solution, planchetted, and beta 

counted in a gas-flow proportional counter. Recovery for all fission products averaged 82% ±5%. 

The detection limit was 5 dpm/sample based on the counting efficiency of the radiologically critical 

isotopes 90Sr and 90

Y. A gamma scan was used for routine detection of certain gamma-emitting 

radionuclides. The gamma scan was performed by counting a sample in the well of a NaI(Tl) crystal 

and feeding the pulses to a multichannel pulse height analyzer. The simultaneous measurement of 

more than one radionuclide using gamma energy analysis precluded making accurate statements 

about detection limits and precision. The laboratory offered a table of detection limits per sample 
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based on experience; this information appears in Attachment F. Lower detection limits than those in 

the table were available at extra cost. Other gamma-emitting radionuclides could be determined.

5.3.1.4 1975 to 1988

The following analytical methods were taken from a series of annual reports (Hart 1979, 1980a,b,c; 

Eggleston 1983, 1984; Tuttle 1985). The measurement type in parentheses (e.g., “UR”) appears in 

many personnel bioassay records. The detection limits should have improved over the years, but a 

listing was not found. The reports list positive results for the year and, in some cases, follow-up 

results. The follow-up results were examined to see if the lowest value reported as nonzero could be 

determined. In general, there were not enough data to provide a satisfactory result. Therefore, dose 

reconstructors will have to use the detection limits in the previous section or, in some cases, on the 

individual results.

Uranium Radiometric (UR) and Uranium Fluorometric (UF)

Uranium was extracted from an acidic solution of ashed urine using aluminum nitrate, tetrapropyl 

ammonium hydroxide, and methyl isobutyl ketone. The uranium was recovered by back-extracting 

into water by evaporating to ketone. The water solution was planchetted for alpha counting for the UR 

analysis. The result was the total alpha activity. 

Fluorometric analysis required removal of an appropriate aliquot of the water solution before 

planchetting for pelletizing with NaF-LiF. The pellet was analyzed for uranium with a fluorometer. 

Most uranium samples were apparently analyzed by using both techniques.

Mixed Fission Products (FP1)

MFPs were precipitated from a basic oxalate media. By adjustment of pH and oxalate concentrations, 

elements that are amphoteric or that form oxalate complexes in the form of excess oxalate were also 

precipitated. Alkali metals such as 137Cs did not precipitate. In addition, volatile fission products such 

as 131I were lost. The precipitate was washed with NAOH and water and planchetted for counting.

Mixed Fission Products (FP2)

FP2 used the same extraction procedure as FP1; however, the soluble oxalate precipitates were 

gamma counted for 137Cs and other gamma emitters. The results from the FP1 analysis and the FP2 

analysis were summed and reported as a single value.

Mixed Fission Products (FP3)

This analysis was the same as FP2 except oxalate-insoluble results were reported separately as 

FP3a and oxalate-soluble results were reported separately as FP3b. The FP3a analysis was 

assumed to indicate 90Sr, but other radionuclides, such as 60Co, might also have been detected. 

Further analysis was used to quantify 90Sr specifically and to identify interfering radionuclides if 

significant quantities occurred. The FP3b analysis was selective for 137Cs using gamma-ray 

spectroscopy (Tuttle 1988a).

Plutonium (PUA, PUB)

After reduction to plutonium (III) and (IV) with hydroxylamine hydrochloride, plutonium was 

precipitated with lanthanum fluoride. This isolated the plutonium from most elements (including 

uranium), except for thorium, rare earths, and actinides. After oxidation of plutonium with sodium 

nitrate in acid media, extraction of plutonium was performed with 0.5 M TTA in xylene. After 
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extraction, the aqueous solution containing plutonium was neutralized and concentrated by heating. 

After oxidation of the plutonium in a basic media, it was electrodeposited on a stainless-steel disc. 

The plutonium activity was determined by autoradiography (PUA) for greater sensitivity or counted for 

alpha radiation with a proportional counter (PUB). Plutonium results would have included activity from 

238Pu, 239Pu, and 240Pu, but not 241Pu. The PUA analysis was chemically selective for plutonium and 

excluded the 241Am that is generally present (Tuttle 1986b).

Gross Beta, High Level (GBH)

The gross sample was evaporated to dryness then organically digested by hydrogen peroxide and 

nitric acid. The natural potassium (40K) correction was determined by diluting the ashed salts to a 

known volume and removing an aliquot for flame spectrophotometry. The remaining solution was 

evaporated to near dryness, planchetted, and counted for beta radiation with a proportional counter. 

The radioactivity in the urine sample due to 40K was subtracted from the gross count.

Gross Alpha (GA1a)

This analysis was specific for uranium and plutonium, which were extracted from ashed urine salts 

using aluminum nitrate, tetrapropylammonium hydroxide, and methyl isobutyl ketone. Transuranic 

elements did not extract to any appreciable extent. Uranium or plutonium (or both) was recovered by 

back-extracting into water by evaporating the ketone. The uranium or plutonium was electrodeposited 

on a stainless-steel disc and autoradiographed.

Gross Alpha (GA1b)

This was the same analysis as GA1a except the extraction solution was planchetted and counted for 

alpha radiation with a proportional counter.

Gross Alpha (GA2)

This analysis was specific for all alpha emitters. Metabolized actinides were converted to states 

suitable for coprecipitation with alkaline earth phosphates by digesting the gross urine sample in 10% 

nitric acid. The actinides were coprecipitated with calcium phosphate by neutralizing the acid solution 

with ammonia. The precipitate was washed, planchetted, and counted for alpha radiation with a 

proportional counter.

Appendix G lists the sensitivities by bioassay code for UST and the other laboratories. These were 

assumed by the SSFL bioassay program to be the defaults. If another value was reported by the 

laboratory (for example, “<2.2 dpm/day”), that value was taken to be the sensitivity (reporting level). 

Dose reconstructors should refer to Attachment G for a comprehensive list of the bioassay codes and 

the SSFL default isotopes and sensitivities for each code during the site’s history. See Section 5.6 for 

guidance on determining MDAs from the sensitivities in Attachment G.

5.3.1.5 1989 to 2005

By 1989, all R&D activities had ended. All work with radioactive materials since then has been in 

conjunction with ongoing D&D activities. Controls for Environmental Pollution (CEP), International 

Technology (IT) (formerly UST), Eberline, and Teledyne (now Teledyne Brown Engineering) were the 

main bioassay vendors. CEP performed bioassay analyses in 1991, 1992, and 1993. Sample results 

from CEP are considered invalid and these results should not be used for determination of intake. 

However, if results of samples analyzed by CEP are found in a worker’s record, the worker should be 

considered as monitored for the purposes of assigning dose. No analytical procedures were 

recovered for this period, but they were probably based on the earlier methods discussed above. The 
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 23 of 82

collection methods, etc., appear to be very similar to those used in previous years. The routine 

sampling strategy was to use generic screening urinalyses (gross alpha and gross beta minus 40K). If 

positive results were found, dose assessments were performed using default radionuclide mixtures or 

more specific analyses were performed (Barnes 1999). Periodic in vivo measurements were 

performed to detect intakes from gamma emitters likely to be present at SSFL (137Cs, 60Co, 154Eu, 

etc.).

Particle size measurements were generally not performed in conjunction with routine workplace 

monitoring. Worker breathing-zone air samples were collected (Barnes 1999), but it is unclear if this 

information was incorporated into individuals’ dosimetry records.

Since 1992, bioassay results returned from the vendor have been entered into a computer database 

(Boeing Canoga Park Internal Dosimetry Tracking System). A database with results back to about 

1960 was developed from results hand-entered for an epidemiology study (Boice 2006). In general, 

the entire sample was not used for the requested analysis; therefore, multiple results for one sample 

are possible.

According to the health physicist responsible for internal dosimetry in 2005, no positive results have 

been reported since 1998. The sensitivities required for each analysis were published (Barnes 1999) 

and are listed in Table 5-6 in Section 5.6.

5.3.2

The methods for individual radionuclides are covered for the specific periods under the sample types 

(i.e., urine and fecal).

In Vitro Methods for Individual Radionuclides

5.3.3

Although fecal sampling was mentioned both as a routine and a special bioassay method in site 

documents, little detail has been found about the analytical methods used. In 1967 all bioassay 

samples were wet ashed with nitric acid and hydrogen peroxide. A salt fusion was added for fecal 

samples to ensure recovery of all radionuclides. An aliquot was apparently analyzed using methods 

similar to those discussed for urine sampling. 

Fecal Sample Analysis

Detection limits for various isotopes in fecal samples were found for 1967. From 1975 to 1988, when 

the number of total and positive tests was well documented, only a few fecal samples were 

mentioned, but this might have been because this series of reports did not always make the 

distinction between fecal and urine sampling, that latter of which was by far the more common 

bioassay medium.

In 1967 bioassay analyses were performed on fecal samples for a case of suspected promethium 

exposure (Alexander 1967b). Fecal samples for uranium were also apparently taken in conjunction 

with the investigation into the uranium aluminide exposures in 1967 (see Attachment A). Tables of 

total samples and positive results for 1975 and 1976 show analysis type codes for uranium fecal 

samples (F-UF and F-UR), but it appears that no such analyses occurred in those years (Hart 1979, 

1980a). In 1982, positive PUA fecal results were reported for two individuals (one sample each) 

(Tuttle 1985).

Due to the lack of information found, the urine MDAs for gross alpha or gross beta listed in Table 5-3 

in Section 5.6 should be converted to dpm/day and used for fecal analyses before 1967 if the 

sensitivity was not reported by the laboratory [8]. From 1967 on, the values in Table 5-4 in Section 

5.6 should provide estimates favorable to claimants because detection limits probably were lower with 

time [9].
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 24 of 82

Table 5-1 lists the frequencies for in vitro monitoring.

Table 5-1. Internal dose control program (in vitro).

Routine monitoring type Period Frequency

Urine, single voida 1958–1988 Quarterly, semiannual, annual, based on job

Urine, single void, H-3 1958–1966 Weekly, based on job

a. Sample requested on Friday; first voiding on Monday morning requested. Positive samples were 

verified and followed with at least one 24-hr sample collected on Sunday.

5.4 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.4.1

Although Helgeson Nuclear Services regularly visited the site to conduct chest counts for 235U (see 

Section 5.3.2), WBCs for fission or activation products were apparently not part of the routine 

bioassay program at SSFL. Between 1975 and 1988, only 25 counts on 25 individuals were 

summarized in annual reports. In the same period, 385 chest counts were completed. All WBCs 

were reported positive for 137Cs [10]. Ten counts were performed in 1977, and 15 were performed in 

1979. The counts were probably performed when intakes were suspected but could have been 

project-ending counts. Although only 137Cs was reported in the summaries, undoubtedly a wide 

spectrum was scanned that would have detected other high-energy gamma emitters such as 60Co 

with a reasonably low detection level [11]. Barnes (1999) stated that WBCs occurred “periodically,” 

but the meaning of that term is not clear.

Whole-Body Counting

5.4.2

In 1967, the first chest (lung) counts for uranium using a medical system were performed at UCLA. 

The 186-keV gamma ray from the decay of 235U was used to quantify the amount of EU in the lung by 

scintillation spectrometry (Tuttle 1986a). Calibration of this system was crude; it used a point source 

rather than a distributed source that would better simulate the radioactivity in a lung. The point source 

contained more 234U than the uranium fuel, so the activity-to-mass conversion was off by about a 

factor of 2. Moreover, no chest-wall thickness corrections were made. The first two corrections 

amounted to initial lung burdens that were about a factor of 4 too high. The correction for chest-wall 

thickness would have depended on each individual. This system was not used after 1967. It is 

uncertain to what extent corrections were made to individual employee records. Dose reconstructors 

could use these data for overestimates that would be favorable to claimants. For more realistic 

calculations, it would be best to rely on urine and fecal data if those data are available.

Chest Counting

Starting in 1968, Helgeson Nuclear Services provided lung-counting services using equipment and 

techniques specifically developed to measure lung deposits of uranium. The first counts in 1968 were 

performed with a 0.5-in. by 8-in. NaI(Tl) detector in the shadow shield used for WBCs. A crude 

Masonite phantom was used for calibration, and the results included a chest-wall thickness correction. 

The individual was raised on an air mattress until the chest was flush with the bottom of the detector. 

The counts took 40 min. The results were reported in milligrams of 235U ±2 sigma, with the uncertainty 

based on counting statistics alone (Helgeson 1968).

By 1977, two 5-in.-diameter, thin-window phoswich detectors were used, which provided a reduction 

in Compton scattered background by a factor of 4 in the 235U region. In addition, the counting 

chamber was fully shielded. These changes enabled reduction of the counting time to 20 min. The 

calibration was similar to that described for 1968 but was cross-checked with the phantom used at 

Oak Ridge National Laboratory (ORNL). The minimum sensitivity was reported to be 30 to 60 μg of 

235U and 1 to 3 mg of natural uranium or DU (Helgeson 1978, 1983). Calibration was performed with 

a phantom that contained a known amount of 235U and a known chest-wall thickness. The resultant 
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 25 of 82

gross pulse height spectrum was adjusted for background by subtracting an assumed background 

spectrum that consisted of a straight line passing through the gross count values just below and just 

above the pulse height spectrum region corresponding to 186 keV. The net count was adjusted for 

chest-wall thickness and converted to mass of 235U based on the calibration (Tuttle 1986a).

Until 1981 the results of lung count measurements were reported as zero if a result was below the 

2-sigma uncertainty based on counting statistics. The results were reported as the actual value if they 

were equal to or greater than the 2-sigma uncertainty. Starting in 1981 Helgeson Nuclear Services 

was asked to report all values regardless of the assigned uncertainty. An analysis of uncensored data 

along with counts of three individuals with no history of uranium exposure was reported in 1986 (Tuttle 

1986a). This analysis estimated that the Helgeson results were biased high by 32.5 μg of 235U when 

only natural levels were present. This suggested that the straight-line approach to subtracting 

background was not appropriate when measuring background subjects. At values approaching the 

maximum permissible lung burden (MPLB), the background subtraction method was less important. 

The MPLB for 235U was 245 μg. Dose reconstructors can consider the values reported by Helgeson 

as overestimates favorable to claimants. For a more realistic estimate, dose reconstructors should

adjust the lung count results by assuming a linear relationship of the bias between 0 and 1 MPLB as 

indicated in Attachment C. Above 1 MPLB, no adjustment in the results is recommended.

Helgeson apparently also performed special counts as requested by the site. These included counts 

for 241Am and might have included counts for plutonium and thorium. Where these counts appear in 

worker records, dose reconstructors should consider a less-than value as the reporting level (decision 

level) and the MDA should be taken as twice the less-than value [12]. Helgeson was purchased by 

Neutron Products around 2003.

Table 5-2 lists the frequencies for in vivo monitoring.

Table 5-2. Internal dose control program (in vivo).

Routine monitoring type Period Frequency

Chest count U-235 (UCLA) 1967–1968a Monthly as long as count indicated >1⁄2 of lung 

burden (0.02 μCi)

Chest count U-235 (ORNL) 1967b One time

Chest count U-235 (Helgeson) 1968–1983 (EU work ended 

in 1983)

Three times/yr, 14–21 personnel selected for 

measurement each time

WBC Cs-137 (Helgeson) 1977, 1979 Uncertainc

WBC (Helgeson/Neutron

Products)

1989–2005 Periodicallyd

a. Counts performed to follow up suspected enriched UAlx exposures in the powder room. The UCLA counter was not 

used after April 1968 (Tschaeche 1968a).

b. Counts of one employee as a cross-check on the UCLA calibration (Alexander 1967a).

c. It is unknown if these counts were routine or special.

d. Frequency not specified in Barnes (1999).

5.5 UNCERTAINTY

At SSFL the uncertainty for a single bioassay measurement was not reported consistently. Reviewed 

statements of work for bioassay services do not contain any specification for reporting uncertainty. 

UST results were apparently reported with a 2-sigma uncertainty if they were above the MDA. 

Bio-Science reported a 95% confidence interval (at least in some cases). The individual analysis 

results sheets (Figures B-5 to B-10, Attachment B) might show the uncertainty for individual 

measurements. Due to the calibration and other problems discussed above, the estimated error in 

the early UCLA lung-counting results for 235U was ±200% at 1 sigma. The stated uncertainty for the 

Helgeson lung counts was about ±25% at 1 sigma (Tschaeche 1968b).
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 26 of 82

5.6 DETECTION LIMITS

Urine results above the lower limit of detection (LLD), sensitivity, MDA, etc., were used to calculate 

the percentage of the MPBB received by the worker. The normalized result (1,500 mL/d) was divided 

by a standard excretion rate for 1 MPBB to produce this percentage. If possible, the worker followed 

with additional urine samples until the results were less than the laboratory LLD, which was the lowest 

value at which radioactivity was considered to be present with reasonable certainty (Tuttle 1989). No 

exact definitions of these terms have been found. In effect, the detection limit value was used as the 

reporting level. Most results appear to have been reported with an uncertainty of 2-sigma error or a 

95% confidence interval. If the uncertainty included zero, the result was considered to be background 

(i.e., no radioactivity detected in the sample). Assuming the uncertainties in the background and the 

sample were equal, the detection limits published in site documents were closer to decision levels 

[13]. That is, there is only an approximate 5% chance that results at this level are really background 

results (false positives). The MDAs would be approximately twice these values to ensure that there is 

only an approximate 5% chance that results at the MDA would not be detected (false negatives). The 

detection limits recovered are listed in the tables as the reporting levels. Because the detection limits 

in the tables were mainly collected from contract documents, the reported sensitivity (less-than 

values) in the worker records should be used to determine the MDA in lieu of the values listed in these 

tables. If no sensitivity is reported, then dose reconstructors should use the MDA listed in Tables 5-3 

to 5-6.

Tables 5-3 to 5-6 list MDAs and reporting levels for periods corresponding to the bioassay methods 

discussed in Sections 5.3 and 5.4. The reporting levels are listed in the units quoted in the 

references, which are generally the units of the results. However, various volumes were used to 

report the results. In general, the excretions assumed when reporting per sample or per day were 

1,500 mL for urine samples and 135 g for fecal samples [14].

Table 5-3. Detection limits, 1958 to 1966.a

Radionuclide

Method/

description MDAb Reporting levelc

Gross alpha Urine 15 dpm/L (U. S. Nuclear) 7.5 dpm/L

Gross beta Urine 150 dpm/L (U. S. Nuclear) 75 dpm/L

Gross beta (minus K-40) Urine 0.4 dpm/mL (Bio-Science)

0.04 dpm/mL (UST)

0.2 dpm/mL

0.02 dpm/mL

H-3 Urine 10,000 dpm/mL (Bio-Science)

10,000 dpm/mL (UST)

5,000 dpm/mL

5,000 dpm/mL

MFP (gross) Urine 0.2 dpm/mL (Bio-Science)

0.04 dpm/mL (UST)

0.1 dpm/mL

0.02 dpm/mL

Po-210 Urine 0.02 dpm/mL (Bio-Science)

0.02 dpm/mL (UST)

0.01 dpm/mL

0.01 dpm/mL

Plutonium Urine 0.0006 dpm/mL -(UST) 0.0003 dpm/mL

Sr-90 Urine 0.2 dpm/mL (BioScience)

0.04 dpm/mL (UST)

0.1 dpm/mL

0.02 dpm/mL

Thorium Urine 0.0004 μg/mL (Bio-Science)

0.0004 μg/mL (UST)

0.0002 μg/mL

0.0002 μg/mL

Uranium UF-1A

Urine

0.004 μg/mL (U. S. Nuclear)

0.0002 μg/mL (Controls for Rad.)

0.0004 μg/mL (Bio-Science)

0.0004 μg/mL (UST)

0.002 μg/mL

0.0001 μg/mL

0.0002 μg/mL

0.0002 μg/mL

Uranium (enriched) UR-1B

Urine

15 dpm/L (U. S. Nuclear)

0.012 dpm/mL (Bio-Science)

0.012 dpm/mL (UST)

7.5 dpm/L

20.006 dpm/mL

0.006 dpm/mL

a. The date of the reference to the reporting levels, which are the best information available for 1958 to 1966.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 27 of 82

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (less-than values) in the worker 

records should be used to determine the MDA in lieu of the values listed in this table. These values should 

also be used for the in-house laboratory.

c. U. S. Nuclear from Shepard (1959), Controls for Radiation from O’Brien (1959), NSEC from Edelmann 

(1959) and NSEC (1957), Bio-Science from Lee (1963), and Tracerlab from Tracerlab (1959). UST results 

that were considered to be less than detectable generally have the reporting level included with the result.

Table 5-4. Detection limits, 1967 to 1974.

a,b

Radionuclide

Method/

description MDAc Reporting leveld

Ca-45 Urine 10 dpm/mL 5 dpm/mL

Ca-45 Urine 0.06 dpm/mL 0.03 dpm/mL

Ca-45 Feces 30 dpm/sample 15 dpm/sample

Cs-137 FP3b

Urine

60 dpm/sample 30 dpm/sample

Gross alpha Urine 0.02 dpm/mL 0.01 dpm/mL

Gross alpha Urine 0.010 dpm/mL 0.005 dpm/mL

Gross alpha Feces 4 dpm/sample 2 dpm/sample

Gross beta (less K-40) Urine 0.08 dpm/mL 0.04 dpm/mL

Gross beta Feces 4 dpm/sample 2 dpm/sample

H-3 Urine 10,000 dpm/mL 5,000 dpm/mL

I-131 Urine 0.12 dpm/mL 0.06 dpm/mL

I-131 Urine 0.4 dpm/mL 0.2 dpm/mL

MFP (gross) Urine 0.04 dpm/mL 0.02 dpm/mL

MFP (gross) Feces 16 dpm/sample 8 dpm/sample

Plutonium Urine 0.8 dpm/sample 

(estimated)

0.4 dpm/sample 

(estimated)

Plutonium Radiographic

Urine

0.0006 dpm/mL 0.0003 dpm/mL

Plutonium Procedure A

Feces

2 dpm/sample 1 dpm/sample

Plutonium Procedure B

Feces

4 dpm/sample 2 dpm/sample

Polonium Urine 0.02 dpm/mL 0.01 dpm/mL

Polonium Feces 6 dpm/sample 3 dpm/sample

Pm-147 Urine 0.10 dpm/sample 0.05 dpm/sample

Radium Urine 0.010 dpm/mL 0.005 dpm/mL

Radium Feces 20 dpm/sample 10 dpm/sample

Sr-90 Urine 0.04 dpm/mL 0.02 dpm/mL

Sr-90 Feces 16 dpm/sample 8 dpm/sample

Sr-90 FP3a 

Urine

60 dpm/day 30 dpm/day 

Thorium Urine 0.002 μg/mL 0.001 μg/mL

Thorium Feces 6 μg/sample 3 μg/sample

Uranium UF Urine 0.006 μg/mL 

(10-mL volume)

0.003 μg/mL 

(10-mL volume)

Uranium UF Urine 0.0004 μg/mL

(100-mL volume)

0.0002 μg/mL

(100-mL volume)

Uranium UF Feces 4 μg/sample 2 μg/sample

Uranium (enriched) UR Urine 0.012 dpm/mL 0.006 dpm/mL

Uranium (enriched) UR Feces 4 dpm/sample 2 dpm/sample

U-235 IVLC (UCLA) 0.8 mg 0.4 mg

U-235 IVLC (Helgeson) 60–120 μg (depends on

chest-wall thickness)

30–60 μg

a. IVLC = in vivo lung count.

b. The date of the reference to the reporting levels, which are the best information available for 1967 

to 1974.
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c. Assumed to be twice the sensitivity (see text). The reported sensitivity (less-than values) in the 

worker records should be used to determine the MDA in lieu of the values listed in this table.

d. U-235 IVLC (UCLA) from Saxe (1967a). In vitro values are primarily from UST contract 

documents (Kellehar 1966b, 1967; Spielman 1968; Bales 1969; Staszesky 1971).

Table 5-5. Detection limits, 1975 to 1988.

a

Radionuclide Method/description MDAb Reporting levelc

Cs-137 FP3b

Urine

60 dpm/sample 30 dpm/sample

Cs-137 WBC 4 nCi 2 nCi

Plutonium PUA

Urine

0.0990 dpm/sample 0.0495 

dpm/sample

Sr-90 FP3a 

Urine

60 dpm/d 30 dpm/d 

U-235 IVLC (Helgeson) 60–120 μg (depends on 

chest wall thickness)

30-60 μg 

Uranium (total) UF 

Urine

0.60 μg/d 0.30 μg/d

Uranium (enriched) UR 

Urine

7.5 dpm/d 3.75 dpm/d

a. The date of the reference to the reporting levels, which are the best information available for 1975 

to 1988.

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (less-than values) in the 

worker records should be used to determine the MDA in lieu of the values listed in this table.

c. Sr-90, U (total), and EU (1987, 1988) from Tuttle (1988b, 1989).

Table 5-6. Detection limits, 1989 to 2005.a

Radionuclide

Method/

description MDAb Reporting levelc

Gross alpha Urine 15 dpm/1,500 mL 7.5 dpm/1,500 mL

Gross beta Urine 30 dpm/1,500 mL 15 dpm/1,500 mL

Gross beta, corrected for K-40 Urine 45 dpm/1,500 mL 22.5 dpm/1,500 mL

Gamma scan Urine 60 dpm/1,500 mL 30 dpm/1,500 mL

Sr-90 Urine 6 dpm/1,500 mL 3 dpm/1,500 mL

UF Urine 15 μg/1,500 mL 7.5 μg/1,500 mL

Uranium alpha spectroscopy Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Thorium alpha spectroscopy Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Plutonium alpha spectroscopy Urine 3 dpm/1,500 mL 1.5 dpm/1,500 mL

Am-241 Urine 0.90 dpm/1,500 mL 0.45 dpm/1,500 mL

H-3, distillation Urine 3000 dpm/1,500 mL 1,500 dpm/1,500 mL

H-3, electrolytic enrichment Urine 60 dpm/1,500 mL 30 dpm/1,500 mL

WBC (Co-60) WBC 4 nCi 2 nCi

a. The date of the reference to the reporting levels, which are the best information available for 1989 

to 2005.

b. Assumed to be twice the sensitivity (see text). The reported sensitivity (less-than values) in the 

worker records should be used to determine the MDA in lieu of the values listed in this table.

c. Sensitivity reported in Barnes (1999, Table D-1).

5.7 SAMPLE KIT INFORMATION

Table 5-7 lists sample kit information summarized from various site documents.

Table 5-7. Excreta sample media and types.

a
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Media Sample description

Urine Routine samples. Single voiding collected on Monday morning before returning to work. Collected in 

single bottle. Also referred to as rate samples. Positive samples were followed up with additional 

rate sample for verification. Time of sample collection and previous voiding were recorded. Normally 

collected in a 16-oz container. By 1999, 1-L polyethylene containers were used. It appears that one 

full container (900–1,000 mL) was collected and collection time noted.

Urine 24-hr samples. Used for follow-up to verify rate samples or for incidents. One or more samples 

could be requested. Single samples were collected at home on Sunday. Collected in 32-oz or 1-L 

polyethylene containers.

Urine Spot samples. For follow-up to incidents, spot sample could be collected as soon as possible. This 

sample was probably collected in 16-oz container.

Feces Could be requested in conjunction with urine samples as follow-up to incidents. No descriptions of 

historical fecal kits were located. By 1999, single voiding samples were collected in 83-oz 

polyethylene containers. The minimum mass that was considered adequate was 30 g.

a. No codes to identify sample kits were used at SSFL. Requests were color-coded for a time, but these do not appear in 

the individual records.

5.8 SOLUBILITY TYPE, ACTIVITY FRACTION, AND PARTICLE SIZE BY FACILITY

In the absence of any measurements or studies, NIOSH guidance requires the use of default solubility 

classes and particle size values from the International Commission on Radiological Protection 

(NIOSH 2002, pp. 15–16). With one exception, facility-specific solubility and particle size data for 

SSFL have not been found. Activity fractions were not available with the exception of those for limited 

fuel fabrication operations. For research reactors using highly enriched fuels, values from Shleien, 

Slaback, and Birky (1998) have been listed as an approximation. In some cases, default assumptions 

from Barnes (1999, Table D-3) have been used to estimate the activation products using a ratio to the 

fission products. Table 5-8 lists this information.

The behavior of UAlx in the powder room workers was found to differ from the standard ICRP 

absorption types. A study in 2005 (Leggett, Eckerman, and Boice 2005) developed a new model that 

fits the excretion data. This model cannot be directly implemented in IMBA. Calculations performed 

using other software indicate that the standard ICRP 66 models (types F, M and S) with IMBA will 

generally provide favorable to claimant dose estimates for chronic intakes. For acute intakes where 

the application of types F, M, and S all yield a POC<52%, alternate software will be required to model 

the behavior of UAlx in the lungs. In this case, contact the Principal Internal Dosimetrist for 

assistance. See Attachments A and H for additional details.

Several non-nuclear energy research programs were conducted in Area IV. The primary non-nuclear 

R&D activities were performed at the Liquid Metal Engineering Center, which became ETEC. The 

facilities that performed non-nuclear research are not included in Table 5-8 unless they were also 

used for nuclear research (Sapere and Boeing 2005, p. 2-12). 

In the 1960s, a test was conducted that involved dropping a 1-kg DU slug from a helicopter. The slug 

was apparently never recovered. No contamination was found in the area (Sapere and Boeing 2005 

p. 2-13). It is unlikely that this event would contribute significantly to the internal dose of 

occupationally exposed workers, and it has not been included [15].
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 30 of 82

Table 5-8. Solubility type, activity fraction, and particle size by facility.

a

Facility

b Compoundc RNd

Particle 

size

e

(μm AMAD) Activity fraction

f

Downey Plant

Building 001

WBNS reactor

1948–1955

Canoga Avenue Facility

Building 038 L-47, L-77 reactors

1956–1960

De Soto Facility

Building 104 L-77 reactor

1960–1979?

Building 4073 

Area IV, SSFL

KEWB reactor

1956–1966

Building 4093 L-85 reactor

1956–1980

Homogeneous water boiler-type reactors – 93%-

enriched uranyl sulfate solution in H

2O. Reactors 

operated at low power levels (4-10 W).

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

De Soto Facility

Building 101

Powder room & new powder 

room 1966–1968 & 1979–

1982

Uranium aluminide fuel fabrication –alloy of uranium 

and aluminum. See Attachment A for information on 

solubility. See Attachment H for the impact of the 

uranium aluminide-specific model on intake and dose 

calculations.

U-2341 See Table 5-9

U-2351 See Table 5-9

U-2361 See Table 5-9

U-2381 See Table 5-9

Building 4143

SRE

1957–1964

Decontaminated

1974–1983

Storage facility

1983–1999

Demolished

1999

High-temperature, sodium-cooled, graphite-

moderated EU reactor (site of loss-of-coolant 

accident in 1959) – unalloyed uranium metal 

thermally bonded by NaK in stainless-steel tubes.

1957-July 1959, core was 2.78% EU

September 1960, 2nd core began operation. Core 

was 7.6 % (weight) Th-232 with 93% EU (AI 1959)

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 31 of 82

Facility

b Compoundc RNd

Particle 

size

e

(μm AMAD) Activity fraction

f

Eu-1545

Th-2325

H-3 NA

Building 4010

SER

1959–1960

S8ER

1963–1965

Building 4024

S2DR

1961–1962

S10FS

1965–1966

SNAP TTF

1971

Building 4028

STR

1961–1964

STIR

1964–1972

Building 4059

S8DR

1968–1969

SNAP reactors – “fully enriched” uranium dispersed 

in zirconium hydride fuel rods

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545

H-3 NA

Building 4012

SNAP CTF

1962–1968

HMRFSR

1970–1972

Building 4373

SNAP CTF

1957–1963

Building 4019

SNAP FSCF

1964–1965

Critical test facilities – SNAP development test 

facilities (fully EU dispersed in zirconium hydride fuel 

rods)

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Building 4009

OMR CF

Critical test facilities – civilian nuclear power test 

facilities, EU low-power reactors. HEU and DU used 

and stored also.

Sr-905 See ORAUT (2007)

Cs-1375

U-2345
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 32 of 82

Facility

b Compoundc RNd

Particle 

size

e

(μm AMAD) Activity fraction

f

1958–1967

SGR CF

1958–1967

High-energy rate forging

1980s

DU storage

Early 1990s

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545

Building 4100

AETR test facility

1960–1974

Radiation Instrument 

Calibration and Radiological 

Sample Counting Laboratory

1985–present

Critical test facilities – thorium- and uranium-fueled 

reactors – 20 different reactor core configurations 

studied. The AETR’s first 9 core configurations 

(through 1965) contained various amounts of U-233 

and Th-232 and were driven by 93% EU fuel 

(Mountford and Morewitz 1965). The activity 

fractions are the medians of the cores studied. Only 

very small amounts of the other radionuclides listed 

were likely to have been produced, but they were 

listed as radionuclides of concern by Boeing.

Sr-905

Cs-1375

U-2335 9.81E-01

U-2345 1.83E-02

U-2355 5.71E-04

U-2385 6.63E-06

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Eu-1525

Eu-1545

Th-2325 1.05E-03

H-3 NA

Bldgs. 4003

4163, 4041, 

4654, 4689,

4653, 4606,

4773

SRE Support Complex

1954–1964

Analysis of SNAP fuel burn-

up & irradiation experiments

1965–1973

Nuclear support operations – SRE fuel assembly 

(uranium and thorium metal slugs), contained 

radioactive “hot cave,” tanks, hoods, and lines until 

1975. SRE fuel loaded in 1960 was 7.6 % (weight) 

Th-232 with 93% EU (AI 1959)

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 33 of 82

Facility

b Compoundc RNd

Particle 

size

e

(μm AMAD) Activity fraction

f

Th-2325

Building 4005

Uranium Carbide Fuel Pilot 

Plant

1958-1993

Nuclear support operations – pilot plant for uranium-

carbide fuel production in 1966-1977, first using DU 

and then EU

U-2345

U-2355

U-2385

Building 4011

Radiation Instrumentation 

Calibration Laboratory

1984–1996

Nuclear support operations – sealed sources. In 

1960, a radioactive liquid spill from OMRE shipping 

cask occurred west of building.

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545

Building 4030

Van deGraaff Accelerator

1960–1964

Nuclear support operations – neutron source, used 

H-3 target

H-3 NA1

Building 4020

Hot Laboratory

1957–1988

D&D Completed

1999

Nuclear support operations – used to examine fuel 

from SRE, SER, S2DR, S8DR, S10FS-3, and 

outside reactors. Fuel disassembled or separated 

from cladding.

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545

Pm-1475

Building 4023

Liquid Metals

Component

Nuclear support operations – tests with sodium loops 

containing radioactive contaminants 

Sr-905

Cs-1375

Co-605
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 34 of 82

Facility

b Compoundc RNd

Particle 

size

e

(μm AMAD) Activity fraction

f

Test Building

1962–1986

Eu-1525

Eu-1545

Building 4029

Radioactive Measurement 

Facility

1959–1974

Nuclear support operations – leaking calibration 

source contaminated source well

Ra-22651

Building 4055

NMDF

1967–1979

Nuclear support operations – development work 

involving Pu

DU work 1967

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Building 4064

Fuel Storage Facility

1958–1993

D&D in 1997

Nuclear support operations – secure storage of 

unirradiated fuel (EU and Pu); after removal of 

fissionable material in mid-1980s, radioactive waste 

(soil) stored until 1993.

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545

Bldgs. 4021, 

4022

RMHF

1959–present

Nuclear support operations – radioactive waste 

processing from onsite programs; 2005 storage area 

from D&D activities

Sr-905 See ORAUT (2007)

Cs-1375

U-2345

U-2355

U-2385

Pu-2385

Pu-2395

Pu-2405

Pu-2415

Am-2415

Co-605

Eu-1525

Eu-1545
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

04/

26/2010 Page 35 of 82

Facilityb Compoundc RNd

Particle 

sizee

(μm AMAD) Activity fractionf

H-3 NA

Building 4363

Mechanical Component 

Development & Counting

1956–1963

Nuclear support operations – in 1962, work done on 

contaminated component from SRE accident

Sr-90 5 0.5

Cs-137 5 0.5

17th Street Drainage

1959?–1999

Areas of known contamination – runoff from SNAP 

facilities

Cs-137 5 1.0

OCY

Late 1960s–late 1970s

Areas of known contamination – spill detected in 

1976 (MFPs suspected)

Sr-90 5 0.5

Cs-137 5 0.5

Building 4886

Former Sodium Disposal 

Facility

1956–1978

Areas of known contamination – storage drums 

contaminated with residual radioactivity. SRE fuel 

loaded in 1960 was 7.6 % (weight) Th-232 with 93% 

EU (AI 1959)

Sr-90 5

Cs-137 5

U-234 5

U-235 5

U-238 5

Pu-238 5

Pu-239 5

Pu-240 5

Pu-241 5

Th-232 5

H-3 NA

a. The Acronyms and Abbreviations list spells out the terms in this table.

b. Facilities were combined for this analysis if they were similar and had a common list of radionuclides of concern.

c. If chemical compounds were not available, the best description found is listed.

d. RN = radionuclide; radionuclides are the radionuclides of concern from Sapere and Boeing (2005, Tables 2-1, 2-2 and Section 2.0). Radionuclides were not included 

if they were an activation product limited to steel and rebar, had a short half-life, were a naturally occurring isotope consistent with natural background, or had not 

been observed in soil measurements. This list may be expanded to include radionuclides that were likely present during operations at the reactor facilities using 

ORAUT (2007).

e. Defaults used except for the powder room.

f. Activity fractions for fission products calculated from available information for non-reactor facilities. ORAUT (2007) provides guidance on the assignment of 

radionuclide-specific intakes of mixed fission and activation products when air sampling or urinalysis data associated with reactors or reactor fuels are available only 

as gross or total beta activity or gross or total gamma activity.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 36 of 82

5.9 FACILITY-SPECIFIC RADIONUCLIDE CONVERSIONS

SSFL fabricated fuel for the ATR and the Engineering Test Reactor (ETR) in 1966, 1967, and 1968 at 

the De Soto Facility in Room 1110-62, Building 101 (known at the time as Building 001); Attachment 

D contains details. A second fuel fabrication campaign for the ATR was completed in 1979 and 1980; 

Attachment E contains details. These are the only two operations for which facility-specific data were 

made available. These data are summarized in Table 5-9.

Table 5-9. Facility-specific radionuclide conversions.

Process description

Activity per unit mass (Bq/g uranium)

U-234 U-235 U-236 U-238

EU fuel fabrication, Room 1110-62 (powder room), Building 

101, De Soto Facility, 1966–1967

1.35E+06 7.44E+04 7.31E+03 7.50E+02

ATR EU fuel fabrication, 1979–1982 Building 101, De Soto 

Facility

2.37E+06 7.44E+04 1.14E+04 6.79E+02

5.10 RADON

For dose reconstruction under EEOICPA, occupational radon exposure is exposure to radon 

emanating from sources other than those that occur naturally in the area. Dose reconstructors must 

subtract the natural background level of radon exposure from any measured values when assessing 

occupational exposure (NIOSH 2002, p. 32). SSFL was not a processing or storage location for large 

quantities of 226Ra or 222Rn. Radon measurements were made from December 1989 to February 

1990 in several facilities; none of the areas measured exceeded 2 pCi/L, and all but one was less 

than 1 pCi/L [18].

5.11 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

This statement is based on the fact that industrial radiography sources are sealed sources 

which have been subject to leak testing from the earliest days of their use.

[2] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

By definition, unrestricted use limits are protective of the general public.

[3] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

This observation is based on the descriptions of the bioassay program in several site 

documents. It appears to have been formalized in Garcia (1963) if not sooner.

[4] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

Examples of these counts were observed in the site records. See Section 5.4.1.

[5] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

See Barnes (1999) for a description of the internal dosimetry program in the later years.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 37 of 82

[6] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

This statement is based on the assumption that all of the early laboratories had essentially the 

same technical capabilities.

[7] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

Ten percent of the amount indicating one MPBB. See the description in Section 5.2.

[8] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

This method provides a reasonable workaround for fecal samples where the sensitivity is not 

reported.

[9] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

This statement is based on the observation that the radiochemistry techniques for analyzing 

bioassay samples have improved over time and counting equipment has become better and 

more reliable.

[10] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

During the time frame discussed, it was common to detect 137Cs in individuals exposed only to 

radioactive fallout from weapons tests. This statement does not imply that the results were 

necessarily due to site activities.

[11] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

It is a common practice when performing gamma spectroscopy to identify all prominent peaks 

in the energy spectrum. The sensitivity statement is based on the fact that 60Co has high 

energy gamma rays that are readily detected in whole body counting systems.

[12] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

See section 5.6 for the rationale behind this statement.

[13] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

The level at which radioactivity is determined to be present with reasonable certainty is termed 

the “decision level” or “critical level.”

[14] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

Determined from an inspection of various site documents.

[15] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

The contamination, if any, resulting from this incident is likely confined to a relatively small 

subsurface area.

[16] Potter, Eugene. M. H. Chew & Associates. Consultant Health Physicist. October 2007.

Results are from a DOE indoor radon study of multiple facilities, reported in 1991.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 38 of 82

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Rocketdyne Division, Canoga Park, California, July 5. [SRDB Ref ID: 19141]

Tuttle, R. J., 1988b, Annual Review of Radiological Controls – 1987, Rockwell International, 

Rocketdyne Division, Canoga Park, California, July 28. [SRDB Ref ID: 19140]

Tuttle, R. J., 1989, Annual Review of Radiological Controls – 1988, Rockwell International, 

Rocketdyne Division, Canoga Park, California, May 12. [SRDB Ref ID: 19139]
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 44 of 82

GLOSSARY

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or zirconium). 

Cladding promotes the transfer of heat from the fuel to the coolant and contains fission 

products and activation products that result from the fissioning of the fuel.

core

That part of the reactor consisting of the fuel and some of the control elements for reactor 

operation.

dosimetry

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

from external or internal sources of radiation.

fission

A nuclear transformation characterized by the splitting of a nucleus into at least two other 

nuclei and the release of a relatively large amount of energy.

fission product

Radionuclides resulting from fission.

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

in vitro

Literally, in glass; outside the living body and in an artificial environment; internal bioassay 

sampling, such as fecal samples or urine samples.

in vivo

Literally, in the living; in the living body of a plant or animal; bioassay sampling by whole-body 

counting.

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but a 

differing number of neutrons (different mass number).

lower limit of detection (LLD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called limit of detection and detection limit or level. 

natural uranium

Uranium that has not been through an enrichment process.

radiation

Energy transferred through air or some other media in the form of particles or waves (see 

ionizing radiation).

radioactivity

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

neutrons from unstable atoms.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 45 of 82

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus specified by 

atomic number (the number of protons) and the mass number (equal to the number of protons 

plus neutrons).

type

The rate of absorption from lung to blood of inhaled radioactive materials and includes types F 

(fast), M (moderate), and S (slow).

U.S. Atomic Energy Commission (AEC)

An agency established by the U.S. Government for oversight of nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for nuclear 

fuel. It is sometimes alloyed in small amounts in the fuel itself.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 46 of 82

SSFL began fabricating reactor fuel elements in the fall of 1965. SSFL fabricated fuel for the ATR 

and the ETR from 1966 to 1968 at the De Soto Facility. In Room 1110-62 of Building 101 (known at 

the time as Building 001), briquettes of an alloy of 93% EU and aluminum known as uranium 

aluminide (UAlx) were formed in an electric arc melting furnace. These briquettes were crushed to 

form a powder, which was cold-pressed into compacts that became the cores of the fabricated fuel 

plates. The room where these activities took place was known as the powder room. The work was 

performed at six workstations: Weighing station 1, a melting station, a crushing and sieving station, 

weighing station 2, a compact-forming station, and a deburring station. The operations were 

performed with containment thought to be appropriate for the type of operation and the hazard of the 

material. From one to four air samplers operated in the room near the workstations.

In 1967 after 15 mo of operation, urine bioassay data indicated that the material was probably 

insoluble and, therefore, that the air activity was not being compared to the appropriate MPC. 

Although the uranium was more than 93% 235U by weight, 234U accounted for more than 96% of its 

activity. The insoluble MPC for 234U was a factor of 6 lower than the soluble MPC. This led to the 

conclusions that the regulatory standard (weekly average MPC) had been exceeded on a number of 

occasions and that equipment and procedures for controlling the airborne uranium were insufficient.

An internal investigation determined that the primary reason for the ineffective confinement of the 

uranium was leakage from the crushing glovebox seal, from the fume hood of weighing station 2, and 

from the arc furnace. Temporary measures were put in place until more permanent controls could be 

implemented. These changes were put in place and the project was completed. Along with the 

engineering changes, workers were required to wear full-face respirators and lapel air samplers (Saxe 

1967b). At some point, operations were relocated to the new powder room.

Of 83 wk of operation, the MPC for insoluble 234U (1 × 10-10 μCi/cm3

) was exceeded in 33 wk. Twentyone personnel who had worked in the powder room were monitored by chest counting. Three of 

these personnel were determined to have exceeded the MPBB and received work restrictions. The 

results for the other workers ranged from no detectable intake to less than 1 MPBB.

On August 15, 1967, a report of regulatory violations was submitted to the AEC. The report included 

the corrective measures taken (Sax 1967b). Inquiries by the AEC led to a review of an earlier fuel

fabrication project, the Japan Atomic Energy Research Institute (JAERI), which took place in 

Room 1110-061. This operation involved uranium metal plates enriched to 20% by weight. The 

activity of the material was still approximately 90% 234U. This operation did not involve crushing, but 

deburring the edges of the plates occurred during the process. This apparently took place outside 

ventilated enclosures until January 1967. An investigation into the JAERI fuel operation led to the 

conclusion that the average weekly concentration for insoluble 234U had been exceeded six times 

during the operation in November and December 1966. The soluble standard (6 × 10-10 μCi/cm3

) 

would not have been exceeded. The insoluble standard was used because of the experience with 

UAlx and the lack of evidence that uranium metal should be treated differently. Using this standard, 

five people had been exposed to concentrations greater than the AEC MPC. None of these people 

were involved in the ATR or ETR project (Remley 1967). Urinary excretion was measured on all five 

workers. The data were consistent with short-term exposures with small lung depositions. However, 

one of the workers initially showed a lung burden 1.5 times the MPBB. As stated in the main text, 

lung counting was its infancy during this time. The estimate for this worker was later reduced to 

0.5 MPBB, based on the difference in enrichments between the JAERI material and the calibration 

source for the lung counter at UCLA.

ATTACHMENT A

SUMMARY OF URANIUM FUEL FABRICATION INTERNAL DOSIMETRY ISSUES

Page 1 of 2
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 47 of 82

Air samples collected before 1966 were reviewed, and no concentrations exceeded 1 × 10-10 μCi/cm3

(Saxe 1967c). A particle size study consisting of two general area air samples from the powder room 

indicated that the particles were less than 1-μm count median diameter (CMD) (Alexander 1967c). 

The actual CMD and geometric standard deviation were not provided in the reference; therefore, the 

AMAD could not be determined. In the only other particle size reference found, Baurmash (1967) 

measured the particle size distribution for a UO2 grinding operation in the “processing room in Bldg. 

1.” The CMD reported for this operation was 0.195 μm with a geometric standard deviation of 1.66. 

The mass median diameter was calculated as 0.42 μm. Using a density of 10.97 g/cm3 for UO2, the 

AMAD for this aerosol is 1.39 μm. While this information is extremely limited, it points out that small 

particle sizes were at least possible during uranium fuel fabrication. Therefore, dose reconstructors 

should consider a 1-μm AMAD particle size for these operations in addition to the default assumption.

A study of the solubility of UAlx in simulated lung fluid was completed by SSFL in 1968. While the 

original study has not been found, a reference to it indicates that UAlx was about 10 times more 

insoluble than the most insoluble uranium compound (U3O8) measured by ORNL (Alexander 1968b). 

ICRP (1979) lists U3O8 as a class Y compound. The quality of the SSFL study is uncertain. ICRP 

(1994) lists the most insoluble compounds of uranium as type S.

As part of an epidemiological study, doses from intakes of radionuclides were estimated for workers 

employed during a 52-yr period at Rocketdyne/SSFL. Based on this work, Leggett, Eckerman, and 

Boice (2005) summarized the behavior of UAlx in the powder room workers and described materialspecific parameter values of the International Commission on Radiological Protection (ICRP) Human 

Respiratory Tract Model (HRTM) (ICRP 1994) needed to fit the data. See Attachment H for an 

evaluation of the impact of the uranium aluminide-specific model on intake and dose calculations.

ATTACHMENT A

SUMMARY OF URANIUM FUEL FABRICATION INTERNAL DOSIMETRY ISSUES

Page 2 of 2
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 48 of 82

LIST OF FIGURES

FIGURE TITLE

B-1 Example of an AI Request for Film Badge and Bioassay Services form, 1963 .......................

PAGE

49

B-2 Example individual personnel McBee Keysort card, 1966 ...................................................... 50

B-3 Example Bioassay Data Sheet, front...................................................................................... 52

B-4 Example Bioassay Data Sheet, back ..................................................................................... 53

B-5 Example Bio-Science Laboratories UF bioassay data sheet .................................................. 54

B-6 Example Bio-Science Laboratories UR bioassay data sheet .................................................. 55

B-7 Example Bio-Science Laboratories UR bioassay data sheets showing bottle tags ................. 56

B-8 Example UST UF bioassay data sheet................................................................................... 57

B-9 Example UST UR bioassay data sheet .................................................................................. 57

B-10 Example UST MFP bioassay data sheet................................................................................ 58

B-11 Example AI Wound Monitoring Report ................................................................................... 59

Figure B-1 shows an example of an AI Request for Film Badge and Bioassay Services form. 

Figure B-2 shows the individual personnel McBee Keysort card. The following numbers correspond to 

the handwritten numbers on the card. The cards were set up for a 4-yr period (Kellehar 1966a).

1. Frequency

2. Week of collection

3. Date of each specimen collection, wound, and nasal smear

4. Indication of analysis results; degree of positive results in any year

5. Type of analyses required

6. Changes in work affecting analyses and/or specimen collection (reverse side)

7. Special bioassays (reverse)

This is the master record for internal dosimetry results. This lists the type of analysis and the per-day 

results of those analyses.

Date is the date the sample was obtained from the individual. Type lists the type of sample obtained. 

Method is a code to identify the procedure used. An incomplete record exists of the specific 

laboratory techniques associated with these methods. When results are listed as “<XXX,” XXX is 

used as the reporting level rather than the value listed in the table.

In general, unless otherwise noted, listed results are in dpm per day (for radiological data) or 

micrograms per day (for fluorometric or other chemical analyses). Results were standardized to a 

1,500-mL/d urine excretion rate or a 135-g/d fecal excretion rate. The results line up with the 

methods. Statistical errors were generally not

The back side of the Keysort card sometimes has additional bioassay data. No results are listed on 

this side of the card, but it can be useful for determining the work location of the individual and

possibly the type of work being done at the time.

provided but can be obtained by manual search of the 

source data.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 1 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 49 of 82

Figure B-1. Example of an AI Request for Film Badge and Bioassay Services form, 1963.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 2 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 50 of 82

Figure B-2. Example individual personnel McBee Keysort card, 1966. (Both sides are shown).

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 3 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 51 of 82

Special bioassays might also be listed on this side of the card; these generally suggest that the 

bioassay was taken in response to a specific incident. However, absence of bioassay data on this 

side of the card should not be interpreted as an indication that the individual was never involved in an 

incident.

Figures B-3 and B-4 show the front and back of the Bioassay Data Sheet. This sheet records 

bioassay results when a determination was made in the SSFL laboratory (it also appears to have 

been used to log in and document vendor results that were sent in tabular form). It was designed to 

be a multipurpose form for use with all types of analyses.

The general concept of the sheet was that all steps of a hand calculation could be performed by 

entering the appropriate data and performing the indicated mathematical operations. The calculations 

on this side of the page were preliminary; the sample results are listed on the second page of this 

form.

Some results are penciled in on the top right corner of the page. The calculations are relatively 

straightforward. They are typical for a radiochemical or radiological laboratory. Multiple count blocks 

were provided; it appears this was used to log in sequential counts to determine half-lives of the 

isotopes. Information blocks are self-explanatory.

Only those blocks on the form pertinent to the type of measurement were filled in, so the dose 

reconstructor can use this to determine the type of sample.

Figure B-5 shows an example Bio-Science Laboratories UF Bioassay Data Sheet. 

These are the record of entry reports from the vendor laboratory that processed urine and fecal 

bioassay results. These records should be correlated to the bioassay summary records on the 

McBee Keysort cards (Figure B-2). The individual's name is usually written in pencil in the upper left 

margin.

The Source Number probably refers to an identification code for SSFL, the Report Date is the date the 

results were available, and the number in the Patient block is the Health Physics ID Number used to 

identify the person. The meaning of the No. Spec. Type entry is not known at this time. It is 

sometimes blank.

Entry Date refers to the date that the sample was received for processing. In general, it is 1 to 2 d 

after the date the sample was obtained by Rocketdyne from the individual. This date is from the 

McBee Keysort card.

Results are generally reported in micrograms of uranium per 100 mL of samples. Positive samples 

usually have values prorated into per-day rates using 1,500 mL/d as a urine excretion rate and 135 

g/d for a fecal excretion rate.

Dividing the UR values by the UF values results in a specific activity for the sample. This specific 

activity can then be used to estimate the general enrichment level of the uranium to which the 

individual was exposed. For simplicity, for low intakes, the SSFL practice was to model the intake as 

pure 234U (which results in overestimation of true dose).

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 4 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 52 of 82

Figure B-3. Example Bioassay Data Sheet, front (used by onsite laboratory).

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 5 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 53 of 82

Figure B-4. Example Bioassay Data Sheet, back (used by onsite laboratory).

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 6 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 54 of 82

Figure B-5. Example Bio-Science Laboratories UF bioassay data sheet.

Figure B-6 shows an example of a Bio-Science Laboratories UR bioassay data sheet. Figure B-7 

shows two more examples that include bottle tags listing the exact times of sample collection. 

These are the record of entry reports from the vendor laboratory that processed urine and fecal 

bioassay results. These records should be correlated to the McBee Keysort card (Figure B-2). Two 

versions were used (see Figure B-7 for a comparison).

The individual's name is usually written in pencil in the upper left margin. The Source No. generally 

lists a sample number. The Report Date is the date the results were available. The number in the 

Patient block is the sample number to identify the person. The meaning of the No. Spec. Type entry 

is not known at this time.

Entry Date refers to the date that the sample was received for processing. In general, it is 1 to 2 d 

after the date the sample was obtained by Rocketdyne from the individual. This date is from the 

McBee Keysort card.

Background refers to the background on the alpha counter (generally counts per hour). Counter 

efficiency is the efficiency of the counter used, and Recovery is the efficiency of the chemical recovery 

during the chemical preparation of the sample.

Urine Vol., ml, is the amount of sample actually analyzed, and Gross is the counter results 

uncorrected for background.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 7 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 55 of 82

Figure B-6. Example Bio-Science Laboratories UR bioassay data sheet.

Net - 0.95 error is the net

Disintegrations/min is the net activity determined for the sample. In some cases, "XX dpm/day" is 

written in. This is the sample result prorated to a daily excretion rate using a 1,500-mL assumption for 

daily urine volume and a 135 g/d rate for feces.

count rate (gross – background) with a 95% statistical error indicated. If the 

error exceeds the net count, it is considered a nondetect. If the net exceeds the error, it is considered 

a detect. The basis for this calculation of error (i.e., count times, normal or Poisson, etc.) is not 

available.

Disintegrations are calculated by (Net CPM) ÷ (Counter Efficiency/100) × (Recovery/100) (assuming 

that the Counter Efficiencies are written in percentage; fractional "percentages" are more likely true 

efficiency fractions and would not need conversion from percentage).

Dividing the UR values by the UF values results in a specific activity for the sample. This specific 

activity can then be used to estimate the general enrichment level of the uranium to which the 

individual was exposed. For simplicity, for low intakes, SSFL practice was to model the intake as pure 

234U (which results in overestimation of true dose).

This is the vendor record for results for UF sampling from UST. The calculations are self-explanatory. 

See the previous discussion of the Bio-Science UR record for additional data.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 8 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 56 of 82

Figure B-7. Example Bio-Science Laboratories UR bioassay data sheets showing bottle tags.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 9 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 57 of 82

Specimen Number contains the individual’s Health Physics Identification Number, and a sample 

number was generally written in the center box.

Figure B-8 shows an example of a UST UF bioassay data sheet. 

Figure B-8. Example UST UF bioassay data sheet.

Date Received is the date the sample arrived at the laboratory. This was generally 1 to 2 d after the 

collection date. The actual date the sample was obtained is on the McBee Keysort card. These 

sheets should be correlated to that card.

Figure B-9 shows an example of a UST UR bioassay data sheet. 

Figure B-9. Example UST UR bioassay data sheet.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 10 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 58 of 82

This is the vendor record for results for UR sampling from UST. The calculations are self-explanatory. 

See the previous discussion of the Bio-Science UR record description for additional data.

Specimen Number contains the individual’s Health Physics Identification Number. A sample number 

was generally written in the center box.

Date Received is the date the sample arrived at the laboratory. This was generally 1 to 2 d after the 

collection date. The actual date the sample was obtained is on the McBee Keysort card. These 

sheets should be correlated to that card.

Figure B-10 shows an example of a UST MFP bioassay data sheet for MFPs. 

Figure B-10. Example UST MFP bioassay data sheet.

The calculations are self-explanatory. Specimen Number contains the individual’s Health Physics 

Identification Number. A sample number was generally written in the center box.

Date Received is the date the sample arrived at the laboratory. This was generally 1 to 2 d after the 

collection date. The actual date the sample was obtained is on the McBee Keysort card. These 

sheets should be correlated to that card.

These measurements were generally defaulted to 137Cs. In addition, one should be aware that 137Cs 

was generally assumed to be accompanied by 90Sr in a 50%/50% mix. Analyses of MFP results 

should therefore account both for 137Cs exposures and 90Sr exposures as being coincident.

Based on laboratory analyses directly comparing 137Cs and 90Sr values, the mixture ratio for 137Cs and 

90Sr during D&D operations after 1991 was approximately 85% 137Cs to 15% 90Sr. In 1999 the ratio 

published in the site TBD was 90% 137Cs to 10% 90Sr for SSFL locations other than Building 4059.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 11 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 59 of 82

Figure B-11 shows an example of an AI Wound Monitoring Report.

Figure B-11. Example AI Wound Monitoring Report.

ATTACHMENT B

EXAMPLE INTERNAL DOSIMETRY RECORD DOCUMENTS

Page 12 of 12
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 60 of 82

ATTACHMENT C

ADJUSTMENT TO HELGESON NUCLEAR SERVICES 235U LUNG COUNTS, 1981 to 1983

Measured

(μg U-235)

Actual

(μg U-235)

32.5 0

40 8.6

50 20.2

60 31.7

70 43.2

80 54.8

90 66.3

100 77.8

110 89.4

120 101

130 112

140 124

150 135

160 147

170 159

180 170

190 182

200 193

210 205

220 216

230 228

240 239

245 245
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 61 of 82

ATTACHMENT D

ISOTOPIC COMPOSITION OF ATR-ETR FUEL USED IN THE POWDER ROOM, 1966 to 1967

Date received

Isotopic composition (weight percent)

U-234 U-235 U-236 U-238

06/27/1966 0.579 93.2 0.304 5.917

06/27/1966 0.571 93.2 0.301 5.928

10/12/1966 0.516 93.15 0.310 6.010

10/13/1966 0.656 93.16 0.295 5.900

11/11/1966 0.563 93.14 0.289 6.000

12/08/1966 0.591 93.11 0.326 5.970

12/08/1966 0.627 93.13 0.286 5.960

01/20/1967 0.563 93.09 0.305 6.020

02/06/1967 0.570 93.15 0.320 5.956

02/08/1967 0.580 93.17 0.290 5.961

02/09/1967 0.550 93.14 0.310 6.001

02/10/1967 0.570 93.19 0.270 5.966

02/11/1963 0.630 93.14 0.300 5.929

03/11/1967 0.660 93.17 0.300 5.872

03/16/1967 0.610 93.13 0.350 5.912

04/16/1967 0.500 93.13 0.340 6.030

06/19/1967 0.557 93.23 0.324 5.860

06/22/1967 0.582 93.16 0.304 5.950

06/23/1967 0.600 93.14 0.290 5.972

06/23/1967 0.580 93.12 0.294 6.010

06/23/1967 0.570 93.14 0.301 6.000

Average 0.582 93.152 0.305 5.958

Standard deviation 0.039 0.033 0.019 0.048

Specific activity (μCi/g) 6,263.14 2.16 64.70 0.340

Fractional specific activity 

(μCi/g U) 36.46 2.01 0.20 0.020

Fractional specific activity 

(Bq/g U) 1.35E+06 7.44E+04 7.31E+03 7.50E+02

Data from Tschaeche (1968c)
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 62 of 82

ATTACHMENT E

ISOTOPIC COMPOSITION OF URANIUM USED FOR ATR FUEL FABRICATION, 1979 to 1982

Date

Isotopic composition (weight percent)

U-234 U-235 U-236 U-238

10/18/1979 1.006 93.145 0.525 5.234

01/29/1980 1.145 93.142 0.498 5.348

05/12/1980 1.007 93.141 0.472 5.379

12/03/1980 1.007 93.142 0.476 5.374

07/28/1981 1.004 93.14 0.455 5.648

11/02/1981 1.002 93.136 0.455 5.407

03/02/1982 0.9977 93.141 0.448 5.414

Average 1.024 93.141 0.476 5.401

Standard deviation 0.0534 0.0027 0.0276 0.1246

Specific activity 

(μCi/g) 6,263.14 2.16 64.70 0.340

Fractional specific 

activity (μCi/g-U) 64.141 2.012 0.308 0.018

Fractional specific 

activity (Bq/g-U) 2.37E+06 7.44E+04 1.14E+04 6.79E+02

Data from Moore (1979, 1980a,b,c, 1981a,b, 1982).
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 63 of 82

ATTACHMENT F

DETECTION LIMITS FOR GAMMA SCANS OF BIOASSAY SAMPLES, 1967 to 1975

Isotope(s)

Detection limit

(dpm/sample)

Na-24 80

Co-60 75

Sc-46 60

Cs-134 175

Ba-La-140 250

Co-58 50

K-40 500

Ru-l06 500

Zn-65 100

Cu-64 300

Mn-56 70

Mn-54 60

Ga-72 140

Zr-Nb-95 20

Cs-137 50

As-76 120

Ru-103 50

Zn-69m 50

I-131 60

Cr-51 400

Np-239 80

Ce-141 60

Ce-Pr-144 500
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 64 of 82

ATTACHMENT G

SENSITIVITY OF BIOASSAY ANALYSES BY BIOASSAY CODE,

UST AND OTHER LABORATORIES

Table G-1. In vitro analyses (Barnes 2006).

Code in 

bioassay 

cards Analysis type

Sensitivity 

per daya

(1,500 mL)

Units for 

sensitivity

Typical 

analysis 

volume (mL)

Default 

isotope Vendorb

AP Gross alpha 15 dpm 100 Pu-239

FP1 Fission products (1) 30 dpm 200 Sr-90 UST

FP2 Fission products (2) 60 dpm 300 Cs-137 UST

FP3A Fission products (3a) 30 dpm 300 Sr-90 UST

FP3B Fission products (3b) 60 dpm 300 Cs-137 UST

GA1A Gross alpha (1a) 1.5 dpm 100 Pu-239 UST

GA1B Gross alpha (1b) 9 dpm 100 Pu-239 UST

GA2 Gross alpha (2) 15 dpm 100 Pu-239 UST

GBH Gross beta high level 750 dpm 50 Sr-90 UST

H3 Tritium 2.25E+06 dpm 10 H-3 UST

I131 Iodine-131 300 dpm 250 I-131 UST

MFP2B Fission products (2) 60 dpm 300 Cs-137 UST

MFPB Fission products (beta) 30 dpm 300 Sr-90

MFPG Fission products (gamma) 30 dpm 300 Cs-137

PUA Plutonium (a) 0.0495 dpm 1,000 Pu-239 UST

PUB Plutonium (b) 0.0495 dpm 1,000 Pu-239 UST

PUB(O) Plutonium (b) (optional) 0.75 dpm 1,000 Pu-239 UST

SR90 Strontium 90 30 dpm 200 Sr-90 UST

TH Thorium 0.99 μg 1,000 Th-232 UST

UF Uranium fluorometric 0.3 μg 10 U-238 UST

U-ISO Uranium isotopic 3.75 dpm 200 U-235

UR Uranium radiometric 3.75 dpm 200 U-235 UST

U-TOT Uranium total 0.3 μg 10 U-238

a. Considered to be the “default MDAs” by AI; values listed on the cards (for example, “<2.2 dpm/day”) were used as the 

MDA, if available. However, these values were apparently used as screening or reporting levels and the MDA is 

assumed to be twice these values for dose reconstruction purposes.

b. Blank indicates that the vendor was not recorded on the bioassay card.

Table G-2. In vivo analyses (Barnes 2006).

Code in 

bioassay 

cards Analysis type Sensitivity 

Units for 

sensitivitya Default isotope Vendor

IVLC In vivo lung count 40 μg U-235 UST

TBC Total-body count 2.8 nCi Cs-137 UST

WBC Whole-body count 2.8 nCi Cs-137 Helgeson

a. The MDA is assumed to be twice these values for dose reconstruction purposes.
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 04/26/2010 Page 65 of 82

ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 1 of 18

TABLE OF CONTENTS

SECTION TITLE

1.0 Purpose ....................................................................................................................................

PAGE

66

2.0 The Model.................................................................................................................................66

3.0 Limiting Solubility Classes......................................................................................................... 67

4.0 Conclusions and Recommendations......................................................................................... 68

LIST OF TABLES

TABLE TITLE

H-1 Parameters for Type S and Type K uranium ..........................................................................

PAGE

66

H-2 Limiting uranium solubility class for chronic intakes................................................................ 69

H-3 Limiting uranium solubility class for an acute intake ............................................................... 70

H-4 Acute intake urine bioassay at 2 days after intake.................................................................. 71

H-5 Acute intake urine bioassay at 5 days after intake.................................................................. 72

H-6 Acute intake urine bioassay at 10 days after intake................................................................ 73

H-7 Acute intake urine bioassay at 40 days after intake................................................................ 74

H-8 Acute intake urine bioassay at 100 days after intake.............................................................. 75

H-9 Acute intake urine bioassay at 400 days after intake.............................................................. 76

H-10 Acute intake urine bioassay at 1000 days after intake............................................................ 77

H-11 Acute intake urine bioassay at 10000 days after intake.......................................................... 78

H-12 Chronic intake (10 days) with urine bioassay on day after end of intake................................. 79

H-13 Chronic intake (100 days) with urine bioassay on day after end of intake............................... 80

H-14 Chronic intake (1000 days) with urine bioassay on day after end of intake............................. 81

H-15 Chronic intake (10000 days) with urine bioassay on day after end of intake........................... 82

LIST OF FIGURES

FIGURE TITLE

H-1 Comparison of urinary excretion patterns following inhalation intakes of 5 μm

AMAD Type M and Type S 234U and 1 μm AMAD Type K 234U...............................................

PAGE

67
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1.0

Leggett, Eckerman, and Boice (2005) recently proposed a modification to the ICRP International 

Commission on Radiological Protection Publication 66 human respiratory tract model (HRTM) that 

specifically addresses the biokinetics of uranium aluminide (UAl) (ICRP 1994). This modification, 

which this Attachment refers to as Type K (versus Type F, M, or S) material, was based on 

occupational bioassay data collected at the Rocketdyne/Atomics International Facility in California. 

This Attachment evaluates the impact of this model on intake and dose calculations.

PURPOSE

2.0

The Type K model is based on the standard Type S aerosol with the modifications to the mechanical 

clearance parameters and dissolution parameters of the HRTM listed on Table H-1. 

THE MODEL

Table H-1. Parameters for Type 

S and Type K uranium.

Parameter S K

sp 0.1 0.0001

spt 100 0.004

st 0.0001 0.004

TAI1 -> Tbb1 0.02 0.006

TAI2 -> Tbb1 0.001 0.006

TAI3 -> Tbb1 0.0001 0.006

It is important to note that the changes in the mechanical clearance parameters are for the 

transformed compartments only, which means that the Type K model cannot be directly implemented 

in the Integrated Modules for Bioassay Analysis (IMBA) computer program. Finally, based on the 

Leggett, Eckerman, and Boice (2005) paper an aerosol size of 1 μm activity median aerodynamic 

diameter (AMAD) is recommended for exposure to UAl at the Rocketdyne/Atomics International 

Facility.

Type K, M, and S models were implemented in the Mathcad® 13 computer program (Mathsoft® 2006) 

to calculate urinary excretion and dose. A comparison of the urinary excretion following unit intakes of 

these materials is shown in Figure H-1, which is a benchmark of Figure 7 in the Leggett, Eckerman, 

and Boice (2005) paper. In Figure H-1, the dots are the values of the Type K excretion curve digitized 

from Figure 7 in Leggett, Eckerman, and Boice and the lines are the excretion rates calculated here. 

This plot provides some evidence that the Type K model was accurately implemented.
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1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04

Time After Intake (days)

Fraction of Intake in 24-hour Urine

Type M

Type S

Type K

Figure H-1. Comparison of urinary excretion patterns following inhalation intakes of 5 

μm AMAD Type M and Type S 234U and 1 μm

3.0

AMAD Type K 234U.

The general approach taken here is to examine the dose to organs from various exposure scenarios 

to determine if and when Type K 234U delivers a dose that is higher than the dose from Type F, M, or S 

234U. If Type K delivers the highest dose, it is “limiting” and the case cannot be evaluated with IMBA. 

Otherwise, the case can be evaluated with a standard type using IMBA. In other words, the standard 

models in IMBA can be used as long as the product of the DCF and 1/IRF exceeds that of Type K.

LIMITING SOLUBILITY CLASSES

Each table in Attachment H consists of three parts:

• The first is a matrix of intake-to-organ-dose conversion factors (DCFs) for 234U. The DCFs give 

the 50-year committed organ dose in units of Sv per total intake. For example, in the case of a 

10-d chronic intake of 1 Bq/d the total intake is (10 d)(1 Bq/d) = 10 Bq. The DCFs are given 

for 1 μm AMAD Type K uranium (referred to as K1) and 5 μm AMAD Type K (referred to as 

K5), F, M, and S uranium. 

• The second matrix contains the urine intake retention fractions (irf) for the given time after an 

acute intake or the start of a chronic intake. For an acute intake the irf(t) is the fraction of a 

unit intake in the 24-hr urine sample collected on day t after intake. For a chronic intake the 

irf(t) is the Bq of 234U in the 24-hr urine sample collected on the day after the end of the chronic 

intake. For example, in the case of a 10-d chronic intake, the urine sample is collected from t 

= 10 days to t = 11 days.

• Finally, in the third matrix the organ dose H in Sv implied by 1 Bq/d of uranium excreted on 

day t is given:
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CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

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1

( )

Bq Sv H DCF

irf t Bq

   =      

For a given intake scenario the organ doses are calculated for all organs and solubility classes of 

uranium. The solubility class with the highest dose is referred to as the “limiting” solubility class. The 

results of Attachment H are summarized in the following tables.

4.0

As can be seen in Table H-2, except for relatively short chronic intakes, the standard solubility types 

F, M, and S are limiting.

CONCLUSIONS AND RECOMMENDATIONS

1

 1 For the stomach wall, 5 μm Type K uranium was limiting for a 10,000-day chronic intake, but it was only 1% greater than 

Type M.

 This means that for typical dose assessments that involve chronic intakes 

one can use standard solubility types and not risk underestimating the dose to organs. The situation 

is more complex for acute intakes, as can be seen in Table H-3. Here Type K uranium can be limiting 

for most organs if urine samples are taken days after an intake and for nonsystemic organs if samples

are taken years after an intake. In these situations custom dose assessments that use software other 

than IMBA might be required, so these cases should be referred to the Principal Internal Dosimetrist 

for evaluation. In summary, standard models may be used for calculating dose to any organ resulting 

from a chronic UAl intake of 365 days or more as long as the urine data were collected more than 365 

days after the start of the intake. The PID should be consulted for guidance in other cases that do not 

fit this criterion or use other types of bioassay data to estimate the dose.
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Table H-2. Limiting uranium solubility class for chronic 

intakes. The time (e.g., 10 d) is the duration of the chronic 

intake. The 24-hr urine sample was collected on the last day 

of the chronic intake.

Chronic Chronic Chronic Chronic

10 days 100 days 1000 days 10000 days

Adrenals K1 K1 S F

UB_Wall K1 K1 S F

Bone_Sur K1 K1 S F

Brain K1 K1 S F

Breasts K1 K1 S F

St_Wall K1 K1 S K5

SI_Wall K1 K1 S S

ULI_Wall K1 S S S

LLI_Wall K1 S S S

Kidneys K1 K1 S S

Liver K1 K1 S F

ET1-bas K1 S S S

ET2-bas S S S S

LN-ET S S S S

BBi-bas K1 S S S

BBi-sec K1 S S S

bbe-sec K1 S S S

AI S S S S

LN-Th S S S S

Muscle K1 K1 S F

Ovaries K1 K1 S F

Pancreas K1 K1 S F

R_Marrow K1 K1 S F

Skin K1 K1 S F

Spleen K1 K1 S F

Testes K1 K1 S F

Thymus K1 K1 S F

Thyroid K1 K1 S F

GB_Wall K1 K1 S F

Ht_Wall K1 K1 S F

Uterus K1 K1 S F

Lung K1 S S S

ET S S S S

Colon K1 S S S

Oesophagus K1 K1 S F

Remainder K1 K1 S F
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Table H-3. Limiting uranium solubility class for an acute intake. The time (e.g., 2 d) is when

the urine sample was collected. 

Acute Acute Acute Acute Acute Acute Acute Acute

2 days 5 days 10 days 40 days 100 days 400 days 1000 days 10000 days

Adrenals K1 K1 K1 S F F F F

UB_Wall K1 K1 K1 S F F F F

Bone_Sur K1 K1 K1 S F F F F

Brain K1 K1 K1 S F F F F

Breasts K1 K1 K1 S F F F F

St_Wall K1 K1 K1 S F F F F

SI_Wall K1 K1 K1 S F F F K5

ULI_Wall K1 K1 K1 S S F F K5

LLI_Wall K1 K1 K5 S S F F K5

Kidneys K1 K1 K1 S F F F F

Liver K1 K1 K1 S F F F F

ET1-bas K1 K1 K1 S F F F K5

ET2-bas S S S S S S S K5

LN-ET S S S S S S S S

BBi-bas K5 S S S S S K5 K5

BBi-sec K1 K5 S S S S K5 K5

bbe-sec K1 K1 K1 S S S K5 K5

AI S S S S S S S K1

LN-Th S S S S S S S S

Muscle K1 K1 K1 S F F F F

Ovaries K1 K1 K1 S F F F F

Pancreas K1 K1 K1 S F F F F

R_Marrow K1 K1 K1 S F F F F

Skin K1 K1 K1 S F F F F

Spleen K1 K1 K1 S F F F F

Testes K1 K1 K1 S F F F F

Thymus K1 K1 K1 S F F F F

Thyroid K1 K1 K1 S F F F F

GB_Wall K1 K1 K1 S F F F F

Ht_Wall K1 K1 K1 S F F F F

Uterus K1 K1 K1 S F F F F

Lung K1 S S S S S S K5

ET S S S S S S S K5

Colon K1 K1 K1 S S F F K5

Oesophagus K1 K1 K1 S F F F F

Remainder K1 K1 K1 S F F F F
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CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

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Table H-4. Acute intake, urine bioassay at 2 days after intake. 

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.305E-03 5.451E-04 2.015E-04 8.531E-05 6.165E-05 K1

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.309E-03 5.469E-04 2.021E-04 8.559E-05 6.185E-05 K1

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 3.747E-02 1.565E-02 6.186E-03 2.434E-03 1.755E-03 K1

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.165E-05 K1

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.305E-03 5.451E-04 2.015E-04 8.531E-05 6.164E-05 K1

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.316E-03 5.585E-04 2.134E-04 8.572E-05 6.168E-05 K1

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.334E-03 5.786E-04 2.313E-04 8.633E-05 6.173E-05 K1

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 1.480E-03 7.479E-04 3.826E-04 9.149E-05 6.217E-05 K1

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 1.820E-03 1.140E-03 7.329E-04 1.034E-04 6.318E-05 K1

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.373E-02 5.732E-03 2.349E-03 8.893E-04 6.405E-04 K1

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 5.138E-03 2.146E-03 8.512E-04 3.333E-04 2.402E-04 K1

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.381E-03 6.688E-04 3.326E-04 8.953E-05 6.239E-05 K1

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 5.093E-01 7.565E-01 1.709E+00 1.115E-02 6.181E-05 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 4.302E-02 6.264E-02 3.021E+00 5.613E-04 6.168E-05 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 9.866E-02 1.101E-01 1.046E-01 2.569E-03 6.172E-05 K5

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 1.583E+00 1.250E+00 1.115E+00 3.365E-02 6.638E-05 K1

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 1.702E+00 7.257E-01 6.814E-01 1.859E-02 6.441E-05 K1

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 9.180E-01 3.607E-01 1.478E+00 6.907E-03 6.175E-05 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 3.237E-01 2.190E-01 1.042E+01 2.304E-03 6.164E-05 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.305E-03 5.451E-04 2.016E-04 8.531E-05 6.165E-05 K1

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 3.867E-03 1.615E-03 6.399E-04 2.513E-04 1.812E-04 K1

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.305E-03 5.451E-04 2.015E-04 8.531E-05 6.164E-05 K1

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.305E-03 5.451E-04 2.016E-04 8.531E-05 6.164E-05 K1

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.153E+00 5.885E-01 9.326E-01 1.452E-02 6.340E-05 K1

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 5.083E-01 7.551E-01 1.709E+00 1.113E-02 6.181E-05 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 1.627E-03 9.166E-04 5.332E-04 9.662E-05 6.260E-05 K1

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.305E-03 5.451E-04 2.014E-04 8.531E-05 6.164E-05 K1

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 1.431E-03 5.981E-04 2.238E-04 9.344E-05 6.748E-05 K1

IRF 3.050E-05 3.874E-05 4.410E-05 1.125E-03 6.380E-03
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

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Table H-5. Acute intake, urine bioassay at 5 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.827E-03 1.027E-03 4.046E-04 1.305E-04 9.337E-05 K1

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.833E-03 1.030E-03 4.059E-04 1.310E-04 9.368E-05 K1

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 5.246E-02 2.948E-02 1.242E-02 3.725E-03 2.658E-03 K1

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.337E-05 K1

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.827E-03 1.027E-03 4.046E-04 1.305E-04 9.336E-05 K1

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.843E-03 1.052E-03 4.286E-04 1.312E-04 9.342E-05 K1

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.867E-03 1.090E-03 4.645E-04 1.321E-04 9.350E-05 K1

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 2.073E-03 1.409E-03 7.683E-04 1.400E-04 9.416E-05 K1

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 2.549E-03 2.147E-03 1.472E-03 1.583E-04 9.569E-05 K1

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.922E-02 1.080E-02 4.716E-03 1.361E-03 9.701E-04 K1

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 7.194E-03 4.042E-03 1.709E-03 5.101E-04 3.638E-04 K1

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.934E-03 1.260E-03 6.679E-04 1.370E-04 9.449E-05 K1

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 7.130E-01 1.425E+00 3.432E+00 1.707E-02 9.361E-05 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 6.024E-02 1.180E-01 6.067E+00 8.589E-04 9.342E-05 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 1.381E-01 2.074E-01 2.101E-01 3.932E-03 9.348E-05 S

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 2.216E+00 2.355E+00 2.240E+00 5.149E-02 1.005E-04 K5

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 2.383E+00 1.367E+00 1.368E+00 2.845E-02 9.755E-05 K1

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 1.285E+00 6.793E-01 2.968E+00 1.057E-02 9.352E-05 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 4.532E-01 4.126E-01 2.093E+01 3.525E-03 9.336E-05 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.827E-03 1.027E-03 4.048E-04 1.305E-04 9.337E-05 K1

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.827E-03 1.027E-03 4.045E-04 1.305E-04 9.336E-05 K1

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.827E-03 1.027E-03 4.045E-04 1.305E-04 9.336E-05 K1

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 5.415E-03 3.043E-03 1.285E-03 3.846E-04 2.745E-04 K1

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.336E-05 K1

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.827E-03 1.027E-03 4.046E-04 1.305E-04 9.336E-05 K1

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.336E-05 K1

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.827E-03 1.027E-03 4.045E-04 1.305E-04 9.336E-05 K1

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.336E-05 K1

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.336E-05 K1

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.827E-03 1.027E-03 4.049E-04 1.305E-04 9.336E-05 K1

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.827E-03 1.027E-03 4.044E-04 1.305E-04 9.336E-05 K1

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.614E+00 1.108E+00 1.873E+00 2.222E-02 9.602E-05 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 7.117E-01 1.422E+00 3.431E+00 1.703E-02 9.361E-05 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 2.277E-03 1.726E-03 1.071E-03 1.479E-04 9.482E-05 K1

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.827E-03 1.027E-03 4.045E-04 1.305E-04 9.336E-05 K1

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 2.003E-03 1.127E-03 4.495E-04 1.430E-04 1.022E-04 K1

IRF 2.178E-05 2.056E-05 2.196E-05 7.349E-04 4.212E-03
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 9 of 18

Table H-6. Acute intake, urine bioassay at 10 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.557E-03 1.098E-03 5.544E-04 1.764E-04 1.462E-04 K1

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.562E-03 1.102E-03 5.562E-04 1.770E-04 1.467E-04 K1

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 4.471E-02 3.153E-02 1.702E-02 5.032E-03 4.162E-03 K1

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.557E-03 1.098E-03 5.544E-04 1.764E-04 1.462E-04 K1

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.571E-03 1.125E-03 5.872E-04 1.772E-04 1.463E-04 K1

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.592E-03 1.166E-03 6.364E-04 1.785E-04 1.464E-04 K1

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 1.767E-03 1.507E-03 1.053E-03 1.892E-04 1.475E-04 K1

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 2.172E-03 2.297E-03 2.017E-03 2.139E-04 1.499E-04 K5

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.638E-02 1.155E-02 6.462E-03 1.839E-03 1.519E-03 K1

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 6.131E-03 4.323E-03 2.342E-03 6.892E-04 5.697E-04 K1

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.649E-03 1.347E-03 9.151E-04 1.851E-04 1.480E-04 K1

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 6.077E-01 1.524E+00 4.703E+00 2.306E-02 1.466E-04 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 5.134E-02 1.262E-01 8.314E+00 1.160E-03 1.463E-04 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 1.177E-01 2.219E-01 2.879E-01 5.312E-03 1.464E-04 S

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 1.888E+00 2.519E+00 3.069E+00 6.957E-02 1.574E-04 S

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 2.031E+00 1.462E+00 1.875E+00 3.843E-02 1.528E-04 K1

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 1.095E+00 7.266E-01 4.067E+00 1.428E-02 1.465E-04 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 3.863E-01 4.413E-01 2.868E+01 4.763E-03 1.462E-04 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.557E-03 1.098E-03 5.547E-04 1.764E-04 1.462E-04 K1

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 4.615E-03 3.254E-03 1.761E-03 5.197E-04 4.299E-04 K1

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.557E-03 1.098E-03 5.543E-04 1.764E-04 1.462E-04 K1

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.557E-03 1.098E-03 5.543E-04 1.764E-04 1.462E-04 K1

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.557E-03 1.098E-03 5.548E-04 1.764E-04 1.462E-04 K1

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.557E-03 1.098E-03 5.542E-04 1.764E-04 1.462E-04 K1

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.375E+00 1.186E+00 2.566E+00 3.003E-02 1.504E-04 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 6.065E-01 1.521E+00 4.702E+00 2.302E-02 1.466E-04 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 1.941E-03 1.847E-03 1.467E-03 1.998E-04 1.485E-04 K1

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.557E-03 1.098E-03 5.543E-04 1.764E-04 1.462E-04 K1

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 1.707E-03 1.205E-03 6.159E-04 1.932E-04 1.601E-04 K1

IRF 2.556E-05 1.923E-05 1.603E-05 5.439E-04 2.690E-03
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 10 of 18

Table H-7. Acute intake, urine bioassay at 40 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.192E-04 S

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 8.208E-04 8.162E-04 1.384E-03 4.408E-04 9.222E-04 S

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 2.349E-02 2.335E-02 4.234E-02 1.254E-02 2.616E-02 S

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.192E-04 S

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 8.254E-04 8.335E-04 1.461E-03 4.415E-04 9.197E-04 S

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 8.362E-04 8.634E-04 1.583E-03 4.446E-04 9.204E-04 S

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 9.283E-04 1.116E-03 2.619E-03 4.712E-04 9.269E-04 S

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 1.141E-03 1.701E-03 5.016E-03 5.327E-04 9.420E-04 S

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 8.606E-03 8.554E-03 1.608E-02 4.580E-03 9.550E-03 S

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 3.221E-03 3.202E-03 5.826E-03 1.717E-03 3.581E-03 S

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 8.662E-04 9.980E-04 2.276E-03 4.611E-04 9.302E-04 S

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 3.193E-01 1.129E+00 1.170E+01 5.745E-02 9.216E-04 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 2.698E-02 9.347E-02 2.068E+01 2.891E-03 9.196E-04 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 6.186E-02 1.643E-01 7.162E-01 1.323E-02 9.202E-04 S

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 9.922E-01 1.866E+00 7.634E+00 1.733E-01 9.897E-04 S

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 1.067E+00 1.083E+00 4.664E+00 9.574E-02 9.604E-04 S

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 5.756E-01 5.382E-01 1.012E+01 3.557E-02 9.207E-04 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 2.030E-01 3.269E-01 7.133E+01 1.187E-02 9.191E-04 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 8.181E-04 8.135E-04 1.380E-03 4.394E-04 9.192E-04 S

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 2.425E-03 2.411E-03 4.380E-03 1.295E-03 2.702E-03 S

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 8.180E-04 8.135E-04 1.378E-03 4.394E-04 9.191E-04 S

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 8.181E-04 8.135E-04 1.380E-03 4.394E-04 9.191E-04 S

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 8.180E-04 8.135E-04 1.378E-03 4.394E-04 9.191E-04 S

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 7.227E-01 8.782E-01 6.384E+00 7.480E-02 9.453E-04 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 3.187E-01 1.127E+00 1.170E+01 5.734E-02 9.216E-04 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 1.020E-03 1.368E-03 3.650E-03 4.976E-04 9.334E-04 S

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 8.181E-04 8.135E-04 1.379E-03 4.394E-04 9.191E-04 S

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 8.970E-04 8.926E-04 1.532E-03 4.812E-04 1.006E-03 S

IRF 4.864E-05 2.596E-05 6.442E-06 2.183E-04 4.279E-04
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 11 of 18

Table H-8. Acute intake, urine bioassay at 100 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 5.907E-04 6.156E-04 2.172E-03 8.373E-04 3.927E-03 F

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 5.926E-04 6.176E-04 2.179E-03 8.401E-04 3.939E-03 F

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 1.696E-02 1.767E-02 6.669E-02 2.389E-02 1.118E-01 F

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 5.907E-04 6.155E-04 2.171E-03 8.373E-04 3.927E-03 F

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 5.907E-04 6.156E-04 2.172E-03 8.373E-04 3.926E-03 F

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 5.959E-04 6.307E-04 2.301E-03 8.414E-04 3.929E-03 F

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 6.038E-04 6.533E-04 2.494E-03 8.473E-04 3.932E-03 F

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 6.702E-04 8.446E-04 4.125E-03 8.980E-04 3.960E-03 S

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 8.241E-04 1.287E-03 7.901E-03 1.015E-03 4.024E-03 S

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 6.214E-03 6.473E-03 2.532E-02 8.728E-03 4.080E-02 F

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 2.326E-03 2.423E-03 9.177E-03 3.272E-03 1.530E-02 F

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 6.254E-04 7.552E-04 3.586E-03 8.788E-04 3.974E-03 F

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 2.306E-01 8.543E-01 1.843E+01 1.095E-01 3.937E-03 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 1.948E-02 7.073E-02 3.258E+01 5.509E-03 3.929E-03 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 4.466E-02 1.244E-01 1.128E+00 2.522E-02 3.931E-03 S

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 7.164E-01 1.412E+00 1.202E+01 3.302E-01 4.228E-03 S

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 7.705E-01 8.195E-01 7.347E+00 1.825E-01 4.103E-03 S

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 4.156E-01 4.073E-01 1.594E+01 6.779E-02 3.933E-03 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 1.465E-01 2.474E-01 1.124E+02 2.261E-02 3.926E-03 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 5.907E-04 6.156E-04 2.173E-03 8.373E-04 3.927E-03 F

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 5.907E-04 6.156E-04 2.172E-03 8.373E-04 3.926E-03 F

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 5.907E-04 6.155E-04 2.172E-03 8.373E-04 3.926E-03 F

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 1.751E-03 1.824E-03 6.899E-03 2.467E-03 1.154E-02 F

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 5.907E-04 6.155E-04 2.171E-03 8.373E-04 3.926E-03 F

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 5.907E-04 6.156E-04 2.172E-03 8.373E-04 3.926E-03 F

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 5.906E-04 6.155E-04 2.171E-03 8.373E-04 3.926E-03 F

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 5.907E-04 6.155E-04 2.172E-03 8.373E-04 3.926E-03 F

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 5.907E-04 6.155E-04 2.171E-03 8.373E-04 3.926E-03 F

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 5.907E-04 6.155E-04 2.171E-03 8.373E-04 3.926E-03 F

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 5.907E-04 6.156E-04 2.174E-03 8.373E-04 3.926E-03 F

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 5.906E-04 6.155E-04 2.171E-03 8.373E-04 3.926E-03 F

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 5.218E-01 6.645E-01 1.006E+01 1.425E-01 4.038E-03 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 2.301E-01 8.527E-01 1.842E+01 1.093E-01 3.937E-03 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 7.364E-04 1.035E-03 5.749E-03 9.483E-04 3.988E-03 S

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 5.907E-04 6.155E-04 2.172E-03 8.373E-04 3.926E-03 F

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 6.477E-04 6.754E-04 2.413E-03 9.171E-04 4.298E-03 F

IRF 6.737E-05 3.430E-05 4.090E-06 1.146E-04 1.002E-04
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 12 of 18

Table H-9. Acute intake, urine bioassay at 400 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.296E-03 1.350E-03 3.482E-03 5.358E-03 8.580E-02 F

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.300E-03 1.355E-03 3.493E-03 5.376E-03 8.608E-02 F

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 3.721E-02 3.876E-02 1.069E-01 1.529E-01 2.442E+00 F

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.296E-03 1.350E-03 3.482E-03 5.358E-03 8.580E-02 F

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.307E-03 1.383E-03 3.688E-03 5.384E-03 8.585E-02 F

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.325E-03 1.433E-03 3.997E-03 5.422E-03 8.592E-02 F

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 1.470E-03 1.852E-03 6.612E-03 5.747E-03 8.652E-02 F

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 1.808E-03 2.824E-03 1.266E-02 6.497E-03 8.793E-02 F

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.363E-02 1.420E-02 4.059E-02 5.586E-02 8.914E-01 F

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 5.103E-03 5.315E-03 1.471E-02 2.094E-02 3.343E-01 F

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.372E-03 1.656E-03 5.747E-03 5.624E-03 8.683E-02 F

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 5.058E-01 1.874E+00 2.953E+01 7.006E-01 8.602E-02 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 4.273E-02 1.551E-01 5.221E+01 3.525E-02 8.584E-02 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 9.798E-02 2.727E-01 1.808E+00 1.614E-01 8.590E-02 S

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 1.572E+00 3.097E+00 1.927E+01 2.113E+00 9.238E-02 S

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 1.690E+00 1.797E+00 1.178E+01 1.168E+00 8.964E-02 S

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 9.117E-01 8.932E-01 2.554E+01 4.338E-01 8.594E-02 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 3.215E-01 5.425E-01 1.801E+02 1.447E-01 8.580E-02 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.296E-03 1.350E-03 3.483E-03 5.358E-03 8.580E-02 F

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.296E-03 1.350E-03 3.481E-03 5.358E-03 8.580E-02 F

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.296E-03 1.350E-03 3.481E-03 5.358E-03 8.580E-02 F

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 3.841E-03 4.001E-03 1.106E-02 1.579E-02 2.522E-01 F

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.296E-03 1.350E-03 3.482E-03 5.358E-03 8.580E-02 F

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.296E-03 1.350E-03 3.481E-03 5.358E-03 8.580E-02 F

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.296E-03 1.350E-03 3.484E-03 5.358E-03 8.580E-02 F

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.296E-03 1.350E-03 3.480E-03 5.358E-03 8.580E-02 F

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.145E+00 1.457E+00 1.612E+01 9.121E-01 8.824E-02 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 5.048E-01 1.870E+00 2.953E+01 6.992E-01 8.602E-02 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 1.615E-03 2.270E-03 9.215E-03 6.069E-03 8.713E-02 F

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.296E-03 1.350E-03 3.481E-03 5.358E-03 8.580E-02 F

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 1.421E-03 1.481E-03 3.868E-03 5.869E-03 9.392E-02 F

IRF 3.071E-05 1.564E-05 2.552E-06 1.790E-05 4.584E-06
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 13 of 18

Table H-10. Acute intake, urine bioassay at 1000 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.832E-02 1.820E-02 5.829E-03 8.597E-02 1.706E-01 F

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.838E-02 1.826E-02 5.849E-03 8.625E-02 1.711E-01 F

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 5.260E-01 5.226E-01 1.790E-01 2.453E+00 4.855E+00 F

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.832E-02 1.820E-02 5.829E-03 8.597E-02 1.706E-01 F

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.848E-02 1.865E-02 6.175E-03 8.638E-02 1.707E-01 F

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.872E-02 1.932E-02 6.692E-03 8.699E-02 1.708E-01 F

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 2.078E-02 2.498E-02 1.107E-02 9.219E-02 1.720E-01 F

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 2.555E-02 3.807E-02 2.120E-02 1.042E-01 1.748E-01 F

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.927E-01 1.914E-01 6.795E-02 8.961E-01 1.772E+00 F

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 7.213E-02 7.166E-02 2.463E-02 3.359E-01 6.646E-01 F

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.939E-02 2.233E-02 9.622E-03 9.022E-02 1.726E-01 F

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 7.149E+00 2.526E+01 4.945E+01 1.124E+01 1.710E-01 S

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 6.040E-01 2.092E+00 8.742E+01 5.656E-01 1.707E-01 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 1.385E+00 3.678E+00 3.027E+00 2.589E+00 1.708E-01 K5

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 2.222E+01 4.176E+01 3.227E+01 3.391E+01 1.837E-01 K5

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 2.389E+01 2.423E+01 1.972E+01 1.873E+01 1.782E-01 K5

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 1.289E+01 1.204E+01 4.276E+01 6.960E+00 1.709E-01 S

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 4.544E+00 7.315E+00 3.015E+02 2.322E+00 1.706E-01 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.832E-02 1.821E-02 5.832E-03 8.597E-02 1.706E-01 F

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 5.429E-02 5.394E-02 1.851E-02 2.533E-01 5.015E-01 F

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.832E-02 1.820E-02 5.829E-03 8.597E-02 1.706E-01 F

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.831E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.832E-02 1.820E-02 5.828E-03 8.597E-02 1.706E-01 F

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.832E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.832E-02 1.821E-02 5.834E-03 8.597E-02 1.706E-01 F

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.831E-02 1.820E-02 5.827E-03 8.597E-02 1.706E-01 F

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.618E+01 1.965E+01 2.698E+01 1.463E+01 1.754E-01 S

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 7.135E+00 2.522E+01 4.943E+01 1.122E+01 1.710E-01 S

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 2.283E-02 3.061E-02 1.543E-02 9.737E-02 1.732E-01 F

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.832E-02 1.820E-02 5.828E-03 8.597E-02 1.706E-01 F

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 2.008E-02 1.997E-02 6.476E-03 9.416E-02 1.867E-01 F

IRF 2.173E-06 1.160E-06 1.524E-06 1.116E-06 2.306E-06
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 14 of 18

Table H-11. Acute intake, urine bioassay at 10000 days after intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv/Bq) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Adrenals 1.612E+00 1.615E+00 8.765E-02 1.671E+00 1.693E+00 F

UB_Wall 3.993E-08 2.119E-08 8.914E-09 9.625E-08 3.946E-07 UB_Wall 1.618E+00 1.620E+00 8.795E-02 1.677E+00 1.699E+00 F

Bone_Sur 1.143E-06 6.062E-07 2.728E-07 2.737E-06 1.119E-05 Bone_Sur 4.630E+01 4.635E+01 2.691E+00 4.768E+01 4.819E+01 F

Brain 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Brain 1.612E+00 1.614E+00 8.762E-02 1.671E+00 1.693E+00 F

Breasts 3.979E-08 2.112E-08 8.884E-09 9.593E-08 3.933E-07 Breasts 1.612E+00 1.615E+00 8.765E-02 1.671E+00 1.693E+00 F

St_Wall 4.015E-08 2.163E-08 9.411E-09 9.640E-08 3.935E-07 St_Wall 1.627E+00 1.654E+00 9.285E-02 1.679E+00 1.694E+00 F

SI_Wall 4.068E-08 2.241E-08 1.020E-08 9.708E-08 3.938E-07 SI_Wall 1.648E+00 1.714E+00 1.006E-01 1.691E+00 1.695E+00 K5

ULI_Wall 4.515E-08 2.897E-08 1.687E-08 1.029E-07 3.966E-07 ULI_Wall 1.829E+00 2.215E+00 1.665E-01 1.792E+00 1.707E+00 K5

LLI_Wall 5.552E-08 4.416E-08 3.232E-08 1.163E-07 4.031E-07 LLI_Wall 2.249E+00 3.377E+00 3.189E-01 2.026E+00 1.735E+00 K5

Kidneys 4.186E-07 2.220E-07 1.036E-07 1.000E-06 4.086E-06 Kidneys 1.696E+01 1.698E+01 1.022E+00 1.742E+01 1.759E+01 F

Liver 1.567E-07 8.312E-08 3.753E-08 3.748E-07 1.532E-06 Liver 6.349E+00 6.356E+00 3.703E-01 6.530E+00 6.596E+00 F

ET1-bas 4.213E-08 2.591E-08 1.467E-08 1.007E-07 3.980E-07 ET1-bas 1.707E+00 1.981E+00 1.447E-01 1.754E+00 1.713E+00 K5

ET2-bas 1.553E-05 2.930E-05 7.536E-05 1.254E-05 3.943E-07 ET2-bas 6.293E+02 2.241E+03 7.436E+02 2.185E+02 1.697E+00 K5

LN-ET 1.312E-06 2.426E-06 1.332E-04 6.311E-07 3.935E-07 LN-ET 5.316E+01 1.855E+02 1.315E+03 1.100E+01 1.694E+00 S

BBi-bas 3.009E-06 4.266E-06 4.614E-06 2.889E-06 3.937E-07 BBi-bas 1.219E+02 3.262E+02 4.552E+01 5.033E+01 1.695E+00 K5

BBi-sec 4.827E-05 4.843E-05 4.918E-05 3.784E-05 4.235E-07 BBi-sec 1.955E+03 3.703E+03 4.852E+02 6.592E+02 1.823E+00 K5

bbe-sec 5.191E-05 2.811E-05 3.005E-05 2.090E-05 4.109E-07 bbe-sec 2.103E+03 2.149E+03 2.965E+02 3.642E+02 1.769E+00 K5

AI 2.800E-05 1.397E-05 6.517E-05 7.767E-06 3.939E-07 AI 1.134E+03 1.068E+03 6.430E+02 1.353E+02 1.696E+00 K1

LN-Th 9.872E-06 8.485E-06 4.595E-04 2.591E-06 3.933E-07 LN-Th 4.000E+02 6.488E+02 4.534E+03 4.513E+01 1.693E+00 S

Muscle 3.980E-08 2.112E-08 8.888E-09 9.593E-08 3.933E-07 Muscle 1.612E+00 1.615E+00 8.770E-02 1.671E+00 1.693E+00 F

Ovaries 3.979E-08 2.112E-08 8.881E-09 9.593E-08 3.933E-07 Ovaries 1.612E+00 1.615E+00 8.763E-02 1.671E+00 1.693E+00 F

Pancreas 3.979E-08 2.111E-08 8.882E-09 9.593E-08 3.933E-07 Pancreas 1.612E+00 1.614E+00 8.763E-02 1.671E+00 1.693E+00 F

R_Marrow 1.180E-07 6.257E-08 2.822E-08 2.826E-07 1.156E-06 R_Marrow 4.779E+00 4.784E+00 2.784E-01 4.924E+00 4.977E+00 F

Skin 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Skin 1.612E+00 1.614E+00 8.763E-02 1.671E+00 1.693E+00 F

Spleen 3.979E-08 2.112E-08 8.883E-09 9.593E-08 3.933E-07 Spleen 1.612E+00 1.615E+00 8.765E-02 1.671E+00 1.693E+00 F

Testes 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Testes 1.612E+00 1.614E+00 8.762E-02 1.671E+00 1.693E+00 F

Thymus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Thymus 1.612E+00 1.614E+00 8.764E-02 1.671E+00 1.693E+00 F

Thyroid 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 Thyroid 1.612E+00 1.614E+00 8.763E-02 1.671E+00 1.693E+00 F

GB_Wall 3.979E-08 2.111E-08 8.881E-09 9.593E-08 3.933E-07 GB_Wall 1.612E+00 1.614E+00 8.763E-02 1.671E+00 1.693E+00 F

Ht_Wall 3.980E-08 2.112E-08 8.891E-09 9.593E-08 3.933E-07 Ht_Wall 1.612E+00 1.615E+00 8.772E-02 1.671E+00 1.693E+00 F

Uterus 3.979E-08 2.111E-08 8.880E-09 9.593E-08 3.933E-07 Uterus 1.612E+00 1.614E+00 8.762E-02 1.671E+00 1.693E+00 F

Lung 3.516E-05 2.280E-05 4.113E-05 1.633E-05 4.045E-07 Lung 1.424E+03 1.743E+03 4.058E+02 2.845E+02 1.741E+00 K5

ET 1.550E-05 2.925E-05 7.534E-05 1.252E-05 3.943E-07 ET 6.281E+02 2.236E+03 7.434E+02 2.181E+02 1.697E+00 K5

Colon 4.961E-08 3.550E-08 2.351E-08 1.087E-07 3.994E-07 Colon 2.010E+00 2.715E+00 2.320E-01 1.893E+00 1.719E+00 K5

Oesophagus 3.979E-08 2.111E-08 8.883E-09 9.593E-08 3.933E-07 Oesophagus 1.612E+00 1.614E+00 8.764E-02 1.671E+00 1.693E+00 F

Remainder 4.363E-08 2.317E-08 9.870E-09 1.051E-07 4.305E-07 Remainder 1.768E+00 1.772E+00 9.738E-02 1.830E+00 1.853E+00 F

IRF 2.468E-08 1.308E-08 1.014E-07 5.740E-08 2.323E-07
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 15 of 18

Table H-12. Chronic intake (10 days) with urine bioassay on day after end of intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv) (Sv (Sv) (Sv) (Sv) Biggest

Adrenals 3.978E-07 2.111E-07 8.881E-08 9.590E-07 3.932E-06 Adrenals 8.146E-04 2.952E-04 9.838E-05 3.249E-05 1.796E-05 K1

UB_Wall 3.991E-07 2.118E-07 8.912E-08 9.622E-07 3.945E-06 UB_Wall 8.172E-04 2.962E-04 9.872E-05 3.260E-05 1.802E-05 K1

Bone_Sur 1.143E-05 6.062E-06 2.727E-06 2.737E-05 1.119E-04 Bone_Sur 2.340E-02 8.477E-03 3.021E-03 9.273E-04 5.111E-04 K1

Brain 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 Brain 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K1

Breasts 3.978E-07 2.111E-07 8.881E-08 9.590E-07 3.932E-06 Breasts 8.146E-04 2.952E-04 9.838E-05 3.249E-05 1.796E-05 K1

St_Wall 4.014E-07 2.163E-07 9.408E-08 9.637E-07 3.932E-06 St_Wall 8.219E-04 3.025E-04 1.042E-04 3.265E-05 1.796E-05 K1

SI_Wall 4.066E-07 2.240E-07 1.020E-07 9.705E-07 3.932E-06 SI_Wall 8.326E-04 3.132E-04 1.130E-04 3.288E-05 1.796E-05 K1

ULI_Wall 4.514E-07 2.897E-07 1.687E-07 1.028E-06 3.965E-06 ULI_Wall 9.243E-04 4.051E-04 1.869E-04 3.483E-05 1.811E-05 K1

LLI_Wall 5.551E-07 4.416E-07 3.231E-07 1.163E-06 4.029E-06 LLI_Wall 1.137E-03 6.175E-04 3.579E-04 3.940E-05 1.840E-05 K1

Kidneys 4.186E-06 2.221E-06 1.036E-06 9.999E-06 4.086E-05 Kidneys 8.572E-03 3.106E-03 1.148E-03 3.388E-04 1.866E-04 K1

Liver 1.567E-06 8.311E-07 3.753E-07 3.748E-06 1.532E-05 Liver 3.209E-03 1.162E-03 4.157E-04 1.270E-04 6.997E-05 K1

ET1-bas 3.982E-07 2.590E-07 1.466E-07 1.007E-06 3.979E-06 ET1-bas 8.154E-04 3.622E-04 1.624E-04 3.412E-05 1.817E-05 K1

ET2-bas 1.553E-04 2.930E-04 7.577E-04 1.254E-04 3.934E-06 ET2-bas 3.180E-01 4.097E-01 8.393E-01 4.249E-03 1.797E-05 S

LN-ET 1.312E-05 2.426E-05 1.332E-03 6.308E-06 3.934E-06 LN-ET 2.687E-02 3.393E-02 1.475E+00 2.137E-04 1.797E- S

BBi-bas 3.012E-05 4.268E-05 4.615E-05 2.890E-05 3.932E-06 BBi-bas 6.168E-02 5.968E- 5.112E-02 9.791E-04 1.796E- K

BBi-sec 4.826E-04 4.843E-04 4.915E- 3.781E-04 3.932E- BBi- 9.882E-01 6.773E-01 5.444E- 1.281E-02 1.796E-05 K1

bbe-sec 5.190E-04 2.811E-04 3.004E-04 2.089E-04 3.932E-06 bbe-sec 1.063E+00 3.931E-01 3.328E-01 7.078E-03 1.796E-05 K1

AI 2.800E-04 1.397E-04 6.540E-04 7.766E-05 3.932E-06 AI 5.733E-01 1.954E-01 7.244E-01 2.631E-03 1.796E-05 S

LN-Th 9.873E-05 8.485E-05 4.594E-03 2.591E-05 3.932E-06 LN-Th 2.022E-01 1.187E-01 5.089E+0 8.778E-04 1.796E-05 S

Muscle 3.978E-07 2.111E-07 8.886E-08 9.590E-07 3.932E-06 Muscle 8.146E-04 2.952E-04 9.843E-05 3.249E-05 1.796E-05 K1

Ovaries 3.978E-07 2.111E-07 8.879E-08 9.590E-07 3.932E-06 Ovaries 8.146E-04 2.952E-04 9.835E-05 3.249E-05 1.796E-05 K1

Pancreas 3.978E-07 2.111E-07 8.879E-08 9.590E-07 3.932E-06 Pancreas 8.146E-04 2.952E-04 9.835E-05 3.249E-05 1.796E-05 K1

R_Marrow 1.179E-06 6.257E-07 2.821E-07 2.826E-06 1.156E-05 R_Marrow 2.414E-03 8.750E-04 3.125E-04 9.574E-05 5.280E-05 K1

Skin 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 Skin 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K1

Spleen 3.978E-07 2.111E-07 8.881E-08 9.590E-07 3.932E-06 Spleen 8.146E-04 2.952E-04 9.838E-05 3.249E-05 1.796E-05 K1

Testes 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 Testes 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K1

Thymus 3.978E-07 2.111E-07 8.880E-08 9.590E-07 3.932E-06 Thymus 8.146E-04 2.952E-04 9.837E-05 3.249E-05 1.796E-05 K1

Thyroid 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 Thyroid 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K1

GB_Wall 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 GB_Wall 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K1

Ht_Wall 3.978E-07 2.111E-07 8.889E-08 9.590E-07 3.932E-06 Ht_Wall 8.146E-04 2.952E-04 9.846E-05 3.249E-05 1.796E-05 K

Uterus 3.978E-07 2.111E-07 8.878E-08 9.590E-07 3.932E-06 Uterus 8.146E-04 2.952E-04 9.834E-05 3.249E-05 1.796E-05 K

Lung 3.515E-04 2.280E-04 4.119E-04 1.632E-04 3.932E-06 Lung 7.198E-01 3.188E-01 4.563E-01 5.529E-03 1.796E-05 K

ET 1.550E-04 2.924E-04 7.576E-04 1.252E-04 3.934E-06 ET 3.174E-01 4.089E-01 8.392E-01 4.242E-03 1.797E-05 S

Colon 4.960E-07 3.550E-07 2.351E-07 1.086E-06 3.993E-06 Colon 1.016E-03 4.964E-04 2.604E-04 3.679E-05 1.824E-05 K

Oesophagus 3.978E-07 2.111E-07 8.880E-08 9.590E-07 3.932E-06 Oesophagus 8.146E-04 2.952E-04 9.837E-05 3.249E-05 1.796E-05 K

Remainder 4.362E-07 2.316E-07 9.867E-08 1.050E-06 4.304E-06 Remainder 8.932E-04 3.239E-04 1.093E-04 3.557E-05 1.966E-05 K

IRF 4.884E-04 7.151E-04 9.028E-04 2.952E-02 2.189E-01
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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 16 of 18

Table H-13. Chronic intake (100 days) with urine bioassay on day after end of intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv) (Sv) (Sv) (Sv) (Sv) Biggest

Adrenals 3.970E-06 2.106E-06 8.859E-07 9.572E-06 3.924E-05 Adrenals 7.559E-04 6.481E-04 5.966E-04 1.969E-04 1.484E-04 K1

UB_Wall 3.983E-06 2.113E-06 8.890E-07 9.603E-06 3.937E-05 UB_Wall 7.583E-04 6.503E-04 5.987E-04 1.976E-04 1.489E-04 K1

Bone_Sur 1.142E-04 6.058E-05 2.725E-05 2.735E-04 1.119E-03 Bone_Sur 2.174E-02 1.864E-02 1.835E-02 5.627E-03 4.233E-03 K1

Brain 3.970E-06 2.106E-06 8.856E-07 9.572E-06 3.924E-05 Brain 7.559E-04 6.481E-04 5.964E-04 1.969E-04 1.484E-04 K1

Breasts 3.970E-06 2.106E-06 8.859E-07 9.571E-06 3.924E-05 Breasts 7.559E-04 6.481E-04 5.966E-04 1.969E-04 1.484E-04 K1

St_Wall 4.005E-06 2.158E-06 9.386E-07 9.618E-06 3.926E-05 St_Wall 7.625E-04 6.641E-04 6.321E-04 1.979E-04 1.485E-04 K1

SI_Wall 4.058E-06 2.236E-06 1.017E-06 9.686E-06 3.929E-05 SI_Wall 7.726E-04 6.881E-04 6.849E-04 1.993E-04 1.486E-04 K1

ULI_Wall 4.506E-06 2.894E-06 1.687E-06 1.027E-05 3.957E-05 ULI_Wall 8.579E-04 8.906E-04 1.136E-03 2.113E-04 1.497E-04 S

LLI_Wall 5.542E-06 4.411E-06 3.229E-06 1.161E-05 4.022E-05 LLI_Wall 1.055E-03 1.357E-03 2.175E-03 2.389E-04 1.521E-04 S

Kidneys 4.186E-05 2.220E-05 1.035E-05 1.000E-04 4.086E-04 Kidneys 7.970E-03 6.832E-03 6.971E-03 2.058E-03 1.546E-03 K1

Liver 1.566E-05 8.309E-06 3.750E-06 3.747E-05 1.532E-04 Liver 2.982E-03 2.557E-03 2.526E-03 7.710E-04 5.795E-04 K1

ET1-bas 4.204E-06 2.585E-06 1.464E-06 1.005E-05 3.971E-05 ET1-bas 8.004E-04 7.955E-04 9.860E-04 2.068E-04 1.502E-04 S

ET2-bas 1.553E-03 2.930E-03 7.536E-03 1.254E-03 3.930E-05 ET2-bas 2.957E-01 9.017E-01 5.075E+00 2.580E-02 1.487E-04 S

LN-ET 1.312E-04 2.426E-04 1.331E-02 6.309E-05 3.926E-05 LN-ET 2.498E-02 7.466E-02 8.964E+00 1.298E-03 1.485E- S

BBi-bas 3.009E-04 4.266E-04 4.614E-04 2.889E-04 3.926E-05 BBi-bas 5.729E-02 1.313E-01 3.107E-01 5.944E-03 1.485E- S

BBi-sec 4.826E-03 4.843E-03 4.918E-03 3.784E-03 4.237E-05 BBi-sec 9.189E-01 1.490E+00 3.312E+00 7.786E-02 1.603E- S

bbe-sec 5.190E-03 2.811E-03 3.005E-03 2.090E-03 4.115E-05 bbe-sec 9.882E-01 8.651E-01 2.024E+00 4.300E-02 1.557E- S

AI 2.800E-03 1.397E-03 6.515E-03 7.766E-04 3.930E-05 AI 5.331E-01 4.299E-01 4.388E+00 1.598E-02 1.487E- S

LN-Th 9.872E-04 8.485E-04 4.588E-02 2.590E-04 3.924E-05 LN-Th 1.880E-01 2.611E-01 3.090E+01 5.329E-03 1.484E- S

Muscle 3.970E-06 2.107E-06 8.864E-07 9.572E-06 3.924E-05 Muscle 7.559E-04 6.484E-04 5.970E-04 1.969E-04 1.484E- K1

Ovaries 3.970E-06 2.106E-06 8.857E-07 9.572E-06 3.924E-05 Ovaries 7.559E-04 6.481E-04 5.965E-04 1.969E-04 1.484E- K1

Pancreas 3.970E-06 2.106E-06 8.857E-07 9.571E-06 3.924E-05 Pancreas 7.559E-04 6.481E-04 5.965E-04 1.969E-04 1.484E- K1

R_Marrow 1.178E-05 6.252E-06 2.818E-06 2.824E-05 1.155E-04 R_Marrow 2.243E-03 1.924E-03 1.898E-03 5.810E-04 4.369E- K1

Skin 3.970E-06 2.106E-06 8.857E-07 9.571E-06 3.924E-05 Skin 7.559E-04 6.481E-04 5.965E-04 1.969E-04 1.484E- K1

Spleen 3.970E-06 2.106E-06 8.859E-07 9.571E-06 3.924E-05 Spleen 7.559E-04 6.481E-04 5.966E-04 1.969E-04 1.484E- K1

Testes 3.970E-06 2.106E-06 8.856E-07 9.571E-06 3.924E-05 Testes 7.559E-04 6.481E-04 5.964E-04 1.969E-04 1.484E- K1

Thymus 3.970E-06 2.106E-06 8.858E-07 9.571E-06 3.924E-05 Thymus 7.559E-04 6.481E-04 5.966E-04 1.969E-04 1.484E- K1

Thyroid 3.970E-06 2.106E-06 8.857E-07 9.571E-06 3.924E-05 Thyroid 7.559E-04 6.481E-04 5.965E-04 1.969E-04 1.484E- K1

GB_Wall 3.970E-06 2.106E-06 8.857E-07 9.571E-06 3.924E-05 GB_Wall 7.559E-04 6.481E-04 5.965E-04 1.969E-04 1.484E- K1

Ht_Wall 3.970E-06 2.107E-06 8.867E-07 9.572E-06 3.924E-05 Ht_Wall 7.559E-04 6.484E-04 5.972E-04 1.969E-04 1.484E-04 K1

Uterus 3.970E-06 2.106E-06 8.856E-07 9.571E-06 3.924E-05 Uterus 7.559E-04 6.481E-04 5.964E-04 1.969E-04 1.484E-04 K1

Lung 3.515E-03 2.280E-03 4.112E-03 1.633E-03 4.042E-05 Lung 6.692E-01 7.017E-01 2.769E+00 3.360E-02 1.529E-04 S

ET 1.550E-03 2.925E-03 7.534E-03 1.252E-03 3.930E-05 ET 2.951E-01 9.002E-01 5.074E+00 2.576E-02 1.487E-04 S

Colon 4.951E-06 3.547E-06 2.350E-06 1.084E-05 3.985E-05 Colon 9.427E-04 1.092E-03 1.583E-03 2.230E-04 1.507E-04 S

Oesophagus 3.970E-06 2.106E-06 8.858E-07 9.571E-06 3.924E-05 Oesophagus 7.559E-04 6.481E-04 5.966E-04 1.969E-04 1.484E-04 K1

Remainder 4.354E-06 2.312E-06 9.846E-07 1.049E-05 4.296E-05 Remainder 8.290E-04 7.115E-04 6.631E-04 2.158E-04 1.625E-04 K1

IRF 5.252E-03 3.249E-03 1.485E-03 4.860E-02 2.644E-01
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

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ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 17 of 18

Table H-14. Chronic intake (1000 days) with urine bioassay on day after end of intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv/intake) (Sv) (Sv (Sv (Sv) (Sv Biggest

Adrenals 3.888E-05 2.063E-05 8.646E-06 9.377E-05 3.845E-04 Adrenals 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

UB_Wall 3.902E-05 2.070E-05 8.677E-06 9.409E-05 3.858E-04 UB_Wall 1.418E-03 1.419E-03 2.414E-03 1.410E-03 1.412E-03 S

Bone_Sur 1.135E-03 6.020E-04 2.700E-04 2.718E-03 1.112E-02 Bone_Sur 4.124E-02 4.126E-02 7.512E-02 4.072E-02 4.068E-02 S

Brain 3.888E-05 2.063E-05 8.643E-06 9.377E-05 3.845E-04 Brain 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Breasts 3.888E-05 2.063E-05 8.646E-06 9.377E-05 3.845E-04 Breasts 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

St_Wall 3.924E-05 2.115E-05 9.174E-06 9.424E-05 3.847E-04 St_Wall 1.426E-03 1.450E-03 2.552E-03 1.412E-03 1.408E-03 S

SI_Wall 3.977E-05 2.193E-05 9.962E-06 9.491E-05 3.850E-04 SI_Wall 1.445E-03 1.503E-03 2.772E-03 1.422E-03 1.409E-03 S

ULI_Wall 4.424E-05 2.849E-05 1.663E-05 1.007E-04 3.878E-04 ULI_Wall 1.607E-03 1.953E-03 4.627E-03 1.509E-03 1.419E-03 S

LLI_Wall 5.461E-05 4.368E-05 3.208E-05 1.141E-04 3.942E-04 LLI_Wall 1.984E-03 2.994E-03 8.925E-03 1.709E-03 1.442E-03 S

Kidneys 4.184E-04 2.219E-04 1.032E-04 9.996E-04 4.085E-03 Kidneys 1.520E-02 1.521E-02 2.871E-02 1.497E-02 1.495E-02 S

Liver 1.561E-04 8.281E-05 3.723E-05 3.735E-04 1.527E-03 Liver 5.672E-03 5.675E-03 1.036E-02 5.595E-03 5.587E-03 S

ET1-bas 4.122E-05 2.542E-05 1.442E-05 9.852E-05 3.892E-04 ET1-bas 1.498E-03 1.742E-03 4.012E-03 1.476E-03 1.424E-03 S

ET2-bas 1.553E-02 2.930E-02 7.535E-02 1.254E-02 3.855E-04 ET2-bas 5.643E-01 2.008E+00 2.096E+01 1.879E-01 1.410E-03 S

LN-ET 1.311E-03 2.426E-03 1.317E-01 6.290E-04 3.847E-04 LN-ET 4.763E-02 1.663E-01 3.664E+01 9.423E-03 1.408E-03 S

BBi-bas 3.008E-03 4.265E-03 4.614E-03 2.887E-03 3.849E-04 BBi-bas 1.093E-01 2.923E-01 1.284E+00 4.325E-02 1.408E-03 S

BBi-sec 4.826E-02 4.843E-02 4.918E-02 3.783E-02 4.147E-04 BBi-sec 1.753E+00 3.319E+00 1.368E+01 5.667E-01 1.517E-03 S

bbe-sec 5.190E-02 2.811E-02 3.005E-02 2.090E-02 4.021E-04 bbe-sec 1.886E+00 1.927E+00 8.360E+00 3.131E-01 1.471E-03 S

AI 2.799E-02 1.397E-02 6.511E-02 7.762E-03 3.851E-04 AI 1.017E+00 9.574E-01 1.811E+01 1.163E-01 1.409E-03 S

LN-Th 9.871E-03 8.484E-03 4.522E-01 2.588E-03 3.845E-04 LN-Th 3.586E-01 5.815E-01 1.258E+02 3.877E-02 1.407E-03 S

Muscle 3.889E-05 2.063E-05 8.651E-06 9.377E-05 3.845E-04 Muscle 1.413E-03 1.414E-03 2.407E-03 1.405E-03 1.407E-03 S

Ovaries 3.888E-05 2.063E-05 8.644E-06 9.377E-05 3.845E-04 Ovaries 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Pancreas 3.888E-05 2.063E-05 8.644E-06 9.377E-05 3.845E-04 Pancreas 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

R_Marrow 1.170E-04 6.208E-05 2.792E-05 2.804E-04 1.147E-03 R_Marrow 4.251E-03 4.255E-03 7.768E-03 4.201E-03 4.197E-03 S

Skin 3.888E-05 2.063E-05 8.644E-06 9.377E-05 3.845E-04 Skin 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Spleen 3.888E-05 2.063E-05 8.646E-06 9.377E-05 3.845E-04 Spleen 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Testes 3.888E-05 2.063E-05 8.643E-06 9.377E-05 3.845E-04 Testes 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Thymus 3.888E-05 2.063E-05 8.645E-06 9.377E-05 3.845E-04 Thymus 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Thyroid 3.888E-05 2.063E-05 8.644E-06 9.377E-05 3.845E-04 Thyroid 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

GB_Wall 3.888E-05 2.063E-05 8.644E-06 9.377E-05 3.845E-04 GB_Wall 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Ht_Wall 3.889E-05 2.063E-05 8.654E-06 9.377E-05 3.845E-04 Ht_Wall 1.413E-03 1.414E-03 2.408E-03 1.405E-03 1.407E-03 S

Uterus 3.888E-05 2.063E-05 8.643E-06 9.377E-05 3.845E-04 Uterus 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Lung 3.515E-02 2.279E-02 4.109E-02 1.633E-02 3.957E-04 Lung 1.277E+00 1.562E+00 1.143E+01 2.446E-01 1.448E-03 S

ET 1.550E-02 2.924E-02 7.534E-02 1.251E-02 3.855E-04 ET 5.632E-01 2.004E+00 2.096E+01 1.874E-01 1.410E-03 S

Colon 4.870E-05 3.502E-05 2.327E-05 1.065E-04 3.906E-04 Colon 1.769E-03 2.400E-03 6.474E-03 1.595E-03 1.429E-03 S

Oesophagus 3.888E-05 2.063E-05 8.645E-06 9.377E-05 3.845E-04 Oesophagus 1.413E-03 1.414E-03 2.405E-03 1.405E-03 1.407E-03 S

Remainder 4.273E-05 2.269E-05 9.632E-06 1.029E-04 4.218E-04 Remainder 1.553E-03 1.555E-03 2.680E-03 1.542E-03 1.543E-03 S

IRF 2.752E-02 1.459E-02 3.594E-03 6.675E-02 2.733E-01
Document No. ORAUT-TKBS-0038-5 Revision No. 01 Effective Date: 

04/

26/2010 Page 82 of 82

ATTACHMENT H

CALCULATION OF DOSES FROM INTAKES OF URANIUM ALUMINIDE

Page 18 of 18

Table H-15. Chronic intake (10000 days) with urine bioassay on day after end of intake.

K1 DCF K5 DCF S DCF M DCF F DCF K1 K5 S M F

(Sv/intake) (Sv/intake (Sv/intake) (Sv/intake (Sv/intake) (Sv) (Sv) (Sv (Sv (Sv) Biggest

Adrenals 3.021E-04 1.603E-04 6.437E-05 7.313E-04 3.002E-03 Adrenals 1.053E-02 1.054E-02 9.600E-03 1.067E-02 1.072E-02 F

UB_Wall 3.034E-04 1.610E-04 6.466E-05 7.345E-04 3.015E-03 UB_Wall 1.058E-02 1.059E-02 9.643E-03 1.072E-02 1.077E-02 F

Bone_Sur 1.042E-02 5.528E-03 2.374E-03 2.500E-02 1.023E-01 Bone_Sur 3.634E-01 3.634E-01 3.540E-01 3.649E-01 3.654E-01 F

Brain 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Brain 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Breasts 3.021E-04 1.603E-04 6.437E-05 7.313E-04 3.002E-03 Breasts 1.053E-02 1.054E-02 9.600E-03 1.067E-02 1.072E-02 F

St_Wall 3.057E-04 1.655E-04 6.964E-05 7.360E-04 3.005E-03 St_Wall 1.066E-02 1.088E-02 1.039E-02 1.074E-02 1.073E-02 K5

SI_Wall 3.109E-04 1.733E-04 7.752E-05 7.428E-04 3.008E-03 SI_Wall 1.084E-02 1.139E-02 1.156E-02 1.084E-02 1.075E-02 S

ULI_Wall 3.557E-04 2.389E-04 1.442E-04 8.008E-04 3.036E-03 ULI_Wall 1.240E-02 1.571E-02 2.151E-02 1.169E-02 1.085E-02 S

LLI_Wall 4.594E-04 3.908E-04 2.986E-04 9.351E-04 3.100E-03 LLI_Wall 1.602E-02 2.569E-02 4.453E-02 1.365E-02 1.107E-02 S

Kidneys 4.148E-03 2.200E-03 9.812E-04 9.914E-03 4.052E-02 Kidneys 1.447E-01 1.446E-01 1.463E-01 1.447E-01 1.447E-01 S

Liver 1.457E-03 7.732E-04 3.284E-04 3.496E-03 1.431E-02 Liver 5.081E-02 5.083E-02 4.898E-02 5.103E-02 5.112E-02 F

ET1-bas 3.255E-04 2.082E-04 1.207E-04 7.788E-04 3.050E-03 ET1-bas 1.135E-02 1.369E-02 1.800E-02 1.137E-02 1.090E-02 S

ET2-bas 1.552E-01 2.930E-01 7.535E-01 1.252E-01 3.013E-03 ET2-bas 5.412E+00 1.926E+01 1.124E+02 1.828E+00 1.076E-02 S

LN-ET 1.303E-02 2.421E-02 1.126E+00 6.083E-03 3.005E-03 LN-ET 4.544E-01 1.592E+00 1.679E+02 8.879E-02 1.073E-02 S

BBi-bas 2.999E-02 4.261E-02 4.611E-02 2.866E-02 3.007E-03 BBi-bas 1.046E+00 2.801E+00 6.877E+00 4.184E-01 1.074E-02 S

BBi-sec 4.824E-01 4.841E-01 4.917E-01 3.781E-01 3.304E-03 BBi-sec 1.682E+01 3.183E+01 7.333E+01 5.519E+00 1.180E-02 S

bbe-sec 5.188E-01 2.810E-01 3.001E-01 2.088E-01 3.179E-03 bbe-sec 1.809E+01 1.847E+01 4.476E+01 3.048E+00 1.136E-02 S

AI 2.799E-01 1.397E-01 6.376E-01 7.744E-02 3.009E-03 AI 9.761E+00 9.185E+00 9.509E+01 1.130E+00 1.075E-02 S

LN-Th 9.863E-02 8.480E-02 3.643E+00 2.568E-02 3.002E-03 LN-Th 3.439E+00 5.575E+00 5.433E+02 3.749E-01 1.072E-02 S

Muscle 3.021E-04 1.603E-04 6.441E-05 7.313E-04 3.002E-03 Muscle 1.053E-02 1.054E-02 9.606E-03 1.067E-02 1.072E-02 F

Ovaries 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Ovaries 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Pancreas 3.021E-04 1.603E-04 6.435E-05 7.313E-04 3.002E-03 Pancreas 1.053E-02 1.054E-02 9.597E-03 1.067E-02 1.072E-02 F

R_Marrow 1.079E-03 5.723E-04 2.478E-04 2.587E-03 1.059E-02 R_Marrow 3.763E-02 3.763E-02 3.696E-02 3.776E-02 3.783E-02 F

Skin 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Skin 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Spleen 3.021E-04 1.603E-04 6.436E-05 7.313E-04 3.002E-03 Spleen 1.053E-02 1.054E-02 9.598E-03 1.067E-02 1.072E-02 F

Testes 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Testes 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Thymus 3.021E-04 1.603E-04 6.436E-05 7.313E-04 3.002E-03 Thymus 1.053E-02 1.054E-02 9.598E-03 1.067E-02 1.072E-02 F

Thyroid 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Thyroid 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

GB_Wall 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 GB_Wall 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Ht_Wall 3.021E-04 1.603E-04 6.444E-05 7.313E-04 3.002E-03 Ht_Wall 1.053E-02 1.054E-02 9.610E-03 1.067E-02 1.072E-02 F

Uterus 3.021E-04 1.603E-04 6.434E-05 7.313E-04 3.002E-03 Uterus 1.053E-02 1.054E-02 9.595E-03 1.067E-02 1.072E-02 F

Lung 3.514E-01 2.278E-01 4.055E-01 1.631E-01 3.114E-03 Lung 1.225E+01 1.498E+01 6.047E+01 2.381E+00 1.112E-02 S

ET 1.549E-01 2.924E-01 7.532E-01 1.249E-01 3.013E-03 ET 5.402E+00 1.922E+01 1.123E+02 1.823E+00 1.076E-02 S

Colon 4.003E-04 3.042E-04 2.106E-04 8.585E-04 3.064E-03 Colon 1.396E-02 2.000E-02 3.141E-02 1.253E-02 1.095E-02 S

Oesophagu 3.021E-04 1.603E-04 6.436E-05 7.313E-04 3.002E-03 Oesophagus 1.053E-02 1.054E-02 9.598E-03 1.067E-02 1.072E-02 F

Remainder 3.411E-04 1.812E-04 7.392E-05 8.241E-04 3.381E-03 Remainder 1.189E-02 1.191E-02 1.102E-02 1.203E-02 1.208E-02 F

IRF 2.868E-02 1.521E-02 6.705E-03 6.851E-02 2.799E-01