Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

SC&A’S Evaluation of RPRT-0090, “Monitoring Feasibility

Evaluation for Exotic Radionuclides Produced by the Oak Ridge

National Laboratory Isotopes Division,” Revision 00

Contract No. 211-2014-58081 

SCA-TR-2018-SEC004, Revision 0 

Prepared by 

Robert Barton, III, CHP

Ron Buchanan, PhD, CHP 

Joe Fitzgerald, MS, MPH

SC&A, Inc.

2200 Wilson Boulevard, Suite 300 

Arlington, VA 22201-3324 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

October 9, 2018 

DISCLAIMER 

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues.

NOTICE: This document has been reviewed to identify and redact any information that is protected by 

the Privacy Act 5 U.S.C. § 552a and has been cleared for distribution. 
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SC&A, INC.: Technical Support for the Advisory Board on Radiation and 

Worker Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: SC&A’s Evaluation of RPRT-0090, “Monitoring Feasibility 

Evaluation for Exotic Radionuclides Produced by the Oak Ridge 

National Laboratory Isotopes Division,” Revision 00

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2018-SEC004

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: October 9, 2018

TASK MANAGER Joe Fitzgerald, MS, MPH [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S): John Stiver, MS, CHP [signature on file]

Record of Revisions

Revision 

Number

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0 (Draft) 10/9/2018 Initial issue
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TABLE OF CONTENTS

List of Tables .................................................................................................................................. 4

List of Figures ................................................................................................................................. 4

Abbreviations and Acronyms ......................................................................................................... 5

1 Introduction and Background ................................................................................................. 8

2 RPRT-0090 Feasiblity Evaluation: Basis for Review ............................................................ 9

3 SC&A’s Review of Scope of Exotic Radionuclide Source Terms ....................................... 10

4 SC&A’s Evaluation of Radionuclide Inventory ................................................................... 12

5 SC&A’s Evaluation of ORAUT-RPRT-0090 with Respect to In Vitro Methods ................ 15

5.1 SC&A’s Evaluation of Monitoring Capability and Bioassay Data for 179 

Radionuclides ................................................................................................................ 15

5.1.1 Gross Alpha Counting ............................................................................................... 16

5.1.2 Gross Beta and Gamma Counting ............................................................................. 17

5.1.3 SC&A’s Evaluation of RPRT-0090 Attachment A .................................................. 19

5.2 SC&A’s Evaluation of Feasibility of DR for 28 Radionuclides ................................... 21

5.2.1 Implications of Lack of Routine Bioassay Program ................................................. 22

5.2.2 Implications of Significant vs. Non-negligible Intakes ............................................ 23

5.3 SC&A’s Evaluation of Feasibility of DR for Five Iodine Radionuclides .................... 24

6 SC&A’s Evaluation of ORAUT-RPRT-0090 With Respect to In Vivo Bioassay Methods. 28

7 Post-1988 Inventory and Cleanup ......................................................................................... 34

8 Summary and Conclusions ................................................................................................... 36

8.1 Findings ........................................................................................................................ 36

8.2 Observations ................................................................................................................. 37

9 References ............................................................................................................................. 39

Appendix A: Example of Evaluation of CM-244 and RU-106 Bioassay Data .............................. 43

Appendix B: Examples of NOCTS ORNL Records ..................................................................... 45

Appendix C: Example of X-10 Database for Am-241 .................................................................. 49

Appendix D: Evaluation of Internal Dose Using a Fractional Intake Based on Amount of 

Material in Process ................................................................................................................ 50
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LIST OF TABLES

Table 1. Examples of Radionuclides Not Included in RPRT-0090 .............................................. 12

Table 2. SC&A’s Evaluation of 12 Radionuclides from Table 7-2 of RPRT-0090 ..................... 16

Table 3. Dosimetric Analysis of 28 Radionuclides ...................................................................... 21

Table A-1. Cm-244 Data ............................................................................................................... 43

Table A-2. Ru-106 Data ................................................................................................................ 44

Table C-1. Excerpt for Part of 1955 and 1956 Applicable to Am-241 from ORNL Database 

Containing 25,162 Entries ............................................................................................................ 49

Table D-1. SC&A’s Summary of Fractional Intake Data in Brodsky 1977 ................................. 50

Table D-2. Summary of Fractional Intake Data from Brodsky 1980 ........................................... 51

Table D-3. Typical Values for Modification Factors Used in Arriving at a More Realistic Intake 

Based on the Quantity of the Material in Process and the Fractional Intake Factor (10-6) ........... 52

LIST OF FIGURES

Figure 1. Excerpt from ATSDR 2008 Showing that the Largest Radioiodine Releases Occurred 

in 1956 .......................................................................................................................................... 26

Figure 2. Commercial Radioiodine Production by Year .............................................................. 27

Figure B-1. 1955 Plutonium, Uranium, and Strontium ................................................................ 45

Figure B-2. 1955 Thorium ............................................................................................................ 46

Figure B-3. 1964 Cm-244 ............................................................................................................. 47

Figure B-4. 1965 Cm-242 and Am-241 ........................................................................................ 48
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ABBREVIATIONS AND ACRONYMS

Advisory Board or 

ABRWH Advisory Board on Radiation and Worker Health 

Am americium

Au gold 

Be beryllium 

Bk berkelium 

BNWL Battelle Northwest Laboratory

C confinement factor 

Cd cadmium

Cf californium

CFR Code of Federal Regulations

Ci curie

Cm curium 

Cr chromium 

Cs cesium

CTW construction trade workers

Cu copper 

D dispersibility

DCF dose conversion factor 

D&D decontamination and decommissioning 

DOE U.S. Department of Energy 

dpm disintegrations per minute.

DR dose reconstruction 

ER petition evaluation report

Fe iron

FPPP Fission Product Pilot Plant

GA0 or GU0 gross alpha in urine sample 

GB0 gross beta in urine sample

GF0 gross alpha in fecal sample

HFIR High Flux Isotope Facility 

Hg mercury
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HP health physics

HTO tritiated water

I iodine 

Ir iridium

LANL Los Alamos National Laboratory 

LBNL Lawrence Berkeley National Laboratory

MAP mixed activation product 

MDA minimum detectable activity

MFP mixed fission product 

mCi millicurie

μCi microcurie

MMES Martin Marietta Energy Systems

Mo molybdenum 

mrem millirem

NaI(Tl) sodium iodide (thallium)

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

OCAS Office of Compensation Analysis and Support 

ORAUT Oak Ridge Associated Universities Team

ORNL Oak Ridge National Laboratory 

Os osmium

pCi/d picocurie per day 

Pd palladium

Pm promethium 

PNL Pacific Northwest Laboratory

POC probability of causation

Pt platinum 

Pu plutonium 

R release fraction

RaLa radioactive lanthanum 

Ru ruthenium 

SEC Special Exposure Cohort 
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Sr strontium 

SRDB Site Research Database

TBD technical basis document

Tc technetium 

Th thorium 

TP0 gross trans-plutonium alpha in urine sample 

TF0 gross trans-plutonium alpha in fecal sample 

U uranium 

WBC whole body counter 

yr year
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1 INTRODUCTION AND BACKGROUND

The National Institute for Occupational Safety and Health (NIOSH) evaluated the internal 

monitoring capability of Oak Ridge National Laboratory (ORNL) for radionuclides that were 

produced by the Isotopes Division (termed “exotic radionuclides”) and its predecessors from 

1955 to 1988 in ORAUT-RPRT-0090, Revision 00, Monitoring Feasibility Evaluation for Exotic 

Radionuclides Produced by the Oak Ridge National Laboratory Isotopes Division (2018; 

hereafter “RPRT-0090”). In RPRT-0090, NIOSH listed 213 radionuclides in Table 6-3, which 

was presented as the final inventory for the Isotopes Division for the period 1955–1988. 

Table 7-2 provided a detailed list of each of the 213 radionuclides and the years they were in 

inventory (representing potential exposure), along with monitoring capability and bioassay data 

availability. NIOSH found that ORNL had adequate monitoring capabilities for 179 of these 

213 radionuclides. Attachment B of RPRT-0090 provided a brief summary of the decay 

characteristics and bioassay methods for each of these 179 radionuclides. Table 7-4 of 

RPRT-0090 summarized the 34 remaining radionuclides that needed additional evaluation. Five 

of these 34 radionuclides were addressed in Attachment C of RPRT-0090 concerning 

radioiodine. Plutonium-241 was removed from the list of consideration because it was located at 

the Y-12 Plant (now Y-12 National Security Complex). In April 2018, the Advisory Board on 

Radiation and Worker Health (Advisory Board) tasked SC&A to evaluate RPRT-0090.
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2 RPRT-0090 FEASIBLITY EVALUATION: BASIS FOR REVIEW

Demonstrating “feasibility” is central to petition evaluation under the Special Exposure Cohort 

(SEC) rule and process and requires that NIOSH determine whether it is “feasible to estimate the 

level of radiation doses of individual members of the class with sufficient accuracy” 

(42 CFR Part 83). To accomplish this determination, NIOSH would need “access to reliable 

information on the identity or set of possible identities and maximum quantity of each 

radionuclide (the radioactive source material) to which members of the class were potentially 

exposed without adequate protection” (42 CFR Part 83). 

In RPRT-0090, NIOSH has identified each nuclide produced by the ORNL Isotopes Division 

within the time period in question (1955–1988) and has derived an estimated maximum annual 

inventory for each radionuclide. A “feasibility evaluation” was performed for each of the 213 

radionuclides identified in the inventory for 1955–1988, as reflected in Table 7-2, by matching 

an “adequate monitoring method” for each radionuclide for each year for which an inventory for 

it was established (and assuming that the method was available thereafter) (NIOSH 2018a). For 

the 28 radionuclides without an applicable monitoring method, NIOSH conducted a dosimetric 

analysis combining the maximum annual inventory for each radionuclide with a maximum organ 

dose conversion factor (DCF) to estimate the “committed dose to the maximally exposed organ 

from inhalation of 1 × 10-5 of the total inventory”1

 (NIOSH 2018a, p. 40). NIOSH’s analysis 

showed doses (50-year committed dose equivalent) from inhalation ranging from 0.3 millirem 

(mrem) to 1,464 mrem to the maximally exposed organ, from which it concluded that “the 

relatively low radiotoxicity of these same nuclides [the 28 without evidence of monitoring] in 

comparison with a bounding potential intake (Table 7-6) lends credence to the position that a 

significant intake of one of these nuclides would not be credible” (NIOSH 2018a, p. 43). 

1 SC&A’s evaluation of the use of the 1×10-5 factor is provided in Appendix 4 of this report.

SC&A has reviewed RPRT-0090 for the general premise and scope of its “feasibility” 

evaluation, as described above, as well as in terms of the adequacy and completeness of its 

review of monitoring capabilities and operational inventories. 

The following findings and observations need clarification or further substantiation in terms of 

actual dosimetric practice and how that translates to demonstrated feasibility to monitor the wide 

range of exotic radionuclides historically present in ORNL operations. 

 
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3 SC&A’S REVIEW OF SCOPE OF EXOTIC RADIONUCLIDE SOURCE 

TERMS

The petition evaluation report (ER) for ORNL (X-10), SEC-00189, reserved “the radionuclides 

created by cyclotrons, accelerators, and reactors for this evaluation report due to the complexity 

and interdependency involved in transferring materials between two DOE-covered facilities 

(X-10 and Y-12)” (NIOSH 2012, p. 40). It is not clear from RPRT-0090 whether its scope 

represents part or all of the reserved radionuclides that are to be evaluated. This needed 

clarification has three components, as follows. 

First, the residual and the decontamination and decommissioning (D&D) periods for specific 

facilities at ORNL involved handling the cleanup of exotic radioisotopes, as well any postoperational waste management associated with them. The extensive accelerator and reactor 

operations at ORNL would have involved a broad spectrum of mixed activation products 

(MAPs) and mixed fission products (MFPs) (according to the ER: “Many facilities have 

exhibited fission and activation product exposure potential” [NIOSH 2012, p. 46]), not all of 

which would have been part of the Isotopes Division inventory that NIOSH addressed. While 

RPRT-0090 addresses what has been identified as being handled by the Isotopes Division, this 

inventory would likely need to be expanded to accommodate operational source terms, including 

D&D and waste management, for these other activities. 

This is not an insignificant distinction. Of the 11 cases where internal exposure guides were 

exceeded at ORNL in 1964–1967, six cases involved iodine-131 (I-131) uptakes from 

decontamination work, four cases involved tritium uptakes due to target preparation and 

decontamination work, and one case involved a strontium-90 (Sr-90) uptake due to work at a 

solid waste disposal burial ground (MacPherson 1968). It can be assumed that the number of 

intakes that did not happen to exceed exposure guides would have been higher (that information 

was not readily available) and exotic radionuclides would likely have been associated with some 

of these activities.

Second, the processing of radioisotopes from beginning to end involved workplace emissions 

and waste streams, as well as byproduct materials, that may have included quantities of the 

exotic radioisotopes in question. Are these also reflected in the inventory NIOSH used in 

RPRT-0090, or is that inventory simply the “finished product” inventory of isotopes produced 

for transfer? That former amount would obviously be larger than the latter, and such a 

discrepancy would directly influence the basis for determining the “maximum annual inventory” 

in Table 7-6, from which a justification is made for 28 radionuclides not having a “bounding 

potential intake” of concern. 

Third, construction trade workers (CTWs), maintenance personnel, and other support personnel 

(such as health physics [HP] technicians) moved around the ORNL site and were potentially 

exposed to a broad range of the exotic radionuclides in question, even beyond those generated by 

the Isotopes Division. This fluidity of work assignment is captured in an interview with a 

maintenance engineer: 

Please understand that we had central maintenance shops that were responsible 

for maintenance all over ORNL. It would not be correct to assume that because a 
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worker was not assigned to a certain building that he or she did not actually do 

work in that particular building. The issue of overtime must also be considered.

Workers from all over the plant site are asked to work overtime on any particular 

job. A worker that would normally work in a clean area could in fact receive an 

exposure on an overtime job. [Bancord 2004] 

This raises three questions. First, was bioassay monitoring for CTWs, maintenance workers, and 

other personnel implemented in the same manner as operating personnel? In other words, did 

area health physicists direct nonroutine bioassays based on the same subjective exposure criteria 

and were such exposures (e.g., due to contamination) identified in the same manner as with 

operational staff? Second, do records exist to document how they were actually monitored during 

the 1955–1988 time period? Third, what radionuclides would these worker categories have been 

exposed to (given their laboratory-wide work locations), and how does that compare with 

Table 7-2 in terms of the adequacy of bioassay monitoring? 

Finding 1: Scope of RPRT-0090 needs to be clearly defined. 

SC&A finds that the scope of RPRT-0090 needs to be clarified in terms of whether (and how) it is 

meant to encompass the “reserved” portion of the ER for “cyclotrons, accelerators, and 

reactors” and whether NIOSH intends to address the full scope of radionuclides involved in 

waste management (including D&D), site-wide construction, and maintenance. 
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4 SC&A’S EVALUATION OF RADIONUCLIDE INVENTORY

SC&A compared several of the radionuclide initial dates listed in Table 7-2 of RPRT-0090 to the 

NIOSH-supplied Microsoft Excel X-10 Exotics Workbook_022015 kwv under the tab “Master 

Summary Data” (NIOSH 2015; hereafter “X-10 Inventory”). While some of the early dates 

matched, some did not. For example, Table 7-2, page 31, of RPRT-0090 shows that curium-244 

(Cm-244) was present beginning in1962; however, the NIOSH X-10 Inventory spreadsheet, 

Column AA, Row 72, lists the first inventory of Cm-244 as 1964. Similar discrepancies were 

found for berkelium-249 (Bk-249), Cm-242, plutonium-238 (Pu-238), Pu-239, and uranium-232 

(U-232) in the early years.

SC&A compared Table 7-2 of RPRT-0090 with the contents of Table 5-15 of the ORNL 

technical basis document (TBD), ORAUT-TKBS-0012-5, Revision 02, Oak Ridge National – 

Occupational Internal Dose (NOISH 2013a) and found several radionuclides listed in Table 5-15 

for years that were listed as “N” (meaning no radionuclide present in inventory in the specified 

year) in Table 7-2. For example, protactinium-233 is listed with a 1961 maximum measured in 

vivo activity of 2,800 nanocuries in ORAUT-TKBS-0012-5, Revision 02, but has a note of “N” 

in Table 7-2 of RPRT-0090. 

Observation 1: Inventory discrepancy. 

A sampling of some of the inventory of the radionuclides for the early years indicated some 

discrepancies in inventory between Table 7-2 in RPRT-0090 and NIOSH’s X-10 Inventory 

spreadsheet.

SC&A also compared the radionuclide listings in Table 7-2 of RPRT-0090 with various ORNL 

records, including a sampling of customer shipment listings (ORNL 1957, ORNL 1965, BNWL 

1977, PNL 1984) and an in-house isotope inventory (Kohring 1990).2

2 Kohring 1990 provided a listing for 1989. While it was dated shortly after the Dec 1988 cutoff, SC&A considered

it illustrative of radionuclides that may have been shipped in the late 1980s (no other inventories were found for 

fiscal year 1987–1988). 

Table 7-2 of RPR-0090 

was derived using source documents from the Site Research Database (SRDB), including isotope 

shipping and sales reports and operational and technical reports, as well as target rupture records 

(for sealed shipping containers for which leakage of radionuclides may have presented an 

exposure source). SC&A’s comparison, based on similar source documents, identified the 

following radioisotopes as generated at ORNL but not included in the NIOSH inventory in 

RPRT-0090. 

Table 1. Examples of Radionuclides Not Included in RPRT-0090 

ORNL 1957 ORNL 1965 BNWL 1977 PNL 1984 Kohring 1990

Europium-154

Iron-55, -59

Lutetium-174

Iodine-128

Europium-149

Europium-154

Thulium-168

Europium-154 Aluminum-26 Clorine-34

Manganese-57

Iodine-128

Europium-154

 
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As these were radioisotopes for offsite shipment, it is unclear whether these particular 

radionuclides were shipped directly off site for processing and were, therefore, not handled in an 

unsealed form at ORNL (and accordingly, excluded from the inventory). If they were shipped in 

a sealed container off site, the only remaining question would be whether any record of target 

rupture exists (which would lead to inclusion in Table 6-2 of RPRT-0090). Clearly, isotopic 

variants of certain radionuclides, such as euriopium-154, would likely have the same 

radiochemical properties as other isotopes of the same radionuclide; this listing is provided as a 

means of verifying completeness.

In reviewing the completeness of the Table 7-2 radioisotope listing, SC&A also reviewed the 

corresponding ORNL facilities involved with their generation. One ORNL report, Oak Ridge 

National Laboratory Isotopes Facilities Shutdown Program Management Plan (ORNL 1992, 

p. 2), noted that: 

The goal of the shutdown program [was] to place 16 formerly utilized isotopes 

facilities at ORNL...in a radiologically and industrially safe condition for routine, 

long-term maintenance and surveillance prior to eventual decommissioning. 

The additional facilities identified by SC&A in this review that handled such radioisotopes, but 

were not listed among the 10 buildings cited by RPRT-0090 (and not including subparts of those 

facilities, e.g., 3038-M and 3038-E), were: 

• Krypton-85 Enrichment Facility Building 3026-C 

• Radioisotope Production Laboratory-H Building 3118 

• Tritium Target Preparation Facility Building 7025 

Building 3026-C was the Krypton-85 Enrichment Facility, and handled a mixture of krypton 

gases of different isotope masses, with receipt of tritium-filled tubes from Building 3033 (Patton 

1988). Building 3118 was constructed in the early 1960s to enclose access to the rear entry area 

for the hot cells in Buildings 3030 and 3031 and provided a storage area for “drums and 

containers of hazardous and radiological waste, radioactive shielding materials, and casks” 

(ORNL 1997a, p. 1). Building 7025 housed a tritium target fabrication system for preparing thin 

oxide targets for domestic and international customers, as well as tritium tritide targets and 

various metallurgical samples (ORNL 1997b). 

Another ORNL facility handling exotic radioisotopes was Building 3515, the “Fission Product 

Pilot Plant” (FPPP),” one of the first facilities at ORNL to extract radioisotopes from liquid 

radioactive wastes. The pilot plant was initially constructed in 1948 and modified in the early 

1950s with additional shielding and a hot cell. FPPP operations included “extracting radioactive 

isotopes from waste liquids from off-site and ORNL activities,” and the “radionuclides removed 

included ruthenium, strontium, cesium, cerium, and others” (Mandry and Snedaker 1994, p. 2). 

The basis for defining the ORNL radioisotope complex as the “10 buildings designed for 

processing, packaging, and shipping radioisotopes” is not clear in RPRT-0090 (p. 6). While these 

10 are clearly key facilities that handled radioisotopes at ORNL, others had similar operations or 

were even co-located and associated with them. If the scope of RPRT-0090 is intended to 

evaluate “the monitoring capability of the ORNL HP program to have monitored all materials 
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that were produced and handled by the Isotopes Division regardless of production location” 

(RPRT-0090, p. 6), then other such facilities and associated radionuclides should be reviewed for 

inclusion. 

Finding 2: Incomplete radionuclide and radioisotope facility inventory. 

A sampling of the radionuclides listed in Table 7-2 found a few missing when compared with 

operational and customer records. Likewise, a few ORNL facilities that historically handled 

radioisotopes are also not included in those cited and addressed in RPRT-0090. Given the 

operational diversity of ORNL accelerator and reactor operations, consideration should be 

given to an inventory scope that encompasses isotopic source terms broader than that of the 

Isotope Division.
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5 SC&A’S EVALUATION OF ORAUT-RPRT-0090 WITH RESPECT TO 

IN VITRO METHODS

5.1 SC&A’S EVALUATION OF MONITORING CAPABILITY AND BIOASSAY DATA FOR 179

RADIONUCLIDES

SC&A selected 12 of the 179 radionuclides to evaluate monitoring capabilities and bioassay 

data. These 12 radionuclides were selected because of the relatively large quantities of the 

radionuclides in inventory for a number of years and their radiotoxicity, which could result in 

potentially significant dosimetric intakes. This mixture included both routine and exotic 

radionuclides. The 12 radionuclides selected were americium-241 (Am-241), Bk-249, 

californium-252 (Cf-252), Cm-242, Cm-244, promethium-147 (Pm-147), Pu-238, Pu-239, 

ruthenium-106 (Ru-106), Sr-90, thorium-230 (Th-230), and U-232. 

For each of the 12 radionuclides SC&A evaluated, the available bioassay data for the years the 

radionuclides were listed in the inventory. SC&A selected bioassay data (as opposed to other 

search criteria) for evaluation because if bioassay data are available, the procedures, bioassays, 

and records were available and functional for dose reconstruction (DR) purposes. For this 

evaluation, SC&A used the following sources of information: 

• Copies of original bioassay records: NIOSH Office of Compensation Analysis and

Support (OCAS) Claims Tracking System (NOCTS). 

• List of Code 000 data in NOCTS: NIOSH’s Excel spreadsheet Type0_archiveR1

(NIOSH 2018b). 

• X-10 bioassay database: NIOSH’s Microsoft Access database, niosh_04282013, 

tab “VPA ornl_invitro” (NIOSH 2013b), which contains 94,988 urine and fecal bioassay 

records for the period 1955–1988 (hereafter “X-10 Database”).

• Radionuclide inventory: SC&A considered the radionuclide inventory information in 

NIOSH’s X-10 Inventory spreadsheet to be the correct inventory list when performing 

comparisons for this evaluation. 

For the 12 radionuclides under evaluation, SC&A compared the information in Table 7-2 of 

RPRT-0090 (using the bioassay codes listed in Table 7-2 for each of the radionuclides) to the 

bioassay data in the NIOSH’s X-10 Database and NIOSH’s X-10 Inventory. Examples of 

SC&A’s summary of the evaluation for Cm-244 and Ru-106 are shown in Appendix A of this 

report. Table 2 summarizes the results for the 12 radionuclides evaluated. Table 7-2 in 

RPRT-0090 used a green cell color and “G” to indicate that a radionuclide was present in 

inventory, a bioassay method was available, and sample results were available for that year. It

used a yellow cell color and “Y” to indicate that a radionuclide was present in inventory and a 

bioassay method was available, but no sample results were available for that year.
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Table 2. SC&A’s Evaluation of 12 Radionuclides from Table 7-2 of RPRT-0090 

Radioisotope

Major 

Radiation 

Emitted

Table 7-2 Bioassay 

Codes for Specific 

Radioisotope

Years Table 7-2 

Listed Green (G)

when Bioassay Data 

Were Not Present

but Radioisotope

Was in X-10 

Inventory (a)

Years Table 7-2 

Listed Yellow

(Y) or Green (G) 

when the 

Radioisotope

Was Not in X-10 

Inventory

Contents of 

Table 7-2 

Compared to

Supporting 

Data

Am-241 Alpha 000(Am-241), AM0, 

GU0, TP0 None None Complete

Bk-249 Alpha 000(Bk-249), TP0 None 1964 Mostly 

complete

Cf-252 Alpha TP0 None 1962 Mostly 

complete

Cm-242 Alpha 000(Cm-242), TP0 None 1961, 1962 & 

1964

Fairly 

complete

Cm-244 Alpha 000(Cm-244), TP0, 

CM0 None 1962 & 1963 Fairly complete

Pm-147 Beta 000(Pm-147), 013, 

PM7, FU0, Pm-147

(a)(1956 & 1957) None Fairly 

complete

Pu-238 Alpha GU0, PU0 None 1961 & 1962 Fairly 

complete

Pu-239 Alpha GU0, PU9, PU0 None 1961 & 1962 Fairly 

complete

Ru-106 Beta 000(Ru-106), 013, 

GB0, RU6

(1956, 1957, 1959)(a),

1975, 1978, 1986, 

1987, & 1988

None Incomplete

Sr-90 Beta SR0 None None Complete

Th-230 Alpha GF0, TF0, Th-230 None 1962 & 1963 Fairly 

complete

U-232 Alpha UR0 None None Complete

(a) Lack of beta records in the X-10 Database for 1955–1959 may be because they were not entered into the X-10 

Database; however, they could be contained in the NOCTS files, as discussed on page 15 of RPRT-0090.

5.1.1 Gross Alpha Counting

SC&A found that, in general, the pre-1965 records for the most prevalent alpha-emitting 

radionuclides (such as plutonium, thorium, and uranium) indicated the specific element assayed 

(e.g., plutonium). For the trans-plutonium elements, such as Am-241, Cf-252, etc., gross alpha in 

urine or fecal samples were used (bioassay codes GA0, GU0, TP0, GF0, or TF0). In the 1950s, 

some specific radionuclides were listed on the bioassay card; however, it became more prevalent 

in approximately 1965 for the specific radionuclide analyzed (such as Am-241) to be specified in 

the bioassay records. Most of the ORNL original bioassay cards available in NOCTS provide 

detailed analytical information, including chemical process yield, counter efficiency, etc., with 

specific radionuclides listed as far back as 1955. Some of these are illustrated in the examples in 

Appendix B of this report. SC&A found that there were a large number of bioassay records for 

each year for the period 1955–1988 containing alpha counting results (coded GA0, GU0, TP0, 

GF0, or TF0) in the X-10 Database from the 94,988 urine and fecal bioassay records. An 

example from this database for Am-241 for a small part of 1955 and 1956 is shown in 

Appendix C of this report. 
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As can be seen from Table 2, the major alpha emitters of the 12 radionuclides evaluated appear 

to have been monitored as was listed in Table 7-2 of RPRT-0090, with the exception that in 

several instances Table 7-2 indicates that inventory was present when the inventory list does not 

indicate that the radionuclides were present during that year. Additionally, it has to be assumed 

that the dose reconstructor has sufficient information in the claimant’s files to assign the correct 

radionuclide the worker was exposed to from gross alpha counts, such as gross alpha in urine 

(GA0, GU0, and TP0 bioassay codes), or gross alpha in fecal samples (GF0 and TF0 bioassay 

codes). The X-10 Database records do not always list the specific radionuclide analyzed, as the 

original ORNL bioassay cards often do. 

Observation 2: Specific alpha-emitting radionuclide needs to Be identified for DR.

The specific radioisotope monitored is not always presented in NIOSH’s X-10 Database as it 

generally is in the NOCTS files. Gross alpha results could be applied to many radionuclides. Is 

the information on the original bioassay cards available in the X-10 Database, and will the X-10 

Database be used in DR or coworker model development? 

Trans-plutonium Radionuclides

According to pages 19 and 20 of ORAUT-TKBS-0012-5, Revision 02, bioassay results for 

plutonium are to be assigned as Pu-239 intakes, and bioassay results for trans-plutonium 

radionuclides (using such methods as gross alpha counting) are to be assigned as Am-241 intakes 

for DR purposes. In general, this is an acceptable DR method; however, with the increased 

potential for concentrating exotic radionuclides at the ORNL Isotopes Division, which may 

contain such radionuclides as Bk-249, Cf-252, etc., this method may, or may not, be applicable.

Observation 3: Trans-plutonium radionuclides may need further analyses.

SC&A is concerned that assigning trans-plutonium gross alpha counting results as Am-241 

intakes without consideration of other potential trivalent alpha-emitting actinides (such as 

Bk-249, Cf-252, Cm-242, Cm-244, etc.) and their individual radiotoxicity could result in 

underestimating the internal dose. It could be beneficial to determine if assigning the intake as 

Am-241 is claimant favorable, considering the exotic trans-plutonium radionuclides at ORNL.

5.1.2 Gross Beta and Gamma Counting

Gross beta counting was sometimes used for strontium, but the bioassay code SR0 or SR9 was 

generally used to specify Sr-90 because of the prevalence of Sr-90, as far back as the 1950s. In 

addition to beta counting specifically for strontium, Table 4-1 and Table 4-2 of RPRT-0090

indicate that ORNL used gross beta counting (153 samples with bioassay code 013 and 26 with

bioassay code GBO), gross gamma counting (9 with bioassay code GGO), fission products (17 

with bioassay code FP0), and rare earths (1,333 with bioassay code FU0 and 70 with RF0). 

While these are small numbers of bioassays compared to the total of 91,867 coded bioassays in 

Table 4-1, some of these gross counting codes are listed as bioassay codes for many

radionuclides in Table 7-2 of RPRT-0090. These gross counting methods are briefly discussed in 

Attachment A of RPRT-0090, but no specific counting efficiencies, correction factors, etc., are

provided. Several SRBD references provide detailed procedures for the radiochemistry and gross 

beta counting of bioassay samples, but they do not include any specific radionuclide-efficiency 
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calibration information (e.g., Henley 1968, PDF page 39, and Henley 1978, PDF page 23). 

Henley 1968 states on PDF page 38: 

ANALYSIS FOR BETA EMITTERS

The object of a mixed beta count is to estimate the beta activity in the urine 

exclusive of the naturally occurring potassium-40. This is largely accomplished 

by using two group separations and counting the product of each. 

Group #1 includes the ions precipitated by or absorbed in a calcium phosphate 

precipitate. These include the alkaline earths, the lanthanides, the actinides and 

most other ions with strong tendencies for hydrolysis.

Group #2 includes the elements which can be precipitated from a dilute acid 

solution as the sulfide. Cu, Mo, Tc, Ru, Pd, Cd, Os, Ir, Pt, Au and Hg are 

precipitated directly from urine in this manner. Alkali metals and anion activities 

are not carried by this precipitation. 

As can be seen from this description, there could be a variety of combinations of radionuclides, 

or a single radionuclide, contributing to the gross beta counts. Attachment B of RPRT-0090 

provides a summary of the decay emissions from the radionuclides at ORNL, which indicate a 

wide range of beta energies and percent of emission (abundance) for the many radionuclides that 

could be analyzed by beta counting. This can present problems when assigning gross beta count 

data to a specific radionuclide. The use of data obtained from gross beta counting of bioassay 

samples to project specific radionuclide intakes (from an assortment of potential radionuclides) 

presents potential issues as outlined in SCA’s recent evaluation of a NIOSH white paper on 

Lawrence Berkeley National Laboratory (SC&A 2018). In that report, SC&A also analyzed 

issues with applying gross alpha counting data, such as from air samples, to project biological

intakes. However, the issues for alpha emitters are not as prevalent for ORNL bioassays because 

of the use of defined urine and fecal counting and the fact the most dosimetric-important alpha 

emitters listed in Attachment B of RPRT-0090 are similar in energy and abundance. This is not 

true for beta emitters listed in Attachment B of RPRT-0090, which are not all similar in energy 

and abundance. For example, Attachment B of RPRT-0090, page 68, shows that copper-67 (Cu67) beta energy and abundance is quite different than those of Ru-97 on page 94; however, both 

could be present in the Group #2 radionuclides as stated in the Henley 1968 quote above. 

Additionally, among radionuclides of the same element, such as silver-110m and silver-111 as 

shown on pages 96 and 97 of RPRT-0090, the beta energies and abundance can be very different. 

A similar analogy applies to gross gamma count data. 

SC&A’s evaluation of the three beta-emitting radioisotopes Pm-147, Ru-106, and Sr-90 is 

summarized in Table 2 of this report. Table 2 indicates the following: 

• Pm-147 – The data in Table 7-2 of RPRT-0090 for Pm-147 are in good agreement with 

the supporting data in the X-10 Database and the X-10 Inventory, considering that the 

1956 and 1957 beta results may not have been recorded in the X-10 Database. 

• Ru-106 – The data in Table 7-2 of RPRT-0090 for Ru-106 are not in good agreement 

with the supporting data in the X-10 Database and the X-10 Inventory, even when 
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considering that the 1956, 1957, and 1959 beta results may not have been recorded in the 

X-10 Database. SC&A could not locate any appropriate beta counting data for the years 

1975, 1978, 1986, 1987, and 1988, although Table 7-2 of RPRT-0090 has those years 

colored in green indicating that bioassay data were available for those years.

• Sr-90 – The data in Table 7-2 of RPRT-0090 for Sr-90 are in good agreement with the 

supporting data in the X-10 Database and the X-10 Inventory. 

Observation 4: Use of gross beta or gamma count data could result in underestimate of

assigned dose. 

Using gross beta or gamma count data without knowledge of the radionuclide the counter was 

calibrated with and the radionuclides in the bioassay sample could result in assigning the 

incorrect radionuclide and radioactivity content because of different counting efficiencies for the 

different energy of beta particles and gamma photons. Has this issue been addressed for DR for 

ORNL claimants? Additionally, bioassay data for at least one beta-emitting radionuclide 

(Ru-106) could not be located for several years that Table 7-2 indicated it was available.

5.1.3 SC&A’s Evaluation of RPRT-0090 Attachment A 

The Attachment A references upon which the in vitro radionuclide monitoring capability cited in 

Table 7-2 are based come down to essentially two key documents: Henley 1968 and Henley 

1978. Both of these comprise radiochemical “recipes” for processing gross alpha, gross beta, rare 

earths, etc. Whether and how they (and the other methods cited) were actually used onsite based 

on field procedures, available dose records, and interviews, is not discussed in RPRT-0090. 

SC&A believes that how the identification and quantification of “associated” radionuclides was 

performed by ORNL, in practice, should be validated. The isotope-specific monitoring capability 

analysis in RPRT-0090 carries with it certain inherent assumptions not addressed in the report, 

including necessary identification of associated radionuclides through process knowledge and 

quantification by radiochemical assays. 

The TBD for ORNL occupational internal dose, ORAUT-TKBS-0012-5, Revision 02 (NIOSH

2013a), observes the following (p. 10): 

Before that time [1989], chemical methods were used to separate radioelements 

as well as practicable, and the materials were assayed in terms of total activity. 

The measured activity would later be assigned to a predominant nuclide. 

Therefore, a result from the early years might indicate 90Sr, when in reality it 

includes 89Sr. The same is true for early plutonium results and results for 

transuranic materials. Therefore, “associated” radionuclides are inherently 

included in such results. Process knowledge of radionuclides present in various 

work areas was used to assign nuclides to sample results. 

What is clear here is that the detailed matrix covering 213 specific radionuclides is dependent on 

a much smaller set of gross alpha and beta/gamma analytic procedures, and an uncertain set (see 

later comments) of in vivo monitoring protocols, all of which are, in turn, dependent on the 

extent to which area health physicists were able to identify the scope of radionuclides involved 

based on process knowledge, and whether and how radiochemical assays were performed to 
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“separate radioelements” and to ascertain how much intake activity to assign a “predominant” 

nuclide. 

In reviewing RPRT-0090, as well as the ER and internal dose TBD, SC&A has not found any 

validation of how ORNL actually implemented the above assays and process reviews such that 

assurance can be given that dose-contributing nuclides were identified and nuclide-specific dose 

contributions were adequately estimated within the constraints of the technological capability of 

the time. If “associated” radionuclides were not adequately addressed in bioassay sample results, 

the credit taken for that capability in Table 7-2, and the results themselves, may not represent 

internal dose in an adequate and sufficiently accurate manner.

In broader terms, a “bridge” between ORNL written procedures, monitoring requirements, and 

capabilities for radionuclide identification, and the recorded bioassay data available for DR, is 

not clear at this time. As SC&A has found in the past, program directives and monitoring 

capabilities do not necessarily equate to what was actually put into practice and how complete 

the data are for DR. For example, a 1988 safety appraisal of ORNL’s High Flux Isotope Facility 

(HFIR) found that “implementing or operating procedures have not been developed for internal 

dosimetry. These procedures should be developed to address the day-to-day activities of the 

internal dosimetry function” (MMES 1988, PDF p. 41). In terms of carrying out internal 

dosimetry procedures, this same 1988 safety appraisal found that “frequency schedules have 

been established and are documented in Appendix 12 of the Health Physics Manual; however, 

the established frequencies for whole-body counting of HFIR personnel are not followed” 

(MMES 1988, PDF p. 39). 

It is also not clear whether some of the 28 radionuclides for which matching monitoring 

capability has not been verified would have required whole-body counting coverage for adequate 

detection and dose assessment. If so, technological and programmatic limitations (including 

higher minimum detectable activity [MDA] benchmarks) in the earlier years may have precluded 

such coverage, bringing into question ORNL’s “ability to develop specialized bioassay methods 

as needed” (RPRT-0090, page 43). Actual radionuclide-specific in vivo monitoring capabilities 

need to be verified to establish feasibility.

If, as indicated in the ORNL site profile (NIOSH 2013a), there were no routine radionuclidespecific in vitro or in vivo bioassays before 1989, how does NIOSH intend to bound routine 

doses when (1) a significant portion of the ORNL workforce was mobile, (2) radiological source 

terms varied widely from operation to operation (and included exotics, MAPs, and MFPs), and 

(3) various area HP technicians made individual judgments on who would be bioassayed for 

cause? Under such circumstances, the “availability” of monitoring technology or procedure 

would almost be a secondary consideration for reviewing feasibility.

Finding 3: Attachment A in vitro bioassay methods lack information about actual 

implementation. 

In vitro bioassay methods are outlined in Attachment A, but it does not include any discussion or 

references regarding their actual field implementation. The exclusion of comparable in vivo

monitoring methods makes a review of ORNL monitoring capability incomplete. 
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5.2 SC&A’S EVALUATION OF FEASIBILITY OF DR FOR 28 RADIONUCLIDES

Section 7.2 of RPRT-0090 discussed the 28 radionuclides that were identified as needing 

additional consideration because of the lack of defined bioassay methods, and some of their 

characteristics are listed in Tables 7-4, 7-5, and 7-6 of RPRT-0090. To evaluate the significance 

of the 28 radionuclides listed in Table 7-6, SC&A ran the Integrated Modules for Bioassay 

Analysis program using a factor of 1E-5 times the maximum annual inventory as listed in 

column 2 of the table for several of the radionuclides and found that the resulting organ doses 

were similar to the doses listed in column 5 of the table, for solubility Type M; other solubility 

types could produce different doses. SC&A’s detailed analyses of using a factor of 1E-5 is in 

Appendix D of this report.

SC&A then compared the maximum annual inventory value for each radionuclide, as listed in 

column 2 of Table 7-6, to the inventory value in NIOSH’s X-10 Inventory for each year there 

was a red symbol in Table 7-2 for that radionuclide. SC&A found a few discrepancies in the 

annual inventory values, but, in general, there was reasonable agreement. However, the dose 

values still indicate that some of radionuclides may be dosimetrically significant. Table 3 

provides a summary of Table 7-6 of RPRT-0090 to include the SC&A radionuclide inventory 

analyses. 

Table 3. Dosimetric Analysis of 28 Radionuclides

Radionuclide 

RPRT0090

Maximum 

Annual 

Inventory 

(mCi) 

RPRT0090 Year 

of 

Maximum 

Inventory 

RPRT0090

Maximum 

Organ 

DCF 

(mrem/

mCi)

RPRT-0090

Organ of 

Maximum 

Dose 

RPRT0090

Committed

Dose 

Equivalent 

(mrem)

X-10 

Inventory 

Maximum 

Annual 

Inventory 

(mCi) 

X-10 

Inventory 

Year of 

Maximum 

Inventory 

Revised 

Committed 

Dose 

Equivalent 

(mrem)

Beryllium-7 340 1957 1,554 Extrathoracic 5 340 1957 5 

Calcium-41 501 1986 9,250 Bone 46 501 1986 46

Chromium-51 46,225 1961 925 Extrathoracic 428 46,225 1961 428

Manganese-54 115 1957 27,010 Extrathoracic 31 115 1957 31

Iron-55 620 1955 27,750 Spleen 172 620 1955 172

Cobalt-57 175 1957 13,690 Lungs 24 175 1957 24

Gallium-67 120 1959 6,290 Extrathoracic 8 120 1959 8 

Selenium-75 2,160 1955 27,750 Kidneys 599 2,160 1955 599

Strontium-85 142 VNL (a) 20,720 Extrathoracic 29 105 1957 21

Strontium-87m 84 1960 2,220 Extrathoracic 2 84 1960 2 

Molybdenum-93 (d) NA 103,600 Bone (d) 39 1962 40

Ruthenium-103 1,020 1955 55,500 Lungs 566 1,020 1955 566

Palladium-103 (d) NA 6,290 Extrathoracic (d) 47.5 1962 3 

Cadmium-109 64 1964 851,000 Kidneys 545 64 1964 545

Tin-113 610 1957 48,100 Lungs 293 610 1957 293

Tin-119m 3,598 1986 40,700 Lungs 1464 3,598 1986 1464

Tellurium-121 13 No data (b) 19,980 Extrathoracic 3 NA No data NA 

Cesium-131 25 No data 2,775 Extrathoracic 1 NA No data NA 
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Radionuclide 

RPRT0090

Maximum 

Annual 

Inventory 

(mCi) 

RPRT0090 Year 

of 

Maximum 

Inventory 

RPRT0090

Maximum 

Organ 

DCF 

(mrem/

mCi)

RPRT-0090

Organ of 

Maximum 

Dose 

RPRT0090

Committed

Dose 

Equivalent 

(mrem)

X-10 

Inventory 

Maximum 

Annual 

Inventory 

(mCi) 

X-10 

Inventory 

Year of 

Maximum 

Inventory 

Revised 

Committed 

Dose 

Equivalent 

(mrem)

Barium-133 80 1961 35,520 Bone 28 80 1961 28

Cerium-139 (d) NA (c) 37,000 Lungs (d) (d) NA (d)

Promethium-145 32 Not listed 151,700 Bone 49 NA NA NA 

Terbium-156 (d) NA 51,800 Extrathoracic (d) (d) NA (d)

Dysprosium-159 60 1961 9,620 Bone 6 60 1961 6 

Tungsten-181 18 No data 1,628 Extrathoracic 0.3 NA No data NA 

Osmium-185 (d) NA 26,270 Extrathoracic (d) (d) NA (d)

Gold-195 5.6 1961 30,710 Lungs 2 5.6 1961 2 

Mercury-197 542 1957 133,200 Lungs 722 542 1957 722

Bismuth-206 (d) NA 88,800 Extrathoracic (d) (d) NA (d)

(a) VNL denotes that this value was not listed in the X-10 Inventory. (b) No data for this radionuclide were listed in the X-10 Inventory. (c) NA denotes not applicable to this radionuclide. (d) No inventory data were available for this radionuclide, per RPRT-0090. 

Observation 5: The results in Table 7-6 depend on inventory used. 

As outlined in Observation 1, there appear to be some discrepancies in the inventory used by 

NIOSH compared to those provided to SC&A for evaluation of RPRT-0090. These discrepancies 

change a few of the results of Table 7-6, as illustrated in Table 3 of this report. 

The feasibility of monitoring for the 28 radionuclide intakes for DR purposes was not 

definitively addressed in RPRT-0090 Section 7.2, or in the Summary in Section 8.0, which states 

(p. 43): 

Evaluation of the ORNL bioassay program indicates the ability to develop 

specialized bioassay methods as needed to adapt to changing conditions and 

emergent events. While evidence has not been found for all of the 28 identified 

nuclides, it is clear that ORNL had the capability to develop methods as needed. 

In addition, the relatively low radiotoxicity of these same nuclides in comparison 

with a bounding potential intake (Table 7-6) lends credence to the position that a 

significant intake of one of these nuclides would not be credible.

The presence of the capability to develop monitoring does not necessarily mean that it was 

implemented, especially since bioassay records do not exist to show that it was used. The 

dosimetric significance of the 28 radionuclide depends on the inventory quantity and the 

inhalation factor used, as illustrated in Table 3 of this report. 

5.2.1 Implications of Lack of Routine Bioassay Program 

Moreover, SC&A questions whether the lack of routine bioassays until the late 1980s may take 

precedence over “monitoring capability” as addressed in RPRT-0090. ORNL did not routinely 

bioassay its workers for day-to-day potential intakes “in the early years,” and isotope-specific 
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analyses for in vitro samples did not become routine until 1989. As noted in the TBD for 

occupational internal dose, ORAUT-TKBS-0012-5, Revision 02 (NIOSH 2013a, p. 12): 

Urine samples were collected in the early years of the bioassay program based on 

the area health physicist’s knowledge of field conditions (e.g., known spills or 

incidents, air and contamination sample results, etc.). This practice of scheduling 

did not use a specified sampling frequency.... A 1961 procedure 

manual...references procedures and practices governing the health physics 

program at that time, including internal and external exposure monitoring. 

Although referenced, the procedure detailing internal dosimetry was not among 

the documents available for review.

Regarding whole body counting frequencies, this same reference noted that: 

Historical information of the in vivo monitoring program...indicates that the 

whole-body counting frequencies in Table 5-3 were not consistently followed. 

Discussions with previous site personnel indicate that no formal counting 

frequency was used at ORNL until the later 1980s. 

In other words, no routine whole-body counting was performed at ORNL until after the time 

period in question (1955–1988). 

Although, an argument can be made for the availability of bioassay monitoring capability, as 

well as the ability to use nonroutine bioassay data to bound routine exposures, several concerns 

need to be addressed. First is the inventory completeness question previously raised; i.e., did the 

nonroutine bioassays performed encompass the full scope of radionuclides (including associated 

“exotic” radionuclides) for which an exposure potential existed? Second, did the area health 

physicists making judgments about which workers received bioassays treat CTWs and 

maintenance personnel the same as operations personnel for internal monitoring purposes? Third, 

how representative were the nonroutine bioassays in terms of potential exposures, as compared 

with other exposures experienced by the full range of workers at ORNL (i.e., including CTWs 

and maintenance personnel)? If intakes could have been missed at ORNL with this approach, as 

indicated by the then-lead ORNL internal dosimetrist (Berger 2004), what are the implications 

from an adequacy and completeness standpoint? 

5.2.2 Implications of Significant vs. Non-negligible Intakes

The only apparent evidence for implementation of analytic procedures are actual bioassay results 

and dose records. For NIOSH to claim that “ORNL had the capability to develop methods as 

needed” (e.g., for the 28 identified nuclides lacking analytic protocols) is questionable in light of 

actual whole-body counting system limitations that existed in the 1960s–1980s. The qualifying 

comment in RPRT-0090, that “the relatively low radiotoxicity of these same nuclides in 

comparison with a bounding potential intake (Table 7-6) lends credence to the position that a 

significant intake of one of these nuclides would not be credible,” appears to acknowledge the 

lack of information on such monitoring. 
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From this concluding statement of RPRT-0090, a key question is whether the lack of a 

“significant intake” is an adequate basis for not considering a potential internal exposure in DR, 

as opposed to a determination that the potential for such an uptake could be considered 

“negligible.” The question of whether a source term is negligible from an exposure potential 

standpoint, and the corresponding implications for DR under the Energy Employees 

Occupational Illness Compensation Program Act have been previously addressed by the 

Advisory Board; that precedent should provide guidance here. 

Finding 4: Feasibility of monitoring 28 radionuclides not adequately addressed. 

While the 28 radionuclides were discussed in Section 7.2 and some of their characteristics were

listed in Tables 7-4, 7-5, and 7-6 of RPRT-0090, the feasibility of monitoring for intakes for DR 

purposes was not completely addressed, particularly given the lack of routine bioassays in the 

earlier years. Methods for accounting for the lack of monitoring of these radionuclides need to 

be addressed in more detail, and an acceptable resolution derived. SC&A finds that it is not 

possible at this time to validate implementation without further onsite review, including 

document review and interviews with health physicists of the time period involved. 

5.3 SC&A’S EVALUATION OF FEASIBILITY OF DR FOR FIVE IODINE RADIONUCLIDES

Five of the 34 radionuclides that needed additional evaluation were radioiodine, as listed in 

Table 7-4 of RPRT-0090. NIOSH provided further evaluation of these five radionuclides in 

Attachment C of RPRT-0090. In that attachment, NIOSH postulates that radioiodine monitoring 

data from the period 1944–1954 can be used to formulate a coworker model to cover 

unmonitored radioiodine exposures during the period from 1955 to 1962, at which point it is 

presumed in vivo monitoring can be used to reconstruct doses. Specifically, NIOSH’s 

methodology utilized data from 1947–1949 to develop a chronic intake model and states the 

following conclusions to justify it as bounding (NIOSH 2018a, p. 119): 

1. The projected urinary excretion (1.7 × 105 pCi/d) is more than an order 

of magnitude greater than the highest measured urinary excretion for 

routine sampling of 4.5 × 103 pCi/d... 

2. The projected whole-body accumulation (1.2 μCi) is a factor of 4 larger 

than the highest measured whole-body accumulation of 0.28 μCi.... 

3. The projected air concentration (1.8[3] μCi/cm3

) is nearly a factor of 2 

greater than the maximum operating level used to control facility air 

concentrations.

3 SC&A assumes that this is a typo and should actually read 1.8×10-8 μCi/cm3 rather than 1.8 μCi/cm3

. 

SC&A has the following concerns with NIOSH’s conclusions: 

• Of the 168 bioassay samples evaluated in RPRT-0090, only 8 were taken prior to 1963 

and only 2 were taken prior to the first use of the whole body counter (WBC) in 1961. 

• Although RPRT-0090 notes that the projected urinary excretion rate is more than an 

order of magnitude higher than the maximum observed routine sample, no information or 

 
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references are provided to indicate when that routine sample was taken. The analysis in 

Section C.7 of RPRT-0090 indicates that the evaluated urinalysis results spanned all the 

way to 1988. 

• Per Table C-8 of RPRT-0090, the highest observed radioiodine urinalysis sample was 

2.2×107 picocuries per day (pCi/d), which is a factor of 130 higher than the projected 

urinary excretion rate using the chronic coworker model. NIOSH indicates this sample 

was categorized as “incident/follow-up/resample” but does not elaborate on the 

timeframe or conditions. 

• Conclusion 2 notes that the projected whole-body accumulation is a factor of 4 larger 

than the highest whole-body accumulation recorded (0.28 microcuries [μCi]). However, 

this whole-body measurement was made in 1962, and no whole-body measurements were 

made until 1961. It has not been established that these data can be back-extrapolated to 

represent prior exposure conditions. 

• Conclusion 3 notes that the projected chronic air concentration (1.8×10-8 μCi/cm3) was 

nearly a factor of 2 higher than the maximum operating level used to control facility air 

concentrations. However, the air sampling data are only available in summary form, and 

neither the quantitative results nor the locations of these air samples are currently known.

• The ORNL site profile (NIOSH 2007, p. 34) notes that the tolerance-level air 

concentration during 1954 (the year just prior to the unmonitored period of interest) was 

actually 3×10-8 μCi/cm3

, 50% higher than the projected air concentration calculated in 

RPRT-0090 (1.8×10-8 μCi/cm3

).

In addition to the above commentary, the fundamental question is whether the monitoring data 

and exposure potential occurring during the period 1947–1949 is representative of the period 

when exposures to radioiodine were not adequately monitored. To establish that the exposure 

potential during the proposed coworker development years is representative and/or bounding, 

RPRT-0090 notes that the relative quantity of radioiodine processed was much greater than 

during the unmonitored period. Specifically, RPRT-0090 states (p. 119): 

The quantity of iodine in process from 1955 to 1961 (1,000 to 3,600 Ci/yr) is 

bounded by the amount in process from 1947 to 1949 (8,800 to 42,000 Ci/yr). 

However, this comparison is between the commercially produced radioiodine (1955–1962) and 

the estimated annual radioactive lanthanum (RaLa) iodine releases for the period from 1944 to 

1956 as provided in ATSDR 2008. This is an “apples to oranges” comparison for at least 1955 

and 1956, when RaLa production at X-10 was still occurring. Unfortunately, the radioiodine 

releases from RaLa production are not necessarily provided on a year-by-year basis in the 

underlying documentation (ATSDR 2008). However, ATSDR 2008 does indicate that the 

highest radioiodine releases occurred during 1956 as a result of the processing of Hanford slugs. 

Figure 1 shows an excerpt from ATDSR 2008 that includes estimates of the highest annual 

releases by operation.
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Figure 1. Excerpt from ATSDR 2008 Showing that the Largest Radioiodine Releases 

Occurred in 1956

Finding 5: 1955 and 1956 intakes may not be bound by earlier coworker data. 

Assessment of RaLa radioiodine releases at X-10 indicates the highest annual releases occurred 

during the campaign to process Hanford slugs during 1956. Therefore, the radioiodine 

production and releases during the years used for coworker development (1947–1949) do not 

appear to bound the production throughput, at least during 1956 and possibly 1955. 

When strictly comparing commercially produced radioiodine by year, the proposed coworker 

years (1947–1949) do not bound the unmonitored period (1955–1962), as shown in Figure 2. 

While commercial radioiodine production was far lower in quantity than the radioiodine 

produced via the RaLa program, the relative exposure potential between the two operations has 

not been discussed. In general, the radioiodine produced during the RaLa campaigns was 

considered byproduct waste material and vented to the atmosphere via either a 200-foot central 

pilot plant stack or the local 30-foot stack. Conversely, commercial production of radioiodine 

would logically involve more direct handling of the material; thus, exposure potential might be 

larger even though the actual production amounts are lower than the RaLa operation. The 

relative exposure potential between the two operations should be evaluated and discussed. 

Observation 6: Additional RaLa production information should be provided. 

NIOSH should provide an evaluation and discussion of any potential differences in exposure 

potential between commercial radioiodine production and the radioiodine produced via the 

RaLa operation to justify the extrapolation of exposures occurring during the years 1947–1949 

to the unmonitored period (1955–1962). 
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Figure 2. Commercial Radioiodine Production by Year
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6 SC&A’S EVALUATION OF ORAUT-RPRT-0090 WITH RESPECT TO 

IN VIVO BIOASSAY METHODS

SC&A finds that the adequacy and completeness of in vivo monitoring is unclear in RPRT-0090. 

Notwithstanding the in vitro methods highlighted in Attachment A, the only treatment of 

available in vivo technology and methods is found in Section 5.0 of RPRT-0090,4 which 

provides a brief history of whole-body counting at ORNL accompanied by Table 5-1, containing 

a tabulation of in vivo counts by year (1961–1988) from the ORNL database versus the number 

of in vivo counts from annual reports. It is not clear from this treatment to what extent wholebody counting was relied upon for specific radionuclide monitoring capability, particularly 

MAPs and MFPs, both of which figured prominently at ORNL. It is also not clear for what 

radionuclides in vivo monitoring would have been the only bioassay capability for adequate 

estimation of internal intakes and dose. 

4 A more detailed description of the history and capabilities of the ORNL in vivo monitoring program is provided in 

the internal dose TBD, ORAUT-TKBS-0012-5, Revision 02 (NIOSH 2013a).

ORNL operated several major accelerators along with several smaller accelerators, such as the 

Van de Graaff accelerator. These included the Oak Ridge Electron Linear Accelerator, the Oak 

Ridge Isochronous Cyclotron, and the Holifield Heavy Ion Facility. These accelerators produced 

or analyzed radionuclides of all types and would have generated localized emissions of both 

short-lived and longer-lived MAPs.

In terms of reactor operations, ORNL has had a long history of operating both research and test 

reactors, including the Graphite Reactor, Low-Intensity Test Reactor, Bulk Shield Reactor/Pool 

Critical Assembly, Oak Ridge Research Reactor, Tower Shielding Facility, Health Physics 

Research Reactor, and the High Flux Isotope Reactor. While NIOSH finds it has insufficient in 

vitro bioassay data for MFPs prior to 1950, this changed with the availability of cesium and 

strontium bioassay data starting in 1950. NIOSH’s qualified conclusion about DR after that date 

is that it “may be feasible” (Taulbee 2012), which seems to suggest that more evaluation of 

available data and monitoring capabilities remains to be done. This raises two questions: (1) how 

were in vitro and in vivo monitoring capabilities together historically applied to MAPs and MFPs 

at ORNL? and (2) are internal dose data adequate between the availability of both capabilities 

(particularly in light of the aforementioned limitations in whole-body counting)? 

For some of the longer-lived activation source terms (e.g., beryllium-7 [Be-7], chromium-51 

[Cr-51], iron-55 [Fe-55] by way of monitoring for Fe-59) for which whole-body counting would 

have been a primary means of monitoring, whole-body counting was not routinely used until 

1989, and a full in vivo program was not in place until as late as 1994 (Berger 2004). Before that 

time, in vivo counting was prescribed by individual area health physicists, who ordered it when 

an intake was suspected, leading to what was described as “inconsistent” selection of workers 

and the likelihood that intakes were missed for this reason (Berger 2004). 

As with other national laboratories, the capability to detect and monitor for MAPs and MFPs was 

a function of technological advancement of WBCs and to what extent monitoring extended 

beyond the “core” radionuclides of concern (e.g., plutonium, uranium, tritium, cesium-137 

[Cs-137], Sr-90). The timeframe for when such in vivo capabilities were realized for what 

 
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radionuclides varied from site to site, but that would need to be established in order to ascertain 

whether those radionuclides requiring WBC monitoring were, in fact, being monitored 

adequately. This is not clear from RPRT-0090. 

Moreover, SC&A finds that monitoring feasibility is not adequately defined by technological 

capability or procedures alone, which has been borne out in past reviews of in vivo monitoring at 

DOE sites. RPRT-0090 provides references for radiochemical analysis protocols available at 

ORNL to process actinides, rare earths, tritium, etc., from in vitro gross alpha and gross beta 

monitoring, but does not provide details from the in vivo (whole-body counting) program to 

support capabilities to monitor for the broad scope of MAPs and MFPs at ORNL. 

For example, evolving WBC detector sensitivity over the time period in question (1955–1988) 

may determine the feasibility and MDA for applying in vivo monitoring for a number of the 

nuclides in question. As noted in RPRT-0090, the WBC system went through a series of 

upgrades since installation in 1959, beginning with a 4×4-inch sodium iodide (thallium) 

[NaI(Tl)] crystal, moving to a 512-channel analyzer and two 8×4-inch NaI(Tl) detectors in 1961. 

A phoswich detector was installed sometime between 1971 and 1974 (Auxier et. al. 1975) 

followed by an 80-centimeter squared HPGe counting array in 1980 (Berger and Lane 1981). It 

is not clear from RPRT-0090 what detection sensitivities were being achieved over time and for 

what radionuclides the WBC would be necessary for adequate monitoring (albeit, for nonroutine 

purposes, as routine monitoring did not begin until 1989, according to 2004).

A WBC “library” distinguishing specific energy spectrums for nuclides of interest is a critical 

tool for the internal dosimetrist to ensure that the relevant radioisotopes can be identified, a yield 

for the particular energies can be calculated, and a deposition estimated. Calibration of the WBC 

for specific target spectrums is also necessary, typically through use of a phantom or similar 

standard. If the former is incomplete or the latter deficient, monitoring capability would be 

hampered and intakes may be missed. 

The 1995 WBC nuclide library for ORNL (ORNL 1995) includes 9 of the 28 radionuclides listed 

in Table 7-6 (Cr-51, manganese-54, cobalt-57, Ru-103, cadmium-109 (Cd-109), tin-113, Cs-131, 

barium-133, cerium-139) as not having corresponding proof of monitoring capability, and three

of the radioiodines addressed in Attachment C (I-125, I-131, I-133).5

5 It is clear that for some of these radionuclides, e.g., the radioiodines, an in vitro method was also available as noted 

in RPRT-0090. However, RPRT-0090 does not address the WBC monitoring capability at ORNL in a manner that 

enables a complete assessment of monitoring feasibility across the broad spectrum of radionuclides for which 

exposure potential existed. 

 These WBC library 

citations are from the ORNL 1995 internal dosimetry technical basis document. Given the lack 

of formality for the ORNL in vivo program before 1989 (as noted in ORAUT-TKBS-0012-5, 

Revision 02), has NIOSH verified what WBC monitoring was actually conducted, in practice, by 

establishing either such radionuclide-specific library listings or in vivo monitoring procedures for 

these earlier time periods? If operational presence for and exposure potential to these 

radionuclides can be demonstrated in the 1955–1988 timeframe, how can ORNL capabilities to 

develop feasible monitoring be demonstrated in the absence of WBC libraries, calibrations, and 

 
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procedures for them? Are there any records of actual monitoring and worker exposure for 

radionuclides that would have been monitored via the WBC (e.g., NOCTS, bioassay records)? 

For five other radionuclides subject to WBC monitoring at ORNL in the Table 7-6 compilation 

(Be-7, gallium-67, selenium-75, osmium-185 [Os-185], mercury-197 [Hg-197]),6

 the 1995 

nuclide library would not likely have included them due to lack of operational prevalence by that 

later time. It would be useful to confirm this by reviewing earlier WBC libraries and likewise 

ascertaining whether actual monitoring was taking place in the 1955–1988 timeframe. All told, 

half (14) of the 28 radionuclides listed in Table 7-6 as lacking evidence of monitoring appear to 

be radionuclides for which WBC monitoring would have been necessary. 

6 For validation, SC&A cross-referenced these with WBC monitoring results at LANL for the post-1975 period.

The technological ability to identify, and the capability needed to analyze, specific radionuclide 

intakes at the national laboratories (e.g., Los Alamos National Laboratory [LANL], Lawrence 

Berkeley National Laboratory [LBNL], Brookhaven National Laboratory, and Lawrence 

Livermore National Laboratory), particularly for MAPs and MFPs, was still in its infancy in the 

1950s and 1960s. In general, these methods were not yet applied to bioassays on a routine basis 

(and the results recorded and rendered useful for DR) until later in the 1970s and 1980s because 

they required considerable development time to improve detector sensitivity, calibration, and 

stability. To illustrate, like ORNL, LANL developed WBC technology in the late 1950s and 

began performing routine in vivo counts in June 1969 (unlike ORNL, which did not perform 

routine WBCs until 1989). However, even when the technological capability was available 

(e.g., in the 1970s and 1980s), some laboratories, such as LANL, did not maintain energy spectra 

information for exotics in their WBC libraries and did not routinely monitor for them. 

As SC&A pointed out in its review of NIOSH’s ER for LANL (SC&A 2010, p. 9):

A preliminary review of this premise, i.e., that the availability of monitoring 

capability alone would have been sufficient to enable detection and determination 

of uptakes of exotics, is not necessarily supported by field evidence. For example, 

an internal audit of the internal dosimetry program by the U.S. Department of 

Energy (DOE) as late as 2001 found that thorium-232 and the short-lived 

radionuclides generated at the Los Alamos Neutron Science Center (LANSCE), 

while required procedurally for routine internal dosimetry evaluation, were not 

included in the in-vivo program library at that time.... Interviews with LANL 

internal dosimetrists indicated that, while they are uncertain about the degree of 

attention afforded the exotic radionuclides in the early part of the program 

(because exposures were rare), they believe that the system was capable of 

detecting them. However, again, no documentation was found or offered that 

would corroborate a LANL practice in the 1970s and 1980s to “look for” these 

exotics beyond an “event driven” circumstance, where they would be targeted 

due to suspected elevated exposure potential. And, again, few data points 

apparently exist to demonstrate that such attention was being given to them. 

An obvious question would be whether ORNL was much more advanced and ahead of its time 

than LANL and these other laboratories in this respect (or not), and was the WBC fully 

 
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operational and capable of monitoring for these exotics for which in vitro methods would not 

have been available? If field evidence indicates it was not, as with LANL, the premise that 

available capability translates to adequate monitoring does not hold true. 

This exact premise was also tested in an SC&A review of the bioassay program at LBNL in 

terms of capability to monitor for specific radionuclides associated with the laboratory’s 

extensive accelerator operations (SC&A 2014). This review addressed the claim in a 2013 

NIOSH white paper supporting the LBNL site profile that “LBNL’s internal bioassay program, 

which was fully operational by 1962, was capable of determining specific alpha, beta, and 

gamma emitting radionuclides. Urinalysis and fecal sampling were performed, as well as whole 

body counting” (NIOSH 2013c, p. 3). SC&A sampled the NOCTS claimant file to validate the 

presence of bioassay data, particularly radionuclide-specific information, for job titles for which 

exposure potential for intakes would have been likely. The following is an excerpt of results of 

that review. 

While LBNL was a forerunner in accelerator health physics, it appears from an 

SC&A preliminary evaluation of NIOSH claimant files for LBNL that the 

capability to analyze specific radionuclides may have remained in a laboratory 

development stage, as opposed to being applied to routine bioassays, especially 

for WBCs. LBNL claim files were searched for POCs <50% (to ensure a complete 

DR) and if there was a DR report on file (to see how the internal doses were 

determined). This resulted in 195 claims. Claims with job titles that indicated 

potential exposure were selected for investigation. This included Physicist, 

Nuclear Physicist, Chemist, Lab Tech, Technician, Researcher, Accelerator 

Operator, Maintenance, HP, Machinist, and Magnet Tester. A total of 25 

claimants that worked some time during the period 1960s–1980s were analyzed 

by reviewing their DOE Response files and DR reports to determine if bioassays 

were recorded, and if so, what bioassay information is available; i.e., frequency, 

urinalyses, WBCs, and radionuclide identification. From this review, there did not 

appear to be many bioassay results recorded, and very few routine bioassays; 

only 4 claimants had any bioassay records out of the 25 reviewed. What bioassays 

were recorded generally did not contain nuclide-specific information (mainly 

gross gamma, beta, and alpha counts) and did not appear to be used in the DR 

process, except for a 1971 P-32 measurement for a potential acute intake. It 

would seem reasonable to expect that some of the personnel that worked at the 

facilities on a routine basis, such as operators and technicians, would have some 

records of routine, or at least periodic, bioassays in their records if the HP 

program was firmly in place and operational by 1962. [SC&A 2014, pp. 1–2].

Again, as the LBNL assessment demonstrated, the existence of monitoring capabilities and 

procedures belied the scope and completeness of the bioassay records themselves and required a 

reexamination of available LBNL bioassay data to determine alternate means to estimate internal 

dose beyond reliance on specific bioassay results.

For ORNL, a basis for a similar concern can be found in internal self-audits or appraisals of the 

laboratory’s internal dosimetry program. In one such self-appraisal of the radiation protection 

program for ORNL’s internal dosimetry program, as implemented at HFIR, the following 
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findings and recommendations, made in 1988, raise questions about the formality, accuracy, and 

effectiveness of the ORNL internal dosimetry program: 

• “There is no technical basis documented for whole-body counting or bioassay” 

(MMES 1988, PDF p. 38). 

• “Procedures for performing internal dose assessments should be developed. The current 

practice of not performing dose assessments until activity levels indicate greater than 

50 percent maximum permissible organ burden (MPOB) appears to be too high” 

(MMES 1988, PDF p. 42). 

• “Frequency schedules have been established and are documented in Appendix 12 of the 

Health Physics Manual; however, the established frequencies for whole-body counting of 

HFIR personnel are not followed” (MMES 1988, PDF p. 39). 

• “Procedures should be developed for establishing Minimum Detectable Activities for the 

whole-body counter” (MMES 1988, PDF p. 43). 

Another illustration of capability and procedures not being a sufficient basis alone to determine

bioassay feasibility was a significant Price-Anderson Amendments Act violation and penalty 

levied by DOE on ORNL in 1998. This was for lack of broad compliance with bioassay 

requirements under 10 CFR Part 835 regulatory requirements by ORNL construction contractor 

MK-Ferguson, as well as by Martin Marietta, the operating contractor, which included this 

finding (DOE 1998, p. 2): 

Further, after identifying problems with the bioassay program in October 1996, 

i.e., that approximately 100 positive bioassay results had been identified as 

positive that had previously been considered negative, results for these two 

workers [the specific subject of the enforcement action] were administratively 

invalidated without further evidence that uptakes had not occurred. These 

repeated failures resulted in additional 10 CFR 835 deficiencies in the areas of 

record keeping and issuance of accurate worker annual exposure reports. Other 

deficiencies identified during the investigation included (1) missed bioassay 

sampling, (2) failure to initiate special follow-up bioassay monitoring as 

required, (3) failures to notify workers of their exposures in a timely manner, and 

(4) failures to implement work restrictions in accordance with written procedures.

If fundamental problems existed with ORNL bioassay program compliance and implementation 

well into the 1990s, it underscores the importance to go beyond the written procedures and 

methods to investigate whether actual practice corresponded to these established capabilities and 

procedures in the earlier time period.

In summary, the stated purpose of RPRT-0090 to evaluate monitoring capability by matching 

identified radionuclides with corresponding bioassay methodologies as a gap analysis for 

demonstrating feasibility seemingly overlooks more significant program shortfalls in the 1950s 

into the 1980s. Notably, the lack of routine bioassay monitoring would lead to missed intakes, 

and limitations on whole-body counting capability for MAPs and MFPs could preclude reliable 
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monitoring of such intakes. Reliance on inventory-based gauges of “radiotoxicity” or ORNL’s 

capability to overcome these and other technological or programmatic limitations by being able 

to develop needed monitoring is questionable and would need some level of corroboration. 

Finding 6: Adequacy and implementation of in vivo bioassay program not addressed. 

Information is lacking for the actual implementation of the ORNL in vivo program, including 

what and how radionuclides were monitored in practice, what and how workers were identified 

and included for counting, and how capability to monitor for MAPs, MFPs, and exotic 

radionuclides paced both technology developments and onsite monitoring practice (e.g., routine 

vs. nonroutine monitoring). SC&A recommends that the Work Group request a review of 

available records, particularly internal dosimetry program records and WBC nuclide libraries, 

and scheduling of interviews with appropriate ORNL dosimetry staff. 
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7 POST-1988 INVENTORY AND CLEANUP

RPRT-0090, page 6, states:

The period under evaluation in this report is August 1, 1955, through December 

31, 1988. The start date was selected to coincide with the end of the SEC period 

(NIOSH 2012). The end date coincides with the end of large-scale isotope 

production at ORNL.

Although large-scale production of some of the radionuclides may have ended at ORNL in 1988, 

the remaining inventory was still present and required storage/disposal, and many facilities 

remained intact awaiting future activities. It is clear that after that date, the facilities involved 

were in various phases of continued operation, abandonment, deactivation, and D&D. During 

this period after 1988, workers, consisting primarily of surveillance, maintenance, and cleanup 

workers (e.g., CTWs), entered these buildings and should have been monitored for the 

radionuclides to which they would have been potentially exposed. However, for various reasons, 

including lack of institutional memory regarding historic facility use of isotopes, state of 

transferrable contamination, and surveillance capabilities, workers may have been exposed to 

facility contamination in the isotope production facilities in question during the postproduction 

period. 

For example, in Building 3026-C (Krypton Enrichment Facility), workers were contaminated in 

1988 with tritium that went initially undetected because routine tritium surveillance had been 

discontinued because of nonfunctioning tritium air monitors, lack of adequate survey 

instruments, and no HP surveillance prior to entry for work activities (Ramey 1988). 

In 1990, for Building 3517 (Fission Product Development Laboratory), “craftspersons” were 

directed to begin wearing special protective clothing upon entry to due to “recent contamination 

incidents” involving transferrable radioactive contamination in “relatively inaccessible areas”

(Patton 1990). 

In the mid-1990s, ORNL proceeded to conduct facility characterization studies and, ultimately, 

D&D for Building 3515, the FPPP, one of the first facilities at ORNL to extract radioisotopes 

from liquid radioactive wastes (Mandry and Snedaker 1994). While this longstanding ORNL 

operation was not administratively part of Isotope Production (as noted earlier), it clearly 

involved a wide range of exotic radionuclides and extensive contamination, and it would have 

entailed corresponding bioassay coverage for workers during D&D activities. 

With the uncertainty about what radionuclides were present in abandoned or deactivated 

facilities, how were surveillance and maintenance, cleanup, and other workers monitored for 

internal emitters that may have been present in these facilities? What facility operational and 

contamination characterization was performed, and what bioassay methods were used?
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Finding 7: Unclear treatment of post-1988 monitoring capability during abandonment, 

deactivation, and decontamination and decommissioning phases.

After radionuclide production ended, the adequacy of monitoring and feasibility of assigning 

intakes from the storage, disposal, and D&D of the facilities has not been addressed. This issue 

is especially important for the ORNL Isotopes Division because it processed and concentrated

unusual radionuclides that would not be encountered during the normal D&D process.
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8 SUMMARY AND CONCLUSIONS

SC&A evaluation of RPRT-0090 indicated the following finding and observations that need 

further clarification and evaluation.

8.1 FINDINGS

Finding 1: Scope of RPRT-0090 needs to be clearly defined.

SC&A finds that the scope of RPRT-0090 needs to be clarified in terms of whether (and how) it is 

meant to encompass the “reserved” portion of the ER for “cyclotrons, accelerators, and 

reactors” and whether NIOSH intends to address the full scope of radionuclides involved in 

waste management (including D&D), site-wide construction, and maintenance.

Finding 2: Incomplete radionuclide and radioisotope facility inventory.

A sampling of the radionuclides listed in Table 7-2 found a few missing when compared with 

operational and customer records. Likewise, a few ORNL facilities that historically handled 

radioisotopes are also not included in those cited and addressed in RPRT-0090. Given the 

operational diversity of ORNL accelerator and reactor operations, consideration should be 

given to an inventory scope that encompasses isotopic source terms broader than that of the 

Isotope Division.

Finding 3: Attachment A in vitro bioassay methods lack information about actual 

implementation.

In vitro bioassay methods are outlined in Attachment A, but it does not include any discussion or 

references regarding their actual field implementation. The exclusion of comparable in vivo

monitoring methods makes a review of ORNL monitoring capability incomplete.

Finding 4: Feasibility of monitoring 28 radionuclides not adequately addressed.

While the 28 radionuclides were discussed in Section 7.2 and some of their characteristics were 

listed in Tables 7-4, 7-5, and 7-6 of RPRT-0090, the feasibility of monitoring for intakes for DR 

purposes was not completely addressed, particularly given the lack of routine bioassays in the 

earlier years. Methods for accounting for the lack of monitoring of these radionuclides need to 

be addressed in more detail, and an acceptable resolution derived. SC&A finds that it is not 

possible at this time to validate implementation without further onsite review, including 

document review and interviews with health physicists of the time period involved.

Finding 5: 1955 and 1956 intakes may not be bound by earlier coworker data.

Assessment of RaLa radioiodine releases at X-10 indicates the highest annual releases occurred 

during the campaign to process Hanford slugs during 1956. Therefore, the radioiodine 

production and releases during the years used for coworker development (1947–1949) do not 

appear to bound the production throughput, at least during 1956 and possibly 1955.

Finding 6: Adequacy and implementation of in vivo bioassay program not addressed.

Information is lacking for the actual implementation of the ORNL in vivo program, including 

what and how radionuclides were monitored in practice, what and how workers were identified 

and included for counting, and how capability to monitor for MAPs, MFPs, and exotic 

radionuclides paced both technology developments and onsite monitoring practice (e.g., routine 
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vs. nonroutine monitoring). SC&A recommends that the Work Group request a review of 

available records, particularly internal dosimetry program records and WBC nuclide libraries, 

and scheduling of interviews with appropriate ORNL dosimetry staff. 

Finding 7: Unclear treatment of post-1988 monitoring capability during abandonment, 

deactivation, and decontamination and decommissioning phases. 

After radionuclide production ended, the adequacy of monitoring and feasibility of assigning 

intakes from the storage, disposal, and D&D of the facilities has not been addressed. This issue 

is especially important for the ORNL Isotopes Division because it processed and concentrated 

unusual radionuclides that would not be encountered during the normal D&D process. 

8.2 OBSERVATIONS

Observation 1: Inventory discrepancy. 

A sampling of some of the inventory of the radionuclides for the early years indicated some 

discrepancies in inventory between Table 7-2 in RPRT-0090 and NIOSH’s X-10 Inventory 

spreadsheet.

Observation 2: Specific alpha-emitting radionuclide needs to be identified for DR. 

The specific radioisotope monitored is not always presented in NIOSH’s X-10 Database as it 

generally is in the NOCTS files. Gross alpha results could be applied to many radionuclides. Is 

the information on the original bioassay cards available in the X-10 Database, and will the X-10 

Database be used in DR or coworker model development? 

Observation 3: Trans-plutonium radionuclides may need further analyses. 

SC&A is concerned that assigning trans-plutonium gross alpha counting results as Am-241 

intakes without consideration of other potential trivalent alpha-emitting actinides (such as 

Bk-249, Cf-252, Cm-242, Cm-244, etc.) and their individual radiotoxicity could result in 

underestimating the internal dose. It could be beneficial to determine if assigning the intake as 

Am-241 is claimant favorable, considering the exotic trans-plutonium radionuclides at ORNL.

Observation 4: Use of gross beta or gamma count data could result in underestimate of 

assigned dose. 

Using gross beta or gamma count data without knowledge of the radionuclide the counter was 

calibrated with and the radionuclides in the bioassay sample could result in assigning the 

incorrect radionuclide and radioactivity content because of different counting efficiencies for the 

different energy of beta particles and gamma photons. Has this issue been addressed for DR for 

ORNL claimants? Additionally, bioassay data for at least one beta-emitting radionuclide 

(Ru-106) could not be located for several years that Table 7-2 indicated it was available. 

Observation 5: The results in Table 7-6 depend on inventory used. 

As outlined in Observation 1, there appear to be some discrepancies in the inventory used by 

NIOSH compared to those provided to SC&A for evaluation of RPRT-0090. These discrepancies 

change a few of the results of Table 7-6, as illustrated in Table 3 of this report. 

Observation 6: Additional RaLa production information should be provided. 

NIOSH should provide an evaluation and discussion of any potential differences in exposure 
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potential between commercial radioiodine production and the radioiodine produced via the 

RaLa operation to justify the extrapolation of exposures occurring during the years 1947–1949 

to the unmonitored period (1955–1962). 

Several of the above issues identified by SC&A may have resulted from NIOSH updating the 

inventory list that was used in preparing RPRT-0090 of 2018, whereas NIOSH’s X-10 Inventory 

spreadsheet was dated February 2015. 
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9 REFERENCES

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Henley 1978. Henley, L.C., Radiochemical Procedures, ORNL/TM-6372, Oak Ridge National 

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Hickey et al. 1993. Air Sampling in the Workplace: Final Report, NUREG-1400, E.E. Hickey, 

G.A. Stoetzel, D.J. Strom, G.R. Cicotte, C.M. Wiblin, and S.A. McGuire, Office of Nuclear 
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Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC. September 1993. 

[SRDB Ref. ID 20129] 

Kohring, Mark W., 1990. Internal memorandum from Mark W. Kohring, MS-6256, to T.M. 

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MacPherson 1968. Internal memorandum from H.G. MacPherson, Union Carbide Nuclear 

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Control Limits. January 8, 1968. [SRDB Ref. ID 18591, PDF pages 125–127] 

Mandry and Snedaker 1994. G.J. Mandry and W. Snedaker, Decontamination and 

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MMES 1988. Action Plans for Recommendations on Radiation Protection, Comprehensive 

Safety Appraisal and Upgrade Program (CSAUP) for the High Flux Isotope Reactor (HFIR), 

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NIOSH 2007. Oak Ridge National Laboratory – Site Description. ORAUT-TKBS-0012-2, 

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NIOSH 2013b. niosh_04282013, National Institute for Occupational Safety and Health, 

Cincinnati, OH. April 28, 2013. 

NIOSH 2013c. White Paper: NIOSH Evaluation of the Internal and External Monitoring 

Programs at the Lawrence Berkeley National Laboratory, National Institute for Occupational 

Safety and Health, Cincinnati, OH. December 4, 2013. 

NIOSH 2015. X-10 Exotics Workbook_022015 kwv, National Institute for Occupational Safety 

and Health, Cincinnati, OH. February 20, 2015. 

NIOSH 2018a. Monitoring Feasibility for Exotic Radionuclides Produced by the Oak Ridge 

National Laboratory Isotopes Division, ORAUT-RPRT-0090, Revision 00, National Institute for 

Occupational Safety and Health, Cincinnati, OH. March 29, 2018. 

NIOSH 2018b. Type0_archiveR1, National Institute for Occupational Safety and Health, 

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ORNL 1957. Radioisotope Production and Process Development Annual Report for 1956, 

ORNL-2303, H.E. Seagren, Oak Ridge National Laboratory, Oak Ridge, TN. June 10, 1957.

[SRDB Ref. ID 103912]

ORNL 1965. List of AEC Radioisotope Customers with Summary of Radioisotope Shipments, 

FY 1964, TID-21620, Oak Ridge National Laboratory, Oak Ridge, TN. March 11, 1965. [SRDB 

Ref. ID 40950]

ORNL 1992. Oak Ridge National Laboratory Isotopes Facilities Shutdown Program 

Implementation Plan, ORNL/TM-11829, L.G. Hill, R.E. Eversole, and R.K. Kibbe, Oak Ridge 

National Laboratory, Oak Ridge, TN. May 1992. [SRDB Ref. ID 118653]

ORNL 1995. Oak Ridge National Laboratory Internal Dosimetry Group In Vivo Monitoring 

Technical Basis Document, Revision 0, Oak Ridge National Laboratory, Oak Ridge, TN.

August 30, 1995. [SRDB Ref. ID 11500]

ORNL 1997a. Final Deactivation Report on the Radioisotope Production Lab-H, Building 3118, 

at Oak Ridge National Laboratory, Oak Ridge, Tennessee, ORNL/ER-407, Oak Ridge National 

Laboratory, Oak Ridge, TN. August 1997. [SRDB Ref. ID 104704]

ORNL 1997b. Final Deactivation Report on the Tritium Target Facility, Building 7025, at Oak 

Ridge National Laboratory, Oak Ridge, Tennessee, ORNL/ER-406, Oak Ridge National 

Laboratory, Oak Ridge, TN. August 1997. [SRDB Ref. ID 32261]

Patton 1988. Internal memorandum from B.D. Patton, MS-6018, to J.R. Hightower, Martin 

Marietta Energy Systems, Inc., Isotope Area Facilities. December 6, 1988. [SRDB Ref. ID 

121570]

Patton 1990. Internal memorandum from B.D. Patton, MS-6019, Martin Marietta Energy 

Systems, Inc., to J.L. Hammontree, Special Protective Clothing Requirements for Building 3517.

November 30, 1990. [SRDB Ref. ID 121697]

PNL 1984. List of DOE Radioisotope Customers with Summary of Radioisotope Shipments, 

FY 1983, PNL-5126, D. A. Baker, Pacific Northwest Laboratory. August 1984. [SRDB Ref.

ID 77784]

Ramey 1988. Internal memorandum from D.W. Ramey, MS 6019, Oak Ridge National 

Laboratory, to J.A. Tompkins, Tritium Contamination in Building 3026-C. June 21, 1988.

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SC&A 2010. SC&A’s Review of NIOSH’s Evaluation Report for Los Alamos National 

Laboratory Special Exposure Cohort Petition SEC-00109: Preliminary Issues – Availability of 

Bioassay Records and Adequacy of In-Vivo Dosimetry, SCA-TR-SEC2010-0004, Revision 0, 

SC&A, Inc., Vienna, VA. April 8, 2010.

SC&A 2014. Memorandum from Joe Fitzgerald and Ron Buchanan, SC&A, Inc., to LBNL Work 

Group, NIOSH December 4, 2013, Response to SC&A Comments. February 4, 2014.
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SC&A 2018. SC&A Evaluation of NIOSH White Paper, “Method to Assess Internal Dose Using 

Gross Alpha, Beta, and Gamma Bioassay and Air Sampling at the Lawrence Berkeley National 

Laboratory,” SCA-TR-2018-SP003, Revision 0, SC&A, Inc., Arlington, VA. April 18, 2018.

Taulbee 2012. Timothy Taulbee, Oak Ridge National Laboratory Special Exposure Cohort 

Petition Evaluation Report SEC-00189, Presentation to the Advisory Board on Radiation and 

Worker Health, National Institute for Occupational Safety and Health, Division of Compensation 

Analysis and Support, Denver, CO. September 2012.
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APPENDIX A: EXAMPLE OF EVALUATION OF CM-244 AND RU-106

BIOASSAY DATA

Table A-1. Cm-244 Data

Year X-10 Inventory

RPRT-0090 

Table 7-2 

Cm-242

Number of Bioassays with 

Code 000, TP0, or CM0 in 

X-10 Database

1955 None No 1

1956 None No 0

1957 None No 0

1958 None No 2

1959 None No 0

1960 None No 0

1961 None No 1

1962 None Green [?] 0

1963 None Green [?] 1

1964 0.002 gm Green Many

1965 None No Many

1966 3.5 gm Green Many

1967 19 gm Green Many

1968 144 gm Green Many

1969 213gm Green Many

1970 29.4 gm Green Many

1971 30.9 gm Green Many

1972 96.7 gm Green Many

1973 416.1 gm Green Many

1974 117 gm Green Many

1975 1,435 mg Green Many

1976 174 mg Green Many

1977 1,125 mg Green Many

1978 51.5 gm Green Many

1979 132745 mg Green Many

1980 10 mCi Green Many

1981 725 mg Green Many

1982 1,622 mg Green Many

1983 151,002 mg Green Many

1984 1,061 mg Green Many

1985 1,516 mg Green Many

1986 61,000 gm Green Many

1987 141 mg Green Many

1988 22 gm Green Many * The Green [?] notation means it is not clear why this cell is marked green in Table 7-2.
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Table A-2. Ru-106 Data

Year X-10 Inventory

RPRT-0090 

Table 7-2 

Ru-106 * 

Number of Bioassays 

with Code 000, 013, 

GB0, or RU6 in X-10 

Database

1955 717 Green 2

1956 330 Green [?] 0

1957 397 Green [?] 0

1958 None No 2

1959 1,360 Green [?] 0

1960 936 Green Some

1961 233 Green Some

1962 338 Green Some

1963 810 Green Some

1964 937 Green Some

1965 500 Green Some

1966 326 Green Some

1967 221 Green Some

1968 201 Green Some

1969 35 Green Some

1970 119 Green Some

1971 85 Green Some

1972 123 Green Some

1973 105 Green Some

1974 610 Green 2

1975 7 Green [?] 0

1976 346 Green 6

1977 133 Green 1

1978 704 Green [?] 0

1979 286 Green Some

1980 350 Green Some

1981 72 Green Some

1982 360 Green Some

1983 438 Green Some

1984 132 Green Some

1985 59 Green Some

1986 81 Green [?] 0

1987 48 Green [?] 0

1988 15 Green [?] 0

* The Green [?] notation means it is not clear why this cell is marked green in Table 7-2.
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APPENDIX B: EXAMPLES OF NOCTS ORNL RECORDS

Figure B-1. 1955 Plutonium, Uranium, and Strontium
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Figure B-2. 1955 Thorium 
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Figure B-3. 1964 Cm-244 
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Figure B-4. 1965 Cm-242 and Am-241 
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APPENDIX C: EXAMPLE OF X-10 DATABASE FOR AM-241

Table C-1. Excerpt for Part of 1955 and 1956 Applicable to Am-241 from ORNL Database 

Containing 25,162 Entries

Bioassay Code (a) dpm per Sample (b) dpm per 24 Hours Date Received

GU0 0 0000000.00 12-Dec-55

GU0 0 0000000.00 13-Dec-55

GU0 0.1 0000000.01 14-Dec-55

GF0 1.5 0000000.15 16-Dec-55

GF0 8.8 0000000.88 17-Dec-55

GF0 2.5 0000000.25 17-Dec-55

GF0 26 0000002.60 17-Dec-55

GU0 0 0000000.00 17-Dec-55

GF0 47 0000004.70 18-Dec-55

GF0 12 0000001.20 18-Dec-55

GF0 4.6 0000000.46 18-Dec-55

GU0 0 0000000.00 19-Dec-55

GU0 0.1 0000000.01 19-Dec-55

GU0 0 0000000.00 19-Dec-55

GF0 3.6 0000000.36 21-Dec-55

GU0 0.2 0000000.02 21-Dec-55

GU0 0 0000000.00 27-Dec-55

GU0 0 0000000.00 27-Dec-55

GU0 0 0000000.00 03-Jan-56

GF0 22 0000002.20 04-Jan-56

GF0 1.2 0000000.12 05-Jan-56

GF0 0.4 0000000.04 05-Jan-56

GF0 15 0000001.50 05-Jan-56

GF0 0.9 0000000.09 06-Jan-56

GU0 0.1 0000000.01 06-Jan-56

GF0 8.7 0000000.87 06-Jan-56

GF0 33 0000003.30 06-Jan-56

GF0 13 0000001.30 06-Jan-56

GF0 0.8 0000000.08 07-Jan-56

GU0 0 0000000.00 08-Jan-56

GF0 12 0000001.20 08-Jan-56

GF0 19 0000001.90 08-Jan-56

GU0 0 0000000.00 09-Jan-56

GU0 0 0000000.00 09-Jan-56

GU0 0 0000000.00 09-Jan-56

(a) GU0 = gross alpha in urine sample. GF0 = gross alpha in fecal sample.

(b) dpm = disintegrations per minute.

In addition to these data, the spreadsheet contains other information, such as the worker’s social 

security number and name associated with each bioassay sample.
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APPENDIX D: EVALUATION OF INTERNAL DOSE USING A 

FRACTIONAL INTAKE BASED ON AMOUNT OF MATERIAL IN 

PROCESS

Section 7.2 of RPRT-0090 describes a method of deriving internal doses for the 28 radionuclides 

requiring further evaluation that is based on assuming the intake potential can be bounded using 

a fraction of the total radionuclide inventory in process during a given year. Specifically, RPRT0090 (p. 40) states:

An evaluation of the dosimetric consequences of each of the 28 radionuclides 

requiring further analysis was conducted. The committed dose to the maximally 

exposed organ from inhalation of 1 × 10-5 of the total annual inventory was 

computed for the year with the maximum recorded inventory.... The factor of 

1 × 10-5 was selected based on the guidance in NUREG-1400, which postulates 

that 1 × 10-6 times the material handled could serve as a reasonable estimate of 

the quantity that could be inhaled (Hickey et al. 1993). A factor of 10 was added 

to ensure a conservative evaluation. 

Although NIOSH cites NUREG-1400 (Hickey et al, 1993), the underlying data forming the basis 

of this factor7

 is found in Brodsky 1977 and 1980. Brodsky 1977 focused the evaluation on the 

potential for tritium intake at various facilities and processes. Although tritium is not one of the 

28 radionuclides evaluated in Section 7.2 of RPRT-0090, the volatile nature of tritium 

compounds (in particular tritiated water [HTO] and tritium gases) could be considered a 

conservative comparison to most other contaminants. Brodsky 1977 evaluated the following 

tritium intake situations summarized in Table D-1. 

7 While the term “factor” is used in RPRT-0090, the underlying literature uses the term “fractional intake” and thus 

that will be used in this evaluation.

Table D-1. SC&A’s Summary of Fractional Intake Data in Brodsky 1977 

Industry or Activity Type Form of Tritium Protective 

Enclosure

Estimated Fractional 

Intake*

Luminous timepiece Luminous paint Hood, bench 6×10-6 per employee*;

1.2×10-4 for 20 employees

Luminous timepiece Luminous paint, 

organic polymer

Open bench, 

modified with 

local exhaust

6×10-7 per employee*;

(2×10-6 for 3 employees)

Luminous paint, mixing 

polymer, phosphor and binder

Luminous paint, 

organic polymer

Glovebox, 

single thickness 

rubber gloves

10-7 per employee

Luminous paint, research 

development

Gas, HTO, various 

organics None 4×10-6 to 3×10-5 per employee

Luminous watch storage area Solid paint on 

timepieces

Poorly 

ventilated room 7×10-7 per employee

Academic laboratories Various Fume hoods <5×10-6 per employee

Academic laboratories – spill HTO Hood 10-2 highest observed 

employee dose

 
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Industry or Activity Type Form of Tritium Protective 

Enclosure

Estimated Fractional 

Intake*

Particle accelerators Adsorbed gas and 

HTO Unspecified <2×10-6 per employee

Tritium gas processing Tritium gas

Special glass 

vessels and 

glove boxes

<10-10 per employee

Reactor operations HTO in reactor 

coolant

Primary coolant 

system 

enclosure

10-5 (based on concentration 

in coolant)

10-10 of activity in reactor

*Total number of employees involved in these scenarios is not currently known.

As seen in Table D-1, the fractional intake based on the amount of tritium material in process for 

different situations varied from a high of 10-2 to a low of 10-10. In the latter (maximizing) case, 

the intake fraction was for the highest exposed individual who was involved in a spill of 

approximately 73 millicuries (mCi) of tritiated water on the benchtop of a hood that was cleaned 

up. The worker involved in the cleanup was only wearing a set of gloves for personal protective 

equipment, and so it is likely a significant portion of the intake was caused by absorption through 

the skin. Several of the remaining examples were bounded by, or within range of, the NIOSHassumed fractional intake value of 10-5. 

Brodsky 1980 also presents four situations that involved elements other than tritium. Table D-2 

summarizes the estimated fractional intakes from these scenarios.

Table D-2. Summary of Fractional Intake Data from Brodsky 1980

Case Description Fraction 

Deposited

I-131 escaping from open vial in hospital environment (145 mCi process) 1.2×10-6

Technician mixed AmO2-Au powder, compacted, sintered, rolled into foils over 

3–5 years (9 Ci in process) 2×10-7 

Two technicians cut into Ir-192 pellets with lathe inside hot cell, hot cell 

ventilation off (2,000 Ci in process) 2×10-7

Two technicians cut into Ir-192 pellets with lathe inside hot cell, hot cell 

ventilation off (75 Ci release) 1.4×10-7 

Hospital worker opening vial of I-131 for therapeutic administration (60 mCi in 

process) 8×10-7 

As seen in Table D-2, each of the five situations involved other radionuclides that were 

sufficiently bounded by the NIOSH assumed fractional intake value of 10-5. 

It should be noted that NUREG-1400 refers to the fractional intake factor of 10-6 as a “rule of 

thumb” and that to arrive at a more realistic estimate of the intake, several modifying factors 

should be applied, including the release fraction (R), confinement factor (C), and the 

dispersibility (D) of the material. The release fraction accounts for the physical and chemical 

properties of the material being evaluated. The confinement factor accounts for the configuration 

of the facility’s engineering controls where the material is being handled (such as a hood or 

glovebox). Finally, the dispersibility accounts for the physical and chemical processing of the 

material being evaluated (such as heating or grinding). 
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Using these modifying factors, a more realistic estimate of the intake can be calculated using the

following formula: 

Estimated Intake = (Quantity of Material in Process) × (R) × (C) × (D) × (10-6) 

Table D-3 provides typical values for each of these modification factors. As seen in the table, 

application of the release fraction and confinement factor in the above equation will either not 

change or decrease the effective fractional intake below 10-6. Application of the dispersibility 

factor will either not change the intake estimate or potentially increase the intake by a factor of 

10 if cutting, grinding, heating or chemical reactions are occurring. However, it is likely that any 

such operations would occur (at a minimum) inside a vented hood, which would effectively 

cancel out the dispersibility factor. Therefore, based on the above equation, the fractional intake 

would only increase above 10-6 if a gas/volatile material were being handled in the open air while 

performing activities that increase the dispersibility. 

Table D-3. Typical Values for Modification Factors Used in Arriving at a More Realistic 

Intake Based on the Quantity of the Material in Process and 

the Fractional Intake Factor (10-6)

Modification Factor Description Value

Release Fraction Solid Material 0.001

Non-Volatile Powders and Liquids 0.01

Gases or Volatile Materials 1

Confinement Factor Glovebox 0.01

Vented Hood 0.1

Open Air 1

Dispersibility All Others 1

Cutting, Grinding, Heating, or Chemical 

Reactions

10

The fractional intake method of estimating potential internal doses described above assumes 

normal routine conditions over a long period of time (i.e., 1 year) and does not necessarily 

address acute incident scenarios. However, Brodsky 1980 (p. 993) does discuss a fractional 

intake method under accident conditions: 

Frank et al. ...have found from data collected in their survey that usually no more 

than 10-6 of material in process will enter the body of a worker in the event of a 

release caused by an explosion or other dispersing incident. Even for volatile 

materials at elevated temperatures, no more than 10-5 of the material in process 

entered the body after release. In several accident cases involving Pu, Am and Ir, 

which the author evaluated at the University of Pittsburgh whole-body counter, 

estimated fractional intakes of material in process were 10-6 or less, even for 

workers handling material at arms’ length at the time of the accident.... 

However, in the case of specific tritium or iodine compounds that may penetrate 

protective gloves and enter the body through the skin, there may be special 

circumstances where higher fractional intakes are possible.
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Clearly, the amount of “material in process” considered for an accident or incident scenario 

would be significantly less than the amount of material considered during non-upset conditions 

over a long period of time (in the RPRT-0090 scenario, 1 year). Although the above quote notes 

that incidents involving iodine could produce higher acute fractional intakes, NIOSH used a 

different method to evaluate iodine, described in Appendix C of RPRT-0090. 

CONCLUSION

Review of NUREG-1400 and the underlying documentation indicates that under routine 

conditions, the potential for intake of radioactive material is generally less than 10-6 multiplied 

by the amount of material in process. When the appropriate modification factors are included 

(release fraction, confinement factor), the value is likely much less than 10-6 under normal 

conditions. Estimates of intakes based on incidents or accidents are not specifically accounted for 

in this method and would have to be added to any routine estimate. Without specific knowledge 

of any incident conditions, frequency, and material amounts, it is difficult to assess how this 

might affect the total fractional intake estimate. However, SC&A agrees that use of a fractional 

intake factor of 10-5 represents a reasonably conservative methodology to estimate potential 

intakes to unmonitored workers under routine conditions in which off-normal occurrences are 

controlled and documented.