Page 1 of 24

DOE Review Release 04/08/2019

Neutron Dose from Highly Enriched Uranium

ORAUT-RPRT-0060 Rev. 00

Effective Date: 03/28/2019

Supersedes: None

Subject Expert(s): Matthew H. Smith, Elizabeth K. Algutifan, Michalene Rodriguez, Thomas R. 

LaBone, and Keith A. McCartney

Document Owner 

Approval: Signature on File

Matthew H. Smith, Document Owner

Approval Date: 03/25/2019

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 03/25/2019

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 03/25/2019

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 03/25/2019

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 03/28/2019

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 2 of 24

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

03/28/2019 00 New document initiated for evaluation of neutron dose from highly 

enriched uranium at the Oak Ridge Gaseous Diffusion Plant (K-25), 

Paducah Gaseous Diffusion Plant, Portsmouth Gaseous Diffusion 

Plant, and Y-12 Plant. Incorporates formal internal and NIOSH 

review comments. Training required: As determined by the 

Objective Manager. Initiated by Matthew H. Smith.
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 3 of 24

 

TABLE OF CONTENTS

SECTION TITLE PAGE

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

1.0 Introduction .............................................................................................................................. 7

2.0 Background.............................................................................................................................. 7

3.0 Neutron Dose Data from Facilities............................................................................................ 9

3.1 Gaseous Diffusion Plants ............................................................................................. 9

3.1.1 Paducah Gaseous Diffusion Plant.................................................................. 9

3.1.2 Portsmouth Gaseous Diffusion Plant.............................................................. 9

3.1.3 K-25 Gaseous Diffusion Plant ...................................................................... 11

3.2 Y-12 Plant................................................................................................................... 14

4.0 Additional Data Analysis of Employee Data Measurements for K-25, Portsmouth, 

and Y-12 Plant ....................................................................................................................... 15

5.0 Photon Summary and Conclusions ........................................................................................ 18

6.0 Attributions and Annotations .................................................................................................. 20

References ........................................................................................................................................ 21

ATTACHMENT A QUANTILE REGRESSION ANALYSIS IMPLEMENTATION WITHIN 

IREP ...............................................................................................................23

LIST OF TABLES

TABLE TITLE PAGE

2-1 Composition of various uranium materials and two specific VHEU materials from the 

SRS and ICPP ......................................................................................................................... 7

2-2 Neutron production in various uranium isotopes and materials ................................................ 8

2-3 Neutron dose fractions for HEU storage areas......................................................................... 9

3-1 Portsmouth employee dosimetry data, 1992 to 2013.............................................................. 11

3-2 K-25 annual photon/beta and neutron badge exchanges from 1989 to 2012.......................... 13

3-3 Summary of Y-12 Plant N:P ratio data, 1992 to 2004............................................................. 15

4-1 Summary of quantile regression fit parameters for K-25, Portsmouth, and Y-12 Plant 

employee data measurements ............................................................................................... 15

5-1 Summary of neutron dose rates, photon dose rates, and N:P ratios associated with 

natural, depleted, and enriched uranium materials at Paducah, Portsmouth, and K-25 

Gaseous Diffusion Plants and Y-12 Plant............................................................................... 19

5-2 Summary of N:P ratios and quantile regression relationships associated with natural, 

depleted, and enriched uranium materials at Portsmouth, K-25, and Y-12 Plants, based 

on personnel dosimetry.......................................................................................................... 20
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LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Quantile regression results for K-25 employee dosimetry data (1989-2012)........................... 16

4-2 Quantile regression results for Portsmouth employee dosimetry data, production era 

(1992–2001)...........................................................................................................................16

4-3 Quantile regression results for Y-12 Plant employee dosimetry data (1955-1959).................. 17

4-4 Quantile regression results for the combined K-25, Y-12, and Portsmouth employee 

dosimetry data sets................................................................................................................17

A-1 Lognormal distribution fit of calculated neutron data based on photon sample and N:P 

ratio sample. ..........................................................................................................................24
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ACRONYMS AND ABBREVIATIONS

cm centimeter

DOE U.S. Department of Energy

DU depleted uranium

ETTP East Tennessee Technology Park

EU enriched uranium

ft feet

g gram

GDP gaseous diffusion plant

HEU highly enriched uranium

hr hour

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

LEU low-enriched uranium

LOD limit of detection

m meter

MeV megaelectron-volt, 1 million electron-volts

mR milliroentgen

mrem millirem

n neutron

NFS Nuclear Fuel Services

NIOSH National Institute for Occupational Safety and Health

N:P neutron-to-photon ratio

NU natural uranium

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

ppb parts per billion

PW product withdrawal

RU recycled uranium

s second

SRDB Ref ID Site Research Database reference identification (number)

SRS Savannah River Site

t ton

TLD thermoluminescent dosimeter

USEC United States Enrichment Corporation
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VHEU very highly enriched uranium

WGU weapons-grade uranium

wt % weight percent

α alpha particle
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1.0 INTRODUCTION

Neutron dose to workers at gaseous diffusion and uranium metals plants has been an issue of 

concern because some workers were unmonitored or monitored for exposure to photon and beta 

radiation only. Information is provided on this topic in the Guide of Good Practices for Occupational 

Radiological Protection in Uranium Facilities (DOE 2009), but the basis for that information is limited. 

Therefore, this report focuses on measurements and calculations of the neutron dose and neutron-tophoton (N:P) ratio from various uranium compounds with an emphasis on exposures to highly 

enriched uranium (HEU) compounds at gaseous diffusion plants and at uranium metal processing 

facilities, such as the Y-12 Plant in Oak Ridge, Tennessee.

2.0 BACKGROUND

From 1944 to 1952, all uranium used by the Manhattan Engineer District and the U.S. Atomic Energy 

Commission [(AEC) U.S. Department of Energy (DOE) predecessor agencies] was derived from 

natural uranium (NU) sources (DOE 2003, p. 29). NU contains the following three isotopes: 238U, 

235U, and 234U. Uranium that has been processed to increase the weight concentration of 235U is 

referred to as enriched uranium (EU). The extent of the enrichment depends on the intended use of 

the uranium (DOE 2009). Commercial light-water reactors are designed for use with the 235U enriched 

from 3% to 5% by weight, or so-called low-enriched uranium (LEU). HEU is required for use in hightemperature gas-cooled reactors, naval nuclear propulsion reactors, most research reactors, and 

weapons (DOE 2009, p. 19). Weapons-grade uranium (WGU) is produced using separation 

techniques that increase the weight concentration of 235U to more than 90% by weight (NAS 2005). 

For example, the concentrations by weight of various uranium isotopes in materials from NU sources 

and from recycled very highly enriched uranium (VHEU) sources such as the Savannah River Site 

(SRS) and the Idaho Chemical Processing Plant (ICPP) are provided in Table 2-1.

Table 2-1. Composition of various uranium materials and two specific VHEU materials from the SRS 

and ICPP.

a 

Material

U-233 

(wt %)

U-234

(wt %)

U-235

(wt %)

U-236

(wt %)

U-238

(wt %)

Non-U:U α 

emission 

rate (%)

NUa <0.001 0.0057 0.7193 <0.001 99.275 ≈0

LEUa <0.001 0.025 3.5 <0.001 96.475 ≈0

HEUa <0.01 0.12 94.0 <0.01 5.88 ≈0

High burn-up recycled VHEU from SRSb 0.01 1.28 52.2 29.2 17.3 0.04

Lower burn-up recycled VHEU from ICPPb 0.0003 1.0 78.0 20.0 1.7 0.08

Recycled VHEU (WGU)b,c <0.01 1.1 93 0.75 6 0.21

Recycled NUb <0.0001 0.0054 0.72 <0.0001 99.27 0.008

Recycled LEUb <0.01 0.02 2.0 <0.01 97.98 0.04

Recycled DUb 0.001 0.2 <0.0001 0.0003 99.8 0.018

a. Uranium composition from NAS (2005, Table A-2).

b. Recycled uranium (RU) composition from ORAUT (2012, Table 5-7). 

c. WGU is produced using separation techniques to increase the concentration of U-235 to more than 90 wt % (NAS 

2005).

In 1952, the AEC started reprocessing uranium from weapon stockpiles and reactors (DOE 2003). 

This reprocessed uranium is commonly referred to as recycled uranium (RU). Uranium that has been 

irradiated in reactors contains transuranic elements (such as plutonium and 237Np), fission products 

(such as 99Tc), and reactor-generated products (such as 236U). After chemical processing to separate 

and extract plutonium, as well as to recover uranium for reuse, trace quantities of plutonium, 

neptunium, technetium, and 236U remain in the RU. These constituents make the RU more 

radioactive than NU. Thus, the handling, reprocessing, and reenrichment of RU may present a 

greater potential for personnel exposure than that normally associated with the processing of 
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unirradiated uranium (BWXT 2000). In the early days of RU processing, the Y-12 Plant set an 

acceptance limit of 10 ppb for the plutonium concentration in RU being sent to the site for subsequent 

processing. This limit has continued in use for the duration of the RU production (DOE 2003). 

Shipments of RU with plutonium concentrations above this limit required site permission before 

shipment. Due to these requirements, nearly all RU coming from these sites contained less than 10 

ppb of plutonium (DOE 2003).

The RU process streams involved the processing of uranium metals, uranium alloys, chemical 

compounds (such as UF4 and UF6), uranium oxides (UO2, UO3, U3O8), uranium chloride (UCl4), and 

sodium diuranate (Na2U2O7) (ORAUT 2005, p. 3). Comparisons of neutron production rates due to 

spontaneous fission and (α,n) reactions in various isotopes and two chemical compounds of uranium 

are provided in Table 2-2. The interaction of alpha particles from uranium with nuclei of fluorine, 

oxygen, and other low-atomic weight atoms generates neutrons with energies of approximately 2 MeV 

(DOE 2009). The magnitude of the neutron flux varies based on total activity of the uranium (a 

function of enrichment) and the chemical compound (combining of uranium with fluorine or oxygen). 

In the case of UF6, the typically measured neutron dose equivalent rates for storage containers are as 

follows (DOE 2009):

NU to 5% EU: 0.01 to 0.2 mrem/hr

VHEU (97%+): 2 to 4 mrem/hr (contact)

1 to 2 mrem/hr (3 ft or 1 m). 

Table 2-2. Neutron production in various uranium isotopes and materials. 

Uranium isotope or 

material

Spontaneous

fission yield 

(n/s-g)

Alpha

particle yield 

(α/s-g)

(α,n) yield

in oxides 

(n/s-g)

(α,n) yield

in fluorides 

(n/s-g)

U-232a 1.3 8.0 × 1011 1.49 × 104 2.6 × 106

U-233a 8.6 × 10-4 3.5 × 108 4.8 7.0 × 102

U-234a 5.02 × 10-3 2.3 × 108 3.0 5.8 × 102

U-235a 2.99 × 10-4 7.9 × 104 7.1 × 10-4 8 × 10-2

U-236a 5.49 × 10-3 2.3 × 106 2.4 × 10-2 2.9

U-238a 1.36 × 10-2 1.2 × 104 8.3 × 10-5 2.8 × 10-2

NUb 5.45 × 10-3 3.07 × 104 3.04 × 10-4 6.66 × 10-2

LEUb 5.31 × 10-3 7.18 × 104 8.55 × 10-4 1.75 × 10-1

HEUb 7.69 × 10-2 3.51 × 105 4.27 × 10-3 7.73 × 10-1

High burn-up VHEUc 4.18 × 10-3 3.69 × 106 4.63 × 10-2 7.43

Lower burn-up VHEUc 1.61 × 10-3 2.83 × 106 3.52 × 10-2 6.44

WGU recycled VHEUc 1.19 × 10-3 2.62 × 106 3.38 × 10-2 6.48

Recycled NUc 1.35 × 10-2 2.44 × 104 2.38 × 10-4 5.74 × 10-2

Recycled LEUc 1.33 × 10-2 5.97 × 104 6.96 × 10-4 1.45 × 10-1

Recycled DUc 1.36 × 10-2 1.48 × 104 1.14 × 10-4 3.39 × 10-2

a. Spontaneous fission neutron yields and (α,n) reactions in oxides and fluorides from Tables 6-4 and 6-5 of DOE (2009). 

b. See Table 2-1 of this report and uranium compositions (wt %) from Table A-2 of NAS (2005). 

c. See Table 2-1 of this report and RU compositions (wt %) from Table 5-7 of ORAUT (2012). 

The potential for significant worker exposures to neutrons generated by (α,n) reactions in uranium 

compounds is not very high unless a worker spends a large amount of time near uranium fluoride or

uranium oxide compounds storage containers, or in processing areas for large quantities of those 

materials (DOE 2009). At very high 235U enrichments, the N:P dose ratio can be as much as 2:1 and 

neutrons can be the limiting radiation source for whole-body exposure (DOE 2009). As stated 

previously, there is a significantly limited basis for this information from DOE (2009). The N:P ratio of 

2:1 should be considered an upper bound of the potential range of ratios seen in facilities handling 
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HEU. The neutron doses from low-enriched 235U compounds or from uranium metals are not limiting 

doses (DOE 2009). 

With respect to neutron energy spectra for HEU, Y-12 National Security Complex – Occupational 

External Dosimetry provides data for HEU storage areas at the Y-12 Plant that is summarized below 

in Table 2-3. 

Table 2-3. Neutron dose fractions for HEU storage areas. 

Neutron energy group Dose fraction

<10 keV 0.012

10–100 keV 0.003

0.1–2 MeV 0.970

2–20 MeV 0.015

Source: ORAUT (2009). 

3.0 NEUTRON DOSE DATA FROM FACILITIES

3.1 GASEOUS DIFFUSION PLANTS

Studies have been published about measured neutron doses and N:P ratios near storage cylinders 

containing any DU, NU, LEU, or HEU in the form of UF6 at the Paducah, Portsmouth, and K-25 

gaseous diffusion plants. In some cases, these studies focused on large (14-t) cylinders in outside 

storage yards, while others were measurements of smaller HEU storage cylinders inside the facilities.

3.1.1 Paducah Gaseous Diffusion Plant

At the Paducah plant, measurements during a UF6 cylinder painting project in the cylinder storage 

yard gave values for neutron-to-photon dose ratios ranging from 0.14 to 0.42 with an approximate 

average of 0.2 (BJC 1999). These cylinders, however, contained either DU or NU.

3.1.2 Portsmouth Gaseous Diffusion Plant

Neutron Dose Rate Data

Cardarelli (1997) quotes values for measurements of the neutron dose near small (5-in.-diameter) 

HEU storage containers that ranged from 3 mrem/hr (contact) to <0.5 mrem/hr (at 1 m). These values 

are in reasonably good agreement with the values in DOE (2009). This report also cites the neutron 

dose component of total external dose as 12.5% based on a monitoring study of uranium material 

handlers in 1995. It is important to note that this study included workers exposed to a variety of 

uranium enrichments.

In 1992, measurements taken in the PW (product withdrawal) vault (Building X-326) at Portsmouth 

indicated a neutron dose rate of 0.8 mrem/hr at a distance of 0.6 to 1.3 m facing a group of HEU 

cylinders (Soldat and Tanner 1992). Another survey was conducted in 1995 of a single 5A cylinder 

(97% HEU) in the X-345 Building. The recorded neutron dose rates at that time were 2.1 mrem/hr 

and 0.77 mrem/hr at distances of 15.2 cm and 30.5 cm, respectively (Scherpelz and Murphy 1995). 

Unfortunately, photon measurements were not made in either of these two studies.

Neutron and photon rate measurements were documented in a 1986 memo for four empty and six full 

5-inch product cylinders stored in the X-326 PW vaults. Surface neutron and photon dose rates for 

the empty cylinders were ≤0.5 mrem/hr and ≤1.0 mR/hr respectively (N:P ratio = 0.5). Neutron and 

photon dose rates for the full cylinders were ≤3.0 mrem/hr and ≤2.4 mR/hr respectively (N:P ratio = 

1.25). At a distance of 1 meter, the combined neutron and photon dose rate was 0.2 mrem/hr for 

empty cylinders and 0.5 mrem/hr for full cylinders (Bassett 1986).
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N:P Ratio Data from Area Monitoring Measurements

A summary memorandum from 2008 (McGuire 2008) provides annual Portsmouth area dosimetry 

data that was collected to meet the requirements of Dosimetry Program Standards (Madder and 

Turner 1997; Dodd, Duncan, and Teeters 2001, 2002). Outdoor annual neutron and photon dose 

measurements for the period from 2004 to 2008 were available for the following radiological facilities: 

X-326, X-330, X-333, X-342, X-343 X-344, X-530, X-705, X-710, and X-745. Data were also provided 

for nonradiological facility areas as well as Perimeter Road near the cylinder yards. Paired positive 

photon and neutron measurements were recorded at the following facilities or areas: X-330, X-344, 

X-343, X-530, X-745, and Perimeter Road near the cylinder yards. The average N:P ratio derived 

from this data was 0.147 ±0.09 with a minimum value of 0.01 and a maximum value of 0.398.

N:P Ratio Data from Employee Dosimetry Measurements, 1992 to 2013

Four databases containing historical dosimetry records for Portsmouth employees were reviewed in 

order to calculate a favorable to claimant N:P ratio. The databases reviewed are as follows: 

• 3_Individual_Dose_Ports (overlaps with REMS_Portsmouth 1992-2013 Exp Recs)

• Extra_REIRS_Ports (overlaps with 3_Individual_Dose_Ports)

• REIRS_Portsmouth 1992-2013 Exp Recs

• REMS_Portsmouth 1992-2013 Exp Recs

The Portsmouth REIRS database was not useful in calculating an N:P ratio because no neutron dose 

records were included. Also, as indicated above, two of the databases were actually subsets of other 

databases. Thus, only the REMS_Portsmouth 1992-2013 database was used. Positive paired 

neutron and photon doses from this database were used to calculate a favorable to claimant N:P ratio 

that can be applied to certain workers in facilities where enriched uranium was handled. It should be 

noted that positive neutron results appeared in the REMS database beginning in 1999 (Bechtel

Jacobs era).

Evaluation of REMS_Portsmouth 1992-2013 Exp Recs Database

A total of 38,964 records are contained in the Portsmouth REMS database. These encompass the 

period 1992 to 2013. A review of these records revealed the following information:

• 10,721 unique individuals were monitored by external dosimetry during this period.

• Of the 10,721 individuals monitored, 722 were also monitored for neutrons (7%) at one time or 

another.

• This database indicates that neutron monitoring began at Portsmouth in 1996, although the 

hardcopy records in the NIOSH Division of Compensation Analysis and Support Claimant 

Tracking System indicate an earlier date: 

– 1987 – sporadic monitoring for a few individuals, and

– Fall 2005 – widespread monitoring.

• It should be noted that United States Enrichment Corporation (USEC) dosimetry records were

included in a separate database beginning in 1997 (REIRS).

• Of the 722 individuals monitored for neutrons, 56 had at least one neutron monitoring result 

that exceeded the neutron limit of detection (LOD) of 10 mrem.
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• Positive neutron results do not appear in the REMS database until 1999. HPRS printouts in

NOCTS, however, indicate positive neutron dose as early as 1987 for one individual (see

[Redacted]).

• USEC personnel wore a combined thermoluminescent dosimeter (TLD).

• DOE personnel wore an albedo neutron dosimeter separate from their regular issued TLD. 

Table 3-1 shows a breakdown of monitored individuals by year, based on a review of the REMS 

database: 

Table 3-1. Portsmouth employee dosimetry data, 1992 to 2013. 

Year

Number of individuals monitored 

for total external dose

Number of individuals 

monitored for neutron dose (ND)

1992 3,628 0

1993 3,589 0

1994 4,404 0

1995 4,485 0

1996 3,957 444 (all ND)

1997a 144 0

1998 176 4 (all ND)

1999 204 36

2000 421 30

2001 639 32

2002 640 39

2003 541 26

2004 650 25

2005 643 64

2006 762 40

2007 731 31

2008 803 33

2009 1,219 95

2010 1,544 144

2011 2,718 142

2012 2,737 184

2013 2,662 147

Source: Portsmouth REMS Database – 1992-2013.

a. Databases split in 1997; REIRS database beginning in 1997 contains all USEC

employees but no neutron monitoring results.

Development of N:P Ratio Based on Portsmouth Dosimetry Data

A total of 161 records in the REMS database indicate neutron dose values greater than the LOD of 

10 mrem. These values were divided by their paired photon dose values to calculate an N:P ratio for 

each line of data. These ratios were then averaged to obtain an overall N:P ratio. If a positive 

neutron dose had a corresponding photon dose of less than its LOD of 10 mrem, the photon dose was 

set equal to 10 mrem. This occurred with 10 records. In addition, a doubling of the radiation 

weighting factor from 10 to 20 was accounted for beginning in 2010. Based on this approach, the 

average N:P ratio was calculated to be 0.369 ±0.2. 

3.1.3 K-25 Gaseous Diffusion Plant

Neutron Dose Rate Data

Discussions of N:P ratio values were found in a variety of documents written to address radiological 

safety issues associated with the K-25 cylinder storage yards. In a May 2000 report, an N:P ratio of 

0.25 was cited with the notation that many of the neutron dose rate measurements were zero (BJC 
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2000). A Final Safety Analysis Report for the K-25 cylinder yards, written in 1997, indicated contact 

dose rate values between 1 to 6 mrem/hr and general area dose rates between 1.5 to 2 mrem/hr. 

The report further indicated that neutron dose was one-sixth of the dose received from photon

radiation (N:P = 0.166) (LMES 1997). Data from an area characterization study done along the 

outside boundary of the K1066E cylinder yard in 2000 indicated an N:P ratio of 0.274. This value was 

based on nine locations that had positive neutron values out of a total of 86 sampling points (DOE 

2000). 

N:P Ratio Data from Modeling Study

A modeling study for the neutron dose and N:P ratio from a typical array of large storage containers 

(14-t cylinders) of DU, NU, and LEU (5% enrichment) at the K-25 Plant can be found in Determination 

of Neutron to Gamma Dose Ratios for the UF6 Cylinder Yards (NISYS 2000). The array of tanks 

consisted of 2 rows of cylinders stacked 2 cylinders high and 25 cylinders long on a 6-in. concrete 

pad. Calculations of the neutron dose and N:P ratio were made at distances ranging from 1 m to 

300 m at the mid-length of the array of tanks. These calculations considered the effects of radiation 

scattering by the 6-in. thick concrete pad and by air with a relative atmospheric humidity of 30%, 60%, 

and 90% (NISYS 2000). For tanks containing DU, NU, and LEU (5% enrichment), at 1 m the N:P

dose ratios were approximately 0.10, 0.22, and 0.88, respectively (NISYS 2000). The calculations at 

larger distances from the tanks containing DU, NU, and LEU (5% enrichment) were used to estimate 

bounding N:P ratios for radiation exposures at the tank yard of approximately 0.2, 0.4, and 1.5, 

respectively (NISYS 2000).

N:P Ratio from Equipment and Area Monitoring Surveys

A summary of K-25 survey data was provided in 2015 by East Tennessee Technology Park [(ETTP) 

previously known as the K-25 Site] personnel. The data cover the period January 1999 through 

November 2006 (at which time the last of the UF6 cylinders were removed). Over 6,000 paired 

neutron and photon survey data points were available that corresponded with equipment and general 

area measurements. Analysis of the equipment survey data yielded N:P ratio values of 0.099, 0.085, 

and 0.122 for distances on contact, 1 foot, and 1 meter respectively. An N:P ratio of 0.605 ±0.408 

was calculated from 445 neutron/photon data points associated with general area measurements 

(Pope 2015).

N:P Ratio Data from Employee Dosimetry Measurements, 1989 to 2012

In the past, dosimetry services for K-25 were provided by either the Oak Ridge National Laboratory 

(ORNL) or the Y-12 Plant. Therefore, to establish an N:P ratio, dosimetry records data for Oak Ridge 

Gaseous Diffusion Plant (K-25), also known as the East Tennessee Technology Park (ETTP), 

employees was extracted based on employee information and collated into a merged copy of the Y-12 

Plant Dosimetry and ORNL Safety and Health Databases (dated February 6, 2013). Although 

integration of the two databases resulted in a version containing a variety of different tables, only 

records extracted from Table 3_Individual_Dose_ETTP (hereafter referred to as Table 3 ETTP) were 

used because the others: (1) did not have neutron dosimetry information, (2) contained duplicate 

data, (3) were limited to extremity dose, or (4) contained only Portsmouth Gaseous Diffusion Plant 

records.

Evaluation of Table 3 ETTP 

Table 3 ETTP initially contained a total of 187,295 records; dates ranged from January 1, 1989, 

through December 31, 2012. Duplicate data were removed, leaving only unique dosimetry records for 

K-25 employees during the period of interest. Also of note:

• Personnel wore the Harshaw four-element beta/photon dosimeter and a separate four-element 

neutron dosimeter. 

• 129,132 unique beta/photon dosimeters remained after duplicates were removed.
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• 16,701 unique neutron dosimeters remained after duplicates were removed.

• 11,599 individuals were monitored for beta/photon radiation from 1989 to 2012.

• 404 individuals were monitored for neutrons from 1989 to 2012.

• Approximately 3.5% of the individuals were monitored for beta/photon and neutron radiation.

• The first positive neutron result occurred in 1993.

• Results from 369 neutron dosimeters were greater than or equal to the limit of detection 

(LOD).

• Neutron monitoring seldom occurred from 1990 to 1996, but seems to have been more 

frequent from 1999 to 2012.

Table 3-2 below tabulates the number of photon/beta and neutron dosimeters exchanged for K-25 

employees by year.

Table 3-2. K-25 annual photon/beta and neutron badge exchanges from 1989 to 2012. 

Year

Photon/beta dosimeter 

exchanges

Neutron dosimeter 

exchanges

Neutron dosimeters (results 

> or = to the limit of detection)

1989 232 0 0

1990 401 7 0

1991 359 5 0

1992 550 2 0

1993 461 4 1

1994 327 3 0

1995 169 6 0

1996 253 12 0

1997 6,005 79 1

1998 6,304 59 0

1999 5,920 183 9

2000 5,980 773 20

2001 7,107 1,138 35

2002 7,002 1,415 11

2003 8,096 1,769 21

2004 10,239 1,923 26

2005 12,515 2,283 20

2006 10,637 2,030 25

2007 9,040 877 60

2008 8,137 959 44

2009 8,419 904 19

2010 8,618 857 54

2011 7,518 852 20

2012 4,843 561 3

Source: K-25_PORTS_Exposure Database. 

A total of 369 records in Table 3 ETTP had individual neutron doses greater than or equal to the LOD 

of 10 mrem. An analysis of dosimeters with paired photon and neutron doses greater than the LOD 

was performed. The N:P ratio was calculated by dividing the neutron dose by the deep dose for each 

badge exchange. A ratio of 0.490 was calculated based on the average of the paired N:P ratios for 

the period January 1, 1989 to December 31, 2012. Starting January 1, 2010, ORNL adopted the 

1990 International Commission on Radiological Protection (ICRP) recommendations on neutron 
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 14 of 24

 

weighting factors (O’Connell 2011). Based on this information, an adjusted N:P ratio of 0.420 was 

calculated for the period January 1, 1989 to December 31, 2012. 

3.2 Y-12 PLANT

The Y-12 Plant site profile discusses historical dosimetry information that can be used to determine 

neutron and photon exposure near HEU operations (ORAUT 2009). For this report, individual worker 

quarterly data for pairs of neutron and photon doses from workers in Y-12 Plant Departments 2301 

(Development Operations) and 2160 (Material Engineering) during the period from 1955 to 1959 were 

combined and used as a basis for obtaining an average N:P ratio. These data were selected because 

unique work activities involving EU were carried out in these departments during this period as 

indicated in Recycled Uranium Mass Balance Project Y-12 National Security Complex (BWXT 2000). 

Excerpts from this report stated (BWXT 2000): 

• In the late 1950s, continuous solvent extraction equipment was installed in the B-1 wing (of 

Building 9212). This period covered the transition from small-scale batch operations to the 

existing continuous recovery operation equipment in use today. 

• The 9206 building has been used extensively over its lifetime for the chemical processing of 

uranium. 

• Enriched uranium processes, activities, and/or missions of the 9206 Facility have included: 

conversion of UF6 to UF4 to uranium metal for weapons (1954 to 1964) and casting and 

machining of HEU metal for weapons (1955 to 1965).

Film-badge data for Departments 2160 (Material Engineering) and 2301 (Project Design Development 

and Analytical Development) appear to be somewhat typical of the N:P ratio for workers who were 

exposed to neutrons from HEU sources in Buildings 9202, 9206, and 9212. The data analysis shown 

in Figure 4-3 is based on 90 paired N:P results for 47 different workers in Departments 2160 and 

2301.

A study was conducted in 1989 of neutron and photon emissions from small storage containers of 

HEU in Building 9212 at the Y-12 Plant in Oak Ridge, Tennessee (Soldat et al. 1990). At the time of 

the measurements, the storage area in Building 9212 held small HEU containers of both UF4 and 

UO3. These small HEU containers were placed on a rack of shelves and arranged in a matrix that 

was criticality safe (Soldat et al. 1990). The exact enrichment value was not given, but the nature of 

the work in Building 9212 (as well as Buildings 9202 and 9206, discussed below) would indicate an 

enrichment value of 90% or greater. The measurements were made at a height of 39 in. (1 m) above 

the floor and 27 in. (0.7 m) from the nearest container. The location of the measurements was also 

selected so that it was only near containers of UF4 (Soldat et al. 1990). The characterization of the 

neutron field consisted of measurements of the neutron energy spectrum and personal dose 

equivalent from both neutrons and photons. The measured values for the neutron dose rate ranged 

from 0.7 to 2.08 mrem/hr. An N:P ratio was derived from this data by comparing the integrated 

neutron dose of 76.2 mrem for a period of 40.1 hours as determined from measurements made with a 

tissue equivalent proportional counter and the integrated photon dose of 68.0 mrem as measured by 

phantom-mounted thermoluminescent dosimeters over the same time period. The resulting N:P ratio 

for these measurements was 1.12.

Table 3-3 summarizes N:P ratio data based on survey information from the Y-12 Plant for the period 

1992 to 2004. The readings were taken either in general work areas or at a distance of 30 cm.
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Table 3-3. Summary of Y-12 Plant N:P ratio data, 1992 to 2004. 

Date Building Room N:P Range N:P Average Reference

06/02/1992 9981/9998 G-2/G-5 0.13–1.0 0.37 Y-12 2013; SRDB Ref ID: 129045

02/28/2001 9212 D-107 0.06–0.50 0.2 Y-12 2013; SRDB Ref ID: 126485

07/11/2002 9212 1004 0.1–0.5 0.3 Y-12 2013; SRDB Ref ID: 128511

12/05/2003 9212 1004 0.25–1.0 0.6 Y-12 2013; SRDB Ref ID: 128511

12/08/2003 9212 1004 0.25–1.0 0.6 Y-12 2013; SRDB Ref ID: 128511

05/25/2004 9212 1014 0.22–0.43 0.31 Y-12 2013; SRDB Ref ID: 126498

4.0 ADDITIONAL DATA ANALYSIS OF EMPLOYEE DATA MEASUREMENTS FOR K-25, 

PORTSMOUTH, AND Y-12 PLANT

An additional method of analysis was applied to the employee dosimetry datasets for K-25 and 

Portsmouth. This method – quantile regression – is discussed in ORAUT-RPRT-0087, Applications of 

Regression in External Dose Reconstruction (ORAUT 2018). In summary, this method uses quantile 

regression to fit a linear model to the neutron and photon doses (not an N:P ratio), giving the 50thand 95th-percentile regression fits directly – with the added benefit of requiring no special treatment of 

data that might yield a traditional N:P ratio of 1. This approach also allows for exploratory data 

analysis (i.e., deriving fits for the 50th- and 95th-percentile exposed populations). 

Table 4-1 provides the results of the quantile regression fits for K-25, Portsmouth, and Y-12 Plant – 

and data for all three sites combined. In the case of K-25, the entire dataset from 1989 to 2012 was 

used since the site was in a single (post-production) mode during that era. The 50th- and 95thpercentile regression fits for K-25 can be seen in Figure 4-1. At Portsmouth, enrichment activities 

ended in 2001, so the fits seen in Figure 4-2 are based on data from 1992 to 2001. Figure 4-3 

illustrates the fits for Y-12 Plant from 1955 to 1959. Finally, Figure 4-4 illustrates the fits for combining 

the datasets for all three sites.

Table 4-1. Summary of quantile regression fit parameters for K-25, Portsmouth, and Y-12 Plant 

employee data measurements.

Site N

50thpercentile 

slope

50th-percentile 

intercept 

(rem)

95thpercentile 

slope

95th-percentile 

intercept 

(rem)

K-25 (1989–2012) 375 0.139 0.01 0.112 0.03

Portsmouth, Production Era 

(1992–2001)

3,727 0.133 0.002 0.231 0.004

Y-12 Plant (1955–1959) 89 0.14 0.03 0.7 0.1

K-25/Portsmouth/Y-12 Plant 

Combined

4,191 0.195 0.002 0.846 0.002

The slope data for K-25, Portsmouth during the production era, and the Y-12 Plant show good 

agreement – especially at the 50th-percentile. Please note that production of HEU product at 

Portsmouth ended in 1991, therefore, this production dataset represents work with mostly LEU 

product. 
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 16 of 24

 

Figure 4-1. Quantile regression results for K-25 employee dosimetry data 

(1989–2012). 

Figure 4-2. Quantile regression results for Portsmouth employee dosimetry 

data, production era (1992–2001). 
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 17 of 24

 

Figure 4-3. Quantile regression results for Y-12 Plant employee dosimetry 

data (1955–1959). 

Figure 4-4. Quantile regression results for the combined K-25, Y-12, and 

Portsmouth employee dosimetry data sets. (Data for K-25 are in red, data for 

Portsmouth are in blue, and data for Y-12 are in black). 
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 18 of 24

5.0 PHOTON SUMMARY AND CONCLUSIONS

Table 5-1 provides a summary of the neutron dose rates, photon dose rates (where available), and 

N:P ratios (where available) associated with the measurements and calculations described above. 

Neutron dose rate measurements were in good agreement with the values in DOE (2009).

At locations where NU and DU were present, an N:P ratio of 0.2 was observed (Paducah and K-25).

At facilities where material with higher uranium enrichment was present, an increase in the N:P ratio 

was observed. A close-proximity (1 m) N:P ratio of 0.88 was calculated for LEU (5%) at the K-25 

storage yard. Survey and area monitoring N:P values at the K-25 and Portsmouth gaseous diffusion 

plants ranged from 0.1 to 0.605. N:P ratios ranging from 0.41 to 1.12 (surveys and stationary 

measurement) were found for HEU material at the Y-12 Plant.
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 19 of 24

 

Table 5-1. Summary of neutron dose rates, photon dose rates, and N:P ratios associated with 

natural, depleted, and enriched uranium materials at Paducah, Portsmouth, and K-25 Gaseous 

Diffusion Plants and Y-12 Plant. 

Location/reference Material

Measurement 

distance

Neutron 

dose rate 

(mrem/hr)

Photon 

dose rate 

(mrem/hr) N:P ratio

K-25 GDP, Cylinder Storage Yard (NISYS 

2000)

DU 1 m 0.07 0.7 0.1

Paducah GDP, Cylinder Storage Yard (BJC 

1999)

DU, NU Not given NAa NA 0.2

(0.14-0.42)

K-25 GDP, Cylinder Storage Yard (NISYS 

2000)

NU 1 m 0.15 0.68 0.22

DOE Good Practices Guide (DOE 2009) NU to 5% 

LEU

Not given 0.01-0.2 NA 2.0

K-25 GDP, Cylinder Storage Yard (NISYS 

2000)

5% LEU 1 m 0.6 0.68 0.88

Y-12 Plant, Bldg. 9212 (Soldat et al., 1990) HEU Not given 1.9 1.7 1.12

Y-12 Plant, Bldg. 9981/9998 (Y-12 2013) HEU 1 ft 2-32 1-4 0.37

(0.13-1.0)

Y-12 Plant, Bldg. 9212, Rm D-107 (Y-12 2013) HEU 1 ft 1-1.8 0.1-0.5 0.2

(0.06-0.50)

Y-12 Plant, Bldg. 9212, Rm 1004 (Y-12 2013) HEU 

(SRS/UF4)

1 ft 0.4-8 0.2-0.4 0.3

(0.1-0.5)

Y-12 Plant, Bldg. 9212, Rm 1004 (Y-12 2013) HEU 1 ft 0.1-0.4 0.1 0.6

(0.25-1.0)

Y-12 Plant, Bldg. 9212, Rm 1004 (Y-12 2013) HEU 1 ft 0.1-0.5 0.1 0.6

(0.25-1.0)

Y-12 Plant, Bldg. 9212, Rm 1014 (Y-12 2013) HEU 1 ft 1.4-1.8 0.4-0.6 0.31

(0.22-0.43)

Portsmouth GDP (Cardarelli 1997) HEU Contact 3 NA 0.125

DOE Good Practices Guide (DOE 2009) 97% HEU Contact 2-4 NA 2

Portsmouth GDP, X-326, Full 5” Cylinders 

(Bassett 1986)

97% HEU Contact <3.0 <2.4 1.25

DOE Good Practices Guide (DOE 2009) 97% HEU 1 m 1-2 NA 2

K-25, FSAR (LMES 1997) DU, NU, 

LEU, HEU

Contact NA NA 0.166

K-25 Equipment and Area Monitoring Surveys 

(Pope 2015)

DU, NU, 

LEU, HEU

Contact NA NA 0.099

K-25 Equipment and Area Monitoring Surveys 

(Pope 2015)

DU, NU, 

LEU, HEU

1 ft NA NA 0.085

K-25 Equipment and Area Monitoring Surveys 

(Pope 2015)

DU, NU, 

LEU, HEU

1 m NA NA 0.122

Portsmouth GDP (Cardarelli 1997) HEU 1 m <0.5 NA 0.125

K-25 Equipment and Area Monitoring Surveys 

(Pope 2015)

DU, NU, 

LEU, HEU

General area NA NA 0.605

K-25, FSAR (LMES 1997) DU, NU, 

LEU, HEU

General area NA NA 0.166

Portsmouth Area Monitoring Results (McGuire 

2008)

DU, NU, 

LEU, HEU

Not given NA NA 0.147

(0.01-0.398)

a. NA = not applicable.

Data based on analyses of personnel dosimeter measurements at these facilities (see Table 5-2) 

yielded N:P values between 0.369 and 0.420 and, for comparison, quantile regression 50th- and 95thpercentile relationships yielding slope values ranging from 0.133 to 0.231 (where the y-intercepts are

close to zero) respectively during eras after active HEU production had ceased. Finally, quantile 

regression 50th- and 95th-percentile relationships yielding slope values ranging from 0.14 to 0.7
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 20 of 24

 

respectively (again, where the y-intercept was close to zero) were found for workers actively handling 

HEU product at the Y-12 Plant.

Although a wide range of survey, modeling, and static dosimetry data is available for all three sites, 

the quantile regression results from each site’s personnel dosimetry measurements would yield the 

most accurate N:P ratio result. In addition, good agreement was seen – especially when comparing 

the slope values at the 50th-percentile level – between the personnel measurement results from all 

three sites. 

Table 5-2. Summary of N:P ratios and quantile regression relationships associated with natural, 

depleted, and enriched uranium materials at Portsmouth, K-25, and Y-12 Plants, based on personnel 

dosimetry. 

Location/reference Material

N:P

ratio Quantile regression relationship

K-25 Personnel 

Dosimetry, 1989–2012

DU, NU, LEU, HEU 0.420 50th neutron = 0.139 [photon (rem)] + 0.01 (rem)

95th neutron = 0.112 [photon (rem)] + 0.03 (rem)

Portsmouth Personnel 

Dosimetry, 1992–2013

DU, NU, LEU, HEU 0.369 50th neutron = 0.133 [photon (rem)] + 0.002 (rem)

95th neutron = 0.231 [photon (rem)] + 0.004 (rem)

Y-12 Plant Personnel 

Dosimetry, 1955–1959

HEU NAa 50th neutron = 0.14 [photon (rem)] + 0.025 (rem)

95th neutron = 0.7 [photon (rem)] + 0.098 (rem)

K-25/Portsmouth/Y-12 

Plant Combined

DU, NU, LEU, HEU NAa 50th neutron = 0.195 [photon (rem)] + 0.002 (rem)

95th neutron = 0.846 [photon (rem)] + 0.002 (rem)

a. NA = not applicable. 

The personnel data take into account the actual distances and residence time personnel would have 

had near potential source terms, whereas the survey data – by their nature – are static measurements 

taken during a snapshot of time. Survey data are typically taken at maximum dose rate locations to 

set bounds for radiological control purposes, thus, the survey data can be considered an upper 

bound. Most of the surveys are indicating N:P ratios less than 0.4:1 with only three slightly greater 

than 1.2:1. This further confirms that the 2:1 upper bound N:P ratio discussed in DOE (2009) is 

indeed a conservative upper bound. Given multiple work locations and personnel movement between 

areas of neutron exposure with N:P ratios less than 0.4 – and areas of no neutron exposure – the 

data resulting from the quantile regression analysis of personnel dosimetry provides the best estimate 

for an overall N:P ratio for unmonitored workers.

An N:P relationship based on the quantile regression analysis of dosimetry data for all three sites 

combined – as summarized in Tables 4-1 and 5-2 – is recommended for assigning neutron dose for 

unmonitored workers in eras as described above. Application of the guidance in Attachment A yields 

a lognormal distribution that reflects the 50th- and 95th-percentile fits for the quantile regression 

analysis of the personnel dosimetry data from K-25, Portsmouth, and Y-12 Plant combined. This 

distribution is to be combined with photon dose (defined as a constant or distribution) – using Monte 

Carlo methods as needed – to yield a distribution of neutron dose.

6.0 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 21 of 24

REFERENCES

Bassett, A. C., 1986, “External Dose Rates from Five-Inch Product Cylinders,” memorandum to File,

Goodyear Atomic Corporation, Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, 

February 17. [SRDB Ref ID: 23926]

BJC (Bechtel Jacobs Company), 1999, Paducah UF6 Cylinder Painting Project Neutron Dose Study, 

Paducah Gaseous Diffusion Plant, Paducah, Kentucky, July 7. [SRDB Ref ID: 13682]

BJC (Bechtel Jacobs Company), 2000, Evaluation of Potential Radiation Doses to Members of the 

Public and ETTP Unmonitored Site Workers East Tennessee Technology Park, Oak Ridge, 

Tennessee, BJC/OR-585/R1, Oak Ridge, Tennessee, May. [SRDB Ref ID: 29760]

BWXT (BWXT Y-12, LLC), 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. [SRDB Ref ID: 16500]

Cardarelli II, J., 1997, Health Hazard Evaluation Report No. 96-1298-2651, Portsmouth Gaseous 

Diffusion Plant, Piketon, Ohio, National Institute for Occupational Safety and Health, 

Cincinnati, Ohio. [SRDB Ref ID: 10913]

Dodd, T., K. Duncan, and M. Teeters, 2001, Dosimetry Program Standards, UE2-HP-RP1034, Rev. 2, 

Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, March 21. [SRDB Ref ID: 75434, p. 93]

Dodd, T., K. Duncan, and M. Teeters, 2002a, Dosimetry Program Standards, UE2-HP-RP1034, Rev. 

2, Change B, Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, October 14. [SRDB Ref ID: 

75434, p. 27]

DOE (U.S. Department of Energy), 2000, Area Characterization Survey, East Tennessee Technology 

Park, Oak Ridge, Tennessee, February 15. [SRDB Ref ID: 93592]

DOE (U.S. Department of Energy), 2003, Recycled Uranium, United States Production, Enrichment, 

and Utilization, DOE/SO-0003, Office of Security, Washington, D.C., May. [SRDB Ref ID: 

23595]

DOE (U.S. Department of Energy), 2009, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2009, Washington, D.C., July. [SRDB Ref 

ID: 97381]

LMES (Lockheed Martin Energy Systems, Inc.), 1997, K-25 Site UF6 Cylinder Storage Yards Final 

Safety Analysis Report, K/D-SAR-29, Oak Ridge K-25 Site, Tennessee, February 28. [SRDB 

Ref ID: 126947]

McGuire, M., 2008, “Area Monitor TLD Results,” letter to T. Taulbee, United States Enrichment 

Corporation, Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, August 11. [SRDB Ref ID: 

131566]

NAS (National Academy of Sciences), 2005, Monitoring Nuclear Weapons and Nuclear-Explosive 

Materials: An Assessment of Methods and Capabilities, National Academies Press, 

Washington, D.C. [SRDB Ref ID: 119945]
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 22 of 24

 

NISYS (NISYS Corporation), 2000, Determination of Neutron to Gamma Dose Ratios for the UF6

Cylinder Yards, Calc. No. 1073-018-C001, Rev. 0, Duluth, Georgia, August 25. [SRDB Ref ID: 

108666]

O’Connell, P., 2011, “DOE Adoption of 1990 ICRP recommendations on neutron weighting factors,” 

email to C. A. Kirkwood (Oak Ridge Associated Universities), U.S. Department of Energy, 

Office of Health and Safety, Washington, D.C., December 20. [SRDB Ref ID: 106069] 

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

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

Oak Ridge, Tennessee, April 7. [SRDB Ref ID: 19445] 

ORAUT (Oak Ridge Associated Universities Team), 2009, Y-12 National Security Complex – 

Occupational External Dosimetry, ORAUT-TKBS-0014-6, Rev. 02, Oak Ridge, Tennessee, 

December 18. [SRDB Ref ID: 77701] 

ORAUT (Oak Ridge Associated Universities Team), 2012, Y-12 National Security Complex – 

Occupational Internal Dose, ORAUT-TKBS-0014-5, Rev. 03, Oak Ridge, Tennessee, 

March 12. [SRDB Ref ID: 109202] 

ORAUT (Oak Ridge Associated Universities Team), 2018, Applications of Regression in External 

Dose Reconstruction, ORAUT-RPRT-0087, Rev. 00, Oak Ridge Tennessee, March 13. 

[SRDB Ref ID: 170100] 

Pekala, T., and M. Philpot, 2014, “Unredacted Electronic Version of Health Physics Records System,”

spreadsheet, Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, November 14. [SRDB Ref 

ID: 138026]

Pope, J., 2015, Spreadsheet of Neutron and Photon Dose Rates at K-25, Equipment and General 

Area Surveys, East Tennessee Technology Park, Oak Ridge, Tennessee, May 14. [SRDB 

Ref ID: 144768]

Scherpelz, R. I., and M. K. Murphy, 1995, Enriched Uranium Cylinder Neutron Characterization 

Measurements at the Portsmouth Gaseous Diffusion Plant, Pacific Northwest Laboratory, 

Richland, Washington, September. [SRDB Ref ID: 109074]

Soldat, K. L., L. W. Brackenbush, M. P. Moeller, R. I. Sherpelz, and J. E. Tanner, 1990, Neutron Dose 

Equivalent and Energy Spectra Measurements at the Oak Ridge National Laboratory and Y-12 

Plant, PNL-7528, Pacific Northwest Laboratory, Richland, Washington. [SRDB Ref ID: 11093]

Soldat, K. L., and J. E. Tanner, 1992, Neutron Dose Equivalent and Energy Spectra Measurements at 

the Portsmouth Gaseous Diffusion Plant, Pacific Northwest Laboratory, Richland, Washington, 

April. [SRDB Ref ID: 8119]

Y-12 (Y-12 National Security Complex), 2013, Neutron and Gamma Survey Results in Buildings 9981 

and 9212, 1992-2006, Y-12 Site, Oak Ridge, Tennessee, November. [SRDB Ref IDs: 

128507, 128508, 128509, 128510, 128511, 128512, 128513, 129045, 129046, 129047, 

129048, 129049, 129050]
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ATTACHMENT A 

QUANTILE REGRESSION ANALYSIS IMPLEMENTATION WITHIN IREP

Using a quantile regression analysis of paired photon / neutron data, an N:P ratio can be derived for 

calculating neutron dose based on photon dose using the following inputs (note that this N:P ratio 

derivation is needed to implement the quantile regression results in the calculation tools used by the 

ORAU Team): 

1. 50th-percentile equation, 

2. 95th-percentile equation, and

3. Known photon dose (normal or lognormal distribution). 

The neutron dose at any photon dose is assumed to have a lognormal distribution, the parameters of 

which are described by the 50th- and 95th-percentile equations.

In this discussion, N50(p) denotes the 50th-percentile neutron dose for a given photon dose p and 

N95(p) denotes the 95th-percentile neutron dose for a given photon dose (p). 

The lognormal parameters for the N:P ratio of a given photon dose are:

(A-1) 50( ) ( ) ( )

N p GM p

p =

where

z95 = 1.64485 (normal standard deviation for the 95th-percentile)

and

(A-2)

95

50

95

( ) ln ( )

( )

N p

N p

z

GSD p e

            =

To obtain the neutron dose, the derived N:P ratio is multiplied by the photon dose distribution. (See 

example below). 

Example Calculation:

N:P ratio from quantile regression relationship from K-25/Portsmouth/Y-12 Plant combined: 

N95(p) = 0.846p + 0.002

N50(p) = 0.195p + 0.002

Photon dose p = 0.05 rem

Calculation of N:P Ratio from Photon Dose and Quantile Regression Neutron Dose

= = (A-3) 0.011366 ( ) 0.22732

0.05 GM p

and
Document No. ORAUT-RPRT-0060 Revision No. 00 Effective Date: 03/28/2019 Page 24 of 24

ATTACHMENT A

QUANTILE REGRESSION ANALYSIS IMPLEMENTATION WITHIN IREP (continued)

(A-4)

            = =

0.0444575 ln

0.011366

1.64485 GSD p e ( ) 2.291481

Monte Carlo Simulation to Assign Neutron Dose

Photon dose (p) = 0.05 rem ±0.015 rem (normal distribution)

N:P ratio: GM = 0.22732, GSD = 2.291481 (lognormal distribution)

A sample from the normal photon dose distribution is multiplied by a sample from the derived N:P

ratio. This process is repeated for 10,000 iterations and a distribution is fit to the resulting data set 

(see results in Figure A-1). 

Figure A-1. Lognormal distribution fit of calculated neutron data based on photon sample and N:P 

ratio sample. Neutron dose lognormal fit from photon sample: 0.05 ±0.015 rem. (The parameters for

this fitted lognormal are: Mean = 0.016 rem, Standard Deviation = 0.015 and, for use with IREP: GM 

= 0.0135 rem, GSD = 2.090, and Shift = -0.00198)