ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 18

Document Title:

Maximum Internal Dose Estimates for Certain 

DOE Complex Claims 

Document Number: ORAUT-OTIB-0002

Revision: 02

Effective Date: 02/07/2007

Type of Document: OTIB

Supersedes: Revision 01

Subject Expert(s): Eugene M. Rollins and Elizabeth M. Brackett

Site Expert(s): N/A

Approval: Signature on File Approval Date: 01/12/2007

John M. Byrne, Document Owner

Concurrence: Signature on File Concurrence Date: 01/26/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 01/24/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 02/07/2007

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-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 2 of 18

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

11/07/2003 00 New document to establish Technical Information Bulletin for 

Maximum Internal Dose Estimates for Certain DOE Complex Claims. 

First approved issue. Initiated by Eugene M. Rollins.

01/10/2004 01 Approved Issue of Revision 01. Initiated by Eugene M. Rollins.

01/23/2004 01 PC-1 Added information on pages 8 and 9 regarding the employee’s initial 

hire date after 1969 and revised hire date for Hanford employees. 

Deleted the following information: The covered employee’s initial hire 

date was 1970 or later. For Hanford employees, the covered 

employee’s initial hire date was 1953 or later. See Attachment B for 

discussion. Approved issue of page change. Initiated by Eugene M. 

Rollins.

05/07/2004 01 PC-2 Lists additional organs to Table 3.1.1-4 on page 8. Adds information 

on page 9 regarding conditions to determine the potential for internal 

exposure – for covered employee’s initial hire date after 1952 and not 

participating in a bioassay program prior to 1970 and the covered 

employees initial hire date prior to 1970 with an evaluation of working 

conditions. Initiated by Pam Olsen. 

02/07/2007 02 Approved Revision 02 revised for biennial review. Constitutes a total 

rewrite of the document. Formatting and terminology changes were 

made to make the document consistent with current practices. 

Attachment A and associated references from Rev. 01 were deleted 

as they are no longer used. No changes occurred as a result of 

formal internal review. Incorporates NIOSH formal review 

comments. This revision results in no change to the assigned dose 

and no PER is required. Initiated by Elizabeth M. Brackett.
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 3 of 18

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

4

1.0 Introduction ..............................................................................................................................5

2.0 Purpose ...................................................................................................................................5

3.0 Background..............................................................................................................................5

4.0 Implausible Undiscovered Intakes............................................................................................6

5.0 Calculation of Internal Dose from Maximum Intakes............................................................... 10

6.0 Applications and Limitations................................................................................................... 10

References ........................................................................................................................................12

ATTACHMENT A, USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL 

 INFORMATION BULLETIN 0002 FOR HANFORD CLAIMS FROM 1953 

 FORWARD ..........................................................................................................14

A.1 Purpose ................................................................................................................................14

A.2 Background............................................................................................................................14

A.2.1 Brief Description of the Radiation Protection Program at the Hanford Site.................. 14

A.2.2 Air Monitoring .............................................................................................................15

A.2.3 Bioassay Monitoring ...................................................................................................16

A.2.4 Other Considerations..................................................................................................17

A.3 Conclusions ...........................................................................................................................17

LIST OF TABLES

TABLE TITLE

4-1 Fractional retention in whole body after initial clearance from lung........................................

PAGE

7

4-2 Radionuclides of interest, MPBBs, assumed intake absorption types, 

derived maximum plausible intake quantities, absorption types, and 

radiation types, used for estimating organ doses at DOE sites.............................................. 9

4-3 Total intakes and possible average air concentrations ........................................................ 10

4-4 Organs appropriate for application of the maximum internal dose for certain 

DOE complex claims........................................................................................................... 10

A-1 Comparison of MPBBs and air MPCs between NBS Handbooks 52 and 69....................... 16
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 4 of 18

ACRONYMS AND ABBREVIATIONS

CFR Code of Federal Regulations

d day

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

g gram

hr hour 

IMBA Integrated Modules for Bioassay Analysis

L liter

MDA minimum detectable activity

mg milligram

mL milliliter

MPBB maximum permissible body burden

MPC maximum permissible concentration

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NTS Nevada Test Site

TIB technical information bulletin

U.S.C. United States Code

wk week

yr year

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 5 of 18

1.0 INTRODUCTION

Technical information bulletins (TIBs) 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). TIBs may be used to assist NIOSH staff in the completion of 

individual dose reconstructions.

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 of 2000 [42 U.S.C. § 7384l(5) and (12)]. 

2.0 PURPOSE

The purpose of this TIB is to provide a method to facilitate timely processing of claims under the 

EEOICPA that involve cancer to an organ with little or no reported dose from internally deposited 

radionuclides that might be associated with work at DOE complex sites. 

3.0 BACKGROUND

Title 42, Part 82 of the Code of Federal Regulations (CFR) dictates the methods to be used for 

radiation dose reconstruction under the EEOICPA; 42 CFR 82.10(k) summarizes the general 

philosophy to be adopted:

Research and analysis will be determined sufficient if one of the following three 

conditions is met:

(1) From acquired experience, it is evident the estimated cumulative dose is sufficient 

to qualify the claimant for compensation (i.e., the dose produces a probability of 

causation of 50% or greater);

(2) Dose is determined using worst case assumptions related to radiation exposure 

and intake, to substitute for further research and analyses;

(3) Research and analysis indicated under steps described in paragraphs (f) - (j) of this 

section have been completed.

Worst-case assumptions will be employed under condition 2 to limit further research 

and analysis only for claims for which it is evident that further research and analysis will 

not produce a compensable level of radiation dose (a dose producing a probability of 

causation of 50% or greater), because using worst-case assumptions it can be 

determined that the employee could not have incurred a compensable level of radiation 

dose.

Worst-case assumption is defined in 42 CFR 82.5(r) as:

A term used to describe a type of assumption used in certain instances for certain dose 

reconstruction conducted under this rule [42 CFR 82]. It assigns the highest 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 6 of 18

reasonably possible value, based on reliable science, documented experience, and 

relevant data to a radiation dose of a covered employee.

A number of EEOICPA claims have been submitted that can be expedited based on paragraph (2) 

above. These cases involve cancer of an organ that does not concentrate any of the primary 

radionuclides expected to be found at DOE sites and with little or no apparent internal dose to that 

organ. 

In accordance with Internal Dose Reconstruction Implementation Guideline (NIOSH 2002), internal 

dose is assigned to employees who were monitored but had no detectable activity in their samples 

and to employees who were not included in a bioassay program because there is some amount of 

intake and associated dose that is not detectable by an internal dosimetry program. To expedite dose 

reconstructions, cases that meet the criteria can be evaluated with the method described in this TIB. 

If the outcome yields a probability of causation greater than 50%, a dose reconstruction using more 

reasonable assumptions will be performed.

4.0 IMPLAUSIBLE UNDISCOVERED INTAKES

For cases in which it is considered likely that the employee had no significant internal radiation 

exposure, a method to expedite claims has been developed in accordance with 42 CFR 82.10(k)(2). 

This method assumes the “largest reasonably possible value” of the source term that consisted of 

radionuclides that are or were typically the more significant radionuclides (by either preponderance or 

by internal dose significance) on a site. For this worst-case estimate of internal dose, it is assumed 

that on the first day of employment the worker had an acute inhalation intake of each of the 

radionuclides in the source term in the amounts listed below. 

Assumptions

Based on historical data, it is believed to be unlikely that such an intake could have occurred without 

being detected by workplace monitoring at the time. It is also believed that this is an overestimate of 

internal dose for an unmonitored worker or a worker with no bioassay results exceeding detection 

thresholds. 

Additional assumptions to develop this method are: 

• All intakes are inhalations of standard 5-μm activity median aerodynamic diameter, except for 

131I, which is assumed to be in vapor form (class SR-1).

• The most soluble form of the radionuclide specified in ICRP (1995) was used to maximize 

dose to systemic organs, except as noted below; dose to lung is not germane to this exercise.

• Because maximum permissible body burdens (MPBBs) were the derived limits for so many 

years, the assumed implausible uptake was based on a percentage of the radionuclidespecific MPBB for soluble chemical forms as defined in National Bureau of Standards (NBS) 

Handbook 69 [also referred to as National Committee on Radiation Protection (NCRP) 

Publication 22] (NBS 1959). It was assumed that an intake that resulted in 10% of an MPBB 

would likely not have occurred to an unmonitored worker or would have likely resulted in a 

readily noticeable bioassay result in a monitored worker, readily noticeable air sample, or other 

indicator of personnel contamination. In other words, an event that provided the possibility of 

an intake that would have resulted in a body burden exceeding 10% of the MPBB would not 

have gone unnoticed and there would be some indication in the worker’s records. This 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 7 of 18

assumption applies to facilities with active radiation protection programs where bioassay or airmonitoring programs were present and able to detect such intakes. The current ICRP 

methodology is used to calculate doses from these implausible intakes.

• For types F and M materials, the associated derived intakes (i.e., intake resulting in a 10% 

MPBB) were assumed to be 10 and 20 times 10% of the MPBB, respectively. The factors of 

10 and 20 are based on the current ICRP models and approximate the differences between an 

intake and the activity that is present in the body after the initial clearance of the short-term 

compartments. These factors are used to relate body burden, the historical quantity of control

that was based on ICRP Publication 2 (ICRP 1959) methods, to intake, the present quantity of 

control that is based on current ICRP methods. These factors were estimated from tables in 

the November 2002 issue of Health Physics that list the intake retention fraction for the whole 

body (without the extrathoracic (ET) region) as a function of time after acute intake for different 

elements and inhalation types (Potter 2002). Because initial deposition in the nonsystemic 

organs was not considered by ICRP (1959) to be part of the body burden, the selected 

retention fractions allowed some time for the rapid clearance components. Table 4-1 lists 

examples. 

Table 4-1. Fractional retention in whole body after initial clearance from lung.

a,b

Element

Inhalation type F Inhalation type M

3 d after intake 4 d after intake 60 d after intake 90 d after intake

Strontium 0.227 0.199 - -

Cesium 0.449 0.440 - -

Cerium - - 0.0683 0.0654

Ruthenium 0.251 0.227 0.0392 0.0392

Barium 0.147 0.104 - -

Lanthanum 0.307 0.291 0.0683 0.0654

Zirconium 0.278 0.252 0.0496 0.0448

Niobium - - 0.0476 0.0409

Cobalt - - 0.0410 0.0337

Technetium 0.217 0.160 0.0321 0.0238

Europium - - 0.0605 0.0568

a. Source: Potter (2002). 

b. - = not applicable. 

• The assumption of type S for 58Co and 60Co is used because it results in larger doses to the 

systemic organs because of the high-energy photons. Although the logic in the paragraph

above does not directly apply to insoluble (type S) material, the fractional retention in the 

whole body is similar for type M and type S at 60 and 90 d, so the derived intake is estimated 

as 20 times the 10% MPBB.

• Zinc-65 is classified as type S in ICRP (1995). As with the 58Co and 60Co type S materials, the 

logic of determining a body burden-to-intake conversion factor does not directly apply, but use 

of a factor of 20 does not seem unreasonable. 

• Manganese-54 type M has a larger dose conversion factor for most organs and tissues and 

was generally more favorable to claimants than type F.

• This method applies only to intakes of particulate radioactive material and radioiodines. These 

intakes are in addition to any intakes of 3

H, 14C, or radon/thoron and their progeny, as 

applicable. 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 8 of 18

• To be generic to most DOE sites, intakes were assumed to involve the most plausible 

radionuclides for all the sites, even though it is implausible that one worker had intakes of all 

the radionuclides. Two groups of radionuclides were considered: those from sites having one 

or more reactors and those from sites without a reactor. The Nevada Test Site (NTS) was 

considered a special case for which these two groups of radionuclides might not be 

appropriate. A review indicated that after 1971 the source term at NTS, except in the tunnels, 

was consistent with that chosen for reactor sites. 

• For sites without a reactor, the actual ratio of radionuclides in mixtures is not relevant because 

each radionuclide is used at its maximum amount. For sites with reactors, the fission products 

were chosen for the list based on their relative abundance in fuel or contamination after 180 d 

of post-irradiation cooling time. This is conservative for any contamination at facilities other 

than the reactors themselves. 

Table 4-2 lists the radionuclides of interest in relation to these assumptions.

A review of the assumed intakes from the method described above and associated possible air 

concentrations for different periods indicates that the uranium intakes based on soluble uranium might

not have been sufficiently conservative for uranium facilities that controlled their programs based on 

consideration of uranium exposure alone. The derived intakes for 234U and 238U were increased by a 

factor of 100 to ensure that this scenario bounds doses to those employees who worked at uranium 

facilities but were not included in a bioassay program or had no detectable occupationally related 

activity in their samples. Derived intake values for 234U and 238U can be lowered to 50 and 5 nCi, 

respectively, for employees who worked in areas that were controlled based on the possible presence 

of other alpha emitters such as 238

Pu and 239Pu. 

After the intakes of the radionuclides were determined, the intakes from the alpha-emitting

radionuclides and the beta/gamma-emitting radionuclides were summed separately to show the total 

intake of each for the reactor and nonreactor sites. These alpha and beta/gamma intakes were then 

used to consider what the average air concentrations would have been if the intakes had occurred 

chronically rather than acutely. A breathing rate of 2.4 × 109

 mL/yr, based on Reference Man, was 

used to calculate the air concentrations for 1, 10, and 30 years that would produce the given intakes. 

Site air concentration criteria (area controls, respirator usage, etc.) were based on the permissible air 

concentrations of radionuclides assumed to be present in significant quantities. At many facilities, air 

concentration controls were specified for alpha emitters, based on 238/239/240Pu, and beta/gamma

emitters based on 90Sr. The typical air concentration controls listed in Table 4-3 are based on the 

radionuclide air concentration values in NBS (1959), AEC (1968), and 10 CFR 835, which for most 

radionuclides did not change significantly over time. In later years, the beta/gamma limit might have 

increased to 2 × 10-9 μCi/mL, but the control criteria also changed and were based on 10% of the limit, 

which would equate to an air concentration of 2 × 10-10 μCi/mL. 

Table 4-3 shows that even chronic intakes at these levels would exceed typical air concentration 

controls that were being used at most sites since the 1950s to manage access to and work practices 

in areas that had increased radioactivity in air. Acute intakes are assumed for this worst-case 

approach because they result in larger assignments of dose.

Table 4-3 also shows that workers who spent significant time in areas where uranium air 

concentration controls were used to manage access and work practices do not appear to be good 

candidates for the nonuranium facility approaches. The nonuranium facility approaches can be used 

for reconstruction of doses for claims that did not involve exposure in uranium process environments 

for more than a few years. The uranium intakes are included with the alpha intakes, because it is 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 9 of 18

Table 4-2. Radionuclides of interest, MPBBs, assumed intake absorption types, derived maximum 

plausible intake quantities, absorption types, and radiation types used for estimating organ doses at 

DOE sites. 

Radionuclide

Soluble (S) 

MPBB (nCi)a

Fractional 

retention based 

on type:

Derived 

intakeb (nCi)

Absorption type 

applied for dose 

determination

Sites without a reactor:

Sr-90 2,000 F 2,000 F

Tc-99 10,000 F 10,000 F

Cs-137 30,000 F 30,000 F

U-234 50 F 5,000c F

U-238 5 F 500c F

Am-241d 50 M 100 M

Cm-244 100 M 200 M

Np-237 60 M 120 M

Pu-238 40 M 80 M

Pu-239 40 M 80 M

Th-230e 50 M 100 M

Cf-252 40 M 80 M

Mn-54 20,000 M 40,000 M

Sites with a reactor–all those above plus:

Co-58 30,000 M 60,000 S

Co-60 10,000 M 20,000 S

Fe-59 20,000 F 20,000 F

Zn-65 60,000 M 120,000 S

Y-91 5,000 M 10,000 M

Nb-95 40,000 M 80,000 M

Zr-95 20,000 F 20,000 F

Ru-103 20,000 F 20,000 F

Ru-106 3,000 F 3,000 F

I-131 700 F 700 F

Ce-141 30,000 M 60,000 M

Ce-144 5,000 M 10,000 M

Pm-147 60,000 M 120,000 M

Eu-154 5,000 M 10,000 M

Eu-155 70,000 M 140,000 M

a. Source: NBS (1959).

b. Derived maximum plausible intake based on 10% of the soluble NBS (1959) MPBB multiplied by factors of 10 and 20 for 

assumed intake absorption types of F and M, respectively.

c. The uranium values for the general approach are increased to ensure that this is a bounding approach for all facilitytypes. The lower values of 50 nCi for U-234 and 5 nCi for U-238 can be used to assign worst-case doses to individuals 

who did not work in areas where there were uranium exposure concerns.

d. Pu-241 is accounted for by the Am-241 intake value.

e. Thorium-230 can be used as a surrogate for Th-232. 

unlikely that the measured alpha air concentration would have excluded uranium. Uranium is shown 

by itself for nonreactor facilities, because the air concentration control in a uranium facility was likely 

to be larger than the cited gross alpha air concentration control. For facilities with reactor source 

terms, the uranium air concentration was not listed because it is believed that the control would have 

been on the gross alpha air concentration and because the uranium concentration would be the same 

as the comparable nonreactor facility.
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 10 of 18

Table 4-3. Total intakes and possible average air concentrations. 

Facility type

Radiation 

type

Intake 

(μCi)

Air 

concentration 

based on 

1-yr intake 

(μCi/mL)

Air 

concentration 

based on 

10-yr intake 

(μCi/mL)

Air 

concentration 

based on 

30-yr intake 

(μCi/mL)

Typical air 

concentration 

controls 

(μCi/mL)

Nonreactor and 

nonuranium

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10

Uranium 0.055 2.3E-11 2.3E-12 7.6E-13 7 E-11

Reactor and 

nonuranium

Alpha 0.815 3.4E-10 3.4E-11 1.1E-11 2 E-12

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10

Nonreactor and 

uranium

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12

Beta/gamma 42 1.8E-08 1.8E-09 5.8E-10 3 E-10

Uranium 5.5 2.3E-09 2.3E-10 7.6E-11 7 E-11

Reactor and 

uranium

Alpha 6.26 2.6E-09 2.6E-10 8.7E-11 2 E-12

Beta/gamma 776 3.2E-07 3.2E-08 1.1E-08 3 E-10

To encompass employees who might have worked in facilities where uranium was the primary 

radionuclide of concern for internal dose, each assumed uranium intake was increased by a factor of 

100. The factor of 100 was chosen to ensure that this approach adequately bounds the level of 

exposure that could have gone unnoticed at uranium facilities.

This approach can also be used for any cancer site for which the highest nonmetabolic organ is 

specified for internal dose assessment in ORAUT (2006). 

Table 4-4 lists the organs that are appropriate for application of maximum internal dose.

Table 4-4. Organs appropriate for application of the 

maximum internal dose for certain DOE complex claims. 

Adrenals Heart Ovaries

Brain Muscle Thymus

Breast Liver U. Bladder

Colon Eye Uterus

Kidneys Pancreas Bone surfaces

Red bone marrow Skin Testes

Esophagus Spleen

G. bladder Stomach

5.0 CALCULATION OF INTERNAL DOSE FROM MAXIMUM INTAKES

The calculation of organ dose assumes that the covered employee had maximum intakes of all the 

radionuclides listed in Table 4-2. While individual exposures to a single radionuclide from the 

mixtures defined in this method possibly could be underestimated, it is believed that the individual 

dose determined from this method would be overestimated because, at a minimum, intakes and 

doses from 12 radionuclides are considered.

6.0 APPLICATIONS AND LIMITATIONS

The following conditions must be met to apply this approach: 

• The covered employee’s initial hire date was after 1969. 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 11 of 18

• The covered employee’s initial start date was before 1970 (for the DOE Hanford Site, before

1953), provided the dose reconstruction report includes an evaluation or explanation that

demonstrates the doses from Table 4-2 intakes overestimate the actual or potential doses 

received by the worker. The following are some examples of potential information to use in 

this evaluation. 

1. Employee job title

2. Bioassay monitoring results

3. Work location or other job conditions

4. Recorded external dose

5. Other site employees with similar jobs 

• For Hanford Site employees, the covered employee’s hire date was after 1952. See 

Attachment A for discussion. 

• The target organ must be listed in Table 4-4 or must be an organ whose dose is based on the 

highest nonmetabolic organ dose; the respiratory tract is not included because the intakes are 

not maximized for insoluble material.

• Employees who would have had no significant exposure to uranium (i.e., no uranium bioassay 

or did not work with uranium) can be assessed with the smaller uranium source term. 

• The employment must have been at a facility with an active radiation protection program.

• Unless it can be demonstrated that the nuclides are appropriate to the site and there was an 

active radiation protection program that conformed to the standards at the time, dose from 

Atomic Weapons Employers must be reconstructed by other methods.

• Dose from NTS can be assigned by this approach if the employee worked at the Site after 

1971, was not involved in tunnel work, and meets the other conditions herein. 

• Dose from tritium is to be considered separately. 

• Potential for exposure to radionuclides that are not included in Table 4-2, unless there is a 

similar radionuclide (i.e., 230Th for 232Th) and supplemental explanation of the substitution.
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 12 of 18

REFERENCES

AEC (U.S. Atomic Energy Commission), 1968, "Chapter 0524, Standards for Radiation Protection," 

AEC Manual, 0524-01, Washington, D.C., November 8. [SRDB Ref ID: 10668]

Cantril, S. T, 1945, "Tolerance Limits," memorandum to file, HW-4.159, 7-2602, Hanford Works, 

Hanford, Washington, October 18. [SRDB Ref ID: 13428]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order 

5480.11, Office of Environment, Safety and Health, Washington D.C., December 21. [SRDB 

Ref ID: 8073]

HPS (Health Physics Society), 1995, An American National Standard for Bioassay Programs for 

Uranium, ANSI Standard N13.22-1995, McLean, Virginia, October.

ICRP (International Commission on Radiological Protection), 1959, Report of Committee II on 

Permissible Dose for Internal Radiation, Publication 2, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

NBS (National Bureau of Standards), 1953, Maximum Permissible Amounts of Radioisotopes in the 

Human Body and Maximum Permissible Concentrations in Air and Water, Handbook 52, U.S. 

Department of Commerce, Washington, D.C., March 20. [SRDB Ref ID: 12332]

NBS (National Bureau of Standards), 1959, Maximum Permissible Body Burdens and Maximum 

Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure, 

Handbook 69, U.S. Department of Commerce, Washington, D.C., June 5. [SRDB Ref ID: 

11110]

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Hanford 

Site – Occupational Internal Dose, ORAUT-TKBS-0006-5, Rev. 01, Oak Ridge, Tennessee, 

November 24.

ORAUT (Oak Ridge Associated Universities Team), 2006, Internal Dosimetry Organ, External 

Dosimetry Organ, and IREP Model Selection by ICD-9 Code, ORAUT-OTIB-0005, Rev. 01 

PC-1, Oak Ridge, Tennessee, February 10.

Parker, H. M., 1948. “Health Physics, Instrumentation, and Radiation Protection,” Advances in 

Biological and Medical Physics, Volume I, Academic Press, New York, New York, p. 957. 

Parker, H. M., 1955. “Radiation Protection in the Atomic Energy Industry, a Ten-Year Review,” 

Radiology, volume 65, number 6, pp. 903–911. 

Patterson, C. M., 1949, “General Operating Tolerances,” letter to L. S. Taylor (National Bureau of 

Standards), HW-12710, Hanford Works, Richland, Washington, March 11. [SRDB Ref ID: 

7756]
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 13 of 18

Potter, C. A., 2002, "Intake Retention Fractions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68—Particulate Inhalation," Health Physics, 

volume 83, number 5, pp. 594–789. [SRDB Ref ID: 12373]

Wilson, R. H., 1987, Historical Review of Personnel Dosimetry Development and its Use in Radiation 

Protection Programs at Hanford, 1944 to the 1980s, PNL-6125, Battelle Memorial Institute, 

Pacific Northwest Laboratory, Richland, Washington, February. [SRDB Ref ID: 262]
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 14 of 18

A.1 PURPOSE 

The purpose of this attachment is to provide the basis for applying the assumptions, conditions, and 

results of this TIB to claims for Hanford Site employees before 1970. 

A.2 BACKGROUND

This TIB assigns intakes of a number of radionuclides based on the assumption that intakes that 

resulted in greater than 10% of the MPBB, as defined in ICRP Publication 2 (ICRP 1959) and NBS 

Handbook 69 (NBS 1959), would not have occurred without being detected by workplace monitoring 

or bioassay or both. For sites with reactors, which would apply to the Hanford Site, each worker is 

assigned intakes of 28 radionuclides that are considered representative of potential sources of intake. 

For situations in which more than one isotope might have been available as a potential source of 

intake, the more restrictive isotope was assigned (e.g., 137Cs for mixtures of 134Cs and 137Cs, and 90Sr 

for mixtures of 89Sr and 90Sr). The intakes and resulting doses are considered conservative in that the 

risk of a worker truly incurring an undetected intake or series of intakes during a career involving all 

the listed radionuclides at 10% of the MPBB for each is unlikely. 

The justification for extending this TIB to Hanford for hire dates earlier than 1970 requires an 

understanding of the status of the radiation protection program for the earlier years.

A.2.1

Until 1965, there was one operating contractor at Hanford and, therefore, one radiation protection 

program. Herbert M. Parker, manager of the Health Instrumentation Division and later manager of 

Hanford Laboratories, was the guiding force for radiation protection at Hanford in the 1940s and 

1950s. Dr. Parker was a leader in the field of health physics who pushed hard for continuously 

improving the science and the practice of health physics at Hanford and in the United States as a 

whole. Although managing radiation protection in general, Dr. Parker’s expertise was dosimetry, both 

external and internal. For example, he was a member of the NCRP subcommittee that developed the 

first set of MPBBs and maximum permissible concentrations (MPCs). 

Brief Description of the Radiation Protection Program at the Hanford Site

Jack Healy led developments in bioassay and internal dosimetry during this time. Dr. Healy was also 

one of the foremost experts and leaders in these disciplines. For example, he was a member of the 

NCRP subcommittee that developed the second set of MPBBs and MPCs.

Because of these men and others, radiation protection was a major concern at Hanford from the start. 

Considerable research was conducted throughout most of Hanford’s history to improve survey 

instruments, effluent cleanup mechanisms, monitoring instrumentation, and workplace air-sampling 

practices, and to understand the biokinetics associated with intakes of radionuclides. Some of these 

developments are described in Dr. Parker’s 1948 article in Advances in Biological and Medical 

Physics, “Health-Physics Instrumentation and Radiation Protection” (Parker 1948), and his 1955 

article in Radiology, “Radiation Protection in the Atomic Energy Industry, a Ten-Year Review” (Parker 

1955).

ATTACHMENT A

USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL INFORMATION 

BULLETIN 0002 FOR HANFORD SITE CLAIMS FROM 1953 FORWARD

Page 1 of 5 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 15 of 18

Nevertheless, the science and practice of monitoring for intakes of plutonium, uranium, and mixed 

fission and activation products were still in their infancy in the 1940s. Knowledge and skills 

associated with routine bioassay for plutonium and uranium were rapidly being developed and 

improved, and the scope of monitored workers was expanding. Limits were based on tolerance doses 

and tolerance air concentrations, which were forerunners of the MPBB and MPC concepts. However,

the tolerance doses were high by today’s standards; for instance, the external dose limit at Hanford 

was 0.1 rep/d and later was revised to 0.3 rep/wk [a rep was a radiation equivalent physical, defined 

as 83 ergs/g energy deposition in tissue and later redefined as 93 ergs/g]. The internal dose limit for 

alpha emitters was 0.01 rep/d (Cantril 1945). [Assuming a quality factor of 20, 0.01 rep/d

approximately equates to 200 mrem/d.] In addition, the science of calculating intake or air 

concentration limits based on a limiting dose rate to a “significant organ” was based on very little 

biokinetic knowledge for radionuclides of concern associated with the Manhattan Project. 

Therefore, applying the default intakes from this TIB to the earliest years at Hanford is not 

recommended. 

A.2.2

Air concentration limits at Hanford followed national guidance and, because of Drs. Parker and 

Healy’s involvement in setting national guidance, changes would have been implemented promptly, if 

not actually before the publication of the guidance. For example, in a 1949 letter from C. M. Patterson 

of the Hanford Health Instrument Division to Lauriston S. Taylor of the National Bureau of Standards, 

in response to a request from Dr. Taylor, Dr. Patterson replied (examples from the letter), “The 

permissible tolerances used at Hanford Works are as follows: . . . internal emitter limits – 0.3 rep per 

week to the significant organ or whatever the National Committee on Radiation Protection 

recommends” (italics not in original letter). The italicized phrase was used in several places. For

drinking water, “mixed fission products – 0.1 μc/liter – to be changed when new figure provided by 

K.Z. Morgan’s Subcommittee” (Patterson 1949). Dr. Morgan’s subcommittee was the NCRP 

Subcommittee on Permissible Internal Dose, of which Dr. Parker was a member, and which 

eventually produced the MPCs published in NBS Handbooks 52 and 69 (NBS 1953, 1959). 

Air Monitoring

Table A-1 compares the MPBBs and MPCs of key radionuclides provided as national guidance in 

NBS Handbooks 52 and 69 (NBS 1953, 1959). This table is instructive because the MPBBs used in 

this TIB are directly linked to the MPCs in NBS Handbook 69 and, therefore, to air concentration limits 

at Hanford since about 1959. However, for before 1959, the table shows what the MPCs were 

starting in 1953 and how those values compared to 1959 values.

Air monitoring at Hanford was based on gross alpha and gross beta measurements corrected for 

radon progeny. The beta MPC used at Hanford since or before 1953 was based on the MPC for 90Sr 

in NBS (1953) applicable at the time. Because, as shown in Table A-1, the MPC for 90Sr actually 

increased between 1953 and 1959, the use of the MPBBs from the 1959 document is applicable and 

conservative for the period from 1953 to 1959. Similarly, the alpha MPC at Hanford was 2 × 10-12

μCi/mL, based on 239Pu, which was unchanged from 1953. Areas with air concentrations exceeding 

either MPC were labeled as airborne contamination areas and required respiratory protection. 

Therefore, although the MPCs varied somewhat for the other radionuclides between 1953 and 1959, 

the 90Sr and 239Pu MPCs were always more limiting. Furthermore, these MPC values for 

ATTACHMENT A

USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL INFORMATION 

BULLETIN 0002 FOR HANFORD SITE CLAIMS FROM 1953 FORWARD

Page 2 of 5 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 16 of 18

Table A-1. Comparison of MPBBs and air MPCs between NBS Handbooks 52 

and 69. 

Radionuclide

MPBB (nCi) MPC air (μCi/mL)

Handbook 52 Handbook 69 Handbook 52 Handbook 69

Pu-239 soluble 40 40 2E-12 2E-12

Pu-239 insoluble 8 NA 2E-12 4E-11

U-nat soluble 200 5 1.7E-11 7E-11

U-238 soluble 5 7E-11

U-nat insoluble 9 7E-11 6E-11

U-238 insoluble NA 1E-10

Am-241 56 50 3E-11 6E-12

Sr-90 1,000 2000 2E-10 3E-10

Cs-137 90,000 30,000 2E-7 6E-8

Ce-144 5,000 5,000 7E-9 1E-8

I-131 300 700 3E-9 9E-9

Co-60 3,000 10,000 1E-6 3E-7

Ru-106 4,000 3,000 3E-8 8E-8

Y-91 15,000 5,000 4E-8 4E-8

90Sr and 239Pu (and therefore gross alpha and gross beta measurements on air samples) were 

continued via AEC (1968) and DOE (1988) into the 1980s.

The point of this discussion is that the MPBB values used for this TIB were directly linked to the air 

monitoring program at Hanford since 1953, and that the air concentration limits and response to 

exposures that potentially exceeded the limits were consistent from 1953 to at least 1989. 

A.2.3

Another key consideration in monitoring for intakes is the bioassay program. The plutonium bioassay 

program was well established and had years of operating experience at Hanford by 1953 (ORAUT

2004). In fact, the plutonium urinalysis methodology changed little from March 1953 through 

September 1983, and the minimum detectable activity (MDA) was 0.05 dpm/24-hr sample. For type 

M plutonium, this MDA translates to a possible missed intake at 365 d after an intake of 5.8 nCi total 

plutonium alpha compared to the intake of 160 nCi plutonium alpha assumed in this TIB. In addition, 

the sensitivity of the program improved throughout the period due to increased use of fecal samples in 

response to potential intake incidents. 

Bioassay Monitoring

Uranium bioassay (elemental analysis) was also well established by 1953 and was stable through 

about 1969 with an MDA of 4 μg/L. For type M recycled uranium, this MDA corresponds to a possible 

missed intake at 180 d after intake [sampling was conducted monthly for a while and semiannually 

part of the time (Wilson 1987)] of 78-nCi total uranium alpha compared to the intakes assumed in this 

TIB of 55-nCi type F uranium for nonuranium facilities and 5,500 nCi for uranium facilities. For type F 

recycled uranium, the corresponding possible missed acute intake at 180 d after intake is 163 nCi, but 

this is equivalent to 179 mg, which greatly exceeds the threshold for permanent renal damage (HPS 

1995) and therefore is an unrealistic estimate. In fact, the 55-nCi intake in this TIB is overly 

conservative for type F uranium from the acute toxicity perspective, but most uranium intakes across 

the DOE complex were chronic. Uranium sampling was usually biweekly or monthly for workers 

ATTACHMENT A

USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL INFORMATION 

BULLETIN 0002 FOR HANFORD SITE CLAIMS FROM 1953 FORWARD

Page 3 of 5 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 17 of 18

exposed to highly soluble forms of uranium, but some sampling was semiannual in the 1960s and it is 

not known under what conditions the semiannual sampling was allowed (e.g., it might have been just 

for workers exposed to insoluble forms of uranium). 

The largest MDA for the mixed fission product urinalysis method was 70 dpm/24-hr sample. When 

considered as pure type F 90Sr, this MDA corresponds to a possible missed intake at 365 d after

intake of 1,500 nCi compared to the intake of 2,000 nCi assumed in this TIB. When considered as 

pure type M 144Ce, this MDA corresponds to a possible missed intake at 365 d after intake of 

72,000 nCi compared to the intake of 10,000 nCi assumed in this TIB. This implies that an intake of 

144Ce at 10% of the MPBB would not have been detectable solely by an annual urinalysis. However, 

an acute intake of 10,000 nCi of mixed fission products would have been associated with readily 

detectable contamination on the worker or in the air. For instance, breathing air at the beta 

concentration limit of 3 × 10-10 μCi/mL for a week would have resulted in an intake of 14 nCi. Allowing 

for a factor of 100 for possible nonrepresentativeness of the air sample versus the air breathed by the 

worker still indicates that a 10,000-nCi intake would have been detected and resulted in prompt 

bioassay. 

The mixed fission product urinalysis did not detect intakes of isotopes of cesium or ruthenium or most 

activation products; however, these radionuclides are factored into the default intakes in this TIB. 

A.2.4

A final consideration is how successful the various monitoring programs were at discovering and 

documenting intakes. In 1955, Parker reported in Radiology on the status of and improvements in 

radiation protection at Hanford for the preceding 10 years (Parker 1955). Included in this journal 

article was a table showing the number of internal depositions per year for three equal parts of the 

10-year period as follows: “positive cases per year, early period – 0; middle period –1; recent period – 

30,” the recent period being about 1952 to 1954. This statistic implies that either there were more 

workers being exposed in the later period or detection capabilities had improved. The truth is 

probably both, but the point Parker was making was that the ability to detect and confirm intakes had 

improved considerably from 1944 to 1952. Since sometime before 1965, there was a policy that 

intakes leading to a body burden that exceeded 5% of the MPBB were recorded in the worker’s 

radiation exposure file. An indication of a possible intake was usually recorded along with the 

analysis that led to the 5% recording decision. Intakes well below the 5% criterion are listed in the 

records back to the 1950s. These points – the improved capabilities of detection and the 5% MPBB 

policy – are not conclusive of themselves, but they lend credence to the strength of the radiation 

protection programs by the early 1950s and the likelihood that intakes exceeding 10% of the MPBB 

would not have been missed. 

Other Considerations

A.3 CONCLUSIONS 

• Air concentration limits of 3 × 10-10 μCi/mL total beta activity and 2 × 10-12 μCi/mL total alpha 

activity were based on the same MPBBs used in this TIB and were in use at least since 1953. 

These afforded adequate protection from acute intakes. 

ATTACHMENT A

USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL INFORMATION 

BULLETIN 0002 FOR HANFORD SITE CLAIMS FROM 1953 FORWARD

Page 4 of 5 
Document No. ORAUT-OTIB-0002 Revision No. 02 Effective Date: 02/07/2007 Page 18 of 18

• Routine bioassay from 1953 forward was adequate for either acute or chronic intakes of 

plutonium or radiostrontium; was marginal for type M uranium, and would not have detected

an intake that corresponded to 10% of the MPBB for type M 144Ce if the intake occurred 365 d 

before the bioassay. However, an intake of 10,000 nCi of 144Ce would have been associated 

with a large contamination incident and would have resulted in prompt bioassay. The same is 

true for large intakes of 141Ce from exposure to short-cooled fuel. 

• Records show that before about 1952 few intakes were being discovered or at least confirmed, 

but the radiation protection program showed a large increase in detections in about 1952. The 

records also show intakes well below 5% of the MPBB starting at about that time. 

The Hanford radiation protection program, which consisted of air sampling using 90Sr and 239Pu MPCs 

for gross beta and gross alpha activities, respectively, workplace contamination monitoring, and 

routine bioassay, was adequate from 1953 forward to ensure that intakes that corresponded to 10% of 

the 1959 MPBBs for 28 radionuclides described in this TIB would have been discovered and 

documented. The assigned dose from these 28 intakes is sufficient to overestimate doses for 

Hanford employees from 1953 forward who meet all the other conditions described in this TIB. 

ATTACHMENT A

USING MAXIMUM INTERNAL DOSE ESTIMATES FROM TECHNICAL INFORMATION 

BULLETIN 0002 FOR HANFORD SITE CLAIMS FROM 1953 FORWARD

Page 5 of 5