ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 21

Document Title:

External Radiation Dose Estimates For 

Individuals Near the 1958 Criticality 

Accident at the Oak Ridge Y-12 Plant

Document Number: ORAUT-OTIB-0057

Revision: 00

Effective Date: 05/15/2006

Type of Document: OTIB

Supersedes: None

Subject Experts: George D. Kerr and William G. Tankersley

Document Owner

Approval: Signature on File Approval Date: 05/04/2006

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 05/11/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 05/15/2006

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 2 of 21

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

05/15/2006 00 New technical information bulletin for external radiation dose 

estimates for individuals near the 1958 criticality accident at the Oak 

Ridge Y-12 plant. First approved issue. Training required: As 

determined by the Task Manager. Initiated by George D. Kerr and 

William G. Tankersley.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 3 of 21

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

4

1.0 INTRODUCTION .....................................................................................................................6

2.0 PURPOSE ...............................................................................................................................7

3.0 IDENTIFICATION AND INTERROGATION OF EXPOSED INDIVIDUALS............................... 7

4.0 DESCRIPTION OF BUILDING 9212 COMPLEX AND ACCIDENT .......................................... 9

5.0 EXTERNAL DOSE ESTIMATES FOR THE EIGHT MOST HIGHLY EXPOSED 

INDIVIDUALS ........................................................................................................................10

6.0 EXTERNAL DOSE ESTIMATES FOR OTHER INDIVIDUALS ............................................... 13

7.0 EXTERNAL DOSE MODIFICATION FOR USE IN IREP........................................................ 15

8.0 SUMMARY ............................................................................................................................18

REFERENCES..................................................................................................................................19

LIST OF TABLES

TABLE TITLE

3-1 Information on individuals near the accident ............................................................................

PAGE

8

5-1 Doses of record for the eight most highly exposed individuals ............................................... 13

5-2 Correction to recorded doses for the eight most highly exposed individuals........................... 13

6-1 Estimated doses for other individuals near the accident......................................................... 14

7-1 Theoretical neutron energy distribution .................................................................................. 16

7-2 Sample calculation of the neutron dose equivalent for Employee H (9).................................. 16

7-3 Dose equivalent estimates for all individuals near the accident .............................................. 17

LIST OF FIGURES

FIGURE TITLE

4-1 Building 9212 with location of accident.....................................................................................

PAGE

9

5-1 Thermal neutron capture probability as a function of the energy of neutrons 

normally incident on the front or back of the human body....................................................... 11

6-1 Comparison of estimated and measured gamma-ray doses for other 

individuals near the accident .................................................................................................. 15
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 4 of 21

ACRONYMS AND ABBREVIATIONS

A projected area of the human body in the neutron direction

cm centimeter

DEn first collision dose equivalent from neutrons

DEγ first collision dose equivalent from gamma rays

Dn first collision absorbed dose from neutrons

Dn(H) first collision absorbed dose from neutrons to a human

DOE U.S. Department of Energy

Dγ first collision absorbed dose from gamma rays

Dγ(H) first collision absorbed dose from gamma rays to a human

eV electron volt

ft feet

g gram

gal gallon

hr hour

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program (a computer program)

keV thousand electron volts

kg kilogram

L liter

MeV million electron volts

min minute

mL milliliter

mrem millirem

n neutron

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

POC probability of causation

r distance from center of reactive solution in barrel to the beltline of an individual 

exposed during the accident

RBE relative biological effectiveness

RPG radiation protection guideline

s second

S specific activity
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 5 of 21

Swb(B) activity of 24Na per mL of whole blood from burro

Swb(H) activity of 24Na per mL of whole blood from human

TIB technical information bulletin

U.S.C. United States Code 

UCCND Union Carbide Corporation – Nuclear Division

V volume of the human body

wk week

wR radiation weighting factor from ICRP Publication 60

yr year

Φ neutron fluence 

μCi microcurie 

§ section

ξ spectrum-average capture probability for neutrons within the human body
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 6 of 21

1.0 INTRODUCTION

Technical information bulletins (TIBs) are not official determinations made by the National Institute for 

Occupational Safety and Health (NIOSH) but are rather working documents that provide historical 

background information and guidance to assist in the preparation of dose reconstructions at particular 

sites or categories of sites. They will be revised in the event additional relevant information is 

obtained. TIBs may be used to assist the NIOSH staff in the completion of individual work required for 

each dose reconstruction.

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

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

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 7 of 21

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

2.0 PURPOSE

The purpose of this TIB is to review the available dosimetric data and its potential application in dose 

reconstruction for Y-12 workers who were near the nuclear criticality accident in Building 9212 of the 

Y-12 Plant in Oak Ridge, Tennessee, in 1958. The accident occurred on Monday, June 16, at 

approximately 2:05 p.m. (UCNC 1958; Callihan and Thomas 1959). Additional data relevant to 

individuals described in this TIB of a personal nature can be found in Y-12 1958 Criticality Accident 

Roster (ORAUT 2006a), which is available to dose reconstructors as needed.

3.0 IDENTIFICATION AND INTERROGATION OF EXPOSED INDIVIDUALS

By 2:45 p.m. on Monday, June 16, 1958, radiation surveys of workers for evidence of personal 

contamination and for indications of neutron activation of indium foils in security badges were 

underway at two control centers. Starting in 1955, strips of indium foil with a mass of approximately 

1 g had been included in the Y-12 security badges for all employees (McLendon 1959). Indium has a 

very large capture cross section for neutrons, and the beta and gamma radiation from the 116mIn 

isotope (half-life of 56.3 min) produced by neutron activation of the stable 115In isotope are easily 

measured. The measured activity of the foils is not very useful in radiation dosimetry, but is useful as 

a quick technique for sorting nuclear accident exposures in order of most to least severe. This quick 

search procedure was helpful in (1) preventing unnecessary overloading of medical facilities by lightly 

exposed workers, (2) providing reassurance to workers who had received small or negligible 

exposures, and (3) providing an immediate indication of the seriousness of the nuclear criticality 

accident to management. 

Interrogation of the workers assembled at the control centers was also begun in an attempt to 

establish the exact location of the accident. Very high readings were detected from the indium foil in 

the security badge of an employee (Employee A) who worked as a chemical operator in C-1 Wing. A 

process supervisor questioned the employee and concluded that the accident must have occurred in 

C-1 Wing of Building 9212. It was eventually determined that eight employees were in close proximity 

to the accident. The above chemical operator was within 3 to 6 ft, and the other seven were at 

distances ranging from 15 to 50 ft. A more complete assessment of the indium foils in the security 

badges of all employees indicated a total of 31 individuals with potentially significant neutron doses 

from the accident (UCNC 1958). Table 3-1 of this TIB provides a summary of certain available data 

for these 31 individuals, and ORAUT (2006a) contains additional personal data on the employees that 

dose reconstructors can access as needed. The identification numbers and letters for individuals in 

Table 3-1 and ORAUT (2006a) are taken directly from UCNC (1958), Y-12 Plant (1958) and Hurst, 

Ritchie, and Emerson (1959).

Only five individuals in the vicinity of the 1958 criticality accident were wearing a film badge dosimeter 

(see Table 3-1). The Y-12 policy at the time was to issue film badge dosimeters to workers if their 

potential dose from external radiation exposure exceeded 10% of established radiation protection 

guidelines (RPGs). The RPGs in 1958 were 1.5 rem/wk (75 rem/yr) to the extremities, 600 mrem/wk 

(30 rem/yr) to the skin, and 300 mrem/wk (15 mrem/yr) to the whole body (ORAUT 2005a,b). Gamma 

radiation from highly enriched uranium is normally not the controlling challenge to radiation protection 

(DOE 2004). However, the interaction of alpha particles from uranium with nuclei of fluorine and other 

low-atomic number atoms generates neutrons with energies of approximately 2 MeV. The magnitude 

of the neutron flux varies, based on the total activity of uranium (a function of enrichment) and the 

chemical compound in question (ORAUT 2005c). In general, an individual’s exposure to neutrons
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 8 of 21

Table 3-1. Information on individuals near the accident.

Employee 

identification

Job title

External

monitoring

(years)

2nd quarter 1958

gamma/beta

dose (mrem)

Distance 

from

Number Letter criticality (ft)

1 A Chemical Operator 1961–81 3–6 

2 B Electrician 1961–88 14

3 C Maintenance Mechanic 1961–72 17

4 D Electrician 1961–72 22

5 E Maintenance Mechanic 1961–80 22

6 F Welder 1961–76 25

7 G Maintenance Mechanic 1961–64 25

8 Equipment Tabulation Clerk 1954–57, 61–67 29

9 H Chemical Operator 1958, 61–65 47/34a 50

10 Chemical Operator 1961–62 71

11 Chemical Operator 1961–85 85

12 Chemical Operator 1961–78 92

13 Chemical Operator 1961–78 100

14 Health Physics Inspector 1951–53, 61–80 137

15 Electrician None 148

16 Process Foreman 1950 (1 qtr.), 61–82 175

17 Chemical Operator 1955, 57–62 333/247 178

18 Draftsman Engineer 1952, 61–87 183

19 Engineer 1961–87 190

20 Chemical Operator 1953–59, 61–80 333/250 212

21 Chemical Operator 1961–86 216

22 Record Clerk 1961–68 224

23 Chemist 1952–54, 61–84 260

24 Stenographer 1961–73 276

25 Chemist Associate 1953–55, 58, 61–88 167/432 280

26 Laboratory Analyst 1961–85 288

27 Laboratory Analyst 1952–54, 61–82 304

28 Development Engineer 1961–72 312

29 Product Inspector 1953–72 29/51 340

30 Road Foreman 1961–87 390

31 Receiving Clerk 1958–62 291/521 412

a. On supplemental film badge program from Feburary 1 to April 15, 1958. Not wearing a film badge at time of accident. 

generated by (alpha,neutron) reactions was low unless the individual was required to spend more 

than a few hours per week in close proximity to large storage containers of uranium fluoride 

compounds or spending time near storage or processing areas for large amounts of uranium fluoride 

compounds (DOE 2004). Therefore, most individuals working in chemical process areas for enriched 

uranium at the Y-12 Plant were only monitored on a routine basis for internal uranium exposures, and 

the contaminants in recycled enriched uranium were controlled at the Y-12 Plant so that both internal 

and external exposures were characteristic of those from uncontaminated enriched uranium (BWXT 

Y-12 2000).

Following the accident, a new film badge was developed for use at all DOE facilities operated by the 

Union Carbide Corporation – Nuclear Division (UCCND). This film badge served as a security badge 

and also provided for monitoring of both routine and accident-related radiation exposures (Thornton, 

Davis, and Gupton 1961; Hurst and Ritchie 1961; McRee, West, and McLendon 1965). Starting in 

January 1961, all employees at the Y-12 Plant and at other DOE facilities operated by UCCND were 

monitored using this combination security badge and radiation dosimeter. 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 9 of 21

4.0 DESCRIPTION OF BUILDING 9212 COMPLEX AND ACCIDENT

The 9212 Building complex at the Y-12 Plant includes Buildings 9212, 9809, 9812, 9815, 9980, and 

9981 (ORAUT 2005d; BWXT Y-12 2003; UCNC 1958). Buildings 9980 (Radiography) and 9981 

(X-Ray) were an integral part of Building 9212 as shown in Figure 4-1 (UCNC 1958). The accident 

occurred in an area in which salvable enriched uranium (approximately 90% 235U) was recovered from 

various materials by physicochemical methods (Callihan and Thomas 1959).

Figure 4-1. Building 9212 with location of accident.

Over 100 different processes could be performed in the Building 9212 complex. Building 9212, the 

largest building, was constructed in the early 1940s. It was a multistory steel frame structure infilled at 

the perimeter with hollow clay tile. The original mission of Building 9212 was to recover highly 

enriched uranium from the electromagnetic separation process. The original structure consisted of a 

central building (the Headhouse) 72 ft wide by 308 ft long (N-S direction) and four parallel wings 

projecting from the east side of the Headhouse, each 36 ft by 254 ft (the A, B, C, and D Wings). The 

open space between the wings was designed to help mitigate the effects of a nuclear criticality 

accident or a chemical explosion. In the late 1940s, Building 9212 was expanded to accommodate 

the increased production of uranium from the Oak Ridge Gaseous Diffusion Plant and to recover 

highly enriched uranium from scrap materials. In the late 1940s, new structures were erected in the 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 10 of 21

spaces between the existing A, B, and C Wings (these were called the A-1, B-1, and C-1 Wings) and 

an adjoining D-Wing (the D-1 Wing). A single-story steel frame structure 133 ft wide and 400 ft long 

was added in the early 1950s (the E-wing) adjacent to the D-1 Wing north of the Headhouse. The EWing was used to cast and machine enriched uranium components. 

The process phase in which the accident occurred was a temporary arrangement involving portions of 

a new installation in the startup stage (B-1 Wing) and an old installation in the shutdown stage (C-1 

and C Wings). This situation arose from delays in the startup of the new facilities in B-1 Wing for the 

conversion of uranium nitrate solution to uranium tetrafluoride. At the time of the accident, the 

uranium processing areas in Building 9212 were concerned with the required monthly accounting of 

uranium in inventory, which necessitated a stoppage of operations. The method of taking inventory 

varied with the form and concentration of the uranium. In certain process areas, where equipment 

contained dilute homogenous uranium solutions, a satisfactory accounting could be made by taking 

samples and computing the contents of known volumes. In the process area where the accident 

occurred, more precise accounting was required because of the higher uranium concentrations and 

the tendency of the solutions to deposit uranium-bearing solids in the equipment. This more precise 

accounting was obtained by processing the contents of several safe-geometry tanks for uranium 

tetrafluoride just before inventory. In addition, the process was to wash, dismantle, and swab out the 

safe-geometry tanks, collecting the washings in portable safe-geometry bottles. After reassembly, the 

safe-geometry tanks were prone to leak at the tank ends. Therefore, the tanks were filled with water, 

checked for leaks, and the water drained from the tanks before their return to operation.

In the interval between reassembly and leak testing, a highly enriched uranium nitrate solution had 

accumulated in the tanks through a valve that was supposed to provide isolation from operating 

equipment in the B-1 and C-1 Wings. The water being drained from the tanks was preceded by 

this solution. The accident occurred when a critical volume of uranium nitrate solution drained into a 

55-gal drum during this operation in which only water was expected to flow from the safe-geometry 

tanks of C-1 Wing. The drum was about 33 in. (84 cm) in height and 22 in. (56 cm) in diameter. A 

volume of 56.2 L containing 2.1 kg of 235U and standing at a height of 23.5 cm was estimated as the 

delayed critical configuration. 

There were no strong ambient neutron fields in C-1 Wing. Therefore, the system might have been 

prompt critical before any fissions occurred in the drum, and the neutron initiating the first fission 

reaction in the drum might have come from an (alpha,neutron) reaction in the uranium nitrate solution. 

Once fission started, the power level rose rapidly to a high level, and the fission energy produced gas 

bubbles by dissociation of the solution, which reduced its density and drove the system subcritical. 

Escape of the gas bubbles allowed the system to return to prompt critical, and with the delayed 

neutrons as a source, the power level rose again. This cycling probably continued for several minutes 

as the temperature of the solution increased. A high solution temperature would cause the system to 

settle into delayed critical and subsequently into a subcritical configuration due to the continued flow 

of water into the drum from the safe-geometry tanks in C-1 Wing. It was estimated that about 1 × 1017

fissions occurred during the first pulse and 1 × 1018 fissions occurred over a period of about 20 min. 

There was no evidence that solution in the drum splashed out of the drum during the accident or 

caused serious contamination to the area surrounding the accident (McLendon 1959). 

5.0 EXTERNAL DOSE ESTIMATES FOR THE EIGHT MOST HIGHLY EXPOSED 

INDIVIDUALS

Very early estimates of the radiation doses received by the eight most highly exposed workers, based 

on their known locations at the time of the accident and an estimate of the number of fissions during 

the accident, were unreasonably high (Hurst, Ritchie, and Emerson 1959). For example, the fast 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 11 of 21

neutron component of the total radiation dose received by Employee A was 1,500 rad. It is assumed, 

therefore, that the location of the eight most highly exposed employees and their exposure geometries 

were unknown during a significant part of their exposure. These facts, together with the fact that the 

eight most highly exposed employees were not wearing personnel monitoring devices, dictated that 

sodium activation in the body be used in the evaluation of their radiation doses (Hurst, Ritchie, and 

Emerson 1959). 

The capture of thermal neutrons by 23Na in the body gives rise to 24Na, which has a half-life of 15 hr 

and emits a 1.37-MeV gamma ray in cascade with a 2.76-MeV gamma ray per 24Na decay (Shleien, 

Slaback, and Birky 1998; Martin 2000). The human body is several mean free paths thick for fast 

neutrons. As a consequence and as shown in Figure 5-1, the probability that a fast neutron will be 

captured within the body as a thermal neutron is not a very sensitive function of the initial energy of 

the fast neutron (Cross 1981). The specific activity S of 24Na in the body is related to the capture 

probability ξ as follows:

7 5 0 61 μBq Na/g Na 24 23 Φ . ξ Φ

V

A S = . ξ ≅ Eq. 5-1

where V is the volume of the body, A is the projected area of the body in the neutron direction, Φ is 

the total neutron fluence at all energies, and ξ is the spectrum-average capture probability for 

neutrons within the body (IAEA 1982). The capture probability data shown in Figure 5-1 are for 

Figure 5-1. Thermal neutron capture probability as a function of the energy of 

neutrons normally incident on the front or back of the human body (Cross 

1981). 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 12 of 21

neutrons normally incident on the front or back of the body (Cross 1981). For neutrons normally 

incident on the sides of the body, both A and S are approximately 20% less. Therefore, there is an 

uncertainty of as much as 20% in an estimate of the neutron dose component if an individual’s body 

orientation in the neutron field is not known or if the orientation changes during the accident.

The neutron and gamma-ray doses to the eight individuals whose indium foils indicated the highest 

exposures were determined by the use of blood sodium activation. At approximately 5:00 pm on the 

day of the accident, 100 mL of blood was collected from several of the most highly exposed 

individuals (Hurst, Ritchie, and Emerson 1959; UCNC 1958). Each blood sample was placed in a 

small polyethylene bottle and the 24Na activity was determined using a scintillation counter. During 

the morning of the next day, a second set of blood samples was collected from all of the highly 

exposed individuals. This time only 50-mL blood samples were collected, an anticoagulant (heparin) 

was added to prevent clotting, and the blood samples were counted as before. The second set of 

measurements was used as the basis for dose estimation. To establish the relationship between 

neutron dose and blood sodium activation, a mock-up of the criticality accident was constructed and 

operated as a low-power reactor in two experiments (UCNC 1958; Hurst, Ritchie, and Emerson 1959). 

During the first experiment, the first collision doses from neutrons and gamma-rays were measured at 

a distance of about 200 cm (6 ft). The neutron dose was measured with an absolute fast-neutron 

dosimeter (Hurst and Ritchie 1961), and the gamma-ray dose was measured with an ionization 

chamber having carbon walls and CO2 gas (Hurst and Ritchie 1962; Ballweg and Meem 1951). After 

making a correction for the difference in the fission product gamma rays from the reactor mockup, a 

value of 2.8 was estimated as the ratio of the first collision doses from gamma rays and neutrons for 

the accident (Hurst, Ritchie, and Emerson 1959). During the second experiment, a burro was 

exposed at the same distance from the mockup to a first collision dose of 48 rad from fast neutrons. 

The burro was chosen because its torso was comparable in size to that of a man and the amount of 

sodium per gram of blood serum is nearly the same for burro and man (Hurst, Ritchie, and Emerson 

1959; Auxier, Sanders, and Snyder 1961). Blood samples were collected from the burro and counted 

for 24Na activity in the same manner as the second set of measurements discussed above. The blood 

sodium activation in the burro was determined to be 2.9 × 10−4 μCi/mL of whole blood from a first 

collision dose of 48 rad from fast neutrons (Hurst, Ritchie, and Emerson 1959). Table 5-1 provides a 

summary of the first collision doses from fast neutrons and gamma rays for the mostly highly exposed 

individuals (i.e., Employees A to H in Table 3-1). The last column of Table 5-1 lists the doses of 

record for these eight individuals (UCNC 1958).

The first collision absorbed doses from fast neutrons and gamma rays in Table 5-1 were calculated 

using the following relationships: 

48 rads, and ( ) 2.8 ( ) ( )

( ) ( ) D H D H

S B

S H D H γ n

wb

wb n = × = × Eq. 5-2

where Swb(B) and Swb(H) are the activities of 24Na per mL of whole blood from the burro and human, 

respectively; 48 rad is the first collision neutron dose to the burro; Dn(H) and Dγ(H) are the first 

collision doses to the human from neutrons and gamma rays, respectively; and 2.8 is the gamma-toneutron dose ratio for the exposures. A previous review of the dosimetry for the accident suggested 

that the estimated first collision doses based on blood sodium activation should be increased by 

approximately 10% (Mole 1984). Because more than 90% of the blood sodium is contained in the 

blood serum, the standard practice today is to use the sodium activation in blood serum rather than 

the sodium activation in whole blood (Kerr and Mei 1993). The haematocrit, or the proportion of a
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 13 of 21

Table 5-1. Doses of record for the eight most highly exposed individuals.

a

Exposed

individuals

Blood sodium 

activation (μCi/mL)

First collision adsorbed dose (rad) First collision dose 

Neutrons Gamma rays Total equivalent (rem)

A 5.8E-4 96 269 365 461

B 4.3E-4 71 199 270 341

C 5.4E-4 89 250 339 428

D 5.2E-4 86 241 327 413

E 3.7E-4 62 174 236 298

F 1.1E-4 18 50.5 68.5 86.5

G 1.2E-4 18 50.5 68.5 86.5

H 3.6E-5 6.0 16.8 22.8 28.8

Burro 2.9E-4 48 N/Ab N/A N/A

a. See UCNC (1958). The first collision dose equivalent assumes a relative biological effectiveness (RBE) of 2 for the 

neutron component of first collision absorbed dose based on deterministic effects related to lethality (Langham 1967; 

ICRP 1990). The estimated uncertainty in both the neutron and gamma-ray components of absorbed dose for these 

eight individuals is 20% (Hurst, Ritchie, and Emerson 1959).

b. N/A = not applicable.

blood sample by volume that consists of red blood cells, was 40% to 47% for the eight most highly 

exposed workers and 36.5% for the burro (Brucer 1958). Therefore, the amount of blood serum in the 

sample from the burro was larger by about 10% than the amount of serum in the human samples, and 

the recorded dose estimates for all exposed individuals at the accident have been increased by 10% 

in this report as recommend in the 1984 review by Mole (see Table 5-2). 

Table 5-2. Correction to recorded doses for the eight most highly exposed individuals.

a

Exposed

individuals

First collision neutron dose (rad) First collision gamma-ray dose (rad)

UCNC (1958) Mole (1984) UCNC (1958) Mole (1984)

A 96 106 269 296

B 71 78 199 219

C 89 98 250 275

D 86 95 241 265

E 62 68 174 191

F 18 20 50.5 55.6

G 18 20 50.5 55.6

H 6.0 6.6 16.8 18.5

a. The use of sodium activation in whole blood rather than the current standard use of sodium activation 

in blood serum caused the recorded neutron and gamma-ray doses for the eight most highly exposed 

individuals at the Y-12 1958 criticality accident (UCNC 1958) to be underestimated by approximately 

10% (Mole 1984). 

6.0 EXTERNAL DOSE ESTIMATES FOR OTHER INDIVIDUALS

The distances in Table 3-1 were used to estimate the first collision absorbed doses to other 

individuals near the accident. These are the distances from the center of the reactant solution in the 

barrel to individuals’ beltlines at the time of first criticality (UCNC 1958). Therefore, the radiation 

doses are based on the neutron dose from fissions occurring during the first pulse and gamma-ray 

dose from the prompt fission gamma rays, neutron capture gamma rays, and fission product gamma 

rays emitted during the first 15 s after the first pulse. This is consistent with the estimation of radiation 

doses in the previous section for the eight most highly exposed individuals (UCNC 1958; Hurst, 

Ritchie, and Emerson 1959). The 15 s represents the time taken for most individuals to exit from 

Building 9212 after the evacuation alarms sounded. The radiation doses for the two individuals who 

were 25 ft from the accident are well established from blood sodium activation (i.e., Employees F and 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 14 of 21

G), so their doses were used to estimate radiation doses to other individuals at larger distances by the 

application of 1/r

2 scaling. The relationships used in this scaling are:

[25 ft/ ] 20 rad, and [25 ft/ ] 55.6 rad 2

γ

2 Dn = r × D = r × Eq. 6-1

where Dn, Dγ, and r are the first collision neutron dose, first collision gamma-ray dose, and exposure 

distance in Table 3-1, respectively, for each individual of interest. Table 6-1 summarizes the 

estimated neutron and gamma-ray doses for other individuals near the accident, and Figure 6-1 

shows the estimated gamma-ray doses in comparison to the available film badge data for five of the 

employees (see Tables 3-1 and 6-1). 

Table 6-1. Estimated doses for other individuals near the accident.

Exposed

individual

Distance from 

criticality (ft)

First collision

neutron dose (rad)

First collision

gamma-ray dose (rad)

8 29 14.9 41.3

10 71 2.48 6.89

11 85 1.73 4.81

12 92 1.48 4.11

13 100 1.25 3.48

14 137 0.666 1.85

15 148 0.571 1.59

16 175 0.408 1.13

17 178 0.395 1.10

18 183 0.373 1.04

19 190 0.346 0.963

20 212 0.278 0.773

21 216 0.268 0.745

22 224 0.249 0.693

23 260 0.185 0.514

24 276 0.164 0.456

25 280 0.159 0.443

26 288 0.151 0.419

27 304 0.135 0.376

28 312 0.128 0.357

29 340 0.108 0.301

30 390 0.0822 0.228

31 412 0.0736 0.205

The buildup and attenuation of the in-air neutron and gamma-ray fields should be offsetting factors at 

the distances of interest here, but there could have been significant shielding of some individuals near 

the accident due to the equipment in the building and the building’s walls. For example, four of the 

gamma-ray doses for individuals wearing film badges showed a potential for significant shielding 

against radiation from the accident (i.e., Employees 17, 20, 25, and 29), while the film badge data for 

Employee 31 was in reasonably close agreement with the estimated gamma-ray doses (Figure 6-1). 

For dose reconstruction purposes, the estimated neutron and gamma-ray doses for Employee 31 in 

Table 6-1 should be used as a default value for other individuals who (1) were in or near Building 

9212 at the time of the accident but (2) whose names and social security numbers are not among the 

31 names and social security numbers listed in ORAUT (2006a). 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 15 of 21

Figure 6-1. Comparison of estimated and measured gamma-ray doses for 

other individuals near the accident (see Tables 3-1 and 6-1).

7.0 EXTERNAL DOSE MODIFICATION FOR USE IN IREP

When historical radiation measurements are used in dose reconstruction, the factors used to account 

for the greater relative biological effectiveness of neutrons compared to gamma rays must be 

removed such that absorbed dose is the fundamental unit and the radiation weighting factors wR from 

the International Commission on Radiological Protection’s Publication 60 (ICRP 1991) must be 

applied before the conversion to organ dose for calculation of the probability of causation of a specific 

cancer using the Interactive RadioEpidemiological Program (IREP) (NIOSH 2002; ORAUT 2006b). 

The emphasis in this TIB has been on the estimation of absorbed doses to individuals near the 

accident. Therefore, it is only necessary to apply the weighting factors to the absorbed doses in 

Tables 5-2 and 6-1 of this report.

IREP uses three gamma-ray energy ranges and five neutron energy ranges to calculate the organ 

dose of interest. The three gamma-ray energy ranges are less than 30 keV, 30 to 250 keV, and more 

than 250 keV; the five neutron energy ranges are less than 10 keV, 10 to 100 keV, 0.1 to 2 MeV, 2 to 

20 MeV, and more than 20 MeV (NIOSH 2002), which are the same as the ones used by ICRP to 

define the wR values for neutrons (ICRP 1991). Table 7-1 lists the calculated neutron fluence and 

absorbed dose distributions published in Table K.11 of Y-12 Plant (1958). Note that distribution data 

from the report are given in terms of six neutron energy ranges from thermal to 10 MeV. As a result, 

the estimated percent of absorbed dose for the IREP neutron intervals was calculated by regrouping 

the energy internals in Table 7-1. 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 16 of 21

Table 7-1. Theoretical neutron energy distribution (UCNC 1958).

Neutron energy range

Percent total 

neutrons, Φ (n/cm2)

Percent absorbed 

dose, Dn (rad)

Thermal (0–0.04 eV) 35.8 13.1

0.04 eV–5 keV 9.2 0.4

5–750 keV 14.0 8.0

0.75–1.5 MeV 8.4 11.5

1.5–2.5 MeV 10.9 18.0

2.5–10 MeV 21.8 49.0

For the first IREP neutron energy interval (less than 10 keV), all of the first two energy intervals in 

Table 7-1 were combined and the total percentage of the absorbed dose at energies greater than 

10 keV was 13.5%. Because the second IREP neutron energy group (10 to 100 keV) is relatively 

small in comparison to the 5- to 750-keV neutron energy range in Table 7-1, the absorbed dose was 

assumed to be evenly distributed across this larger neutron energy group or approximately 1% per 

100 keV. This corresponds to a 1% contribution to the absorbed dose for the second IREP neutron 

energy group of 10 to 100 keV. The remaining 7% of the absorbed dose from the 5- to 750-keV 

interval was combined with the 11.5% from the 0.75- to 1.5-MeV energy range and 9% (or half) of the 

1.5- to 2.5-MeV energy range for a total of 27.5% in the third IREP neutron energy group of 0.1 to 2 

MeV. The remaining 9% in the 1.5- to 2.5-MeV energy range was then combined with the 49% from 

the 2.5- to 10-MeV energy range for a total of 58% in the fourth IREP neutron energy group from 2 to 

20 MeV. There was no absorbed dose specified for the fifth IREP neutron energy group more than 20 

MeV. 

Table 7-2 lists the sample dose equivalents for each IREP neutron energy range calculated from the 

neutron absorbed dose of 6.6 rad for Employee H (9) (see Tables 3-1and 5-2), the ICRP 

Publication 60 (ICRP 1991) radiation weighting factors, and the dose fractions for the neutron energy 

ranges. The calculated total neutron dose equivalent of 79.7 rem in Table 7-2 is significantly greater 

than the recorded neutron dose equivalent of 12 rem for Employee H. However, the recorded neutron 

dose equivalent of 12 rem is based on a neutron RBE of 2 for deterministic effects related to lethality 

(Langham 1967; ICRP 1990), while the calculated neutron dose equivalent in Table 7-2 is based on 

the Publication 60 weighting factors for stochastic effects such as cancer that are of concern at the 

low doses and dose rates normally encountered in radiation protection (ICRP 1991, 2003). The 

Publication 60 radiation weighting factors yield dose equivalent estimates for both neutrons and 

gamma rays that are extremely claimant-favorable at the large dose rates and doses involved in the 

accident (i.e., doses from neutrons were delivered in a few seconds and the doses from gamma rays 

were delivered over about 15 s).

Table 7-2. Sample calculation of the neutron dose equivalent for Employee H (9).

IREP neutron 

energy range

ICRP 60 weighting 

factor, wR

Dose

fraction (%)

Estimated absorbed 

dose, Dn (rad)

Estimated dose

equivalent, DEn (rem)

<10 keV 5 13.5 6.6 4.46

10–100 keV 10 1.0 6.6 0.66

0.1–2 MeV 20 27.5 6.6 36.30

2–20 MeV 10 58.0 6.6 38.28

>20 MeV 5 0 6.6 0

Total 79.7

Table 7-3 lists the dose equivalent estimates for both neutrons and gamma rays for reconstruction 

of organ doses for the cancer of interest. The neutron dose equivalents are broken down into the 

neutron energy groups used in IREP, and the gamma-ray dose equivalent is based on a radiation 

weighting factor of 1 for gamma rays of all energies (ICRP 1991). For individuals located at 50 feet 
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 17 of 21

or less from the criticality (see Table 3-1), the gamma rays are lumped together in the one IREP 

energy group of more than 250 keV because of the very high energies of the prompt gamma rays 

from fission, the capture gamma rays from fission neutron interactions, and the delayed gamma rays

from the highly radioactive fission products (IAEA 1982). At larger distances, it is expected that there

would be scattered gamma rays in the 30- to 250-keV energy range due to shielding of an individual 

by walls and other equipment in the building. Thus, 25% of the gamma rays at distances of more than 

50 feet from the criticality are assumed to be in the 30- to 250-keV energy group and 75% in the more 

than 250 keV energy group. Because of the way the IREP computer program treats gamma rays with 

energies of 30-250 keV, this would be more favorable to the claimant than the assumption of 100% 

gamma rays with energies of more than 250 keV. As recommended previously, the dose equivalent 

estimates for Employee 31 should be used as claimant favorable dose estimates for other employees 

who were in or near Building 9212 at the time of the criticality accident but are not among the 31 

individuals listed in ORAUT (2006a).

Table 7-3. Dose equivalent estimates for all individuals near the accident.

Exposed

individual

Neutron dose equivalent, DEn (rem) Gamma dose 

<10 keV 10–100 keV 0.1–2 MeV 2–20 MeV Total equivalent, DEγ (rem)

1 71.6 10.6 583.0 614.8 1,280 296

2 52.7 7.8 429.0 452.4 942 219

3 66.2 9.8 539.0 568.4 1,183 275

4 64.1 9.5 522.5 551.0 1,147 265

5 45.9 6.8 374.0 394.4 821 191

6 13.5 2.0 110.0 116.0 242 55.6

7 13.5 2.0 110.0 116.0 242 55.6

8 10.1 1.5 82.0 86.4 180 41.3

9 4.46 0.66 36.30 38.28 79.7 18.5

10 1.67 0.25 13.64 14.38 29.9 6.89

11 1.17 0.17 9.52 10.03 20.9 4.81

12 1.00 0.15 8.14 8.58 17.8 4.11

13 0.84 0.12 6.88 7.25 15.1 3.48

14 0.450 0.067 3.663 3.863 8.04 1.85

15 0.385 0.057 3.141 3.312 6.89 1.59

16 0.275 0.041 2.244 2.366 4.93 1.13

17 0.267 0.040 2.173 2.291 4.77 1.10

18 0.252 0.037 2.052 2.163 4.5 1.04

19 0.234 0.035 1.903 2.007 4.18 0.963

20 0.188 0.028 1.529 1.612 3.36 0.773

21 0.181 0.027 1.474 1.554 3.24 0.745

22 0.168 0.025 1.370 1.444 3.01 0.693

23 0.125 0.019 1.018 1.073 2.23 0.514

24 0.111 0.016 0.902 0.951 1.98 0.456

25 0.107 0.016 0.875 0.922 1.92 0.443

26 0.102 0.015 0.831 0.876 1.82 0.419

27 0.091 0.014 0.743 0.783 1.63 0.376

28 0.086 0.013 0.704 0.742 1.55 0.357

29 0.073 0.011 0.594 0.626 1.30 0.301

30 0.0555 0.0082 0.4521 0.4768 0.993 0.228

31 0.0497 0.0074 0.4048 0.4269 0.889 0.205
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 18 of 21

8.0 SUMMARY

This TIB provides an important update to the radiation dosimetry for the accident (UCNC 1958; Hurst, 

Ritchie, and Emerson 1959). The dose estimates for the eight most highly exposed individuals were 

based on measurements of their blood sodium activation and a mockup of the criticality using a burro 

as a surrogate for man. Measurements of sodium activation in blood from the burro made it possible 

to estimate the first collision absorbed dose from neutron and gamma rays for these eight workers. 

An RBE of 2 based on deterministic effects related to lethality was then used to estimate the first 

collision dose equivalent for the eight workers. The uncertainty of the first collision doses was 

estimated to be approximately 20% (Hurst, Ritchie, and Emerson 1959).

A 1984 review of the radiation dosimetry for the accident suggested that the estimated first collision 

doses based on blood sodium activation should be increased by approximately 10% (Mole 1984). 

Because more than 90% of the blood sodium is contained in the blood serum, the standard practice 

today is to use the sodium activation in blood serum rather than sodium activation in whole blood 

(Kerr and Mei 1993). The haematocrit or the proportion of a blood sample by volume that consists of 

red blood cells was 40% to 47% for the eight most highly exposed workers and 36.5% for the burro 

(Brucer 1958). Therefore, the amount of blood serum in the sample from the burro was larger by 

about 10% than the amount of serum in the human blood samples, and the recorded doses for all 

eight of the most highly exposed workers were corrected to account for the 10% increase 

recommended in the 1984 review by Mole (1984). The estimated uncertainties of 10% by Mole and 

20% by Hurst, Ritchie, and Emerson (1958) suggest an overall uncertainty of approximately 25% in 

the estimated dose to workers from the accident. 

Information on distances of workers from the accident at the time of exposure was available for 

another 23 workers (Y-12 Plant 1958). The first collision doses from neutrons and gamma rays for

these workers were estimated using 1/r

2 scaling of the radiation fields and normalized to the first 

collision doses for two of the eight highly exposed workers who were 25 ft from the accident. The 

dose estimates were then compared with available film badge data for 5 of the 23 workers. Four of 

the gamma-ray doses for workers wearing film badges showed a significant potential for shielding 

against radiation from the accident, while the film badge data for the individual at the greatest distance 

from the criticality was found to be in reasonably close agreement with the estimated gamma-ray 

dose. Therefore, the estimated gamma-ray and neutron doses for this individual should be used as 

the default dose values for any other workers in or near Building 9212 at the time of the accident. 

To obtain the first collision dose equivalent as required by the IREP computer program (NIOSH 2002), 

ICRP Publication 60 radiation weighting factors were applied (ICRP 1991). The dose equivalent 

estimates for the eight most highly exposed workers are quite large in comparison to the recorded 

dose equivalents in the literature (UCNC 1958; Hurst, Ritchie, and Emerson 1959) because the 

recorded dose equivalents are based on an RBE of 2 for deterministic effects in relation to lethality 

(Langham 1967; ICRP 1990), while the dose equivalent estimates calculated here are based on 

radiation weighting factors for stochastic effects such as cancer that are of concern at the low doses 

and dose rates normally encountered in routine radiation protection (ICRP 1991, 2003). The 

Publication 60 radiation weighting factors yield dose equivalent estimates for both neutrons and 

gamma rays (Table 7-3) that are extremely claimant-favorable at the large dose rates and doses 

involved in the accident.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 19 of 21

REFERENCES

Auxier, J. A., F. W. Sanders, and W. S. Snyder, 1961, “24Na Activation in the Dosimetry of Nuclear 

Accidents,” Radioactivity in Man, G. R. Meneely, editor, Charles C. Thomas, Springfield, 

Illinois, pp. 201–218.

Ballweg, L. H., and J. L. Meem, 1951, A Standard Gamma-Ray Ionization Chamber for Shielding 

Measurements, ORNL-1028, Oak Ridge National Laboratory, Oak Ridge, Tennessee.

Brucer, M., editor, 1958, A Medical Report of the Y-12 Accident, ORINS-25, Oak Ridge Institute of 

Nuclear Studies, Oak Ridge, Tennessee.

BWXT Y-12, 2000, Recycled Uranium Mass Balance Project Y-12 National Security Complex Site 

Report, Y/LB-16,036, Rev. 1, Y-12 National Security Complex, Oak Ridge, Tennessee, 

December.

Callihan, D., and J. T. Thomas, 1959, “Accidental Radiation Excursion at the Oak Ridge Y-12 Plant—

I, Description and Physics of the Accident,” Health Physics, volume 1, number 4, pp. 363–372.

Cross, W. G., 1981, “Neutron Activation of Na in Phantoms and the Human Body,” Health Physics, 

volume 41, pp. 135–153.

DOE (U.S. Department of Energy), 2004, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2004, Washington, D.C., December.

Hurst, G. S., and R. H. Ritchie, 1962, “Applications to Radiation Dosimetry” Nuclear Instruments and 

Their Uses, A. H. Snell, editor, John Wiley & Sons, New York, New York, pp. 367–390. 

Hurst, G. S., and R. H. Ritchie, editors, 1961, Radiation Accidents: Dosimetric Aspects of Neutron 

and Gamma-Ray Exposure, ORNL-2748, Part A, 2nd Issue, Union Carbide Corporation, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee.

Hurst, G. S., R. H. Ritchie, and L. C. Emerson, 1959, “Accidental Radiation Excursion at the Oak 

Ridge Y-12 Plant—III, Determination of Radiation Doses,” Health Physics, volume 2, pp. 121–

133.

IAEA (International Atomic Energy Agency), 1982, Dosimetry for Criticality Accidents: A Manual, 

Technical Reports Series No. 211, Vienna, Austria, March.

ICRP (International Commission on Radiological Protection), 1990, RBE for Deterministic Effects, 

Publication 58, Annals of the ICRP, Volume 20, Number 4, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 2003, Relative Biological Effectiveness 

(RBE), Quality Factor (Q), and Radiation Weighting Factor (wR), Publication 92, Pergamon 

Press, Oxford, England.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 20 of 21

Kerr, G. D. and G. T. Mei, 1993, Technical Basis for Nuclear Accident Dosimetry at the Oak Ridge 

National Laboratory, ORNL/TM-12224, Martin Marietta Energy Systems, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee.

Langham, W. H., editor, 1967, “Radiation Effects in Man: Early Effects,” Radiobiological Factors in 

Manned Space Flight, National Research Council, National Academy Press, Washington, 

D.C., pp. 59–133.

Martin, J. E., 2000, Physics for Radiation Protection, John Wiley & Sons, New York, New York.

McLendon, J. D., 1959, “Accidental Radiation Excursion at the Oak Ridge Y-12 Plant—II. Health 

Physics Aspects of the Accident,” Health Physics, volume 2, number 1, pp. 21–29.

McRee, P. C., C. M. West, and J. D. McLendon, editors, 1965, Y-12 Radiation Safety Manual, Y1401–Revised, Union Carbide Corporation, Nuclear Division, Oak Ridge, Tennessee, May 11.

Mole, R. H., 1984, “Sodium in Man and the Assessment of Radiation Dose After Critical Accidents,” 

Physics in Medicine and Biology, volume 29, pp. 1307–1327.

NIOSH (National Institute for Occupational Safety and Health), 2002, NIOSH-Interactive 

RadioEpidemiological Program (NIOSH-IREP) Technical Documentation, Final Report, Office 

of Compensation Analysis and Support, Cincinnati, Ohio.

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Y-12 

National Security Complex – Site Description, ORAUT-TKBS-0014-2, Rev. 00 PC-1, Oak 

Ridge, Tennessee, October 11.

ORAUT (Oak Ridge Associated Universities Team), 2005a, Historical Evaluation of the Film Badge 

Program at Y-12 Facility in Oak Ridge, Tennessee: Part 1 – Gamma Radiation, ORAUTRPRT-0032, Rev. 00, Oak Ridge, Tennessee, April 13.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Historical Evaluation of the Film Badge 

Program at Y-12 Facility in Oak Ridge, Tennessee: Part 2 – Neutron Radiation, ORAUTRPRT-0033, Rev. 00, Oak Ridge, Tennessee, April 13.

ORAUT (Oak Ridge Associated Universities Team), 2005c, Estimation of Neutron Dose Rates from 

Alpha-Neutron Reactions in Uranium and Thorium Compounds, ORAUT-OTIB-0024, Rev. 00, 

Oak Ridge, Tennessee, April 7.

ORAUT (Oak Ridge Associated Universities Team), 2006a, Y-12 1958 Criticality Accident Roster, 

Rev 00 (Official Use Only), Oak Ridge, Tennessee.

ORAUT (Oak Ridge Associated Universities Team), 2006b, Technical Basis for Conversion from 

NCRP Report 38 Neutron Quality Factors to ICRP Publication 60 Radiation Weighting Factors 

for Respective IREP Input Neutron Energy Ranges, ORAUT-OTIB-0055, Rev. 00-A, Oak 

Ridge, Tennessee.

Shleien, B., L. A. Slaback, Jr., and B. K. Birky, editors, 1998, Handbook of Health Physics and 

Radiological Health, Third Edition, Williams & Wilkins, Baltimore, Maryland, January.
Document No. ORAUT-OTIB-0057 Revision No. 00 Effective Date: 05/15/2006 Page 21 of 21

Thornton, W. T., D. M. Davis, and E. D. Gupton, 1961, The ORNL Badge Dosimeter and Its Personnel 

Monitoring Applications, ORNL-3126, Union Carbide Corporation, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee.

UCNC (Union Carbide Nuclear Company), 1958, Accidental Radiation Excursion at the Y-12 Plant, 

June 16, 1958, Final Report, Y-1234, Y-12 Plant, Oak Ridge, Tennessee.

Y-12 Plant, 1958, Accident of June 16, 1958, Outline for Interim Report, (Official Use Only - NIOSH 

OR 03-066), Oak Ridge, Tennessee, June 20.