Draft

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

DOSE RECONSTRUCTION TEMPLATE REVIEW FOR THE 

PEEK STREET FACILITY, SCHENECTADY, NEW YORK

Contract No. 211-2014-58081 

SCA-TR-2019-PR004, Revision 0 

Prepared by 

Douglas Farver, CHP 

SC&A, Inc.

2200 Wilson Boulevard, Suite 300 

Arlington, VA 22201-3324 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

January 29, 2019 

DISCLAIMER 

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

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

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

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

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

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

issues.

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

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

Worker Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: Dose Reconstruction Template Review for the Peek Street Facility, 

Schenectady, New York

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2019-PR004 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: January 29, 2019

TASK MANAGER Kathy Behling [signature on file]

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

DOCUMENT 

REVIEWER(S):

John Stiver, MS, CHP [signature on file]

Kathy Behling [signature on file]

Record of Revisions

Revision 

Number

Effective 

Date

Description of Revision

0 (Draft) 1/29/2019 Initial issue
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TABLE OF CONTENTS

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

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

1 Summary Background Information ........................................................................................ 7

1.1 SC&A DR Template Approach ...................................................................................... 7

2 Facility Description ................................................................................................................. 8

3 External Dose ........................................................................................................................ 11

3.1 Dosimeter ...................................................................................................................... 11

3.2 Recorded Photon Dose .................................................................................................. 11

3.3 Electron Doses .............................................................................................................. 14

3.4 Unmonitored Neutron Dose .......................................................................................... 14

3.5 Missed Dose .................................................................................................................. 16

3.6 Onsite Ambient Dose .................................................................................................... 17

3.7 Occupational Medical Dose .......................................................................................... 18

4 Internal Dose ......................................................................................................................... 20

4.1 Bioassay Detection Limits ............................................................................................ 20

4.2 Fission Product Dose .................................................................................................... 21

4.3 Plutonium Dose ............................................................................................................. 22

4.4 Uranium Dose ............................................................................................................... 23

4.5 Rare Earth Dose ............................................................................................................ 26

4.6 Tritium Dose ................................................................................................................. 27

5 Summary Conclusions .......................................................................................................... 28

6 References ............................................................................................................................. 29
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LIST OF TABLES

Table 3-1. Neutron-to-Photon Ratios for Reactors and Critical Assemblies ................................ 15

Table 3-2. Maximum Ambient Dose (from ORAUT-PROC-0060, Attachment B) ..................... 17

Table 4-1. PSF Bioassay Coworker Study .................................................................................... 20

Table 4-2. Table 5 from PSF DR Template .................................................................................. 24
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ABBREVIATIONS AND ACRONYMS

ABRWH Advisory Board on Radiation and Worker Health 

AEC U.S. Atomic Energy Commission 

AP anterior-posterior 

Ce cerium

DOE U.S. Department of Energy 

DCF dose conversion factor 

dpm disintegrations per minute

DR dose reconstruction 

EE energy employee 

HRE Homogeneous Reactor Experiment 

GDP Gaseous Diffusion Plant 

GM geometric mean

GSD geometric standard deviation

HRT Homogeneous Reactor Test

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules of Bioassay Analysis

IREP Interactive RadioEpidemiological Program

KAPL Knolls Atomic Power Laboratory 

keV kiloelectron volt

LOD limit of detection

MDA minimum detectable activity

MED Manhattan Engineer District

MeV mega-electron volt 

μCi/L microcuries per liter

MTU metric ton(s) of uranium 

Nb niobium 

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH/OCAS Claims Tracking System

OCAS Office of Compensation Analysis and Support 

ORAUT Oak Ridge Associated Universities Team
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ORNL Oak Ridge National Laboratory 

OTIB ORAUT technical information bulletin

pCi picocurie 

Pm promethium 

ppb parts per billion

ppm parts per million

Pr praseodymium 

PSF Peek Street Facility

PSL physically significant level

Pu plutonium 

Ru ruthenium 

RU recycled uranium 

SPRU Separations Process Research Unit

Sr strontium 

SRDB Site Research Database

SRS Savannah River Site 

TBD technical basis document

Tc technetium 

TRU transuranic

U uranium 

UO3 uranium oxide 

Y yttrium

Zr zirconium
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1 SUMMARY BACKGROUND INFORMATION

On December 3, 2018, the Advisory Board on Radiation and Worker Health tasked SC&A with 

a technical review of the current dose reconstruction (DR) template used by the National Institute 

for Occupational Safety and Health (NIOSH) to reconstruct radiation doses for employees of the 

Peek Street Facility (PSF) located in Schenectady, New York. No technical basis document has 

been developed for DRs of workers at PSF. In lieu of a technical basis document, NIOSH has 

developed a DR methodology guideline document and DR template that contain facility-specific 

data, assumptions, and list of documents that provide the basis for the facility-specific data and 

assumptions. 

This report presents SC&A’s review of the current PSF DR template, “DR Draft PSF 3.0.doc” 

(hereafter the “DR Template”) and the DR methodology guideline report, Dose Reconstruction 

Methodology for the Peek Street Facility (NIOSH 2009). 

1.1 SC&A DR TEMPLATE APPROACH

In the PSF DR Template, NIOSH has color-coded the text to indicate the type of information or 

to provide direction to the dose reconstructor. The color schemes are: 

• Black text is wording that is typically unchanged. 

• Blue text is wording that is optional and may need to be deleted.

• Red text indicates wording where a change is needed. 

• Blue text with yellow highlight indicates a description of an optional section. 

• Red text with yellow highlight indicates directions to the dose reconstructor. 

• Turquoise highlight over black or blue text is a field code that is populated when a claimspecific field is imported from workbook tools. 

• Text enclosed in angle brackets (< >) represents fields in the report that must be filled in 

by the dose reconstructor. 

SC&A reviewed the statements and assumptions pertaining to the facility information, external 

dose determination, and the internal dose determination. In this report, each statement is 

presented, followed by SC&A’s evaluation of the statement and a determination of the adequacy 

of the statement. When an issue was identified, SC&A consulted NIOSH 2009 for additional 

information that might resolve the concern.
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2 FACILITY DESCRIPTION

SC&A reviewed the facility information in the DR Template and was able to verify all of the 

PSF site information contained in the template. The specific statements are discussed below.

The DR Template states the main purpose of the PSF was to provide a temporary location for the 

work of the Knolls Atomic Power Laboratory (KAPL) until facilities being constructed at the 

KAPL Site in Niskayuna, New York, were ready to accept the Peek Street Facility work (KAPL 

1994, KAPL 1979). SC&A verified the purpose of PSF based on the information on page 99 of 

KAPL 1994. 

The DR Template states the PSF operated during the years 1947–1954. The Certification Docket 

for the Remedial Action Performed at the Peek Street Site Property in Schenectady, New York

(KAPL 1994) states the General Electric Company operated the PSF as a temporary home for the 

KAPL during the period 1947–1955. The operational period is further clarified on PDF pages 22 

and 23 of KAPL 1994: 

Transfer of activities to the current KAPL facility located on River Road in 

Niskayuna, New York, occurred during the period from early 1949 until late 

1954.... 

In October 1955, the site was sold by the US Government to the private sector. 

SC&A concurs with the DR Template that the operational period at PSF was from 1947 through 

1954. 

The DR Template (page 5) gives the facility description of the PSF as: 

The Peek Street Facility Site consisted of approximately 4.5 acres of land that was 

located at 425 Peek Street in Schenectady, New York. The major structure on the 

site was a 75,000 square foot L-shaped building. In 1947, the main building was 

renovated to house most elements of the Peek Street Facility, including 

administrative offices, library, laundry, security, drafting, design, engineering, 

maintenance, health physics, chemistry laboratories, metallurgy areas, machine 

shops, pile reactor, vault, metal oxidation area, etc. There were also several other 

smaller buildings that were separate from the main building. Those buildings were 

used for security, storage, garage, carpentry, laboratory, and fire protection. Three 

of the buildings included a Laboratory Building, Waste Storage Building, and a 

Fire House.[KAPL 1994, KAPL 1979]

SC&A found this facility description given on PDF page 22 of KAPL 1994. 

The DR Template (page 5) describes the purpose and types of work performed at the PSF: 

The Peek Street Facility was used for two basic purposes: (1) the design of an 

intermediate breeder reactor concept, which was later converted to the design of 

the S1G/S2G submarine reactor plan for the Navy; and (2) the design of a 

chemical process for the recovery of uranium and plutonium from irradiated 
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nuclear reactor fuel. As related to the work of developing and supporting these 

designs, beryllium and radioactive materials were used in the Chemistry and 

Metallurgy Areas of the main building. Radioactive material was also worked 

with in the Metal Oxidation Facility, which was attached to the easterly side of 

the main building, and a Laboratory Building that was located away from the 

main building on its westerly side. Pending preparation for disposal, radioactive 

liquids were stored in the Laboratory Building; solid wastes were stored in the 

Waste Storage Building and alongside the exterior of the Waste Storage Building 

in large metal-lined wood boxes.[KAPL 1979]

SC&A reviewed History and Radiological Status of the KAPL Peek Street Site and the KAPL 

Sacandaga Site (KAPL 1979) and found information describing the function and use of the PSF 

on PDF page 99. 

The DR Template (page 5) provides a description of the types and quantities of radioactive 

materials used at PSF:

Based on record reviews, the maximum quantities of radioactive materials that 

were present at the Peek Street Facility at any one time were estimated to be 

approximately 500 kg of natural uranium, 65 kg of enriched uranium, 300 g of 

plutonium, 3 g of encapsulated radium, a few grams of polonium, and as much as 

a few curies of fission products.[KAPL 1979]

SC&A found the information confirming the DR Template’s description on PDF page 100 of 

KAPL 1979: 

It is reasonable to estimate, based on record reviews, that the maximum quantities 

of radioactive materials mentioned in A, B, and C above at the Peek Street Site at 

any one time consisted of about 500 kilograms of natural uranium, 65 kilograms 

of enriched uranium, 300 grams of plutonium, a few grams of polonium, as much 

as a few curies of fission products, 3 grams of radium (this was contained in 

capsules and not altered), and 100 kilograms of beryllium. 

Finally, the DR Template (pages 5–6) describes the material removal and facility survey and 

release. 

By the end of October 1949, the entire area previously occupied by the Chemistry 

Division at the Peek Street Facility had been surveyed and released for future 

use.... In February 1950, the radioactive material and beryllium in the vault area 

were moved to the KAPL Site.... In April 1950, the Metallurgy Area was 

unconditionally released.... In June 1950, the Laboratory Building was 

unconditionally released. In July 1950, the Waste Storage Building was removed 

and the Oxidation Building that was attached to the main building was also 

removed.... In November 1953, the vault area was unconditionally 

released.[KAPL 1979]
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On PDF pages 101–102 of KAPL 1979, SC&A found the chronology of decontaminating and 

decommissioning events from the phase out of the Chemistry Area operations in October 1949 

through the final release of the site in October 1955.
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3 EXTERNAL DOSE

SC&A reviewed the external dose information in the DR Template regarding dosimeter type, 

recorded, missed, and unmonitored photon, neutron, and electron doses, ambient doses, and 

occupational medical doses. During review of the external dose information, SC&A identified 

six findings and one observation. The specific assumptions and SC&A’s evaluation are discussed 

below. 

3.1 DOSIMETER

The DR Template states:

Because information provided in the dosimetry records indicated that the Hanford 

type dosimeters were used at the Peek Street Facility, the distribution of [the 

energy employee’s] exposure geometry was assumed to be consistent with the 

specific dosimetry parameters applicable to the Technical Basis Document for the 

Hanford Site – Occupational External Dose.[ORAUT-TKBS-0006-6] This exposure 

assumes 100% anterior-posterior (AP) geometry.

SC&A reviewed “Excerpts from KAPL Radiological History Report” (ORAUT 1997). Table 4-1 

on PDF page 73 of that document shows the history of personnel film dosimeters used at KAPL 

facilities. For the PSF operational period, 1947–1954, Table 4-1 indicates a Hanford-type holder 

with a 39 mil silver filter was used. Section 6.4.2.1 of ORAUT-TKBS-0006-6 describes the twoelement dosimeter used at Hanford from 1944 to 1957 (page 25): 

Two-Element Film Dosimeter, October 1944 to March 1957. Hanford 

implemented a two-element beta/photon dosimeter in 1944 based on the design 

developed by Pardue, Goldstein, and Wollan (1944) at the Metallurgical 

Laboratories. This basic dosimeter design was used at the Clinton Laboratory 

(now ORNL) and later by many other MED/AEC/DOE laboratories. The Hanford 

design consisted of an open window and a 1-mm silver shield.

SC&A agrees that the PSF used Hanford-type dosimeters during the operational period from 

1947 to 1954. 

3.2 RECORDED PHOTON DOSE

The DR Template addresses the photon energy distribution used for unmonitored workers, 

monitored plutonium workers, and monitored workers (other than plutonium). Unmonitored 

works are assigned onsite ambient dose and occupational medical doses assuming 100% 30–250 

kiloelectron volts (keV) photons. Monitored plutonium workers are assigned recorded photon 

doses assuming 100% of the nonpenetrating dose is attributed to <30 keV photons and 100% of 

the penetrating dose is attributed to 30–250 keV photons. The photon doses for monitored 

workers not handling plutonium are assigned assuming 100% 30–250 keV photons. 

SC&A agrees that assigning ambient and occupational medical doses using the energy 

distribution of 30–250 keV photons is consistent with the guidance in ORAUT-PROC-0060, 
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Revision 01 (2006), Occupational Onsite Ambient Dose Reconstruction for DOE Sites, and 

ORAUT-PROC-0061, Revision 04 (2017), Occupational Medical X-Ray Dose Reconstruction. 

SC&A also agrees that the nonpenetrating dose for plutonium workers should be attributed to 

<30 keV photons as described in ORAUT-OTIB-0017, Revision 01 (2005), Interpretation of 

Dosimetry Data for Assignment of Shallow Dose. 

The DR Template (page 9) contains the following statement regarding the penetrating photon 

energy assumption: 

Because there was more than one photon energy distribution associated with the 

photon radiation source terms at the Peek Street Facility and because it cannot be 

determined which source term a worker was exposed to at this small facility, a 

photon energy distribution of 100% 30–250 keV photons was applied to the 

measured and missed penetrating photon doses. 

No basis is provided or referenced. Table 6-7 of ORAUT-TKBS-0006-6 lists the photon energy 

distribution for reactor and plutonium processing facilities as attributed to 25% 30–250 keV and 

75% >250 keV photons. NIOSH 2009 (page 4) contains a similar paragraph:

Because there was more than one photon energy distribution associated with the 

photon radiation source terms at the Peek Street Facility and because it cannot be 

determined which source term a worker was exposed to at this small facility, all 

exposures to photon radiation will be assumed to have the most claimantfavorable photon energy distribution that was possible at the site. Therefore, a 

photon energy distribution of 100% 30–250 keV photons will be used for all Peek 

Street Facility photon exposures. 

While NIOSH 2009 states that a photon energy of 30–250 keV is the most claimant favorable, no 

reference is provided in NIOSH 2009 or the DR Template, and SC&A did not find this guidance 

in OCAS-IG-001, Revision 3 (2007), External Dose Reconstruction Implementation Guideline. 

SC&A believes NIOSH should include either the basis or a reference for the basis of the 

penetrating photon energy distribution. 

Finding 1: The assumption of 100% 30–250 keV for the penetrating photon energy distribution 

is unsupported and inconsistent with assumptions used in the Hanford technical basis document. 

Also related to the penetrating photon dose, the DR Template goes on to state:

In addition, an uncertainty factor of 1.3 has been applied to the measured 

penetrating photon doses to account for the uncertainty in dosimeter 

response.[ORAUT-TKBS-0006-6]

Table 6-25 of ORAUT-TKBS-0006-6 lists the overall bias and uncertainty of the Hanford 

dosimeters. The systematic uncertainty for the two-element film dosimeter, like those used at the 

PSF, is listed as 1.2. The Hanford Calculation Workbook 4.16 also list the dosimeter uncertainty 

as 1.2 through 1956. NIOSH 2009 does not provide information on dosimeter uncertainty. 

However, NIOSH 2009 (page 4) does state the following: 
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There is no site-specific information regarding dosimeter limits of detection, 

uncertainty, and bias; therefore, it will be assumed that the Hanford Site’s 

information regarding those dosimeter parameters is applicable to the Peek Street 

Facility. The dose reconstructors shall obtain the applicable dosimeter parameters 

from the Technical Basis Document for the Hanford Site – Occupational External 

Dosimetry.

[ORAUT-TKBS-0006-6, Rev. 02]

SC&A believes NIOSH should provide a basis for the uncertainty factor of 1.3 or use the 

uncertainty factor of 1.2 that is given in their cited reference.

Finding 2: The assumption of an uncertainty factor of 1.3 is unsupported and inconsistent with 

the cited reference.

The DR Template also contains two notes, directions to the dose reconstructor, regarding the 

photon doses. The first note states (page 8): 

(NOTE: For the <30 keV photons energy group, the organ DCFs in Table 4.1a of 

IG-001 [i.e., for the <20 keV Pu photons] should be used instead of the App B 

organ DCFs. Because Table 4.1a does not provide 20 keV photon organ DCF 

values for the esophagus or the ovaries, the <30 keV photon DCFs should be 

used, in accordance with the Table 4.1a recommendations. It should also be noted 

that the Table 4.1a based values for the <30 keV photon energy group have been 

programmed into the tool that was created for the PSF.) 

Table 4.1a of OCAS-IG-001, Revision 3, contains special dose conversion factors (DCFs) for 

plutonium that were calculated assuming AP geometry and 20 keV mono-energetic photons. 

SC&A agrees that applying the 20 keV photon DCFs is appropriate for PSF plutonium workers. 

However, SC&A was unable to locate any PSF-specific tool containing the DCFs from 

Table 4.1a. 

Observation 1: SC&A did not locate a PSF-specific tool containing the preprogrammed 

plutonium DCFs. 

The second note (page 8) reads:

(NOTE: Also, a correction factor to account for the dosimeter’s over-response to 

low energy photons should not be used for the <30 keV photons energy group in 

PSF cases, because the Pu workers were likely exposed to a combination of 

electron and low energy photon emitting source terms. Due to the small size of 

the PSF, the Pu workers’ exposures were not likely limited to Pu. In addition, the 

bioassay data for some of the Pu workers indicate that some were also exposed to 

sources of electron radiation, such as uranium. Applying such a correction factor 

for a PSF worker could result in a worker’s non-penetrating doses being 

underestimated.) 

According to ORAUT-TKBS-0006-6 and the Hanford Calculation Workbook 4.16, the 

dosimeter correction factor is also applied to the <30 keV photon doses. However, SC&A agrees 

with NIOSH’s basis for not applying the correction factor to the <30 keV photon doses. 
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3.3 ELECTRON DOSES

For non-plutonium workers, the DR Template (page 10) attributes the recorded nonpenetrating 

doses to >15 keV electrons:

An electron energy distribution of 100% >15 keV electrons was applied to the 

measured and missed electron doses. Because both photon and electron doses 

were measured by the same dosimeter, when both a zero electron dose and a zero 

photon dose were reported, the zeros were assigned as single missed photon dose, 

in accordance with the Technical Information Bulletin: Interpretation of 

Dosimetry Data for Assignment of Shallow Dose.[ORAUT-OTIB-0017]

This is a common DR practice and is consistent with the guidance in ORAUT-OTIB-0017. 

3.4 UNMONITORED NEUTRON DOSE

As previously described, the PSF used a two-element beta/photon film dosimeter to monitor 

employees. Unmonitored neutron doses are assigned using an assumed neutron energy 

distribution and applying a neutron-to-photon ratio to the penetrating dosimeter reading. The DR 

Template (page 9) states:

Because the neutron energy distribution was mostly unknown, a neutron energy 

distribution of 100% 0.1 to 2 MeV neutrons was applied to all of the unmonitored 

neutrons doses that were estimated using a neutron-to-photon ratio, in accordance 

with the recommendations in the Technical Information Bulletin: Technical Basis 

for Conversion from NCRP Report 38 Neutron Quality Factors to ICRP 

Publication 60 Radiation Weighting Factors for Respective IREP Input Neutron 

Energy Ranges.[ORAUT-OTIB-0055] In accordance with the recommendation in this 

same document, the unmonitored neutron doses were multiplied by a factor of 2. 

SC&A finds this consistent with guidance found on page 11 of ORAUT-OTIB-0055, 

Revision 00 (2006): 

If sufficiently detailed information on neutron energies is not available, dose 

reconstructors should use the following claimant-favorable recommendations: 

(1) assume the corrected dose equivalent to be twice the measured dose, missed 

dose, or dose based on neutron-to-photon ratios; and (2) assume the neutron 

energies are in the range from 0.1 to 2 MeV. 

The DR Template (page 11) states the neutron-to-photon ratio to be used for the PSF was 

determined from facilities with similar neutron producing activities. 

The technical basis documents for the following reactor and critical assembly 

sites were reviewed to determine an appropriate neutron-to-photon ratio for the 

Peek Street Facility: Hanford, Savannah River Site, Oak Ridge National 

Laboratory, Idaho National Engineering Laboratory, Los Alamos National 

Laboratory, Argonne National Laboratory East, Brookhaven National Laboratory, 

and the Energy Technology Engineering Center.... Based on the information 
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provided in these technical basis documents, a neutron-to-photon ratio of 1.2:1 is 

likely claimant-favorable for facilities with critical assemblies and small-scale 

research reactors. It should also be noted that the technical basis documents for 

the following sites did not provide any neutron-to-photon ratios: Idaho National 

Engineering Laboratory, Argonne National Laboratory East, Brookhaven 

National Laboratory, and the Energy Technology Engineering Center. 

SC&A reviewed the neutron-to-photon ratios in the external technical basis documents for 

Hanford (ORAUT-TKBS-0006-6, Revision 04), Savannah River Site (ORAUT-TKBS-0003, 

Revision 03), Oak Ridge National Laboratory (ORAUT-TKBS-0012-6, Revision 01), and Los 

Alamos National Laboratory (ORAUT-TKBS-0010-6, Revision 03). Table 3-1 lists the neutronto-photon ratios for reactors and critical assemblies.

Table 3-1. Neutron-to-Photon Ratios for Reactors and Critical Assemblies

Site Description GM GSD Reference

Hanford Site B, D, F, H, DR, C, KW, 

KE Reactors 0.8 3.0 ORAUT-TKBS-0006-6, 

Table 6-22

Hanford Site N Reactor 0.06 3.1 ORAUT-TKBS-0006-6, 

Table 6-22

Hanford Site 300 Area Test Reactors 

(305, 309, 318, 326) 0.8 3.0 ORAUT-TKBS-0006-6, Table 6-22

Hanford Site Critical Mass Laboratory 

120, 209E 1.1 2.3 ORAUT-TKBS-0006-6, Table 6-22

Savannah River Site 105C, K, L, P, R 0.18 2.52 ORAUT-TKBS-0003, 

Table E-9

Oak Ridge National 

Laboratory

3005 Low Intensity Test 

Reactor 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

3010 Bulk Shielding 

Reactor 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

3042 Oak Ridge 

Research Reactor 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

7500 Homogeneous 

Reactor Experiment 

(HRE) Facility

1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

7503 Molten Salt Reactor 

Experiment 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

7710 Health Physics 

Research Reactor Facility 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory 7500 HRE/HRT Reactors 1.2 2.2

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12

Oak Ridge National 

Laboratory

9213 Critical 

Experiments Facility 

neutron source

1.2 2.0

ORNL-TKBS-0012-6, Section 

6.3.4.2.3.9, X-10 Calculation 

Workbook 2.12
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Site Description GM GSD Reference

Los Alamos National 

Laboratory Reactor Operations 2.3 1.05

ORAUT-TKBS-0010-6, 

Table 6-22, LANL Calculation 

Workbook 3.04

Los Alamos National 

Laboratory Criticality Experiments 2.3 1.05

ORAUT-TKBS-0010-6, 

Table 6-22, LANL Calculation 

Workbook 3.04

Using the information in Table 3-1, SC&A calculated average neutron-to-photon ratios of 1.05 

and 1.53 for reactor facilities and critical assemblies, respectively. The average of these values is 

1.29. 

NIOSH 2009 also specifies the neutron-to-photon ratio of 1.2, citing the same references as the 

DR Template.

SC&A was unable to verify the neutron-to-photon ratio of 1.2 using the cited references and 

believes NIOSH should provide a basis for their decision. The basis may be as simple as citing 

the ORNL neutron-to-photon information given in Section 6.3.4.2.3.9 of ORAUT-TKBS-0012-

6, Revision 01 (2007). 

Finding 3: SC&A was unable to verify the neutron-to-photon ratio of 1.2 using the cited 

references. 

3.5 MISSED DOSE

The missed photon and neutron doses are determined based on the number of zero photon 

dosimeter cycles. Depending on the type of DR, the maximum or actual number of zero 

dosimeter cycles may be used. For the maximum number of dosimeter cycles, the DR Template 

(page 12) states:

The total number of dosimeter cycles assigned was 338 for photons. This number 

was based on a claimant-favorable assumptions of 52 (1947, 1948, and 1952–

1954)[ORAUT 1997] and 26 (1949–1951)[ORAUT 1997]

 photon dosimeter exchanges for 

each full or partial year of monitored employment, up to the date of diagnosis, to 

ensure that all possible instances of a zero dosimeter reading were accounted for

in this dose reconstruction. Based on the minimum reporting level information 

associated with the Hanford type dosimeters,[ORAUT-TKBS-0006-6] a potential missed 

photon dose of 8.409 rem was calculated. 

Table 4-9 of ORAUT 1997 provides the dosimeter exchange frequencies used at KAPL facilities. 

A biweekly exchange frequency was used from 1949–1951 and a weekly exchange frequency for 

the other years from 1947 to 1954, resulting in a maximum of 338 dosimeter cycles. Applying 

the two-element dosimeter limit of detection (LOD) of 0.040 rem, shown in Table 6-13 of 

ORAUT-TKBS-0006-6, results in a maximum missed photon dose of 6.760 rem. 

NIOSH 2009 does not address the dosimeter exchange frequency or dosimeter LOD. 
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Although not stated in the DR Template, it appears NIOSH used a dosimeter LOD of 0.050 rem 

to calculate the missed photon dose. As shown above, this is not consistent with the dosimeter 

LOD information in the Hanford technical basis document. 

Finding 4: The dosimeter LOD used in the DR Template is not specified in the template, and the 

value of 0.050 rem assumed based on NIOSH’s calculation is not consistent with the Hanford 

dosimeter information. 

3.6 ONSITE AMBIENT DOSE

Onsite ambient doses are assessed as part of the DR for monitored and unmonitored periods of 

employment. The maximum ambient dose would be assigned for an unmonitored period 

covering the entire PSF operating period from 1947 to 1954, 8 years. The DR Template 

(page 13) states:

Therefore, annual on-site ambient doses were assigned for each full or partial year 

of employment, up to the date of diagnosis. The doses assigned were based on a 

review of the on-site ambient radiation levels at other sites with similar 

activities,[ORAUT-PROC-0060] which determined that the maximum annual on-site 

ambient radiation levels did not likely exceed 0.423 rem per year at the Peek 

Street Facility. Therefore, a likely claimant-favorable <OR> reasonable annual 

dose of 0.423 rem was assigned for each full or partial year of employment at the 

Peek Street Facility. This results in a total on-site ambient dose of 3.384 rem 

being assigned to the <organ>. 

SC&A reviewed the “Maximizing Dose Summary,” Attachment B of ORAUT-PROC-0060. 

Table 3-2 shows the 1947–1954 annual maximum ambient doses for the Hanford Site, Oak 

Ridge National Laboratory, and Idaho Engineering Laboratory. 

Table 3-2. Maximum Ambient Dose (from ORAUT-PROC-0060, Attachment B) 

Year Hanford Site Oak Ridge National 

Laboratory

Idaho Engineering 

Laboratory

1947 0.157 0.473 NA

1948 0.150 0.980 NA

1949 0.205 1.555 NA

1950 0.072 0.270 NA

1951 0.072 0.372 NA

1952 0.125 0.946 0.159

1953 0.082 0.710 0.159

1954 0.282 0.473 0.159

NA = not listed in Attachment B. 

Using the data above, SC&A calculated the average maximum annual ambient dose for each site 

as 0.143 rem, 0.772 rem, and 0.159 rem for Hanford, Oak Ridge, and Idaho, respectively. The 

average of these three values is 0.342 rem. The average of the Oak Ridge and Hanford values is 

0.433 rem. 
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NIOSH 2009 (page 6) states:

During the years of 1947–1954, the Hanford Site and the Oak Ridge National 

Laboratory (ORNL) were the only sites performing similar types of activities and 

working with similar radionuclides. Using the doses provided in 

Attachment B,[ORAUT-PROC-0060]

 an annual on-site ambient dose assignment of 

0.423 rem was calculated for the Peek Street Facility, and will be assigned for 

each full year of Peek Street Facility employment. A dose of 0.423 rem represents 

average of the annual maximizing on-site ambient doses for the Hanford Site and 

the ORNL for the years of 1947–1954. 

SC&A was unable to verify the annual maximum ambient dose value of 0.423 rem stated in the 

DR Template.

Finding 5: SC&A was unable to verify the PSF annual maximum ambient dose value using the 

cited reference.

3.7 OCCUPATIONAL MEDICAL DOSE

For the occupational medical doses, the DR Template references ORAUT-OTIB-0006, 

Revision 04 (2011), Dose Reconstruction from Occupational Medical X-Ray Procedures, and 

ORAUT-PROC-0061, Revision 03 (2010), Occupational Medical X-Ray Dose Reconstruction 

for DOE Sites. The DR Template states the basis for determining the occupational medical dose 

for PSF employees as:

Based on the information in Table 6-5 of the Technical Information Bulletin: 

Dose Reconstruction from Occupationally Related Diagnostic X Ray 

Procedures[ORAUT-OTIB-0006, Rev. 04] and an assumed annual chest X-ray procedure for 

each full or partial year of employment up to the date of diagnosis... 

NIOSH assumed an annual chest x-ray frequency for each year of employment at the PSF. 

SC&A found information supporting this assumption on PDF page 80 of ORAUT 1997, which 

indicates that in November 1947, KAPL implemented safety rules that included “Chest X-Ray 

and urine examination are to be taken once a year and as indicated by pertinent findings.” 

Revision 04 of ORAUT-OTIB-0006, Dose Reconstruction from Occupational Medical X-Ray 

Procedures, became effective June 20, 2011, and does not contain a Table 6-5 as stated in the 

DR Template. The text in the DR Template most likely refers to Table 6-5, “Organ doses for 

default entrance kerma values,” in Revision 03 PC-1, effective December 21, 2005. SC&A 

believes the correct reference is ORAUT-OTIB-0006, Revision 05, Attachment B, “Organ Dose 

Equivalent Tables,” effective August 13, 2018. 

The DR Template also states that the medical doses include a factor of 1.3 to account for 

uncertainty or an uncertainty of ± 30%. 

This X-ray dose incorporates a factor of 1.3 to account for uncertainty and likely 

exceeds the true X-ray dose to the <organ>.[ORAUT-PROC-0061, Rev. 03] <OR> The 
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assigned X-ray doses were applied as a normal distribution with an uncertainty of 

± 30%.[ORAUT-PROC-0061, Rev. 03]

Upon review of Revision 03 of ORAUT-PROC-0061, SC&A found that Section 6.2.1, 

“Overestimate Approach,” contains the following guidance: 

Unless otherwise noted in Table A-1, doses from the TBDs should be multiplied 

by 1.3 (1.35 for some TBDs) to account for uncertainty in an overestimate (and 

the doses treated as constants in IREP), or doses may be assigned as a normal 

distribution in IREP Parameter 1 and multiplied by 0.3 (0.35 for some TBDs) in 

Parameter 2.

Section 6.2.2, “Best-Estimate Approach,” states:

Unless otherwise noted, assigns the doses from the TBD as a normal distribution 

in IREP Parameter 1 and assigns the doses from the TBD multiplied by 0.3 in 

Parameter 2. 

However, SC&A believes the correct reference is ORAUT-PROC-0061, Revision 04 (2017), 

which does not include guidance to use factor of 1.3 to account for uncertainty. For 

overestimates, best estimates, and underestimates, the Revision 04 guidance states (page 7): 

Assigns the doses in IREP with a normal distribution in IREP Parameter 1 and an 

uncertainty of the dose multiplied by 0.3 in Parameter 2. 

The PSF occupational medical dose methodology described in NIOSH 2009 is similar and 

references ORAUT-OTIB-0006, Revision 03 PC-1. 

SC&A believes the DR Template information regarding occupational medical doses is not 

consistent with current approved guidance. 

Finding 6: The DR Template occupational medical dose basis contains incorrect information 

and outdated references. 
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4 INTERNAL DOSE

SC&A reviewed the internal dose information in the DR Template regarding detection limit and 

the parameters used to assign internal doses from fission product, plutonium, enriched uranium, 

natural uranium, and tritium. During review of the internal dose information, SC&A identified 

two findings and two observations. The specific assumptions and SC&A’s evaluation are 

discussed below.

4.1 BIOASSAY DETECTION LIMITS

According to ORAUT 1997, KAPL employees submitted urine samples that were analyzed for 

fission products, plutonium, enriched uranium, natural uranium, and tritium, as appropriate.

NIOSH complied the bioassay data and calculated intakes for 29 employees in a PSF coworker 

study.1

1 NIOSH’s Excel files for the coworker study are stored on the Division of Compensation Analysis and Support 

internal server at . 

Table 4-1 shows the analyses and number of bioassay samples contained in the PSF 

coworker study. 

Table 4-1. PSF Bioassay Coworker Study

Bioassay Analysis Sample 

Dates

Number 

Samples

PSL

(dpm/day)

Results 

>PSL

Enriched uranium 1950–1961 29 5 1

Fission products 1953–1955 16 50 0

Plutonium 1949–1956 53 0.33 1

PSL = physically significant level. 

PDF page 94 of ORAUT 1997 describes the physically significant level (PSL) as follows: 

The term physically significant level refers to the minimum practical lower limit 

of reliable analytical and counting detectability of uranium in urine 

The DR Template (page 15) discusses the PSL and its relationship to the minimum detectable 

activity (MDA): 

The internal dosimetry records indicated that the physically significant level 

(PSL) for a given analysis technique was an empirically derived value at the 95% 

confidence level (i.e., at a level providing a 5% false positive rate), which is 

similar to currently used decision levels (Lds). However, given that the minimum 

detectable activity (MDA) and a decision level are related by the equation MDA = 

Ld/Ec, where Ec is the counting efficiency, and given the units of measure reported 

for the PSL, the PSL values are more closely related to the currently used MDA 

values. Because all measurement results for non-naturally occurring radionuclides 

showed an activity less than the PSL for the given radionuclides and bioassay 

methods, only missed internal doses were assessed.

 
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The PSLs stated in the DR Template are 50 disintegrations per minute (dpm)/day, 0.33 dpm/day, 

5 dpm/day, 5 μg/day, and 1 microcurie per liter (μCi/L) for fission products, plutonium, enriched 

uranium, natural uranium, and tritium, respectively. With the exception of the PSL for natural 

uranium, SC&A found these same values in Figure 5-23 on PDF page 120 of ORAUT 1997. The 

table on PDF page 120 of ORAUT 1997 lists the PSL for natural uranium at 3 μg/day.

NIOSH 2009 provides a discussion of PSL and references ORAUT 1997. However, it does not 

provide specific PSLs for fission products, plutonium, enriched uranium, natural uranium, or

tritium.

Observation 2: The natural uranium PSL in the DR Template is not consistent with information 

in ORAUT 1997 and is not referenced.

4.2 FISSION PRODUCT DOSE

For the fission product dose assessment, the DR Template (pages 15–16) states:

A 10-day decay period was used for the activation and fission product intake 

calculations, in accordance with the Technical Information Bulletin: Fission and 

Activation Product Assignment for Internal Dose-Related Gross Beta and Gross 

Gamma Analyses[ORAUT-OTIB-0054, Rev. 00 PC-1] because the Peek Street Facility was 

equipped with a zero pile reactor and a critical assembly. The fraction of the gross 

beta radioactivity in the urine sample that was attributable to strontiumn-90 was 

assumed to be 0.037, based on the information in Table 7-2 of the Technical 

Information Bulletin: Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses.[ORAUT-OTIB-0054, Rev. 00 PC-1]

Using that fraction, a missed strontium-90 intake rate was calculated using the 

IMBA computer model and an excretion rate of 0.93 dpm/day. Because lung 

absorption Type F (i.e., very soluble) was likely the only form of strontium-90 

that was likely present at the Peek Street Facility, the evaluation was only 

performed for Type F material. Based on the information and assumptions above, 

a missed internal strontium-90 intake rate of 2.5 dpm/day was calculated.

Intakes for the other radionuclides that may have been associated with the fission 

product analysis were estimated using the strontium-90 intake and the associated 

radionuclide ratios provided in Table 7-3 of the Technical Information Bulletin: 

Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses.[ORAUT-OTIB-0054, Rev. 00 PC-1] A summary of the 

estimated fission product and activation product radionuclide intake rates is 

provided in Table 1 below.
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Table 1. Radionuclide Intake Rates for Urine Samples Analyzed for Fission 

Products 

Fission Product or Activation Product Ratio Applieda Intake Rate (dpm/day)

Strontium-90 1:1 2.5E+00

Barium-140 26:1 6.5E+01

Cerium-141 30:1 7.5E+01

Cerium-144 16:1 4.0E+01

Cesium-134 0.30:1 7.5E-01

Cesium-137 1.3:1 3.2E+00

Europium-155 0.072:1 1.8E-01

Iron-55 1.7:1 4.2E+00

Iodine-131 532:1 1.3E+03

Lanthanum-140 30:1 7.5E+01

Niobium-95 28:1 7.0E+01

Promethium-147 3.4:1 8.5E+00

Ruthenium-103 23:1 5.7E+01

Ruthenium-106 5.5:1 1.4E+01

Strontium-89 19:1 4.7E+01

Yttrium-91 23:1 5.7E+01

Zirconium-95 31:1 7.7E+01 a. Ratios applied are relative to strontium-90 for 10-day old reactor fuel, per ORAUT-OTIB0054.[Rev. 00 PC-1] 

SC&A was able to find the Tables 7-2 and 7-3 information cited in the DR Template. ORAUTOTIB-0054, Revision 00 PC-1, became effective on November 19, 2007. Since that date, there 

have been several revisions. The current version is ORAUT-OTIB-0054, Revision 04, effective 

August 27, 2015; this version does not contain the fission product information in the tables cited. 

NIOSH 2009 contains the same fission product information and references ORAUT-OTIB-0054, 

Revision 00, effective May 11, 2007. 

Finding 7: The fission product information in the DR Template is not consistent with current 

guidance in ORAUT-OTIB-0054, Revision 04. 

4.3 PLUTONIUM DOSE

The assessment of plutonium dose is described in the DR Template (page 17), in part, as follows: 

It was assumed that the plutonium at the Peek Street Facility was weapons-grade 

plutonium, because the Peek Street Facility was developing methods to separate 

plutonium from irradiated nuclear reactor fuel for nuclear weapons production. 

Based on this assumption and given that the plutonium in the urine was freshly 

separated and analyzed for alpha radioactivity, the composition of the plutonium 

in the detected in urine sample only consisted of plutonium-238, plutonium-239, 

plutonium-240, and plutonium-242. The isotopic composition of the alphaemitting plutonium in the urine is provided in Table 2 below... 
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Because the radionuclide composition of plutonium is dependent on its age since 

separation and because the Peek Street Facility only existed for 7 years, the 

maximum age of the plutonium that the Peek Street Facility workers were 

exposed to was assumed to be 7 years old. Table 3 provides the potential 

radionuclide composition of the plutonium source term that the Peek Street 

Facility workers were exposed to. For this dose reconstruction, a plutonium-239 

intake was calculated based on the bioassay data and Table 2 information, and the 

intakes of the other radionuclides in the plutonium source term were estimated 

from the plutonium-239 intake and the ratios provided in Table 3. 

SC&A reviewed the information in Tables 2 and 3 of the DR Template. The reference cited for 

both tables is ORAUT-TBKS-0006-5, Revision 04, effective October 20, 2010. The current 

revision of that document is ORAUT-TBKS-0006-5, Revision 06, effective November 16, 2015. 

The specific activities of Hanford weapons-grade plutonium mixture shown in Tables 2 and 3 are 

contained in Table 5-5 of Revision 06. 

The plutonium information in NIOSH 2009 references ORAUT-TKBS-0006-5, Revision 01, 

effective November 24, 2004. 

Observation 3: The plutonium composition information is correct. However, the reference cited 

is outdated and needs updating. 

4.4 URANIUM DOSE

The DR Template (page 19) discusses the assumptions and parameters used to assess doses from 

enriched uranium as follows: 

Because the Peek Street Facility was involved with developing processes to 

recycle enriched uranium from irradiated nuclear fuel, [the energy employee] 

might have been exposed to other non-uranium radionuclides that were associated 

with recycled uranium. Therefore, potential intakes of the other radionuclides in 

recycled uranium were estimated from the uranium intakes using the ratios for 

each of the associated radionuclides provided in Table 5 below. 

The DR Template (page 20) also addresses doses from natural uranium as follows:

Because the Peek Street Facility was involved with developing fast breeder 

reactor technologies, it may have been involved with developing processes to 

recycle the natural uranium from an irradiated breeder reactor blanket. Therefore, 

[the energy employee] might have been exposed to other non-uranium 

radionuclides that were associated with recycled uranium. The potential intakes of 

the other radionuclides in recycled uranium were estimated from the uranium 

intakes using the ratios for each of the associated radionuclides provided in 

Table 5 above. 

Table 5 of the DR Template is reproduced below in Table 4-2. No reference is cited for the 

radionuclide fractions presented. 
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Table 4-2. Table 5 from PSF DR Template

Radionuclide Intake Activity Fraction Relative to Total Uranium

Neptunium-237 1.4E-03

Plutonium-239 8.5E-04

Technetium-99 3.0E-01

SC&A reviewed ORAUT-TKBS-0006-5, Revision 06; ORAUT-TKBS-0017-5, Revision 03, 

Feed Materials Production Center – Occupational Internal Dose (2017); and Review of 

Generation and Flow of Recycled Uranium at Hanford (DOE 2000) but was unable to verify the 

recycled uranium activity fractions in Table 5. 

NIOSH 2009 (pages 14–15) provides a discussion of the recycled uranium basis: 

Basis for Associated Radionuclides in Recycled Uranium

The basis for the associated radionuclides and ratios being used in Table 3 has 

been excerpted from the draft of the Technical Basis Document titled Dose 

Reconstruction Considerations for Recycled Uranium Contaminants (OTIB0053). The excerpted information is intended to provide a stand alone basis for 

the ratios in Table 3, since plans to issue this technical basis document may have 

been cancelled.

Recycled uranium (RU) is contaminated with trace levels of actinide and fission 

product isotopes that could not be completely removed during chemical 

separation. Contaminants primarily included: Sr/Y-90, Tc-99, Zr/Nb-95, Ru-103, 

Ru-106, thorium, neptunium, plutonium, and other gamma emitting radionuclides. 

The major contaminants contaminates with a possible dose impact include: Tc-99, 

plutonium, and neptunium, since they existed at concentrations that were 

sufficient to result in internal doses to some organs that were considered to be 

significant.

The default ratios listed in Table 3 are intended to be applied to workers of the 

category of primary chemical separation plants or the primary shipping sites 

represented only by the four sites that extracted uranium from spent fuels and 

constituted the supply of recycled uranium to the system.[1] The default ratios are 

based upon Hanford data, and a more detailed summary of how the default ratios 

were derived are provided in Table 4.[1] Because the Peek Street Facility was 

involved with developing the chemical processes that were used by the chemical 

separation plants, the ratios for the chemical separation plants are representative 

of the composition of the material at the Peek Street Facility.”
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Table 4 

Recycled Uranium Contaminant Levels 

for Primary Production Sites[1]

Contaminant

Range of

Average

Concentrations

in Uranium

(ppb)

Range of

Observed

Concentrations

in Uranium

(ppb)

Default Values

(ppb) (pCi/μg U) (pCi/pCi U)

Np-237 100–400 3–800 800 5.7E-04 1.4E-03

Plutonium 3–5 0.1–5 5 3.5E-04 8.5E-04

Tc-99 10–5E+03 1–6E+03 6E+03 0.1 3.0E-01

[Note: Original table has been slightly reformatted to comply with Section 508 of the 

Rehabilitation Act.]

Reference [1] in the above quote is given in NIOSH 2009 (pages 19–20) as:

[1] Bryce, Rich L. ORAU Team. Senior Health Physicist. May 2008. 

For the category of facilities defined as primary production sites this includes the 

four chemical processing plant sites, which dissolved spent fuels and extracted 

plutonium (in the case of the weapons production plants) and the remaining 

unused uranium. Of necessity the radiation protection programs for these facilities 

were extensive as exposure potential was to high levels of fission products, TRU, 

uranium, and others. For RU considerations the primary exposure source would 

have been the process to convert uranyl nitrate to UO3 for shipment. In the case of 

Hanford and SRS they had fuel fabrication facilities in which RU was received 

and processed. Notes in the Hanford Mass Balance report indicate that the trace 

levels of contaminants did not result in detectable levels of TRU, etc. in the 

bioassay analyses, which indicate in turn that claimants will likely have direct 

bioassay upon which to perform DR. However, in cases in which uranium 

bioassay is available and a default for contaminant ratio addition is necessary.

The recommended default of 5 ppm U for Pu is based upon the upper level of the 

observed range. The limits of 10 ppm U were established from the beginning and 

throughout the history of the RU period and the analyses were performed on the 

uranyl nitrate product from the extraction columns and returned if the levels were 

above or approaching the limits. Records in DOE/RL-2000-43[DOE 2000] indicate 

that rare shipments (6 in number) above 10 ppb U (12, 13, 16, 19, 22, & 30) in 

193 MTU of the total 93,316 MTU (approx. 0.2% by mass) were sent to the 

Gaseous Diffusion Plants (GDPs). During a 3 yr period (5/64 to 6/67) out of 352 

shipment lots 6 exceeded 5 ppb U (6, 6, 6, 7, 7, & 8) for a weight percentage of 

approx. 1.7%. The average was 2.2 ppb U with a range of <1 – 8). Since the 

recorded experience of the shipment contaminant levels for Pu indicated that a 

trivial percentage of the mass exceeded 10 ppb U and <2% exceeded 5 ppbU, 

with a clear indication that those lots were sporadic in nature rather than 

extending over a significant time, and with the average a factor of 2 or so below 

5 ppb U, a default of 5 ppb U appears to be adequately favorable to claimants 

with insignificant risk of exposure above this value. Also, the fuel fabrication 
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processes were with materials that had been processed through the GDP process 

with significant contaminant removal and/or through Fernald also with 

contaminant removal and or dilution mechanisms. 

SC&A could not locate a draft, issued, or archived version of ORAUT-OTIB-0053, and NIOSH 

2009 did not provide a Site Research Database (SRDB) identifier for the Rich citation. SC&A 

performed a search of the SRDB for the Rich document and ORAUT-OTIB-0053 but did not 

find either document. 

SC&A believes the DR Template should contain a discussion of the basis for the determination 

of the recycled uranium contaminants and/or an appropriate reference. 

Finding 8: No basis or reference is cited for the recycled uranium activity fractions in Table 5 of 

the DR Template.

4.5 RARE EARTH DOSE

Although rare earth urinalyses were not performed at PSF, the DR Template provides a 

discussion of potential doses from rare earth radionuclide produced at the Separations Process 

Research Unit (SPRU) located at KAPL. The DR Template specifies using the same PSL as for 

fission products, 50 dpm/day, and also contains two notes relevant to rare earth bioassay 

evaluation: 

(NOTE: A rare earth urinalysis result is an indicator that there are some potential 

problems with the covered employment for this case, since this sample was most 

likely collected as a result of the EE’s work at the KAPL or SPRU Site. (SPRU is 

located on the KAPL Site.) If a rare earth urinalysis is reported for PSF 

employment period, the DR should carefully review the available claim records to 

determine if there is an error in NOCTS regarding the EE’s covered employment.)

[pp. 20–21] 

(NOTE: Because Ce/Pr 144 and Pm-147 are both rare earths and fission products, 

doses attributable to intakes of these radionuclides are accounted for in both the 

assessment of the mixed fission products and rare earth urinalyses. Because the 

rare earth urinalysis implies that these radionuclides may have been separated 

from the other fission products, the OTIB-0054 approach should not be used for 

them. Therefore, the potential intakes of Ce/Pr-144 and Pm-147 should be 

assessed based on the rare earth urinalysis data, for the periods that are covered by 

that data.) [p. 21] 

NIOSH 2009 provides similar information regarding rare earth radionuclides. 

SC&A found information on SPRU located at KAPL in ORAUT 1997 and concurs with the 

information described in the DR Template and NIOSH 2009 regarding rare earth radionuclides. 
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4.6 TRITIUM DOSE

Should an employee have tritium bioassay data, the tritium dose is assessed assuming a PSL of 

1 μCi/L, verified and described in Section 4.1, and the guidance in ORAUT-OTIB-0011, 

Revision 00, Technical Information Bulletin: Tritium Calculated and Missed Dose Estimates 

(2004). SC&A concurs with the method described in the DR Template to assess the tritium doses 

at PSF.
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5 SUMMARY CONCLUSIONS

The DR Template for PSF is used by the dose reconstructors in lieu of a technical basis 

document. The DR Template contains facility-specific data, assumptions, and list of documents 

that provide the basis for the facility-specific data and assumptions. SC&A’s review of the DR 

Template for PSF identified eight findings and three observations. 

• Finding 1: The assumption of 100% 30–250 keV for the penetrating photon energy 

distribution is unsupported and inconsistent with assumptions used in the Hanford 

technical basis document. 

• Finding 2: The assumption of an uncertainty factor of 1.3 is unsupported and 

inconsistent with the cited reference.

• Finding 3: SC&A was unable to verify the neutron-to-photon ratio of 1.2 using the cited 

references.

• Finding 4: The dosimeter LOD used in the DR Template is not specified in the template, 

and the value of 0.050 rem assumed based on NIOSH’s calculation is not consistent with 

the Hanford dosimeter information. 

• Finding 5: SC&A was unable to verify the PSF annual maximum ambient dose value 

using the cited reference.

• Finding 6: The DR Template occupational medical dose basis contains incorrect 

information and outdated references. 

• Finding 7: The fission product information in the DR Template is not consistent with 

current guidance in ORAUT-OTIB-0054, Revision 04. 

• Finding 8: No basis or reference is cited for the recycled uranium activity fractions in 

Table 5 of the DR Template.

• Observation 1: SC&A did not locate a PSF-specific tool containing the preprogrammed 

plutonium DCFs. 

• Observation 2: The natural uranium PSL in the DR Template is not consistent with 

information in ORAUT 1997 and is not referenced. 

• Observation 3: The plutonium composition information is correct. However, the 

reference cited is outdated and needs updating. 
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6 REFERENCES

DOE 2000. Review of Generation and Flow of Recycled Uranium at Hanford, DOE/RL-2000-

43, U.S. Department of Energy, Richland Operations Office, Richland, WA. June 30, 2000. 

[SRDB Ref. ID 16494]

KAPL 1979. History and Radiological Status of the KAPL Peek Street Site and the KAPL 

Sacandaga Site, Knolls Atomic Power Laboratory, Schenectady, New York. August 23, 1979. 

[SRDB Ref. ID 6258, PDF pp. 96–118].

KAPL 1994. Certification Docket for the Remedial Action Performed at the Peek Street Site 

Property in Schenectady, New York, Knolls Atomic Power Laboratory, Schenectady, New York, 

for U.S. Department of Energy Schenectady Naval Reactors Office. December 1994. [SRDB 

Ref. ID 6258, PDF pp. 11–34]

NIOSH 2009. Dose Reconstruction Methodology for the Peek Street Facility, National Institute 

for Occupational Safety and Health, Cincinnati, OH. November 20, 2009.

OCAS-IG-001. 2007. External Dose Reconstruction Implementation Guideline, Revision 3, 

National Institute for Occupational Safety and Health, Cincinnati, OH. November 21, 2007.

ORAUT 1997. Excerpts from the KAPL Radiological History Report. December 1997. [SRDB 

Ref. ID 34947]

ORAUT-OTIB-0006. 2005. Dose Reconstruction from Occupationally Related Diagnostic 

X-Ray Procedures, Revision 03 PC-1, National Institute for Occupational Safety and Health, 

Cincinnati, OH. December 21, 2005.

ORAUT-OTIB-0006. 2011. Dose Reconstruction from Occupational Medical X-Ray 

Procedures, Revision 04, National Institute for Occupational Safety and Health, Cincinnati, OH. 

June 20, 2011.

ORAUT-OTIB-0006. 2018. Dose Reconstruction from Occupational Medical X-Ray 

Procedures, Revision 05, National Institute for Occupational Safety and Health, Cincinnati, OH. 

August 13, 2018.

ORAUT-OTIB-0011. 2004. Technical Information Bulletin: Tritium Calculated and Missed 

Dose Estimates, Revision 00, National Institute for Occupational Safety and Health, Cincinnati, 

OH. June 29, 2004.

ORAUT-OTIB-0017. 2005. Interpretation of Dosimetry Data for Assignment of Shallow Dose, 

Revision 01, National Institute for Occupational Safety and Health, Cincinnati, OH. October 11, 

2005.

ORAUT-OTIB-0054. 2007. Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gross Gamma Analyses, Revision 00, National Institute for 

Occupational Safety and Health, Cincinnati, OH. May 11, 2007.
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ORAUT-OTIB-0054. 2007. Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gross Gamma Analyses, Revision 00 PC-1, National Institute for 

Occupational Safety and Health, Cincinnati, OH. November 19, 2007.

ORAUT-OTIB-0054. 2015. Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gross Gamma Analyses, Revision 04, National Institute for 

Occupational Safety and Health, Cincinnati, OH. August 27, 2015.

ORAUT-OTIB-0055. 2006. Technical Basis for Conversion from NCRP Report 38 Neutron 

Quality Factors to ICRP Publication 60 Radiation Weighting Factors for Respective IREP Input 

Neutron Energy Ranges, Revision 00, National Institute for Occupational Safety and Health, 

Cincinnati, OH. June 5, 2006. [SRDB Ref. ID 29980]

ORAUT-PROC-0060. 2006. Occupational Onsite Ambient Dose Reconstruction for DOE Sites, 

Revision 01, National Institute for Occupational Safety and Health, Cincinnati, OH. June 28, 

2006.

ORAUT-PROC-0061. 2010. Occupational Medical X-Ray Dose Reconstruction for DOE Sites, 

Revision 03, National Institute for Occupational Safety and Health, Cincinnati, OH. March 3, 

2010.

ORAUT-PROC-0061. 2017. Occupational Medical X-Ray Dose Reconstruction, Revision 04, 

National Institute for Occupational Safety and Health, Cincinnati, OH. June 6, 2017.

ORAUT-TKBS-0003. 2005. Savannah River Site, Revision 03, National Institute for 

Occupational Safety and Health, Cincinnati, OH. April 5, 2005.

ORAUT-TKBS-0006-5. 2004. Technical Basis Document for the Hanford Site – Occupational 

Internal Dose, Revision 01, National Institute for Occupational Safety and Health, Cincinnati, 

OH. November 24, 2004.

ORAUT-TKBS-0006-5. 2010. Hanford Site – Occupational Internal Dose, Revision 04, 

National Institute for Occupational Safety and Health, Cincinnati, OH. October 20, 2010.

ORAUT-TKBS-0006-5. 2015. Hanford Site – Occupational Internal Dose, Revision 06, 

National Institute for Occupational Safety and Health, Cincinnati, OH. November 16, 2015.

ORAUT-TKBS-0006-6. 2006. Hanford Site – Occupational External Dose, Revision 02, 

National Institute for Occupational Safety and Health, Cincinnati, OH. November 21, 2006.

ORAUT-TKBS-0006-6. 2010. Hanford Site – Occupational External Dose, Revision 04, 

National Institute for Occupational Safety and Health, Cincinnati, OH. January 7, 2010.

ORAUT-TKBS-0010-6. 2013. Los Alamos National Laboratory – Occupational External Dose,

Revision 03, National Institute for Occupational Safety and Health, Cincinnati, OH. March 21, 

2013.
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ORAUT-TKBS-0012-6. 2007. Oak Ridge National Laboratory – Occupational External Dose, 

Revision 01, National Institute for Occupational Safety and Health, Cincinnati, OH. 

September 10, 2007.

ORAUT-TKBS-0017-5. 2017. Feed Materials Production Center – Occupational Internal Dose,

Revision 03, National Institute for Occupational Safety and Health, Cincinnati, OH. March 30,

2017.

Pardue, Goldstein, and Wollan 1944. L.A. Pardue, N. Goldstein, and E.O. Wollan, Photographic 

Film as a Pocket Radiation Dosimeter, CH-1553-A-2223, University of Chicago, Metallurgical 

Laboratory, Chicago, IL. April 8, 1944. [SRDB Ref. ID 8599]