DRAFT 

REVIEW OF THE 

REVISED HUNTINGTON PILOT PLANT SITE PROFILE 

Contract No. 200-2009-28555 

SCA-TR-SP2013-0043

Stephen F. Marschke 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182

June 2013 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted. 
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S. Cohen & Associates:

Technical Support for the Advisory Board on

Radiation & Worker Health Review of

NIOSH Dose Reconstruction Program

Document No. 

SCA-TR-SP2013-0043

Effective Date: 

Draft – June 4, 2013

Review of the Revised Huntington Pilot Plant 

Site Profile Page 2 of 25

Task Manager:

________________________ Date: ___________

John Mauro, PhD, CHP

Supersedes:

N/A

Project Manager:

________________________ Date: ___________

John Stiver, MS, CHP

Reviewer:

 John Mauro

 John Stiver

Record of Revisions

Revision 

Number Effective Date Description of Revision

0 (Draft) 06/04/2013 Initial issue.
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TABLE OF CONTENTS 

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

1.0 Background and Introduction ..............................................................................................6

2.0 Independent Evaluation of Exposure Estimates ..................................................................8

2.1 Internal Dose during Operations ..............................................................................8

2.2 External Dose during Operations .............................................................................9

2.2.1 Gamma Energy Spectrum ........................................................................... 9

2.2.2 Beta Energy Spectrum ............................................................................... 11

2.2.3 Penetrating Dose Rate ............................................................................... 12

2.2.4 Dose Rate Energy Distribution ................................................................. 15

3.0 Airborne Nickel Concentration ..........................................................................................17

4.0 Shallow Dose – Typographical Errors ...............................................................................19

4.1 UCNC 1958 Contact Beta Dose Rate ....................................................................19

4.2 Annual Dose Rate to Hands and Forearms ............................................................19

4.3 Hourly Dose Rate at 30 Centimeters .....................................................................19

5.0 Summary of Findings and Observations ............................................................................21

5.1 Finding 1 – Recycled Uranium Isotopes of Concern .............................................21

5.2 Finding 2 – Uranium Specific Activity ..................................................................21

5.3 Finding 3 – Unit Conversion..................................................................................21

5.4 Observation 1 – Gamma Energy Spectrum ...........................................................21

5.5 Observation 2 – Beta Energy Spectrum .................................................................21

5.6 Observation 3 – Residue Storage Container ..........................................................22

5.7 Observation 4 – Penetrating Dose Rate .................................................................22

5.8 Finding 4 – Dose Energy Distribution ...................................................................22

5.9 Finding 5 – Including Modern Airborne Nickel Concentrations ...........................22

5.10 Finding 6 – Excluding Data from the Crushing, Grinding, Handling, and 

Calciner Areas ........................................................................................................22

5.11 Observation 5 – Use of Site Concentration Distributions ......................................22

5.12 Finding 7 – Numerical Typographical Errors ........................................................22

References ......................................................................................................................................24
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LIST OF FIGURES 

Figure 1: Uranium Gamma Spectrum, using MicroShield .......................................................... 10

Figure 2: Uranium Gamma Spectrum, using the National Nuclear Data Center ......................... 11

Figure 3: Uranium Beta Spectrum, using the National Nuclear Data Center .............................. 12

Figure 4: MicroShield Dose Rate from a 20-Gallon Drum by Gamma Energy .......................... 14

Figure 5: MicroShield Dose Rate from a 20-Gallon Drum – Cumulative over Gamma 

Energy .......................................................................................................................... 15

Figure 6: Energy Distribution for Photon Emissions ................................................................... 16
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ABBREVIATIONS AND ACRONYMS 

BNL Brookhaven National Laboratory 

Ci Curies

cm centimeter 

DOE Department of Energy 

DRSC Dose Reconstruction Subcommittee 

g gram 

g/cc gram per cubic centimeter 

HPP Huntington Pilot Plant 

keV kiloelectron volts 

MCNPX Monte Carlo N-Particle eXtended 

MeV mega electron volts, million electron volts 

mg Ni/m3 milligrams of nickel per cubic meter 

mrep/hr milli-Röntgen equivalent physical per hour 

nCi/g U nanocurie per gram of uranium 

NIOSH National Institute for Occupational Safety and Health 

NNDC National Nuclear Data Center 

OCAS Office of Compensation Analysis and Support 

ORAUT Oak Ridge Associated Universities Team 

pCi/mg picocurie per milligram 

ppm part per million 

R/hr Roentgen per hour 

rad/hr radiation absorbed dose per hour 

RADAR RAdiation Dose Assessment Resource 

RPP Reduction Pilot Plant 

RU recycled uranium 

SC&A S. Cohen and Associates (SC&A, Inc.) 

TBD Technical Basis Document 

UCNC Union Carbide Nuclear Company 
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1.0 BACKGROUND AND INTRODUCTION 

During the meeting of the Dose Reconstruction Subcommittee (DRSC) held on February 4, 

2013, one of the topics discussed was SC&A’s focused review of the Huntington Pilot Plant 

(HPP) exposure matrix (SC&A 2008, Attachment 3), as originally provided in ORAUT 2004a. 

During that meeting, NIOSH indicated that the original SC&A findings are now addressed in a 

revised version of the exposure matrix; i.e., Technical Basis Document for the Huntington Pilot 

Plant, Huntington, West Virginia, OCAS-TKBS-0004 (OCAS 2008). SC&A was thereby 

directed by the DRSC to review the revised exposure matrix to evaluate the degree to which our 

original findings have been addressed. 

SC&A presented the results of that evaluation in a report dated March 21, 2013 (SC&A 2013). 

Because of the relatively short time period between authorization to perform the review and the 

fact that NIOSH developed “an entirely revised strategy for deriving the external penetrating and 

non-penetrating doses to workers during operations” (SC&A 2013), SC&A was not able to fully 

address the responsiveness of the revised site profile to each of its original 12 findings. 

However, we were able to recommend the following: 

 Closing Findings 1, 4, 6, 9, and 10 

 Withdrawing Findings 2 and 5 

 Keeping open Findings 3, 7, 8, 11, and 12, because SC&A requires additional time to 

investigate the new strategies that NIOSH has incorporated into the revised site profile 

In Section 3 of SC&A 2013, we summarized the new strategy that NIOSH adopted in its revised 

site profile. A summary of that strategy is repeated here because it serves as a useful 

introduction to the new material that is reviewed in this report: 

...in the earlier version of the site profile, NIOSH assumed that the level of 

enrichment was 36%, because this was the contractual limit on the level of 

enriched uranium that was permitted at Huntington. SC&A concurred with 

NIOSH at the time that this was a very conservative assumption, because the 

actual data showed that the level of enrichment in the actual uranium processed 

at the facility was only a few percent. Because of this extremely conservative 

assumption, NIOSH concluded that it was not necessary to explicitly account for 

the internal exposures that were associated with recycled uranium (RU), which 

was handled at the plant. SC&A agreed with this simplifying assumption. In the 

revised site profile, NIOSH no longer assumes 36% enrichment, but instead uses 

the more realistic enrichment of 2%, and now the site profile explicitly addresses 

RU by employing the RU composition used in the K-25 Gaseous Diffusion Plant 

TBD (ORAUT 2006). As part of this review, we did not review the models and 

assumptions employed in the K-25 TBD, but leave that to the Gaseous Diffusion 

Plant Work group. 
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This report presents the completion of our review of the revised site profile for HPP, including 

explicitly addressing the new material provided in the revised site profile and the five original 

findings that we identified as requiring additional review. 
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2.0 INDEPENDENT EVALUATION OF EXPOSURE ESTIMATES 

2.1 INTERNAL DOSE DURING OPERATIONS 

Section 5.2 of OCAS-TKBS-0004 (OCAS 2008) states, “Mass of 2% enriched uranium is 

estimated to be 1,200 pCi/mg (DOE 2004).” DOE (2004), Table 2-5, gives the specific activity 

of 2% enriched uranium as 1,000 pCi/mg (1 × 10-6 Ci/g). Alternatively, the DOE (2004) 

equation for calculating uranium specific activity (i.e., [0.4 + 0.38E + 0.0034E2] × 10-6 Ci/g, 

where E = % U-235 by weight) gives 1,174 pCi/mg at 2% U-235 enrichment. The differences 

between these three values are all likely to be due to round-off. 

Section 5.2 of OCAS-TKBS-0004 states, “Pu-239 and Np-237 are the contaminants that are 

likely to have contributed to significant dose at the RPP” [Reduction Pilot Plant]. No basis or 

reference for making this statement is provided. Additionally, the site profiles for Paducah 

[ORAUT-TKBS-0019-5 (ORAUT 2012)], Portsmouth [ORAUT-TKBS-0015-5 (ORAUT 

2004b)], and K-25 [ORAUT-TKBS-0009-5 (ORAUT 2006)] all indicated that Technetium-99 is 

the fission product of concern from a dosimetry standpoint for recycled uranium. Paducah also 

indicated that americium-241 was also “present in small amounts in the feed material produced 

from RU [recycled uranium].”

Section 5.2 of OCAS-TKBS-0004 indicates that the default isotopic ratios for Pu-239 and 

Np-237 were taken from the K-25 site profile (ORAUT 2006). SC&A assumes that ORAUTTKBS-0009-5, Table 5-6, was the source for these ratios. Using the data from Table 5-6 of 

ORAUT-TKBS-0009-5 and the three uranium specific activities discussed above, SC&A 

attempted to match the Pu-239 and Np-237 contaminant activities presented in OCAS-TKBS0004, Table 4. As shown in the table below, we were not able to exactly match the Table 4 

values, although our values are close to and on either side of them. 

Nuclide ORAUT-TKBS-0009, 

Table 5-6 (nCi/g U)

Containment Activity (unit-less)

Table 4

Specific Activity (pCi/mg)

1,200 1,000 1,174

Pu-239 67.5 0.063 0.056 0.068 0.058

Np-237 5.4 0.0050 0.0045 0.0054 0.0046

Am-241 67.5 N.G. 0.056 0.068 0.058

U-236 0.93 N.G. 0.00078 0.00093 0.00079

U-235 43.9 N.G. 0.037 0.044 0.037

U-234 702 N.G. 0.59 0.70 0.60

U-238 337.5 N.G. 0.28 0.34 0.29

Th-230 18.9 N.G. 0.016 0.019 0.016

Tc-99 0.12 N.G. 0.00010 0.00012 0.00010

Based on the above discussion, the following two findings related to the internal dose have been 

generated. 
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Finding 1: Since the three diffusion plants (the source of the HPP nickel) had additional 

isotopes of concern, NIOSH should clearly provide the basis for only specifying Pu-239 and 

Np-237 as isotopes of concern for recycled uranium. 

Finding 2: NIOSH should clearly state which uranium specific activity was used in the analysis 

and ensure that it was used consistently throughout the analysis. 

2.2 EXTERNAL DOSE DURING OPERATIONS 

2.2.1 Gamma Energy Spectrum 

The external dose review focused on Attachment A. The first item to be checked was the 

Table A2 gamma ray spectra. 

Finding 3: There is a unit conversion error in going from Table A2 column 3 (Photons per 

decay 238U) to column 4 (Photons per second per Ci 238[U]). 

The values in column 4 are a factor of 3,600 larger than they should be. For example, for the 

0.001 to 0.01 MeV row: 

 

 

 

 

 

 

When this calculated value is compared to the value from Table A2, column 4, the difference is a 

factor of 3,600, as shown: 

As will be demonstrated below, this unit conversion error only affects column 4 of Table A2 

(and Table A3), but does not affect the calculated dose rates.

Next, SC&A checked the gamma energy spectra given in Table A2. First, MicroShield (Grove 

2009) was used to generate gamma energy spectra from the decay of U-238, U-235, and U-234 

and their decay products. MicroShield contains its own library of radionuclide decay data, 

including decay schemes and photon energies. It was initially assumed that the decay products 

were in full equilibrium with their uranium parents. The three uranium MicroShield spectrums 

were combined using the drum activities given in Table A1 (i.e., U-234: 1,420 Ci; U-235: 

78 Ci; and U-238: 295 Ci), and then normalized to the U-238 activity. The results are shown in 

Figure 1 (the purple dashed line) and compared to the Table A2 spectra (the blue solid line). 

 

 
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Figure 1: Uranium Gamma Spectrum, using MicroShield 

As Figure 1 shows, when full equilibrium is assumed, there is considerable disagreement 

between the MicroShield and Table A2 gamma spectra. SC&A next used MicroShield to create 

gamma spectra at decay times of 10, 50, and 100 years. These would be the conditions before 

Th-230 and its progeny could grow in for the U-238 series and before any Pa-231 and its 

progeny could grow in the U-235 series. These would be the conditions one would expect from 

the types of uranium handled at HPP (not including any recycled uranium). The same procedure 

was followed to develop the gamma spectra, and the results are shown in Figure 1 as the red, 

green, and light blue dashed lines. Figure 1 now shows that there is much better agreement 

between the Table A2 and MicroShield gamma spectrums. Although there are differences 

between the Table A2 and MicroShield gamma spectrums, those differences are not considered 

significant. The differences are likely due to different decay times, different energy binning, etc. 

Attachment A indicated that NIOSH utilized data from the National Nuclear Data Center 

[NNDC at Brookhaven National Laboratory (BNL 2013)] in the development of the Table A2 

gamma energy spectrum. SC&A likewise went to the NNDC and attempted to reproduce the 

Table A2 spectrum. Figure 2 shows the results of that attempt. 

1.E-06

1.E-0 5

1.E-0 4

1.E-03

1.E-0 2

1.E-0 1

1.E+00

1.E+01

1 1 0 100 1000 10000 Intensity (Photons per Decay U-238)

Gamma Energy (keV)

NIOSH, Table A2

MicroShield, Full Equilibrium

MicroShield, 10 yr Decay

MicroShield, 50 yr Decay

MicroShield, 100 yr Decay
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Figure 2: Uranium Gamma Spectrum, using the National Nuclear Data Center

With the possible exception of 600 and 1,000 keV, Figure 2 shows that SC&A was able to 

almost exactly match the Table A2 gamma spectrum with data from the NNDC. As with the 

MicroShield comparison, the Figure 2 agreement between the SC&A and Table A2 gamma 

spectrums is considered good. The differences are likely due to different assumed decay times, 

different energy binning, etc.

Observation 1: Consistent with HPP operations, the gamma energy spectrum used by NIOSH is 

not based on full equilibrium of U-238, U-235, and U-234 with their decay products; rather it is 

consistent with a decay period where only short-lived progeny of U-238 and U-235 would have 

had an opportunity to grow in.

2.2.2 Beta Energy Spectrum

Similar to Table A2, Table A3 presents the beta energy spectra. Since it does not include 

information on beta decay, SC&A could not use MicroShield to check the Table A3 beta 

spectrum. Attachment A indicates that NIOSH obtained the beta spectrum from the RADAR 

website (Stanford 2013). SC&A likewise obtained beta data from the RADAR website in order 

to check the Table A3 spectra.

The SC&A-developed beta spectra are shown in Figure 3 (green line). Similar to the gamma 

spectra, there are significant differences between the SC&A-developed beta spectra and the 

Table A3 spectra (shown as the blue line in Figure 3).

1.E-0 6

1.E-05

1.E-0 4

1.E-0 3

1.E-02

1.E-0 1

1.E+00

1.E+01

1 10 100 1000 10000 Intensity (Photons per Decay U-238)

Gamma Energy (keV)

NIOSH, Table A2

NNDC, Full Equilibrium

NNDC, 10 yr Decay

NNDC, 50 yr Decay

NNDC, 100 yr Decay
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Figure 3: Uranium Beta Spectrum, using the National Nuclear Data Center

As a further check, SC&A performed a second beta spectrum analysis, only this time the upper 

energy range was selected as the mid-point between the upper limits given in Table A3. For 

example, Table A3 has upper energy bins from 0.02 to 0.03 and from 0.03 to 0.04; instead of 

these bins, SC&A used 0.025 to 0.035 and 0.035 to 0.045, i.e., the revised SC&A bins were 

centered on the Table A3 upper energy. The results of this revised binning are shown in Figure 3

(red line). What this exercise demonstrates is the effect that simple changes in assumptions can 

have on the resulting energy spectrum.

Regardless of the manner in which the beta energy bins are defined, the agreement between the 

two SC&A-calculated beta energy spectra and the Table A3 energy spectra is considered good.

Observation 2: Consistent with HPP operations, the beta energy spectrum used by NIOSH is 

not based on full equilibrium of U-238, U-235, and U-234 with their decay products; rather, it is 

consistent with a decay period where only short-lived progeny of U-238 and U-235 would have 

had an opportunity to grow in.

2.2.3 Penetrating Dose Rate

Observation 3: In the current site profile (OCAS 2008), NIOSH assumed that residues were 

stored in 20 gallons drums, while in the previous site profile (ORAUT 2004a), it was assumed 

that the residues were stored in “birdcages.” The basis for this change in assumption is not 

provided in OCAS 2008.

To calculate the penetrating dose rate, NIOSH assumed that the “residues were placed into 

20 gallon drums and then stored on the ground floor of the plant spaced over a 40 ft. by 105 ft. 

area” (OCAS 2008, page 8), which is consistent with the information provided in UCNC 1958: 

0

0.05

0.1

0.15

0.2

0.25

0.01 0.1 1 1 0 Intensity (Betas per Decay of U-238)

Beta Energy (MeV)

NIOSH, Table A3

SC&A Check, 50 yr Decay, Table A3 Bins

SC&A Check, 50 yr Decay, Revised Bins
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“The residue from 150,000 lb. of feed material is spaced uniformly on the ground floor over an 

area 45 ft. × 105 ft. with the residue from each batch being in a single 20-gal. drum.” These 

assumptions regarding the configuration of the residues differ from what had previously been 

assumed, i.e., “This process residue was most likely decanted into storage units known as 

‘birdcages,’” and “’Birdcage’ container most likely used for storage and shipping of enriched 

uranium at the Huntington Pilot Plant” (ORAUT 2004a). Notice that the previous site profile 

(ORAUT 2004a) never presents documentation that the “birdcages” were used at the HPP, it 

only asserts that they were most likely used, as opposed to the current site profile (OCAS 2008), 

which based the use of 20-gallon drums on a document that is contemporary with HPP operation 

(i.e., UCNC 1958).

SC&A used MicroShield (Grove 2009) to perform a simple check on the MCNPX (LANL 2008) 

calculation that was performed by NIOSH and explained in Attachment A of OCAS 2008. 

Rather than attempt to model an array of 20-gallon residue drums, SC&A modeled a single 

drum. Therefore, it is expected that the SC&A MicroShield dose rates should be smaller than 

those presented in Table A4.

The SC&A MicroShield model was an 18.25-inch diameter by 14-inch high cylindrical source 

composed of nickel with a density of 0.4 g/cc, as described in Attachment A (OCAS 2008, page 

23). The dose receptor position was 30 cm from the center of the drum, and 77.9 cm above the 

surface.

In the first MicroShield run, SC&A entered the gamma spectrum from Table A2, column 4 of 

OCAS 2008. The resulting dose rate, when adjusted for the amount of U-238 present in the 

drum, was 0.044 R/hr. Once the dose rate was corrected for the Table A2 unit conversion error 

(see Finding 3), the single drum MicroShield-calculated dose rate, integrated over the energy 

spectrum, was 1.2 × 10-5 R/hr, which agrees well with the 2.56 × 10-5rad/hr drum array dose rate 

given in Table A4. This is considered confirmation that NIOSH did not include the unit 

conversion error in the MCNPX calculation. It is also considered confirmation of the MCNPXcalculated dose rate, with the difference between 1.2 × 10-5 R/hr and 2.56 × 10-5

rad/hr attributed 

to the difference between a single drum and an array of drums.

SC&A made a second MicroShield run, but instead of entering the Table A2 gamma spectra, 

SC&A allowed MicroShield to calculate the gamma spectra from its internal library and the 

entered Table A1 drum U-238, U-235, and U-234 inventory. The dose rate from the second 

MicroShield run was 1.4 × 10-5 R/hr, which also agrees with the Table A4 dose rate.
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Figure 4: MicroShield Dose Rate from a 20-Gallon Drum by Gamma Energy

Figure 4 shows the dose rate as a function of the gamma energy spectrum, while Figure 5 shows 

the cumulative dose rate as a function of the spectrum. As expected based on the above 

discussion, there is good agreement between the calculated dose rates.

Figure 5 also shows the Table A4 dose rate of 2.56 × 10-5rad/hr (the dashed blue line). If it is 

assumed that the difference between the MicroShield dose rate and the Table A4 dose rate is due 

to a single drum versus an array of drums, and if it is assumed that this same difference will be 

maintained for the MicroShield-generated gamma spectrum, then the resulting dose rate due to 

an array of drums is estimated to be 2.9 × 10-5 R/hr (the dashed red line on Figure 5). 

Considering the differences between MicroShield and MCNPX, the gamma spectrums, one drum 

versus an array of drums, is considered to be good agreement.

1.E-1 3

1.E-12

1.E-11

1.E-10

1.E-0 9

1.E-0 8

1.E-0 7

1.E-06

1.E-05

0.001 0.01 0.1 1 10 Dose Rate at 30 cm from Gamma Sources (R/hr)

Gamma Energy (MeV)

NIOSH, Table A2, MicroShield

SC&A Check, 50 yr Decay
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Figure 5: MicroShield Dose Rate from a 20-Gallon Drum – Cumulative over Gamma 

Energy

Observation 4: The penetrating dose rate calculated by NIOSH using MCNPX and modeling an 

array of drums is consistent with the dose rate calculated by SC&A using MicroShield and 

modeling a single drum. Furthermore, the MicroShield-calculated dose rate using the Table A2 

gamma energy spectrum is consistent with the dose rate calculated using the gamma energy 

spectrum from the MicroShield radionuclide decay library.

2.2.4 Dose Rate Energy Distribution

Finally, Figure 6 is an attempt to reproduce OCAS-TKBS-0004, Figure A1. As shown, when the 

Table A2 Gamma energy spectrum is used, there is good agreement between Figure 6 and 

Figure A1, and the dose breakdown is approximately a 50/50 split for 0–250 keV/>250keV. 

However, when the MicroShield spectrum is used, the dose breakdown is ~70% for 0–250 keV 

and ~30% for >250 keV.

Finding 4: The dose breakdown between 0–250 keV and >250 keV varies from 50/50 to about 

70/30, depending on the gamma spectrum.

1.E-1 1

1.E-1 0

1.E-0 9

1.E-0 8

1.E-0 7

1.E-06

1.E-05

1.E-04

0.001 0.01 0.1 1 10 Dose Rate at 30 cm from Gamma Sources (R/hr)

Gamma Energy (MeV)

NIOSH, Table A4, Tally 5

NIOSH, Table A2, MicroShield

SC&A Check, 50 yr Decay

SC&A Check, Adjusted for Row of Drums
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Figure 6: Energy Distribution for Photon Emissions

(OCAS-TKBS-004, Figure A1)

0.0

0.2

0.4

0.6

0.8

1.0

1.2

0 0.5 1 1.5 2 2.5 Cummulative Dose from Photons

Gamma Energy (MeV)

NIOSH, Table A2, MicroShield

SC&A Check, 50 yr Decay
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3.0 AIRBORNE NICKEL CONCENTRATION

In our March 21, 2013, evaluation report (SC&A 2013), SC&A questioned the use of modern 

sample data to estimate historical airborne nickel concentrations. In brief, Enterline and Marsh 

(1982) presented airborne nickel concentrations for various areas of the HPP (see Enterline and 

Marsh 1982, Table 8). These nickel concentrations were a combination of both modern and 

historical measurements. Enterline and Marsh recognize that the modern data are not directly 

compatible with historical conditions and made an attempt to correct for that fact, i.e., 

“Whenever possible and/or applicable, the modern data were adjusted on the basis of a 

knowledge of process changes and environmental controls that were implemented over the 

years.” However, Enterline and Marsh further recognize that, “extrapolation of modern sample 

data to historical exposures is imperfect, but it can be assumed that the historical exposures were 

the same or, in most cases, of greater magnitude.” In other words, the modern nickel 

concentrations given in Enterline and Marsh 1982, Table 8, are likely smaller than what the 

historical concentrations would be.

Nevertheless, NIOSH utilized all of the nickel airborne concentration from Enterline and Marsh 

1982, Table 8 (both modern and historical), to develop geometric mean (0.046 mg Ni/m3) and 

95th percentile (0.44 mg Ni/m3) concentrations. As stated above, SC&A took exception to 

utilizing the modern data, and recalculated the mean (0.242 mg Ni/m3) and 95th percentile 

(2.01 mg Ni/m3) concentrations using only the 10 historical concentrations from Enterline and 

Marsh 1982, Table 8.

Finding 5: Provide justification for including modern airborne nickel concentrations in the 

concentration distribution, when Enterline and Marsh 1982 indicate that the historical 

concentrations were (in most cases) of greater magnitude.

Furthermore, at the beginning of their report, Enterline and Marsh state that the concentration of 

airborne nickel was estimated to range from 20 to 350 mg Ni/m3in areas where the matte was 

crushed, ground, and handled, and from 5 to 15 mg Ni/m3around the calciners. These 

concentrations are significantly larger than any of the values given in Enterline and Marsh 1982, 

Table 8, and no explanation is provided as to why they have not been included.

The NIOSH 95th percentile concentration (0.44 mg Ni/m3) implies that a worker spends only 

about 11 minutes per work-day in an area with a concentration of 20 mg Ni/m3. If a worker 

spent more time in such an area, or if the concentration were 350 mg Ni/m3, then use of the 

NIOSH 95th percentile concentration would be claimant unfavorable.

Finding 6: Provide justification for excluding from the concentration distribution the airborne 

nickel concentration in the crushing, grinding, and handling areas and the area around the 

calciners reported by Enterline and Marsh (1982).

Finally, it should be noted that Enterline and Marsh do not use their Table 8 nickel airborne 

concentration to develop a single site distribution. Rather, they couple their Table 8 

concentrations with an individual worker’s work location and duration to develop a unique 

exposure concentration for each individual. For example, in Table 4 of their report, Enterline 
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and Marsh present exposure concentrations for four workers. The table below compares the 

Enterline and Marsh 1982, Table 4, average exposure concentrations to the NIOSH and SC&A 

concentration distributions.

Enterline and Marsh 

1982, Table 4

Mean 95th Percentile

NIOSH SC&A NIOSH SC&A

Hire mg Ni/m30.046 0.242 0.44 2.01

1924 0.03 < < < <

1928 0.24 > < < <

1940 0.20 > < < <

1941 0.94 > > > <

< indicates that the Table 4 value is less than the distribution value

> indicates that the Table 4 value is greater than the distribution value

As shown, the NIOSH mean concentration is favorable for only one of the four workers (or 25% 

of the time), while the SC&A mean and the NIOSH 95th percentile are favorable for three of the 

four workers (or 75% of the time). Only the SC&A 95th percentile concentration is favorable for 

all four of the Enterline and Marsh 1982, Table 4, average exposure concentrations.

Observation 5: When appropriately used, a site airborne nickel concentration distribution can 

be used to make favorable exposure estimates when compared to the individual worker locationspecific estimates made by Enterline and Marsh (1982, Table 4).
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4.0 SHALLOW DOSE – TYPOGRAPHICAL ERRORS

Finding 7: There are three typographical errors in the numerical values given in Section 6.2 of 

OCAS 2008. Despite the erroneous numerical values, the annual doses are reported correctly, 

thus SC&A has characterized them as “typographical,” rather than “numerical” errors. 

Nonetheless, because the erroneous numerical values make it difficult to understand how the 

annual doses were calculated, SC&A has identified these three typographical errors as a finding 

rather than an observation.

Each of the numerical typographical errors is described below.

4.1 UCNC 1958 CONTACT BETA DOSE RATE

Section 6.2 states that UCNC (1958) “estimated an upper bound [beta] dose rate from the 

residues at 0.024 mrep/hour on contact”. The full quotation from UCNC 1958 is:

The beta dose rate at the surface of an effectively infinite slab of normal uranium 

is 240 mrep/hr., and the dose rate at the surface of material with a uniform 

daughter-product concentration corresponding to 1000 ppm. of uranium (or 1 

part per 1000) would accordingly be 1/1000 of this, or 0.24 mrep/hr.

With the production of 50 pounds of residue from a 4000-lb. batch of material, the 

daughter products will be concentrated by a factor of 80, with a resulting dose 

rate of approximately 20 mrep/hr. This beta dose rate, which is the maximum that 

can be obtained from the material described, would be found only at the surface 

of the residue, and the beta radiation would be reduced to essentially zero by the 

shielding effect of the drum in which the material will be stored.

Thus, Section 6.2 should indicate that UCNC (1958) “estimated an upper bound dose rate from 

the residues at 20. mrep/hour on contact.” As described below, a contact beta dose rate of 

20 mrep/hr produces the calculated annual doses reported elsewhere in Section 6.2 and Table 6, 

whereas a contact dose rate of 0.024 mrep/hr produces significantly different annual beta doses.

4.2 ANNUAL DOSE RATE TO HANDS AND FOREARMS

Next, the fourth sentence of the second paragraph of Section 6.2 states that “some workers could 

have had direct contact with the residues for fifty hours per year resulting in a shallow dose to 

the hands and forearms of 1.2 rem per year.” Instead of 1.2 rem/yr, the correct product of 

20 mrep/hr times 50 hr/yr is 1.0 rem/yr, as is correctly shown in Table 6 (OCAS 2008, page 17).

4.3 HOURLY DOSE RATE AT 30 CENTIMETERS

The third sentence of the third paragraph of Section 6.2 states that the “contact dose rate 

estimated for the residues is ... divided by 75 to determine beta dose rates at 30 cm from the 

surface of the residues, resulting in a beta dose rate of 0.00027 mrem per hour.” The quotient of 

dividing 20 mrep/hr by 75 is 0.27 mrem/hr, rather than 0.00027 mrem/hr. Although the 30-cm 
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beta dose rate given in the text is 3 orders of magnitude too small, the calculated annual beta 

doses are correct. For example, with an exposure duration of 2,000 hours per year, the annual 

beta dose would be [0.27 mrem/hr × 2,000 hr/yr × 0.001 rem/mrem] = 0.54 rem/yr, as reported in 

both Section 6.2 and Table 6.
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5.0 SUMMARY OF FINDINGS AND OBSERVATIONS

The findings and observations that have been made in the preceding discussion are summarized 

in the following subsections. Generally, in order to resolve a finding, SC&A expects that 

NIOSH will need to make a change or correction or provide an explanation to OCAS-TKBS0004 (OCAS 2008). However, an observation does not necessarily require any action or 

response by NIOSH.

5.1 FINDING 1 – RECYCLED URANIUM ISOTOPES OF CONCERN

Since the three diffusion plants (the source of the HPP nickel) had additional isotopes of 

concern, NIOSH should clearly provide the basis for only specifying Pu-239 and Np-237 as 

isotopes of concern for recycled uranium.

5.2 FINDING 2 – URANIUM SPECIFIC ACTIVITY

NIOSH should clearly state which uranium specific activity was used in the analysis and ensure 

that it was used consistently throughout the analysis.

5.3 FINDING 3 – UNIT CONVERSION

There is a unit conversion error in going from Table A2 column 3 (Photons per decay 238U) to 

column 4 (Photons per second per Ci 238[U]). This same unit conversion error also appears in 

Table A3 for the beta spectrum.

As explained above, this unit conversion error only affects the values presented in Tables A2 and 

A3, and is not carried through to the dose rate calculations.

5.4 OBSERVATION 1 – GAMMA ENERGY SPECTRUM

Consistent with HPP operations, the gamma energy spectrum used by NIOSH is not based on 

full equilibrium of U-238, U-235, and U-234 with their decay products. Rather it is consistent 

with a decay period where only short-lived progeny of U-238 and U-235 would have had an 

opportunity to grow in.

Although SC&A found that the Table A2 gamma energy spectrum is consistent with a decay 

period of 50 to 100 years, NIOSH may want to consider identifying in OCAS-TKBS-0004 

(OCAS 2008) the specific decay period that was used, with an explanation as to why that period 

was selected.

5.5 OBSERVATION 2 – BETA ENERGY SPECTRUM

Consistent with HPP operations, the beta energy spectrum used by NIOSH is not based on full 

equilibrium of U-238, U-235, and U-234 with their decay products. Rather it is consistent with a 

decay period where only short-lived progeny of U-238 and U-235 would have had an 

opportunity to grow in.
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5.6 OBSERVATION 3 – RESIDUE STORAGE CONTAINER

In the current site profile (OCAS 2008), NIOSH assumed that residues were stored in 20 gallons 

drums, while in the previous site profile (ORAUT 2004a) it was assumed that the residues were 

stored in “birdcages”. The basis for this change in assumption is not provided in OCAS 2008.

5.7 OBSERVATION 4 – PENETRATING DOSE RATE

The penetrating dose rate calculated by NIOSH using MCNPX and modeling an array of drums 

is consistent with the dose rate calculated by SC&A using MicroShield and modeling a single 

drum. Furthermore, the MicroShield calculated dose rate using the Table A2 gamma energy 

spectrum is consistent with the dose rate calculated using the gamma energy spectrum from the 

MicroShield radionuclide decay library.

5.8 FINDING 4 – DOSE ENERGY DISTRIBUTION

The dose breakdown between 0–250 keV and >250 keV varies from 50/50 to about 70/30, 

depending on the gamma spectrum.

5.9 FINDING 5 – INCLUDING MODERN AIRBORNE NICKEL 

CONCENTRATIONS

Provide justification for including modern airborne nickel concentrations in the concentration 

distribution, when Enterline and Marsh 1982 indicate that the historical concentrations were (in 

most cases) of greater magnitude.

5.10 FINDING 6 – EXCLUDING DATA FROM THE CRUSHING, GRINDING, 

HANDLING, AND CALCINER AREAS

Provide justification for excluding from the concentration distribution the airborne nickel 

concentration in the crushing, grinding, and handling areas and the area around the calciners 

reported by Enterline and Marsh (1982).

5.11 OBSERVATION 5 – USE OF SITE CONCENTRATION DISTRIBUTIONS

When appropriately used, a site airborne nickel concentration distribution can be used to make 

favorable exposure estimates when compared to the individual worker location-specific 

estimates made by Enterline and Marsh (1982, Table 4).

5.12 FINDING 7 – NUMERICAL TYPOGRAPHICAL ERRORS

There are three typographical errors (identified above in Section 4) in the numerical values 

given in OCAS-TKBS-0004 (OCAS 2008), Section 6.2. Despite the erroneous numerical values, 

the annual doses are reported correctly, thus SC&A has characterized them as “typographical,” 

rather than “numerical” errors. Nonetheless, because the erroneous numerical values make it 
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difficult to understand how the annual doses were calculated, SC&A has identified these three 

typographical errors as a finding rather than an observation.
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REFERENCES

BNL 2013. National Nuclear Data Center, Brookhaven National Laboratory, 

http://www.nndc.bnl.gov/chart/, Last accessed May 3, 2013.

DOE 2004. DOE Standard, Guide of Good Practices for Occupational Radiological Protection 

in Uranium Facilities, DOE-STD-1136-2004, Washington, DC. December 2004.

Enterline P.E, and G.M. Marsh, 1982. Mortality Among Workers in a Nickel Refinery and Alloy 

Manufacturing Plant in West Virginia, Journal of the National Cancer Institute, 1982; 6:925-933.

Grove 2009. MicroShield®, User’s Manual, Grove Software, Inc., Version 8.03, Copyright 

1992–2009.

LANL 2008. MCNPX User’s Manual, Version 2.6.0, Los Alamos National Laboratory, LA-CP07-1473, April 2008.

OCAS 2008. Technical Basis Document for the Huntington Pilot Plant, Huntington, West 

Virginia, OCAS-TKBS-0004, Rev. 00. Office of Compensation Analysis and Support, National 

Institute for Occupational Safety and Health, Cincinnati, Ohio. August 13, 2008.

ORAUT 2004a. Technical Basis Document: Basis for Development of an Exposure Matrix for 

Huntington Pilot Plant, Rev. 01, ORAUT-TKBS-0004. Oak Ridge Associated Universities 

Team, Cincinnati, Ohio. January 16, 2004.

ORAUT 2004b. Technical Basis Document for Portsmouth Gaseous Diffusion Plant –

Occupational Internal Dose, ORAUT-TKBS-0015-5, Rev. 00. Oak Ridge Associated 

Universities Team, Cincinnati, Ohio. November 24, 2004.

ORAUT 2006. K-25 Gaseous Diffusion Plant - Occupational Internal Dose, ORAUT-TKBS0009-5, Rev. 01 PC-1. Oak Ridge Associated Universities Team, Cincinnati, Ohio. October 4, 

2006.

ORAUT 2012. Paducah Gaseous Diffusion Plant – Occupational Internal Dose, ORAUTTKBS-0019-5, Rev. 03. Oak Ridge Associated Universities Team, Cincinnati, Ohio. August 24, 

2012.

SC&A 2008. “Audit of the Eighth Set of 30 Dose Reconstruction Cases,” Contract No. 200-

2004-03805, Task Order No. 4. Attachment 3: Review of the Huntington Pilot Plant Technical 

Basis Document. S. Cohen & Associates, Vienna, Virginia. May 30, 2008.

SC&A 2013. “Preliminary Review of the Revised Site Profile for the Huntington Pilot Plant,” 

Contract No. 200-2009-28555, S. Cohen & Associates, Vienna, Virginia. March 2013.

Stanford 2013. RAdiation Dose Assessment Resource (RADAR), Stanford Dosimetry, LLC, 

http://www.doseinfo-radar.com/RADARHome.html, Last accessed May 3, 2013.
Effective Date:

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

UCNC 1958. H. F. Henry, to E. C. Bollinger, Union Carbide Nuclear Company, Oak Ridge 

Gaseous Diffusion Plant, “Study of INCO Radiation Problem,” January 22, SRDB Ref ID 

18863.