NIOSH Responses to SC&A Review of 

ORAUT-TKBS-0041, Rev. 02: 

Site Profile for Nuclear Materials and 

Equipment Corporation, Apollo and 

Parks Township, Pennsylvania
 

Response Paper 

Rev. 0-A 

National Institute for Occupational 

Safety and Health 

May 14, 2015 

Dennis Strenge 

Oak Ridge Associated Universities Team
 

-

Reviewed by Jim Neton, Lara Hughes
 

Division of Compensation Analysis and Support
 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 1 of 19 
OVERVIEW
 

In April of 2013, S. Cohen & Associates (SC&A) completed a review of the NIOSH Site Profile 

for Nuclear Materials and Equipment Corporation (NUMEC) Parks Township and Apollo Sites 

(ORAUT 2012a) at the request of the Advisory Board on Radiation and Worker Health (Advisory 

Board). Twenty one findings were reported in the SC&A report. Responses to these findings 

are provided in this response paper. 

NIOSH RESPONSES TO SC&A PRIMARY FINDINGS 

The following presents SC&A’s primary findings for the NUMEC Apollo and Parks Township site 

profile, ORAUT-TKBS-0041 (ORAUT 2012a) and NIOSH responses. 

FINDING 1: CLARIFICATION IS NEEDED ABOUT THE START AND END DATES OF 

PARKS TOWNSHIP SITE OPERATIONS. 

NIOSH Response: The dates provided in the site profile have been reviewed and updates will 

be made in the next revision of the Technical Basis Document (TBD). The following table 

indicates the proposed revisions (changes underlined). 

Table 2-3. Parks Township site area descriptions. 

Building or area Operations Radionuclides 

Operation 

period 

Building A–Plutonium 

processing facility 

Fabrication of plutonium 

reactor fuel pellets, 

blankets, rods 

PuO2; Pu nitrate and oxalates 

(AmBe, PuBe, 1959–1970); 

alpha, beta, and gamma sources 

1961–1980 

A–Fab 1 

Plutonium conversion Plutonium nitrate, plutonium 

oxide, depleted UO2 

1961–1980 

Fuel fabrication for FFTF Plutonium nitrate, plutonium 

oxide, depleted UO2 

1972–1980 

A–East Side of Fab 1 

Routine repair and 

maintenance of 

contaminated equipment 

All 1961–1980 

A–Fab 2 

Fuel fabrication for 

ZPR-III 

Plutonium nitrate, plutonium 

oxide, depleted UO2 

1964–1966 

Fuel fabrication for ZPPR Plutonium nitrate, plutonium 

oxide, depleted UO2 

1966-1970 

A–Fab 3 

Manufacturing 

operations, 

metallography 

Plutonium nitrate, plutonium 

oxide 1963–1980 

Quality control of FFTF 

fuel 

Plutonium nitrate, plutonium 

oxide 1972–1980 

A–Fab 4 

Alpha, beta, gamma, and 

neutron source 

fabrication 

AmBe double encapsulated, 

PuBe compacted powder, 

Ir-192, Cs-137, Be-7, Po-210, 

Co-60, Am-241, Pu-238, 

Pu-239, PuO2, plutonium and 

americium metal 

1963–1980 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

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Page 2 of 19 
Building or area Operations Radionuclides 

Operation 

period 

A–Fab 5 

Scrap recovery Plutonium – various forms 1963– 

08/01/1967 

Analytical laboratory work All, small quantities of 

radioactive samples 1979–1980 

A–Fab 6 Scrap recovery Plutonium – various forms 1968–1980 

A–Fab 7 

Fuel rod quality control 

tests, nonradioactive 

processes 

All, clean and contaminated 

items 1968–1980 

A–Fab 8 Storage All, clean and contaminated 

items 1970–1980 

A–Fab 9 FFTF fuel pin finishing Encapsulated nuclear material 1970–1980 

Building A–Hot Cell 

Room 

Examination of irradiated 

samples, high-activity 

source fabrication 

PuBe compacted powder, 

Co-60, Ir-192 1961–1969 

Storage of sources 

Sealed sources, Ir-192, Cs-137, 

Be-7, Po-210, Co-60, Am-241, 

Pu-238, Pu-239, PuO2, 

1969–1980 

Plutonium Facility 

Trailer Storage Area 

Storage of large 

quantities of plutonium 

and uranium 

Plutonium (nonpyrophoric), DU, 

NU, and EU (to 5% 235U) 1961–1980 

Building B– 

Multipurpose 

fabrication building 

DU, NU, thorium, 

plutonium 

DU metal or alloy, U3O8, Pu-238, 

ThO2 

1961–1980 

Building B 

Hafnium Plant Metal production Hafnium and Zr-Be alloy (nonradioactive) 1961–1980 

Building B 

Plutonium Annex 

Conversion of Pu-238 

nitrate to oxide 

Pu-238 nitrate, Pu-238 oxalate, 

PuO2 powder or alloys 1963–1980 

Building B–Hot Cell 

Room 

Large source production: 

Postirradiation 

examination of test 

capsules and fuel pins 

Co-60, Cs-137, Ir-192, and 

PoBe, irradiated, uranium and 

plutonium and other TRU 

elements and FPs 

1961–1980 

Building B–Metals 

Plant 

First floor: small-scale 

metals production, fuel 

pellet production, 

materials testing 

First floor: DU, UO2, U3O8, UF4, 

fully clad U-233, U-235, and 

Pu-239 

1963–1974 

Second floor: Pu-238 

pacemakers 

Second floor: Pu-238-powered 

heart pacemakers 1963–1970 

Building B–Machine 

Shop 

Occasional machining of 

clad or unclad uranium, 

and clad plutonium and 

U-233; fabrication and 

repair of new and 

contaminated equipment 

from Parks Township and 

Apollo; machining of DU 

Fully clad U-233 and Pu-239 

and clad or unclad U-235 (any 

enrichment), primarily DU 

contamination; could include 

HEU, plutonium, thorium, and 

mixed FPs 

1964–1980 

Building C, Type II 

Facility or T-2 Plant 

HEU processing to form 

sintered product 

HEU (1973–1978), soluble 

chloride/oxide complexes, SNM 

oxides (UO3, UO2 and U3O8) 

1973–1978 

Outdoor Scrap 

Storage Area Storage UF6 cylinders 1971–1980 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 3 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

FINDING 4: URANIUM INHALATION RECOMMENDATIONS FOR THE APOLLO SITE NEED 

TO TAKE INTO CONSIDERATION THE METHOD DISCUSSED BY DAVIS AND STROM 

(2008) FOR DEALING WITH UNCERTAINTIES IN DAILY WEIGHTED EXPOSURE (DWE). 

THIS TECHNIQUE WAS EVALUATED AND FOUND TO BE APPROPRIATE IN THE SITE 

PROFILE REVIEW FOR THE FERNALD FEED MATERIALS PRODUCTION CENTER 

(FERNALD). 

The updates were based on review of current and new information obtained in the February 

2015 data capture effort. Reports from the data capture effort include Shields 1990, 

Scarlata 1996, and B&W 1995. 

FINDING 2: THE SITE PROFILE SHOULD PROVIDE GUIDANCE ON WHAT LEVEL OF 

URANIUM ENRICHMENT SHOULD BE ASSUMED FOR THOSE URINE BIOASSAY 

RESULTS THAT ARE EXPRESSED IN UNITS OF MICROGRAMS (G/L) PER LITER. 

NIOSH Response: A new section (Section 5.2.2.4, Uranium Fluorimetric Bioassay Evaluations) 

will be added to the site profile to provide guidance on conversion from mass units to activity 

units, as follows. 

5.2.2.4 Uranium Fluorimetric Bioassay Evaluations 

The evaluation of intakes based on uranium fluorimetric bioassays requires conversion from 

units of mass to activity. This requires information on the uranium-235 enrichment level 

because the specific activity of the uranium material varies with enrichment. If the enrichment is 

known for the exposed worker, the specific activity for that enrichment should be used as 

defined in Table 5-2. If the enrichment is not known, an estimate of activity that is favorable to 

claimants can be made by assuming the material to be highly enriched (93%). This is a 

reasonable approach because highly enriched uranium was used frequently at the Apollo 

facilities in the early years when urine bioassays were performed by fluorimetric analysis. 

If the bioassay results contain both fluorimetric and radiometric results, the radiometric results 

should be used as the radiometric analysis method provided a more sensitive estimate of 

uranium activity and no conversion of units is required. 

FINDING 3: SOME GUIDANCE IS NEEDED ON HOW TO PERFORM DOSE 

RECONSTRUCTIONS PRIOR TO 1959, AND WHAT APPROACH SHOULD BE USED FOR 

MISSED AND UNMONITORED EXPOSURES. 

NIOSH Response: The evaluation of external and internal dose for NUMEC sites has been 

addressed in the Special Exposure Cohort (SEC) and was found to be infeasible. Therefore, 

evaluation of missed and unmonitored exposures is not feasible. Partial dose reconstructions 

for non-presumptive claims are assigned based on the guidance in the TBD. 

NIOSH Response: Davis and Strom (2008) evaluated several HASL reports and concluded that 

the daily weighted exposures have a lognormal distribution with a geometric standard deviation 

of 5. HASL data is available for the Apollo site and has been used to evaluate exposure 

concentrations. As discussed in Finding 18, the air concentration used for the beginning of the 

residual period has been re-evaluated and is now based on General area (GA) monitoring 

rather than daily weighted average exposures (see response to Findings 5 and 18 below). The 

Page 4 of 19 
 

geometric standard deviation (GSD) for the revised air concentrations has been set to 5.0 based 

on Davis and Strom (2008). 

FINDING 5: INADEQUATE INFORMATION IS GIVEN TO REPLICATE NIOSH’S 

DETERMINATION OF MEDIAN INHALATION CONCENTRATION OF URANIUM. THE NIOSH 

RESULT COULD NOT BE REPLICATED, AND IT APPEARS THAT RELEVANT 

INFORMATION HAS BEEN OMITTED FROM THE HEALTH AND SAFETY LABORATORY 

(HASL) STUDIES REPORTED IN APPENDIX A TO THE SITE PROFILE. 

NIOSH Response: All data listed in the HASL tables has been used in the NIOSH evaluation 

(reported in the Table II entries in the HASL reports). The median concentration was evaluated 

as described in the Site Profile based on the assumed 95% (maximum = 6300 dpm/m3) and 5% 

(minimum = 7 dpm/m3) values. The median is evaluated as the square root of the product of the 

maximum and minimum values. This method assumes the distribution to be lognormal. The 

actual distribution is skewed to high values above the median. A re-evaluation of all of the data 

in the HASL reports indicates the estimated value of 210 dpm/m3 is slightly smaller than the 

arithmetic mean (286 dpm/m3) but much greater than the median based on a review of all data, 

using actual 95% (910 dpm/m3) and 5% values (9 dpm/m3) which gives a median of 90 dpm/m3. 

Using the 1 sigma values, a median value of 59 dpm/m3 is obtained. Therefore, the value 

reported is favorable to claimants. When using a lognormal distribution, the correct value to use 

is the median and not the arithmetic mean. 

The SC&A discussion indicated that a concentration of 563 dpm/m3 (Schnell 1966) is not 

reflected in the NIOSH median air concentration of 210 dpm/m3. The value of 563 dpm/m3 was 

taken on September 21, 1966 (most likely date), and represents a small period. The memo by 

Schnell was written to demonstrate that the air concentrations in the locker room were 

becoming too high. Data presented indicated that from late August to late September 1966, the 

locker room air concentration increased from 16 dpm/m3 on August 25, 1966, to 321 dpm/m3 on 

September 20, 1966, with the 563 dpm/m3 value being reported in the memo of September 22, 

1966. The memo suggested that the high air concentrations were a result of workers tracking 

contamination into the locker room on clothing. Remedial actions were suggested to reduce the 

locker room air concentrations. The value of 563 dpm/m3 is a value that occurred over a short 

period of time and does not represent an average that would be representative of the general air 

concentration. Also, the NIOSH value of 210 dpm/m3 and GSD of 7.91 can be used to estimate 

the 84th percentile of the distribution as 1661 dpm/m3, which is much higher than the Schnell 

value. Therefore, the Schnell value is well represented in the NIOSH analysis when the 

lognormal distribution is considered. 

Based on Finding 18, the concentration used for the residual period should be based on general 

air monitoring and not on breathing zone values (as was used for the HASL studies). Based on 

this finding, the air concentration has been re-evaluated using GA monitoring results (see 

Finding 18). The revised air concentration will be used to update the site profile. 

FINDING 6: THE SITE PROFILE WOULD BENEFIT FROM A DISCUSSION 

DEMONSTRATING THAT THE HANFORD SITE FUEL GRADE MIX, AS OPPOSED TO THE 

WEAPONS-GRADE OR COMMERCIAL-GRADE PLUTONIUM, IS LIMITING FOR THE FULL 

RANGE OF PLUTONIUM MIXES AND AGES THAT WERE USED AT NUMEC. IN ADDITION, 

GIVEN THE COMPLEXITY OF THIS SUBJECT, A REVIEW OF ACTUAL DOSE 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 5 of 19 
 

 

 

 

 

 

RECONSTRUCTIONS WOULD PROVIDE GREATER INSIGHT INTO HOW THIS MATTER IS 

ACTUALLY BEING ADDRESSED. 

NIOSH Response: The Fast Flux Test Facility (FFTF) program used only 12% plutonium-240 

fuel, so that fuel grade should be used when it is known that the worker was involved with FFTF 

fuel fabrication. The Zero Power Plutonium Reactor (ZPPR) fuel program used primarily 12% 

plutonium-240 (11,500 fuel plates), but also used some 27% plutonium-240 (700 fuel plates). 

The 27% plutonium-240 material was not received on the site until about April of 1969 based on 

shipping receipt records (NUMEC 1968–1970). 

Additional information was obtained from a review of new documents provided by BWTX in 

February 2015. A fuel fabrication contract with Japan (PNC abbreviation in NUMEC 

documents) was performed from 1966 to 1970. A fuel fabrication contract with Argonne 

National Laboratory for preparation of fuel wafers for the Zero Power Reactor (ZPR-III 

abbreviation) was performed from 1964 to 1966. Both of these contracts involved fuel with a 

plutonium-240 content of 8.1% to 8.5%. Based on this information, Table 5-3 has been revised 

to include the following four plutonium types: Hanford 6% weapons grade, DOE 8.5% material, 

Hanford 12% fuel grade, and 27% commercial fuel. The Hanford and commercial specific 

activity data is from Table 5-3 through Table 5-4 of the Internal Dosimetry section of the Hanford 

TBD. The specific activity of the 8.5% material was based on the reported activity of 91.47% for 

the sum of the plutonium-239 and -241 activities (fissile material). The activities of 

plutonium-238, -241, and -242 were estimated by ratio of the isotopic content between the 6% 

and 12% inventory values. The plutonium-240 content was set at 8.3% as 100%–91.47%. The 

activity of americium-241 was based on the ratio of americium-241 to plutonium-241 in the 

Hanford fuels. This ratio was a constant for the two Hanford fuels, as expected because the 

americium-241 is a decay product of the plutonium-241. 

Table 7-4 will be updated to include the commercial reactor-grade fuel mix fractional activity 

data. Guidance will be added regarding selection of the appropriate inventory component for 

evaluation of internal doses based on available information. 

The revised plutonium isotopic table is as follows. 

Table 5-3. Activity composition of plutonium mixtures. 

Mixture 

designation Hanford 6% Weapons-grade Plutonium Specific activity (Ci/g)a 

Years of agingb 0 5 10 15 20 

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 

Am-241 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 6 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

Mixture 

designation DOE 8.5% Plutonium Specific activity (Ci/g)c 

Years of agingb 0 5 10 15 20 

Pu-238 1.22E-02 1.17E-02 1.12E-02 1.08E-02 1.04E-02 

Pu-239 5.55E-02 5.55E-02 5.55E-02 5.55E-02 5.55E-02 

Pu-240 1.94E-02 1.94E-02 1.94E-02 1.94E-02 1.94E-02 

Pu-241 1.78E+00 1.40E+00 1.10E+00 8.64E-01 6.80E-01 

Pu-242 2.80E-06 2.80E-06 2.80E-06 2.80E-06 2.80E-06 

Am-241 0 1.26E-02 2.25E-02 3.00E-02 3.58E-02 

Hanford 12% Fuel-grade Plutonium Specific activity (Ci/g)a 

Years of agingb 0 5 10 15 20 

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 

Am-241 0 2.19E-02 3.89E-02 5.22E-02 6.24E-02 

Commercial 27% Fuel-grade Plutonium Specific activity (Ci/g)a 

Years of agingb 0 5 10 15 20 

Pu-238 1.71E-01 1.64E-01 1.58E-01 1.52E-01 1.46E-01 

Pu-239 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 

Pu-240 5.90E-02 5.89E-02 5.89E-02 5.89E-02 5.89E-02 

Pu-241 1.34E+01 1.05E+01 8.28E+00 6.51E+00 5.12E+00 

Pu-242 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 

Am-241 0 9.49E-02 1.69E-01 2.26E-01 2.79E-01 

a. Source: ORAUT (2012b). 

b. Time since separation of Am-241 from the plutonium mix. 

c. NUMEC records (Gerrish 1965). 

FINDING 7: THE MINIMUM DETECTABLE ACTIVITIES (MDAS) FOR AM-241 LUNG 

COUNTING ARE VERY LOW. THE COUNTING METHOD SHOULD BE FURTHER 

EXPLORED IN ORDER TO GIVE THEM CREDIBILITY. 

NIOSH Response: The values listed for americium-241 are consistent with contemporary 

minimum detectable activity (MDA) values for other sites, such as Hanford. For example, both 

sites used sodium iodide crystal detectors in the late 1960s; the Hanford americium-241 MDA 

is given as 0.33 nCi and the value for NUMEC is listed as 0.13 to 0.38 nCi. 

The MDA values in the NUMEC site profile are listed as minimum and maximum values. 

Because the detection limit is a function of body size and chest wall thickness, several data 

points are needed to estimate a reasonable upper bound on the detection limit. The MDA to be 

used in dose reconstructions (when a detection limit is not provided in the bioassay records) 

should account for large persons and provide an intake estimate that is favorable to the 

claimant. The variation in MDA with chest wall thickness that was reported for the University of 

Pittsburg system in 1975 is as follows (Author unknown 1977): 

Chest wall thickness MDA Pu-239 (nCi) 

 1 cm 3–4 

2 cm 9–10 

 3 cm 17–20 

Page 7 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

The review of the February 2015 BWXT data capture documents for MDA values for lung 

counting resulted in minor changes to the previous data. The changes are indicated in the 

following proposed revision to Table 5-9 in the NUMEC site profile (AEC 1974a, 1974b, 

Caldwell 1969a). 

Table 5-9. In vivo MDAs for plutonium-239 and americium-241.

Year 

Pu-239 MDA (nCi) Am-241 MDA (nCi) 

Minimum Maximum Counts Minimum Maximum Counts 

1968 NRb

 NRb NRb 0.13 0.38 17

1969 NRb

 NRb NRb 0.16 0.16 1 

1970 10 10 1 NRb

 NRb NRb

1971 NRb NRb NRb NRb NRb NRb

1972 9.0 11.5 3 0.13 0.13 1

1973 5.6 15.6 46 0.11 0.21 28

1974 5.44 21.3 122 0.06 0.22 96

1975 4.8 19.9 133 0.11 0.21 104

1976 5.0 20.3 109 0.11 0.19 91

1977 4.4 19.6 113 0.09 0.19 88

1978 4.7 19.0 132 0.10 0.19 100

1979 5.16 24.3 168 0.08 0.26 132

1980 5.03 28.2 132 0.09 0.21 94

1981 7.21 27.8 55 0.12 0.20 31

1982 7.12 34.3 77 0.12 0.21 44

1983 9.41 15.6 6 0.12 0.16 4

1984 8.67 22.32 9 0.12 0.15 5

1985 8.84 31.07 31 0.11 0.22 29

a. From a review of worker dosimetry records (AEC 1974a, 1974b, Caldwell 1969a, Boyd 2006a, 2006b, 

2006c, 2006d, 2006e, 2006f). Values for 1968 through 1971 are based on the Helgeson system, with 

remaining values for the University of Pittsburgh system. 

b. NR = none reported. 

Because the data is sparse for the early years, the upper bound of the MDA is not likely 

reflected in the data. Based on the above information, guidance will be added to the site profile 

directing use of the MDA values in the following table when the bioassay records do not provide 

the MDA. 

Period Pu-239 MDA (nCi) Am-241 MDA (nCi) 

1968–1985 35 0.40

a 

FINDING 8: THE SITE PROFILE WOULD BENEFIT FROM A MORE THOROUGH 

DISCUSSION OF THE POSSIBLE USE OF AIR SAMPLING DATA TO RECONSTRUCT 

INTERNAL PLUTONIUM EXPOSURES AND TO TAKE INTO CONSIDERATION THE 

ADDITIONAL DATA PROVIDED BY CROSBY 1967 AND NUMEC 1967. 

NIOSH Response: The two references give results of air sampling in the plutonium scrap 

recovery area of Parks Township. The evaluation of internal dose for NUMEC sites has been 

addressed in SECs and was found to be infeasible. Therefore, evaluation of internal plutonium 

exposures for claims with no bioassay monitoring data is not feasible. Internal plutonium 

exposures can only be included in dose reconstructions when valid bioassay monitoring data is 

available. Air sampling data can only be used when it represents sampling for the energy 

employee being evaluated in a claim. Such data has been used in past dose reconstructions. 

Page 8 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

The records typically provide the exposure time and air concentration which can be used to 

estimate an intake. The internal dose from the intake can be evaluated using the IMBA 

program. Partial dose reconstructions for non-presumptive claims are assigned based on the 

guidance in the site profile. 

FINDING 9: IT DOES NOT SEEM APPROPRIATE TO USE ORAUT-OTIB-0054 (ORAUT 

2007a) TO RECONSTRUCT THE INTERNAL EXPOSURES OF WORKERS AT NUMEC WHO 

MIGHT HAVE BEEN EXPOSED TO MIXED FISSION PRODUCTS BECAUSE ORAUT-OTIB0054 STATES THAT ITS GUIDANCE “DOES NOT APPLY TO DETERMINATION OF 

INTAKES WHERE RADIONUCLIDES HAVE BEEN PURPOSELY EXTRACTED AND 

CONCENTRATED AS FOR HEAT GENERATION SOURCES, MEDICAL USES, OR WASTE 

HANDLING OPERATIONS THAT CAUSED SIGNIFICANT ALTERATION TO THE SOURCE 

TERM TO WHICH WORKERS WERE EXPOSED.” FOR EXAMPLE, TABLE 5-1 OF THE SITE 

PROFILE INDICATES THAT THE FISSION PRODUCTS HANDLED AT NUMEC ARE 

SOURCES USED FOR VARIOUS RESEARCH AND OPERATIONS PURPOSES AND ARE 

NOT ACTUALLY FUEL OR SPENT FUEL AND WOULD NOT NECESSARILY BE PRESENT 

IN THE SAME RATIOS AS IN REACTOR FUEL. ALSO, THE FISSION PRODUCT MIX GIVEN 

IN ORAUT-OTIB-0054 DOES NOT CONTAIN THE SAME RADIONUCLIDES AS THE 

FISSION PRODUCT MIXES GIVEN FOR THE NUMEC LAUNDRY IN THE 1975 EFFLUENT 

RELEASE REPORT (WILLIAMS 1967) AND FOR IN-VIVO COUNT RESULTS IN CALDWELL 

1969B. THE NUMEC MIXES INCLUDE CO-60, WHICH THE ORAUT-OTIB-0054 MIX OMITS, 

FURTHER DEMONSTRATING THE INAPPLICABILITY OF ORAUT-OTIB-0054 TO NUMEC. 

NIOSH Response: Use of OTIB-0054 (ORAUT 2015) is appropriate when the source of activity 

is from reactors. As indicated in this finding, the NUMEC laundry processed materials from 

many sources in addition to the commercial reactor laundry. The site profile will be modified to 

remove guidance on use of OTIB-0054 for evaluation of dose from associated fission products 

and activation products. The guidance will indicate that doses are to be evaluated only for the 

radionuclides included in the bioassay results. 

FINDING 10: INTERNAL DOSE RECONSTRUCTIONS PERFORMED FOR NUMEC 

PERSONNEL MIGHT NEED TO BE REVISITED IN LIGHT OF CHANGES TO THE FERNALD 

SITE PROFILE (ORAUT 2004) WITH RESPECT TO RECYCLED URANIUM (RU). ALSO, 

ADDITIONAL DIRECTION IS NEEDED WITH RESPECT TO WHICH WORKERS OR 

OPERATIONS SHOULD BE ASSIGNED RU INTAKES. 

NIOSH Response: Changes to the Fernald site profile have not been finalized at this time so 

changes to guidance on evaluation of contaminant activity for RU are not yet necessary. The 

source of the uranium used at NUMEC is not known for many of the activities, so the 

assumption of RU for operations involving enriched or depleted uranium is reasonable. 

Operations involving natural uranium would not be expected to have RU contaminants. Current 

guidance for NUMEC is to include the components associated with recycled uranium 

(plutonium-239, neptunium-237, and technicium-99) using activity fractions in Table 5-12 of the 

current (ORAUT 2004) Fernald internal dose TBD. The NUMEC site profile will be updated to 

indicate the need to include all contaminants in the comparison of types F, M, and S uranium 

materials in order to decide which absorption type of uranium is most favorable to the 

claimant. The most claimant-favorable absorption type will also vary, depending on the organ of 

interest. The dose reconstructor is instructed to follow guidance in ORAUT-OTIB-0060 (ORAUT 

2014a) for which material types to use when assigning recycled uranium contaminants. These 

Page 9 of 19 
contaminants must be added to all uranium intakes regardless of the monitoring method (urine, 

chest counts, or breathing zone samples). 

FINDING 11: NIOSH SHOULD EXPLAIN WHETHER THE CONCERNS EXPRESSED IN THE 

PANTEX SITE PROFILE (ORAUT 2007d) ABOUT THE HELGESON CHEST COUNT DATA 

MIGHT ALSO APPLY TO CHEST COUNT DATA AT NUMEC PERFORMED BY HELGESON 

FOR NUMEC WORKERS. 

NIOSH Response: The in vivo counting at NUMEC was performed by Helgeson only during 

1968, 1971, and once in 1975. The Helgeson counts included primarily plutonium-239 with a 

few results for uranium and fission products. The majority of in vivo counting was performed at 

the University of Pittsburg hospital. The issue with Pantex was that the counts for uranium were 

biased high and represented false positives (Brake 1989). The current Pantex internal dose 

TBD has eliminated all reference to Helgeson in vivo measurements. Using the Helgeson 

results for uranium is favorable to the claimant because of the high bias in the reported values. 

There was no issue identified related to Helgeson in vivo counting for fission products, 

plutonium, or americium. No changes to the NUMEC site profile are planned at this time related 

to the Helgeson chest counts for uranium. 

FINDING 12: TABLE 6-2 AND THE ASSOCIATED TEXT IN SECTION 6.3.2 OF THE SITE 

PROFILE SHOULD BE REVIEWED AND MODIFIED AS NEEDED TO CORRECT ANY 

OVERSIGHTS, INCONSISTENCIES, OR ERRORS. 

NIOSH Response: The indium foil criticality dosimeters were not included in Table 6-2 because 

they were not used for routine workplace exposures. They were included only to determine 

dose from a criticality in the event that a criticality occurred. No criticality incidents were 

reported at the NUMEC facilities. The text was reviewed as suggested and the information in 

the text is consistent with information provided in Tables 6-2 and 6-3.. 

FINDING 13: GIVEN OUR UNDERSTANDING THAT IT IS NIOSH’S POSITION THAT 

EXTERNAL EXPOSURES AT THE PARKS TOWNSHIP SITE CAN BE RECONSTRUCTED 

WITH SUFFICIENT ACCURACY, IT APPEARS THAT THE DESCRIPTION OF THE 

SOURCES AND CIRCUMSTANCES RESPONSIBLE FOR EXTERNAL EXPOSURES NEED 

TO BE BETTER DEVELOPED, IF POSSIBLE. 

NIOSH Response: External dose can be evaluated only when dosimetry records exist. There 

was one dosimetry department for all NUMEC facilities that provided dosimetry for both the 

Parks Township and Apollo facilities. The limitations stated for the Apollo facilities also apply to 

the Parks Township facilities (see response to Finding 16). 

FINDING 14: THE SITE PROFILE SHOULD PROVIDE JUSTIFICATION FOR WHY 

ADJUSTMENT FACTORS ARE NOT REQUIRED FOR NEUTRON EXPOSURES ESTIMATED 

USING NUCLEAR TRACK EMULSION TYPE A (NTA) FILM, CONSIDERING THAT IT 

APPEARS THAT THE NEUTRON ENERGY SPECTRUM LIKELY EXTENDED TO WELL 

BELOW 1 MEGA-ELECTRONVOLT (MEV). FOR EXAMPLE, TABLE 6-8 OF THE SITE 

PROFILE INDICATES THAT THE ENERGY RANGE OF NEUTRON EXPOSURES 

EXTENDED FROM 0.1 TO 2 MEV. 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 10 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Neutron-to-Photon Ratio on Personnel Dosimeters 

Geometric Mean 0.34 

GSD 1.71

NIOSH Response: NUMEC used NTA film to monitor for neutron exposure until approximately 

June 1968 when thermoluminsecent dosimeters (TLDs) were introduced at the site. A study 

was performed in 1975 using TLDs to determine the photon and neutron doses while working in 

the FFTF fuel fabrication area of the Parks Township Plutonium Facility. Additionally, 

dosimeters were placed at various fixed locations. The results of this study resulted in the 

following estimate of neutron to photon ratio (Author unknown 1977): 

This ratio is supported in the Parks Township Plutonium Facility 1979–1981 Health and Safety 

ALARA Reports (Corridoni 1982) which indicate neutron to photon ratios varied from 0.23 to 

0.42 with an average of 0.33. 

Neutron-to-Photon Ratio Glovebox Dosimetersa 

Geometric Mean 1.00 

GSD 1.49 

a. The difference between the ratios is likely a result of shielding on the gloveboxes 

that would reduce the photon component reaching the dosimeter. 

In addition, information was obtained from a September 1968 event where a worker involved 

with manufacturing neutron sources had a neutron-to-photon ratio of 2.33 (determined using 

estimated neutron dose values) (Caldwell 1968). A review of neutron-to-photon ratios from 

neutron sources was evaluated and a broad range of ratios were found. Without additional 

information to indicate otherwise, the project determined ratio of 2.33 will be used. 

It is proposed that NUMEC workers, prior to 1969, with neutron dosimetry receive a neutron 

dose assigned using the ratio most appropriate for their work and job location (i.e., use a ratio of 

0.34 for typical workers, a ratio of 1.00 for glovebox workers, and a ratio of 2.33 for workers 

involved with manufacturing sources). This dose would be assigned using a lognormal 

distribution with the GSD provided, except for the manufacturing source ratio which would be 

applied as a constant. If the worker’s recorded neutron dose is higher than the neutron dose 

calculated from applying the ratio to the photon dose, the recorded neutron dose should be 

assigned. 

This approach will be included in the NUMEC site profile. 

FINDING 15: THE MARKEDLY DIFFERENT PHOTON ENERGIES ASSOCIATED WITH THE 

OPERATIONS AT NUMEC WOULD INDICATE THE POSSIBLE NEED FOR ADJUSTMENT 

FACTORS FOR THE RESULTS OF FILM BADGE DOSIMETERS, WHICH ARE NOT 

PROVIDED IN THE SITE PROFILE. IT IS UNCLEAR WHAT THE AUTHORS INTENDED BY 

DESCRIBING ALL BETA ENERGIES FOR AM-241 AS BEING >15 KEV, AS AM-241 AND U233 ARE ALPHA EMITTERS. 

NIOSH Response: Film badge dosimeters, while over responding to radiation recorded in the 

open window, may under respond to low energy photons (16 keV and 59 keV photons are a 

particular concern) (Wilson et al. 1990). Although the films and filters at NUMEC were different 

than the dosimeters discussed in the reference, a reasonable comparison between the film 

Page 11 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

dosimeters is expected (AEC 1955). The site acknowledged this deficiency in a 1966 report 

(Caldwell and Judd 1966) and made corrections to the dosimetry to account for this under 

response. Prior to the report being issued (i.e., prior to 1966), to account for under response of 

film dosimetry to low energy photons, the result in the open window should be assigned as 

<30 keV photons for workers at plutonium facilities while the deep dose response is assigned in 

accordance with the worker location. 

This approach will be included in the NUMEC site profile. 

Americium-241 and uranium-233 have no beta decay but do emit many internal conversion 

electrons and Auger electrons which are essentially beta particles. Therefore, reference to beta 

radiations is appropriate in Table 6-5. While some of the electron emissions are less than 

15 keV, the low energy electrons are not as likely as the higher energy electrons to reach the 

skin and cause dose to the skin. Therefore, assigning the electron dose as 100% >15 keV is 

appropriate, except as noted above for plutonium workers. 

FINDING 16: NIOSH SHOULD CONSIDER DEVELOPING A UNIVERSAL COWORKER 

MODEL BASED ON NUMEC CLAIMANT RECORDS, OR SPECIFY A MORE CONSISTENT 

BASIS FOR ASSIGNING EXTERNAL DOSES BEYOND THE MEDICAL X-RAYS 

ASSOCIATED WITH THE SITE. 

NIOSH Response: The NUMEC sites are already covered by an SEC based on external and 

internal infeasibility. Partial dose reconstructions for non-presumptive claims are assigned 

based on the guidance in the NUMEC site profile. 

Both SEC evaluation reports state that external doses are assigned to partial DRs in cases 

where monitoring data is available. The Apollo Evaluation Report states that external dose 

reconstruction is not feasible, but that in some cases available dosimetry data can be used. 

The Parks ER states that external data is generally available, however, it also states that the 

SEC decision was based NIOSH’s determination that internal dose reconstruction was not 

feasible (83.14) and that the external feasibility was not evaluated. This is reasonable because 

the internal and external monitoring programs were performed under the guidance of the same 

occupational safety organization for both sites. The Parks ER also states that any available 

external dosimetry data will be used for partial dose reconstructions. 

There are 89 post SEC claims for Apollo and 27 for Parks, some of these have concurrent 

employment at both sites. Thirty randomly selected claims were reviewed for Apollo and 25 for 

Parks (all post SEC period) for inconsistencies in their dose reconstruction approach. No 

inconsistencies were noted. In cases where no external dosimetry is available during the SEC 

period, no external dose is assigned, only medical X-rays. In cases where dosimetry is 

available, external dose is assigned. However, note that current guidance is to not assign 

medical X-ray dose based on OTIB-0079 (ORAUT 2011). 

There is an issue with potential external unmonitored dose occurring at NUMEC as outlined in 

the Apollo ER. There are claims with internal dosimetry that clearly show that the energy 

employee worked in a radiation area resulting in detectable intakes but no external dosimetry 

data is available. However, this issue was recognized during the SEC evaluation process and 

contributed to the finding that it is not feasible to reconstruct external doses for Apollo (Parks 

was not evaluated, but the issue is most likely the same as was the case with internal). 

Page 12 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Organ(s) Dose assignment Radiation

All except skin, breast, 

testes, and penis 

Listed whole body dose times organ isotropic 

exposure dose conversion factor 

Photons 30–250 keV 

Skin Listed skin dose Electrons >15 keV 

Breast, testes, and penis Listed whole body dose times organ isotropic 

exposure dose conversion factor 

Photons 30–250 keV 

30% of listed skin dose Electrons >15 keV 

Therefore, NIOSH advises against any effort to develop a coworker model for unmonitored 

workers for NUMEC. 

FINDING 17: THE SITE PROFILE SHOULD INCLUDE GUIDANCE FOR DERIVING NONPENETRATING DOSES TO SKIN AND OTHER ORGANS FROM BETA EMITTERS 

ASSOCIATED WITH SURFACE CONTAMINATION DURING THE RESIDUAL PERIOD. 

NIOSH Response: The dose values presented in the site profile for exposure to surface 

contamination are currently assigned as all photon dose (30–250 keV). While this provides a 

favorable to claimant estimate of probability of causation, it does not include consideration of the 

beta radiation contribution to the dose to breast, testes, and penis as described in OTIB-0017 

(ORAUT 2005). The dose conversion factors used in generation of Table 7-1 include all 

radiation. The doses to organs other than skin are primarily from photons and represent the 

deep dose equivalent in rem. This is in contrast to the dose factors presented in BMI 2011 

which are exposure dose factors with 71.8% of the dose from photons <30 keV. The skin dose 

factor from Table 7-1 is primarily from electron radiation and is much higher than the photon 

dose. In the revision to the site profile in response to Finding 20, the external dose from 

residual thorium will be added to Table 7-1. In the revision, Table 7-1 entries will be limited to 

whole body exposure and skin dose. This will make application of the chronic lymphocytic 

leukemia (CLL) evaluations in the workbooks practical because that calculation is performed 

using one dose value and then applying the isotropic exposure dose conversion factors. The 

value given for whole body exposure in the revised Table 7-1 will be such that application of the 

isotropic exposure dose conversion factors provides a correct estimate of organ doses for the 

CLL evaluation and other cancer organs and does not underestimate the dose to any organ. 

The method for assignment of external dose from residual materials at the Apollo Plant will be 

as indicated in the following table. The following is based on guidance in the technical 

information bulletin for shallow dose (ORAUT 2005) and the procedure for occupational onsite 

external dose (ORAUT 2006). This method includes application of the exposure isotropic dose 

conversion factors and results in a best estimate of dose. 

FINDING 18: GENERAL AIR (GA) SAMPLES, AS OPPOSED TO BREATHING ZONE (BZ) 

SAMPLES, SHOULD BE USED AS THE STARTING POINT FOR RECONSTRUCTION OF 

RADIONUCLIDE INTAKE RATES DURING THE RESIDUAL PERIOD. 

NIOSH Response: GA data rather than the BZ data is appropriate to use as a starting point for 

the residual period. The BZ values are generally taken when employees were working with 

radioactive materials and do not represent an average over all periods. The BZ values are 

therefore skewed to high values. Evaluation of the median GA based on the 5 HASL reports 

gives a value of 33 dpm/m3 (GSD = 5.45). Using the 95% value gives a medium of about 

Page 13 of 19 
This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

60 dpm/m3 (GSD = 4.14). The guidance in the site profile will be updated using additional 

information extracted from site reports. The most appropriate GA data is that collected using 

continuous air monitors that include all variations in air concentration over the year. Such data 

was found for the year 1967 for several of the operations at the Apollo uranium facility. Limited 

data was also found for 1966 for the Hammer Mill, and for a few periods in 1962 and 1963 for 

the CRP-3 beryllium glove box area and process area. The results from these measurements 

are summarized in the following table. 

Data Source Period Work Location 

Median 

dpm/m3 GSD 

Mean 

dpm/m3 

Samples 

taken 

HASL Reports 1959–1961 All U processing 53 6.95 196 332 

Site records 1967 Depleted GB 61 2.02 80 76 

Site records 1967 Hammer Mill 35 2.07 43 223 

Site records 1967 Furnace GB 98 1.34 102 86 

Site records 1967 Change Room 90 2.07 93 284 

Site records 1967 End Shake 51 2.26 63 132 

Site records 1967 Crystallizing Hood 50 1.92 56 375 

Site records 1967 Dissolving Hood 45 1.73 51 311 

Site records 1967 Hammer mill 38 1.97 48 303 

Site records 1967 UO2 Glove Box 34 1.89 41 331 

Site records 1967 Aisleway Sampler 19 53 2.01 64 645 

Site records 1967 Aisleway Sampler 20 43 2.21 51 368 

Site records 1966 Hammer Mill 198 3.40 307 57 

Site records 1966 Hammer Mill 222a NAb 415 ~878 

Site records 1962–1963 

CRP-3 process area 119a NAb 150 181 

CRP-3 Be glove box 

area 

119a NAb 150 95 

Averages 1967 — 54 1.95 63 — 

All data — 82 6.95 max 119 — 

a. Value estimated by ratio of median/mean for other data sets. 

b. NA = value not available from data provided. 

Based on the data in the above table, an air concentration value that is favorable to claimants to 

use for the residual period is represented by the maximum median value of 222 dpm/m3. The 

maximum estimated GSD for all data sets is 6.95 for the HASL GA data. Therefore, the 

residual activity will be based on a lognormal distribution with a median of 222 dpm/m3 and a 

GSD of 5.0. This represents a slight increase in median air concentration from the previous 

concentration of 210 dpm/m3 and GSD of 7.91. 

FINDING 19: SC&A RECOMMENDS THAT NIOSH USE A RESUSPENSION FACTOR OF 

ABOUT 1E-5 PER METER TO DERIVE THE AIRBORNE DUST LOADING FOR THE 

BEGINNING OF THE RESIDUAL PERIOD, OR PERHAPS SIMPLY ASSUME THAT THE 

AVERAGE GENERAL AIR DUST LOADING OBSERVED DURING THE OPERATIONAL 

PERIOD IS APPLICABLE TO THE BEGINNING OF THE RESIDUAL PERIOD. 

NIOSH Response: The NIOSH guidance is to use 10-6 m-1 as a resuspension factor for 

undisturbed areas. As stated in OTIB-0070 (ORAUT 2012c): 

Application of resuspension factors in dose assessment has been studied by a number 

of authors. Generally, early conclusions of a value of 10-6 m-1 under quiescent conditions 

Page 14 of 19 
and a factor of 10 higher (10-5 m-1) under conditions of moderate activity (Stewart 1964) 

have been supported by later analysis (Brodsky 1980). 

Because the Apollo and Parks Township facilities continued to have some operations during the 

residual period, the use of a resuspension factor of 10-5 m-1 is reasonable for the uranium 

facilities at Apollo and Parks Township. The value of 10-6 m-1 is reasonable for the Parks 

Township plutonium facility because the facility was cleaned prior to continued use. 

The resuspension rate for the Apollo facility will be changed to 10-5 m-1 and updates will be 

made to Tables 7-2, 7-3, and 7-5. The ingestion rates will be updated as well, but only for the 

change in initial air concentration. In addition, per Finding 18, the initial air concentration and 

deposition activity will be re-evaluated for uranium at Apollo to reflect GA data rather than BZ 

data. 

FINDING 20: THE SITE PROFILE MAKES NO REFERENCE TO RADIONUCLIDES OTHER 

THAN URANIUM DURING THE RESIDUAL PERIOD AT APOLLO. 

NIOSH Response: The issue has been further reviewed; however, the vast majority of activity 

processed at Apollo was uranium. General air concentration data for thorium at the Apollo CP-1 

facilities during 1964 and 1965 were reviewed. The GA samples were not continuous, but were 

taken for periods from a few minutes to about an hour. The 71 data points indicated a median 

thorium concentration of 112 dpm/m3, a mean of 160 dpm/m3, and a GSD of 2.4. The samples 

were counted immediately and after several hours to allow for decay of natural atmospheric 

radon and progeny. The results used in this analysis were those taken after decay. Based on 

this data, residual activity for thorium can be evaluated using a lognormal distribution of 

112 dpm/m3 and a GSD of 5.0 (as per the uranium analysis indicated in the Finding 19 

response.) The thorium at Apollo was received as thorium oxide which had undergone only one 

separation. Because the residual period is approximately 20–30 years after the initial 

processing of thorium at NUMEC, the thorium is assumed to be in equilibrium with all progeny at 

the time of exposure (ORAUT 2014b). The measured activity (1964 and 1965) is assumed to 

be for freshly separated thorium with a minimum activity of thorium-228 of 42% of the thorium232 activity. This resulted in an initial activity of 79 dpm/m3 thorium-232. The activity of 

thorium-232 is also assigned to progeny radium-228, actinium-228, radium-224, and lead-210 

(radon is assumed to remain in the thorium matrix). 

FINDING 21: THERE IS CONFLICTING GUIDANCE ON HOW AGED PLUTONIUM 

MIXTURES SHOULD BE TREATED DURING THE RESIDUAL PERIOD AT PARKS 

TOWNSHIP. 

NIOSH Response: The site profile does not include a callout or discussion of Table 7-4. This 

will be added along with guidance to apply the appropriate activity fractions to the total alpha 

values in Table 7-3. In Table 7-3, the values will be indicated as total alpha. Text will be added 

to indicate values in Table 7-4 are to be multiplied by the isotopic fractions to get the intakes of 

each radionuclide. The most claimant favorable mix should be used for each analysis (by fuel 

type and age). 

The main purpose of Table 7-4 is to account for the amount of americium-241 in the material. 

The dose can then be evaluated for the americium-241 and plutonium. Although the dose from 

the plutonium isotopes can be evaluated for each isotope, the dose is usually evaluated using 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 15 of 19 
the sum of the alpha emitting isotope values (excluding plutonium-241, a beta emitter) 

represented as plutonium-239 as a claimant-favorable approach. 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 16 of 19 
REFERENCES
 

AEC (U.S. Atomic Energy Commission), 1955, “Intercomparison of Film Badge Interpretations,” 

Isotopics, volume 5, number 2, pp. 8–33. [SRDB Ref ID: 3640] 

AEC (U.S. Atomic Energy Commission), 1974a, RO Investigation Report No: 70-364/74-02, 

NUMEC Plutonium Facility, October 8–10, 1974,. [SRDB Ref ID: 142772] 

AEC (U.S. Atomic Energy Commission), 1974b, IE Inspection Report No: 70-364/75-15, 

NUMEC Plutonium Facility, September 2–4, 1975, [SRDB Ref ID: 143005] 

Author unknown, 1977, Summary of B&W and NRC Review Team Meeting, January 19, 1977,. 

[SRDB Ref ID: 142792] 

B&W, 1995, Section 3 of Parks Facilities Characterization Plan, Babcock & Wilcox, March 13, 

1995. [SRDB Ref ID: 142708] 

BMI (Battelle Memorial Institute), 2011, Site Profiles for Atomic Weapons Employers That 

Worked Uranium Metals, Rev 1, Battelle-TBD-6000, June 17. [SRDB Ref ID: 101251] 

Boyd, C. C. Jr., 2006a, “BWXT Records Request,” letter with attachments (collection of external 

dosimetry records 1960–1976, internal dosimetry records 1960–1981, in-vivo counting records 

1968–1985, Occupational Exposure Histories 1978) to S. L. Hinnefeld (NIOSH Taft 

Laboratories), BWXT Services, Lynchburg, Virginia, October 25. [SRDB Ref ID: 27060] 

Boyd, C. C. Jr., 2006b, “BWXT Records Request,” letter with attachments (collection of internal 

dosimetry records 1972–1975, 1979–1987, report of exposure to airborne radioactivity 1976, invivo reports 1980–1981) to S. L. Hinnefeld (NIOSH Taft Laboratories), BWXT Services, 

Lynchburg, Virginia, October 25. [SRDB Ref ID: 27062] 

Boyd, C. C. Jr., 2006c, “BWXT Records Request,” letter with attachments (collection of external 

dosimetry records 1973, internal dosimetry records 1974, 1977, 1978, occupational exposure 

histories 1982, 1983, 1985, incident exposure reports 1975) to S. L. Hinnefeld (NIOSH Taft 

Laboratories), BWXT Services, Lynchburg, Virginia, October 25. [SRDB Ref ID: 27063] 

Boyd, C. C. Jr., 2006d, “BWXT Records Request,” letter with attachments (collection of in-vivo 

counting results, 1970–1976) to S. L. Hinnefeld (NIOSH Taft Laboratories), BWXT Services, 

Lynchburg, Virginia, October 25. [SRDB Ref ID: 27066] 

Boyd, C. C. Jr., 2006e, “BWXT Records Request,” letter with attachments (collection of 

occupational exposure histories 1970–1976, work restrictions 1975, external exposure records 

1975, 1995–1998) to S. L. Hinnefeld (NIOSH Taft Laboratories), BWXT Services, Lynchburg, 

Virginia, October 25. [SRDB Ref ID: 27067] 

Boyd, C. C. Jr., 2006f, “BWXT Records Request,” letter with attachments (collection of in-vivo 

count results 1968, 1971–1980, external dosimetry reports 1970, 1976) to S. L. Hinnefeld 

(NIOSH Taft Laboratories), BWXT Services, Lynchburg, Virginia, October 25. [SRDB Ref ID: 

27069] 

This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 17 of 19 
Brake, R., 1989, untitled letter to G. Runkle (U.S. Department of Energy), Los Alamos National 

Laboratory, Los Alamos, New Mexico, November 9. [SRDB Ref ID: 11091] 

Caldwell, R. D., 1968, untitled letter to L. D. Low (U.S. Atomic Energy Commission), Nuclear 

Materials and Equipment Corporation, Pennsylvania, September 23. [SRDB Ref ID: 142650, 

pp 90-92] 

Caldwell, R. D., 1969a, untitled letter to L. D. Low (U.S. Atomic Energy Commission), Nuclear 

Materials and Equipment Corporation, Pennsylvania, March 18. [SRDB Ref ID: 142650, pp 76-

78] 

Caldwell, R. D., 1969b, “Excessive Exposures, SNM-414 and SNM-145, October, 1969” letter to 

L. D. Low (U.S. Atomic Energy Commission), Nuclear Materials and Equipment Corporation, 

November 4. [SRDB Ref ID: 19970] 

Caldwell, R. D., and W. C. Judd, 1966, Gamma Spectrum Measurements and the Interpretation 

of Absorbed Dose During Plutonium Fuel Fabrication, Nuclear Materials and Equipment 

Corporation, Apollo, Pennsylvania. [SRDB Ref ID: 22776] 

Corridoni, R. D., 1982, untitled memorandum to the B&W Safety Advisory Board, Babcock & 

Wilcox, March 23. [SRDB Ref ID: 142790] 

Davis, A.J., and D.J. Strom, 2008, “Uncertainty and Variability in Historical Time-Weighted 

Average Exposure Data,” Health Physics, volume 94, number 2, pp. 145–160. 

Gerrish, R. W., 1965, “Missing ZPR Plutonium,” memorandum to J. G. Collett, Nuclear Materials 

and Equipment Corporation, Pennsylvania, January 14. [SRDB Ref ID: 142918] 

NUMEC (Nuclear Materials and Equipment Corporation), 1968–1970, Monthly Status Reports 

May 1968 to July 1970, Pennsylvania. [SRDB Ref ID: 142895] 

ORAUT (Oak Ridge Associated Universities Team), 2004, ORAUT-TKBS-0017-5, Fernald 

Environmental Management Project (FEMP) – Occupational Internal Dose, Rev 00, May 28. 

[SRDB Ref ID: 19485] 

ORAUT (Oak Ridge Associated Universities Team), 2005, ORAUT-OTIB-0017, Interpretation of 

Dosimetry Data for Assignment of Shallow Dose, Rev 01, October 11. [SRDB Ref ID: 19434] 

ORAUT (Oak Ridge Associated Universities Team), 2006, ORAUT-PROC-0060, Occupational 

Onsite Ambient Dose Reconstruction for DOE Sites, Rev 01, June 28. [SRDB Ref ID: 29986] 

ORAUT (Oak Ridge Associated Universities Team), 2011, ORAUT-OTIB-0079, Guidance on 

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ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

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information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

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This working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions 

with the ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or 

ABRWH (or their technical support and review contractors) positions unless specifically marked as such. This document represents preliminary 

positions taken on technical issues prepared by NIOSH or its contractor. NOTICE: This report has been reviewed to identify and redact any 

information that is protected by the Privacy Act 5 USC §552a and has been cleared for distribution. 

Page 19 of 19