Draft 

REVIEW OF SPECIAL EXPOSURE COHORT ISSUES 

FOR THE HANFORD SITE FOR THE PERIOD 

JULY 1, 1972, TO DECEMBER 31, 1990

Volume I – Main Report 

Revision 1 

Contract No. 200-2009-28555 

Prepared by 

Arjun Makhijani 

Bob Barton 

Joyce Lipsztein

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

September 2011 

Disclaimer

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

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

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

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

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

information only and that premature interpretations regarding its conclusions are unwarranted. 
Effective Date: 

September 30, 2011 

Revision No. 

 1 – Draft 

Document No. 

Draft – Hanford SEC Issues Review – Vol. I 

Page No. 

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

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

Document No. 

Hanford SEC Issues Review 

Effective Date: 

Draft – August 18, 2011 

S. Cohen & Associates: 

Technical Support for the Advisory Board on 

Radiation and Worker Health Review of NIOSH 

Dose Reconstruction Program Revision No: 

1 – Draft 

Review of Special Exposure Cohort Issues for the 

Hanford Site for the Period July 1, 1972, to 

December 31, 1990 

Page 2 of 71 

Task Manager: 

_____________________ Date: ____________ 

Arjun Makhijani, PhD 

Supersedes: 

Rev. 0 

Project Manager: 

_____________________ Date: ____________ 

John Mauro, PhD, CHP 

Reviewers: 

 John Mauro (Entire report) 

 Joyce Lipsztein (Main report) 

 Arjun Makhijani (Appendix A) 

Record of Revisions 

Revision 

Number 

Effective 

Date Description of Revision 

0 (Draft) 07/28/2011 Initial issue 

1 09/30/2011 Vol. I: Section 20 regarding Building 324 incidents has been expanded and modified, 

along with changes to Findings 27-1 and 27-2. 

Vol. II: “Appendix D: Summary of Interviews Related to Building 324 Incidents” has 

been added. 

 
Effective Date: 

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Page No. 

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

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

TABLE OF CONTENTS 

List of Acronyms and Abbreviations...............................................................................................6 

1.0 Overview..............................................................................................................................8 

1.1 Findings....................................................................................................................9 

2.0 Matrix Issue 3: Thorium-232 Exposures from 1960 Onwards .........................................14 

2.1 Thorium-232 Handling and Processing .................................................................14 

2.2 Thorium Monitoring Data – Quantity....................................................................17 

2.3 Thorium-232 Monitoring Data – Quality ..............................................................18 

2.4 NIOSH’s Approach for Thorium-232....................................................................19 

2.5 Findings on Matrix Issue 3 – Thorium-232 ...........................................................20 

3.0 Matrix Issue 4: Highly Enriched Uranium Exposures from July 1, 1972, to 

December 31, 1990 ............................................................................................................21

3.1 Uranium Monitoring Data......................................................................................22 

3.2 NIOSH Proposed Approach for HEU....................................................................23 

3.3 Findings Regarding HEU.......................................................................................24 

4.0 Matrix Item 6: Uranium Coworker Model for Unmonitored Workers.............................24 

4.1 Uranium Data Quality............................................................................................27 

4.2 NIOSH’s Proposed Approach for Uranium...........................................................28 

5.0 Issue 7: Uranium-233........................................................................................................29 

5.1 NIOSH’s Proposed Approach for U-233...............................................................29 

6.0 Issue 8: Recycled Uranium ...............................................................................................30

6.1 Recycled Uranium Findings ..................................................................................31 

7.0 Matrix Issue 9: Neptunium-237........................................................................................31 

7.1 NIOSH’s Approach to Np-237 ..............................................................................32 

7.2 Neptunium Issue Findings .....................................................................................32 

8.0 Issue 10: Tritium...............................................................................................................33 

8.1 NIOSH’s Approach to Tritium ..............................................................................34 

8.2 Tritium Issue Findings ...........................................................................................34 

9.0 Issue 11: Promethium-147 ................................................................................................35

10.0 Issue 12: Strontium-90, Cesium-137, and Mixed Fission Product Dose Estimation........37 

10.1 Periods of Operation and Exposure Potential ........................................................37 

10.2 In-Vivo and In-Vitro Data for Sr-90, Cs-137, and Mixed Fission Products..........38 

10.3 Quality of Data.......................................................................................................40 

10.4 NIOSH’s Approach to Coworker Models..............................................................40 
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10.5 Findings for Mixed Fission Products, Sr-90, and Cs-137......................................41 

11.0 Matrix Issue 13: Tank Farm Alpha Monitoring................................................................42 

12.0 Matrix Issue 14: Plutonium Data Adequacy.....................................................................42 

12.1 Plutonium Data – Quantity ....................................................................................42 

12.2 Plutonium Data – Quality ......................................................................................44 

12.3 NIOSH Approach to Coworker Model for Plutonium...........................................45 

12.4 Findings for Plutonium Modeling..........................................................................46 

13.0 Issue 16: Curium-244 and Americium-241 ......................................................................46 

13.1 Curium-244............................................................................................................46 

13.2 Curium-244 Data ...................................................................................................47 

13.3 NIOSH Approach to Curium-244..........................................................................47 

13.4 Americium-241......................................................................................................47 

13.5 Americium-241 Data .............................................................................................48 

13.6 NIOSH Approach to Americium-241....................................................................49 

13.7 Findings Relating to Curium-244 and Americium-241.........................................49 

14.0 Matrix Issue 19: Lack of Adequate Monitoring (Petitioner Issue)...................................49 

15.0 Matrix Issue 20: Skin Contamination ...............................................................................50 

16.0 Matrix Issue 21: Missing and Destroyed Records ............................................................50 

16.1 Findings for Matrix Issue 21: Destroyed and Missing Records ...........................51 

17.0 Missing Incident Records ..................................................................................................51 

17.1 Incidents in REX Incident Database compared to Individual Records..................52 

17.2 Findings Relating to Missing Incident Records.....................................................61 

18.0 Issue 25: Other Radionuclides ..........................................................................................61 

18.1 Quality of Data.......................................................................................................61 

18.2 NIOSH’s Approach to Other Radionuclides..........................................................61 

18.3 Findings for Other Radionuclides..........................................................................62 

19.0 Matrix Issue 26: Data Completeness ................................................................................62 

19.1 Findings from Appendix A ....................................................................................63 

20.0 Matrix Issue 27: Building 324 Leaks................................................................................63 

20.1 Findings for Issue 27..............................................................................................66 

21.0 References..........................................................................................................................68 
Effective Date: 

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NOTICE:

LIST OF APPENDICES 

Appendix A: Internal Dose Data Completeness Review 

Appendix B: November 2009 Updated Hanford SEC Issues Matrix 

Appendix C: Petitioner/Worker Interview Summary for the Hanford Site 

Appendix D: Summary of Interviews Related to Building 324 Incidents 

LIST OF TABLES 

Table 1. Buildings with Thorium Processing..............................................................................14 

Table 2. Thorium-232-Related Measurements at Hanford, July 1, 1972, to December 31, 

1990...............................................................................................................................18 

Table 3. Percentage of Workers Monitored for Uranium during the Period 

of July 1, 1972, to December 31, 1990 .........................................................................25 

Table 4. Number of Workers Monitored for Uranium during the July 1, 1972, to 

December 31, 1990, Period by Area .............................................................................26 

Table 5. Number of Positive In-Vitro Uranium Results during the Period of July 1, 1972, 

to December 31, 1983 ...................................................................................................27 

Table 6. Impurities in RU at Hanford .........................................................................................31 

Table 7. Promethium-147 Sampling Frequency .........................................................................36 

Table 8. Number of Workers Monitored for Mixed Fission Products, July 1, 1972, to 

December 31, 1990 .......................................................................................................39 

Table 9. Number of Workers Monitored In Vivo for Cs-137 .....................................................39 

Table 10. Number of Workers with Plutonium In-Vitro or InVivo Monitoring Data by Year.....................................................................................43 

Table 11. Positive Plutonium Bioassay Results by Year..............................................................43 

Table 12. Number of Workers Monitored for Am-241 by Area...................................................48 

Table 13. Summary of Incidents Reviewed ..................................................................................53 
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LIST OF ACRONYMS AND ABBREVIATIONS 

Advisory Board 

or Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

ARHCO Atlantic Richfield Hanford Company 

BNW Battelle Northwest 

CATI Computer Assisted Telephone Interview 

c/m or cpm counts per minute 

CSMO Central Scrap Management Office 

DNFSB Defense Nuclear Facilities Safety Board 

DOD Department of Defense 

DOE Department of Energy 

DPM disintegrations per minute 

FFTF Fast Flux Test Facility 

FP Fission Products 

FRG Federal Republic of Germany 

FY Fiscal year 

GSD geometric standard deviation 

HEPA High-Efficiency Particulate Air 

HEDL Hanford Engineering Development Laboratory 

HEU highly enriched uranium 

HT tritiated hydrogen gas 

HTO tritiated water 

IMBA Integrated Modules for Bioassay Analysis 

kgs kilograms 

L liter 

μCi/lb microcurie per pound 

MDA minimum detectable activity 

MeV million electron volts 

MFP Mixed Fission Product 

mL milliliter 

mrem millirem 
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NOTICE:

mR/hr milliroentgen per hour 

nCi nanocurie 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH OCAS Claims Tracking System 

OBT organically bound tritium 

ORAUT Oak Ridge Associated Universities Team 

pCi/d picocurie per day 

PFP Plutonium Finishing Plant 

PNNL Pacific Northwest National Labs 

ppb parts per billion 

ppm parts per million 

PUREX Plutonium-Uranium Extraction 

R&D research and development 

RCRA Resource Conservation and Recovery Act 

REX Radiological EXposure (Hanford exposure database) 

RM Radiation Monitoring 

RTD Resistance Temperature Detector 

RWP Radiation Work Permit 

SC&A S. Cohen and Associates 

SEC Special Exposure Cohort 

SR Savannah River 

SRDB Site Research Database 

SRS Savannah River Site 

TBD Technical Basis Document 

TLD thermoluminescent dosimeter 

TOFDL Thorium Oxide Fuel Development Laboratory 

UST U.S. Testing
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1.0 OVERVIEW 

This is a review of the issues remaining in the Hanford Special Exposure Cohort (SEC) petition, 

SEC-00057-2. On October 20, 2009, the Advisory Board on Radiation and Worker Health 

(ABRWH) accepted NIOSH’s recommendation to extend the Hanford SEC to all Hanford 

workers who met the eligibility criteria for an SEC. The period between July 1, 1972, and 

December 31, 1990, which was the balance of the period in SEC petition SEC-00057-2, was 

reserved for future examination. NIOSH was to publish a revised Technical Basis Document 

(TBD) for Hanford to describe how it was going to approach dose reconstruction for this 

remaining period. 

SC&A prepared an update to the SEC issues matrix and sent it to the Hanford Work Group in 

November 2009. This matrix update specified the remaining outstanding SEC issues 

(Appendix B) and also noted the issues that were resolved. The following issues were resolved 

so far as the SEC review was concerned, because they involved dose reconstruction prior to June 

30, 1972, the period for which Hanford employees with qualified employment have been added 

to the SEC class, or because they do not involve feasibility of dose reconstruction in the July 1, 

1972, to December 31, 1990, period, but rather the details of how it might be approached with 

available data: 

 Matrix Issue 1. Thorium-232 internal exposure from September 1, 1946, up to December 

31, 1959. 

 Matrix Issue 2. Americium-241 internal exposure, January 1, 1949 to December 31, 

1968. 

 Matrix Issue 5. Uranium intake estimation prior to 1948. 

 Matrix Issue 15. Hot particle ingestion (only relevant to the 1940s and 1950s). 

 Matrix Issue 17. Neutron doses to December 31, 1971 (after which TLDs were 

introduced). 

 Matrix Issue 18. External exposure geometry. This was determined during the Work 

Group process to be a site profile and not an SEC issue. 

 Matrix Issue 24. Po-210. (Not relevant after June 30, 1972) 

NIOSH published a revision of the Hanford site profile in 2010 [six volumes covering 

Introduction, Site Description, Occupational Medical Dose, Occupational Environmental Dose, 

Occupational Internal Dose, and Occupational External Dose (ORAUT 2010a through ORAUT 

2010f]. One reason this revision was published was to demonstrate its approach to dose 

reconstruction for the rest of the SEC period from July 1, 1972, to December 31, 1990. The 

revision also specifies dose reconstruction approaches for the SEC period that enable partial dose 

reconstruction in the period up to June 30, 1972, for which an SEC has been granted, as well as 

dose reconstruction for 1991 and beyond. 

The Hanford Work Group asked SC&A to review the 2010 version of the Hanford Site Profile to 

examine issues only as they relate to the feasibility of dose reconstruction during the remaining 
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period of the SEC petition (July 1, 1972, to December 31, 1990). Issues relating to details of 

dose reconstruction methods once feasibility of dose reconstruction is established were to be left 

for later review as Site Profile issues at the direction of the Work Group. This enables as rapid a 

review of the remaining SEC issues as possible. 

In addition, this report also addresses an additional issue brought to the attention of the Advisory 

Board during the course of its deliberations. This issue relates to whether the high radiation 

levels found under Building 324 after 1990 are relevant for the SEC period from July 1, 1972, to 

December 31, 1990. 

It should be noted that while some issues relating to data quality have been addressed as relevant 

and necessary for this report, the issue of the quality of the data as it related to analyses done by 

U.S. Testing (UST) from January 1, 1987, to December 31, 1989, for workers in the 200 Area 

Plutonium Finishing Plant is not covered in this report. That issue is the subject of a separate 

SEC Petition (SEC-00155) and is being addressed in a parallel, but distinct, process of 

investigation by NIOSH and by SC&A (at the direction of the Board). SC&A was asked to 

begin review of this petition during the May 23–25, 2011, meeting in St. Louis, Missouri. 

For ease of reference, the finding numbers are keyed to the matrix issue numbers. 

1.1 Findings 

Finding 3-1: There were Th-232 production and handling operations in the July 1, 1972, to 

December 31, 1979, period; these operations may have been intermittent. Specific examples of 

such operations in the 1970s have been documented above. There was also some intermittent 

exposure potential during surveys and some construction operations in the 1970s and 1980s, but 

the extent of such exposure potential is unclear. 

Finding 3-2: NIOSH’s interpretation of the Th-232 data as being based on Ac-228 counts is 

based on “an educated guess,” and appears to be at variance with Hanford internal dosimetry 

protocols. NIOSH should investigate whether there is a documentary basis for interpreting the 

available in-vivo thorium data. 

Finding 3-3: There are scattered thorium in-vivo counts available for the July 1, 1972, to 

December 31, 1990, period, but the data are sparse and insufficient for dose reconstruction and 

coworker models. 

Finding 3-4: NIOSH’s approach of using uranium bioassay data for thorium intake estimation 

for the WR Vault and Buildings 3722 and 3732 is technically unsupported and not appropriate 

for dose reconstruction. 

Finding 4-1: Highly enriched uranium (HEU) was present in some parts of Hanford, notably, but 

not only, in the Hanford Engineering Development Laboratory (HEDL), throughout the July 1, 

1972, to December 31, 1990, period. Some handling and packaging was periodically done, and 

there may also have been some processing; hence, some HEU exposure potential existed. 

Finding 4-2: NIOSH’s approach to interpretation of fluorometric bioassay data for enriched 

uranium depends on knowledge of workers’ locations at the general times when the samples 

were taken. A limited SC&A examination indicates that such information is usually available. 
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However, a consultation with NIOSH Hanford dose reconstructors, as well as a more detailed 

verification from individual worker files, is needed before the method can be applied with 

confidence to HEU exposure in the SEC period under consideration (July 1, 1972, to December 

31, 1990). 

Finding 4-3: There are in-vivo data that may indicate HEU exposure for the period in question. 

NIOSH should examine these data to see if they are relevant for the HEDL laboratory. Prior to 

use of the U-235 in-vivo data, the discrepancies in minimum detectable activity (MDA) between 

the technical basis document (TBD) and the REX (Radiological EXposure) database need to be 

explained and resolved. 

Finding 6-1: Overall, NIOSH’s method for uranium coworker modeling for the July 1, 1974, to 

December 31, 1988, period is sound. However, there are some caveats noted in Findings 6-2 to 

6-4. NIOSH needs to better justify extending 1988 coworker data to 1989 and 1990, instead of 

using data from those years for facilities other than the 100 Area. 

Finding 6-2: The number of in-vitro measurements in the July 1, 1972, to December 31, 1973, 

period is small. NIOSH might consider using a 95th percentile or some other upper bound value 

for the coworker dose assignment for this period. 

Finding 6-3: The fit for the early part of the 1975–1983 period needs to be revised to be 

claimant favorable; at present it is not. 

Finding 6-4: The MDAs from 1982 onward to at least 1984 and possibly beyond were higher 

than assumed by NIOSH, since the contractual and target values were not being met. Since 

values below the MDA dominate the results, the coworker model in the TBD is not claimant 

favorable for at least part of the period considered here. SC&A has not been able to locate data 

quality audits for the 1972–1979 period. These need to be located, if they exist. 

Finding 7-1: NIOSH should determine the last date for U-233 production, since thoria pellets 

were manufactured at least until the end of 1979 at Hanford. SC&A has not found any 

documentary evidence for thoria target irradiation or U-233 separation in 1979 or later. 

Finding 7-2: If thoria targets were irradiated and U-233 was separated in the July 1, 1972, to 

December 31, 1990, period, NIOSH should determine a bounding intake estimation procedure 

that is specific to that period or establish that the plutonium intake value used for the pre-July 1, 

1972, period is technically appropriate. 

Finding 8-1: Trace contaminant data exist for the 1970–1972 period. They can be used for 

estimating a claimant-favorable value for plutonium and neptunium trace contamination ratios. 

These data appear to be a more appropriate basis than late 1980s and early 1990s data for 

estimating recycled uranium doses suggested by NIOSH. NIOSH should analyze the 1970–1972 

data to determine a better bounding value to plutonium trace contamination in recycled uranium. 

For neptunium, it will be necessary to find the calibration factor to convert counts per minute 

into nanocuries of neptunium per gram of uranium. Other trace contaminant values should also 

be checked against the available raw data. 

Finding 9-1: Neptunium-237 separation (or at least handling) lasted until December 1972 at 

Hanford, about 6 months after the beginning of the SEC period being reviewed here. Neptunium 
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data are too sparse (and, moreover, appear to be mainly incident-related) to allow for a reliable 

coworker model to be developed for the July 1, 1972, to December 31, 1972, period. 

Finding 9-2: Neptunium recovery at Hanford was discussed in the context of processing highlevel waste from the Barnwell Nuclear Fuels plant. While the Barnwell plant was never operated 

for reprocessing spent fuel, NIOSH should make a more complete effort to determine whether 

neptunium was separated in the 1983–1990 period when the PUREX plant was periodically 

operated. 

Finding 9-3: NIOSH has not provided a technical basis for its choices for Np-237 intake 

assignment in relation to plutonium from 1973 onward. 

Finding 10-1: Tritium was produced at Hanford until June 1973. NIOSH has not specified a 

coworker model for tritium for the period of July 1, 1972, to June 30, 1973, that takes this into 

account. The approach specified by NIOSH of calibrating tritium intake to total uranium burnup 

is not likely to be claimant favorable for this period. 

Finding 10-2: NIOSH has not provided a quantitative justification for using a small set of data 

from 1982 and 1983 to derive a conclusion that the chosen coworker intake value for tritium is 

also valid for July 1, 1972, to the end of 1981. 

Finding 10-3: NIOSH should provide a technical basis for using the same value for N Reactor 

workers from July 1, 1972, to the end of 1981 as for 1950s’ unmonitored reactor workers. 

Finding 10-4: NIOSH has pointed to the presence of organically bound tritium (OBT) in 108-B 

and potentially of metal tritides starting in 1988. However, NIOSH has not specified which 

workers had exposure potential or a basis upon which OBT or tritide dose would be assigned. 

Finding 10-5: NIOSH has pointed to the possible presence of metal tritides starting in 1988. 

However, NIOSH has neither confirmed the presence of metal tritides, nor specified a dose 

reconstruction method for metal tritides. NIOSH needs to identify the periods of tritide exposure 

potential and the group of workers who might have been exposed to metal tritides. 

Finding 11-1: Contrary to an inference in the NIOSH site profile, Pm-147 processing occurred 

in the 1972–1975 period and appears to have continued into the late 1970s. 

Finding 11-2: NIOSH’s assigning the same Pm-147 intakes throughout the 1970s is not claimant 

favorable, given that processing operations occurred during portions of this period. NIOSH has 

ample Pm-147 data for the 1960s that may be more appropriate as a basis for coworker models, 

since that was also a period of production. This is indicated by NIOSH’s coworker analysis in 

the TBD. NIOSH should investigate whether this is appropriate for demonstrating dose 

reconstruction feasibility during the times in the 1970s when Pm-147 processing occurred. 

Specifically, NIOSH should investigate whether working conditions were comparable enough to 

allow the use of earlier data for the period of production in the 1970s. 

Finding 11-3: NIOSH should investigate how it is going to address possible intakes, as a result 

of documented incidents in 1974 and 1975. NIOSH should also investigate other Pm-147 

incidents in the SEC period, especially during 1972–1979. 
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Finding 12-1: Overall, there seem to be ample data for Sr-90 and Cs-137 to create coworker 

models except in certain years. 

Finding 12-2: NIOSH’s approach for assigning the same Sr-90 intake values for the period from 

1971 onward is not claimant favorable. A more claimant-favorable approach needs to be 

developed. The Cs-137 and mixed fission product models should also be reviewed in light of the 

Sr-90 and Cs-137 recovery operations and the processing to condition Cs-137 and Sr-90 for 

encapsulation out to the completion of the encapsulation process in October 1983 for Cs-137 and 

January 1985 for Sr-90. 

Finding 12-3: The mixed fission product data for tank farm workers for July 1, 1972, to the end 

of 1977 are very sparse. NIOSH needs to examine whether its present approach is adequate for 

tank farm workers or whether a special coworker model is needed for them for this period. 

Finding 12-4: The implications of the failure to meet contractual MDA limits for several fission 

products (though not Sr-90) are unclear. NIOSH has not discussed this issue. 

Finding 12-5: NIOSH has specified coworker models for several specific fission products. The 

method specified by NIOSH is applicable to reactor areas, but not to separations or waste 

handling and processing areas. An explicit approach for unmonitored workers in reprocessing 

and waste handling and processing areas needs to be developed for radionuclides that have not 

yet been covered for the 200 Area and for certain buildings in the 300 Area. 

Finding 14-1: NIOSH has not considered Type Super S plutonium for intake estimation. 

Finding 14-2: NIOSH should take the different, non-contractual MDA for 1981 into account in 

its coworker model. 

Finding 14-3: NIOSH should ensure that extending 1988 values to 1989 and 1990 is claimant 

favorable for separations workers. 

Finding 14-4: NIOSH’s method of redacting samples relating to incidents may bias coworker 

intake estimates towards a lower intake level. 

Finding 16-1: Curium-244 exposure potential existed due to experimental operations in 1973 

and due to other operations later in the SEC period, probably related to refurbishment of 

facilities. SC&A has documented exposures due to incidents in 1978. It is unclear whether this 

potential continued into the 1980s. Routine sampling appears to have begun in November 1983. 

Finding 16-2: Curium-244 data do not exist for 1973, and are sparse thereafter until October 

1983. 

Finding 16-3: A coworker model for Cm-244 may be feasible after 1984 (inclusive). 

Finding 16-4: NIOSH has not provided a technical rationale for using plutonium data for Cm244 dose reconstruction. This does not appear to correspond to documented situations with Cm244 exposure potential. 

Finding 16-5: NIOSH should develop a coworker model that is specific to Am-241, apart from 

its association as a trace contaminant in plutonium, especially for the period up to 1976. This 
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may be difficult, given that much of the Am-241 in-vitro monitoring was in the context of 

incidents. 

Finding 21-1: A large number of records have been destroyed as a matter of government recordkeeping practices. The vast majority of these records were administrative records of no 

relevance to dose reconstruction. 

Finding 21-2: A small portion of these boxes appear to have contained data relevant to dose 

reconstruction, but most appear to have been prior to the period under consideration. Some 

boxes that may be relevant have no dates indicating the period of the documents in the box. 

Finally, a small portion of the boxes have labels that indicate they are relevant and a date that 

places some or all of the contents in the period under review here. SC&A has not been able to 

establish whether the destroyed boxes contained data that could fill in some of the gaps described 

above for internal dosimetry. 

Finding 22-1: The individual records, consisting of a combination of the records supplied by 

DOE and Computer Assisted Telephone Interview (CATI) reports, appear to document the vast 

majority of incidents. There does not appear to be a detectable pattern of not recording incidents. 

Finding 22-2: The REX databases do not appear to contain the detail of information or the 

quantity of information available in the claimant records. 

Finding 25-1: The MDA for several of the radionuclides in question in this section were not met 

in 1981. There are no other audit reports for the 1972–1986 period that address these 

radionuclides. (SC&A is examining some quality of data issues for the 1987 to 1989 period 

separately, as noted in the Overview section.) 

Finding 25-2: Some of the unusual radionuclides, such as Co-60, only have a significant 

quantity of in-vivo data from about 1983 or 1984. 

Finding 25-3: NIOSH should discuss the pre-1987 quality issues in the TBD relating to several 

fission products, as well as sodium-24, and specify how it will handle MDA and other problems 

that may have existed in years other than 1981. (SC&A is in agreement with NIOSH that 

sodium-24 is only relevant to the 400 Area for the period under consideration here (ORAUT 

2010e, p. 53) and only during the period of operation of the Fast Flux Test Facility (FFTF). 

Finding 25-4: While NIOSH has specified coworker models for a number of radionuclides and 

areas, it has not specified a coworker model for mixed fission and activation products in the 200 

Area and in the waste-handling and processing buildings in the 300 Area. 

Finding 27-1: There were at least two incidents in which radioactive materials migrated from 

presumably contained hot cells and/or pipes to reach the soil. The material recently discovered 

under B-Cell is unlikely to have created internal exposure potential during the leak; however, the 

stabilization operations that began in the late 1980s may have created internal exposure potential. 

Furthermore, we note that the source of the leaks discovered under A and C cells is not well 

understood. 

Finding 27-2: In general, there appear to be ample internal monitoring data for 300 Area 

workers for Cs-137 and mixed fission products (MFPs), and overall for Sr-90 during the 1980s, 
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which is the period in question for the B-Cell incidents. However, it is important for NIOSH to 

verify whether the specific employees who worked in Building 324 at the time of the leaks, 

including the A-Cell and C-Cell leaks and their management, and during the B-Cell stabilization 

operations have adequate internal monitoring data for dose reconstruction. A detailed review of 

the logbooks and incident records relating to B-Cell operations in the mid- to late-1980s and of 

the operations in A-Cell and C-Cell may also be warranted given the scale of the contamination. 

For quality of data comments prior to 1987, see Section 10. 

Finding 27-3: NIOSH should investigate whether there were ad hoc mixtures of fission products 

for which there are scant data for the period used in B-Cell feasibility studies during the mid- to 

late-1980s. If such operations were done, NIOSH should specify a dose reconstruction approach 

for those operations. 

2.0 MATRIX ISSUE 3: THORIUM-232 EXPOSURES FROM 1960 ONWARDS 

This issue was resolved from 1960 up to June 30, 1972, by the extension of the Hanford SEC. 

The balance of the period from July 1, 1972, to 1990 remains to be addressed. 

2.1 Thorium-232 Handling and Processing 

ORAUT 2010e, the updated Site Profile, provides the following information regarding thorium 

handling and processing at Hanford: 

Table 1. Buildings with Thorium Processing 

(Gerber 1992; Isochem 1967; Walser 1978) 

Building, Area Date 

313, 300 1945–1970 

314, 300 1945–1970 

305, 300 1945–1946 

306, 300 1955–1970 

3706, 300 1954–1963 

3707 A&B, 300 1945–1970 

3732, 300 1965–1970 

3722, 300 1946–1970 

325, 300 1963a

321, 300 1964–1970 

202 A, PUREX, 200 East 1965–1970 

241-WR vault, 200 West 1965–1978 

224-U, 200 West 1966–1967 

a – Hot cell work only. Worker exposure unlikely. 

Source: Reproduced from ORAUT 2010e, p. 48 

The only item that is relevant to this report in the above table is the handling of thorium in the 

241-WR vault in the 200 West Area, since it includes the period from July 1, 1972, to 1978, 

which is part of the SEC period under review in this report. Of the publications referenced in the 

above table, Isochem 1967 is not relevant to the period under consideration; neither is Walser 

1978, which only concerns a 1970 thoria campaign, and does not address any thorium handling 

or processing beyond 1970. The TBD further notes that the main thorium projects ended in 

1970, though there may have been some limited projects: 
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Except for small-scope, limited-time projects that might have occurred 

occasionally (such as shipping of stored thorium from the PUREX campaigns), 

exposure to thorium at Hanford stopped in 1970. [ORAUT 2010e, p. 49] 

According to a December 1976 National Lead of Ohio memorandum, National Lead (also known 

as the “Fernald Plant”) expected to receive 100, 150, and 100 metric tons of thorium nitrate 

stored at the time of the writing of the memorandum in tank cars at Hanford in FY 1977, FY 

1978, and FY 1979, respectively (Audia 1976). The shipments did take place, since a 1979 

Fernald memorandum states that the thorium nitrate from Hanford was processed into thoria gel 

at Fernald between May 1977 and January 1979 (Gessiness 1979). However, it should be noted 

that the dates of the filling of the tank cars at Hanford were not researched by SC&A. It is 

unlikely that there was routine internal exposure potential after the tank cars were filled.1

 But 

there may have been internal exposure potential during the filling of the tank cars and due to any 

incidents, such as spills, that may have occurred. There was also thorium nitrate on site after the 

1976 shipments, as noted during the interviews that SC&A did with Hanford employees: 

In 1979, 222-S Laboratory analyzed some thorium nitrate for the purpose of 

disposal. There were some outside underground tanks of thorium nitrate at the 

east end of U Plant. Thorium in these tanks was left over from the thoria runs 

conducted in the late 1960s. Samples were collected from the tanks and brought 

to the 222-S Laboratory. Aliquots of the samples were taken, and destructive 

analysis was run on the aliquots. The 222-S Laboratory also did Pm-147 analysis 

in the 1960s and worked on several different types of fission products. There was 

a shed on the east side of 222-S that was contaminated by Pm-147. There may 

have been a spill. [See Appendix B of this report, p. 19] 

Furthermore, thorium processing did take place after June 30, 1972. Specifically, the monthly 

activities report for November 1979 describes thorium processing in Building 306 as follows: 

A new RWP was issued to cover decontamination and removal of two ventilation 

hoods from existing duct work and transporting and installing them in the 

Thorium Oxide Fuel Development Laboratory (TOFDL). 

Three batches of thorium oxide have been made into pellets in the TOFDL. 

Thorium oxide with 20% UO2 oxide is presently being processed in TOFDL. 

[Monthly Activities Report – November 1979, in SRDB 67785, pdf p. 71] 

[Emphasis added.] 

The December 1979 activities report describes more on the set of thorium pellets that were 

fabricated in that month: 

JA Jones construction is still in the process of renovating room 120 in 

preparation for installing new equipment. The floor and walls in the room need 

to be painted; the floor is scheduled to be covered with tile after it is painted. The 

fixed uranium contamination on the floor has been painted over with yellow paint. 

1

 SC&A has not explicitly researched the topic, but did not come across any incident related to a thorium 

nitrate spill from tank cars once they were filled during the document review for this report. 
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Maintenance continues to machine depleted uranium rods for DOD in the hot 

shop. The last batch of thorium pellets is presently being fabricated in TOFDL 

(approximately 7 kgs). The pellets will have to be weighed, sintered, ground, 

reweighed, and ultrasonic cleaned before they can be loaded into 36" zirconium 

rods. A radiation level of 0.6 mR/hr has been detected at the side of a rod loaded 

with thorium pellets. [Monthly Activities Report – December 1979, in SRDB 

67785, pdf, p. 68] [Emphasis added.] 

It is unclear from these two monthly reports whether the December batch was the last one in the 

series of four batches described therein, or whether they were the last ever fabricated at Hanford. 

Certainly, the date of processing is beyond that described in the TBD (reproduced in the table 

from the TBD above). In fact, the Thorium Oxide Fuel Development Laboratory was built or 

refurbished in 1978: 

Work on the Thorium Oxide Fuel Development Laboratory (TOFDL) in 306 is 

continuing with the installation of new piping, new duct work, and new 

equipment. [Jech 1978b, pdf p. 335] 

Since a new or refurbished facility was built in 1978, there may have been an intent to do 

thorium oxide-related work for some time. As noted above, thorium pellets were fabricated in 

1979. SC&A has not investigated if the thorium oxide work continued into the 1980s and, if so, 

for how long. 

In addition, there was extensive thorium contamination in parts of the 300 Area in the 1970s. 

For example, volume 2 of the TBD notes that, after the 1960s, there was thorium (and uranium) 

contamination in Building 321 (ORAUT 2010b, pp. 23–24). 

Gerber’s history of the 300 Area provides evidence of contamination in other 300 Area buildings 

as well: 

 In 1976, it was found that “much uranium and thorium contamination lies in and beneath 

the 306 Building floor, slab, pipe trenches, and sewer lines” (Gerber 1992, p. 16). 

 In 1977, thorium contamination was found in the West process trench lime pit (Gerber 

1992, pp. 16–17). 

 The 3732 Building was a process development laboratory where work ended in 1970. 

However, contamination from the thorium process development existed “throughout and 

near the building” when Gerber wrote the 300 Area history in 1992 (Gerber 1992, p. 55) 

 When a new 3719 Building was built in 1977 and 1978, “contamination readings of 

50,000 c/m [counts per minute] of U-235 and thorium were found in the sewer line... 

Some contaminated soil was removed at that time” (Gerber 1992, p. 95). According to 

Gerber, the building served as a first aid building, a photography processing shop, and 

headquarters for fire protection. While it was sometimes referred to as the “old laundry 

building,” Gerber states that no laundry was done there (Gerber 1992, p. 95). No 

explanation for the thorium and uranium contamination is provided. 
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 According to Gerber, “every process drain, pipe, pipe trench, and sewer manhole in the 

300 Area also must be expected to contain some degree of process waste buildup, the 

amounts varying depending on the age and type of facility... For example, surveys in the 

late 1980’s found thorium and uranium contamination in a manhole southeast of the 321 

Building...” (Gerber 1992, pp. 114–115). 

Evidently there was some exposure potential to the personnel conducting the surveys in the 

1970s and 1980s. The extent of the exposure potential cannot be deduced from these 

descriptions, which were recorded as guides to decommissioning operations in the post-1992 

period. However, two points can be noted. First, the exposure potential from such survey 

activities was likely very intermittent. Second, the internal exposure potential may have been 

limited, except when soil or residual matter was disturbed. This latter observation derives from 

the fact that the above list compiled from Gerber is connected with residual, rather than 

production-related, exposure potential. 

2.2 Thorium Monitoring Data – Quantity 

There are some in-vivo thorium monitoring data for the post-June 30, 1972, period. The data are 

reported as Th-232, but according to the TBD, the monitoring was based on Ac-228 counting 

(ORAUT 2010e, p. 49): 

Although the in vivo records show the results as 232Th, the measured activity was 

228Ac... The two radionuclides would not have been in equilibrium; both 

disequilibrium at intake and separate biological processes on the 232Th parent and 

the 228Ra progeny mean that the 232Th activity in the body would not have been the 

same as the measured 228Ac. [ORAUT 2010e, p. 49] 

Thorium-232 in-vivo data are sparse for the period of July 1, 1972, to December 31, 1990. 

There are only two in-vitro measurements in 1981. Nine in-vitro measurements in all are labeled 

“TH.” Until 1979, a few in-vivo counts for Th-232 are listed with a detection limit of 0.5 nCi. 

There are 21 Th-232 counts recorded in the 1980s with count-specific detection limits that vary 

from 0.19 to 0.93 nCi, with a median of 0.43 nCi (ORAUT 2010e, pp. 48–50). Table 2 shows 

the number of available measurements by year. 

The measurements are scattered over a number of job titles. In about half the cases, the job title 

is unknown. The data do indicate thorium intakes, for instance in 1985. SC&A has not 

investigated whether some special decommissioning or processing or other work led to the 

thorium counting in 1983, 1985, and 1986. 

Overall, these Th-232 data would be insufficient to reconstruct doses during the period of 

processing and handling of Th-232, which appears to extend at least to the end of 1979. 

However, it should be noted that such handling or processing may have been intermittent, rather 

than continuous. SC&A has not investigated whether the in-vivo Th-232 samples were 

associated with actual processing or with exposure related to decommissioning activities. 
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Table 2. Thorium-232-Related Measurements at Hanford, July 1, 1972, 

to December 31, 1990 

# of Workers Monitored 

by In Vivo 

# of Workers Monitored 

Year by In Vitro Total # 

Workers Th-228 Th-232 Th (see note) Th-232 

1972 8,494 – 1 – – 

1973 9,051 – 1 – – 

1974 10,740 – – – – 

1975 11,374 – – – – 

1976 12,164 – – – – 

1977 14,751 – 2 – – 

1978 16,063 – – – – 

1979 16,626 – 7 1 – 

1980 16,281 – – 7 – 

1981 16,161 – – 1 1 

1982 15,220 – – – 

1983 16,657 1 4 – – 

1984 16,949 – – – – 

1985 17,826 – 10 – – 

1986 18,200 – 4 – – 

1987 18,318 – – – – 

1988 16,542 – – – – 

1989 16,016 – – – – 

1990 16,920 – – – – 

Note: Measurements labeled “Th” appear to be total thorium (ORAUT 2010e, p. 39), i.e., combined 

Th-232 and Th-228. 

2.3 Thorium-232 Monitoring Data – Quality 

Thorium cannot be measured directly; it is the decay products of Th-232 that are measured. 

Therefore, it is necessary to know which decay products were measured and the age of the 

thorium to assess the extent of the disequilibrium. NIOSH assumed that in-vivo thorium 

measurements were made by measuring Ac-228, and that the thorium age from the time of last 

processing was 0.5 years (ORAUT 2010e, p. 49). NIOSH has provided no technical basis or 

data for this choice. For instance, NIOSH has not cited typical storage times before processing. 

Further, NIOSH provides a note attributed to Don Bihl, the Battelle Principal Health Physicist, 

that the conclusion that Ac-228 was monitored was not based on actual documentation: 

This is based on discussion with T.P. Lynch [the In Vivo Monitoring Project 

Manager at Pacific Northwest Laboratory]; however this is an educated guess by 

Lynch based on present practices and his knowledge of the detection systems at 

the time. Documentation that the Th-232 body burdens were based on Ac-228 has 

not been found. [ORAUT 2010e, p. 74] 

In addition, the 1989 Hanford Internal Dosimetry Program Manual states that Th-232 was 

measured via Tl-208 gamma radiation; the manual gives 2 nanocuries as the whole-body limit of 

detection, and 1.5 nCi as the lung burden detection limit for the “Preview Counter” (Carbaugh et 

al. 1989, p. 6.10, pdf pg. 77). We note that the TBD lists the same 2 nCi detection limit, but 

attributes it to Ac-228. It is unclear why “an educated guess” has been chosen over 

documentation in the Hanford internal dosimetry manual. At the same time, the accuracy of Th-
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232 detection using six germanium planar detectors and counting Pb-212 is listed as 0.36 nCi 

(Carbaugh et al. 1989, Table 6.8). The dates of use of the various counters and methods are not 

clear and have not been further investigated by SC&A. The Hanford Whole Body Counting 

Manual is in agreement with this last statement [Th-232 based on Pb-212 for lung counting using 

6 germanium detectors with a 0.36 nCi detection limit (Table 7.1, Palmer 1990)]. It also lists 

Th-232 being measured through the 2.612 MeV (Tl-208) on the Preview Counter and when the 

shadow shield counter is used. 

In summary, data for Th-232 are not only sparse, they are difficult to interpret because of 

conflicting statements and documentation about how they are to be interpreted. It is possible that 

document research going back into various editions of internal dosimetry procedures may be able 

to clarify the issues sufficient to enable reliable information as to which decay product was 

actually measured in the in-vivo count. Furthermore, we note that reliable in-vivo monitoring for 

Th-232 requires separate lung counting and a bone count, such as a cranial or knee count, that 

would distinguish the lung from the bone count. This enables a determination of the distribution 

of activity between lung and skeleton, which is very important to correctly estimate organ doses. 

Generally, whole-body counting is not enough to accomplish this. While the TBD notes that 

head counts and other specialized counts were done at Hanford (ORAUT 2010e, Section 5.3.5), 

SC&A has found no indication that these methods were applied to Th-232 detection. On the 

contrary, NIOSH notes that Hanford used whole-body counting contrary to the practice at the 

time: 

Note: Use of whole-body counting for thorium monitoring was apparently 

different than at other AEC sites that used chest counting. Be sure to recognize 

that the following discussion and Tables 5-26 to 5-28 apply only to whole-body 

counts. [ORAUT 2010e, p. 49] 

NIOSH then states that, “[c]hest counting for thorium replaced whole body counting in 1979” 

(ORAUT 2010e, p. 59). However, the Th-232 intakes that NIOSH has specified for in-vivo 

counting in the TBD do not correspond to chest counting, but to whole-body counting (ORAUT 

2011e, Table 5-27, p. 49); therefore, the NIOSH method would not apply to in-vivo 

measurements after 1979. 

2.4 NIOSH’s Approach for Thorium-232 

NIOSH proposes to interpret the Th-232 data assuming that Ac-228 was the measured 

radionuclide. We have already detailed the problems with this approach above. Furthermore, 

for unmonitored workers who were in the WR Vault and Buildings 3722 and 3732 in 

decontamination activities, NIOSH proposes to use uranium bioassay data for thorium intake 

estimation: 

Only five whole body counts specific for 232Th were conducted from 1970 through 

1978 and there were a few urinalyses for “total actinides” during this period, but 

it is possible that workers involved with shipping of the thorium from the WR 

Vault (primarily storage and transfer of liquid thorium nitrate solution) or 

involved with decontamination of the 3722 and 3732 Buildings did not receive 

bioassay monitoring. If there is evidence of exposure to thorium from shipping of 

the material in the WR Vault or decontamination of the 3722 or 2732 Buildings 
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without bioassay, unmonitored workers should be assigned intakes of thorium 

(with progeny per Section 5.3.4) using the 95th percentile of the uranium 

coworker intakes (constant distribution). [ORAUT 2010e, pp. 59–60] 

Based on the analysis in the prior sections, these approaches are inadequate for several reasons: 

 Thorium in-vivo data are very sparse, difficult to interpret reliably, and of uncertain 

quality. 

 There was thorium processing at least up to the end of 1979. The NIOSH approaches do 

not address the workers who were involved in it. 

 NIOSH has provided no evidence to show that uranium bioassay data would provide a 

satisfactory basis for thorium dose reconstruction in the areas where it has proposed this 

approach. NIOSH had proposed this approach in the 300-M Area for the Savannah River 

Site for the 1953–1965 period. In that case, NIOSH had provided a rationale for such 

use. Nonetheless, SC&A found this to be inappropriate and not well founded 

scientifically for a variety of reasons (SC&A 2011). In the present case, the approach has 

been suggested for a specific group of workers without any technical basis for the 

specific circumstances and for the period under consideration.2

 For instance, NIOSH has 

not analyzed exposure conditions, location, and timing of thorium and uranium 

processing to show a similarity between the two. While such technical information may 

not be sufficient, it is required as a necessary starting point for justifying the use of 

uranium bioassay for thorium intake estimation. At present, it is difficult to see how 

NIOSH’s instruction that there should be “evidence” of exposure to thorium can be 

implemented with only five whole-body counts for thorium between 1970 and 1978 (four 

of which were between July 1, 1972, and 1978). It is even more difficult to see how a 

uranium bioassay can be shown to be a scientifically supportable bounding estimate, 

given that bone surface committed dose for the same mass intake is roughly two orders of 

magnitude greater than that for uranium. This would magnify small errors for certain 

cancers. It is not apparent how errors can be kept small enough to meet the goal of a 

scientifically reasonable bounding dose. 

2.5 Findings on Matrix Issue 3 – Thorium-232 

Finding 3-1: There were Th-232 production and handling operations in the July 1, 1972, to 

December 31, 1979, period; these operations may have been intermittent. Specific examples of 

such operations in the 1970s have been documented above. There was also some intermittent 

exposure potential during surveys and some construction operations in the 1970s and 1980s, but 

the extent of such exposure potential is unclear. 

2

 The internal dose TBD states that “[t]horium exposure was associated with uranium exposure in 

300 Area facilities from 1950 to 1970” (ORAUT 2010e, p. 131). However, this statement is not relevant 

to the period under consideration, which is July 1, 1972, to December 31, 1990. Moreover, it is also 

insufficient to establish scientific reasonableness of using uranium bioassay data, much less a bounding 

dose estimate.
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Finding 3-2: NIOSH’s interpretation of the Th-232 data as being based on Ac-228 counts is 

based on “an educated guess,” and appears to be at variance with Hanford internal dosimetry 

protocols. NIOSH should investigate whether there is a documentary basis for interpreting the 

available in-vivo thorium data. 

Finding 3-3: There are scattered thorium in-vivo counts available for the July 1, 1972, to 

December 31, 1990, period, but the data are sparse and insufficient for dose reconstruction and 

coworker models. 

Finding 3-4: NIOSH’s approach of using uranium bioassay data for thorium intake estimation 

for the WR Vault and Buildings 3722 and 3732 is technically unsupported and not appropriate 

for dose reconstruction. 

3.0 MATRIX ISSUE 4: HIGHLY ENRICHED URANIUM EXPOSURES FROM 

JULY 1, 1972, TO DECEMBER 31, 1990 

The use of HEU for production purposes (irradiation of target rods for the production of tritium 

and U-233) continued into the mid-1960s (Gerber 1992, p. 9), but appears to have ended prior to 

the SEC period under review in this report. However, as noted in matrix issue 4, some R&D 

uses of HEU continued into the 1980s in the HEDL, which was created in 1972 from a part of 

Building 306 (Gerber 1992, p. 15). 

Specifically, work on plutonium and enriched uranium fuels, including HEU, seems to have 

continued in the HEDL. SC&A has not been able to find detailed descriptions of these activities, 

but has not conducted a complete search. However, that these activities took place can be 

inferred from documents in the late 1970s and early 1980s referring to continued accumulation 

of scrap, including HEU scrap. These documents are collected in a single file in the Site 

Research Database (HEU Documents SRDB 66599). There are references to packaging and 

shipping the scrap for evaluation and possible processing to the Savannah River Site (SRS). For 

instance, a 1981 document describes the problem as follows: 

HEDL’s scrap storage problem is obstructing program objectives. Request SR 

[Savannah River] to provide scrap disposition instructions by 3/19/1982, to 

establish whether any relief is imminent through retrievable storage or earlier 

recovery because of highly enriched uranium associated with some of the scrap. 

[HEU Documents SRDB 66599, pdf p. 10]. 

The problem apparently was excessive radiation exposure due to crowding of stored items. A 

November 1981 document describes the problem as follows: 

The accumulation of scrap in HEDL's vault has been a long-standing problem. 

SR has not been able to receive HEDL scrap since 1978. CSMO [Central Scrap 

Management Office] disposition instructions have been to hold much of the scrap 

designated for SR recovery until 1987. Funds have not become available for 

added HEDL vault capacity. Because of the crowding of stored items, HEDL has 

a serious problem with personnel radiation exposure caused by the necessity for 

double handling of items to gain access to material needed for work. 
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We urge that SR reassess the possibility of receiving some of the scrap in the near 

future. [SRDB 66599, pdf p. 15] 

It is unclear when the backlog of HEU was packaged and shipped in its entirety. However, there 

is some evidence that scrap shipments may have occurred well into the decommissioning period. 

A search for HEU documents conducted by ORAUT has brief, one-phrase descriptions of 

documents in various boxes. These descriptions have several mentions of HEU (HEU Search 

Results 2008, pp. 2 and 3 for example). 

The HEU vulnerability report published by the Department of Energy in 1996 describes the 

Hanford HEU holdings as follows: 

The total Westinghouse Hanford inventory is 79 kilograms of HEU, 

198 kilograms of commingled plutonium and HEU, and 37 grams of U-233. 

Three site facilities contain HEU materials within the scope of this assessment: 

the Fast Flux Test Facility (FFTF), the Plutonium Finishing Plant (PFP), and the 

222-S Laboratory. Highly enriched uranium holdings at FFTF include 

approximately 42.7 kilograms in one stainless steel fuel test assembly in Building 

405, and 24 grams within 12 fission chambers in Building 4713-C. Holdings at 

PFP include approximately 36.3 kilograms of U-235 and 37 grams of U-233. 

These are in the form of stable powder, pellets, or pins encased in triplecontainment packages. This facility also contains 198 kilograms of commingled 

plutonium and HEU that is the object of a scheduled stabilization program in 

response to DNFSB 94-1 and the plutonium vulnerability assessment. Holdings 

of the 222-S Laboratory consist of 7 grams of HEU in three samples. [DOE 1996, 

pp. 43–44]

It is clear from the above quote that HEU holdings continued to exist at Hanford past the end of 

the SEC period under examination here. Some exposure potential is likely in circumstances of 

processing or handling for packaging for the purpose of shipment, but it appears that it was likely 

confined to the HEDL and to a much smaller extent elsewhere, as for instance in the 222-S 

Laboratory mentioned in the above quote from DOE 1996. 

3.1 Uranium Monitoring Data 

There are extensive uranium bioassay data in the SEC period (see Section 4). However, until 

1983, measurements were made by the fluorometric method, from which isotopic composition 

cannot be determined. Even after 1983 to the end of the SEC period under consideration here 

(December 31, 1990), alpha spectrometry was only selectively used: 

Starting in 1983 an alpha spectrometry procedure has been used for some 

workers, but these data were less robust and were not representative of the 

overall Hanford workforce. The results were recorded as micrograms per liter 

from 1947 to July 1, 1982, and as micrograms per sample from July 2, 1982, to 

present. The latter were converted to micrograms per liter using the sample 

volume that was recorded as part of the information in the database... Elemental 

uranium samples were not usually 24-hour samples; a mix of sample collection 

periods was used from 1947 to 1988 with overnight sampling being one of the 
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more frequent methods. Therefore, samples of less than 400 mL were not 

removed from the uranium data. [ORAUT 2010e, pp. 119–120] 

The TBD further states the following: 

When alpha spectrometry was introduced in 1983, two uranium urinalyses 

procedures were offered: the elemental procedure above and the alpha 

spectrometric procedure to provide isotopic results. In general, the elemental 

procedure was used for workers who were exposed to natural or slightly enriched 

forms of uranium, and the isotopic procedure was used for depleted or more than 

slightly enriched forms of uranium. In general, personnel who worked in the 

production facilities were monitored by the elemental analysis, whereas PNL 

workers were monitored by the isotopic analysis because of the wide scope of 

research projects that occurred over the years. [ORAUT 2010e, p. 30] 

3.2 NIOSH Proposed Approach for HEU 

Since the vast majority of workers in the period under review were exposed to uranium from 

depleted to slightly enriched (up to 1.25%3), the bioassay data interpreted as 1.25% would be 

appropriate for estimating uranium dose. For those who had exposure potential to HEU, NIOSH 

has proposed the following approach for estimating dose: 

There was exposure to highly enriched uranium (HEU) at Hanford, starting in 

July 1948, occurring heavily in 1949 through 1956 and sporadically after that. 

Almost all the known work was done in the 300 Area although there is indication 

that some work might have occurred in 231-Z. See the Hanford Site Description 

(ORAUT 2010c) for details. The workers exposed to HEU were most likely also 

exposed to natural or slightly enriched uranium. There does not seem to be a way 

to determine from measurements in workers’ files whether an intake of uranium 

included HEU. If exposure to HEU seems likely, the mass urinalysis results 

should be interpreted as HEU using the 93.5% mixture in IMBA.... If claimspecific evidence is found for HEU from July 1, 1972 through October, 1983, the 

elemental uranium urinalysis and chest counts will be used to reconstruct dose. 

From October 1983 on, isotopic urinalysis results that indicate HEU (235U) will 

be used to reconstruct dose. [ORAUT 2010e, p. 33] 

This method requires knowledge of the location of the worker in the context of bioassay 

sampling; in other words, for the period from July 1, 1972, up to October 1983. SC&A has 

conducted audits of individual dose reconstructions for Hanford workers that include this period. 

These audits indicate that the location of the worker is often, but not always, available in 

individual claimant DOE files after about 1972 in both internal and external dose records. 

However, SC&A has not performed a statistically valid check of worker data in this regard. 

NIOSH should consult with its team of dose reconstructors who have the most experience on 

Hanford for the period in question to determine whether the method it has set forth in the TBD, 

quoted above, is a valid one. This is especially important for work in a very specific area like 

3

 The highest enrichment of uranium used in the N reactor, the only one to operate in the period under 

review, was 1.25% (see ORAUT 2010b, p. 12). 
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HEDL that contained HEU that was not handled in any other area, so far as SC&A can 

determine, in the period under consideration. 

SC&A points out that there are also in-vivo data for uranium. Specifically, there are more than 

10,000 U-235 in-vivo measurements starting in 1974, up to and including 1990. Of these, 67 

were above the MDAs specified in the REX database. The measurements greater than the MDA 

may indicate HEU exposure; they are associated with locations in the 300 Area where the HEDL 

was located. However, the MDAs indicated in the TBD are different than those in the REX 

database. The MDA, which varies (for instance according to chest wall thickness of the 

individual), is stated to be 0.17 to 0.37 nCi for the 1968 to 1983 period in the TBD (ORAUT 

2010e, p. 46). However, the MDA values for the same period in the REX database are 0.07 to 

0.22 nCi. NIOSH specifies a single calculated MDA value for the 1983–1986 period of 0.08 nCi 

(ORAUT 2010e, p. 46; the 1983 overlap between the two periods for which MDA is specified is 

not explained). There is a wide range of MDAs indicated in the REX database. SC&A has not 

investigated the reasons for the discrepancies in in-vivo uranium MDAs between the REX 

database and the TBD. 

3.3 Findings Regarding HEU 

Finding 4-1: Highly enriched uranium (HEU) was present in some parts of Hanford, notably, 

but not only, in the HEDL, throughout the July 1, 1972, to December 31, 1990, period. Some 

handling and packaging was periodically done, and there may also have been some processing; 

hence, some HEU exposure potential existed. 

Finding 4-2: NIOSH’s approach to interpretation of fluorometric bioassay data for enriched 

uranium depends on knowledge of workers’ locations at the general times when the samples 

were taken. A limited SC&A examination indicates that such information is usually available. 

However, a consultation with NIOSH Hanford dose reconstructors and a more detailed 

verification from individual worker files are needed before the method can be applied with 

confidence to HEU exposure in the SEC period under consideration (July 1, 1972, to December 

31, 1990). 

Finding 4-3: There are in-vivo data that may indicate HEU exposure for the period in question. 

NIOSH should examine these data to see if they are relevant for the HEDL laboratory. Prior to 

use of the U-235 in-vivo data, the discrepancies in MDA between the TBD and the REX 

database need to be explained and resolved. 

4.0 MATRIX ITEM 6: URANIUM COWORKER MODEL FOR UNMONITORED 

WORKERS 

Potential for exposure to uranium in various enrichments from depleted to slightly enriched 

existed throughout the period from July 1, 1972, to December 31, 1990. SC&A has compiled the 

characteristics of monitoring data for uranium (and other radionuclides). The details are 

provided in Appendix A. Table 3 shows the percentage of workers monitored for uranium (in 

vitro only) by year and area, and Table 4 shows the absolute number of workers monitored for 

uranium by year. The values in bold indicate that the samples were taken in areas judged to have 

some exposure potential—generally the areas with operating facilities using or processing 

uranium. It should be noted that identification of workers with a particular area was done 
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because they worked there at some point in time during their employment. It does not mean that 

that particular worker has worked there throughout their term of employment at Hanford. This is 

a limitation of the analysis (explained in more detail in Appendix A, Attachment 1) and should 

be kept in mind when reviewing the data on percentages and absolute numbers of workers 

monitored in any particular area. It should also be noted that the count of the workers and the 

percentages is a raw count and not adjusted for records where the internal dose numbers could 

not be interpreted, for instance because the units of measurement were missing. See Appendix A 

for more details as to the way in which the data were compiled. 

Table 3. Percentage of Workers Monitored for Uranium during the Period 

of July 1, 1972, to December 31, 1990 

Uranium In-Vitro Overview 

Percentage of Workers Monitored In Vitro for Uranium 

Year U-233 U-234 U-235 U-238 Elemental U U Nat† Combined In-Vitro 

Samples 

1972 – – – – 0.58% 0.15% 0.72% 

1973 – – – – – 0.69% 0.69% 

1974 – – – – – 1.43% 1.43% 

1975 – – – – 0.18% 1.20% 1.20% 

1976 – – – – 0.62% – 0.62% 

1977 – – – – 0.96% 0.01% 0.96% 

1978 – – – – 1.67% – 1.67% 

1979 – – 0.01% – 1.29% – 1.29% 

1980 – – – – 1.41% – 1.41% 

1981 – – – – 1.48% – 1.48% 

1982 – – 0.01% – 1.33% – 1.33% 

1983 – – – – 1.09% – 1.09% 

1984 – – – – 2.04% – 2.04% 

1985 0.02% – 0.02% 0.02% 2.06% – 2.09% 

1986 0.05% 0.01% 0.05% 0.05% 1.47% – 1.51% 

1987 0.01% – 0.01% 0.01% 1.73% – 1.74% 

1988 0.02% – 0.02% 0.02% 1.37% – 1.38% 

1989 0.02% – 0.02% 0.02% 1.48% – 1.50% 

1990 – 0.44% 0.44% 0.44% 1.35% – 1.77% 

 † This urinalysis type is likely also an elemental uranium analysis, as spectrometric uranium analyses were not 

introduced to the Hanford site until 1983. 

Dashes (-) indicate that no data are available for that sampling type and year. 
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Table 4. Number of Workers Monitored for Uranium during the July 1, 1972, to 

December 31, 1990, Period by Area 

Number of Workers Monitored for Uranium (In Vitro + In Vivo) 

Year 

All 

Workers 100 Area 100N Area 200 Area 

200 Tank 

Farms 300 Area 400 Area 700 Area 

1972 (July 1 

onward) 56 [redacted] [redacted] 19 [redacted] 19 [redacted] [redacted]

1973 81 [redacted] [redacted] [redacted] 0 49 0 [redacted]

1974 658 48 41 110 [redacted] 264 13 [redacted]

1975 150 10 10 17 [redacted] 64 0 [redacted]

1976 747 42 36 121 [redacted] 346 22 9 

1977 878 61 56 158 [redacted] 353 46 [redacted]

1978 1258 81 71 293 17 492 116 15 

1979 1450 105 87 386 20 489 235 [redacted]

1980 1573 86 71 430 24 627 262 21 

1981 1519 117 96 496 24 697 303 17 

1982 1437 91 70 440 32 783 405 12 

1983 1424 91 60 543 47 711 314 15 

1984 1627 142 110 658 52 768 308 14 

1985 1745 191 148 764 66 788 267 21 

1986 1847 226 183 892 86 755 223 13 

1987 1868 172 135 906 82 716 232 15 

1988 1758 321 285 1066 78 658 224 25 

1989 2020 382 342 1294 75 681 206 27 

1990 2655 563 519 1799 103 767 210 34 

All Yrs 8171 1161 1057 3233 188 2226 848 130 

The results presented in bold italics indicate areas and years with exposure potential (see Appendix A).

Tables 3 and 4 show that there are potentially a large number of measurements from which to 

construct a coworker model, except for 1972 and 1973, when data are relatively sparse. Table 3 

also shows that there were almost no isotopic analysis in-vitro results (at least in the REX 

database) until 1985, even though the capability may have been established in 1983. However, 

as noted above, U-235 in-vivo measurements are available for most of the period under review 

starting in 1974. In addition, there are Th-234 in-vivo measurements, which measure U-238. 

These could supplement other uranium data. 

NIOSH proposes to use a coworker model based only on in-vitro results. Table 5 shows the 

number of positive in-vitro uranium bioassay results. Note that the number of positive results for 

1972 and 1973 are sparse. The number of positive results for 1983 is also low. 
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Table 5. Number of Positive In-Vitro Uranium Results during the Period 

of July 1, 1972, to December 31, 1983 

Year U U Nat U-233/U-234 U-235 U-238 

1972 20 0 0 0 0 

1973 0 3 0 0 0 

1974 0 160 0 0 0 

1975 26 95 0 0 0 

1976 50 0 0 0 0 

1977 48 0 0 0 0 

1978 27 0 0 0 0 

1979 36 0 0 0 0 

1980 22 0 0 0 0 

1981 57 0 0 0 0 

1982 148 0 0 5 0 

1983 10 0 0 0 0 

1984 619 0 0 0 0 

1985 731 0 15 (See note) 7 7 

1986 731 0 10 0 10 

1987 740 0 1 0 1 

1988 461 0 1 1 2 

1989 204 0 2 0 0 

1990 275 0 10 0 8 

Note: Six positive samples came from the same person, probably involved in an accident. The first 3 

samples were very high, but the MDA was also very high, giving some doubts about the results. 

The same is true for seven U-235 results. The U-238 results were much smaller, indicating 

exposure to enriched uranium (possibly HEU). 

4.1 Uranium Data Quality 

As noted in Section 3.0, the main measurements for uranium were mass measurements made by 

the fluorometric method. This is not directly usable for uranium exposures that may have 

involved low enriched uranium, but it could be adjusted to take enrichment into account without 

stretching scientific reasonableness. Hanford had low or slightly enriched uranium in its 

reactors. The N Reactor used 0.95% or a combination of 0.95% and 1.25% enriched uranium. 

Fluorometric data can reasonably be used if an adjustment is made for the enrichment. This was 

the only production reactor operating at Hanford during the period under review in this report. 

NIOSH’s choice of 0.97% enrichment for defense operations (ORAUT 2010e, Table 5-11), 

while reasonable for a typical situation, is not claimant favorable. A choice of 1.25% would be 

reasonable as a claimant-favorable value except for commercial fuel. 

A 1982 audit of the Hanford bioassay program found that the performance met the requirements, 

except that the MDA was 0.37 micrograms per sample, instead of the required 0.2 micrograms 

per sample of 500 mL (Fleishman 1982, p. 4 of the pdf file). But NIOSH is using an MDA value 

of 0.05 to 0.25 micrograms per liter, with 0.1 being the “most probable” value (ORAUT 2010e, 

p. 31) for 1982. The NIOSH choice of 0.1 micrograms per liter is not claimant favorable for 

1982. A choice of 0.1 micrograms per liter is justified for 1983–1984, since the audit for 1983–

1984 indicates that 0.1 micrograms per liter would be met as the MDA (Spitz 1984). However, 

NIOSH assumes that the target MDA of 0.03 microcuries per liter was met (ORAUT 2010e, 

p. 31). The MDA in the 1981 audit appears to have been met (Hickman 1982). SC&A has not 
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been able to locate earlier audits. As noted in the Overview section of this report, issues related 

to data quality for the 1987 to 1989 period (inclusive) will be covered separately. 

4.2 NIOSH’s Proposed Approach for Uranium 

NIOSH proposes to estimate intakes for Types F, M, and S solubilities for its coworker model 

based on bioassay data. This provides intakes by year. NIOSH notes that only 65 samples were 

collected in 1973 (ORAUT 2010e, p. 121), but included the statistical parameters derived from 

that data, since the values were “comparable to neighboring years.” However, this result is 

heavily dependent on the below MDA values. As is clear from Table 5, there were only 3 above 

MDA values in 1973. This would not provide an adequate basis for providing sufficient 

confidence that all workers were being treated in a claimant-favorable manner. Furthermore, of 

the workers that had any monitoring at all in 1973, only a small fraction (0.69%) were monitored 

(see Table 3); the same was true in the last half of 1972, when the PUREX plant was still 

operating.4 

NIOSH’s choice analysis for the early part of the 1975–1983 period, as shown in Figure C-23 

(ORAUT 2010e, p. 151), does not appear to be claimant favorable. The fitted line is 

systematically below the 50th percentile of the measured values throughout the first part of this 

period.5

Tables C-11 through C-13 specify intakes for Types F, M, and S solubilities of uranium 

(ORAUT 2010e, pp. 131–132). However, the last date in these tables is 1988 (and also in Table 

C-2, p. 121), though the last date for the period under review is 1990. NIOSH has provided the 

following rationale for this: 

For assignment of coworker intakes after 1988, the intake rates for 1988 for each 

radionuclide may be assigned. Given that the internal dosimetry and radiation 

safety programs generally improved with time, especially after 1988, the 

assignment of the 1988 intake rates for periods after 1988 is considered to be 

favorable to claimants. For radionuclides with no assigned intake rate in 1988, 

no intake post-1988 should be assigned. [ORAUT 2010e, p. 136] 

This appears to be a reasonable approach in general, and is likely to be claimant favorable in 

some cases, for instance for N Reactor workers, since that reactor was not operated after 1987. 

However, it may not be appropriate in other cases, for instance for PUREX plant workers, since 

that plant operated intermittently from November 1983 to its final closure at the end of 1992 

(ORAUT 2010b, p. 15). 

Finding 6-1: Overall, NIOSH’s method for uranium coworker modeling for July 1, 1974, to 

December 31, 1988, is sound; however, there are some caveats, as noted in Findings 6-2 to 6-4. 

NIOSH needs to better justify extending 1988 coworker data to 1989 and 1990, instead of using 

data from those years for facilities other than the 100 Area. 

4

 The TBD states that the PUREX plant was closed in June 1972 and did not reopen until November 1983 

(ORAUT 2010b, p. 15). However, as discussed in Section 7, there is clear evidence of PUREX plant operation in 

the latter half of 1972. 5

 It is difficult to read the number of years in the first part of the period to which this statement applies, due 

to the scale at which the graph is drawn. 
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Finding 6-2: The number of in-vitro measurements in the July 1, 1972, to December 31, 1973, 

period is small. NIOSH might consider using a 95th percentile or some other upper-bound value 

for the coworker dose assignment for this period. 

Finding 6-3: The fit for the early part of the 1975–1983 period needs to be revised to be 

claimant favorable; at present, it is not. 

Finding 6-4: The MDAs from 1982 onward to at least 1984, and possibly beyond, were higher 

than assumed by NIOSH, since the contractual and target values were not being met. Since 

values below the MDA dominate the results, the coworker model in the TBD is not claimant 

favorable for at least part of the period considered here. SC&A has not been able to locate data 

quality audits for the 1972–1979 period. These need to be located, if they exist. 

5.0 ISSUE 7: URANIUM-233 

According to the site description volume of the TBD, as well as the history of the 300 Area, the 

last date for thorium target manufacture for irradiation and later separation of U-233 was 1970 

(ORAUT 2010b, p. 31; Gerber 1992, p. 51). The internal dose volume of the TBD states that U233 was separated in the Hanford PUREX plant until 1971 (ORAUT 2010e, p. 28). However, as 

noted in Section 2.1, thoria pellets were manufactured at Hanford at least until the end of 1979. 

It is not clear from the documents cited for 1979 whether these pellets were irradiated. As a 

result, we do not know whether there was any exposure potential for U-233 after 1971, due to 

production-related activities. 

5.1 NIOSH’s Proposed Approach for U-233 

NIOSH has discussed the dose reconstruction method for U-233 as follows: 

For the period for which there was exposure potential, NIOSH proposes the 

following approach to estimating uranium-233 dose: 

Operators, radiation monitors, and equipment maintenance workers at PUREX 

who were involved in the 233U campaigns might have had intakes without specific 

monitoring for 233U. The same might be true for chemists, chemical technicians, 

and radiation monitors associated with the Plutonium Chemistry Laboratory in 

Z-Plant in 1965. The highest urinalysis result (4.39 pCi/d) from a batch of 1970 

PUREX worker records was used to estimate potential chronic intakes for the 

period from 1965 through 1970. [ORAUT 2010e, p. 59] 

While the analogy to plutonium exposure conditions is appropriate for the period for 

which the SEC has been granted (prior to July 1, 1972), it is unclear whether this would 

be claimant favorable or bounding for the SEC period that is under review here (July 1, 

1972, to December 31, 1990) if U-233 was produced or separated in this period. As 

noted in Section 2, thoria targets were manufactured up to 1979, but it is unclear if they 

were shipped offsite or if some were irradiated at Hanford. NIOSH should establish that 

U-233 separation exposure potential in the separations area was equivalent to plutonium 

separation exposure potential if there was U-233 separation in the period under review 

here. 
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Finding 7-1: NIOSH should determine the last date for U-233 production, since thoria pellets 

were manufactured at least until the end of 1979 at Hanford. SC&A has not found any 

documentary evidence for thoria target irradiation or U-233 separation in 1979 and later years. 

Finding 7-2: If thoria targets were irradiated and U-233 was separated in the July 1, 1972, to 

December 31, 1990, period, NIOSH should determine a bounding intake estimation procedure 

that is specific to that period, or establish that the plutonium intake value used for the pre-July 1, 

1972, period is technically appropriate. 

6.0 ISSUE 8: RECYCLED URANIUM 

NIOSH has specified trace contaminant ratios to be applied to uranium bioassay results to 

estimate dose due to recycled uranium handling and processing at Hanford. NIOSH’s table for 

these impurities ratios is reproduced here as Table 6, along with the note that is referenced in the 

table. It is clear that the measurements on which these ratios are based were all made in the late 

1980s and early 1990s. Plutonium and neptunium would be the most important trace 

radionuclides contributing to internal dose. Hence, SC&A performed a check as to whether 

these ratios were claimant favorable when back extrapolated to the earlier part of the 1980s and 

to the 1970s. SC&A deemed such a check necessary not only to attempt to check the validity of 

the back extrapolation, but also because the PUREX plant was shut down periodically during the 

period of the cited measurements (ORAUT 2010b, p. 15). 

SC&A compiled data for the 1970–1972 period relating to plutonium and neptunium trace 

contamination of uranium. These are the closest years to the start of the SEC period under 

review in the present report. The raw datasheets reviewed by SC&A also contain data prior to 

the 1970–1972 period. 

The data used here consist of over 340 PUREX uranium samples taken in the 1970–1972 period. 

The plutonium data are in terms of parts per billion (ppb) of uranium, while the neptunium data 

are in counts per minute (cpm) per gallon. Trace contaminant data for zirconium/niobium-95, 

ruthenium-103 and ruthenium-106 are also available on these datasheets. 

The empirical lognormal average is 2.7 ppb and the empirical lognormal 95th percentile is about 

41 ppb. The maximum measurement is 1,550 ppb. Hence, this is a distribution with a 

particularly long tail. The contaminant concentration of Pu suggested by NIOSH equals about 

13 ppb of plutonium, which is above the empirical lognormal mean, but more than 100 times less 

than the maximum value. The trace contaminant value for plutonium chosen by NIOSH is 

(arguably) not claimant favorable. The available data should allow NIOSH to determine a more 

claimant-favorable value. There are also some fission product trace contaminant data. 

SC&A has not found the calibration data for converting Np-237 counts into parts per billion of 

uranium or nanocuries per gram of uranium. But the data do exist to estimate a claimantfavorable value if the calibration data can be located. 
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Table 6. Impurities in RU at Hanford 

Constituent Maximum Alloweda Observed Rangeb Recommend Levelc

Plutonium 10 ppb U <1–2 ppb U 0.8 nCi Pu-alpha/g U 

Neptunium Not established 0.04–0.16 ppm U 0.4 nCi Np-237/g U 

Thorium 750 ppm U 8–10 ppb U 5 pCi Th-232/g U 

Tx-99 Not established 3–4 ppb U 0.2 μCi Tc-99/g U 

Ru-103, 106 <20 μCi/lb U <6 μCi/lb U 40 nCi Ru-106/g U 

ZrNb-95 <10 μCi/lb U <4 μCi/lb U 20 nCi ZrNb-95/g Ud

Other Gamma Emitters <2 μCi/lb U 0.09–0.75 μCi/lb U Negligible 

a – From UO3 Plant operating specifications (Sula, Carbaugh, and Bihl 1991). 

b – From analysis of uranium lots 88-1, 88-2, and 88-3 that were processed in 1988, and lots 93-01, 93-02, 93-03, 

93-04, and 93-05, processed in 1991. 

c – The recommended levels are expected to result in a slight overestimation of dose compared to levels actually 

observed as presented in the DOE investigation of RU at Hanford (DOE 2000). The plutonium reference 

level was based on the 10-ppb specification, which was reached or exceeded in a few drums throughout 

Hanford history (DOE 2000). 

d – Interpret as 10 nCi each of the two radionuclides [39]. 

Reproduced from ORAUT 2010e, Table 5-12. 

[39] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2006. Because both the 

DOE Hanford RU document (DOE 2000) and the Hanford internal dosimetry manual (PNNL 2003a) express the 

entity as 95Zr/95Nb, it is hard to know for sure if the activity is for each radionuclide or a total of both; however, 

DOE (2000) also refers to 106Ru/106Rh. In the latter case, it is clear that the activity refers to the sum of the parent 

and progeny, so it was deemed reasonable to assume that the 20 nCi of 95Zr/95Nb also refers to the sum of the parent 

and progeny. Because the radionuclides are usually found close to equilibrium, the recommendation was made to 

use 10 nCi 95Zr and 10 nCi 95Nb. 

6.1 Recycled Uranium Findings 

Finding 8-1: Trace contaminant data exist for the 1970–1972 period. They can be used for 

estimating a claimant-favorable value for plutonium and neptunium trace contamination ratios. 

These data appear to be a more appropriate basis than late 1980s and early 1990s data for 

estimating recycled uranium doses suggested by NIOSH. NIOSH should analyze the 1970–1972 

data to determine a better bounding value to plutonium trace contamination in recycled uranium. 

For neptunium, it will be necessary to find the calibration factor to convert counts per minute 

into nanocuries of neptunium per gram of uranium. Other trace contaminant values should also 

be checked against the available raw data. 

7.0 MATRIX ISSUE 9: NEPTUNIUM-237 

Neptunium-237 was separated at Hanford and targets fabricated for irradiation in Hanford 

reactors to produce Pu-238. Subsequently, the neptunium and Pu-238 were separated. The site 

description volume of the TBD states that the last date for neptunium nitrate in the 200 Area 

where separation was done was 1972, since the PUREX plant was shut in that year. The 

monthly report for December 1972 states that the last shipment of neptunium prior to shutdown 

occurred in December 1972: 

Approximately 450 grams of neptunium were dispatched from the Purex Plant on 

December 12, for offsite shipment. This was the last shipment of Purex produced 

neptunium until the plant is restarted. [Monthly report for December 1972, 

p. A-1] 
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There are only four neptunium bioassay samples in the REX database, and all are for the same 

date (September 6, 1972), possibly related to an incident. In any case, the data are not sufficient 

to create a coworker model for neptunium for the last half of 1972, when neptunium appears to 

have been separated in the Hanford PUREX plant (presuming the shipment referred to freshly 

separated material). 

There were 10 other Np-237 bioassay samples and 5 in-vivo counts in total taken in the 1980s, 

scattered throughout that decade. The majority of these also appear to be incident related. 

Hence, there appear to be essentially no data that would capture routine exposure during the 

period of production in the last half of 1972. SC&A has not done an investigation to locate the 

possible Np-related incidents. 

Neptunium recovery from high-level waste was also considered in the context of Cm-244 

recovery as part of the effort involving the development of a production flowsheet for the 

Barnwell Nuclear Fuels Plant (Curium Recovery Flowsheet 1974, Section V). SC&A has not 

found any evidence that the 1974 flowsheet was implemented at Hanford.6

7.1 NIOSH’s Approach to Np-237 

NIOSH has specified the following approach for dose reconstruction in the period under review: 

No evidence of significant exposure to neptunium after 1972 has been found, but 

small research projects with the material can not be ruled out. If an intake 

estimate is required from July 1, 1972 on, unmonitored workers should be 

assigned intakes at the 95th percentile of the plutonium coworker intakes (constant 

distribution) prior to 1984 and at the 50th percentile (with GSDs) from 1984 on. 

[ORAUT 2010e, p. 60] 

Given that neptunium was separated, or at least shipped (and hence handled), in the last half of 

1972, the method specified by NIOSH is not appropriate for those 6 months. While one would 

expect plutonium and neptunium to be associated with each other in reprocessing operations 

after Np-237 separation stopped, NIOSH has not explained its choice of the 95th percentile value 

of plutonium for neptunium dose prior to 1984; nor has it explained the choice of the 50th

percentile after that. This may be associated with the restart of the PUREX plant in November 

1983, but NIOSH should make the technical case for its choices. 

7.2 Neptunium Issue Findings 

Finding 9-1: Neptunium-237 separation (or at least handling) lasted until December 1972 at 

Hanford, about 6 months after the beginning of the SEC period being reviewed here. Neptunium 

data are too sparse (and, moreover, appear to be mainly incident-related) to allow for a reliable 

coworker model to be developed for the July 1, 1972, to December 31, 1972, period. 

Finding 9-2: Neptunium recovery at Hanford was discussed in the context of processing highlevel waste from the Barnwell Nuclear Fuels plant. While the Barnwell plant was never operated 

for reprocessing spent fuel, NIOSH should make a more complete effort to determine whether 

6

 The Barnwell Nuclear Fuels Plant was never operated to reprocess spent fuel. 
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neptunium was separated in the 1983–1990 period, during which the PUREX plant was 

periodically operated. 

Finding 9-3: NIOSH has not provided a technical basis for its choices for Np-237 intake 

assignment in relation to plutonium from 1973 onward. 

8.0 ISSUE 10: TRITIUM 

According to the TBD, tritium data were included as part of external dose records until about 

1987: 

Tritium intakes were accounted for as part of external dose until about 1987, 

when they were entered in the dose database as an internal dose... Section 5.6 

contains a more detailed discussion and guidelines for assigning intakes. Tritium 

exposure was assumed to be chronic during the exposure period, unless a very 

large acute intake was known to occur. [ORAUT 2010e, p. 27] 

However, NIOSH also states that an examination of the external dose records show that external 

dose data do not contain tritium data in many years, including the years between 1973 and 1984 

(ORAUT 2010e, p. 56). 

The REX database contains over 1,650 bioassay measurements for tritium from 1978 onwards to 

the end of 1990, with the vast majority being between 1983 and 1990. 

The site description volume of the TBD states that target rods were irradiated for tritium 

production until 1967 (ORAUT 2010b, p. 12), but the DOE record shows that tritium was 

produced until 1973 (Westinghouse 1994). Therefore, specifying a dose reconstruction method 

for workers who were not monitored for tritium from July 1, 1972, to December 31, 1973, 

appears to be especially important. 

Besides tritiated water [HTO], NIOSH points to the presence of OBT at Hanford: 

Tritium was present in an organic form in the 108-B facility in the pump oil, 

which became contaminated over time. According to a retired radiation monitor 

at the facility, the contamination in pump oil was orders of magnitude below 

general HT and HTO levels in the facility... [ORAUT 2010e, p. 27]

Finally, the TBD also points to the possible presence of metal tritides: 

Metal tritides were potentially present as part of the Tritium Target Program 

work starting in 1988, the metal mostly likely being zirconium. The irradiated 

targets were examined and the tritium driven off in hot cells. Samples were taken 

for analysis, but where the analyses were performed has not been discovered yet. 

[ORAUT 2010e, p. 27] 
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8.1 NIOSH’s Approach to Tritium 

NIOSH dose reconstruction approach for the 1984 to 1986 period is based on a coworker 

analysis of the tritium data. These values are used for later years for 100 Area and 200 

Area workers on the hypothesis that the N Reactor was shut down in 1987, and since 

reprocessing was intermittent in the 1987–1992 period (ORAUT 2010e, p. 58). There are 

data available for the later years, including above MDA results, for instance for 1989 and 

1990. The rationale provided for extending the 1984–1986 analysis to later years for 200 

Area workers in particular seems rather weak, given that data are available for the years 

in question. 

For the 1973–1983 period for which no tritium data exist in the external dose records 

(according to NIOSH either because measurements were not made or were below MDA 

and not recorded), NIOSH has calibrated tritium doses to uranium fuel dissolution in a 

rather complex way. 

First it has assumed that reactor and fuel dissolution plant workers had 10% of the intakes 

of the workers in the Tritium Extraction Facility, which shut down in 1955 (ORAUT 

2010e, pp. 56–58). This is simply an assumption based on the judgment of one site 

expert (ORAUT 2010e, p. 77). Secondly, this value is calibrated to the total burn-up of 

uranium fuel dissolved. Thirdly, the resulting value of an intake of 3.2 × 10-5 pCi/day, is 

extrapolated to the 1973–1983 period. This value was much higher than indicated by the 

1982–1983 bioassay results of a “small set” of reactor workers; NIOSH notes that the set 

was “too small for a complete statistical analysis” (ORAUT 2010e, p. 58). Since the 

suggested value is almost 27 times higher than the actual measurements, NIOSH’s 

approach seems reasonable for 1982 and 1983. However, the N Reactor was operating 

throughout the 1973–1981 period; in view of that, a technical basis that relates N Reactor 

exposures to earlier 1950s exposure estimates should be provided. This is especially true 

for the period until June 30, 1973, when tritium production still continued at Hanford. 

Finally, NIOSH has not specified a dose reconstruction method for either OBT or for metal 

tritides (if their presence is confirmed). The main problem here is not so much determining the 

model to be applied for metal tritides, which NIOSH has specified in ORAUT-OTIB-0066 

(ORAUT 2007a), but whether and where the tritides were used and which of the workers at 

Hanford had exposure potential. Some quantitative basis for determining exposure potential is 

needed; so far, it is lacking. 

8.2 Tritium Issue Findings 

Finding 10-1: Tritium was produced at Hanford until June 1973. NIOSH has not specified a 

coworker model for tritium for the period of July 1, 1972, to June 30, 1973, that takes this into 

account. The approach specified by NIOSH of calibrating tritium intake to total uranium burnup 

is not likely to be claimant favorable for this period. 

Finding 10-2: NIOSH has not provided a quantitative justification for using a small set of data 

from 1982 and 1983 to derive a conclusion that the chosen coworker intake value for tritium is 

also valid for July 1, 1972, to the end of 1981. 
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Finding 10-3: NIOSH should provide a technical basis for using the same value for N Reactor 

workers from July 1, 1972, to the end of 1981 as for 1950s’ unmonitored reactor workers. 

Finding 10-4: NIOSH has pointed to the presence of OBT in 108-B and potentially of metal 

tritides starting in 1988. However, NIOSH has not specified which workers had exposure 

potential or a basis upon which OBT or tritide dose would be assigned.

Finding 10-5: NIOSH has pointed to the possible presence of metal tritides starting in 1988. 

However, NIOSH has neither confirmed the presence of metal tritides, nor specified a dose 

reconstruction method for metal tritides. NIOSH needs to identify the periods of tritide exposure 

potential and the group of workers who might have been exposed to metal tritides. 

9.0 ISSUE 11: PROMETHIUM-147 

Hanford separated Pm-147 from fission products for use as a heat source. Major production 

started in 1964 (ORAUT 2010b, p. 40). The site description volume of the TBD does not 

mention Pm-147 work past 1969 (ORAUT 2010b, p. 41); however, the internal dose volume 

indicates sporadic work well into the 1970s: 

Specific bioassay for 147Pm was initiated when PNNL began manufacturing 147Pm 

heat sources in the 325 Building in 1966. The number of workers in the bioassay 

program and the number of samples were small in comparison to the numbers for 

the plutonium, uranium, or strontium bioassays, with a high of 65 workers in 

1968. There was almost no sampling from 1972 to 1975, which probably 

indicates a cessation of the original heat source program, and only 20 workers 

were sampled from 1976 to 1979. The exact end date of the heat source program 

has not been determined. For purposes of this study, it was assumed that 

exposure occurred from 1966 to 1979, although it is likely that any exposure from 

1972 to 1975 was only due to residual activity. [ORAUT 2010e, p. 122] 

The NIOSH inference from scarce bioassay data that promethium work may have stopped from 

about 1972 to 1975 appears to be incorrect. Specifically, an attachment to a February letter from 

Carl Unruh, Occupational and Environmental Safety Manager, describes an incident in which 

Pm-147 in solution was released in the 325-A Building and caused contamination “of the floor 

over a wide area.” While the clothing of a technician was contaminated, the technician was not 

(Attachment to Unruh 1975). The incident took place on January 29, 1975. 

Hence, there is clear documentation of Pm-147 processing in early 1975. Interestingly, none of 

the Pm-147 bioassay sample dates in 1975 are in the early part of the year. Five of the six 

Pm-147 bioassay samples in that year were taken on the same date, in mid-September 1975, and 

the remaining one was in November 1975. Perhaps this is because a spot check at the time of the 

incident seems to have led to the conclusion that the technician involved as well as another 

technician in the room were not contaminated. It is also interesting to note that 1975 seems to 

have been a year when bioassay sampling frequency dipped for other radionuclides as well. 

The description of this 1975 incident also notes that, “[a] similar release from the hot cell was 

detected on 9-12-74...” (Unruh 1975, pdf, p. 10). There are no Pm-147 bioassay samples in 

1974. 
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These two incidents make it clear that some Pm-147 work was ongoing in the 1972–1975 period. 

The number of bioassay samples available in the REX database is shown in Table 7. 

Table 7. Promethium-147 Sampling Frequency 

Year Number of Bioassay 

Samples Comments 

1972 from July 1 onward 16 

1973 None 

1974 1 (labeled “PM”) Sample was in April; Pm-147 incident was in September. 

1975 6 Pm-147 Incident in January; no samples until September 

1976 23 

1977 39 

1978 54 

1979 25 

1980 6 

1981–1990 5 One in 1981, three in 1985 and one in 1989 

Table 7 shows that the frequency of Pm-147 measurements increased in 1976, as recognized in 

the TBD. This may have been due to a larger number of workers involved in 1976 and 

continuing through to 1979. However, NIOSH’s inference that there was no Pm-147 processing 

in the 1972–1975 period appears to be incorrect. SC&A has not found any documentation 

regarding processing of Pm-147 in the 1980s, so it may have stopped then. However, an April 

1981 memorandum from the DOE refers to the preparation of a plan for the processing of 

Am-241 and Pm-147 (Heusser 1981). SC&A has not found any documentation as to whether or 

not this project was implemented. 

NIOSH’s assumption that there was no processing in the 1972–1975 period appears to have led 

to the proposed approach of assigning the same Pm-147 intakes throughout the 1970s as a 

claimant-favorable approach (ORAUT 2010e, Table C-28, p. 139). However, as noted above, 

there was processing in the 1972–1975 period, but little data. NIOSH has a significant amount 

of data from the 1960s that could possibly be used for the 1972–1975 period when data were 

sparse. NIOSH should examine whether the use of this data is appropriate. The coworker model 

analysis in the TBD (ORAUT 2010e, pp. 153–155) indicates that the use of data from the 1960s 

would be more appropriate, in view of the fact that there was production in the 1972–1975 

period. Furthermore, NIOSH also needs to address the issue of doses incurred during the 

incidents in the July 1, 1972, to 1975 period. 

Finding 11-1: Contrary to an inference in the NIOSH site profile, Pm-147 processing occurred 

in the 1972–1975 period, and appears to have continued into the late 1970s. 

Finding 11-2: NIOSH’s assigning the same Pm-147 intakes throughout the 1970s is not 

claimant favorable, given that processing operations occurred during portions of this period. 

NIOSH has ample Pm-147 data for the 1960s that may be more appropriate as a basis for 

coworker models, since that was also a period of production. This is indicated by NIOSH’s 

coworker analysis in the TBD. NIOSH should investigate whether this is appropriate for 

demonstrating dose reconstruction feasibility during the times in the 1970s when Pm-147 

processing occurred. Specifically, NIOSH should investigate whether working conditions were 

comparable enough to allow the use of earlier data for the period of production in the 1970s. 
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Finding 11-3: NIOSH should investigate how it is going to address possible intakes, as a result 

of documented incidents in 1974 and 1975. NIOSH should also investigate other Pm-147 

incidents in the SEC period, especially during 1972–1979. 

10.0 ISSUE 12: STRONTIUM-90, CESIUM-137, AND MIXED FISSION PRODUCT 

DOSE ESTIMATION 

Hanford operated the PUREX reprocessing plant during the July 1, 1972, to December 31, 1990, 

period, which created exposure potential in the 200 Area, including the high-level waste tank 

farms. Furthermore, Sr-90 and Cs-137 were separated in the B Plant from 1968 to 1978, and 

then conditioned for encapsulation between 1978 and 1984. Hence, there was specific exposure 

potential to MFPs during the separation period, as well as to Sr-90 and Cs-137 during the 

conditioning from 1978 to 1984 (B Plant 1985). 

10.1 Periods of Operation and Exposure Potential 

According to the TBD, the PUREX plant was shut down from June 1972 to November 1983 

(ORAUT 2010b, p. 15). However, as noted above, neptunium was separated at least until 

December 1972. According to a 1993 Westinghouse report, the PUREX plant was temporarily 

closed in early 1990, but was never reopened for production (Westinghouse 1993, p. 6). Hence, 

exposure to MFPs in PUREX plant associated facilities (200 Area) would have continued in the 

last half of 1972, which is the starting point of the SEC period under review in this report. 

Furthermore, strontium and cesium separation from high-level waste continued throughout the 

first part of the SEC period until 1978. Figure 1, taken from a B-Plant history (B-Plant 1985, p. 

12 of the pdf file), shows the history of Sr-90 and Cs-137 recovery. 

Figure 1. B-Plant Strontium and Cesium Recovery History 

As is evident from the figure, which shows cumulative recovery, very little cesium recovery took 

place during the 1976–1978 period, and very little Sr-90 recovery took place during 1978. From 

1974 to 1978, the recovered Cs-137 and Sr-90 were purified for encapsulation. From 1978 to 

November 1983, there was no separation of Sr-90 and Cs-137, but both were conditioned for 
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encapsulation. The cesium capsule production run ended in October 1983, and the Sr-90 capsule 

production run ended in January 1985 (B Plant 1985, pdf file p. 13; ORAUT 2010b, p. 16). 

Hence, except for the tank farms, the main exposure potential from reprocessing and Hanford 

high-level waste activities was to Sr-90 and Cs-137 during 1978 to 1983 period. The dates can 

be summarized as follows: 

 July 1, 1972, to 1976 or 1977: MFPs, Cs-137, and Sr-90 

 1977 or 1978 to October 1983: mainly Sr-90 and Cs-137, except possibly in the highlevel waste tank farms 

 November 1983 (PUREX restart date) to January 1985 (Sr-90 encapsulation ends): Sr-90 

and MFPs (due to restart of the PUREX plant) 

 February 1985 to February 1990: MFPs 

These dates should be kept in mind when evaluating adequacy of data for coworker models. It 

should also be remembered that other work on fission product mixes for waste treatment and 

vitrification development occurred in Building 324/325 in the 1980s. This is discussed 

separately in Section 19 (“Building 324 Incident”). 

10.2 In-Vivo and In-Vitro Data for Sr-90, Cs-137, and Mixed Fission Products 

Table 8 shows the number of workers who were monitored for MFPs from July 1, 1972, to 

December 31, 1990, by year and area of work. Table 9 shows the in-vivo monitoring for Cs-137. 

As can be seen from Tables 8 and 9, monitoring data for MFPs are rather sparse for the 1975–

1977 period, and for Cs-137 for 1975 and 1985. It should be noted that the Cs-137 

measurements in vitro that are designated as indicative of MFPs were also sparse in 1975 and 

1985 (See Appendix A). NIOSH should examine the reasons for the sudden decline in 

monitoring, and whether the earlier data can be used to fill in the gaps for this period for MFPs 

and Cs-137. Furthermore, we note that tank farm workers had exposure potential for fission 

products throughout this period. The mixed fission product data for tank farm workers appear to 

be sparse in terms of the absolute number of workers during the period 1972 to 1977 (see 

Table 8). However, as noted in Appendix A, the percentage of workers designated at some time 

during their employment as tank farm workers who were monitored for fission products at some 

time during their employment was the highest of any category. But this does not alleviate the 

concern about tank farm worker monitoring during the 1972 to 1977 period, which was much 

lower than in the subsequent period for the same group (see Appendix A, Table 21). 

There are ample Sr-90 in-vivo data for 1973 and 1974 (over 1,000 and 2,000 measurements, 

respectively). After that time, there are almost no in-vivo Sr-90 measurements in the REX 

database; measurements for Sr-90 from 1975 onwards were made by analyzing urine samples. 

There are just over 40 measurements for 1976, and the data adequacy for that year needs to be 

examined. There are more than 500 bioassay records for 1978 and each year after that until 1990 

(inclusive). 
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Table 8. Number of Workers Monitored for Mixed Fission Products, July 1, 1972, to 

December 31, 1990 

Number of Workers Monitored for Fission Products (In Vitro + In Vivo)* 

Year All Workers 100 Area 100N Area 200 Area 200 Tank 

Farms 300 Area 400 Area 700 Area

1972 110 20 19 37 [redacted] [redacted] [redacted] 0 

1973 1403 149 131 179 [redacted] 291 22 22 

1974 2008 238 224 240 13 477 40 29 

1975 104 [redacted] [redacted] 46 [redacted] 25 [redacted] [redacted]

1976 32 [redacted] [redacted] [redacted] 0 22 0 [redacted]

1977 114 [redacted] [redacted] 58 [redacted] 25 [redacted] 0 

1978 421 27 25 305 35 47 [redacted] [redacted]

1979 682 65 55 518 86 51 12 [redacted]

1980 668 88 72 535 85 46 [redacted] [redacted]

1981 729 106 92 611 93 52 12 [redacted]

1982 492 68 60 422 73 34 12 [redacted]

1983 455 48 34 417 82 36 12 [redacted]

1984 744 70 48 715 92 56 22 [redacted]

1985 818 142 107 755 112 53 23 [redacted]

1986 908 251 214 757 113 88 28 [redacted]

1987 991 217 181 872 122 73 30 [redacted]

1988 1234 369 339 1052 125 221 63 29 

1989 1304 352 325 1124 137 206 55 44 

1990 1340 325 297 1194 123 186 70 36 

All Years: 6588 1328 1218 3091 275 1149 225 152 

* Includes a small number of Cs-137 in-vitro results, driving sampling for in-vitro FP was for Sr-90. 

The results presented in bold italics indicate areas and years with exposure potential (see Appendix A). 

Table 9. Number of Workers Monitored In Vivo for Cs-137 

Number of Workers Monitored In Vivo for Cesium-137 

Year All Workers 100 Area 100N Area 200 Area 200 Tank 

Farms 300 Area 400 Area 700 Area

1972 993 91 84 122 [redacted] 211 16 14 

1973 1879 208 185 203 [redacted] 408 31 33 

1974 2080 253 236 226 [redacted] 487 43 33 

1975 [redacted] [redacted] [redacted] 0 0 [redacted] 0 0 

1976 1926 305 281 253 [redacted] 556 47 41 

1977 2761 429 393 694 74 661 82 33 

1978 3479 534 492 854 75 841 175 61 

1979 4323 746 681 1225 104 959 318 58 

1980 4575 819 748 1447 125 1305 419 90 

1981 5332 984 914 1874 152 1495 551 105 

1982 5402 1438 1349 2359 187 1716 631 165 

1983 5507 1657 1528 2532 196 1693 606 205 

1984 441 307 295 148 14 106 26 21 

1985 52 25 24 28 [redacted] 9 0 [redacted]

1986 3562 1312 1236 1649 115 998 326 247 

1987 7109 2512 2344 3472 283 2058 678 308 

1988 6463 2232 2098 3381 280 2086 702 241 

1989 7974 2002 1863 4565 299 2392 876 406 

1990 8474 1922 1788 4949 309 2451 918 414 

All Years: 23629 4530 4209 7600 335 4407 1414 1078 

The results presented in bold italics indicate areas and years with exposure potential (see Appendix A). 
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10.3 Quality of Data 

The 1982 data quality report indicates considerable problems with meeting contractual MDA’s 

for a number radionuclides important to MFPs (though not Sr-90): 

Based on analysis of audit samples submitted to UST [U.S. Testing] with zero 

activity of the following radionuclides, it is concluded that UST is unable to 

achieve the contractual detection level (Ld) for the following radionuclides: 

140Ba-La, 22Na, 24Na, 137Cs, 59Fe, 106Ru,134Cs, 239Np, 95Nb-Zr, 131I. As with 

uranium, this is determined by analysis of audit samples which did not contain 

any of these radionuclides. (Fleishman 1982) 

None of the other audit reports reviewed (Bihl 1986; Hickman 1982) contained a substantive 

discussion of these radionuclides. Stronium-90 limits were apparently being met. SC&A has not 

located similar audit reports for the 1970s. 

NIOSH has not discussed the implications of this problem for dose reconstruction where data for 

these radionuclides are available. 

10.4 NIOSH’s Approach to Coworker Models 

NIOSH proposes to assign the same intake rate for Sr-90 in the 1971–1988 period as in the 

period extending from 1989 to the present, when there was essentially no exposure potential due 

to production operations (ORAUT 2010e, Table C-27, p. 139). This does not appear to be 

claimant favorable in view of the extensive operations involving Sr-90 until January 1985. The 

year-by-year data from the 200 Area should preferably be used for the coworker models not only 

for Sr-90, but also for mixed fission products and Cs-137. There appear to be ample data for 

most years. NIOSH will have to examine how to approach the selection of an intake level for the 

years in the mid-1970s, when there are far fewer data points than in other years in the time 

periods examined here. 

NIOSH does not propose to use mixed fission product data for coworker dose assignment: 

A large number of different radionuclides were present at Hanford at various 

times, but the available bioassay data for radionuclides in addition to those in this 

attachment were considered to be too few to be statistically reliable for intake 

estimation. Workers who were exposed to 137Cs, 90Sr, 24Na, and 65Zn could also 

have been exposed to other fission and activation products. From 1960 to 1988, 

intakes of most fission or activation products would have been detectable in 

whole-body counts, but the recording practice for fission and activation products 

other than 137Cs, 24Na, and 65Zn was not amenable to statistical analysis. 

Guidance for assigning intakes of additional radionuclides is provided in 

ORAUT-OTIB-0054 [ORAUT 2007b] and for specific buildings in Section 5.6. 

Assignment to unmonitored workers of radionuclides that were not analyzed in 

this study should be in accordance with guidance in Section 5.6. [ORAUT 2010e, 

p. 141] 
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The coworker model approach in ORAUT-OTIB-0054 (ORAUT 2007b) states that it applies “to 

intakes of fission and activation products associated with most reactor operations, destructive 

fuel examination, fuel dissolution, and high-level waste management.” However, it only 

specifies source terms for reactors. A procedure for estimating dose from air monitoring data is 

also provided, but the Hanford TBD does not provide such data for mixed fission product dose 

reconstruction. 

Hence, ORAUT 2007b could be applied to N Reactor and other reactor workers who were not 

monitored for fission products. However, it would not apply to PUREX operations (until the end 

of 1972, and then from November 1983 onward), or to source terms related to the processing of 

waste, notably those related to the separation of Sr-90 and Cs-137 for high-level waste until 

1978. 

Since ORAUT 2007b cannot be used for unmonitored separations and waste-handling workers, 

there is a circular reference between Attachment C and Section 5.6 for these operations. As the 

quote above shows, Attachment C of ORAUT 2010e refers to Section 5.6, which is for 

unmonitored worker dose estimation procedures; yet Section 5.6 refers back to Attachment C for 

fission and activation product dose estimation for operations relating to the 200 Area: 

Use fission and activation products (from September 1, 1946 on), plutonium (from 

September 1, 1946 on, types M, S, or Super S), and uranium (type F from January 

1, 1949 on) from Attachment C (except 147Pm) and [ORAUT (2007b] but may 

ignore 24Na from Attachment C for best estimates. [ORAUT 2010e, p. 64] 

Similarly, NIOSH refers back to Attachment C of ORAUT 2010e for fission product dose 

estimation for unmonitored workers for several 300 Area buildings where fission products were 

handled and processed [including Buildings 324, 325, 327, and 3706 (ORAUT 2010e, p. 65)]. 

This is puzzling, since Attachment C of ORAUT 2010e does not specify a coworker model for 

mixed fission products in general, though it does have coworker intake assignments for several 

specific ones, such as Sr-90, Cs-137 and Pm-147. 

10.5 Findings for Mixed Fission Products, Sr-90, and Cs-137 

Finding 12-1: Overall, there seem to be ample data for Sr-90 and Cs-137 to create coworker 

models, except in certain years. 

Finding 12-2: NIOSH’s approach for assigning the same Sr-90 intake values for the period 

from 1971 onward is not claimant favorable. A more claimant-favorable approach needs to be 

developed. The Cs-137 and mixed fission product models should also be reviewed in light of the 

Sr-90 and Cs-137 recovery operations and the processing to condition Cs-137 and Sr-90 for 

encapsulation out to the completion of the encapsulation process in October 1983 for Cs-137 and 

January 1985 for Sr-90. 

Finding 12-3: The mixed fission product data for tank farm workers for July 1, 1972, to the end 

of 1977 are very sparse. NIOSH needs to examine whether its present approach is adequate for 

tank farm workers, or whether a special coworker model is needed for them for this period. 

Finding 12-4: The implications of the failure to meet contractual MDA limits for several fission 

products (though not Sr-90) are unclear. NIOSH has not discussed this issue. 
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Finding 12-5: NIOSH has specified coworker models for several specific fission products. The 

method specified by NIOSH is applicable to reactor areas, but not to separations or waste 

handling and processing areas. An explicit approach for unmonitored workers in reprocessing 

and waste handling and processing areas needs to be developed for radionuclides that have not 

yet been covered for the 200 Area and for certain buildings in the 300 Area. 

11.0 MATRIX ISSUE 13: TANK FARM ALPHA MONITORING 

SC&A has analyzed data completeness and adequacy by area, and broken out tank farm areas 

separately. Hence, this issue is subsumed under the other headings discussed in this report, as 

well as in Appendix A. 

12.0 MATRIX ISSUE 14: PLUTONIUM DATA ADEQUACY 

Hanford had extensive operations involving plutonium and has extensive monitoring data for 

plutonium in the period under consideration. SC&A compiled the data from the REX database 

to analyze for completeness as to quantity; SC&A also analyzed the quality of the available data 

and the REX data entries for plutonium. The REX database reports results in microcuries per 

sample of dpm/sample. Most of the plutonium results are below the MDA. NIOSH uses these 

results in its dose reconstruction to assign missed dose. 

12.1 Plutonium Data – Quantity 

Table 10 shows the number of workers monitored each year for plutonium. It is clear that a large 

number of workers had either bioassay or in-vivo records for plutonium for this period, 

especially in the 200 and 300 Areas where there was exposure potential. 

It is clear from Table 10 that there are ample plutonium monitoring data available. However, 

since NIOSH is using only bioassay data for its coworker model, we also reviewed how many 

positive results are available by year. The majority of the plutonium monitoring consisted of invitro measurements. 

Table 11 shows the positive plutonium bioassay results by year. 

It is important for NIOSH to evaluate whether only positive results should be used for coworker 

modeling, especially for unmonitored workers in the PUREX (202-A) reprocessing plant and the 

plutonium finishing facilities (231-Z and 234-5Z) during the years of their operation in the 

period under review here (see Section 12.2). 

It should be kept in mind that in compiling the information in Table 11, we have inserted an 

interpretation of what the units are in many of the measurements. A large number of these 

measurements are entered into the REX database without units. However, since we know the 

MDA, a reasonable inference about units can be made. Without making such an inference, there 

would be a few years (1975, 1977, and 1979) when there would be no interpretable in-vitro data. 
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Table 10. Number of Workers with Plutonium In-Vitro or InVivo Monitoring Data by Year 

Number of Workers Monitored for Plutonium (In Vitro + In Vivo) 

Year All Workers 100 

Area 

100N 

Area 

200 

Area 

200 Tank 

Farms 

300 

Area 

400 

Area 

700 

Area 

1972 965 42 38 316 14 204 16 13 

1973 1514 72 60 483 21 332 38 20 

1974 1645 84 72 539 21 363 44 19 

1975 1750 85 72 567 28 425 47 22 

1976 1866 104 89 636 45 431 56 23 

1977 1795 92 78 694 44 432 74 19 

1978 1486 122 106 475 19 485 147 14 

1979 1159 85 71 559 39 337 81 16 

1980 1591 136 116 555 37 579 269 25 

1981 1443 110 94 620 49 576 291 25 

1982 1327 73 58 531 43 602 325 13 

1983 1572 57 39 721 68 640 297 19 

1984 1514 85 51 904 76 505 228 17 

1985 1811 100 55 1030 86 567 261 22 

1986 1767 103 59 1002 90 542 220 11 

1987 1903 105 63 1124 84 492 238 16 

1988 2079 419 381 1359 86 586 235 36 

1989 2122 409 371 1459 78 615 214 54 

1990 2047 438 403 1496 62 535 193 50 

All Years 10996 1240 1124 3964 233 2057 839 224 

The results presented in bold italics indicate areas and years with exposure potential (see Appendix A).

Table 11. Positive Plutonium Bioassay Results by Year 

Year Pu Pu-239 Pu-238 Total 

1972 105 41 1 147 

1973 0 253 6 259 

1974 0 181 5 186 

1975 0 88 3 91 

1976 0 89 0 89 

1977 0 53 3 56 

1978 0 106 0 106 

1979 0 106 1 107 

1980 7 183 68 258 

1981 27 5 0 32 

1982 239 0 0 239 

1983 53 17 5 75 

1984 0 196 81 277 

1985 0 217 199 416 

1986 0 195 65 260 

1987 0 109 15 124 

1988 0 265 9 274 

1989 0 178 27 205 

1990 0 44 4 48 
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12.2 Plutonium Data – Quality 

The MDA value for plutonium in-vitro data from October 1955 to September 30, 1983, was 

0.05 dpm/sample (ORAUT 2010e, Table 5-1, p. 18). The value reported in case of results below 

MDA is called the decision level or reporting level equal to half this value (reported as 1.1 × 

10-8 μCi/sample until 1974, and 0.025 dpm/sample after that in the REX database). After 

October 1983, the results for Pu-238 and Pu-239+240 have been reported separately. 

NIOSH also did fecal sampling for plutonium: 

Fecal sampling was normally done in response to suspected intakes, but routine 

fecal sampling was used for some high-risk plutonium workers, mostly operators 

of Plutonium–Uranium Extraction (PUREX) facility and the Plutonium Finishing 

Plant, from 1986 through June 1989. The special study showed that, when 

considered as a group, the mean fecal excretion was statistically significantly 

different from control. Enhanced air sampling, which was initiated in response to 

the study, showed frequent but intermittent releases of plutonium in the 

workplaces, at levels below the detectability of normal air sampling. MDAs are 

listed for nonroutine plutonium excreta analysis. When modeled as chronic 

intake, the intakes and doses were low (less than 10 mrem committed effective 

dose equivalent)... [ORAUT 2010e, pp. 19–20] 

Since the modeled doses were low, NIOSH suggests modeling these intakes as routine rather 

than as being due to incidents (ORAUT 2010e, p. 20). 

Plutonium at Hanford was a mixture of several isotopes, notably Pu-238, Pu-239, Pu-240, and 

Pu-241. Pu-238 and Pu-239/240 were measured separately only from 1983 onwards. Their ratio 

can be used to estimate the composition of the plutonium mixture, as noted by NIOSH (ORAUT 

2010e, p. 21), and also to infer the activities of Pu-241 and Am-241. The ratios suggested by 

NIOSH in the TBD appear to be reasonable and claimant favorable (see ORAUT 2010e, pp. 20–

22). Specifically, the assumption of 12% Pu-240 when there is no information about isotopic 

composition is generally reasonable and claimant favorable. For the coworker model, a set of 

default assumptions about the age of the plutonium have been suggested by NIOSH (ORAUT 

2010e, Table C-8, p. 129). These assumptions are scientifically reasonable and claimant 

favorable. Before 1983, the measured quantity was total alpha from plutonium, which means the 

total of Pu-238 and Pu-239+240. In such cases, the information can be supplemented by 

available Am-241 data to infer how much Pu-241 and Am-241 were present at intake. 

Finally, NIOSH also did comparisons of in-vitro versus in-vivo counts. There were 100 positive 

Am-241 chest count records analyzed together with the corresponding in-vitro bioassay records 

for Pu-238, 239 (100 sets). Each record was examined to determine if a positive urine or fecal 

sample were associated with the positive chest count. 

Just over half (52) of the 100 sets of data points showed negative in-vitro bioassay results when 

there were positive chest counts. There were also cases where the in-vitro bioassay discovered 

the uptake and that was the reason that the chest count was initiated. This study made it apparent 

that in-vivo counting and in-vitro bioassay programs are complementary. They work together to 

identify incidents and evaluate internal doses due to them. 
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While SC&A will review issues associated with the U.S. Testing data quality for the 1987–1989 

period separately (as stated at the beginning of this report), it is important to note that audits of 

the bioassay program were conducted routinely at least during the 1980s. Some results from the 

period prior to the 1987–1989 period (inclusive) covered by SEC-00155 are important for the 

plutonium program (and some others as well). Specifically, a 1982 audit report noted that 

plutonium measurements were systematically low in 1981 and that, despite improvements, the 

problem persisted into 1982. NIOSH uses the contractual value of 0.05 dpm/sample as the 

MDL; however, the actual MDL was closer to 0.1 dpm/sample. The underestimates of 

plutonium in urine extended to values higher than 0.1 dpm as well (Fleishman 1982, pdf p. 4 and 

p. 6). The problem for plutonium sampling appears to have been corrected by the 1984 audit 

(Spitz 1984). SC&A has not located similar audits for the 1970s. The data in the audits may 

provide a basis for corrections. 

12.3 NIOSH Approach to Coworker Model for Plutonium 

NIOSH has presented its approach to coworker modeling based on the plutonium measurements 

that are available for the period. NIOSH includes below MDA measurements in determining the 

intake values corresponding to various assumptions about solubility. The coworker model 

approach suggested is a standard one used by NIOSH and is not an SEC issue as such. NIOSH’s 

coworker model for plutonium uses a linear distribution to fit the data below detection limit. The 

TBD notes that many plutonium samples are related to incidents; in such cases, workers were 

often sampled multiple times. NIOSH redacted the data in the following way for developing its 

coworker model: 

If there were a large number of samples from a single worker, usually bunched 

after a specific time, these were removed. A specific number was not rigorously 

applied as the definition of large, but it was generally greater than about 20 in 

1 year. In addition, once a person was selected to be removed, data for 

subsequent years in that worker’s history were removed until the excretion was 

near or less than the reporting level. [ORAUT 2010e, p. 118] 

Removing incident-related samples for coworker modeling is appropriate. However, SC&A is 

concerned that in removing samples from subsequent years until “the excretion was near or less 

than the reporting level” may miss routine exposures that were high, for instance during 

intermittent production or processing campaigns or during waste-handling operations. This 

approach may bias the coworker model towards lower intake values. 

Since the measurements for 1981 did not meet the contractual MDA limit of 0.05 dpm/sample, 

but rather only achieved an MDA closer to 0.1 dpm/sample, NIOSH should adjust its coworker 

model to reflect this fact. Similarly, NIOSH needs to investigate the correction factors for 

plutonium results in the period up to June 1984, and how a claimant-favorable approach to dose 

reconstruction for those with data and to unmonitored workers needs to be created. The TBD 

mentions exposures to Super S, but the coworker model only considers Types M and S (ORAUT 

2010e, p. 137). Specifically, the TBD mentions in several places that Super S plutonium was 

present at Hanford (ORAUT 2010e, pp. 22 and 53). 
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Finally, NOISH’s coworker model and intake estimation only goes up to 1988 (ORAUT 2010e, 

Tables C-9 and C-10, p. 131), though the SEC period extends to the end of 1990. It is important 

for NIOSH to ensure that this approach is claimant favorable, especially for separations workers. 

12.4 Findings for Plutonium Modeling 

Finding 14-1: NIOSH has not considered Type Super S plutonium for intake estimation.

Finding 14-2: NIOSH should take the different, non-contractual MDA for 1981 into account in 

its coworker model. 

Finding 14-3: NIOSH should ensure that extending 1988 values to 1989 and 1990 is claimant 

favorable for separations workers. 

Finding 14-4: NIOSH’s method of redacting samples relating to incidents may bias coworker 

intake estimates towards a lower intake level. 

13.0 ISSUE 16: CURIUM-244 AND AMERICIUM-241 

13.1 Curium-244 

Major work with curium appears to have ended in the late 1960s. There was a 1974 proposal to 

separate transplutonium radionuclides, including curium and americium isotopes, from waste 

generated at the Barnwell Nuclear Fuel Plant, which was being built at the time (Battelle 1974). 

However, the Barnwell plant never reprocessed spent fuel, and it appears that this project was 

not implemented. 

There were some experiments with Cm-244 up to and including 1973, according to the TBD: 

PNNL’s Space Nuclear Systems Division reported on laboratory work in March 

1969 on phase transformations in curium oxide in which a small sample of 

curium oxide was heated in a vacuum in a differential thermal analysis apparatus 

(Hansen 1969c, p. 4). The building where the work took place was not 

mentioned. One of the authors was interviewed briefly by phone. He said that he 

and a few others worked with milligram quantities for special studies in the 308 

Building between about 1969 and 1973. All the work was done in gloveboxes. 

Each experiment was of short duration (a month or so), but there were multiple 

experiments. The curium was not from Hanford. Because quantities of curium 

and other higher order transplutonium elements were scarce, who, when, and 

how much of these special isotopes was made available to researchers was 

controlled by the AEC-wide Transplutonium Research Coordinating Committee. 

That committee arranged for delivery of the samples, which were used then 

shipped to the next researcher on the priority list. The researcher said there were 

no contamination spreads with the curium samples (Bihl 2009b). [ORAUT 

2010b, p. 49] 

However, Hanford documentation indicates Cm-244 exposure potential and possible 

handling and processing well beyond 1973. For instance, this is indicated in the Monthly 

Report for June 1978: 
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One radiation occurrence involving the loss of containment and potential 

inhalation of 244Cm was formally reported to RL. Other occurrences documented 

internally included a repeat of the above occurrence at 231-Z and inhalation of 

90Sr and 137Cs by a technician at the 324 Building. [Jech 1978a] 

Similarly, the monthly report for October 1978 reported intakes of curium: 

A series of inhalations of curium-244 has begun at 331 and is expected to 

continue for several weeks as are the Chromium 51 inhalations. This is not 

creating any undue problems. [Jech 1978 b] 

The context of these monthly reports indicates that these intakes may have occurred 

during cleanup or refurbishment of facilities. It may explain a single curium urinalysis 

done in June 1978. However, none were reported for October 1978. 

13.2 Curium-244 Data 

There is only one in-vivo data point for Cm-244 in the REX database in 1978. There were 15 invitro measurements reported in the 1970s (starting on July 1, 1972); but none of the 

measurements were made in 1973, when experimental work was done at Hanford. As noted, 

above, there was a measurement in June 1978, when an incident was noted; no measurement 

corresponding to the October inhalation incidents is in the REX database. 

It is possible that the 1974 to 1979 and the scattered samples thereafter until 1981 (inclusive) 

were related to incidents during refurbishment and cleanup operations. There are no data for 

1982 and also none for 1983 until October (inclusive). SC&A has not researched the reasons for 

the sampling in detail, but notes that exposures occurred well after the reported ending date of 

exposure potential related to curium processing. There appear to be ample routine urine data in 

the REX database for Cm-244 starting in November 1983. 

13.3 NIOSH Approach to Curium-244 

The TBD provides the following approach for Cm-244 after July 1, 1972: 

Limited use of routine urinalysis was conducted starting in 1974. If an intake 

estimate is required, unmonitored workers should be assigned intakes at the 95th

percentile of the plutonium coworker intakes (constant distribution) prior to 1974 

and at the 50th percentile (with GSDs) from 1974 on. [ORAUT 2010e, p. 60) 

NIOSH has not provided any analysis to show how the various situations of curium 

exposure potential could reasonably be modeled by using the 95th percentile of plutonium 

intakes. Furthermore, as noted above, Cm-244 data are too sparse for coworker 

modeling. 

13.4 Americium-241 

Americium-241 is created by the decay of Pu-241 (which is associated with plutonium-239/240) 

and hence can be inferred from plutonium data if the age and isotopic composition of the 

plutonium is known or can be assumed in a claimant-favorable manner. NIOSH has specified its 
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approach for Am-241 in plutonium mixtures in the TBD (ORAUT 2010e, Table 5-5).7

 However 

Am-241 separation operations were also conducted in Building 325 during the period under 

review in this report. The radiation dose due to such operations cannot be inferred from 

plutonium in-vivo or in-vitro data, since the objective of the process is to produce pure Am-241 

unmixed with plutonium. According to the TBD, about 3 kilograms of pure Am-241 was 

separated in the A-Cell of Building 325 during the 1973 to 1976 period and shipped to Oak 

Ridge National Laboratory. This statement was based on a telephone interview, which referred 

to a set of slides that is not available (ORAUT 2010b, pp. 34–35). 

13.5 Americium-241 Data 

It should be noted that 3 kilograms of Am-241 represent a very large amount of alpha-emitting 

material—more than 10,000 curies. Table 12 shows the number of workers who were monitored 

for Am-241. 

Note that there are very few measurements in 1975. Furthermore, as we have noted in some 

other cases, tank farm workers have very sparse monitoring data until 1977 or so. The reason for 

the sudden decline in monitoring in 1975 is unclear at this time. It is possible that it was related 

to a strike at Hanford in that year. We note that there were many measurements in the 200 and 

300 Areas, in parts of which there was potential for americium exposure. The measurements 

appear to be a mix of routine and incident-related samples. 

Table 12. Number of Workers Monitored for Am-241 by Area 

Number of Workers Monitored for Americium (In Vitro + In Vivo) 

Year All Workers 100 Area 100N Area 200 Area 200 Tank 

Farms 300 Area 400 Area 700 Area

1972 218 12 10 61 [redacted] 70 [redacted] [redacted]

1973 450 29 24 111 [redacted] 190 16 [redacted]

1974 618 45 38 110 [redacted] 256 12 [redacted]

1975 31 [redacted] [redacted] 12 [redacted] 9 [redacted] [redacted]

1976 730 39 33 124 [redacted] 335 21 [redacted]

1977 863 62 57 159 [redacted] 353 46 [redacted]

1978 1224 83 73 295 17 486 117 16 

1979 1398 102 85 386 20 485 234 [redacted]

1980 1491 87 72 429 24 628 262 21 

1981 1535 116 95 494 24 699 303 16 

1982 1380 83 62 437 31 778 404 12 

1983 1346 85 54 544 47 711 314 15 

1984 1398 111 81 614 50 719 304 11 

1985 1469 159 117 715 64 738 266 17 

1986 1582 219 177 859 87 738 223 13 

1987 1517 169 132 886 82 690 233 13 

1988 1563 314 278 1044 78 627 222 24 

1989 1819 374 336 1279 75 663 205 27 

1990 2338 550 507 1768 103 732 208 29 

All Years 7370 1136 1031 3203 187 2199 849 119 

* The results presented in bold italics indicate areas and years with exposure potential (see Appendix A). 

 7

 The selection of the Am-241 to Pu-239 ratio is not an SEC issue and hence has not been reviewed in this 

report, since plutonium data for dose reconstruction are available. 
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13.6 NIOSH Approach to Americium-241 

NIOSH has specified the following approach to estimating Am-241 dose for unmonitored 

workers: 

Significant quantities of pure 241Am or mixtures enriched in 241Am were present in 

several Z-Plant facilities through 1976 and at times at a few other facilities.... 

Exposure prior to July 1, 1972 is covered by the SEC because it was determined 

to be infeasible to reconstruct doses. Bioassay monitoring via chest counts was 

used for significantly exposed workers from 1968 on and urinalysis was used for 

special monitoring after incidents with potential for intake. If an intake estimate 

is required from July 1, 1972 on, unmonitored workers should be assigned intakes 

at the same values as the plutonium coworker intakes using the 50th percentile 

values and the GSDs from Attachment C. [ORAUT 2010e, p. 60] 

Using plutonium as the reference radionuclide for Am-241 when the latter is a contaminant in 

plutonium is a reasonable approach (though NOISH should justify its choice of the 50th

percentile value). However, plutonium is not an appropriate reference radionuclide for 

operations involving separation of americium or its processing apart from plutonium. NIOSH 

needs to specify a coworker model for unmonitored workers in such cases, especially until 1976 

(inclusive), when the last major operation in which americium was not associated with plutonium 

took place. 

13.7 Findings Relating to Curium-244 and Americium-241 

Finding 16-1: Curium-244 exposure potential existed due to experimental operations in 1973 

and due to other operations, probably related to refurbishment of facilities, later in the SEC 

period. SC&A has documented exposures due to incidents in 1978. It is unclear whether this 

potential continued into the 1980s. Routine sampling appears to have begun in November 1983. 

Finding 16-2: Curium-244 data do not exist for 1973 and are sparse thereafter until October 

1983.

Finding 16-3: A coworker model for Cm-244 may be feasible after 1984 (inclusive). 

Finding 16-4: NIOSH has not provided a technical rationale for using plutonium data for Cm244 dose reconstruction. This does not appear to correspond to documented situations with Cm244 exposure potential. 

Finding 16-5: NIOSH should develop a coworker model that is specific to Am-241, apart from 

its association as a trace contaminant in plutonium, especially for the period up to 1976. This 

may be difficult, given that much of the Am-241 in-vitro monitoring was in the context of 

incidents. 

14.0 MATRIX ISSUE 19: LACK OF ADEQUATE MONITORING (PETITIONER 

ISSUE) 

NIOSH has an extensive discussion of internal monitoring data in its revised TBD (ORAUT 

2010e). SC&A has done an extensive review of the available data for completeness and 
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adequacy. The detailed analysis is presented in Appendix A of this report. The above discussion 

analyzes the data by radionuclide for the major radionuclides at the Hanford Site and details 

findings in each case. 

As noted above, data are adequate for dose reconstruction and for coworker models in some 

cases. In other cases, they are not. In yet other cases, there are specific findings and suggestions 

for NIOSH; specifically to examine the feasibility of using available data to fill in gaps that exist. 

Some radionuclides other than those discussed above are discussed below in the order in which 

they appear in the SEC Issues Matrix (Appendix B). 

15.0 MATRIX ISSUE 20: SKIN CONTAMINATION 

Since external dose data records would reflect routine skin exposure, the issue of skin 

contamination is related to that of incidents and is discussed with Matrix Issue 22 below. 

16.0 MATRIX ISSUE 21: MISSING AND DESTROYED RECORDS 

SC&A made a request to DOE for a list of destroyed records. Since records are routinely 

destroyed according to a prescribed protocol, a list was made available to SC&A. It consists of a 

list of about 71,000 boxes of records that were destroyed. Almost all these records are either 

prior to the period being reviewed here (i.e., earlier than July 1, 1972) or are administrative 

records that are not relevant to this review. Some of the labels used for describing the contents 

of such boxes are as follows: 

 Purchase orders 

 Magnetic tapes (possibly backup tapes for records) 

 Vouchers – Acct Pay 

There are general ledger files and other records that appear to be administrative. There are a 

large number of boxes with dosimetry data, labeled “personnel badge film” and “BIOASSAY 

ELECTRO CA” that also bear the label “General Electric.” These are presumably records from 

the time that General Electric was the main contractor at Hanford (until 1964). These records are 

not directly relevant to the investigation of completeness and adequacy of data for this SEC 

review, since NIOSH has already stated that dose reconstruction is not feasible until June 30, 

1972, on a number of counts. However, we note that for radionuclides that have a long residence 

time in the body, loss of records relating to the pre-1972 period could affect dose estimates in the 

period under consideration here if the incidents are not noted in the individual’s raw dose 

records. However, comparison of raw individual records to the electronic incident database 

reveals that the former is much more complete (See Section 17.0). 

Unfortunately, a large number of the box descriptions have no date indicated in that column. 

There is a “To Date” column in the records list, but this appears to be the date on which the 

documents were boxed, rather than the latest document date of materials in the box. 

A small number of boxes appear to have data relevant to the period under consideration, as well 

as a year indicating the date of the documents in the box. For example, there is a box of tritium 

data with the date: 

 Box #53082 – “TRITIUM DATA 1976” 
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In other cases, which are more frequent, the box descriptions appear relevant, but there are no 

dates indicated as to the period in which the documents were created, as shown by the following 

examples: 

 Five boxes labeled “Unusual incident” with no dates (box numbers 37722, 53692, 58218, 

64877, and 64882) 

 Two boxes titled “VAULT INVENTORIES” with no dates (box numbers 75016 and 

75017) 

There are some boxes that may be relevant, but whose titles are difficult to interpret. For 

instance, there are several boxes with the labels “WASTE MANAGEMENT ENVIRON W”; 

none of them have dates. 

SC&A’s review of the box titles, dates, and other date fields in the spreadsheet of destroyed 

documents indicates that some boxes with relevant personnel dosimetry data and incident data 

were destroyed. While most do not have date ranges for the documents in them, some box titles, 

such as Box #53082 mentioned above, are from the period being reviewed here. 

It is not possible for SC&A to determine whether the data that may be relevant in the destroyed 

boxes is in the personnel files or in the REX database or both. Generally, external dose data 

seem to be reasonably complete in the REX database for the period under consideration and were 

declared by the Work Group to not be an SEC issue. Internal data are not complete and are 

inadequate in some cases, as discussed above. It is not possible, based on the review done here, 

to determine whether there are more data that could fill some of the essential gaps. NIOSH has 

done extensive data capture and SC&A has reviewed the available data, so it is likely that any 

duplicates would have been discovered in the extensive effort that NIOSH has made to compile 

information relative to internal dose reconstruction for the period under review. 

16.1 Findings for Matrix Issue 21: Destroyed and Missing Records 

Finding 21-1: A large number of records have been destroyed as a matter of government 

record-keeping practices. The vast majority of these records were administrative records of no 

relevance to dose reconstruction. 

Finding 21-2: A small portion of these boxes appear to have contained data relevant to dose 

reconstruction, but most of that, in turn, appears to have been prior to the period under 

consideration. Some boxes that may be relevant have no dates indicating the period of the 

documents in the box. Finally, a small proportion of the boxes have labels that indicate they are 

relevant and a date that places some or all of the contents in the period under review here. 

SC&A has not been able to establish whether the destroyed boxes contained data that could fill 

in some of the gaps described above for internal dosimetry. 

17.0 MISSING INCIDENT RECORDS 

As discussed above, there are some incident records that have been destroyed. The contents of 

the destroyed boxes, and even the dates of those incidents, were not discovered during this 

investigation. SC&A took a couple of different routes to determining whether personnel records 

contain data on all incidents. 
Effective Date: 

September 30, 2011 

Revision No. 

 1 – Draft 

Document No. 

Draft – Hanford SEC Issues Review – Vol. I 

Page No. 

 52 of 71 

 This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

NOTICE:

First, when researching the SRDB, SC&A came across descriptions of incidents. As noted in 

Observation 1 in Appendix A, SC&A found that workers were generally scheduled for bioassay 

when there were incidents. SC&A did not connect these cases to actual bioassay data to see if 

the measurements were actually done. We know from the Cm-244 examples that incidents were 

not always followed up by bioassay, at least as reflected in the REX database. 

SC&A also attempted to determine whether the REX incident database contains information 

about incidents that are either mentioned in the individual claimant files supplied to NIOSH by 

the DOE or in the CATI conducted by NIOSH as part of the dose reconstruction process. The 

review was based on claimant records. 

17.1 Incidents in REX Incident Database compared to Individual Records 

To examine the extent and validity of the REX incident database, hardcopy records were 

examined for a subset of claimants who had documented incidents in the database during the 

SEC period. There were 40 claimants selected at random for examination; however, once the 

hardcopy records were examined, it was found that 10 of these workers spent some or all of their 

SEC employment at Pacific Northwest National Labs (PNNL). Therefore, an additional 10 

claimants were selected who were identified as working only at Hanford. In addition, 1 of the 

original 40 random claims could not be opened via the NIOSH claimant database, so this 

claimant was removed from the analysis and replaced with another. 

The incident database contained 63 separate incidents for the 50 claimants examined. Of the 63 

incidents, 59 of these incidents (or ~94%) were reflected in the hardcopy records for the 

claimants. In the four cases where the incident was not reflected in the hardcopy records, the 

DOE_Response files appeared to be incomplete compared to the other claims reviewed. The 

files only contained x-ray records, internal monitoring, and annual dose totals, and did not 

contain the other documents typically included in the claimant records for Hanford, such as 

training certifications, visitor forms, mask fittings, monthly external dose reports, and lost 

dosimeter investigation forms. In three of the four cases, the incident was mentioned in the 

CATI report. 

Conversely, the hardcopy records documented 92 separate incidents during the SEC period for 

the 50 claims examined. Of these 92 incidents, 33 (or about 36%) were not found in the 

electronic database. It is not clear why such a large portion of incidents were omitted from the 

database; it did not appear as though the severity of the incident had any effect on whether the 

incident was entered into REX. Table 13 contains specific information about the incidents 

identified in the electronic database, including date of the incident, whether it was included in the 

hardcopy records, a description of the incident (transcribed from the hardcopy record or CATI), 

and any additional comments.
Effective Date: 

September 30, 2011 

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Table 13. Summary of Incidents Reviewed 

Incident Date 

Incident 

Included in 

Hardcopy 

File? 

Description of Incident From Hardcopy Record/CATI Addition

[redacted]/1973 No 

From CATI report: "[Location] In the lab in 325 Bldg... Something got loose and he got 'crapped up' (i.e., contaminated) and he lost all his clothes and he had to be scrubbed until he was raw." 

Incident is mentioned Hardcopy records app

[redacted]/1974 Yes 

"While working in laboratory [redacted] 308 Building received coverall 

contamination... The concerned employee had inspected the gloves prior to starting his work but after 10 minutes of work he removed his hands from the glove box and noted that both gloves had ruptured. He remained at this location and had Operational Health Physics notified." 

Claimant had 3 docum

[redacted]/75, [redacte

were not included in talso had a documente

[redacted]/79 that was 

[redacted]/1980 Yes 

"A [redacted] and [redacted] Operator entered the 221-T Canyon to observe 

removal of a B Plant jumper pan from the Multi-purpose transfer box for 

placement on the canyon deck at Section 12. Dose rates from the jumper pan were to be obtained after it had been removed to the deck. The CAM's in the canyon alarmed when the crane operator started lifting the jumper pan out of the Multi-purpose box. Personnel immediately left the canyon. CAUSE: Unexpected loose contamination in jumper pan that became airborne when pan was lifted by the crane." 

The first DOE responrecords for the wrong the same, but the first differ. 

[redacted]/1983 Yes 

"Involved in an exposure incident at KE Basin. The worker is a [redacted]

operator who runs a terminal in the KE Basin Radiation Zone. During slack 

periods [redacted] had manipulated the hooks and tongs used to access fuel drums in the basin. The tongs were contaminated. [Redacted] face and 

hairline were contaminated." 

[redacted]/1972 Yes 

"Employee punched end of scissors through hood glove, through surgical gloves and into the heel of his left hand at the base of his thumb. Very little bleeding. No airborne contamination thought to be probable." 

Claimant had a docum

[redacted]/72 that was 

database. Claimant al

incident on [redacted]

the database. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

September 30, 2011 

Revision No.  1 – Draft 

Document No. 

Draft – Hanford SEC Issues Review – Vol. I 

Page No.  54 of 7

Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1985 Yes 

"Workers were moving contaminated waste and sorting it at 1C5N, 14-foot left front area. Respiratory protection was not being worn. Nasal 

contamination was detected on [redacted], skin contamination on [redacted]

workers. A third worker was checked because he had been in the general area while the work was occurring." 

Claimant had docume

[redacted]/79 and [red

included in the databa

[redacted]/1974 Yes 

"A [redacted] operator was found contaminated during a personnel survey in 

the 271-CR Change Room at completion of taking East Area Tank Farm catch tank readings and checking out FIC units." 

Claimant had a docum

[redacted]/85 that was 

database. 

[redacted]/1978 Yes 

"Technician was polishing 244Cm doped waste fixation metallography 

sample in the glovebox in [redacted]. He had worked about 20 minutes on 

about six different samples of the same type. Upon removing hands from gloves, gross contamination on surgeon gloves (~1 M DPM) was detected. The surgeons gloves were removed, and the bare hands were removed. There was gross evidence of skin contamination and RM was notified. The glove break was caused by the metallography sample perforating the right glove." 

Claimant had a docum

[redacted]/1978 that wdatabase. NOCTS lis

during the SEC period

[redacted]/1986 Yes 

"While working in [redacted], 100 Area, the worker incurred facial 

contamination of 500 cpm beta-gamma activity. A nose blow indicated 900 cpm activity; however, nasal smears did not indicate detectable activity (<200 cpm). Nuclides of concern were mixed fission and activation products, primarily Co60." 

Date should actually b

to the hardcopy record

[redacted]/1983 Yes 

"A Certified [redacted] Operator and a [redacted] Technician were dispatched to flush the V-9-13 valve. The [redacted] operator installed the 

drain connecter between the V-9 valve and the drain header. As the 

[redacted] performed the required pre-flush survey, the [redacted] operator opened the V-9 flush valve and the connector Ball Valve as the [redacted]

left the immediate work scene. A back spray of water from the drain contaminated both employees." 

Claimant had docume

[redacted]/73 and [red

included in the databa

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

September 30, 2011 

Revision No.  1 – Draft 

Document No. 

Draft – Hanford SEC Issues Review – Vol. I 

Page No.  55 of 7

Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1974 Yes 

"BNW [Battelle Northwest] Staff Members were demonstrating a 210Po 

process to two AEC Canadian visitors. The 210Po process demonstration 

was conducted within a glove box located in Room 604 of the 325 Building. The process, in part, involved removal of 210Po samples from the glove box 

intermittently during the evening. Upon exit from Room 604, and upon completion of the work, at 11 PM, the Eberline Air Monitor alarmed. The operations personnel checked the room for a possible cause of the alarm. Finding no smearable contamination, both the BNW staff members and the visitors performed a personal survey and exited the room to go home... On 3/5/74 HEDL was requested to have all fans running. Word did not get to the power operators and the fans were shut down as a normal routine. The 

210Po glove box exhaust via four HEPA filters in a series prior to exhausting 

the 324 Building's main stack. It is assumed the reduction of exhaust resulted in less negative pressure within the glove box to an unsafe level for opening the glove box air lock." 

Worker had a docume

[redacted]/75 that was 

database. Worker als

incidents on [redacted

that were included in 

the [redacted] inciden[redacted] and [redactlisted for [redacted]in 

[redacted]/1973 Yes 

"[Names Removed] were working in Glove Box [redacted]. DELETED 

finished connecting a thermocouple and removed his hands from the glove box and discovered contamination on his surgeon's gloves. [Claimant] removed his hands from the glove box and found both surgeon's gloves contaminated. Radiation Monitoring was contacted and placed the room on assault mask. The room air sample was counted and was less than assault mask levels... By all indications, one or more glove box gloves were ruptured from working with sharp objects inside the box. The apparent cause of the spread of contamination to eleven glove box gloves was failure of personnel using the gloves to always perform a survey of their hands and forearms each time the hands were withdrawn from the glovebox." 

Worker had documen

[redacted]/73, [redacte[redacted]/75, and [red

contained in the datab

worker was at [redact

period. 

[redacted]/1973 Yes "A hood glove ruptured while [redacted] maintenance craftsmen were repairing the crucible cutter in the button line, hood 16." Claimant also had an 

that was also included 

[redacted]/1973 Yes "Contamination was found in the duct area (2nd floor) of 231-Z bldg. where [Name Removed] had been working." Claimant spent some 

[redacted]. 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

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Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1984 Yes 

The individual received a sliver in his right index finger while pulling a 

thermocouple out of a hood [redacted] facility during [redacted] refurbishing 

work. No contamination was detected on the sliver or on his finger; however, parts of the thermocouple were observed to be contaminated." 

Claimant had a docum

[redacted]/72 that was 

Claimant also had a d

[redacted]/83 that was 

database. 

[redacted]/1973 No 

From CATI report (see comments): "Vacuum pump oil sprayed over the area when being maintained... [protective equipment] Double set of coveralls and mask... [duration of incident] 2 hours." 

The incident is mentiointerview with the cla

claimant was an empl

the SEC period. The 

that the incident occur

Response files only coand x-ray records. 

[redacted]/1983 Yes 

"With manipulator and boot removed from "[redacted]" hot cell, employee 

reached through the penetration port and tightened a light bulb inside the cell. Lab coat sleeve contamination." 

Claimant had a docum

[redacted]/76 that was 

database. 

[redacted]/1981 Yes "Working on RTD's [Resistance Temperature Detectors] (thermocouples) No respiratory protection required or worn... Facial contamination." [redacted]/1985 Yes 

"While preparing a standards solution in a hood, a glass vile containing Pu shattered rupturing the workers glove and contaminating her left thumb. Skin was intact but substantial decontamination was required." 

Claimant was also inv

[redacted]/85 that was 

[redacted]/1973 Yes 

"Employees became contaminated while bagging Pu waste out of a glove 

box. The tape did not seal and the bag dropped to the floor. Radeco air monitor about 20 ft. away from work area had no increase in activity." 

Claimant had docume

[redacted]/72 and [red

included in the databa

[redacted]/1972 Yes 

"Employee received a puncture to his left ring finger from a syringe while cleaning a sludge pit. Puncture just barely broke the skin and no bleeding occurred. There was a visible mark on his finger." 

Hardcopy records ind

occurred in Area [reda

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

Health for factual accuracy or applicability within the requirements of 42 CFR 82.
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Document No. 

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Page No.  57 of 7

Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1990 Yes 

"[Redacted] workers entered the canyon for photography and observation of 

L cell. All wore respiratory protection. A portable air sampler was used to monitor airborne activity. The air sample indicated an air concentration of alpha activity (interpreted as plutonium) about 120% of that allowed by the protection factor for the respirators." 

Hardcopy records ind

occurred on [redacted]documented incident 

not included in the da

[redacted]/1974 Yes 

"The employees changed out a filter in a small exhaust line that serves the inhalation toxicology dog inhalation facility. No masks were worn. The filter and line are located in the 331 bldg. duct level. Upon removal from its holder the filter fell apart into small pieces. These pieces were picked up with a vacuum cleaner. Either a piece of the filter or a smear from the upstream side of the filter holder is required to obtain isotopic and particle size information." 

[redacted]/1974 Yes 

"A technician who had been removing bolts from a flange inside a glove 

box in [redacted], discovered contamination on his coveralls when 

performing a personnel survey. He immediately evacuated the room and contacted operations and Radiation Monitoring." 

Records indicate claim

during the SEC period

[redacted]/1972 Yes "A Senior Chemical Technologist received a small injury to his right forefinger while bagging waste out of hood [redacted] of the [redacted]." [redacted]/1974 Yes "Employee became contaminated while removing V-12 batch covers in the outlet pipe spaces. No mask worn." Worker also had docu

[redacted]/74, [redacte

that were also include

[redacted]/1984 Yes 

"The worker was repairing a primary coolant return valve in a steam 

generator cell [redacted] Area. Respiratory protection was not worn. Upon 

exiting, skin contamination was detected." 

[redacted]/1980 Yes "Employee was working in a greenhouse wearing a power air purifying air respirator, and the filter was knocked off." 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

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Page No.  58 of 7

Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1978 Yes 

"Beta-gamma contamination of 4,000 c/m was detected on the exterior of a 

welder craftsman's nose during personnel surveys following exit from the 221-B Canyon. Investigation revealed the respirator worn by the employee was a Scott facepiece modified to provide deep-tinted lenses for welding activities as well as a clear lenses which could be exposed for normal viewing when the tinted lens was raised. The clear lens was inadequately secured/sealed causing air leakage around the lens to the employee's face rather than all the inlet air being filtered through the appropriate canister." 

[redacted]/1974 Yes 

"[Claimant] was breaking up plutonium castings inside glove box 

[redacted]. He inadvertently cut through the glove box glove and his 

surgical glove breaking the skin of the ring finger of his left hand." 

Claimant had docume

[redacted]/75 and [red

included in the databa

claimant was at [redac

period. 

[redacted]/1982 Yes 

"Switched work from one non-contaminated valve to a contaminated valve. 

One person bumped hose on valve. Moving old valve to replace new. In with camera crew. Bumped hose on dirty pipe." 

[redacted]/1973 Yes 

"The [redacted] technicians routinely bagged in a plutonium part in Room [redacted] conveyer system. One of the [redacted] technicians discovered 

contamination on his right coverall sleeve when performing a personnel check. He checked inside the can where the part was stored and discovered it contaminated. Radiation Monitoring was contacted and placed the room on assault masks." 

Claimant had a docum

[redacted]/74 that was Records indicate claim

during the SEC period

[redacted]/1988 Yes 

"They were trying to unplug a line in SD valve pit using hot water. The operators sprayed cold water in to the pit on top of the hot water, this caused vapor to rise and some of it got into the cab of the crane contaminating [claimant's] coveralls. He was not wearing a mask at the time which is 

normal for this type of operation and is covered in RWP F1. He received a small amount of contamination in his nostrils and mouth." 

[redacted]/1982 Yes "On [redacted]/82 at 0810 hours, a uranium fire occurred in the Peel Test Saw Area." 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

Health for factual accuracy or applicability within the requirements of 42 CFR 82.
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Page No.  59 of 7

Table 13. Summary of Incidents Reviewed 

Incident 

Included in Incident Date Description of Incident From Hardcopy Record/CATI Addition

Hardcopy 

File? 

[redacted]/1980 Yes 

"A [redacted] technician, working in [redacted] making mounts of samples, 

delivered the mounts and the analytical cards via the sample elevator to the counting room without first making radiological status surveys. The counting room attendant surveyed these finding contamination and notified Radiation Monitoring." 

[redacted]/1973 Yes "An air sample that was removed from laboratory [redacted] Building on [redacted]/73 was found to be positive for plutonium contamination." Claimant has a docum

[redacted]/74 that was 

database. 

[redacted]/1973 Yes 

"[Workers] were segregating and loading out polystyrene cubes into a shipping container for transporting to ARHCO for PuO2 recovery. The cubes had been surveyed by Radiation Monitoring. [Claimant] attempted to free two cubes that were stuck together but was unsuccessful and inserted them in a plastic bag. He surveyed his gloves immediately after bagging the cubes and found them contaminated... The technician checked the table and lead shielding they were using and found both contaminated. The table top was covered with a towel, and the contamination on the lead shielding was taped over." 

Records indicate claim

during the SEC period

[redacted]/1983 Yes "Worker became contaminated while inspecting a valve at [redacted]." [redacted]/1989 Yes "[Claimant] incurred slight facial contamination and had detectable activity in nasal smears as a result of moving contaminated scaffolding in a pit." [redacted]/1974 Yes 

"The subject employee's foot slipped while he was working from a ladder 

inside the primary side of [redacted] steam generator. When he slipped his 

right wrist was scratched on several places." 

Claimant had a docum

[redacted]/77 that was 

database. 

[redacted]/1989 Yes 

"On [redacted]-89, at approximately 1050 hrs. a millwright knelt down on a contaminated damp rag to reinstall the impeller on the [redacted]. Upon exiting the [redacted] storage room, contamination was discovered on his 

right knee." 

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

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

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

September 30, 2011 

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Document No. 

Draft – Hanford SEC Issues Review – Vol. I 

Page No.  60 of 7

: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

Ho is report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Work

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

Table 13. Summary of Incidents Reviewed 

NOTICE

wever, th

Incident Date 

Incident 

Included in 

Hardcopy 

File? 

Description of Incident From Hardcopy Record/CATI Addition

[redacted]/1974 Yes 

"Employee became contaminated while working in the [redacted] of 327 

bldg. Contamination was noticed at the beginning of work by a shoe cover survey." 

Claimant had a docum

[redacted]/79 that was 

database. 

[redacted]/1973 Yes "A [redacted] operator incurred a contaminated injury to his thumb, left hand, while working in hood gloves at the sorting hood in [redacted]." [redacted]/1985 Yes 

"Worker was contaminated while working on a check valve in the [redacted]Building, [redacted]. Beta-gamma contamination was found in nose, 

moustache, and cheek area." 

[redacted]/1981 Yes "Steam generator inspection performed by employee, primary coolant sprayed." Claimant had a docum

[redacted]/77 that was 

database. 

[redacted]/1972 No CATI report mentions an incident but it was in Building [redacted] not 232- Z, the date of the incident is not known [redacted]/1974 Yes 

"Employee became contaminated during the changing of a line. A liter of distilled water was drained from a line connected to the tank. The contaminated water (thought to be nonradioactive) was then dumped into a crib sink located in Rm 601. Contamination was found on personal survey."

Claimant had docume

[redacted]/73, [redacte[redacted]/82, and [red

included in the databa

[redacted]/1973 Yes "Employee incurred a small cut to the base (palm) of his right hand. He was attempting to rearrange a package of waste in a burial drum." Claimant had docume

[redacted]/76 and [red

included in the databa

[redacted]/1978 Yes 

"A pipefitter was removing lead wool packing from a pipe sleeve. He pried it out from sleeve. He became contaminated. Employee was wearing 2 pairs of coveralls, assault mask, and 2 pair of gloves." 

[redacted]/1979 Yes 

"The subject employee was installing piping under the discharge elevator... with no respiratory protection... Personnel survey revealed [contamination]" 

Employee also had a d

[redacted]/79 that was 

database. 
Effective Date: 

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Page No. 

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

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

17.2 Findings Relating to Missing Incident Records 

Finding 22-1: The individual records, consisting of a combination of the records supplied by 

DOE and the CATI reports, appear to document the vast majority of incidents. There does not 

appear to be a detectable pattern of not recording incidents. 

Finding 22-2: The REX databases do not appear to contain the detail of information or the 

quantity of information available in the claimant records. 

Note: SC&A has not attempted to review logbooks or to otherwise determine if there were 

incidents records in other records at Hanford that are not in the personnel files. 

18.0 ISSUE 25: OTHER RADIONUCLIDES 

A number of other radionuclides were handled, produced, or present at Hanford to which some 

employees had exposure potential. These include Cr-51 and Co-60, as well as fission products 

Ru-106 and Ce-144 that were separated. In addition, some personnel in the 400 Area had 

exposure potential to sodium-24, as noted in the TBD. Monitoring data for a number of 

radionuclides other than those discussed in detail above exist in the REX database; some, like 

sodium-24, more extensively than others, like Ce-144 or Ru-106. However, for radionuclides 

like Ce-144 and Ru-106, it should be possible to use MFP data. The in-vivo counter had the 

capability of detecting these radionuclides and others. 

Sodium-24 measurements are available for the 1970s as well as the 1980s. However, some 

radionuclides, such as cobalt-60, only have in-vivo data from about 1984 onward. SC&A notes 

that more complete radionuclide monitoring seems to have been initiated in 1983 or 1984 to 

cover the more unusual radionuclides present at Hanford. 

18.1 Quality of Data 

As noted in Section 10.3, a 1982 audit of the quality of urinalysis indicates a failure to meet 

MDA levels for several radionuclides: 140Ba-La, 22Na, 24Na, 137Cs, 59Fe, 106Ru,134Cs, 239Np, 

95Nb-Zr, 131I (Fleishman 1982). Further, SC&A did not find any other audit report (Bihl 1986, 

Spitz 1984, Hickman 1982) that contained a discussion of the quality of monitoring for these 

radionuclides. Finally, SC&A has not located similar audit reports for the 1970s. 

18.2 NIOSH’s Approach to Other Radionuclides 

NIOSH has proposed the following generic approach for dose reconstruction for the large 

number of miscellaneous radionuclides present at Hanford: 

A large number of different radionuclides were present at Hanford at various 

times, but the available bioassay data for radionuclides in addition to those in this 

attachment were considered to be too few to be statistically reliable for intake 

estimation. Workers who were exposed to 137Cs, 90Sr, 24Na, and 65Zn could also 

have been exposed to other fission and activation products. From 1960 to 1988, 

intakes of most fission or activation products would have been detectable in 

whole-body counts, but the recording practice for fission and activation products 

other than 137Cs, 24Na, and 65Zn was not amenable to statistical analysis. 
Effective Date: 

September 30, 2011 

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Page No. 

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

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

Guidance for assigning intakes of additional radionuclides is provided in 

ORAUT-OTIB-0054 (ORAUT 2007a) and for specific buildings in Section 5.6. 

 

Assignment to unmonitored workers of radionuclides that were not analyzed in 

this study should be in accordance with guidance in Section 5.6. [ORAUT 2010e, 

p. 141]. 

As noted in Section 10.4, this approach is suitable for reactor workers. NIOSH has also 

specified dose reconstruction approaches in Attachment C of ORAUT 2010e for sodium-24, 

zinc-65, and the specific fission products Sr-90, Cs-137 and Pm-147. However, as also noted in 

Section 10.4, NIOSH has not specified a coworker model for unmonitored 200 Area separations 

area workers or workers who handled wastes, including those in the 300 Area, including 

Buildings 324, 325, 327, and 3706. 

18.3 Findings for Other Radionuclides 

Finding 25-1: The MDA for several of the radionuclides in question in this section were not 

met in 1981. There are no other audit reports for the 1972–1986 period that address these 

radionuclides. (SC&A is examining some quality of data issues for the 1987 to 1989 period 

separately, as noted in the Overview section.) 

Finding 25-2: Some of the unusual radionuclides, such as Co-60, only have a significant 

quantity of in-vivo data from about 1983 or 1984. 

Finding 25-3: NIOSH should discuss the pre-1987 quality issues in the TBD relating to several 

fission products, as well as sodium-24, and specify how it will handle MDA and other problems 

that may have existed in years other than 1981. (SC&A is in agreement with NIOSH that 

sodium-24 is only relevant to the 400 Area for the period under consideration here (ORAUT 

2010e, p. 53), and only during the period of operation of the FFTF. 

Finding 25-4: While NIOSH has specified coworker models for a number of radionuclides and 

areas, it has not specified a coworker model for mixed fission and activation products in the 200 

Area and in the waste handling and processing buildings in the 300 Area. 

19.0 MATRIX ISSUE 26: DATA COMPLETENESS 

SC&A did an extensive analysis of the completeness of Hanford internal dose data. The main 

parts of this analysis are in Appendix A. This appendix contains a compilation of the internal 

dose data by radionuclides for each year in the period under review—January 1, 1972, to 

December 31 1990. Hence, it should be remembered that the label “1972” only refers to the last 

half of that year. SC&A also sorted the data by area and by job type. Additionally, SC&A 

defined the areas according to whether there was exposure potential or not. This term was 

loosely defined in this context as the area where the radionuclide was handled or processed 

during that year in a form that workers might inhale relative to other areas where such a potential 

was relatively unlikely. 

The percent of workers who were monitored for that category, as well as the median values of 

the monitoring results, are shown by radionuclide. In addition, the above discussion also 

contains the total number of workers who were monitored in each year by Area. 
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The discussion above by matrix issue also contains conclusions regarding whether the data are 

adequate for dose reconstruction. The findings from Appendix A are reproduced here for 

convenience. 

19.1 Findings from Appendix A 

Finding A-1: For the main radionuclides analyzed (Am, Cs, MFP, Pu, Th, and U), workers 

associated with the 200 Tank Farms were the most likely to be monitored during their 

employment.8

Finding A-2: ‘Radiation monitors,’ ‘electricians,’ ‘operators,’ ‘pipefitters,’ and ‘science 

technicians’ were consistently among the five job titles most likely to be monitored during their 

SEC employment. 

Finding A-3: In general, the most commonly monitored job titles by area and year are as 

follows: ‘managers and administrators’ (100, 100-N, 200, and 300 Areas), ‘operators’ (200, 200 

Tank Farms, and 300 Areas), and ‘scientists’ (300 Area). 

Finding A-4: Records analysis for americium, cesium, mixed fission products, iodine, and 

uranium monitoring showed a significant decrease in worker sampling in 1975 (generally less 

than 1% of the worker population was monitored). Other significant decreases in worker 

monitoring include 1974 (iodine), 1976–1977 (mixed fission products), and 1985 (cesium). 

Thorium-232 was sparsely monitored throughout the period, and there are very few data points 

overall. No significant decreases in worker monitoring were identified for plutonium. 

Finding A-5: Section 7 presents radionuclides not included in the main analysis because of the 

sparse available records; these include polonium, neptunium, radium, curium, 

californium/berkelium, and ‘total actinides.’ Polonium, curium, and ‘total actinides’ were 

mostly periodic sampling, while radium and neptunium were likely incident related.9

Observation A-1: Based on a review of the radionuclide-specific records in the SRDB, SC&A 

found that contamination incidents generally tended to be followed up by urinalysis and/or invivo counting of the individuals involved even if the initial nasal smears indicated no intake 

potential. However, it should be borne in mind that SC&A did not attempt to correlate incident 

records with individual personnel files. See also the discussion of Matrix Items 21 and 22 

regarding missing or destroyed records in the main section of this report. 

20.0 MATRIX ISSUE 27: BUILDING 324 LEAKS 

This issue was added to the matrix in 2011 as a result of a comment made to the Advisory Board 

by an individual with knowledge of ongoing Hanford decommissioning operations. The concern 

was initiated by the discovery in November 2010 of very high radiation rates in the soil under 

Building 324 during decommissioning activities. The Board asked SC&A to review the issue for 

its relevance to the SEC period, July 1, 1972, to December 31, 1990, in case there were incidents 

8

 The only exception to this is for iodine, because no workers identified with the Tank Farms had iodine 

monitoring associated with them. 9

 It is assumed that samples designated as ‘special’ via their respective ‘reason codes’ were incident related. 
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during that period that led to the contamination that would affect the feasibility of dose 

reconstruction with sufficient accuracy. 

SC&A conducted a number of interviews with workers who had personal knowledge of the 

operations in Building 324 and the related operations in Building 325. A summary of these 

interviews is presented in Appendix D. These interviews complement the review that SC&A has 

done of the completeness and adequacy of monitoring records as described in this report and in 

Appendix A. They also complement interviews done earlier in the SEC process, which are 

provided in Appendix C. 

The essentials of the incident are reviewed here for context, while more detail is available in 

Appendix D. The B-Cell in Building 324 was the location of a number of experimental and pilot 

programs in high-level waste management, notably the vitrification of high-level wastes. The 

wastes were sometimes artificially created by mixing specific fission products. In other cases, 

high-level wastes from Hanford were used and in yet other cases, high-level wastes from other 

locations were used. The high radiation levels found in 2010 appear to have originated in leaks 

of high-level waste into the soil in the mid-1980s. Onsite high-level waste, reportedly containing 

large amounts of Cs-137 from an on-site Hanford source, was being processed into glass logs as 

part of a feasibility demonstration for the Federal Republic of Germany (FRG). The material 

apparently reached the soil via a breach in the stainless steel liner of B-Cell. The breach was 

discovered during B-Cell remediation activities. 

For one project in the 1970s, spent fuel from commercial reactors was dissolved in Building 324 

B-Cell. The material was sent via underground pipes to Building 325 for recovery of uranium 

and plutonium; the high-level waste from the separations was returned via underground pipe to 

Building 324 B-Cell for vitrification. Interviewees are not aware of any leaks associated with the 

inter-building transfer line. Geoprobe measurements have not been performed. 

Spills and gross contamination within B-Cell were anticipated, due to the nature of the work—

wet chemistry processes with extremely high activity materials. Workers could not enter the hot 

cell, due to the dose rates; work was done remotely (e.g., with cranes and manipulators) or 

outside the cell. During the 1980s, B-Cell was used to fabricate isotopic radiation and heat 

sources for the Federal Republic of Germany (FRG). “Within the borosilicate glass matrix 

radiochemical constituents (137Cs and 90Sr) were immobilized to yield a product with a 

predetermined decay heat and surface radiation exposure rate” (Holton et al. 1989). A major 

spill of materials associated with this project reportedly occurred within B-Cell in 1986. This 

incident, which would have been documented as a contained spill (not a release) at the time, is 

considered to be the primary source of the high dose rates that have been detected in the soil 

under B-Cell. The spill was described in the interviews as follows: 

Most of the contamination in B-Cell came from the FRG program. The biggest 

spill occurred in 1986, when 500 L of liquid cesium nitrate was spilled in B-Cell. 

This spill is considered the most likely source of the soil contamination. Liquid 

cesium nitrate was being transferred from one vessel to another. It was a vacuum 

transfer, so there was no way to measure the liquid volumes during the transfer. 

After the transfer, workers observed that the volume was lower than it should 

have been. They found out there was a crack in a weld on a jumper. About 
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200,000 Ci of cesium had spilled on the floor of the hot cell. This was very well 

characterized material. [Appendix D, Vol. II, of this report.] 

The hot cell floor, lined with stainless steel, sloped towards a sump where spilled liquids could 

be collected and transferred to waste tanks in the high level vault. During decommissioning 

activities, the contractor identified a hole in the stainless steel liner at the floor of the sump 

(WCH 2010). This breach in the liner is believed to be the route by which contamination from 

inside the hot cell reached the soil. Recognizing the potential for soil contamination after the 

breach was identified, the contractor used underground radiation detectors (Geoprobes) to survey 

dose rates along various trajectories in the soil below the B-Cell. Both the B-Cell and Building 

324 are intact. 

Decontamination of B-Cell began in the late 1980s. Early efforts involved size-reducing 

contaminated equipment to fit in waste casks. Grout was used to stabilize materials within the 

casks; some of the grout spilled on the floor and became contaminated. It was during removal of 

contaminated grout from the floor and sump that the breach in the stainless steel liner was 

detected (WCH 2010). The building mission moved decisively into decontamination and 

decommissioning when PNNL transferred Building 324 to Babcock and Wilcox in the late 

1990s. Residual materials in B-Cell and other areas of Building 324 were designated as waste, 

but the nature and form of the material raised regulatory concerns. DOE was found to have 

RCRA (Resource Conservation and Recovery Act) mixed waste in a non-permitted storage area. 

Of particular concern were residual materials from the FRG program, including the 1986 spill. 

Some of the spilled material had been absorbed into a layer of dust that had accumulated on the 

floor of B-Cell due to years of operation under negative pressure for radiological safety reasons. 

It was conservatively assumed that the spilled material had been absorbed into the dust layer 

(rather than draining to the sump) and had dried to a dispersible form (Hobart 1998). A specific 

Closure Plan was created and executed to remove the dispersible mixed waste from the 

Radiochemical Engineering Cells and waste vaults (DOE 1998). 

While conducting a historical review of hot cell logbooks from Building 324, DOE discovered 

evidence of a prior soil contamination event (Langstaff 2001). 

The discovery was an apparent leak from one of the lined hot cells or from piping 

embedded in the cell walls. When facility personnel were modifying the facility to 

accommodate a new process in B-Cell, they entered the crawl space below the 

airlock and between the High Level Vault and B-Cell. They needed to core the 

concrete and install stainless steel piping to the pipe trench where connections 

were then made to the process equipment in B-Cell. What they saw, when they 

entered the crawl space and looked into the open spaces on either side under 

A-Cell and C-Cell, were cracks in the A-Cell and C-Cell concrete floors and 

evidence of radioactive liquid leaking into the crawl space from above, through 

the cracks. They reported that extensive testing was done in the suspect hot cells 

to try to locate the source of the leak, but the testing was unsuccessful. 

Contaminated dirt was dug out and safely disposed. Catch trays were installed to 

contain the effluent from further leaks, should they occur. 
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The contamination in the soil underneath B-Cell appears not to have created exposure potential 

for workers until the decommissioning operations after the SEC period under review here. With 

the building and hot cell intact, workers would not have had physical access to the external dose 

rates, and the material reportedly has not had a pathway to reach the air or water. Potential 

personnel exposures during decontamination, waste classification, and waste 

stabilization/removal efforts (from the late-1980s to the present) have not been assessed. Details 

of the earlier contamination event from the historical logbook review were not available to 

SC&A. This incident may justify further investigation. The timeframe of this discovery was 

described as “before the Point Beach fuel reprocessing,” which would place it sometime in the 

mid- to late-1970s. The exposure potential and monitoring records of workers who discovered 

and managed the leaks under A-Cell and C-Cell also need to be investigated. 

The interviewees reported a good safety culture in the B-Cell operations and good health physics 

practices. Workers generally wore their badges, and there was internal monitoring as well. 

SC&A reviewed the Cs-137, Sr-90, and mixed fission product in-vivo and in-vitro monitoring 

records in light of these incident descriptions (see Section 10). There appear to be ample internal 

monitoring data for these items in the 300 Area in the early to mid-1980s during the period when 

the spills are thought to have occurred, and also during the late-1980s when hot cell 

decontamination, waste characterization, and stabilization operations began. However, SC&A 

has not done a more fine-grained investigation of monitoring records of employees who were 

working in Building 324 at the time of the spill and decontamination. Similarly, monitoring data 

for personnel involved in the historical logbook incident have not been evaluated. 

Two other things should be noted in this context. First, B-Cell operations appear to have 

involved ad hoc mixtures of fission products. Since monitoring records for individual fission 

products other than Cs-137 and Sr-90 are far less complete for the period in question (there are 

few Pm-147 bioassay records for the mid-to-late 1980s for instance), it may be important to 

understand the specific operations involved in B-Cell in this period, to find some way of 

enveloping the intake estimates based on available Sr-90, Cs-137, and mixed fission product 

data. Second, it is important to remember that the issue of the quality of data in relation to the 

U.S. Testing measurements from 1987 to 1989 is being separately investigated and is not 

covered here (see Section 1.0). 

20.1 Findings for Issue 27 

Finding 27-1: There were at least two incidents in which radioactive materials migrated from 

presumably contained hot cells and/or pipes to reach the soil. The material recently discovered 

under B-Cell is unlikely to have created internal exposure potential during the leak; however, the 

stabilization operations that began in the late 1980s may have created internal exposure potential. 

Furthermore, we note that the source of the leaks found under A and C cells is not well 

understood. 

Finding 27-2: In general, there appear to be ample internal monitoring data for 300 Area 

workers for Cs-137 and mixed fission products and overall for Sr-90 during the 1980s, which is 

the period in question for the B-Cell incidents. However, it is important for NIOSH to verify 

whether the specific employees who worked in Building 324 at the time of the leaks, including 

the A and C cell leaks and their management, and during B-Cell the stabilization operations have 

adequate internal monitoring data for dose reconstruction. A detailed review of the logbooks and 
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NOTICE:

incident records relating to B-Cell operations in the mid- to late-1980s and of the operations in 

A-Cell and C-Cell may also be warranted, given the scale of the contamination. For quality of 

data comments prior to 1987, see Section 10. 

Finding 27-3: NIOSH should investigate whether there were ad hoc mixtures of fission 

products, for which there are scant data for the period, used in B-Cell feasibility studies during 

the mid-to-late 1980s. If such operations were done, NIOSH should specify a dose 

reconstruction approach for those operations. 
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Health for factual accuracy or applicability within the requirements of 42 CFR 82.

NOTICE:

Westinghouse 1994. Plutonium and Tritium Produced in the Hanford Site Production Reactors. 

WHC-SD-CP-RPT-014, Westinghouse Hanford Company. SRDB 36571.