Upper-Bound Radiation Dose Assessment for

Military Personnel at McMurdo Station, 

Antarctica, between 1962 and 1979

Revision 2 

DISTRIBUTION A. Approved for public release: distribution is unlimited.

September 2017

Prepared by:

McMurdo Station Radiation Dose Assessment

Integrated Project Team

For:

Assistant Secretary of Defense for Health Affairs

and Deputy Assistant Secretary of the Navy for Safety

DTRA-TR-12-003 (R2)

TECHNICAL REPORT

Defense Threat Reduction Agency

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4. TITLE AND SUBTITLE

Upper-Bound Radiation Dose Assessment for Military Personnel at 

McMurdo Station, Antarctica, between 1962 and 1979, (Revision 2)

5a. CONTRACT NUMBER

5b. GRANT NUMBER

5c. PROGRAM ELEMENT NUMBER

6. AUTHOR(S)

Dunavant, Jason D.1, Chehata, Mondher1, Morris, William J.2, Fairchild, 

Gregory J.3, Blake, Paul K.4.

5d. PROJECT NUMBER

5e. TASK NUMBER

5f. WORK UNIT NUMBER

7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES)

 1 Leidos, 11951 Freedom Drive, Reston, VA 20190

 2 McMurdo Station Crew Member (Veteran)

 3 Naval Dosimetry Center, Bethesda, MD

 4 Defense Threat Reduction Agency, Fort Belvoir VA

8. PERFORMING ORGANIZATION

REPORT NUMBER

9. SPONSORING/MONITORING AGENCY NAME(S) AND ADDRESS(ES)

Nuclear Technologies Department, Attn: Dr. Paul Blake 

Defense Threat Reduction Agency 

8725 John J. Kingman Road, Mail Stop 6201 

Fort Belvoir, VA 22060-6201

10. SPONSOR/MONITOR'S ACRONYM(S)

DTRA RD-NTS

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AGENCY REPORT NUMBER

DTRA-TR-12-003(R2)

12. DISTRIBUTION AVAILABILITY STATEMENT

DISTRIBUTION A. Approved for public release: distribution is unlimited.

13. SUPPLEMENTARY NOTES

14. ABSTRACT

U.S. veterans who were assigned at McMurdo Station, Antarctica, from 1962 to1979, have expressed concern that their 

health may have been affected by radiological releases from an onsite nuclear power plant. This report presents the 

results of the Department of Defense's radiation dose assessment for McMurdo Station veterans who were not 

monitored for occupational radiation exposure. This assessment determined that the radiation doses were low and the 

associated probability that disease could have arisen from these doses is also low. Finally, recommendations are 

provided for McMurdo Station veterans, their dependents, the Department of Veterans Affairs (VA), and the Naval 

Dosimetry Center regarding the VA radiogenic disease claims process.

15. SUBJECT TERMS

McMurdo Station, Veterans, Radiation Dose Assessment, Nuclear Test Personnel Review Program

16. SECURITY CLASSIFICATION OF: 17. LIMITATION OF 

ABSTRACT

 U

18. NUMBER

OF PAGES

129 Pages

19a. NAME OF RESPONSIBLE PERSON 

Paul K. Blake, PhD

a. REPORT

 U

b. ABSTRACT

 U

c. THIS PAGE

 U

19b. TELEPONE NUMBER (Include area code ) 

 (703) 767-3384
UNIT CONVERSION TABLE

U.S. customary units to and from international units of measurement*

U.S. Customary Units Multiply by International Units

Divide by†

Length/Area/Volume

inch (in) 2.54 × 10–2 meter (m)

foot (ft) 3.048 × 10–1 meter (m)

yard (yd) 9.144 × 10–1 meter (m)

mile (mi, international) 1.609 344 × 103 meter (m)

mile (nmi, nautical, U.S.) 1.852 × 103 meter (m)

barn (b) 1 × 10–28 square meter (m2)

gallon (gal, U.S. liquid) 3.785 412 × 10–3 cubic meter (m3) 

cubic foot (ft3) 2.831 685 × 10–2 cubic meter (m3) 

Mass/Density

pound (lb) 4.535 924 × 10–1 kilogram (kg)

atomic mass unit (AMU) 1.660 539 × 10–27 kilogram (kg)

pound-mass per cubic foot (lb ft–3) 1.601 846 × 101 kilogram per cubic meter (kg m–3)

Pound-force (lbf avoirdupois) 4.448 222 Newton (N)

Energy/Work/Power

electronvolt (eV) 1.602 177 × 10–19 joule (J)

erg 1 × 10–7 joule (J)

kiloton (kT) (TNT equivalent) 4.184 × 1012 joule (J)

British thermal unit (Btu) (thermochemical) 1.054 350 × 103 joule (J)

foot-pound-force (ft lbf) 1.355 818 joule (J)

calorie (cal) (thermochemical) 4.184 joule (J)

Pressure

atmosphere (atm) 1.013 250 × 105 pascal (Pa)

pound force per square inch (psi) 6.984 757 × 103 pascal (Pa)

Temperature

degree Fahrenheit (

o

F) [T(oF) − 32]/1.8 degree Celsius (oC)

degree Fahrenheit (

o

F) [T(oF) + 459.67]/1.8 kelvin (K)

Radiation

activity of radionuclides [curie (Ci)] 3.7 × 1010 per second (s

–1‡

)

air exposure [roentgen (R)] 2.579 760 × 10–4 coulomb per kilogram (C kg–1

)

absorbed dose (rad) 1 × 10–2 joule per kilogram (J kg–1§)

equivalent and effective dose (rem) 1 × 10–2 joule per kilogram (J kg–1**) *

Specific details regarding the implementation of SI units may be viewed at http://www.bipm.org/en/si/. †

Multiply the U.S. customary unit by the factor to get the international unit. Divide the international unit by the factor to get the U.S. 

customary unit.

‡

The special name for the SI unit of the activity of a radionuclide is the becquerel (Bq). (1 Bq = 1 s–1). §

The special name for the SI unit of absorbed dose is the gray (Gy). (1 Gy = 1 J kg–1). **The special name for the SI unit of equivalent and effective dose is the sievert (Sv). (1 Sv = 1 J kg–1).
i

Revision 1 Notes

Revision 1 corrects minor errors in the original technical report, DTRA-TR-12-003, dated 

June 1, 2013. These include corrections of some dosimetry statistics in Table 5, and corrections 

to some of the dose values expressed in SI units in Table 30. In addition, selected text in Tables 

11 and 12 was revised to be consistent with the purpose of those tables, and the text in Section 

5.1.1 was rearranged so that table citations are in the correct sequence. Finally, minor editorial 

revisions were made in various discussions. None of these revisions affect the doses used by the 

Naval Dosimetry Center in its performance of dose assessments for claims filed with the 

Department of Veterans Affairs by McMurdo Station Support personnel.

Revision 2 Notes

Revision 2 corrects a typo in Equation 2 in Appendix B-1.1 in the original technical 

report, DTRA-TR-12-003, dated June 1, 2013. The parameter HD is deleted from the 

denominator, which was a typo. All calculations using this equation were checked and are 

correctly implemented.
ii

Acknowledgments

This work has been conducted through the cooperation of McMurdo Station veterans, the U.S. 

Navy, the U.S. Army, the Department of Veterans Affairs (VA), the Veterans’ Advisory Board 

on Dose Reconstruction (VBDR), and the Defense Threat Reduction Agency (DTRA)’s Nuclear 

Test Personnel Review (NTPR) Integrated Project Team (IPT). 

McMurdo Station Veterans

Charles Fegley

James M. Kelly

Arthur C. Tate

Bill Westerfield

U.S. Navy

CDR Gregory R. Kahles, USN, Radiological Affairs Support Office

Gina Nichols, Archivist, U.S. Navy Seabee Museum

CDR James Cassata, USN, Uniformed Services University of Health Sciences

U.S. Army

Brian Hearty, U.S. Army Corps of Engineers

Phil Shubert, Army Reactor Office

VA

Bradley Flohr, Jacqueline Imboden, and Jerry Steele; Veterans Benefits Administration

Paul Ciminera and Daniel McClung; Veterans Health Administration

VBDR

Harold L. Beck

Paul G. Voillequé

Gary H. Zeman

DTRA NTPR IPT

LCDR Gerald Burke, USN, DTRA

Brian Malik, L-3 Stratis

Michael McKenzie-Carter, SAIC

Stephen D. Egbert, SAIC
iii

Executive Summary

U.S. veterans who were assigned at McMurdo Station, Antarctica, from 1962 to 1979, 

have expressed concern that their health may have been affected by radiological releases from an 

onsite nuclear power plant (NPP). In response to this concern, the Under Secretary of Defense 

for Personnel and Readiness stated that the Department of Defense (DoD) would work with the 

Department of Veterans Affairs (VA) and the Veterans’ Advisory Board on Dose Reconstruction 

(VBDR) to assess these risks. A DoD integrated program team was formed of military, civilian, 

and contract radiation health experts from the Defense Threat Reduction Agency, the Naval 

Dosimetry Center, and McMurdo Station veteran volunteers. This team calculated veteran 

radiation doses and assessed the probability that these radiation doses may have caused 

subsequent disease. During this assessment process, VA and VBDR subject matter experts 

provided critical peer-review.

The McMurdo Station NPP was designed to generate electricity and distill drinking water 

in Antarctica thus reducing the significant challenge of transporting and storing fuel oil at this 

remote location. It was assembled at McMurdo Station in 1962 and operated through 1972. It 

was placed in standby mode in October 1972 when shield water leakage into the insulation 

around the pressure vessel and primary coolant piping created the possibility of chloride stress 

corrosion. The high cost of performing a full inspection resulted in the decision to permanently 

shut down operations and begin dismantling. The decommissioning and shipping of the NPP and 

contaminated soil back to the U.S. occurred between 1974 and 1979. 

This assessment focuses on veterans who were not externally and internally monitored 

for radiation exposure. These non-monitored veterans were defined as support personnel to 

differentiate them from veterans who worked at the NPP and received external and internal 

radiation monitoring. Dose assessment for NPP workers is performed on an individual basis 

owing to the availability of individual radiation monitoring records, and is only briefly addressed 

in this report. 

In this assessment for McMurdo Station support personnel, doses from sources outside 

the body and those inside the body are calculated. The main sources of radiation outside the body

include sources from the NPP, e.g., the reactor, radioactive waste, radioactive effluent, 

radioactive materials in passing plumes, and contaminated soil deposited on the ground. The 

sources of radiation inside the body are radioactive materials that deposit in tissues and organs 

after breathing air and ingesting water, and soil and dust that have been contaminated with 

radioactive materials. For external dose calculations, estimated radiation dose rates are used for 

various locations where veterans could have been exposed. For internal dose calculations, 

measurements of radioactivity in air, water, and soil are used with internationally accepted dose 

models recommended by the International Commission on Radiological Protection (ICRP) 

(ICRP, 1994a-b).

The quantities calculated in this report are whole body effective dose and total equivalent 

doses to the thyroid gland and red bone marrow as defined in ICRP Publication 60 (ICRP, 1991). 

The thyroid and red bone marrow selections were considered the best representatives of the 

human body organs that might be affected by the NPP radiation sources. The effective dose is a 
iv

radiation protection quantity that allows external and internal doses to be combined to obtain an 

estimate of overall possible health risk from exposure to radiation. The total equivalent dose to 

the thyroid gland is from external and internal radiation sources and is important in assessing the 

potential disease of this organ. The total equivalent dose to the red bone marrow is similarly 

from external and internal radiation sources and is important in assessing diseases of the blood 

and bone marrow. 

The calculated upper-bound total effective doses for support personnel at the McMurdo 

Station range from 0.2 to 0.6 rem (2 to 6 mSv) per tour. The calculated upper-bound total 

equivalent doses to the thyroid and red bone marrow, the two organs that could have received 

the highest organ doses, range from 0.2 to 0.6 rem (2 to 6 mSv) per tour. These doses are low 

and the associated probability that disease could have arisen from these doses is also low. For 

comparison, at effective doses less than 5 to 10 rem (50 to 100 mSv), risks of health effects are 

either too small to be observed or nonexistent (HPS, 2010).

Finally, to assist McMurdo Station veterans, their dependents, the VA, and the Naval 

Dosimetry Center, this report includes recommendations for the filing of VA compensation 

claims for diseases potentially arising from radiation exposure, and information and procedures 

for processing these claims. Specific recommendations are provided for first time filers, veterans 

previously denied VA service connection, and surviving spouses or children of a deceased 

McMurdo Station veteran.
v

DTRA-TR-12-003(R2): Upper-Bound Radiation Dose Assessment for Military Personnel at 

McMurdo Station, Antarctica, between 1962 and 1979, Revision 2

Table of Contents

Revision 1 Notes......................................................................................................................... i

Revision 2 Notes......................................................................................................................... i

Acknowledgments...................................................................................................................... ii

Executive Summary .................................................................................................................. iii

Table of Contents........................................................................................................................v

List of Figures.............................................................................................................................x

List of Tables............................................................................................................................ xi

1. Introduction.............................................................................................................................1

1.1 Background..........................................................................................................1

1.2 Objectives............................................................................................................2

1.3 Technical Approach .............................................................................................3

2. History of the McMurdo Station Nuclear Power Plant .............................................................5

2.1 Installation and Early Operation...........................................................................5

2.1.1. Installation ............................................................................................ 5

2.1.2. Early Operation..................................................................................... 5

2.2 Operations and Maintenance ..............................................................................11

2.2.1. Environmental Radiation Surveillance Program ...................................11

2.2.2. Radiation Safety Standards...................................................................12

2.2.3. Refueling and Maintenance ..................................................................12

2.2.4. Water Distillation.................................................................................13

2.2.5. Malfunction History and Characteristics...............................................13

2.3 Decommissioning ..............................................................................................15

2.3.1. Removal and Shipment of Plant Components and Contaminated Soil...15

2.3.2. Radioactive Waste................................................................................16

2.3.3. Final Surveys .......................................................................................17

3. Potential Sources of Radiation Exposure for Support Personnel............................................19

3.1 External Gamma and Neutron Radiation ............................................................19

3.1.1. Radiation Sources inside the NPP.........................................................19

3.1.2. Radiation Sources while Transiting Near the NPP................................20
vi

3.1.3. Immersion in Radioactive Noble Gases................................................20

3.1.4. Movement of Packages Containing Radioactive Materials outside the 

NPP .....................................................................................................20

3.1.5. Movement of Radioisotope Thermal Generators...................................20

3.1.6. Loading of Radioactive Waste..............................................................21

3.1.7. Spilled or Residual Contaminated Soil .................................................21

3.2 Internal Radiation...............................................................................................21

3.2.1. Inhalation of Airborne Radioactive Materials from Emissions Released 

through the NPP Stack and Worldwide Fallout.....................................22

3.2.2. Inhalation of Suspended Soil Contaminated with Radioactive Materials

.............................................................................................................22

3.2.3. Ingestion of Tritium in Drinking Water ................................................22

3.2.4. Ingestion of Radioactive Materials in Drinking Water other than Tritium

.............................................................................................................22

3.2.5. Incidental Ingestion of Soil and Dust Contaminated with Radioactive 

Materials..............................................................................................23

4. Radiation Monitoring ............................................................................................................24

4.1 Exposure Records of Reactor Personnel .............................................................24

4.2 External Radiation Measurements......................................................................25

4.3 Airborne Activity Concentrations.......................................................................27

4.4 Radioactivity Concentrations in Water...............................................................30

4.5 Radioactivity Concentrations in Soil ..................................................................32

4.6 Removable Radioactivity on Surfaces................................................................33

4.7 Radiation Levels of Radioactive Waste Shipments.............................................33

5. Dose Assessment Methodology and Assumptions for Support Personnel...............................34

5.1 External Dose Methods and Assumptions...........................................................34

5.1.1. Touring or Visiting the NPP.................................................................35

5.1.2. Transits in the Proximity of the NPP ....................................................37

5.1.3. Immersion in a Cloud of Noble Gases..................................................39

5.1.4. Proximity to Radioactive Waste Packages............................................40

5.1.5. Proximity to Radioisotope Thermal Generators....................................41

5.1.6. Contaminated Soil on the Ground.........................................................42

5.1.7. Ship Loading of Radioactive Waste......................................................43

5.2 Internal Doses–Methods and Assumptions.........................................................44
vii

5.2.1. Inhalation of Airborne Radioactive Materials.......................................44

5.2.2. Inhalation of Suspended Contaminated Soil during Decommissioning..47

5.2.3. Inhalation of Suspended Contaminated Soil Spilled on the Ground, 

Roads and Ship Loading Areas.............................................................48

5.2.4. Ingestion of Tritium in Drinking Water ................................................49

5.2.5. Ingestion of Long-Lived Beta Emitters in Drinking Water other than 

Tritium.................................................................................................50

5.2.6. Incidental Ingestion of Contaminated Soil and Dust .............................51

5.3 Uncertainties and Upper-Bound Doses...............................................................52

5.3.1. External Doses.....................................................................................52

5.3.2. Internal Doses......................................................................................52

6. Support Personnel Dose Assessment Results and Discussion.................................................54

6.1 External Dose ....................................................................................................54

6.2 Internal Dose......................................................................................................54

6.3 Discussion..........................................................................................................54

6.4 Other Exposure Scenarios..................................................................................55

6.5 Comparative Radiation Exposures in Daily Life.................................................60

6.6 Potential Health Effects......................................................................................62

7. Radiation Dose Assessment for Reactor and Decommissioning Crew Members....................63

7.1 Sources of External Radiation Exposure.............................................................63

7.1.1. Reactor Crew Members during NPP Operation ....................................63

7.1.1.1 Radiation Sources at the NPP Facility while on Duty ............63

7.1.1.2 Radiation Sources while off Duty..........................................63

7.1.2. Decommissioning Crew Members........................................................64

7.1.2.1 Radiation Sources at the NPP Facility while on Duty ............64

7.1.2.2 Radiation Sources while off Duty..........................................64

7.2 Sources of Internal Radiation Exposure..............................................................64

7.2.1. Reactor Crew Members during NPP Operation ....................................64

7.2.1.1 Inhalation of Airborne Radioactive Materials while inside the 

NPP ......................................................................................64

7.2.1.2 Inhalation of Airborne Radioactive Materials while outside the 

NPP ......................................................................................64

7.2.1.3 Ingestion of Drinking Water..................................................65

7.2.2. Decommissioning Crew Members........................................................65
viii

7.2.2.1 Inhalation of Airborne Radioactive Materials while on Duty .65

7.2.2.2 Inhalation of Airborne Radioactive Materials while off Duty.65

7.2.2.3 Ingestion of Drinking Water..................................................65

7.2.2.4 Incidental Ingestion of Contaminated Soil and Dust ..............65

7.3 Dose Estimation Methodology ...........................................................................66

7.3.1. External Dose.......................................................................................66

7.3.2. Inhalation of Airborne Radioactive Materials.......................................66

7.3.3. Ingestion of Radioactive Materials .......................................................67

8. Recommendations for Radiogenic Disease Claim Process....................................................68

8.1 Guidance for the Radiogenic Disease Claim Process..........................................68

8.1.1. For McMurdo Station Veterans who are First Time Filers....................68

8.1.2. For McMurdo Station Veterans Previously Denied VA Service 

Connection...........................................................................................69

8.1.3. For a Surviving Spouse of a Deceased McMurdo Station Veteran ........69

8.1.4. For the Veterans Benefits Administration (VBA) .................................69

8.1.5. For Director, Environmental Agents Service, Veterans Health 

Administration (VHA) .........................................................................70

8.1.6. For the Naval Dosimetry Center...........................................................70

8.2 Proposed Guidelines for Individual Radiation Dose Assessment for McMurdo 

Station Veterans.................................................................................................70

8.2.1. Assignment to Appropriate Group and Collection of Information.........70

8.2.2. Individual Dose Assessment for Support Personnel..............................70

8.2.3. Individual Dose Assessment for Reactor Crew Members......................71

9. Conclusions...........................................................................................................................73

9.1 Summary of Assessment Approach....................................................................73

9.2 Dose Assessment Results and Health Risks........................................................74

9.3 Recommendations for Radiogenic Disease Claims Process ................................74

10. References...........................................................................................................................76

Appendix A. Congressional Inquiries Regarding Veterans’ Exposures to Radiation from the 

McMurdo Station Nuclear Power Plant ..............................................................81

A-1. Letter from Senator Brown to Secretary of Defense ...........................................82

A-2. DoD Response to Senator Brown. ......................................................................83

A-3. Letter from Senator Blumenthal to Assistant Secretary of the Navy (Installations 

and Environment)...............................................................................................84
ix

A-4. Department of the Navy Response to Senator Blumenthal..................................85

Appendix B. Radiation Dose Calculation Methods and Parameter Definitions ..........................87

B-1. External Dose Assessment Methods and Parameter Definitions..........................88

B-1.1. Average Hourly Dose Rate inside the NPP........................................88

B-1.2. External Dose from Touring or Visiting the NPP ..............................88

B-1.3. External Dose from Transiting in Proximity to the NPP....................89

B-1.4. External Dose from Immersion in a Cloud of Noble Gases ...............89

B-1.5. External Dose from Movement of Radioactive Waste Packages outside 

the NPP ............................................................................................90

B-1.6. External Dose from Shipments of Radioisotope Thermal Generators 90

B-1.7. External Dose from Contaminated Soil on the Ground......................91

B-2. Inhalation Dose Assessment Methods and Parameter Definitions.......................91

B-2.1. Inhalation of Airborne Radioactive Materials....................................91

B-2.2. Inhalation of Suspended Contaminated Soil during Decommissioning

.........................................................................................................92

B-2.3. Inhalation of Contaminated Soil on the Ground, Roads, and Ship 

Loading Areas ..................................................................................92

B-3. Ingestion Dose Assessment Methods and Parameter Definitions ........................93

B-3.1. Ingestion of Tritium in Drinking Water.............................................93

B-3.2. Ingestion of Other Long-Lived Beta Emitters in Drinking Water ......93

B-3.3. Incidental Ingestion of Contaminated Soil and Dust..........................93

B-4. Dose Uncertainty Methods and Parameter Definitions........................................94

B-4.1. Uncertainty in External Doses...........................................................94

B-4.2. Uncertainty in Internal Doses............................................................94

Appendix C. McMurdo Station Radiation Dose Assessment Questionnaire ...............................95

Abbreviations, Acronyms, and Unit Symbols..........................................................................119

Authors’ Biographies..............................................................................................................122
x

List of Figures

Figure 1. Location of McMurdo Station .....................................................................................6

Figure 2. McMurdo Station in 1963 with the NPP on the side of Observation Hill......................7

Figure 3. The McMurdo Station NPP facility (pre-1973)............................................................7

Figure 4. Plan view of the McMurdo Station NPP ......................................................................8

Figure 5. General arrangement of the equipment in the McMurdo Station NPP ..........................9

Figure 6. Effluent discharge area at the McMurdo NPP............................................................10

Figure 7. Monthly average long-lived beta activity in 24-hour air samples ...............................29

Figure 8. Long-lived beta activity in monthly drinking water samples......................................31

Figure 9. Monthly average tritium activity in drinking water samples.......................................32
xi

List of Tables

Table 1. Environmental release limits.......................................................................................11

Table 2. ERSP sampling schedule ............................................................................................11

Table 3. Categories and numbers of malfunctions.....................................................................14

Table 4. Definitions of health physics, personal safety, and chemistry deviation types.............15

Table 5. Compiled dosimetry records for reactor crew members...............................................26

Table 6. Self-reading dosimeter results for ship loading operations ..........................................27

Table 7. Monthly average of long-lived beta activity for 24-hour air samples* (Bq L−1)............28

Table 8. Air samples taken during removal and loading of bulk contaminated soil ...................28

Table 9. Long-lived beta activity in monthly drinking water samples (Bq L−1) .........................30

Table 10. Monthly average tritium activity in drinking water samples (Bq L−1)........................32

Table 11. Parameter values and assumptions for estimating average dose rates inside the NPP 

facility.......................................................................................................................................36

Table 12. Parameter values and assumptions for external dose while visiting the NPP..............36

Table 13. Estimated average dose rate inside the NPP ..............................................................37

Table 14. Parameter values and assumptions for external dose for support personnel transiting 

near the NPP fence line .............................................................................................................38

Table 15. Parameter values and assumptions for external dose from immersion in radioactive 

noble gases................................................................................................................................40

Table 16. Parameter values and assumptions for external dose from proximity to radioactive 

waste packages..........................................................................................................................41

Table 17. Parameter values and assumptions for exposure to external radiation from shipments 

of RTGs....................................................................................................................................42

Table 18. Parameter values and assumption for the external dose from contaminated soil 

deposited on the ground during decommissioning .....................................................................42

Table 19. Parameter values and assumptions for external dose for ship loading of radioactive 

waste.........................................................................................................................................43

Table 20. Parameter values and assumption for the internal dose from inhalation of airborne 

radioactive materials from stack emissions or worldwide fallout ...............................................46

Table 21. Parameter values and assumption for the internal dose from inhalation of suspended 

contaminated soil during decommissioning ...............................................................................47

Table 22. Parameter values and assumptions for the internal dose from inhalation of suspended 

contaminated soil that was spilled on the ground, roads and ship loading areas..........................48

Table 23. Parameter values and assumptions for the internal dose from ingestion of tritium in 

drinking water...........................................................................................................................49
xii

Table 24. Parameter values and assumptions for the internal dose from ingestion of long-lived 

beta emitters in drinking water other than tritium ......................................................................50

Table 25. Parameter values and assumptions for the internal dose from incidental ingestion of 

contaminated soil and dust ........................................................................................................51

Table 26. External doses and upper-bounds for winter-over support personnel .........................56

Table 27. External doses and upper-bounds for austral summer support personnel ...................56

Table 28. Internal doses and upper-bounds for winter-over support personnel ..........................57

Table 29. Internal doses and upper-bounds for austral summer support personnel.....................58

Table 30. Upper-bound external, internal, and total doses for McMurdo Station support 

personnel during Deep Freeze years 1962 to 1979.....................................................................60

Table 31. Doses from radiation in daily life..............................................................................61

Table 32. Doses from radiological exams.................................................................................61
1

1.

Introduction

This report presents the results of the radiation dose assessment (RDA) for the few 

thousand U.S. Department of Defense (DoD) military personnel who were stationed at McMurdo 

Station, Antarctica, during Operation Deep Freeze (DF) between 1962 and 1979. During this 

period, an onsite nuclear power plant (NPP) was built, operated to generate electricity and steam 

for water distillation, and finally dismantled and decommissioned. There are two groups or 

populations that are addressed in this report. The first group is the support personnel comprised 

of individuals who were stationed at McMurdo Station but not assigned to the NPP. The second 

group consists of the reactor crew members who are defined as those individuals who were 

stationed at McMurdo Station and assigned to the NPP. The two groups differ since reactor crew 

members were individually monitored for occupational exposure to ionizing radiation whereas 

support personnel were not monitored.

In this assessment, upper-bound doses were estimated for the support personnel group. 

The dose estimation included an assessment of potential external and internal sources of 

radiation. Also, for this report, a review of reported doses for reactor crew members was carried 

out. Specific individualized dose assessments for support personnel and reactor crew members 

can be performed through a proposed RDA procedures framework introduced in Section 8 of this 

report, which will be further detailed in a separate document.

1.1 Background

The McMurdo Station is located 850 miles from the South Pole. The U.S. Navy operated 

the NPP from March 1964 to October 1972. The NPP was built and initially operated by 

engineers from the Martin Company, Atomic Energy Commission personnel, and military 

representatives. The NPP arrived at McMurdo Station in December 1961, achieved criticality on 

March 4, 1962, and was transferred to the control of the U.S. Navy in March 1964. The reactor 

was shut down in 1972 when water was detected in the normally dry canning surrounding the 

primary plant piping. It was determined that this water was due to leaks from the primary shield 

tank into the canning. Because the insulation in the canning contained chlorides, the possibility 

of chloride stress corrosion created the need for a detailed inspection that confirmed that the 

insulation has been wetted. The high cost of performing a full inspection of the primary plant for 

stress corrosion resulted in the decision to permanently shut down operations. The U.S. Navy 

operated the NPP in standby mode, i.e., the reactor was subcritical until the decision was made to 

decommission it. The defueling and decommissioning occurred from October 1973 to March 

1979. A detailed history of the NPP is found in Foster and Jones (1982) and Naval Nuclear 

Power Unit (NNPU) (1973a).

In March 2011, Senator Brown of Ohio requested that the Secretary of Defense work 

with the Department of Veterans Affairs (VA), the Veterans' Advisory Board on Dose 
2

Reconstruction (VBDR), and any other relevant parties to review existing methods used to 

evaluate the probability of radiation exposure at McMurdo Station causing an existing cancer 

(probability of causation) in veterans (Brown, 2011). The Under Secretary of Defense for 

Personnel and Readiness (USD(P&R)) responded in May 2011 and stated that the DoD would 

work with these organizations to address any questions related to radiation exposures at 

McMurdo (USD/P&R, 2011). Appendix A includes copies of the letters received from Senator 

Brown and the response from the DoD.

In July 2011, Senator Blumenthal of Connecticut requested the Assistant Secretary of the 

Navy for Energy, Installations, and Environment (ASN(E, I&E)) to work with the VA, the 

VBDR, and other relevant parties to review potential radiation exposures at McMurdo Station 

(Blumenthal, 2011). Later in July 2011, the ASN(E, I&E) responded that the Navy would work 

with the VA, the VBDR, and the Defense Threat Reduction Agency (DTRA) to address the 

McMurdo issue (ASN(E, I&E), 2011). Appendix A includes copies of the letters received from 

Senator Blumenthal and the response from the U.S. Navy.

In response to veterans’ concerns, the DoD assembled an integrated project team of 

military, civilian, and contract radiation dose assessment experts from DTRA and the U.S. Navy,

and knowledgeable McMurdo Station crew members to review the available information and to 

determine the potential health effects to military personnel exposed to ionizing radiation at 

McMurdo Station. In addition, it was agreed that the assessment team would work with the VA 

and VBDR throughout this process. Biographies of this report's authors are provided at the end 

of this document.

The results of the DoD preliminary radiation dose assessment were presented at the 

VBDR’s annual meeting held on March 23–24, 2012, in San Antonio, TX. Lieutenant 

Commander (LCDR) Gregory R. Fairchild, Naval Dosimetry Center, briefed the VBDR on the 

initial results of the radiation dose assessment (Fairchild, 2012a). During the briefing, LCDR 

Fairchild mentioned to the audience that whoever would be willing to provide additional input to 

the study on any experience they may have had during their time at McMurdo Station should 

contact the members of the radiation dose assessment team (VBDR, 2012). Several members of 

the audience related personal experiences at McMurdo Station during the public comment 

session and several veterans provided input following the meeting. Veterans’ experiences and 

input were used in revising the dose assessment. Prior to the VBDR meeting, the Subcommittee 

on Dose Reconstruction Procedures (Subcommittee 1) of the VBDR conducted a preliminary 

review of the dose assessment effort for the McMurdo Station. The members of Subcommittee 1 

found that the methodology developed was sound and agreed to review the revised version of the 

report (VBDR, 2012; Fairchild, 2012a).

1.2 Objectives 

The main objective of the assessment documented in this report is to estimate 

upper-bound radiation doses for McMurdo Station military support personnel that includes an 

evaluation of potential exposures while working and living at the Station, and exposures that 

would have resulted from sources of radiation associated with the NPP during its construction, 

operation, and decommissioning. The upper-bound radiation doses were estimated for groups 

stationed at McMurdo during winter-over (12–14 months) and austral summer (6 months or less) 

starting in December 1961 (DF 62) until March 1979 (DF 79 summer).
3

In addition to the estimating doses accrued by support personnel, a procedural framework 

for individualized dose assessments for both support personnel and reactor crew members was 

developed. Furthermore, the potential health effects due to radiation exposures at the McMurdo 

Station were evaluated.

It is important to note that the present dose assessment does not address exposures from 

radioactive sources at other stations or facilities in Antarctica. It does not include exposures to 

crewmembers aboard the ships that brought in the nuclear fuel rods to the McMurdo Station and 

carried away radioactive waste or spent fuel rods back to the United States. Also, the present 

assessment does not include exposures to radiation from sources used by civilian researchers at 

the station. The other sources of possible exposure to radiation could be addressed in 

individualized dose assessments, if needed.

1.3 Technical Approach

The assessment described in this report is designed to estimate very conservative external 

and internal radiation doses, and corresponding upper-bound doses received by support 

personnel who served at McMurdo Station for the years and activities mentioned above. Highsided conservative parameter values are selected to reflect the top of the range of plausible 

values. High-sided parameter values are not considered to be the worst case. The upper-bound 

dose is estimated to be at least as high as the 95th percentile dose based on comparisons of 

similar assessments using a probabilistic analysis that accounts for uncertainties in the 

determination of dose distributions.

Standard dose reconstruction techniques used in the dose assessment are based on 

standard procedures and methods developed for other veterans’ radiation dose assessment 

programs, e.g., the DTRA Nuclear Test Personnel Review Program (DTRA, 2010). The internal 

doses calculated are the 50-year committed effective dose, the 50-year committed equivalent 

dose to the thyroid, and 50-year committed equivalent dose to the red bone marrow.

The following steps were adopted as part of the approach to develop the upper-bound 

radiation dose assessment for support personnel:

1) Review historical information of reactor operations at the NPP to include design, routine 

operations, incidents, and occupational exposures.

2) Collect information from veterans and U.S. Navy organizations that were involved in 

either the NPP operation or decommissioning with emphasis on finding information on 

radiation measurements, effluent releases, incidents, and radiation safety procedures.

3) Compile and evaluate available occupational dosimetry data of exposed reactor 

personnel.

4) Use the historical data and information provided by veterans to develop exposure 

scenarios and pathways of exposure for all potential NPP radiation sources for both 

winter-over and austral summer personnel.

5) Estimate very conservative external and internal doses and corresponding upper-bounds 

for McMurdo Station support personnel for all exposure pathways using standard dose 

reconstruction techniques.
4

6) Compare the estimated total doses to other types of radiation exposures and determine 

potential health effects.

7) Propose dose assessment procedures for the determination of an individual service 

member’s dose for either the support personnel or reactor crews.
5

2.

History of the McMurdo Station

Nuclear Power Plant

2.1 Installation and Early Operation

2.1.1. Installation

The McMurdo Station NPP was the PM-3A, a portable medium size type 3A, that used a 

pressurized water nuclear reactor with a thermal output of 11.27 megawatts and an electrical 

output of 1.8 megawatts. The reactor used fuel elements that were pellet in rod construction and 

used water as the moderator. (Foster and Jones, 1982)

The McMurdo Station NPP was one of several NPPs constructed and operated by the 

DoD. Other reactors included the SM-1 at Fort Belvoir, Virginia; the SL-1 in Idaho Falls, Idaho; 

the PM-2A at Camp Century, Greenland; the PM-1 at Sundance, Wyoming; the SM-1A at Fort 

Greeley, Alaska; and the MH-1A operated on the USS Sturgis in the Panama Canal Zone. (DOE, 

2001)

The NPP was located on the side of Observation Hill about 300 feet above McMurdo 

Station. The geographical situation and layout of the NPP facility are shown in Figure 1, Figure 2

and Figure 3. A plan view of the NPP and a schema of the equipment are given in Figure 4 and 

Figure 5. The layout of the effluent discharge area is shown in Figure 6.

The nuclear reactor of the NPP was installed in the Primary Building in a below ground 

tank and the steam generator was co-located with the reactor vessel in the same building. The 

radioactive waste disposal system was located in the Primary Building Addition. The electrical 

generators, control room, and the administrative offices were located in the Secondary Building. 

The condensers were in the Condenser Building. The Maintenance and Supply Building and the 

water distillation plant were located outside the main facility. The plant was designed to confine 

the areas of high radiation and high levels of radioactive contamination to the primary building, 

the primary building addition, and to the primary yard inside the perimeter fence. (Foster and 

Jones, 1982)

2.1.2. Early Operation 

Construction of the NPP started in December 1961 on the side of Observation Hill east of 

McMurdo Station. The plant components, including the first reactor core, arrived at McMurdo 

Station on December 13, 1961, with completion of the unloading and hauling of the NPP reactor 

components occurring on December 29, 1961. The reactor achieved criticality for the first time 

on March 4, 1962. A detailed description of the planning and construction of the NPP can be 

found in Foster and Jones (1982). The plant started producing electricity in July 1962 and was 

operated by Navy crews under the direction of employees of the Martin Company and 
6

representatives of the Atomic Energy Commission (AEC) (now superseded by the U.S. Nuclear 

Regulatory Commission). Since the reactor was the first of its type designed for field use, it took 

two years during early operation to complete evaluation and debugging before the U.S. Navy 

assumed operational control and took custody of the plant in March 1964 (Foster and Jones, 

1982).

 

Figure 1. Location of McMurdo Station

McMurdo is 850 miles 

from the South Pole
7

Figure 2. McMurdo Station in 1963 with the NPP

on the side of Observation Hill

Figure 3. The McMurdo Station NPP facility (pre-1973)

Secondary Building

Water Distillation Building

Date of Picture: Pre-1973

“Downtown” McMurdo

Primary Building Maintenance & Supply Building

Condenser Building Primary Building 

Addition
8

Figure 4. Plan view of the McMurdo Station NPP
9

Figure 5. General arrangement of the equipment in the McMurdo Station NPP
10

Figure 6. Effluent discharge area at the McMurdo NPP
11

2.2 Operations and Maintenance

2.2.1. Environmental Radiation Surveillance Program

Operation of the McMurdo Station NPP followed the National Science Foundation 

recommendations for an Environmental Radiation Surveillance Program (ERSP). The operating 

criteria limits for environmental releases of radioactive materials from the NPP are listed in 

Table 1 (Foster and Jones, 1982). Many of the limits were intentionally set below the regulatory 

maximum release criteria that existed at the time of construction. 

Table 1. Environmental release limits

Type of Release Criteria

Incidence of neutrons 1 neutron m−2 min−1 at one mile from the NPP

Gaseous release of radioactive 

materials exclusive of Ar-41 in effluent 4 × 10−14 μCi mL−1 (1.5 × 10−9 Bq mL−1

)

Liquid radioactive waste in effluent 1 × 10−7 μCi mL−1 (4 × 10 3 Bq mL−1)

Ar-41 radioactive gas in effluent 1 × 10−8 μCi mL−1 (4 × 10−4 Bq mL −1

)

In addition to the environmental release limits listed in Table 1, an ERSP environmental 

sampling program was developed by the U.S Public Health Service and continued by the U.S. 

Navy to determine if radiation and radioactivity levels were increasing due to the operation of 

the reactor. This program included monitoring melted snow and galley water and, starting in 

1967, seawater and the drinking water that was produced in the distillation plant adjacent to the 

NPP. The water distillation plant used steam from the NPP. The types of samples and the 

frequency of their collection are listed in Table 2. (Foster and Jones, 1982)

Table 2. ERSP sampling schedule

Type of Sample Frequency

Air sample – long lived beta activity Continuous 24-hour samples at Building 63 in 

the living quarters area

Drinking water – long-lived beta activity 1 sample monthly at the galley 

Sea water-long lived beta activity 1 sample weekly at the distillation plant

Drinking water – tritium 1 sample weekly at the distillation plant (started 

in 1967)

Drinking water – tritium 1 sample weekly at the galley

(started in 1967)

Smear test Weekly at the galley and living quarters
12

2.2.2. Radiation Safety Standards

The radiation safety standards published by the AEC were adhered to at the reactor 

facility even though such standards were not legally binding outside of the United States. The 

standards addressed release of radioactive materials into the environment and control of radiation 

levels in restricted and unrestricted areas. (USPHS, 1963)

To address radiation doses to personnel other than reactor crew, releases of radioactive 

materials from the reactor facility were not to exceed the limits set in Title 10, Code of Federal 

Regulations, Part 20 (10 CFR 20), Radiation Safety. Discharges of radioactive materials into the 

environment were only allowed if individuals at McMurdo Station were not likely to be exposed 

to concentrations in excess of the limits specified in 10 CFR 20, and an individual’s exposure 

could not be averaged over periods greater than one year to determine compliance. (USPHS, 

1963)

The external radiation levels in unrestricted areas had to be sufficiently low to ensure that 

no one would receive an external radiation dose in a year of more than 0.5 rem (5 mSv). 

Radiation levels had to be below 0.002 rem h−1 (0.02 mSv h−1

) or less than 0.1 rem (1 mSv) in 

any seven consecutive days. (USPHS, 1963)

Personnel monitoring was required for individuals who entered any radiation area defined 

as any area where an individual could receive a dose of greater than 0.005 rem (0.05 mSv) in one 

hour or 0.1 rem (1 mSv) in seven consecutive days. Personnel monitoring was also required for 

individuals who entered any high radiation areas defined as any area where an individual could 

receive a dose greater than 0.1 rem (1 mSv) in one hour (USPHS, 1963).

The annual dose limits for reactor crew members and support personnel are not 

mentioned in the historical reports. During the time that the NPP was operational, the standard 

for occupational exposure to ionizing radiation was 3 rem (30 mSv) per 13 weeks with a limit 

such that an occupational worker could not exceed a cumulative lifetime dose of 5 × (N − 18)

rem, where N is the age of the occupational worker. The standard for exposures for members of 

the general population to ionizing radiation was 0.5 rem (5 mSv) per year (Jones, 2005).

2.2.3. Refueling and Maintenance

The reactor was refueled three times during its operational life. The refuelings occurred 

in October 1964, October 1967, and June 1970. The reactor was permanently defueled in July 

1973. The NPP was built with a storage containment vessel to keep spent fuel before shipping it 

to the U.S. (Foster and Jones, 1982). Annual reactor and generator maintenance was usually 

performed during the austral summer when summer support personnel were available (Foster and 

Jones, 1982).

After the first refueling in October 1964, samples of the reactor spent fuel were sent to 

the Battelle Memorial Laboratory in Columbus, Ohio, for post-irradiation examination. The 

spent fuel rod examination was requested due to the increasing fission product accumulation in 

the primary coolant system. The examination determined that the fission product buildup was 

due to leakage through inter-granular cracks in the outer cladding of the fuel rods 

(Brown et al., 1967).
13

The Battelle Memorial Laboratory report on the reactor spent fuel rods (Brown et al., 

1967) was reviewed to determine if the leaking fuel rods could have created new, different, or 

unexpected sources of or pathways for radiation exposure for support personnel at McMurdo 

Station. The leaking fuel rods could have caused the amount of radioactive waste generated at 

the NPP to be relatively greater than otherwise expected or could have caused increased radiation 

releases of radioactive materials through the stack. These potential exposure pathways are 

addressed in the dose assessment using data from 1964. No additional pathways or sources of 

exposure were discovered that are unique to the leaking fuel rods. 

2.2.4. Water Distillation

The McMurdo Station water distillation plant started using steam from the NPP in 

January 1967 as a heat source to produce drinking water from seawater. The reactor replaced an 

oil boiler as the heat source for water distillation. The water distillation plant continued to use 

steam from the NPP until the reactor was shut down in October 1972. The water distillation plant 

produced over 12 million gallons of water using steam from the NPP from 1967 to 1972. The use 

of steam from the NPP as a heat source for water distillation should not have resulted in 

increased exposure to individuals assigned to McMurdo Station. A drinking water monitoring 

program was started in January 1961 to measure beta-gamma radioactivity concentrations other 

than tritium in drinking water made from melted snow. The monitoring program continued after 

the distillation plant was built and began operations. Tritium monitoring was added to the 

monitoring program in 1967. (Foster and Jones, 1982)

2.2.5. Malfunction History and Characteristics

There were a total of 438 malfunctions and abnormal events of the NPP from March 12, 

1964, to September 30, 1973, due to a variety of reasons (NNPU, 1973a). The categories and 

numbers of malfunctions are listed in Table 3. These malfunctions occurred after AEC turned 

over control of the NPP over to the U.S. Navy. There are no detailed records of malfunctions or 

abnormal events before the handover.

A review was made of the records that could be found for the 438 malfunctions to 

determine if any of the malfunctions could have led to increased exposures to support personnel 

or reactor personnel. The accounts of the malfunctions are found in Foster and Jones (1982) and 

give enough detail to determine if the malfunction could have increased radiation levels.

Several of the malfunctions would have increased the doses that reactor crew personnel 

received when extended maintenance on the reactor was needed, or production of more 

radioactive waste required handling and disposal. None of the incidents would have increased 

radiation levels outside of the NPP facility. Any increased doses to reactor personnel would have 

been measured and recorded by the film badges worn by reactor personnel. The malfunctions 

reviewed would have not increased the doses for support personnel above the doses calculated in 

this report.
14

Table 3. Categories and numbers of malfunctions

Category of Malfunction Number of Malfunctions

Nuclear instrumentation 85

Process instrumentation 48

Electrical 83

Mechanical 69

Control rod drive mechanism 68

Operator error 40

Miscellaneous 45

Total 438

During operations at the NPP, reactor crews used the Health Physics Personal Safety and 

Chemistry (HPSC) Reports system to document reportable deviations from HPSC standards 

(Foster and Jones, 1982). There were eight types of reportable deviations that had to be 

documented using the HPSC reporting system. The definitions of the reportable deviations are 

listed in Table 4 (Foster and Jones, 1982). The deviations were reviewed to determine if they 

would have increased doses to support personnel. It was found that some of the deviations could 

have led to increased occupational doses to the reactor personnel. However, none of the 

deviations would have led to doses for the support personnel other than those due to the 

pathways considered in this report. 

The four reports of radioactivity release to the environment in excess of operating limits 

(Type III reports) included: one incident of Xe-133 and Xe-135 the release that exceeded selfimposed plant limits but did not exceed 10CFR20 limits, one incident of contaminated water 

release from holdup tank number 4; one incident of the release of 200 gallons of water with a 

gross beta activity of 4.6 × 10−2 μCi mL−1 (2 × 103 Bq mL−1

); and one incident where 2 gallons 

of water containing gross beta activity of 3 × 10−2 μCi mL−1 (1 × 103 Bq mL−1

) spilled on the 

loading dock (NNPU, 1973a).

Furthermore, shielding water contaminated with radioactive materials leaked into the 

backfill surrounding the reactor vessel through cracks in the containment vessels. The leaks 

contaminated the surrounding soil that had to be removed during decommissioning. (Foster and

Jones, 1982)

In addition to the malfunctions and deviations listed above, the reactor experienced a 

hydrogen fire in the containment tanks in October 1962. The fire led to equipment modifications 

and procedural changes (Foster and Jones, 1982). Since the fire occurred before the Navy 

assumed control of the reactor, the incident was not recorded or counted by the HPSC Reports 

system. Any increased dose from this incident would have been recorded by the film badges 

worn by reactor crew.
15

Table 4. Definitions of health physics, personal safety, 

and chemistry deviation types

Type Definition No. of Reports

Type I Injury to plant personnel or visitors 14

Type II Exposure to personnel greater than 0.350 

rem (3.5 mSv) in seven consecutive days 123

Type III

Release of any radioactivity to the 

environment in excess of Title 10, Code of 

Federal Regulations, Part 20

4

Type IV

Increase of radiation or radioactivity levels 

within the plant by more than a factor of 3 

from normal operating conditions

11

Type V Water chemistry or radiochemistry analysis 

outside of a limiting condition 61

Type VI Inability to perform a required chemistry or 

radiochemistry analysis 5

Type VII

Airborne particulate exposure to personnel 

greater than 3 × 10−10 μCi mL−1

(1 × 10−5 Bq mL−1) for any 40-hour period 

in seven consecutive days

5

Type VIII

Detection of airborne alpha activity greater 

than 2 × 10−12 μCi mL−1

(7 × 10−8 Bq mL−1)

0

Total 223

2.3 Decommissioning 

2.3.1. Removal and Shipment of Plant Components and Contaminated Soil

The decommissioning of the McMurdo Station NPP was planned in 1973 and started at 

the beginning of DF 74. The decommissioning plan included the removal of the reactor vessel 

and reactor components, decontamination and removal of the primary building, and cleanup of 

the site to include removal of contaminated soil and rock. The original plan called for the NPP to 

be cleaned up during the austral summers of DF 74, 75, and 76. The actual cleanup lasted from 

DF 74 to 79 because decommissioning could only occur during the austral summers and because 

of the time needed to receive the analysis results of soil samples. (Foster and Jones, 1982)

During DF 74, the NPP’s secondary systems were removed. In addition, the spent reactor 

fuel and 70 metric tons of radioactive waste were shipped to the United States. The reactor 

pressure vessel was also prepared for shipment by reactor crew personnel. (Foster and Jones, 

1982)

During DF 75, the pressure vessel was prepared and shipped to the United States. To 

complete the removal, 2,300 metric tons of crushed rock backfill was removed from around the 
16

reactor containment tanks by reactor crew members. Over 365 metric tons of radioactive waste 

and equipment contaminated with radioactive materials were shipped to the United States. 

(Foster and Jones, 1982)

During DF 76, the remaining equipment in the Secondary Building and the chemistry 

laboratory were dismantled. The foundation of the Primary Building was removed and the 

backfill cooling system around the reactor tanks was removed. The contaminated soil at the 

facility was surveyed and categorized in three groups based on levels of cesium-137 (Cs-137):

soil contaminated with more than 2,000 pCi g−1 (74 Bq g−1

), soil with less than 2,000 pCi g−1

(74 Bq g−1) but greater than 10 pCi g−1 (0.4 Bq g−1), and soil with levels below 10 pCi g−1

(0.4 Bq g−1). Over 370 metric tons of radioactive waste to include boxed soil contaminated with 

radioactive materials and the steam generator containment vessel were shipped to the United 

States as radioactive waste (Foster and Jones, 1982). The radioactive waste and soil had to be 

shipped to the U.S. since radioactive waste could not be buried for disposal in Antarctica due to 

the Antarctic Treaty (Foster and Jones, 1982). 

During DF 77, radiological surveys were completed for the remaining buildings and 

surveys of the contaminated soil continued. Over 5,000 m3 of contaminated bulk soil was 

shipped to the United States. Since the amount of contaminated soil exceeded the carrying 

capacity of the supply ship, 2,700 m3 of contaminated soil could not be shipped and was stored at 

the NPP facility until DF 78 when it was shipped to the United States (Foster and Jones, 1982).

During DF 78, surveys of the remaining contaminated soil were performed and 3,500 m3

of soil was shipped to the United States. The NUS Corporation conducted a final site survey of 

the NPP facility and other areas such as the effluent release areas (Foster and Jones, 1982).

During DF 79, soil with the highest level of Cs-137 contamination identified by the NUS 

Corporation survey during DF 78 was removed. More than 540 m3 of contaminated soil were 

removed and shipped in bulk to the United States (Foster and Jones, 1982). 

The soil and rock at the NPP site were contaminated by several processes. First, there 

were three instances during plant operations when cracks in the containment vessels occurred. 

The cracks allowed shield water to leak into the crushed rock backfill surrounding the 

containment tanks. Second, backfill and soil near the buildings were contaminated by radioactive 

waste and water handling operations and other operations in the primary yard. Third, during 

normal plant operations, liquid effluent containing activity below maximum permissible 

concentration (MPC) set by the AEC was discharged and allowed to flow over the ground. 

Radioactive materials suspended in the effluent were absorbed into the rock and soil in the 

discharge area and concentrated in the first few centimeters. (Foster and Jones, 1982)

2.3.2. Radioactive Waste

Radioactive waste was generated during the operation of the NPP from 1962 to 1973 and 

during the decommissioning of the facility from 1973 to 1979 (Foster and Jones, 1982). 

According to the Antarctic Treaty, no radioactive waste could be disposed of in Antarctica 

(Foster and Jones, 1982). All radioactive waste had to be transported to the United States for 

disposal (USPHS, 1963).

Solid radioactive waste consisting of resins, cartridges, damaged reactor components, 

rags, paper, and contaminated laboratory equipment was produced over the course of operation 
17

of the NPP. The radioactive waste was shipped for disposal each austral summer during the 

operational phase of the plant. The radioactive waste was shipped to the United States in 

approved shipping containers such as Type A waste barrels and shielded shipping casks (Foster 

and Jones, 1982). Detailed records of the amounts and types of radioactive waste generated each 

year are not available.

An example of the type of radioactive waste shipped from McMurdo Station is the 

packing listings from DF 72. Such waste would be similar to waste generated and shipped to the 

United States in other operational years to include the shipment of a spent fuel core. The list of 

radioactive waste for that year includes 23 Department of Transportation drums, 17 H drums in a 

Conex container, and Type A normal waste containers with a total weight of 2.27 ton (2100 kg 

and a volume of 10 m3 (U.S. Navy, 1972a). Solid waste resins were also shipped during DF 72 in 

two Type A, Special Form containers with a total weight of 25.4 metric tons. In addition, a spent 

fuel core of 17.2 metric tons and 18.5 m3 was shipped to the United States in 1972. (U.S. Navy, 

1972a)

Liquid radioactive waste from the NPP was collected in a waste tank and processed using 

an evaporator. The condensate from the evaporator was held in a second tank for sampling. If the 

condensate was above effluent discharge concentration limits, it was recycled back to the 

evaporator. If the condensate was below release limits, it was discharged to the effluent 

discharge area. (Foster and Jones, 1982)

The gaseous radioactive waste from the reactor was vented through the discharge stack 

after passing through particulate and charcoal filters. The stack radiation detectors were in place 

to monitor the release of radioactive gases (Foster and Jones, 1982). However, no records of 

specific stack monitoring are available.

Four irradiated NPP reactor cores were returned to the United States. In addition, three 

cores were shipped to McMurdo Station but were not used in the NPP (Foster and Jones, 1982). 

The four irradiated, used cores were shipped to the Savannah River Plant, South Carolina. The 

three unused cores were returned to either the Savannah River Plant, South Carolina; the Nuclear 

Materials and Equipment Corporation in Apollo, Pennsylvania; or the National Lead Company 

in Crosby, Ohio. (Foster and Jones, 1982)

No major incidents or violations of federal or international regulations concerning 

radioactive waste occurred during the operation of the NPP (Foster and Jones, 1982). Seven 

incidents involving the shipment of radioactive waste were recorded. These include a leaking 

radioactive waste container, damaged radioactive waste containers, mismarked radioactive waste 

containers, and the mixing of radioactive waste with other dangerous goods, all occurring after 

the waste left McMurdo Station. These incidents are described in Foster and Jones (1982). 

A review of the incidents for this assessment concluded that none of these would have led to 

increased radiation exposure levels to support personnel since the incidents occurred after the 

packages left McMurdo Station or occurred inside the NPP.

2.3.3. Final Surveys

Two final surveys the original plan survey and the follow up survey, were performed 

during the decommissioning of the NPP. The first final survey was performed during DF 78 by 

the NUS Corporation (Jentz, 1978). The NUS report states that out of the 465 survey points, 285 
18

had soil concentration levels between 10 and 100 pCi g−1 (0.4 Bq g−1 and 4 Bq g−1

). Fourteen 

survey points were above 100 pCi g−1 (4 Bq g−1) with the highest being 400 pCi g−1 (16 Bq g−1). 

The NUS report was used to plan final decommissioning activities during DF 79 (Foster and 

Jones, 1982). A second final survey was performed during DF 79 by the U.S. Navy after the 

decommissioning efforts of DF 79 were completed. The average concentration in the soil over 

the site was 8 pCi g−1 (0.32 Bq g−1) with highest concentration of Cs-137 in soil being 29 pCi g−1

(1.2 Bq g−1). (Foster and Jones, 1982)
19

3.

Potential Sources of Radiation Exposure

for Support Personnel

All support personnel stationed at McMurdo Station were potentially exposed to external 

gamma and neutron radiation, and internal radiation from the intake of radioactive materials by 

inhalation or ingestion. The applicability of potential sources of exposure for support personnel 

is based on their activities and their work and living locations. In this section, potential sources 

of external exposures are discussed first, followed by potential sources of exposures from 

internally-deposited radioactive materials. 

3.1 External Gamma and Neutron Radiation

The support personnel at McMurdo Station worked and lived outside the NPP and could 

have been exposed to external gamma radiation from sources inside and outside the reactor 

facilities. The following sources of potential exposure to external radiation are considered for the 

support personnel at McMurdo Station: 

1) Radiation sources inside the NPP while on a tour or visit of the plant. Such sources 

include the reactor vessel, contaminated reactor facility components and equipment, 

radioactive waste, packages containing radioactive materials, radioactive gases, and 

radioactive materials and contaminated soil in the effluent discharge area, and NPP yard.

2) Radiation sources inside the NPP facility while near the fence, such as when driving, 

hiking or walking by.

3) Immersion in indoor and outdoor air containing radioactive noble gases released from the 

NPP during operations.

4) Radioactive packages transported to or from the ship loading area outside the NPP 

facility.

5) Radioisotope thermal generators while being moved.

6) Radioactive waste during loading onto ships.

7) Contaminated soil spilled on the ground, roads and loading areas during 

decommissioning.

3.1.1. Radiation Sources inside the NPP

The support personnel at McMurdo were not routinely exposed to external gamma and 

neutron radiation while inside the NPP facility. However, some support personnel could have 

been exposed during visits to the NPP as part of their duties such as fire inspection (Armstrong, 
20

2012), or during occasional tours of the facility according to a previous officer-in-charge of the 

NPP (Fegley, 2012). To be conservative, all support personnel are assumed to have been exposed 

to this radiation source. Exposure to external radiation while inside the NPP could have occurred 

during all DF years.

3.1.2. Radiation Sources while Transiting Near the NPP

Support personnel could have been exposed to external gamma radiation from sources in 

or near the NPP facility while driving, hiking or walking near the NPP fence (Armstrong, 2012). 

As a conservative assumption, all support personnel are presumed to have been exposed to this 

radiation. Also, exposure to external radiation from this source is assumed to have occurred 

during any DF year an individual was at McMurdo Station. 

3.1.3. Immersion in Radioactive Noble Gases

Support personnel could have been exposed to external gamma radiation during 

immersion in air containing radioactive noble gases outside of the NPP facility. The NPP 

released air containing Ar-41 at concentrations not greater than 1 × 10−2 pCi mL−1

(4 × 10−4 Bq mL−1). Ar-41 was produced in the NPP from the neutron activation of Ar-40 that is 

naturally present as 0.9% of air. The level of radioactive Xe-135 emissions exceeded the 

operating technical limits one time, but Ar-41 emissions never exceeded applicable limits (Foster 

and Jones, 1982). All support personnel who were assigned to McMurdo Station while the 

reactor was operational are assumed to have been exposed to this source of external radiation. 

3.1.4. Movement of Packages Containing Radioactive Materials 

outside the NPP

Support personnel could have been exposed to external gamma radiation during the 

movement of radioactive packages to and from the NPP facility. These packages consisted 

primarily of radioactive waste. Except for the seven incidents referred to in Section 2 and 

reviewed separately, the radioactive waste packages are assumed to have been shipped in 

accordance with International Atomic Energy Agency (IAEA) requirements and were limited to 

external exposure rates acceptable for the labeling of the package (Foster and Jones, 1982). All 

support personnel are assumed to have been exposed to this radiation source. The radioactive 

waste packages would have contained fission products removed from the demineralizers of the 

primary coolant loop of the NPP. Thus, any fission products that leaked into the primary coolant 

through fractures in the cladding of the reactor fuel rods would have been removed by the 

demineralizer and processed by the radioactive waste control system (Brown et al., 1967). The 

exposure of support personnel to radiation from packages containing radioactive materials is 

presumed to have occurred during any DF year.

3.1.5. Movement of Radioisotope Thermal Generators 

Support personnel could have been exposed to external x-ray radiation consisting of

bremsstrahlung radiation produced by the radioactive sources in Radioisotope Thermal 

Generators (RTGs) stored at the station awaiting transport. RTGs are portable electrical power 

units that use the decay heat from radioactive materials as their power source. They were not 
21

used at McMurdo Station, but were shipped to McMurdo Station during transit to other locations 

and were temporarily stored at the NPP. The RTGs are assumed to have been shipped in 

accordance with IAEA requirements and their radiation levels were limited to external exposure 

rates acceptable for the labeling of the package (Foster and Jones, 1982). All support personnel 

are assumed to have been exposed to this radiation source. The exposure of support personnel to 

radiation from RTGs is presumed to have occurred during any DF year.

3.1.6. Loading of Radioactive Waste

Support personnel could have been exposed to external gamma radiation from shipments 

of radioactive waste or bulk soil contaminated with radioactive materials as part of their duties. 

Self-reading (self-indicating) dosimeters were issued to McMurdo Station support personnel and 

ship crews during the loading of some of the shipments of radioactive waste or large quantities of 

soil contaminated with radioactive materials (NNPU, 1976 and 1977a) Self-reading dosimeter 

records are available for one year and are assumed to reflect the external radiation exposures 

received by support personnel during the radioactive waste loading operations. The loading of 

the radioactive waste is presumed to have occurred during any DF year. The loading of bulk 

contaminated soil would have occurred during the decommissioning years from DF 77 to 79.

3.1.7. Spilled or Residual Contaminated Soil

Support personnel could have been exposed to residual radiation from bulk contaminated 

soil left over after ship loading or spilled onto the roads leading to the ship loading area 

(NNPU, 1977a). Support personnel would have been exposed to higher levels of radiation from 

spilled soil during loading and shipping operations than after spill cleanup had occurred. These 

exposures would have occurred during DF 75, 76, 77, 78, and 79.

3.2 Internal Radiation

The support personnel at McMurdo Station worked and lived outside the NPP and could 

have been exposed to internal radiation resulting from the inhalation or ingestion of radioactive 

materials released outside the plant. The following sources of potential exposure to internal 

radiation are considered for support personnel at McMurdo Station:

1) Inhalation of airborne radioactive materials from the emissions of fission and activation 

products released through the NPP stack or worldwide fallout.

2) Inhalation of suspended soil contaminated with radioactive materials during the 

decommissioning of the NPP; suspended soil spilled during transport to the ship loading 

area and ship loading and suspended residual soil contamination.

3) Ingestion of tritium in drinking water.

4) Ingestion of radioactive materials in drinking water other than tritium.

5) Incidental ingestion of contaminated soil and dust during decommissioning.
22

3.2.1. Inhalation of Airborne Radioactive Materials from Emissions Released 

through the NPP Stack and Worldwide Fallout

Support personnel could have inhaled airborne radioactive materials to include gases and 

aerosols released from the main stack of the NPP and worldwide fallout from past and ongoing 

atmospheric nuclear testing. The inhalation could have occurred both during duty and non-duty 

times in areas outside of the NPP facility (Foster and Jones, 1982). All support personnel are 

assumed to have inhaled airborne radioactive materials in effluent from the NPP during the 

operation phase (DF 62 to 72) and airborne worldwide fallout during any DF year. 

3.2.2. Inhalation of Suspended Soil Contaminated with Radioactive Materials

Support personnel could have inhaled suspended soil contaminated with radioactive 

materials during the excavation and removal of backfill surrounding equipment and building 

installations at the NPP. Suspended soil contaminated with radioactive materials could have 

drifted beyond the NPP boundary and been inhaled by support personnel. Support personnel are 

assumed to have been exposed to suspended contaminated soil during DF 75 to 79.

Support personnel could have been exposed to suspended contaminated soil during the 

decontamination of the effluent release areas of the NPP. Support personnel who participated in 

the decontamination of the effluent release areas adjacent to the NPP would have been exposed 

to the suspended residual soil contaminated with radioactive materials for their entire time at 

McMurdo Station. Support personnel are assumed to have been exposed to suspended 

contaminated soil during decontamination of the effluent release areas adjacent to the NPP 

during DF 77 to 79.

Support personnel could have inhaled suspended soil contaminated with radioactive 

materials that was spilled during the transport of bulk soil to the ship loading areas or during ship 

loading. The support personnel are assumed to have assisted in the bulk soil loading. In addition, 

support personnel could have inhaled suspended residual contaminated soil after the spilled bulk 

soil contaminated with radioactive materials was cleaned up. Support personnel could have been 

exposed to suspended spilled contaminated soil and residual spilled soil during DF 75 to 79.

3.2.3. Ingestion of Tritium in Drinking Water

Support personnel could have ingested tritium in the drinking water at McMurdo Station. 

Measurements in drinking water showed low levels of tritium that possibly could have come 

from the water effluent released from the NPP or originated from worldwide fallout, which then 

entered the water supply sources. Support personnel are assumed to have ingested tritium in 

drinking water during any DF year. (Foster and Jones, 1982)

3.2.4. Ingestion of Radioactive Materials in Drinking Water other than 

Tritium

Support personnel could have ingested radioactive materials other than tritium in the 

drinking water at McMurdo Station. The drinking water could have contained long-lived beta 

activity other than tritium released from the NPP through water or gas emissions that found its 

way to the source of drinking water. Radioactive contamination in drinking water could also be 
23

from naturally occurring radioactivity and fallout from atmospheric testing of nuclear weapons 

(Foster and Jones, 1982). Support personnel are assumed to have ingested radioactive material in 

drinking water during any DF year.

3.2.5. Incidental Ingestion of Soil and Dust Contaminated with Radioactive 

Materials

Support personnel at McMurdo Station could have incidentally ingested soil and dust 

contaminated with radioactive materials. The soil could have been from spills in the ship loading 

area or on the roads leading to the ship docks (NNPU, 1976). Support personnel are assumed to 

have incidentally ingested contaminated soil and dust during DF 77 to 79.
24

4.

Radiation Monitoring 

During the operation and decommissioning of the McMurdo Station NPP, measurements 

were made to characterize radiation exposures of reactor crew members as well as for 

environmental monitoring. This includes external radiation levels inside and in areas adjacent to 

the NPP facility, and in the effluent discharge area. Measurements were also made of radioactive 

materials concentrations in emissions from the plant stack. Furthermore, measurements were 

made to monitor air activity concentrations near the living quarters of McMurdo Station 

personnel and water activity concentrations in the drinking water. This section will review the 

available data that can be used to estimate radiation doses for support personnel.

4.1 Exposure Records of Reactor Personnel

Film badge dosimetry records covering the periods when reactor crew members were 

assigned to the NPP have been located for individuals who served during DF 62, 63, 64, 65, 72 

and 73. Partial dosimetry results or results for one of the two calendar years of a winter-over tour 

have been located for some reactor crew members for DF 66, 67, 68, 69, 70, and 71. Partial 

austral summer film badge results are available for DF 74. For all other austral summers during 

decommissioning activities, sufficient film badge results were not available to perform a 

statistical analysis because reactor crew personnel were only present at McMurdo Station during 

the austral summer. (NDC, 1962–1979)

The compiled gamma radiation dosimetry records for reactor crew members for DF years 

for which dosimetry results were located are summarized in Table 5. For the years where 

complete dosimetry results are available, the doses varied based on the Navy occupational 

ratings of the individuals wearing the film badges (ASF, 1962–1979; NDC, 1962–1979). The 

highest film badge results were typically those for utilitiesmen and construction electricians. The 

lowest film badge results were typically for those storekeepers or logistics specialists.

Table 5 does not include neutron doses because they were not reported for all DF years 

for which complete gamma radiation results are available. The neutron doses are considered to 

represent less than 5% of the total gamma doses received by reactor crew members (Johnson, 

1971). The neutron doses for support personnel are considered to be insignificant and not 

included in the dose assessment.

Internal monitoring records have been located for some of the reactor crew members. The 

internal monitoring consisted of whole-body counting of returning crew members at the National 

Naval Medical Center in Bethesda, Maryland (now called Walter Reed National Military 

Medical Center). Whole-body counting consists of using an array of photon detectors to 

measure the gamma emitters that have been deposited inside one’s body (Cember, 1983). The 

internal monitoring measurements consisted of estimates of the amount of Cs-137, Mn-54, and 

Co-60 in the body. The U.S. Navy compared the results, at the time of measurement, to the 
25

National Bureau of Standards Handbook 69 (U.S. Navy 1972b; NBS, 1959) values for 

permissible body burdens to determine if federal guidance had been exceeded. For the 

whole-body count data available for review, no results exceeded the maximum permissible body 

burdens (U.S. Navy, 1972b).

Self-reading dosimeters were used to monitor short-term exposures to reactor crew 

members in the NPP during the operation of the reactor. They were worn in addition to a reactor 

crew member’s film badge, and were used to assist in assessing compliance with the Type II 

HPSC Reports of radiation doses in excess of 0.350 rem (3.5 mSv) in seven consecutive days as 

well as to ensure compliance with radiation work permits that were used to document work 

activities in areas of high radiation levels (Foster and Jones, 1982; Martin Company, 1964).

Self-reading dosimeters were also used to monitor the doses of support personnel who 

assisted in the transport and ship loading of radioactive waste and contaminated soil. The selfreading dosimeters were issued for the days when ships were being loaded with radioactive 

waste or contaminated soil (NNPU, 1977a). Dosimeter results were located for two dates and are 

shown in Table 6. The self-reading dosimeter results were for DF 76, the year during which a 

large amount of radioactive waste—including the contaminated pressure vessel—was returned to 

the U.S. (Foster and Jones, 1982).

4.2 External Radiation Measurements

Measurements of dose rates inside the NPP and the radioactive waste storage areas 

during operational and decommissioning years have not been located. However, it is well 

documented that exposure rate and other environmental monitoring measurements were made 

inside the NPP (Foster and Jones, 1982; Martin Company, 1964). Discussions with veterans 

indicated that gamma and neutron radiation survey instruments were available for use 

(Fegley, 2012). A review of the NNPU Health Physics Operations Logs indicated that external 

radiation measurements were made to include dose rates and contamination activity 

measurements. However, only a small number of records was available for review 

(NNPU, 1969). Currently, insufficient external dose/exposure rate data is available to be able to 

perform detailed dose assessments for support personnel using those types of measurements. 

During DF 78, 43 environmental thermoluminescent dosimeters (TLD) were exposed for 

85 days at NPP on-site locations and for 76 days in off-site locations. The TLDs in the on-site 

locations measured radiation doses between 0.005 rad (0.05 mGy) to 0.048 rad (0.48 mGy). The 

TLDs in the off-site locations measured radiation levels of 0.004 rad (0.04 mGy) to 0.032 rad 

(0.32 mGy).

External radiation measurements were made in the effluent discharge areas during the 

decommissioning of the NPP from 1974 to 1979. Records of some measurements have been 

located (NNPU, 1976; NNPU 1977b). Radiation level measurements were made to locate the 

areas with high levels of soil contamination during the initial phase of the decommissioning. 

Additional measurements were made after the removal of contaminated soil during subsequent 

phases of decommissioning. The results of the radiation level rate measurements were not 

suitable for use in the dose reconstruction because the measurements were not available for all 

years, the measurements were reported in units of counts per minute, and they were spot 

measurements.
26

Table 5. Compiled dosimetry records for reactor crew members

DF No. Minimum

(rem (mSv))

Maximum

(rem (mSv))

Mean

(rem (mSv)) Remarks

62 13 0.137 (1.37) 0.716 (7.16) 0.316 (3.16) Less than full year of 

operation

63 21 0.160 (1.60) 2.878 (28.8) 0.890 (8.90)

63 Summer 6 0.220 (2.20) 0.835 (8.35) 0.496 (4.96)

64 17 0.060 (0.60) 1.965 (19.7) 0.860 (8.60)

64 Summer 11 0.247 (2.47) 1.915 (19.2) 0.946 (9.46)

65 16 0.572 (5.72) 3.968 (39.7) 2.048 (20.5)

65 Summer 10 0.854 (8.54) 3.499 (35.0) 1.863 (1.86)

66 23 0.060 (0.60) 3.814 (38.1) 1.696 (17.0) CY 66 only*

66 Summer 10 0.067 (0.67) 2.705 (27.1) 0.949 (9.49) CY 66 only*

67 21 0.096 (0.96) 6.145 (61.5) 3.624 (36.2) CY 67 only*

67 Summer 11 0.067 (0.67) 3.187 (31.9) 1.334 (13.3) CY 67 only*

68 20 0.168 (1.68) 3.072 (30.7) 1.409 (14.1) CY 68 only*

68 Summer 10 0.285 (2.85) 3.057 (30.6) 1.702 (17.0) CY 68 only*

69 21 0.110 (1.10) 7.551 (75.5) 3.421 (34.2) CY 69 only*

69 Summer 4 1.635 (16.4) 3.620 (36.2) 2.836 (28.4) CY 69 only*

70 22 4.602 (46) 10.800 (108) 7.913 (79.1) CY 70 only*

70 Summer 5 0.965 (9.65) 5.895 (59.0) 2.817 (28.2) CY 70 only*

71 18 0.120 (1.20) 2.840 (28.4) 1.681 (16.8) CY 70 only*

71 Summer 10 0.087 (0.87) 3.804 (38.0) 1.912 (19.1) CY 70 only*

72 22 0.185 (1.85) 9.955 (99.6) 7.034 (70.3) Reactor shut down in Oct 

1972

72 Summer 4 0.564 (5.64) 1.879 (18.8) 1.316 (13.2) Reactor in standby mode

73 9 0.693 (6.93) 2.564 (25.6) 1.502 (15.0) Reactor not operational

73 Summer 18 0.024 (0.24) 5.766 (57.7) 2.295 (22.9) Reactor not operational

74 Summer 7 0.716 (7.16) 4.905 (49.1) 2.663 (26.6) Reactor not operational

*

Dosimetry located for one of two calendar years that comprised a DF tour (see Section 5 for discussion of how 

partial dosimetry records are used in the dose assessment for support personnel).

Exposure rate measurements in the vicinity of radioactive waste packages were made. 

Records of 11 radioactive waste shipments made in 1970 through 1972 have been located. These 

records include dose rates at one meter from outside surface of the radioactive waste packages. 

The average dose rate at one meter of the radioactive waste packages is 0.008 rad h−1 

(0.08 mGy h−1). Also, records for 21 shipments of RTGs were located. The average dose rate at 

one meter from the outside surface of the RTG packages is 0.003 rad h−1 (0.03 mGy h−1). 

(NNPUD, 1972)
27

Table 6. Self-reading dosimeter results for ship loading operations

DF Date

Number of 

Dosimeters 

issued

Minimum 

Self-Reading 

Doses

Maximum 

Self-Reading 

Doses

Average of all

Self-Reading 

Doses

76 Feb 8, 76 26 0.001 rem

(0.01 mSv)

0.036 rem

(0.36 mSv)

0.012 rem

(0.12 mSv)

76 Feb 9, 76 35 0.001 rem

(0.01 mSv)

0.084 rem

(0.84 mSv)

0.023 rem

(0.23 mSv)

4.3 Airborne Activity Concentrations

The monthly averages of the continuous 24-hour air sampling results are available from 

January 1961 before the arrival and installation of the reactor (Busdiecker and Maxwell, 1964) 

until December 1972, after the shutdown of the reactor. The air samples were taken at 

Building 63 in the main building area of McMurdo Station. The air samples were collected for 

24 hours each day and are assumed to represent the average air concentration inhaled by all 

individuals at McMurdo Station, including support personnel. The air sample results are reported 

as gross long-lived beta activity in the air. The air samples are assumed to contain I-131, Cs-137, 

Ru-106, Zr-95, and Ce-144 (Foster and Jones, 1982). Also, the air sample results are assumed to 

include natural background activity (Foster and Jones, 1982; NNPU, 1973b). The samples were 

analyzed using a Geiger-Mueller detection system (Martin Company, 1964). The monthly 

averages are shown in Table 7 and Figure 7.

Some of the variations in monthly measured air activity can be attributed to atmospheric 

nuclear testing by several countries to include atmospheric tests in 1961–1962, 1966, 1968, and 

1970. Some of the variation in monthly measured air activities have also been attributed to 

seasonal mixing of the stratosphere and troposphere as seen in 1965 and early 1970 (Foster and 

Jones, 1982). The periods marked with “FO” in Figure 7 are associated with worldwide fallout 

while “AM” designates the periods of increased measured activity due mainly to atmospheric 

mixing. The periods from August 1965 to September 1966, March 1968 to February 1969, and 

June 1969 to March 1970 are considered the months when the measured long-lived beta activity 

in air samples were least affected by worldwide fallout.

Air sampling data collected during the removal of contaminated soil from the effluent 

discharge area are available for November and December 1975, January 1976, and January 1977. 

The air samples were taken in the area where contaminated soil was removed near the NPP or in 

the area of bulk soil ship loading area. For the dose calculations, the air samples are assumed to

measure airborne Cs-137. The results of the air samples during soil decontamination are given in 

Table 8. (NNPU, 1976; NNPU, 1977b)
28

Table 7. Monthly average of long-lived beta activity for 24-hour air samples* (Bq L−1

)

Year Jan Feb Mar Apr May June Jul Aug Sep Oct Nov Dec

1961 2.22 1.85 0.74 0.74 0.74 0.37 0.74 0.37 0.74 0.74 0.74 0.74

1962 0.74 0.74 1.11 0.74 0.74 1.11 2.59 1.85 2.59 2.96 3.70 1.85

1963 4.81 7.40 5.92 5.55 4.07 2.59 2.59 2.59 2.22 1.48 2.22 2.96

1964 3.33 3.33 2.59 2.59 2.96 2.96 3.33 2.22 2.96 3.70 4.07 4.44

1965 4.07 5.55 3.70 2.59 2.96 2.96 2.22 [1.85 2.22 1.48 0.74 1.11

1966 1.11 1.11 1.11 0.74 0.74 0.74 1.11 1.48 1.48] 2.22 4.07 4.44

1967 6.29 5.92 4.81 3.33 2.96 2.96 2.96 3.70 4.07 4.07 3.33 2.59

1968 2.22 1.85 [1.48 1.48 1.48 1.48 1.11 1.11 1.48 1.48 1.85 1.48

1969 1.85 1.48] 4.81 4.81 2.59 [1.11 1.11 1.85 1.11 1.11 1.48 0.74

1970 0.74 2.59 1.11] 2.22 7.03 6.29 1.85 1.48 2.59 1.85 2.22 6.66

1971 0.74 4.07 1.85 11.84 8.88 4.07 2.96 2.96 3.70 7.40 5.18 5.18

1972 5.92 5.18 6.29 2.96 3.33 1.85 2.22 1.48 2.22 1.11 0.74 1.85

*

The highlighted cells correspond to the periods that were least affected by worldwide fallout and most 

likely representing mostly emissions from the NPP. The start and stop of each period is indicated with 

square brackets.

Table 8. Air samples taken during removal and loading

of bulk contaminated soil

DF Date (Location) Location Activity 

μCi mL−1 (Bq mL−1)

76 November 25, 1975 NPP 2 × 10−11 ( 7 × 10−7)

76 December 1, 1975 NPP 4 × 10−12 (1 × 10−7)

76 January 31, 1976 NPP 9 × 10−13 (4 × 10−8)

77 January 30, 1977 Ship loading area 9 × 10−13 (4 × 10−8)
29

Figure 7. Monthly average long-lived beta activity in 24-hour air samples

FO = fallout AM = atmospheric mixing Aug-65 Sep-66 Mar-68 Feb-69

Jun-69

Mar-70

0.0E+00

2.0E-06

4.0E-06

6.0E-06

8.0E-06

1.0E-05

1.2E-05

1.4E-05

Jan-61Jan-62Jan-63Jan-64Jan-65Jan-66Jan-67Jan-68Jan-69Jan-70Jan-71Jan-72Jan-73

Air Sample Long-Lived Beta Activity (Bq L-1)

Date

AM

FO

AM

FO FO

FO
30

4.4 Radioactivity Concentrations in Water

Monthly drinking water samples were collected at the galley. The water samples were 

analyzed by evaporation of the water on a planchet followed by measurement using a shielded 

Geiger-Mueller detector capable of detecting gross beta-gamma activity (Martin Company, 

1964).The activity concentration results are available from January 1961 to December 1972. 

These samples were analyzed for gross long-lived beta activity. The NNPU raised concerns 

about some abnormally high levels reported during 1969. The NNPU concluded that the 

unusually higher concentration results were due to inadequate measurement techniques (Foster 

and Jones, 1982). However, for this assessment, the reported results are used because they result 

in higher, more conservative dose estimates. In addition, the results are assumed to include 

background (Foster and Jones, 1982; NNPU, 1973a). 

The results of monthly drinking water sampling are shown in Table 9 and Figure 8.

Monthly drinking water sample data was not available for April and September, 1961 and for 

July, 1962. To account for the possibility of residual radioactive materials in the drinking water 

after the end of the drinking water sampling program, the December 1972 result is used for all 

months from January 1973 to March 1979 for the dose assessment. 

Table 9. Long-lived beta activity in monthly drinking water samples (Bq L−1

)

Year Jan Feb Mar Apr May June July Aug Sep Oct Nov Dec

1961 0.04 0.44 0.15 ND* 0.15 0.15 0.11 0.26 ND 0.11 0.22 0.04

1962 0.02 1.30 0.22 0.19 0.22 0.26 ND 0.48 0.26 0.48 0.30 0.26

1963 0.93 0.15 0.44 0.95 0.26 0.59 0.59 0.33 0.22 0.09 0.31 0.17

1964 0.71 0.67 0.45 0.47 0.32 0.30 0.45 0.80 0.55 0.70 0.66 0.64

1965 0.68 1.09 0.81 1.03 1.15 0.26 0.68 0.64 0.33 0.45 0.65 0.67

1966 0.43 0.21 0.42 0.40 0.28 0.21 0.42 0.24 0.26 0.06 0.00 0.00

1967 0.00 0.00 0.30 0.28 0.28 0.23 0.23 0.17 2.18 0.19 0.63 0.19

1968 0.36 0.08 0.05 0.05 0.05 0.05 0.05 0.07 0.04 0.41 0.27 0.14

1969 0.43 1.58 1.20 1.20 2.06 4.18 0.44 1.55 0.47 0.46 0.46 0.70

1970 1.96 0.28 0.15 0.16 0.31 0.10 0.17 0.29 0.10 0.64 0.16 0.15

1971 0.24 0.17 0.15 0.19 0.20 0.22 0.23 0.23 0.23 0.22 0.22 0.19

1972 0.36 0.14 0.14 0.29 0.22 0.19 1.45 0.38 1.48 0.27 0.21 0.31

*

Results were not available for the months marked with ND. For the dose assessment calculation, the higher value 

from the two adjacent months was assigned as the activity for those months.
31

Figure 8. Long-lived beta activity in monthly drinking water samples

Weekly samples of the distillate (drinking water) from the water distillation plant were 

taken and analyzed for tritium starting in March 1967 and ending in November 1972. The results 

of the monthly average activity of the weekly tritium drinking water monitoring are shown in 

Table 10 and Figure 9. Results are not available for April 1967, December 1968, December 

1969, April 1970, June 1970, and December 1972. The higher of the preceding and following 

month’s results are used in the dose assessment as an estimate for the missing months (Foster 

and Jones, 1982; NNPU, 1973a). The value of 16,000 Bq L−1 for the first month tritium was 

measured, which was followed by a month with no measurements, appears to be an outlier 

possibly due to first implementation of a new analysis procedure. This value was still used in the 

dose calculation for the months of March and April 1967.

Additionally, weekly seawater samples were analyzed for tritium starting in March 1967 

and ending in November 1972 (Foster and Jones, 1982 and NNPU, 1973). The tritium in 

seawater results were not used in the dose assessment since the tritium monitoring of the 

drinking water from the water distillation plant is used to estimate tritium ingestion doses.

0

0.5

1

1.5

2

2.5

3

3.5

4

4.5

Jan-61Jan-62Jan-63Jan-64Jan-65Jan-66Jan-67Jan-68Jan-69Jan-70Jan-71Jan-72Jan-73

Long-Lived Beta Activity in Drinking Water (Bq L-1) 

Date
32

Table 10. Monthly average tritium activity in drinking water samples (Bq L−1

)

Year Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec

1967 16,000 ND* 330 23 2,400 200 120 70 480 2,800

1968 1,500 340 1,800 600 300 300 200 170 170 190 3,000 ND

1969 180 160 180 730 160 180 180 ND 230 220 500 640

1970 250 440 1,000 ND 300 ND 260 220 260 210 370 100

1971 80 90 80 150 40 40 40 40 40 40 150 50

1972 70 100 130 140 150 100 150 280 120 120 140

*

Results were not available for the months marked with ND. For the dose assessment calculation, the 

higher value from the two adjacent months was assigned as the activity for those months.

Figure 9. Monthly average tritium activity in drinking water samples

4.5 Radioactivity Concentrations in Soil

Contaminated soil was analyzed during the decommissioning of the NPP. In 1978, soil 

and rock samples were analyzed by the U.S. Navy Civil Engineering Laboratory, the Department 

of Energy, and the NUS Corporation for radioactive contamination. Samples of crushed rock 

0.0E+00

2.0E+03

4.0E+03

6.0E+03

8.0E+03

1.0E+04

1.2E+04

1.4E+04

1.6E+04

1.8E+04

Jan-67Jan-68Jan-69Jan-70Jan-71Jan-72Jan-73

Tritium Activity in Drinking Water (Bq L-1) 

Date
33

were taken in December 1977 and January 1978 from the crushed rock and soil piles that were 

awaiting transport as part of the 3,500 m−3 of contaminated soil shipped to the United States. The 

results for Cs 137 contamination ranged from a low of less than 1 pCi g−1 (0.04 Bq g−1

) to a 

maximum of 310 pCi g−1 (12 Bq g−1). In addition, analyses for Tl-208, Cs-134, and Bi-214 were 

performed and showed detectable quantities, but these concentrations were usually an order of 

magnitude or more lower than those of Cs-137. (Jordan, 1978)

A final report summarizes the residual contamination in the soil and rock in the areas 

around the NPP facility during DF 78 (Jentz, 1978). During the residual contamination survey, 

121 soil samples were collected and analyzed. The Cs-137 contamination in the soil and rock 

samples ranged from a low of less than 1 pCi g−1 ( 0.04 Bq g−1) to a maximum of 400 pCi g−1

(15 Bq g−1). Analysis was also made for Tl-208, Cs-134, Bi-214, Co-60, and Sr-90. The soil 

samples generally contained detectable amounts of Cs-134, Co-60, and Sr-90 but the 

concentrations were usually an order of magnitude or more lower than that of Cs-137. The total 

measured activity for all byproduct radioactive material (fission products) in the soil samples 

was less than 500 pCi g−1 (19 Bq) (Jentz, 1978).

4.6 Removable Radioactivity on Surfaces 

Weekly smear tests for gross beta contamination were made at the galley and crew 

member living quarters (Foster and Jones, 1982). The only reported smear test data available for 

review was for 1972. The available data reported no detectable levels of removable 

contamination for 1972 (NNPU, 1973a).

4.7 Radiation Levels of Radioactive Waste Shipments

Documentation of some radioactive waste shipments was located for 11 shipments from 

DF 70 through DF 72. Records of 21 shipments of RTGs for DF 70 to 72 were also located. The 

average dose rate at one meter from the outside surface of the container for the 11 shipments of 

radioactive waste was 0.008 R h−1 (0.08 mGy h−1). The average dose rate at one meter from the 

outside surface of the container for the 21 RTG shipments was 0.003 R h−1 (0.03 mGy h−1

). 

(NNPUD, 1972)
34

5.

Dose Assessment Methodology and Assumptions

for Support Personnel

When individuals are not monitored for radiation through the use of personal or cohort (a 

group with common characteristics) measurements, the external and internal radiation doses are 

estimated using dose reconstruction methods. The methods include the use of high-sided 

estimates of parameters affecting the dose to personnel along with uncertainty factors that lead to

upper-bound doses. Such reconstructed dose estimates should be considered generic upperbound doses for support personnel whose reported activities and scenarios of exposures fall 

within those that are assumed in the dose assessment described in this report. This section 

describes the methods used for assessing the dose for each external and internal exposure 

pathway identified in Section 3. The parameter values selected for the calculation of doses and 

uncertainty factors are also included. The results of the radiation dose assessment for support 

personnel are presented and discussed in Section 6.

5.1 External Dose Methods and Assumptions

For the support personnel who were located at the McMurdo Station during the 

installation, operation and decommissioning of the NPP, the following potential exposure 

scenarios and pathways are considered in the assessment of external radiation exposures:

1) Exposure to radioactive sources inside the reactor facility while either touring or visiting 

the NPP.

2) Exposure to radioactive sources inside the reactor while near the fence of the NPP, such 

as when driving, hiking, or walking.

3) Exposure to airborne radioactivity by immersion in indoor and outdoor air containing 

radioactive noble gases released from the NPP during operations.

4) Exposure to radioactive packages transported to and from the ship loading area outside 

the NPP facility. 

5) Exposure to radiation sources due to the movement of RTGs.

6) Exposure to radioactive materials due to loading of radioactive waste onto ships.

7) Exposure to soil contaminated with radioactive materials that was spilled on the ground, 

roads, and ship loading areas. 
35

5.1.1. Touring or Visiting the NPP

To estimate the external dose for support personnel who may have toured or visited the 

NPP facility, the film badge results of the winter-over reactor crew members are used (NDC, 

1962–1979). These film badge results are used since they are the only data available to estimate 

the average external dose rates inside the NPP facility for all years of reactor operation. This 

method of estimating an average dose rate inside the NPP for each DF year is very conservative

since the reactor crew results include the larger doses received by some reactor crew members 

who performed activities inside the primary building and in areas below ground level, where 

visitors would not have gone. The parameter values and assumptions shown in Table 11 are used 

to estimate the average dose rates inside the NPP. For the years where dosimetry data were not 

available, the average dose rates are estimated based on the mean dose rate values calculated for 

years where dosimetry data exist even though the average dose rates will be higher than actual 

dose rates due to the change in operations. The average dose rates are calculated using the 

methods presented in Appendix B.

The parameter values and assumptions shown in Table 12 are used to estimate the 

external gamma dose for support personnel touring or visiting the NPP. Table 13 shows the 

estimated average dose rate for each DF year and the basis for the estimate. For the years where 

the only major source of external radiation dose at the NPP was exposure to residual radiation in 

bulk contaminated rock and soil, the average dose rate was based on TLD measurements made 

during a radiological survey in the austral summer of DF 78. The dose estimates for support 

personnel touring or visiting the NPP are calculated using the methods presented in Appendix B. 

The external dose for touring or visiting the NPP is assumed to apply for both winterover and austral summer personnel assigned to the McMurdo Station from 1962 to 1979. 

Neutron doses were not calculated for support personnel since those reported for the reactor crew 

members were less than 5% of the external gamma doses (Johnson, 1971). Therefore, the 

reconstructed conservative external gamma doses are assumed to implicitly include a neutron 

dose estimate along with relevant uncertainty factors.

This exposure scenario does not apply to individuals who made frequent visits to the NPP 

to perform their specific duties. Doses for frequent visitors to the NPP are discussed in 

Section 6.4 below.
36

Table 11. Parameter values and assumptions for estimating average dose rates

inside the NPP facility

Parameter Value Rationale/Reference/Comment

Work day 10 h d−1 Veteran statement

Work week 6 d wk−1 Veteran statement

Full monitoring period 59 wk

Veteran statement

Naval Dosimetry Center 

(ASF, 1962–1979)

Partial monitoring period

13 wk (if October to 

December)

46 wk (if January to 

November)

Naval Dosimetry Center

Austral summer 

monitoring period 26 wk Veteran statementNaval Dosimetry Center

TLD measurement 

results during DF 78

0.038 rem

(0.38 mSv) NUS Radiological Survey

TLD measurement 

duration 76 d NUS Radiological Survey

Average film badge 

results or surrogate value 

for each monitoring year

Values shown in 

Table 5

Dose rates are high-sided by 

including film badge results of 

crew members who worked in high 

dose rate areas that visitors would 

have not accessed

Table 12. Parameter values and assumptions for external dose

while visiting the NPP

Parameter Value Rationale/Reference/Comment

Time inside the NPP 

facility per visit 2 h Reasonable assumption for a tour or other type of non-work related visit

Number of visits per 

year or per duty tour 1 Assumes each person visited once

Dose rate while inside See Table 13 Values calculated using parameter 

values shown in Table 11 
37

Table 13. Estimated average dose rate inside the NPP

DF Average Dose Rate

(rem h−1 (mSv h−1))

Basis for Estimate of Average 

Dose Rate

62 0.0002 (0.002) Full year dosimetry data 

63 0.0002 (0.002) Full year dosimetry data

64 0.0002 (0.002) Full year dosimetry data

65 0.0006 (0.006) Full year dosimetry data

66 0.0007 (0.007) Partial year dosimetry data

67 0.0014 (0.014) Partial year dosimetry data

68 0.0005 (0.005) Partial year dosimetry data

69 0.0013 (0.013) Partial year dosimetry data

70 0.0031 (0.031) Partial year dosimetry data

71 0.0025 (0.025) Partial year dosimetry data

72 0.002 (0.02) Full year dosimetry data

73 0.0005 (0.005) Full year dosimetry data

74 0.0017 (0.017) Austral Summer dosimetry data

75 0.0011 (0.011) Mean of average dose rates from 

all years with film badge data

76 0.0011 (0.011) Mean of average dose rates from 

all years with film badge data

77 0.00002 (0.0002) TLD measurements from austral 

summer of DF 78

78 0.00002 (0.0002) TLD measurements from austral 

summer of DF 78

79 0.00002 (0.0002) TLD measurements from austral 

summer of DF 78

5.1.2. Transits in the Proximity of the NPP

The support personnel were potentially exposed to radiation while moving near the NPP 

fence, e.g., walking or hiking to Observation Hill as reported by McMurdo veterans (Armstrong, 

2012). The external dose for this scenario is estimated by extrapolating the average dose rates 

inside the NPP (Table 14) to the facility fence line. The estimated dose rates at the facility fence 

line account for support personnel being exposed outside of the NPP to sources of radiation from 

the reactor, radioactive components, radioactive waste, and radioactive contamination in the 

effluent discharge areas. The parameter values and assumptions shown in Table 14 are used to 

estimate the potential external dose to support personnel while in the proximity of the NPP fence. 

The doses corresponding to this exposure scenario are calculated using the methods presented in 

Appendix B.
38

Table 14. Parameter values and assumptions for external dose for support personnel 

transiting near the NPP fence line

Parameter Value Rationale/Reference/Comment

Dose rate inside NPP See Table 13 Parameter values above applied to Naval 

Dosimetry Center records

Distance to representative film 

badge location 21 m

Distance from radioactive sources to 

reactor office and break room (Foster and 

Jones, 1982)

Distance to fence line 32 m Closest distance from radioactive sources 

to fence line (Foster and Jones, 1982)

Number of transits (per year) 200 Veteran statement over course of a 1-year 

tour (Armstrong, 2012)

Time of transits 0.25 h Reasonable assumption

This exposure scenario is assigned for both winter-over and austral summer support 

personnel from 1962 to 1979. This is based on the assumption that the majority of the transits 

near the NPP fence line would have occurred in the austral summer months and it overestimates 

the dose for most veterans. 

The use of reactor crew film badge results to estimate doses to support personnel is 

considered conservative. Because of their duty location, the reactor crew members received a 

significant portion of their dose while working in close proximity to the reactor containment 

vessel during maintenance operations rather than while in the reactor office and the break room. 

In addition, the assumed number of transits is based upon veterans’ recollections and should be 

considered high-sided. 

During the VBDR’s annual meeting in San Antonio, Texas, on March 23–24, 2012, a 

McMurdo Station veteran expressed concerns that he was exposed to levels of external radiation 

above the limits for unrestricted areas in November or December of 1972 while visiting the 

aviation supply officer who was assumed to be in the area of the helicopter pad that is located 

approximately 150 yards from the NPP at the base of Observation Hill, even though the location 

is not certain and could have been as far as 500 yards from the reactor in “downtown McMurdo.”

The veteran stated that a radiation survey meter that he had just picked up was reading “3” but he 

did not remember the units. In discussions with the veteran, he identified the Victoreen 440 

ionization chamber as the most likely survey meter that he was using (VBDR, 2012; Fairchild, 

2012b).

A review was made of the October to December 1972 film badge results for the reactor 

crew members to determine the possibility of radiation levels above the limits for unrestricted 

areas in the area of the helicopter landing pad. Several reactor crew members who were issued 

film badges had results indicating that they received external radiation doses of less than 0.2 rem 

(2 mSv) for the last three months of 1972. If the reactor was creating radiation exposure rates of 

0.003 R h−1 (0.03 mGy h−1 corresponding dose rate) at the 150 yard distance of the helicopter 

landing pad, which is in the range of the meter the veteran was using, was from the NPP, then the 

exposure rate in the control room, at 15 yards from the reactor and other areas of the Secondary 
39

Building, would have been approximately 0.3 R h−1 (3 mGy h−1

) using the inverse square law 

(Cember, 1983). Such a dose rate would result in a quarterly film badge dose of over 200 rem 

(2000 mSv) if sustained for the entire three month period for which film badge results are 

available. The film badge results do not support a sustained dose rate of 0.003 R h−1 at 150 yards 

or further from the NPP. Section 4 includes a review of the dosimetry results for reactor crew 

members. Also, a review of the incident reports was made and no HPSC reportable incidents 

were recorded in November or December 1972 to indicate that high radiation alarms were 

activated inside the facility (Foster and Jones, 1982).

5.1.3. Immersion in a Cloud of Noble Gases 

The external dose for support personnel who were potentially exposed to radioactive 

noble gases released from the NPP is estimated by assuming that Ar-41 was emitted at the MPC

during the entire period of NPP operations. The parameter values and assumptions shown in 

Table 15 are used to estimate the dose from this pathway using the methods presented in 

Appendix B. The Ar-41 immersion dose is assumed to include any dose from radioactive xenon 

or krypton gases because the release limit for those gases was set well below their 10 CFR 20 

limits for unrestricted areas and no report was found that mentioned the limit was exceeded 

during any operational year (Foster and Jones, 1982).

The noble gas immersion dose is calculated for both winter-over and austral summer-only 

personnel who were assigned to McMurdo Station from 1962 to 1972 since Ar-41 was released 

only when the reactor was operational. Other than the time at McMurdo Station, the parameter 

values used to estimate the doses for winter-over and austral summer personnel are the same.

The external dose estimate for immersion in radioactive noble gases is considered very 

conservative since the reactor did not emit Ar-41 during the entire time that any individual was at 

McMurdo Station. Historical reports show that the reactor was shut down on multiple occasions 

every year, which eliminated the production of Ar-41 during those periods. In addition, Ar-41 

gas released from the NPP was diluted and dispersed in the air as it moved away from the stack, 

but the effects of dispersion were not included in the dose estimates. The very conservativeness 

of the Ar-41 immersion dose is assumed to include any dose from the release of other noble 

gases. 
40

Table 15. Parameter values and assumptions for external dose from immersion

in radioactive noble gases

Parameter Value Rationale/Reference/Comment

Release concentration 

of Ar-41

1 × 10−8 μCi mL−1

(4 × 10−4

Bq mL−1

)

Maximum permissible emissions of Ar-41 even 

though Ar-41 limit was never exceeded. 

Operational limits for other noble gases exceeded 

once, but release was below federal effluent 

concentration limits, for Xe-133 and Xe-135 

(Foster and Jones, 1982)

Effluent concentration 

limit for Ar-41

1 × 10−8 μCi mL−1

(4 × 10−4

Bq mL−1)

Nuclear Regulatory Commission Standards for 

Protection Against Radiation (USNRC, 2012b)

General public effective 

dose from a full year 

exposure to effluent 

concentration limit 

0.05 rem

(0.5 mSv)

Nuclear Regulatory Commission Standards for 

Protection Against Radiation (USNRC, 2012b)

Fraction of time that 

wind was blowing from 

NPP facility toward 

“downtown” McMurdo

0.5 McMurdo Station wind rose (Lazzara, 2006) and 

NCRP Commentary 8 (NCRP, 1993)

Time at McMurdo 

Station for winter-over 

personnel

1.2 y Veteran statement and dosimetry records (ASF, 

1962–1974)

Time at McMurdo 

Station for austral 

summer personnel

0.5 y Veteran statement and dosimetry records (ASF, 

1962–1974)

5.1.4. Proximity to Radioactive Waste Packages

The external dose accrued by support personnel while in the proximity of radioactive 

waste packages is estimated by assuming that such packages were shipped in accordance with 

IAEA Safety Series Number Six requirements (Foster and Jones, 1982). It is also assumed that 

the average dose rate at one meter from the outside surface of the radioactive packages shipped 

from 1970 to 1972 is representative of the average for all Yellow-III packages. It is assumed that 

the dose rate at one meter for the Yellow-II packages is the maximum permissible dose rate since 

no records of the shipping of Yellow-II packages were available (USNRC, 2011a). The 

parameter values and assumptions shown in Table 16 are used to estimate the external dose from 

being close to the radioactive waste packages. The doses corresponding to this exposure scenario 

are calculated using the methods presented in Appendix B.

The parameter values and methods used to estimate the doses for winter-over and austral 

summer support personnel are the same since radioactive packages were only shipped during the 

austral summer when both sets of personnel overlapped.

The parameter values used in the external dose estimate for being near radioactive waste 

packages are considered high-sided because support personnel did not normally handle 

radioactive waste packages (reactor crew members generally handled the shipping of radioactive 

waste packages). Also, the average radioactive waste package dose rate includes dose rates 
41

involving the shipment of spent fuel (U.S. Navy, 1972a; NNPU, 1976), which did not occur 

every year. 

Table 16. Parameter values and assumptions for external dose from

proximity to radioactive waste packages

Parameter Value Rationale/Reference/Comment

No. of Yellow-III 

shipments 10 Reasonable assumption based on shipping documents 

No. of Yellow-II 

shipments 10 Reasonable assumption based on shipping documents 

Dose rate for 

Yellow-III

0.008 R h−1

at 1 m

(0.08 mGy h−1 

at 1 m)

Average dose rate at one meter for 

Yellow-III radioactive packages 

shipped from McMurdo Station from 

1970 to 1972 (NNPUD, 1972)

Dose rate for 

Yellow-II

0.001 R h−1at 1 m

(0.01 mGy h−1 at 1 m)

Maximum of U.S. Department of 

Transportation danger goods package 

specifications (USNRC, 2011a)

Veteran’s distance 

from package 3 m Reasonable assumption for high-sided average distance 

Time near 

radioactive waste 

packages

1 h Reasonable maximizing assumption, 

not necessarily continuous 

5.1.5. Proximity to Radioisotope Thermal Generators

The external dose that support personnel potentially accrued from being in the proximity 

of RTGs during their shipment is estimated by assuming that packaging and shipping was done 

in accordance with IAEA Safety Series Number Six requirements (NNPU, 1976). It is also 

assumed that the average dose rate at one meter from RTG packages measured for the 1970–

1972 shipments is representative for all years (NNPUD, 1972). The parameter values and 

assumptions shown in Table 17 are used to estimate the external dose from being in the 

proximity to RTGs packages. The doses corresponding to this exposure scenario are calculated 

using the methods presented in Appendix B.

The parameter values used to estimate the dose for winter-over support and the austral 

summer support personnel are the same since RTGs were generally shipped during the austral 

summer when both types of crews overlapped. The parameter values used to estimate the dose 

for exposure to RTGs are considered high-sided since many support personnel were not likely 

exposed to radiation from RTGs.
42

Table 17. Parameter values and assumptions for exposure to external radiation

from shipments of RTGs

Parameter Value Rationale/Reference/Comment

Number of RTG 

shipments 1

Reasonable assumption based on shipping 

documents during decommissioning 

(NNPUD, 1972)

Dose rates for RTG 

shipments

0.003 R h−1 at 1 m

(0.03 mGy h−1 at 1 m)

Average dose rate measured at one meter 

for RTG packages shipped from 

McMurdo Station from 1970 to 1972 

(NNPUD, 1972)

Veteran’s distance 

from RTG shipment 3 m Reasonable high-sided assumption for average distance

Time near shipments 1 h Reasonable high-sided assumption, not 

necessarily continuous

5.1.6. Contaminated Soil on the Ground

The support personnel at McMurdo Station potentially accrued an external dose from 

exposure to contaminated soil deposited on the ground from spills that occurred during transport 

to the loading area or during ship loading. This dose is estimated using the maximum measured 

activity level of any sample of crushed rock and soil that was shipped in bulk. The external dose 

from residual soil contamination, after spills were cleaned up, is estimated using the maximum 

allowed concentration for residual Cs-137 contamination. The parameter values and assumptions 

shown in Table 18 are used to estimate the dose from contaminated soil on the ground. The doses 

corresponding to this exposure scenario are calculated using the methods presented in 

Appendix B.

Table 18. Parameter values and assumption for the external dose from

contaminated soil deposited on the ground during decommissioning

Parameter Value Rationale/Reference/Comment

Contamination level of 

spilled soil

500 pCi g−1

(19 Bq g−1) of Cs-137

Maximum contamination 

measured in all soil samples in 

1978 (Jentz, 1978)

Time exposed to 

contaminated soil spills

7 d

10 h d−1

Decommissioning 

Health Physics Logs 

(NNPU, 1976)

Residual contamination 

level after soil spills 

were cleaned

10 pCi g−1 (0.4 Bq g−1)

of Cs-137

Maximum permissible residual 

contamination level of remaining 

soil (NNPU, 1973b)
43

Parameter Value Rationale/Reference/Comment

Time exposed to residual 

contamination

420 d

4 h d−1

Reasonable assumption for highsided average given that personnel 

spent most of their time indoors. 

Depth of contamination 0.01 m Assumed spread over an infinite 

plane 

Dose coefficient for 

exposure to soil 

contaminated with Cs137 to a depth of 1 cm. 

3.2 × 10−13

Sv per Bq m−3

Environmental Protection Agency 

Federal Guidance Report 12 

(Eckerman and Ryman, 1993)

Thickness of soil 0.01 m Reasonable assumption

Density of soil 1.3 × 106 g m−3 NTPR SOP (DTRA, 2010, SM 

ID01)

Other than the length of time spent at McMurdo Station, the parameter values used to 

estimate the doses for winter-over and austral summer personnel are the same. The parameter 

values used to estimate the external dose for exposure to contaminated soil on the ground are 

considered high-sided because the maximum measured activity concentration in soil is used to 

calculate the dose. In addition, the assumption that a layer 1-centimeter thick over an infinite 

plane was formed from the spilled soil and that the spilled soil remained on the ground for seven 

days is highly conservative.

5.1.7. Ship Loading of Radioactive Waste

The external dose for support personnel potentially exposed to radiation during the ship 

loading of radioactive waste is estimated using self-reading pocket dosimeter results that were 

recorded in February 1976. The parameter values and assumptions shown in Table 19 are used to 

estimate the external dose during ship loading. The doses corresponding to this exposure 

scenario are calculated using the methods presented in Appendix B.

Table 19. Parameter values and assumptions for external dose

for ship loading of radioactive waste

Parameter Value Rationale/Reference/Comment

Time exposed Two days a year for each 

year (1962–1979) Operation Logs (NNPU, 1977a)

Loading dose for two 

days 

0.035 rem

(0.35 mSv)

Mean self-reading dosimeter results 

from radiation work permit in 

Operation Logs (NNPU, 1977a)

The external dose from ship loading of radioactive waste is calculated for all winter-over 

and austral summer personnel since the ship loading occurred during the austral summer over 1 
44

or 2 days and all veterans are assumed to have been present and received the same dose. The 

parameter values and method used to estimate the external dose from exposure to radioactive 

waste during ship loading are considered high-sided because the self-reading dosimeter results 

were from the year when a large quantity of radioactive waste was shipped (Foster and Jones, 

1982). Furthermore, each support personnel member is assumed to have carried out this activity 

over a 2-day period.

5.2 Internal Doses–Methods and Assumptions

The support personnel at McMurdo could have inhaled or ingested radioactive materials 

released from the NPP or suspended during soil excavation, and transport and loading on ships. 

For the support personnel who were located at the McMurdo Station during the installation, 

operation and decommissioning of the NPP, the following potential exposure scenarios and 

pathways are considered in the assessment of radiation doses from internally-deposited 

radioactive materials:

1) Inhalation of beta-gamma emitting airborne radioactive materials in emissions of fission 

and activation products released through the NPP stack or in worldwide fallout.

2) Inhalation of suspended contaminated soil and dust during decommissioning of the NPP.

3) Inhalation of suspended contaminated soil and dust that was spilled on the ground, roads 

and ship loading areas during transport and loading. Also, inhalation of suspended 

residual contamination after spill cleanup. 

4) Ingestion of drinking water contaminated with tritium either from effluent release or from 

the use of steam from the NPP.

5) Ingestion of other long-lived beta emitters in drinking water, other than tritium, that could 

have been in the water from effluent releases from the NPP. 

6) Incidental ingestion of contaminated soil and dust that was spilled during the ship-loading 

process or was residual after spill cleanup. 

5.2.1. Inhalation of Airborne Radioactive Materials 

The internal dose for support personnel from the inhalation of airborne radioactive 

materials released from the NPP stack and worldwide fallout is estimated using the gross beta air 

activity concentration monitoring data provided in Table 7. The parameter values and exposure 

scenario assumptions shown in Table 20 are used to estimate the internal dose from the 

inhalation of airborne radioactive materials calculated using the methods presented in 

Appendix B.

The discussion of airborne activity concentration data included in Foster and 

Jones (1982) reveals that radioactive materials in air at McMurdo is consistent and strongly 

correlated with worldwide fallout from past and on-going nuclear tests that were carried out by 

the U.S., the Soviet Union, France and China during the years 1961 to 1970. To account for the 

presence of emissions from the McMurdo NPP, it is assumed that the measured activity in air 

samples is due both to fission products from worldwide fallout and releases from the NPP. 
45

Examination of the air concentration chart in Table 7 shows that the activity in air sample 

measurements follow a baseline trend with increases that peak following fallout events as 

explained in Foster and Jones (1982).

In this dose assessment, the baseline long-lived beta activity from airborne radioactive 

materials is associated with releases from the NPP and the remainder of the measured air sample 

activity is assumed to be from worldwide fallout. To estimate the baseline activity threshold that 

is due to NPP operation, the mean value of the average monthly activity for periods of low 

worldwide fallout is used. Three periods of low worldwide fallout are identified and are shown in 

in Table 7 and Figure 7. In the months where the threshold is exceeded, the threshold value is 

assigned to NPP operation and the remainder is assigned to worldwide fallout. For months where 

airborne activity is lower than or equal to the threshold, the total value of activity is assigned to 

the NPP. The baseline activity in air samples is assumed to all be from I-131 aerosols, which 

would result in higher doses. This is a reasonable assumption based on nuclear power plant 

operation data, e.g., Calvert Cliffs Nuclear Power Plant, LLC (2009). 

Furthermore, because only the activity from particulate radioactive materials is measured 

in air samples, unmeasured activity in gaseous iodines and hypoiodous acid is accounted for by 

adding corresponding activity that is estimated using a non-particulate to particle ratio of 9. This 

means that 9 times the activity from the air sample measurement, which is presumed to be due to 

NPP air emissions, is added to account for releases of non-particulate iodines. The ratio of 9 is 

approximated from data published by Chandrasekaran et al., (1985) and reflects that most 

particulates are absorbed by ventilation and air effluent filters (Martin, 1964). It is further

indicated that, of the iodines released in effluents from pressurized water reactors, about 

10 percent of the activity is from particulates and 90 percent is from non-particulates. 

As for the component of activity from airborne radioactive materials associated with 

worldwide fallout, given that no radionuclide analysis is available, the dose is calculated by 

assuming that all activity from worldwide fallout is from Cs-137. This assumption maximizes the 

doses from this source and pathway.

The inhalation dose for airborne radioactive materials is calculated for all support 

personnel at McMurdo Station for all DF years. The inhalation dose for airborne radioactive 

materials is different for winter-over and austral summer personnel due to differences in length 

of presence at McMurdo Station.

The parameter values and methodology used to estimate the inhalation dose are 

considered high-sided in part because background radioactivity was not subtracted from the filter 

activity concentration results (Foster and Jones, 1982). Furthermore, the activity attributed to 

reactor operations was assumed to be from I-131because the thyroid dose coefficient for I-131 is 

higher than the dose coefficients for other long-lived beta emitters that could have been released 

in the stack emissions from the NPP. Also, the assumption that all activity in worldwide fallout is 

from Cs-137 high-sides the internal dose to organs, Cs-137 being representative of long-lived 

fission product in worldwide fallout. Finally, the infiltration factors, which are the ratio of inside 

concentration to outside concentration, used for this dose calculation are selected to be highsided.
46

Table 20. Parameter values and assumption for the internal dose from inhalation 

of airborne radioactive materials from stack emissions or worldwide fallout

Parameter Value Rationale/Reference/Comment

Baseline threshold 

of air activity due to 

plant operations

0.04 pCi m−3

(1.5 Bq m−3

)

Average air activity for all months 

during periods of low worldwide fallout

Air activity 

concentration for 

worldwide fallout 

during plant 

operations

Monthly average values

from Table 7 less the 

baseline threshold

(pCi m−3 (Bq m−3

))

Values available from Jan 1962 to Dec 

1972 (Foster and Jones, 1982) 

Air activity 

concentration for 

worldwide fallout 

after plant 

operations ceased

0.04 pCi m−3

(1.5 Bq m−3

)

Average of monthly air monitoring data 

for 1972 less baseline threshold

Air activity 

concentration due 

to air emissions

from NPP during 

plant operation

The smaller of the monthly 

average from Table 7 or 

the baseline threshold

(pCi m−3 (Bq m−3))

Values available from Jan 1962 to 

Dec 1972 (Foster and Jones, 1982)

Fraction of time 

outside 0.0167

Reasonable high-sided assumption of 

four hours per day as an average for an 

entire year given that personnel spent 

most of their time indoors

Breathing rate 0.9 m

3

h

−1 Value for adult averaged over 24 hours 

(ICRP, 1994a) 

Ratio of gaseous 

iodine to aerosol 

iodine

9 NUREG-0017 (Chandrasekaran et al., 

1985) 

Infiltration factor 

for aerosols 0.5 Zhu et al., 2005; Cassata et al., 2012

Infiltration factor 

for gas 1 Zhu et al., 2005; Cassata et al., 2012

Time exposure to 

airborne radioactive 

material (winterover)

24 h d−1 for 420 d Reasonable assumption

Time exposure to 

airborne radioactive 

material (austral 

summer)

24 h d−1 for 180 d Reasonable assumption
47

Parameter Value Rationale/Reference/Comment

Dose coefficients 

Depends on organ

5-micrometer (μm) particle 

size, I-131 and Cs-137 for 

aerosol and elemental 

iodine for gaseous form 

(rem Ci−1 (Sv Bq−1))

ICRP Publication 68 (ICRP, 1994b)

Results in highest doses for most organs

5.2.2. Inhalation of Suspended Contaminated Soil during Decommissioning

The internal dose for support personnel from their inhalation of suspended contaminated 

soil during decommissioning is estimated using air sampling activity concentration data taken 

during the decommissioning process (NNPU, 1976; NNPU, 1977b). The parameter values and 

assumptions shown in Table 21 are used to estimate the internal dose from the inhalation of 

suspended contaminated soil during decommissioning. This internal dose is calculated using the 

methods presented in Appendix B.

Table 21. Parameter values and assumption for the internal dose from inhalation of

suspended contaminated soil during decommissioning

Parameter Value Rationale/Reference/Comment

Air activity 

concentration

Ci m−3 (Bq m−3

)

(Cs-137) derived from 

Table 8

Used highest value from air sampling 

results reported in the 

Decommissioning Health Physics 

Logs (NNPU, 1976; NNPU, 1977b)

Time outside

10 h d−1

6 d wk−1 for

6 months (summer tour)

Veteran’s statements about workday 

and activities carried out only during 

austral summers

Breathing rate 1.2 m3 h−1 ICRP Publication 66 (ICRP, 1994a)

Dose coefficients

Depends on organ

(5-μm particle size, 

all Cs-137) (rem Ci−1

(Sv Bq−1))

- ICRP Publication 68 (ICRP, 1994b)

- Results in highest doses for most 

organs

The inhalation dose from suspended contaminated soil is calculated for all support 

personnel who were at McMurdo Station during decommissioning of the NPP for DF 75 to 79. 

The methodology should also be used for reactor crew members who were at McMurdo Station 

during DF 75 to 79.

The parameter values and methodology used to estimate the internal doses from the 

inhalation of suspended contaminated soil during decommissioning are considered high-sided 

because the highest air sampling result is used. Also the time spent outside during 

decommissioning was assumed to be the maximum possible length. 
48

5.2.3. Inhalation of Suspended Contaminated Soil Spilled on the Ground, 

Roads and Ship Loading Areas

The internal dose that support personnel potentially accrued from the inhalation of 

suspended contaminated soil spilled during bulk soil transport and ship loading operations is 

estimated using the maximum measured soil concentration levels of Cs-137. The internal dose 

from inhalation of suspended residual contaminated soil is estimated using the maximum soil 

concentration levels of Cs-137 that may have remained after cleanup. The parameter values and 

exposure scenario assumptions shown in Table 22 are used to estimate the internal dose from the 

inhalation of suspended soil using the methods presented in Appendix B.

Table 22. Parameter values and assumptions for the internal dose from inhalation of 

suspended contaminated soil that was spilled on the ground, roads and ship loading areas 

Parameter Value Rationale/Reference/Comment

Activity concentration of 

contaminated soil

500 pCi g−1

(20 Bq g−1

) of Cs-137

Maximum contamination measured 

in all soil samples in 1978 

(Jentz, 1978)

Time exposed to suspended 

contaminated soil

7 d

10 h d−1

Health Physics Logs (NNPU, 

1977b) and veteran’s statements 

about workday

Residual contamination 

level after cleanup

10 pCi g−1 (0.4 Bq g−1) of 

Cs-137

Decommissioning Plan and Health 

Physics Logs (NNPU, 1976; 

NNPU, 1977b)

Time exposed to residual 

contamination (winter-over) 4 h d−1 for 420 days Reasonable assumption for residual contamination

Time exposed to residual 

contamination (austral 

summer)

4 h d−1 for 180 days Reasonable assumption for residual 

contamination

Suspension factor 1 × 10−5 m−1 NTPR SOP (DTRA, 2010, 

SM ID01)

Depth of contamination 1 cm NTPR SOP (DTRA, 2010, 

SM ID01)

Dose coefficients

Depends on organ

(5-μm particle size, 

all Cs-137) 

(rem Ci−1 (Sv Bq−1))

ICRP Publication 68 (ICRP, 1994b)

Results in highest doses for most 

organs

The inhalation dose from spilled contaminated soil is calculated for all support personnel 

who were at McMurdo Station during the decommissioning of the NPP from DF 75 to 79. The 

lengths of time at McMurdo Station are different for winter-over and austral summer personnel. 
49

This pathway is also applicable to reactor decommissioning crew members who were at 

McMurdo Station during DF 75 to 79.

The parameter values and methods used to estimate the internal doses from inhalation of 

suspended contaminated soil are considered high-sided because most of the spilled soil would 

have contained less than 500 pCi g−1 (10 Bq g−1). Also, any minor spills during transportation of 

the bulk soil would not have left contaminated soil to a depth of one centimeter everywhere a 

person walked for the length of time assumed.

5.2.4. Ingestion of Tritium in Drinking Water

The internal dose that support personnel potentially accrued from the ingestion of tritium 

in the drinking water is estimated using concentration measurements of the tritium in drinking 

water concentration measurements from 1967 to 1972 or the average monthly concentration for 

all months before 1967 or after 1972 (Table 10). The parameter values and assumptions shown in 

Table 23 are used to estimate the internal dose from the ingestion of tritium in drinking water. 

This dose is calculated using with the methods presented in Appendix B.

Table 23. Parameter values and assumptions for the internal dose from ingestion

of tritium in drinking water

Parameter Value Rationale/Reference/Comment

Tritium 

concentration

in drinking water

Monthly average of 

weekly samples

pCi L−1 (Bq L−1)

Data available from Mar 1967 to 

Nov 1972 (Foster and Jones, 1982)

Tritium 

concentration in 

drinking water 

during unmonitored 

months

1 × 105 pCi L−1

(3,900 Bq L−1)

Average of all tritium monitoring 

results from May 1967 to 

November 1972 (Foster and Jones, 

1982).

Water intake

6 L d−1

420 d (winter-over)

180 d (austral summer)

Mean value for an adult is 1.5 L d−1

(USEPA, 2011). Increased water 

intake in the extreme cold and dry 

environmental conditions in 

Antarctica (Montain and Matthew, 

2010)

Dose coefficients Depends on organ

(Sv Bq−1)

ICRP Publication 68 (ICRP, 

1994b)

The winter-over and austral summer personnel would accrue different doses from the 

ingestion of tritium in drinking water due to the length of presence at McMurdo Station. This 

dose pathway is also applicable to reactor crew members because drinking water consumption is 

a pathway for all personnel.

The parameter values and methods used to estimate the internal doses for the ingestion of 

tritium in drinking water are considered high-sided in part because background radiation is 

included in the tritium activity concentration; the natural background levels for tritium are 
50

approximately 50 to 100 pCi L−1 (2 to 4 Bq L−1

) (ASTSWMO, 2009). In addition, the water 

intake value is conservative to take into account the possibility of increased water consumption 

caused by the cold and dry environmental conditions in Antarctica.

5.2.5. Ingestion of Long-Lived Beta Emitters in Drinking Water other than 

Tritium

The internal dose that support personnel potentially accrued from the ingestion of longlived beta emitters in drinking water, other than tritium, is estimated using the results of the 

drinking water monitoring activity concentration data from 1962 to 1972 along with surrogate 

data for 1973 to 1979 (Table 9). The parameter values and assumptions given in Table 24 are 

used to estimate the internal dose from the ingestion of drinking water using the methods 

presented in Appendix B.

Table 24. Parameter values and assumptions for the internal dose from ingestion

of long-lived beta emitters in drinking water other than tritium

Parameter Value Rationale/Reference/Comment

Beta activity 

concentration in 

drinking water

Monthly average

(Bq L−1)

Values from January 1962 to December 1972

(Foster and Jones, 1982)

Assumed December 1972 drinking water 

concentration for months between January 1973 

to March 1979

Water intake 6 L d−1

Mean value for an adult is 1.5 L d−1 (USEPA, 

2011). Increased water intake in the extreme 

cold and dry environmental conditions in 

Antarctica (Montain and Matthew, 2010)

Dose coefficients

Depends on organ

(all Cs-137) 

(rem Ci−1(Sv Bq−1))

ICRP Publication 68 (ICRP, 1994b)

Results in highest doses for most organs

The winter-over and austral summer personnel would have accrued different doses from 

the ingestion of drinking water due to the length of presence at McMurdo Station. This exposure

pathway is also applicable to reactor crew members because the drinking water consumption is a 

pathway for all personnel. 

The parameter values and methods used to estimate the internal doses for ingestion of 

beta emitters in drinking water are considered high-sided in part because background radiation is 

included in the water activity concentration data. Also, the dose estimate assumes all 

radioactivity is from Cs-137 instead of assuming that some of the activity is due to beta emitters 

with lower dose coefficients. In addition, the water intake value is conservative to take into 

account the possibility of increased water consumption in the cold and dry environmental 

conditions in Antarctica. 
51

5.2.6. Incidental Ingestion of Contaminated Soil and Dust

The internal doses that support personnel potentially accrued from the incidental 

ingestion of contaminated soil and dust during decommissioning is estimated using the results of 

soil monitoring. The parameter values and assumptions provided in Table 25 are used to estimate 

the internal dose from the incidental ingestion of soil and dust. This dose is calculated using the 

methods presented in Appendix B.

Table 25. Parameter values and assumptions for the internal dose from

incidental ingestion of contaminated soil and dust

Parameter Value Rationale/Reference/Comment

Concentration of 

contaminants in soil

500 pCi g−1

(20 Bq g−1) 

of Cs-137

Maximum contamination measured in all 

soil samples in 1978 (Jentz, 1978)

Ingestion rate 0.05 g d−1

Value from USEPA (2011) and military 

guidance. Also based on NTPR standard 

operating procedure ID01 (DTRA, 2010, 

SM ID01)

Days ingesting 

contaminated soil 7 d Decommissioning Health Physics Logs (NNPU, 1977b)

Residual soil 

contamination level 

after spills were cleaned

10 pCi g−1

(0.4 Bq g−1

)

of Cs-137

Maximum permissible residual 

concentration in soil (Decommissioning 

Plan (NNPU, 1973b) and Health Physics 

Logs (NNPU, 1977b) )

Days ingesting soil with 

residual contamination 

(winter-over personnel)

420 d Reasonable assumption based on 14 

month tour

Days ingesting soil with 

residual contamination 

(austral summer 

personnel)

180 d Reasonable assumption based on 6 month 

tour

Dose coefficients Depends on organ

(all Cs-137)

ICRP Publication 68 (ICRP, 1994b)

Results in highest doses for most organs

The dose from the incidental ingestion of contaminated soil and dust is calculated for all 

support personnel who were at McMurdo Station during the decommissioning of the NPP for 

DF 75 to 79. The parameter values used for the doses for the winter-over and austral summer 

personnel are different since the winter-over personnel would have been exposed to residual soil 

contamination for a longer time. The dose pathway is also applicable to reactor crew members 

who were at the McMurdo Station during DF 75 to 79.

The parameter values and methods used to estimate the internal doses from the incidental 

ingestion of contaminated soil and dust are considered high-sided because spilled contaminated 

soil would not generally have remained on the ground for seven days after the spill and before 
52

cleanup. In addition, not all of the spilled contaminated soil would have contained the maximum 

measured level of contamination.

5.3 Uncertainties and Upper-Bound Doses

Several sources of uncertainty in estimating external and internal doses to veterans in this 

assessment are attributed to imperfection in measuring instruments, procedural errors, and data 

recording and processing uncertainties, among others. The following is a non-comprehensive list 

of potential sources of uncertainties: 

• Instrument precision, operator measurement and recording errors.

• Uncertainties due to data processing tools.

• Spatial variability when only average values are reported or a few measurements are 

taken.

• Variability in the exposure times.

• Uncertainties in breathing rates and water consumption rates.

• Uncertainty in dose coefficients for iodine and cesium.

• Uncertainties in the isotopic mix of radioactive materials.

5.3.1. External Doses

Following the procedures used for NTPR dose calculations, an uncertainty factor of 3 is 

assigned to each external dose component calculated for the McMurdo Station support personnel 

(Kocher, 2009; DTRA, 2010, Standard Method ED02). Also, the components of the external 

dose are assumed to be uncorrelated i.e., they vary independent of each other. Therefore, to 

determine an upper-bound external dose, the uncertainties of the external dose components are 

combined in quadrature (DTRA, 2010, Standard Method UA01) as described in Appendix B. 

Using this uncertainty approach, the upper-bound dose is considered to exceed the 95th percentile 

dose determined from a hypothetical distribution of film badge doses for individuals exposed to 

the same sources of radiation. In addition, the uncertainty factor accounts for relatively small 

doses that are less than a few percent of the overall external dose, e.g., dose from proximity to 

radioactive sample packages shipped as White-I.

5.3.2. Internal Doses

Following the procedures used for NTPR dose calculations, an uncertainty factor of 10 is 

assigned to each internal dose calculated for the McMurdo Station support personnel. The 

uncertainties of the internal dose are assumed to be correlated. The upper-bounds of each 

component of the internal dose are summed to estimate the upper-bound internal dose for either 

the committed effective dose or the organ dose as described in Appendix B. The upper-bound 

internal dose is considered to exceed the 95th percentile dose determined from a hypothetical 

distribution of doses for individuals estimated from internal monitoring measurements 
53

(NAS/NRC, 2003; DTRA, 2010, Standard Method ID01). In addition, the uncertainty factor for 

internal doses accounts for relatively small doses that are less than a few percent of the overall 

internal dose, e.g., occasional consumption of locally-caught seafood.
54

6.

Support Personnel Dose Assessment

Results and Discussion

6.1 External Dose 

The external doses and upper-bounds for winter-over and austral summer support 

personnel who were assigned to McMurdo Station during DF 62 to 79 are presented in Table 26

and Table 27, respectively. The results include the dose estimates for each component of the 

external dose for each year or austral summer, the total external dose from all components, and 

the corresponding upper-bound doses. The external dose and upper-bounds are the same for all 

organs.

6.2 Internal Dose 

The internal doses and upper-bounds for winter-over and austral summer support 

personnel who were assigned to McMurdo Station during DF 62 to 79 are provided in Table 28

and Table 29, respectively. The results include the 50-year committed effective dose and the 

50-year committed equivalent dose to the thyroid and the red bone marrow for each pathway and 

each year or austral summer, the total internal dose from all pathways, and the corresponding 

upper-bound internal dose. The thyroid was chosen as an organ of interest because of the 

assumption that much of the airborne radioactive materials were made of iodines. The red bone 

marrow was chosen due to leukemia being the most common radiation induced cancer. The 50-

year committed effective dose is the sum of the products of the committed equivalent dose 

delivered over a 50-year period after exposure and the appropriate weighting factor for all organs 

and tissues (ICRP, 1991).

6.3 Discussion

A summary of the upper-bound external and internal dose results and the corresponding 

total doses for McMurdo Station support personnel during DF 62 to 79 is given in Table 30.

The highest external upper-bound dose of 0.4 rem (4 mSv) is calculated for DF 70, which 

was the year when there was a core change during the middle of winter that lead to higher 

average film badge doses and resulted in the highest dose component for transiting near the NPP. 

The lowest external upper-bound dose is 0.2 rem (2 mSv) and was estimated for the 

decommissioning period when the largest sources of external exposure were from the movement 

of radioactive waste.

The highest upper-bound internal dose for the organs considered is 0.4 rem (4 mSv) and 

is for DF 67, which is the year with the highest tritium doses. The lowest upper-bound internal 
55

dose for the organs considered is 0.02 rem (0.2 mSv) and is calculated for DF 77, 78, and 79, 

which is the period during decommissioning after the reactor core had been returned to the U.S. .

The highest total dose for the organs considered is 0.6 rem (6 mSv)and corresponds to

DF 67. The lowest total upper-bound dose for any organ considered is 0.2 rem (2 mSv) and is for 

DF 77, 78 and 79 during decommissioning after the reactor core had been returned to the U.S.

6.4 Other Exposure Scenarios

All estimated doses and upper-bounds reported here are generic cohort doses that can be 

assigned to any McMurdo Station support personnel veteran who had activities and exposures 

covered in this assessment for the period between 1962 and 1979. However, some support 

personnel may have been exposed to additional or distinctly different sources of radiation that 

were associated with activities and functions that did not fall into the scenarios and pathways 

assessed in this report. For these veterans, proposed guidelines are provided in Section 8 for 

performing individualized radiation dose assessments that would include exposure scenarios that 

were not evaluated in this report. 
56

Table 26. External doses and upper-bounds for winter-over support personnel

External Dose (rem*)

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Plant tour/visit – – – 0.002 0.002 0.003 0.001 0.003 0.007 0.005 0.004 0.001 0.004 0.003 0.003 – – –

Transit near NPP 0.005 0.005 0.005 0.02 0.02 0.04 0.02 0.03 0.07 0.06 0.05 0.02 0.04 0.03 0.03 – – –

Noble gas 

Immersion 0.02 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 0.03 – – – – – –

Spilled 

contaminated soil – – – – – – – – – – – – – – 0.004 0.004 0.004 0.004

Radioactive waste 

package 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01

RTGs – – – – – – – – – – – – – – – – – –

Ship loading 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04

Total† 0.07 0.08 0.08 0.09 0.1 0.2 0.09 0.2 0.2 0.2 0.2 0.09 0.09 0.08 0.08 0.05 0.05 0.05

Upper-Bound 0.2 0.2 0.2 0.2 0.2 0.3 0.2 0.3 0.4 0.3 0.3 0.2 0.2 0.2 0.2 0.2 0.2 0.2

*

1 rem = 10 mSv, "–" means <0.001 rem

†

Total doses may not match the sum of pathway dose components because all results are rounded up to one significant digit.

Table 27. External doses and upper-bounds for austral summer support personnel

External dose (rem*)

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Plant tour/visit – – – 0.002 0.002 0.003 0.001 0.003 0.007 0.005 0.004 0.001 0.004 0.003 0.003 – – –

Transit near NPP 0.005 0.005 0.005 0.02 0.02 0.04 0.02 0.03 0.07 0.06 0.05 0.02 0.04 0.03 0.03 – – –

Noble gas 

Immersion 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 0.02 – – – – – –

Spilled 

contaminated soil – – – – – – – – – – – – – 0.003 0.003 0.003 0.003 0.003

Radioactive 

waste package 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01 0.01

RTG movement – – – – – – – – – – – – – – – – – –

Ship loading 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04

Total† 0.07 0.07 0.07 0.08 0.08 0.1 0.07 0.09 0.2 0.2 0.2 0.07 0.09 0.08 0.08 0.05 0.05 0.05

Upper-Bound 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.3 0.3 0.3 0.2 0.2 0.2 0.2 0.2 0.2 0.2

*

1 rem = 10 mSv, "–" means <0.001 rem

†

Total doses may not match the sum of pathway dose components because all results are rounded up to one significant digit.
57

Table 28. Internal doses and upper-bounds for winter-over support personnel

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Effective (Whole-Body) Dose (rem*)

Inhalation of 

airborne 

radioactivity

– – – – – – – – – – – – – – – – – –

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion 0.002 0.002 0.002 0.003 0.002 0.002 – 0.004 0.002 – 0.002 0.001 0.002 0.002 0.002 0.002 0.002 0.002

Ingestion of 

water (tritium) 0.003 0.005 0.005 0.005 0.005 0.03 0.009 0.007 0.005 0.002 0.002 0.004 0.005 0.005 0.005 0.005 0.005 0.005

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.004 0.006 0.006 0.007 0.006 0.04 0.01 0.02 0.007 0.003 0.003 0.005 0.006 0.007 0.007 0.007 0.007 0.007

Upper-Bound 0.04 0.06 0.06 0.07 0.06 0.4 0.1 0.2 0.07 0.03 0.03 0.05 0.06 0.07 0.07 0.07 0.07 0.07

Thyroid (rem*)

Inhalation of 

airborne 

radioactivity

0.004 0.005 0.005 0.005 0.004 0.005 0.005 0.005 0.005 0.005 0.005 – – – – – 0.004 0.005

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion 0.002 0.002 0.002 0.003 0.002 0.002 – 0.004 0.002 – 0.002 0.001 0.002 0.002 0.002 0.002 0.002 0.002

Ingestion of 

water (tritium) 0.003 0.004 0.004 0.004 0.004 0.03 0.009 0.007 0.005 0.002 0.002 0.004 0.004 0.004 0.004 0.004 0.003 0.004

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.008 0.02 0.02 0.02 0.009 0.04 0.02 0.02 0.02 0.007 0.008 0.006 0.006 0.006 0.006 0.006 0.008 0.02

Upper-Bound 0.08 0.2 0.2 0.2 0.09 0.4 0.2 0.2 0.2 0.07 0.08 0.06 0.06 0.06 0.06 0.06 0.08 0.2

Red Bone Marrow (rem*)

Inhalation of 

airborne 

radioactivity

– – – – – – – – – – – – – – – – – –

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion 0.002 0.002 0.002 0.003 0.002 0.002 – 0.004 0.002 – 0.002 0.001 0.002 0.002 0.002 0.002 0.002 0.002

Ingestion of 

water (tritium) 0.003 0.004 0.004 0.004 0.004 0.03 0.009 0.007 0.005 0.002 0.002 0.004 0.004 0.004 0.004 0.004 0.003 0.004
58

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.004 0.006 0.006 0.007 0.006 0.03 0.01 0.02 0.006 0.002 0.003 0.005 0.006 0.006 0.006 0.006 0.004 0.006

Upper-Bound 0.04 0.06 0.06 0.07 0.06 0.3 0.1 0.2 0.06 0.02 0.03 0.05 0.06 0.06 0.06 0.06 0.04 0.06

*

1 rem = 10 mSv, "–" means <0.001 rem

†

Total doses may not match the sum of pathway dose components because all results are rounded up to one significant digit.

Table 29. Internal doses and upper-bounds for austral summer support personnel

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Effective (Whole-Body) Dose (rem*)

Inhalation of 

airborne 

radioactivity

– – – – – – – – – – – – – – – – – –

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion – – – 0.002 – – – 0.001 0.001 – – – – – – – – –

Ingestion of 

water (tritium) 0.002 0.002 0.002 0.002 0.002 0.02 0.006 0.006 0.003 – – 0.002 0.002 0.002 0.002 0.002 0.002 0.002

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.003 0.003 0.003 0.003 0.003 0.02 0.006 0.007 0.004 0.002 – 0.002 0.003 0.004 0.004 0.004 0.004 0.004

Upper-Bound 0.03 0.03 0.03 0.03 0.03 0.2 0.06 0.07 0.04 0.02 0.009 0.02 0.03 0.04 0.04 0.04 0.04 0.04

Thyroid (rem*)

Inhalation of 

airborne 

radioactivity

0.002 0.003 0.003 0.003 0.002 0.003 0.003 0.003 0.002 0.002 0.003 – – – – – – 0.002

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion – – – 0.002 – – – 0.001 0.001 – – – – – – – – –

Ingestion of 

water (tritium) 0.002 0.002 0.002 0.002 0.002 0.02 0.006 0.005 0.003 – – 0.002 0.002 0.002 0.002 0.002 0.002 0.002

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.004 0.005 0.005 0.005 0.004 0.02 0.008 0.009 0.005 0.003 0.003 0.003 0.003 0.004 0.004 0.004 0.004 0.004

Upper-Bound 0.04 0.05 0.05 0.05 0.04 0.2 0.08 0.09 0.05 0.03 0.03 0.03 0.03 0.04 0.04 0.04 0.04 0.04
59

Pathway DF62 DF63 DF64 DF65 DF66 DF67 DF68 DF69 DF70 DF71 DF72 DF73 DF74 DF75 DF76 DF77 DF78 DF79

Red Bone Marrow (rem*)

Inhalation of 

airborne 

radioactivity

– – – – – – – – – – – – – – – – – –

Inhalation of 

suspended soil – – – – – – – – – – – – – – – – – –

Inhalation of 

spilled soil – – – – – – – – – – – – – – – – – –

Water ingestion – – – 0.002 – – – 0.001 0.001 – – – – – – – – –

Ingestion of 

water (tritium) 0.002 0.002 0.002 0.002 0.002 0.02 0.006 0.005 0.003 – – 0.002 0.002 0.002 0.002 0.002 0.002 0.002

Soil ingestion – – – – – – – – – – – – – – – – – –

Total† 0.003 0.003 0.003 0.003 0.003 0.02 0.006 0.006 0.004 0.002 – 0.002 0.003 0.004 0.004 0.004 0.004 0.003

Upper-Bound 0.03 0.03 0.03 0.03 0.03 0.2 0.06 0.06 0.04 0.02 0.008 0.02 0.03 0.04 0.04 0.04 0.04 0.03

*

1 rem = 10 mSv, "–" means <0.001 rem

†

Total doses may not match the sum of pathway dose components because all results are rounded up to one significant digit.
60

Table 30. Upper-bound external, internal, and total doses for McMurdo Station 

support personnel during Deep Freeze years 1962 to 1979

Reported 

Period

Effective Dose

or

Organ

Range of

Upper-Bound Doses

(rem (mSv))

Range of

Upper-Bound Total Doses

(rem (mSv))*

External Internal

Winter-over

Effective dose 0.2–0.4

(2–4)

0.03–0.4

(0.3–4) 

0.2–0.6

(2–6)

Thyroid 0.2–0.4

(2–4)

0.06–0.4

(0.6–4)

0.2–0.6

(2–6)

Red bone marrow 0.2–0.4

(2–4)

0.02–0.3

(0.2–3)

0.2–0.6

(2–6)

Austral 

summer

Effective dose 0.2–0.3

(2–3)

0.009–0.2

(0.09–2)

0.2–0.4

(2–4)

Thyroid 0.2–0.3

(2–3)

0.03–0.2

(0.3–2)

0.2–0.4

(2–4)

Red bone marrow 0.2–0.3

(2–3)

0.008–0.2

(0.08–2)

0.2–0.4

(2–4)

*

The upper-bound total doses are not the sum of the corresponding external and internal doses in this table. 

The minimum and maximum values for the upper-bound external and internal doses do not correspond to the 

same period (winter-over or summer); whereas, the total doses are combined from external and internal 

doses from the same winter-over or summer period. In addition, results are rounded up to one significant 

digit. 

6.5 Comparative Radiation Exposures in Daily Life

To understand the relative magnitude of the maximum total effective doses listed in 

Table 26 through Table 30, these doses can be compared to those from exposures to radiation 

that are common to most veterans and the general public. Table 31 lists various sources of 

radiation exposures in daily life (NCRP, 2009). Table 32 lists common diagnostic radiological 

exams and the doses that a patient receives from those exams (Mettler et al., 2008; NCRP, 2009).

A comparison of doses shows that the estimated dose from immersion in an Ar-41 cloud 

from the NPP is much higher than at similar reactors. For example, the annual Ar-41 immersion 

dose for the University of Wisconsin TRIGA research reactor in Madison, WI, has been 

estimated to be less than 0.001 rem (0.01 mSv) (USNRC, 2012a). The Ar-41 immersion dose for 

the TRIGA reactor at Washington State University in Pullman, WA, has been estimated to be 

less than 0.0001 mrem (0.001 mSv) (USNRC, 2011b). The calculated doses due to immersion in 

a cloud of Ar-41 are much higher at McMurdo Station due to the high-sided assumptions used 

and the NPP operating a higher percentage of the time versus a research nuclear reactor on a 

university campus.
61

Table 31. Doses from radiation in daily life

Description Effective Dose

(rem (mSv))

Upper limit of radiation dose permitted for radiation workers 5 (50) per year

Radiation dose from naturally occurring radioactive material in 

Guaparai, Brazil (One of the highest background radiation 

locations in the world)

1 (10) per year

Natural radiation dose per year (U.S. average)

External:

Space (cosmic ray): 0.033 rem

Terrestrial: 0.021 rem

Internal:

Ingestion: 0.029 rem 

Inhalation (Radon): 0.228 rem

0.31 (3.1) per year

Dose limit for public (except for medical care) 0.1 (1) per year

A round-trip flight between Los Angeles and New York (dose 

due to increased cosmic radiation at high altitude) < 0.01 (< 0.1)

Average annual dose around a nuclear plant < 0.01 (< 0.1) per year

Evaluated dose of radiation from radioactive substances

emitted from a nuclear fuel reprocessing plant per year < 0.01 (<0.1) per year

Table 32. Doses from radiological exams

Examinations and Procedures Effective Dose 

(rem (mSv))

Fluoroscopic x-ray examination

Upper gastro-intestinal 0.6 (6)

Barium enema 0.7 (7)

Computed tomography examination

Head 0.2 (2)

Chest 0.7 (7)

Abdomen/pelvis 1.0 (10)

Whole-body screening 1.0 (10)

Biopsy 0.1 (1)

Conventional x-ray examination

Chest 0.01 (0.1)

Cervical spine 0.02 (0.2)

Thoracic spine 0.1 (1)

Lumbar spine 0.15 (1.5)

Pelvis 0.07 (0.7)

Abdomen or hip 0.06 (0.6)
62

6.6 Potential Health Effects

The maximum total upper-bound equivalent dose (external and internal combined) for 

McMurdo Station support personnel for the organs considered is 0.6 rem (6 mSv) for any tour. 

The maximum total upper-bound equivalent dose is the thyroid dose that veterans accrued during 

winter-over DF 67. The total upper-bound equivalent dose for the red bone marrow was 

calculated to be 0.5 rem (5 mSv) for any tour. The doses to all other organs would be similar to 

the red bone marrow because the dose coefficients are comparable/similar. These organ doses are 

low and the associated probability that disease could have arisen from these doses is also low. 

For comparison, at effective doses less than 5 to 10 rem (50 to 100 mSv), risks of health effects 

are either too small to be observed or nonexistent (HPS, 2010).

The doses listed above are well below the level that would cause any deterministic health 

effects (health effects caused by cell death, e.g., cataracts, skin burns, etc.). For comparison, the 

dose limit for radiation workers of 5 rem (50 mSv) per year shown in the Table 31 above is 

intended to be protective against deterministic effects.

To determine the potential for stochastic effects (health effects that are random or 

probabilistic in nature where their occurrence cannot be predicted, e.g., a cancer), the upperbound doses can be compared to screening doses. The screening doses are equivalent doses to 

the organ or tissue in which the cancer type occurs that are intended to correspond to a 

probability of causation of 50 percent at an upper credibility limit of 99 percent. The lowest 

screening doses for any organ for which a solid tissue cancer can develop is for cancers of the 

thyroid, liver, or gall bladder that may develop at 10 or more years after exposure. (Kocher and 

Apostoaei, 2007)

The maximum calculated total upper-bound dose thyroid dose of 0.6 rem for the support 

personnel who were stationed at McMurdo from 1962 to 1979 is well below the screening dose 

of 7.5 rem for cancers of the thyroid, and the maximum calculated total upper-bound doses of 

0.6 rem for liver, gall bladder, and red bone marrow are well below the screening doses of 

4.0 rem, 6.0 rem, and 4.4 rem for cancers of the liver, gall bladder, and red bone marrow 

(leukemia), respectively. Therefore, it can be concluded that the maximum total upper-bound 

organ doses for support personnel at McMurdo Station have associated probabilities of causation 

that are far less than 50 percent for any cancers that might arise in McMurdo veterans. Because 

all other organs have higher screening doses and lower expected total upper-bound organ doses 

than those for the thyroid, liver, gall bladder and red bone marrow, it follows that the probability 

of causation for cancer of any of the other organs would be much less than 50 percent. The 

screening doses for all organ considered are reported in Kocher and Apostoaei (2007).
63

7.

Radiation Dose Assessment for Reactor and 

Decommissioning Crew Members

Reactor and decommissioning crew members were exposed to sources of external and 

internal radiation while working at the McMurdo Station NPP either during reactor operations or 

during decommissioning. In addition, reactor crew members would have been exposed to sources 

while outside of the NPP and while off-duty that were similar to the exposures of support 

personnel. Records show that some crew members who were at the NPP during operational years 

also participated in the decommissioning. 

The compiled gamma radiation dosimetry records for reactor crew members for DF years 

for which dosimetry results were located are summarized in Table 5. Since reactor crew 

members wore dosimeters that only recorded external doses and had diverse exposure scenarios, 

the total upper-bound doses that include internal doses cannot be estimated generically. Dose 

assessments for these individuals should be performed in accordance with the proposed 

procedure guidelines described in Section 8. This section provides information on sources of 

radiation exposure and pathways needed to support individualized dose assessments for reactor 

and decommissioning crew members.

7.1 Sources of External Radiation Exposure

7.1.1. Reactor Crew Members during NPP Operation

7.1.1.1 Radiation Sources at the NPP Facility while on Duty

The reactor crew members who worked at the NPP were exposed to external gamma and 

neutron radiation from sources inside and outside the facility. The following sources of potential 

exposure to external radiation are applicable for reactor personnel during the operational years of 

the NPP while on duty: the reactor, contaminated reactor facility components and equipment, 

radioactive waste, packages containing radioactive materials, radioactive gases, and radioactive 

contamination in the effluent discharge area and NPP yard.

While on duty, each reactor crew member was monitored for external gamma and 

neutron exposures with an individual film badge. For certain periods, TLDs were issued in 

addition to film badges. The film badges were exchanged and processed monthly whereas the 

TLDs were worn for the entire tour before processing (Johnson, 1971).

7.1.1.2 Radiation Sources while off Duty

Reactor crew members could have been exposed to external gamma radiation when 

transiting near the NPP fence line, such as hiking on Observation Hill, driving, or walking while 
64

off-duty. Reactor crew members would have also been exposed to gamma radiation from 

immersion in clouds of radioactive noble gases such as Ar-41. They are assumed to have not 

been wearing their dosimeters while off-duty (Armstrong, 2012).

7.1.2. Decommissioning Crew Members

7.1.2.1 Radiation Sources at the NPP Facility while on Duty

The decommissioning crew members who worked mainly inside the NPP were exposed 

to external gamma radiation from sources inside and outside the facility. The following sources 

of potential exposure to external radiation while on duty are applicable for crew members during 

the decommissioning period of the NPP: spent reactor cores, components and equipment 

contaminated with residual radioactive materials, radioactive waste, radioactive waste packages, 

radioactive contamination in the effluent discharge areas and the NPP yard, and bulk 

contaminated soil during movement of the soil.

While on duty, external gamma doses of decommissioning crew members were 

monitored with individually issued film badges. For certain periods, TLDs were issued in 

addition to film badges. The film badges were exchanged and processed monthly

(Johnson, 1971). The TLD exchange period is not known.

7.1.2.2 Radiation Sources while off Duty

Decommissioning crew members could have received external gamma radiation doses 

while hiking, driving, or walking close to the NPP fence. In addition, they could have been 

exposed to residual contamination in soil spilled during transport and ship loading. The crew 

members are assumed to have not been wearing their dosimeters while off-duty. (Armstrong, 

2012)

7.2 Sources of Internal Radiation Exposure

7.2.1. Reactor Crew Members during NPP Operation

7.2.1.1 Inhalation of Airborne Radioactive Materials while inside the NPP

Reactor crew members could have inhaled airborne radioactive materials while working 

inside the NPP. The potential sources of airborne radioactivity include radioactive gases and 

airborne particulates coming from the containment purging system, the radioactive waste 

disposal system, the primary sampling system, the chemistry laboratory, the decontamination 

pad, and the primary building and primary building addition (NNPU, 1967). Entry into areas 

with airborne radioactive materials would have been controlled using the radiation work permit 

process (Martin Company, 1964).

7.2.1.2 Inhalation of Airborne Radioactive Materials while outside the NPP

Reactor crew members could have inhaled airborne radioactive effluents to include gases 

and aerosols released from the main stack of the NPP and from worldwide fallout. The inhalation 
65

could have occurred during off-duty time and would have occurred in areas outside the NPP 

(Foster and Jones, 1982).

7.2.1.3 Ingestion of Drinking Water

Reactor crew members could have ingested radioactive materials in the drinking water at 

McMurdo Station. The drinking water could have contained radioactive materials released from 

the NPP to include tritium and other long-lived beta emitters. Also, drinking water could have 

contained traces of naturally-occurring radioactive materials and radioactive fallout from 

atmospheric testing of nuclear weapons (Foster and Jones, 1982).

7.2.2. Decommissioning Crew Members

7.2.2.1 Inhalation of Airborne Radioactive Materials while on Duty

Crew members could have inhaled airborne radioactive material during the 

decommissioning of the NPP. Such inhalation could have occurred during facility 

decontamination, equipment disassembly, and building dismantling. Crew members could have 

inhaled suspended contaminated soil during surveys, excavation, and decontamination of the 

effluent discharge areas and the NPP yard. In addition, crew members could have inhaled 

suspended soil contaminated with radioactive materials during transportation of bulk soil to the 

ship loading area and while loading the ships. Also, the air could have contained traces of 

naturally occurring radioactive materials and radioactive fallout from atmospheric testing of 

nuclear weapons. (Foster and Jones, 1982)

7.2.2.2 Inhalation of Airborne Radioactive Materials while off Duty

Crew members could have inhaled residual radioactive material in suspended 

contaminated soil left over after ship loading or spilled onto the roads leading to the ship loading 

area. The inhalation would have occurred while off duty in addition to exposures while on duty.

7.2.2.3 Ingestion of Drinking Water

Crew members could have ingested radioactive materials in the drinking water during the

decommissioning phase of the NPP. The drinking water could have contained tritium and 

radioactive materials released from the NPP. Also, drinking water could have contained traces of 

naturally occurring radioactive materials and radioactive fallout from atmospheric testing of 

nuclear weapons. (Foster and Jones, 1982)

7.2.2.4 Incidental Ingestion of Contaminated Soil and Dust

Decommissioning crew members could have incidentally ingested soil and dust 

contaminated with radioactive materials. The contaminated soil would have been disturbed and 

spread during decommissioning operations such as demolition work, soil excavation, surveys, 

transportation of bulk soil to the ship loading areas, and ship loading (Foster and Jones, 1982). In 

addition, crew members could have incidentally ingested residual soil contaminated with 

radioactive materials that was left over in the ship loading area or spilled onto the roads leading 

to the ship loading area (NNPU, 1976).
66

7.3 Dose Estimation Methodology

7.3.1. External Dose

An external dose assessment for individual reactor crew members, including 

decommissioning personnel, requires film badge results. Since doses among the members of the 

reactor crews varied with length of tour, year(s) of assignment, and specific duties, generic 

external doses for these cohorts cannot be estimated. 

If individual film badge records cannot be located for an individual reactor crew member, 

dosimetry for other crew members who were at McMurdo Station during the same time and who 

served in similar positions could be used as surrogate (cohort) data (DTRA, 2010, Standard 

Method ED01).

A further complication for individual dose assessments for crew members is that many of 

them were assigned to other reactors or other locations with potential exposures to external 

radiation during the same calendar years that they were assigned to McMurdo Station. The 

potential sources of exposures at other reactors or facilities could be different than the sources of 

exposure at McMurdo Station, and could have been combined with McMurdo Station exposures 

in their dosimetry records. 

It is assumed that reactor personnel did not wear their badges while off duty. Therefore, 

the external doses for off work time while outside the NPP are similar to the external dose 

estimates for support personnel, which are addressed earlier in the report.

7.3.2. Inhalation of Airborne Radioactive Materials 

Detailed air monitoring records and records of respiratory protection measures would 

normally be used to reconstruct the internal exposure from inhalation of radioactive materials. 

Although air monitoring data were typically collected for operational purposes, these data were 

not archived and are not available. Some air monitoring results have been located for operations 

involving removal of contaminated soil and rock from the reactor facility, storage yard, areas 

adjacent to the NPP, and effluent discharge areas. However, the air monitoring results are not 

sufficient to form the primary basis of an inhalation dose reconstruction (NNPU, 1976; NNPU, 

1977b).

An internal dose assessment for individual reactor crew members would require internal 

monitoring results. Since internal doses from inhalation of radioactive materials while at the NPP 

would have varied among the members of the reactor crews because of differences in length of 

tour, DF year(s) of assignment, and specific duties, a generic internal inhalation dose cannot be 

estimated. Therefore, if individual internal monitoring results cannot be located for a reactor 

crew member, internal monitoring results for other crew members who were at McMurdo Station 

during the same time and who served in similar positions could be used as surrogate data.

Two complications arise from using internal monitoring results from whole-body 

counting. First, since whole-body counting measures the internal deposition of long-lived 

gamma-ray emitters, it does not distinguish between inhaled or ingested radioactive materials. 

Also, whole-body counting does not distinguish between an intake during work inside the NPP

or off-duty outside the NPP. Furthermore, many of the reactor crew members were assigned to 
67

other reactors or other locations where they had the potential for the intake of radioactive 

materials during the same calendar year(s) as their tours at McMurdo Station.

The inhalation doses that a reactor crew member would have received while off duty and 

while outside the NPP would have been similar to the McMurdo Station support personnel. 

However, the dose from any airborne long-lived gamma-ray emitters and the Cs-137 in 

contaminated soil inhaled while off duty would be reflected in any individual internal monitoring 

results.

7.3.3. Ingestion of Radioactive Materials

The ingestion doses due to short-lived radioactive isotopes and tritium are generally not 

reflected in internal monitoring results obtained using a whole-body counter, which cannot detect 

tritium’s low-energy beta particle radiations (Cember, 1983). Those doses would need to be 

estimated using dose reconstruction methods and added to the doses determined from internal 

monitoring results.

For long-lived gamma emitting radioisotopes, dose assessments for individual reactor 

crew members would require individual internal monitoring results. Internal monitoring results 

from whole-body counts would measure the internal deposition of radioactive materials from 

both inhalation and ingestion.
68

8.

Recommendations for Radiogenic Disease 

Claim Process

8.1 Guidance for the Radiogenic Disease Claim Process

The Department of Veterans Affairs (VA) offers health registry exams, health care, 

disability compensation and other benefits to eligible veterans for diseases caused by exposure to 

radiation. Their dependents and survivors also may be eligible for benefits.

The following is provided to assist veterans who believe that they have a disease or 

ailment caused by possible exposure to ionizing radiation at McMurdo Station from the NPP 

between 1962 and 1979. For more information on how to submit a claim for radiogenic diseases, 

visit the U.S. Department of Veterans Affairs at the following webpages:

a) Understanding the Disability Claims Process:

http://www.vba.va.gov/bln/21/Topics/claims.htm

b) Disability Compensation for Radiation-Related Disease:

http://www.publichealth.va.gov/exposures/radiation/benefits/index.asp

8.1.1. For McMurdo Station Veterans who are First Time Filers

First-time filers, who are veterans claiming VA health care, benefits and disability 

compensation for a radiogenic disease arising from exposure to radiation from the NPP while on 

active duty service at McMurdo Station, should take the following steps:

• Complete a VA Form 21-526, Veterans Application for Compensation and/or Pension, and 

attach the following:

- Discharge or separation papers (DD214 or equivalent)

- Medical evidence (doctor and hospital reports)

- Dependency records if applying as dependent or survivor (marriage and children's birth 

certificates)

• This documentation can be submitted online or via a local VA Regional Office (VARO)

• For assistance, call toll-free: (800) 827-1000 
69

Veterans may also fill out a disability compensation application online at:

http://vabenefits.vba.va.gov/vonapp/.

Alternatively, veterans may print and mail-in Form 21-526 or call the VA at (800) 827-1000 to 

ask to have the form mailed to them.

8.1.2. For McMurdo Station Veterans Previously Denied VA Service 

Connection

Veterans previously denied VA service connection for disability compensation due to 

radiogenic disease arising from active duty service at McMurdo Station may re-file based on new 

information presented in this technical report. The veteran should contact their VA Regional 

Office and request that the claim for service connection be reopened. The veteran should cite this 

technical report, DTRA-TR-12-003(R2), as the basis for the request to reopen the claim, i.e., new 

and material evidence. 

8.1.3. For a Surviving Spouse of a Deceased McMurdo Station Veteran

The surviving spouse of a deceased McMurdo Station veteran should complete a VA 

Form 21-534 for an original claim. If an original claim has been filed in the past, and denied, 

then the surviving spouse can reopen the previously denied claim by submitting the basis for new 

evidence, i.e., cite this technical report (DTRA-TR-12-003(R2)). 

8.1.4. For the Veterans Benefits Administration (VBA)

The VBA has centralized the processing of radiogenic disease claims at VARO Jackson, 

MS (VBA, 2006).

• Upon receipt of a McMurdo Station veteran claim for radiogenic disease, VARO Jackson 

should request that the veteran complete a McMurdo Station Radiation Exposure 

Questionnaire (see Appendix C). 

• Upon receipt of a completed questionnaire, forward a request to Naval Dosimetry Center for 

the veteran’s occupational radiation exposure records. 

• Upon receipt of the Naval Dosimetry Center response, VARO Jackson should submit the 

records to the Under Secretary for Benefits (Director, Compensation Service), for review. 

The Under Secretary for Benefits may request an advisory medical opinion from the Under 

Secretary for Health (Director, Environmental Agents Service). 
70

8.1.5. For Director, Environmental Agents Service, Veterans Health 

Administration (VHA)

Use the scenarios and doses cited in DTRA-TR-12-003(R2) as guidance in performing 

medical opinions. Chief, Nuclear Test Personnel Review, DTRA is available for technical 

consultation. 

8.1.6. For the Naval Dosimetry Center

Upon receipt of VARO Jackson’s request for a McMurdo Station veteran’s radiation 

exposure records, first verify if occupational radiation exposure records exist for the veteran. 

Then ensure a complete radiation dose assessment using the guidelines with Section 8.2. The 

radiation dose evaluation would include information from any pertinent veteran occupational 

exposure records. Upon completion of this assessment, forward results to VARO Jackson.

8.2 Proposed Guidelines for Individual Radiation Dose Assessment 

for McMurdo Station Veterans

The following is provided as guidelines that should be used to create detailed procedures 

for performing individual radiation dose assessment for McMurdo Station veterans. 

8.2.1. Assignment to Appropriate Group and Collection of Information

Veterans of military service at the McMurdo Station are characterized by two groups of 

exposure scenarios: (1) reactor and decommissioning crew members of the NPP, and 

(2) McMurdo Station support personnel.

The first step in performing an individual dose assessment is to determine the veteran’s 

group. A recommended McMurdo Station questionnaire is included in Appendix C as a starting 

point in determining the group to which the veteran belongs.

The next step is to gather veteran-specific information about his or her participation in 

Operation Deep Freeze at McMurdo Station. The veteran’s personnel and medical records from 

the National Personnel Records Center, St Louis, MO, should be obtained and reviewed. 

Additional information should be collected based on the records reviewed from other official 

records. The veteran should be given the opportunity to provide additional input. This would be 

in the form of a questionnaire or other statements based on his or her recollection. Veterans

should also be requested to submit any documentation in their possession that contains 

information about their time at McMurdo Station.

8.2.2. Individual Dose Assessment for Support Personnel

Based on the veteran’s recollection and an analysis of relevant individual data and 

historical record, a scenario of participation and radiation exposure (SPARE) would be 

developed to include the veteran’s relevant activities and all possible exposure sources and 

pathways. The evaluation of exposure to radiation should be based on the pathways identified 

and assessed in this report. The dose calculations for each previously analyzed exposure pathway 
71

would then be adjusted to account for veteran-specific information. Changes to the standard 

calculations should only be considered if the adjustment to an exposure pathway results in a 

higher overall dose. An example of a potential adjustment to a known exposure pathway that 

may result in a higher overall dose would be for someone with duties that required visiting the 

NPP several times.

Additional sources of exposure for some McMurdo Station support personnel might need 

to be added to the exposure pathways already identified in this report. Such exposure pathways 

could include occupational exposure to x-ray radiation for individuals assigned to the health 

clinic or dental clinic (NDC, 1962–1979). Others could have assisted civilian researchers in 

experiments that included the use of radioactive materials such as described in the turn-in of 

radioactive waste to the facility (NNPU, 1977b).

If additional sources of exposures are identified, those specific doses should be calculated 

using standard dose reconstruction techniques such as those used in the NTPR program (DTRA, 

2010) or equivalent approved procedures. An upper-bound total external dose and total internal 

dose would then need to be estimated using the dose calculation methods provided in 

Appendix B.

8.2.3. Individual Dose Assessment for Reactor Crew Members

Based on a reactor or decommissioning crew member’s records and personal recollection 

along with historical records, a scenario of participation and radiation exposure (SPARE) would 

be developed to include the veteran’s relevant activities and all possible exposure sources and 

pathways. A dose for each source and pathway would then be calculated.

Members of reactor crews were monitored for radiation exposure using film badges, 

TLDs, whole body counting, and possibly other measuring devices. Therefore, doses from some 

of the exposure pathways would be based on individual monitoring records. The doses for 

sources and pathways not reflected in individual monitoring results would be estimated using 

dose reconstruction techniques (DTRA, 2010). 

Currently, sufficient in-plant monitoring and survey data have not been found to allow for 

the reconstruction of doses from internal exposure pathways for reactor crew members. Since 

their duties varied widely and historical records are limited, the reconstruction of doses from the 

intakes of radioactive material will need to be based on the veteran’s internal monitoring records 

and bioassay results. When individual monitoring records are not available, appropriate surrogate 

data can be used.

Several types of individual monitoring records exist that could be used for an individual 

dose assessment. A veteran’s record of occupational exposure can be found at the Naval 

Dosimetry Center or the Army Dosimetry Center. Also, a veteran’s DD Form 1141 “Record of 

Occupational Exposure” that can be found in the veteran’s medical record or in his or her

possession can be used to supplement the individual dosimetry results from a service dosimetry 

center. Whole-body counting results can be found in the veteran’s medical records or in his or 

her possession that could be used to determine the internal doses due to the intake of long-lived 

gamma emitters.

For those veterans whose film badge records cannot be located or who have gaps in their 

personal results, the dosimetry data of other veterans could be used as cohort dosimetry. The 
72

cohort film badge data would need to be from individuals who were assigned to the same tour 

who had the same or similar duties and military specialties. Since the film badge results would 

not be for the specific veteran, higher uncertainty factors should be used when calculating upperbound doses. (DTRA, 2010, Standard Method ED01 and UA01)

To determine the upper-bound external dose for an individual veteran, an uncertainty 

would be assigned to each film badge result. The upper-bound external dose would be calculated 

by combining the uncertainties of all film badge results following standard methods developed 

for the Nuclear Test Personnel Review Program or developed by the Navy (DTRA, 2010, 

Standard Method UA01; Blake, 2004).

To determine the internal dose for an individual veteran, procedures will need to be 

developed to use whole-body counting results to calculate committed effective doses or 

committed equivalent doses for the organs or tissues of interest. For example, the National 

Institute of Occupational Safety and Health (NIOSH) Dose Reconstruction Program uses the 

Integrated Modules for Bioassay Analysis (IMBA) software application to model input data that 

would produce whole body count equivalent results. IMBA has the capability of using input data 

like whole body counting results to determine intakes and internal doses to the organs of interest. 

(Brackett et al., 2008)

For a veteran whose whole-body counting results cannot be found or who did not receive 

a whole-body measurement upon returning to the U.S., other whole-body counting results can be 

used. The cohort monitoring data would need to be from the same DF year and from veterans 

with similar duties and military specialties.

Some radiation exposure pathways for reactor crew members were not monitored or 

accounted for by the veteran’s film badge records or whole-body counting results. These 

unmonitored doses could require a dose estimation based on existing measurement data. 

Examples of such doses are immersion in a cloud of noble gas or inhalation of airborne 

radioactive iodine or ingestion of tritium in the drinking water. However, such unmonitored 

doses are likely to be similar to the corresponding doses calculated for support personnel.

A complicating factor for dose reconstructions of reactor crew members who served at 

McMurdo Station is that many of them received radiation doses at other nuclear facilities during 

the same calendar years. Many veterans received training on the PM-1A at Fort Belvoir before 

being assigned to the McMurdo Station NPP. In addition, after leaving McMurdo Station, several 

returned to the PM-1A or were assigned to other military NPPs. 
73

9.

Conclusions

9.1 Summary of Assessment Approach

The main goal of the radiation dose assessment documented in this report was to estimate 

conservative upper-bound external and internal radiation doses for McMurdo Station support 

personnel. The goal was met by accomplishing the steps described below.

Operating histories, operator’s manuals, health physics logs, radioactive waste disposal 

documents, decommissioning reports, and other available records were reviewed. These 

historical records were used to develop a detailed description of the installation, operations, and 

decommissioning of the McMurdo Station NPP. In addition, all documented incidents and 

unusual deviations to normal operations at the NPP were reviewed.

Information was collected from U.S. Navy historical archives; verbal and written 

comments from veterans of McMurdo Station; and other governmental, historical, and academic 

sources. Information was located on environmental monitoring at McMurdo Station, on incidents 

and accidents that occurred, and on the radiation safety procedures that were followed. Reports 

were located on the results of the decommissioning efforts, the amount and types of radioactive 

waste generated during decommissioning, and the final radiological conditions at McMurdo 

Station.

External dosimetry results for Operation Deep Freeze 62 through 74 were located for 

reactor crew members. External dosimetry results were located for both winter-over and austral 

summer personnel. Internal monitoring results were located for several reactor crew members. In 

addition, environmental monitoring data based on dosimeter measurements during 

decommissioning were found and evaluated.

Seven potential sources of external exposure were identified to include the reactor, 

radioactive waste, radioactive effluent, radioactive materials in passing plumes, and 

contaminated soil deposited on the ground. Six potential sources of internal exposure were 

identified to include radioactive materials deposited in the tissues and organs after breathing air 

and ingesting water and soil and dust that contained radioactive material. The potential sources 

of both internal and external exposure were reviewed for any possible differences between 

winter-over and austral summer personnel. Where specific information on a potential source of 

exposure was unavailable, other methods of dose estimation, such as the use of surrogate data, 

were employed.

Using high-sided, conservative assumptions, both from exposure and contamination 

conditions that could have existed, radiation dose estimates were calculated for the McMurdo 

Station support personnel for each DF year. The radiation doses were estimated for winter-over 

and austral summer support personnel separately. The total effective whole body doses and the 

total equivalent dose to the thyroid and red bone marrow were estimated for winter-over and 
74

austral summer tours for each DF year between 1962 and 1979. The thyroid and red bone 

marrow were considered the best representatives of the human body organs that might be 

affected by the NPP radiation sources. 

9.2 Dose Assessment Results and Health Risks

The calculated upper-bound total effective doses for support personnel at the McMurdo 

Station for DF years 62 to 79 ranged from 0.2 to 0.6 rem (2 to 6 mSv) per winter-over or austral 

summer tour. The highest estimated upper-bound total effective dose of 0.6 rem (6 mSv) for any 

single tour at McMurdo Station was observed for the winter-over personnel during DF 67. The 

highest upper-bound dose had the highest tritium ingestion dose due to the reported tritium levels 

measured in the drinking water. For all DF years, the austral summer personnel had estimated 

upper-bound total doses that are smaller than for the winter-over personnel for the same year.

The estimated upper-bound total equivalent doses of the support personnel ranged from 

0.2 to 0.6 rem (2 to 6 mSv) for the thyroid and from 0.2 to 0.5 rem (2 to 5 mSv) for the red bone 

marrow per winter-over or austral summer tour. The highest estimated upper-bound total 

equivalent doses for any single tour are for the winter-over personnel during DF 67. The highest 

estimated upper-bound total equivalent dose had the highest tritium ingestion dose. For all DF 

years, the austral summer support personnel had estimated upper-bound total equivalent organ 

doses that were smaller than the winter-over personnel for the same year.

The upper-bound total doses for the support personnel in all DF years are comparable to 

the doses that veterans would have received from medical diagnostic procedures or from other 

sources encountered during their lives after leaving McMurdo Station. These doses are similar in 

magnitude to those that a person residing in the United States would receive, on average, in one

year from background or medical sources. 

The calculated doses are low and the associated probability that disease could have arisen 

from these doses is also low. For comparison, at effective doses less than 5 to 10 rem (50 to 

100 mSv), risks of health effects are either too small to be observed or are nonexistent (HPS, 

2010).

All estimated doses and upper-bounds reported here are generic cohort doses that can be 

assigned to any McMurdo Station support personnel veteran who had activities and exposures 

covered in this assessment for the period between 1962 and 1979. However, some support 

personnel may have been exposed to additional or distinctly different sources of radiation. For 

these veterans, proposed guidelines are provided in Section 8 for performing individualized 

radiation dose assessments that would include exposure scenarios that were not evaluated in this 

report.

9.3 Recommendations for Radiogenic Disease Claims Process

Recommendations and guidance are provided in this report to assist the McMurdo Station 

veterans and their dependents, the Secretary of Veterans Affairs and the Naval Dosimetry Center

in the radiogenic disease claims process. Guidance is provided to develop procedures that can be 

used to perform individual dose assessments for McMurdo Station veterans. Separate procedures 

are proposed for support personnel, and reactor and decommissioning crew members.
75

Using the available historical information, occupational dosimetry results and exposure 

pathways identified in this report, individual dose assessment procedures are proposed. The 

individualized assessments should be based on the pathways identified in this report along with 

an individual’s specific history. If additional sources of exposures are identified, those specific 

doses should be calculated using standard dose reconstruction techniques such as those used in 

the NTPR program.
76

10.

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Processing”. VBA Fast Letter 06-20, Washington, DC. October 16

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Freeway Ultrafine Particles into Indoor Environments”. Journal of Aerosol Science,

Volume 36, pp. 303–322.
81

Appendix A.

Congressional Inquiries Regarding Veterans’ Exposures to Radiation from 

the McMurdo Station Nuclear Power Plant
82

A-1. Letter from Senator Brown to Secretary of Defense
83

A-2. DoD Response to Senator Brown. 
84

A-3. Letter from Senator Blumenthal to Assistant Secretary of the Navy (Installations 

and Environment)
85

A-4. Department of the Navy Response to Senator Blumenthal
86
87

Appendix B.

Radiation Dose Calculation Methods and

Parameter Definitions
88

B-1. External Dose Assessment Methods and Parameter Definitions

B-1.1. Average Hourly Dose Rate inside the NPP

The average hourly dose rate inside the NPP is estimated using Equation 1 (DTRA, 2010, 

SM ED01):

Badge

average

average T

FB

D = (1)

where: 

Daverage

 = Average hourly dose rate inside the NPP (mSv h−1

)

FBaverage = Average annual film badge dose for each DF winter-over year (mSv)

TBadge = Time duration that the reactor crew members were monitored using film 

badges while inside the NPP (h)

For the years when no film badge data exist but environmental monitoring data exist, the 

average hourly dose rate inside the NPP is estimated using Equation 2 (DTRA, 2010, ED01):

monitor

highest

average T

TLD

D = (2)

where: 

TLDhighest = Highest TLD measurement during environmental monitoring (mSv)

Tmonitor = Time duration that the TLD was used as an environmental monitor (h) 

B-1.2. External Dose from Touring or Visiting the NPP 

For a specific Deep Freeze year, the dose while touring or visiting the NPP is estimated 

using Equation 3 (DTRA, 2010, SM ED02):

Dtour Daverage Ttour Ntour =  (3)

where: 

Dtour = Dose due to touring or visiting the NPP (mSv)

Ttour = Time duration that support personnel were inside the NPP for each tour or 

visit (h) 

Ntour = Number of visits to (or tours of) the NPP
89

B-1.3. External Dose from Transiting in Proximity to the NPP

For a specific DF year, the dose while transiting in proximity to the NPP is estimated 

using Equation 4 (Cember, 1983) based upon the reduction in dose rate with the square of the 

distance given the assumed distances to the film badge and the NPP boundary:

2

2

boundary

badge

transit average boundary boundary d

d

D = D T N (4)

where: 

Dtransit = Dose due to transiting in proximity to the NPP (mSv)

Tboundary = Time that a veteran would have been at the boundary during any transit 

near the NPP (h)

Nboundary = Number of transits near the NPP boundary

dbadge = Distance from the NPP radiation sources to the assumed location of the 

average film badge (m)

dboundary = Distance from the NPP radiation sources to the NPP boundary (m)

B-1.4. External Dose from Immersion in a Cloud of Noble Gases

The dose due to immersion in a cloud of noble gases emitted from the NPP is estimated 

using Equation 5 (USNRC, 2012b):

Ar ECL

Ar Ar

ng ECL Wind ECL T

C T D D F

41

41 41 = (5)

where: 

Dng = Dose due to immersion in a cloud of noble gases (mSv)

DECL = Dose to a member of the general public from one year exposure to a cloud 

containing the effluent concentration limit for Ar-41

FWind = Wind direction factor (the fraction of the time that the wind was blowing 

toward the area of interest)

CAr41 = Maximum permissible concentration for Ar-41 that could be released from 

the NPP stack (Bq mL−1

)

ECLAr41 = Effluent concentration for Ar-41 that results in a permissible dose limit of 

50 mrem in a year to a member of the general public (Bq mL−1

)

TAr41 = Time duration that a veteran was immersed in a cloud of noble gases (y)

TECL = Duration of exposure to effluent concentration for Ar-41 that results in a 

permissible dose limit of 50 mrem in a year to a member of the general 

public (y)
90

B-1.5. External Dose from Movement of Radioactive Waste Packages outside the NPP

The dose due to movement of radioactive waste packages outside the NPP is estimated 

using Equation 6 (Cember, 1983; USNRC, 2011a):

2

2

( )

Dose

Index

Pack II II III III Package d

d D = I N + I N T (6)

where: 

Dpack = Dose due to exposure during the movement of radioactive waste packages 

outside the NPP (rem)

III = Exposure rate from a Yellow-II radioactive package at the distance that 

the transportation index is determined (mSv h−1)

IIII = Exposure rate from a Yellow-III radioactive package at the distance that 

the transportation index is determined (mSv h−1)

NII = Number of Yellow-II packages to which a veteran would have been 

exposed

NIII = Number of Yellow-III packages to which a veteran would have been 

exposed 

TPackage = Time duration that a veteran was near any one radioactive waste 

package (h)

dIndex = Distance at which the transportation index was determined (m)

dDose = Assumed distance from the surface of the radioactive waste packages to 

the veteran (m)

B-1.6. External Dose from Shipments of Radioisotope Thermal Generators

The dose due to shipment of RTGs is estimated using Equation 7 (Cember, 1983;

USNRC, 2011a):

2

2

Dose

Index

RTG RTG RTG RTG d

d D = I N T (7)

where: 

DRTG = Dose due to shipments of RTGs (mSv)

IRTG = Exposure rate from a Yellow-III radioactive package containing an RTG 

at the distance that the transportation index is determined (mSv h−1

)

NRTG = Number of RTG packages to which a veteran could have been exposed 

TRTG = Time duration that a veteran was near an RTG (h)
91

B-1.7. External Dose from Contaminated Soil on the Ground

The dose due to spilled contaminated soil on the ground before cleanup and residual 

contaminated soil on the ground after cleanup is estimated using Equation 8 (DTRA, 2010, SM 

ED02):

( ) Dsoilext = ρ Fdose CHigh THigh + CLow TLow (8)

where: 

Dsoilext = Dose from exposure to contaminated soil on the ground (i.e., Shine) (mSv)

ρ = Density of the soil (g m−3)

Fdose = Dose coefficient assuming an infinite plane of Cs-137 contamination at a 

depth of 1 cm (mSv m3 h−1 Bq−1)

CHigh = Maximum activity concentration for contaminated soil on ground (Bq g−1)

CLow = Maximum permissible activity concentration of residual contaminated soil 

on ground (Bq g−1)

THigh = Time duration that a veteran was exposed to spilled contaminated soil on 

the ground before cleanup (h)

TLow = Time duration that a veteran was exposed to residual spilled contaminated 

soil on the ground after cleanup (h)

B-2. Inhalation Dose Assessment Methods and Parameter Definitions

B-2.1. Inhalation of Airborne Radioactive Materials

The dose due to inhalation of airborne radioactive materials is estimated using 

Equations 9a, b and c (DTRA, 2010, SM ID01):

Dbetainh = DFO + DNPPeff (9a)

FO Cesium Cesium out inside inf TAA D = BR C DC ( F + F F ) (9b)

[ ] DNPPeff BR CIodine DCIgas FIodine DCIaero Fout Finside F TAA ( ) = + + inf

(9c)

where: 

Dbetainh = Dose from inhalation of gross long-lived, beta-emitting airborne 

radioactive materials (mSv)

DFO = Dose from inhalation of gross long-lived, beta-emitters in worldwide 

fallout (mSv)

DNPPeff = Dose from inhalation of gross long-lived, beta-emitting radioactive 

material in the air effluents released from the NPP (mSv)

BR = Breathing rate (m3 h−1)

TAA = Time duration for inhalation of airborne radioactive materials (h)
92

CIodine = Average concentration of long-lived, gross beta-emitting radioactive 

aerosols for duration of intake, assumed to all be from aerosol I-131 

(Bq m−3

)

CCesium = Average concentration of long-lived, gross beta-emitting radioactive 

aerosols assumed to all be from aerosol Cs-137 (Bq m−3)

DCIgas = Inhalation dose coefficient of elemental gaseous I-131 (effective dose or 

organ dose) (mSv Bq–1)

DCIaero = Inhalation dose coefficient of aerosol I-131 (effective dose or equivalent

dose for an organ) (mSv Bq–1)

DCCesium = Inhalation dose coefficient of aerosol I-131 (effective dose or equivalent 

dose for an organ) (mSv Bq–1)

FIodine = Ratio of gaseous I to aerosol I

Fout = Fraction of the day that a veteran was outside

Finside = Fraction of the day that a veteran was inside

Finf = Infiltration factor for aerosols for indoor locations

B-2.2. Inhalation of Suspended Contaminated Soil during Decommissioning

The dose due to inhalation of suspended contaminated soil during decommissioning is

estimated using Equation 10 (DTRA, 2010, SM ID01): 

Dsoilinh = BR CCsaero DCCsaero Tsoil (10)

where: 

Dsoilinh = Inhalation dose from suspended contaminated soil during 

decommissioning (mSv)

CCsaero = Highest measured concentration of Cs-137 in air samples taken during soil 

decontamination (Bq m−3)

DCCsaero = Inhalation dose coefficient of aerosol Cs-137 (mSv Bq–1)

Tsoil = Time duration for soil intake at the specified concentration (h)

B-2.3. Inhalation of Contaminated Soil on the Ground, Roads, and Ship Loading Areas

The dose due to inhalation of spilled contaminated soil before clean-up and residual 

contaminated soil after spill clean-up on the ground, roads, and ship loading areas is estimated 

using Equation 11 (DTRA, 2010, SM ID01):

D BR K DC Thick (C T C T ) Ispillsoil = Csaero Cshigh High + Cslow Low (11

where: 

DIspillsoil = Inhalation dose from exposure to contaminated soil on the ground, roads, 

and ship loading area (mSv)

K = Suspension factor for soil contamination on the ground (m−1)

CCshigh = Maximum measured concentration of Cs-137 in contaminated soil 

(Bq g

−1)
93

CCslow = Maximum permissible concentration of Cs-137 in residual soil 

contamination (Bq g−1)

Thick = Assumed thickness of the spilled or residual contaminated soil (m)

B-3. Ingestion Dose Assessment Methods and Parameter Definitions

B-3.1. Ingestion of Tritium in Drinking Water

The dose due to ingestion of tritium in drinking water is estimated using Equation 12

(DTRA, 2010, SM ID01): 

DH 3ing = DRwater CH3 water DCH3 ing TWater (12)

where: 

DH3ing = Ingestion dose from tritium in the drinking water for each year (mSv)

DRwater = Ingestion volume of drinking water (L d−1)

CH3water = Average concentration of tritium in drinking water for duration of intake 

(Bq L−1

)

DCH3ing = Dose coefficient for ingestion of tritium (mSv Bq–1

)

TWater = Time duration for water intake at the specified concentration (d)

B-3.2. Ingestion of Other Long-Lived Beta Emitters in Drinking Water

The dose due to ingestion of long-lived beta emitters in drinking water is estimated using 

Equation 13 (DTRA, 2010, SM ID01): 

Dbetaing = DRwater Cbwater DCCsing TWater (13)

where: 

Dbetaing = Dose due to ingestion of long-lived beta emitters in the drinking water 

(mSv)

Cβwater = Average concentration of Cs-137 in drinking water for duration of intake 

(Bq L−1

)

DCCsing = Dose coefficient for ingestion of Cs-137 (mSv Bq–1)

B-3.3. Incidental Ingestion of Contaminated Soil and Dust

The dose due to incidental ingestion of spilled contaminated soil and dust before clean-up 

and residual contaminated soil and dust after clean-up is estimated using Equation 14 (DTRA, 

2010, SM ID01):

( ) Dsoiling = DS soil DCCSing THigh CCshigh + TLow CCslow (14)

where: 
94

Dsoiling = Dose from incidental ingestion of contaminated soil and dust (mSv)

DSsoil = Mass of soil and dust incidentally ingested each day (g d−1)

THigh = Time duration for incidental ingestion of residual contaminated soil before 

clean up (d)

TLow = Time duration for incidental ingestion of residual contaminated soil after 

clean up (d)

B-4. Dose Uncertainty Methods and Parameter Definitions

B-4.1. Uncertainty in External Doses

The upper-bound total external dose for McMurdo Station support personnel is estimated 

using Equation 15, which assumes that the dose components are uncorrelated (DTRA, 2010, SM 

UA01): 

∑ ∑[ ( )] = =

= + −

NDE

j

ext , j ext

NDE

j

DUBext Dext , j D UF

1

2

1

1 (15)

where: 

DUBext = Upper-bound total external dose due to all sources of external radiation 

(mSv)

Dext,j = The external dose component from each source/pathway (mSv)

j = j

th component of the total external dose

UFext = Uncertainty factor for external doses

NDE = Number of external dose components 

B-4.2. Uncertainty in Internal Doses

The upper-bound total internal dose for the McMurdo Station support personnel is 

estimated using Equation 16, which assumes that the dose components are fully correlated

(DTRA, 2010, SM UA01): 

int

NDI

j

UBint Dint,j UF 













 = ∑=1

(16)

where: 

UBint = Upper-bound internal dose due to all sources of internal radiation for each 

year (mSv)

Dint,j = The internal dose component from each source/pathway (mSv)

j = j

th component of the total internal dose

NDI = Number of internal dose components 

UFint = Uncertainty factor for internal doses
95

Appendix C.

McMurdo Station Radiation Dose Assessment Questionnaire
NDC 6470/20 (31 Dec 2012) Page 1 of 12

U.S. Naval Dosimetry Center

McMurdo Station Radiation Dose Assessment Questionnaire

TO START, SELECT ONE OF THE FOLLOWING TWO CATEGORIES:

 You were a military person who was NOT ASSIGNED to the McMurdo Nuclear Power Plant (NPP) 

(Support person or support personnel, hereafter)  [Please complete Sections I, II, III and V]

 You were a military person who was ASSIGNED to the McMurdo NPP 

(Reactor crew member, hereafter)  [Please complete Sections I, II, IV and V]

SECTION I: PARTICIPANT CONTACT INFORMATION

Name: (Last, First, Middle Initial) Service Number: Social Security Number: 

Mailing Address: 

Telephone: Cell Phone: Email: 

If this questionnaire is completed by someone other than the participant, please provide:

Name: (Last, First, Middle Initial)

Mailing Address:

Telephone: Cell Phone: Email:

Relationship to veteran:

SECTION II: ASSIGNMENT SUMMARY (At McMurdo Station)

Dates Assigned Rate/Rank Name(s) of Person(s) who Severed with You Arrival Date Departure Date
NDC 6470/20 (31 Dec 2012) Page 2 of 12

SECTION III: MCMURDO STATION SUPPORT PERSON

The following questions are intended to assess your potential for radiation exposure for a support person who 

was NOT ASSIGNED to the McMurdo NPP. Please provide detailed answers to the best of your recollection.

Qualify as “approximate” where necessary. If more space is needed for any question, use additional sheets and 

include reference to section and question numbers. If you are unable to answer a question, state “Unknown”.

If you were assigned to McMurdo Station from 1962 to 1972, please complete the questions in A, C and D.

If you were assigned to McMurdo Station from 1973 to 1979, please complete the questions in B, C and D.

If you were assigned to McMurdo Station during both periods, please complete the questions in A, B, C and D.

A. Assigned to McMurdo Station while the NPP was operational (1962-1972)

1. Did you ever enter the NPP facility such as the reactor building, supply building, or the fenced yard? 

Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o Approximately, how many times did you enter the NPP facility? 

o What were the reasons for going into the NPP facility? 

o What areas and buildings were you in and what activities did you carry out while at the NPP?

2. Did you handle radioactive waste or packages such as helping move radioactive waste or helping load 

packages containing radioactive material on a ship? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o What was your role in the handling of waste or packages containing radioactive material? 

o How many times approximately? 

o For how long on average each time? 
NDC 6470/20 (31 Dec 2012) Page 3 of 12

3. Did you ever handle, transport, or work in close proximity to radioisotope thermal generators (RTGs)?

Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o Describe your activities in handling, transporting, working near RTGs: 

o Approximately how many times did you handle RTGs? 

o How long did each time take on average? 

4. In the course of your duties or during off time, did you transit near the NPP, such as walked/hiked

or drove/rode near the fenced area of the facility? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o Describe the purpose of your transit(s) near the fenced area of the NPP: 

o How many times did you transit near the fenced area of the NPP? 

o How long each transit lasted on average? 

5. Are there any other duties, actions or locations that you think may have caused you to be exposed to 

radiation from the McMurdo NPP? 

6. Where were your living quarters while at McMurdo Station? 

7. Where did you have your meals while at McMurdo Station? 
NDC 6470/20 (31 Dec 2012) Page 4 of 12

B. Assigned to McMurdo Station while the NPP was decommissioned (1973-1979)

1. Did you ever enter the NPP facility during decommissioning? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o Approximately, how many times did you enter the facility? 

o What were the reasons for entering the NPP facility? 

o What areas and buildings were you in and what activities did you carry out in the NPP?

2. Did your duties require you to assist in the decommissioning of the NPP? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o What did you do to assist in decommissioning activities? 

o How many hours/days did you spend inside the NPP facility assisting in its decommissioning? 

3. Did you handle radioactive waste such as helped move radioactive waste or load packages containing 

radioactive materials on a ship? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o What was your role in the handling of waste or packages containing radioactive materials? 

o How many times approximately? 

o For how long on average each time? 
NDC 6470/20 (31 Dec 2012) Page 5 of 12

4. Did you ever handle or work in close proximity to radioisotope thermal generators (RTGs)? 

Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-questions:

o Describe your activities in handling, transporting or working near RTGs: 

o Approximately, how many times did you handle RTGs? 

o How long did each time take on average? 

5. In the course of your duties or during off time, did you transit near the NPP, such as walked/hiked or 

drove/rode near the fenced area of the facility? Yes ___ No ___

If No, skip to the next numbered question. If Yes, answer the following sub-question:

o Describe the purpose of your transit(s) near the fenced area of the NPP: 

o How many times total did you transit near the fenced area of the NPP? 

o How long each transit lasted on average? 

6. Are there any other duties, actions or locations that you think may have caused you to be exposed to 

radiation from the decommissioning of the McMurdo NPP? 

7. Where were your living quarters while at McMurdo Station? 

8. Where did you have your meals while at McMurdo Station? 
NDC 6470/20 (31 Dec 2012) Page 6 of 12

C. Radiological Monitoring

1. Were you issued a film badge (radiation dosimeter) during your assignment at McMurdo Station? 

Yes ___ No ___

If No, skip to Section D If Yes, answer the following sub-question:

o Provide details (i.e., when did you wear the dosimeter, in what areas, etc.):

2. Were you ever advised of the results of the badge readings? Yes ___ No ___

If No, skip to D. If Yes, answer the following sub-question:

o Please give any details you remember about your film badge doses:

D. Additional Comments

1. Add any information you believe was not covered under the questions in this section:

SECTION IV: McMurdo Station REACTOR CREW MEMBERS

The following questions are intended to assess your potential for radiation exposure as an individual who was 

ASSIGNED to the NPP as a reactor crew member. Please provide detailed answers to the best of your 

recollection. Qualify as “approximate” where necessary. If more space is needed for any question, use 

additional sheets and include reference to section and question numbers. If you are unable to answer a question, 

state “Unknown”.

If you were assigned to McMurdo Station from 1962 to 1972, please complete the questions in A, C and D.

If you were assigned to McMurdo Station from 1973 to 1979, please complete the questions in B, C and D.

If you were assigned to McMurdo Station both periods, please complete the questions in A, B, C and D.
NDC 6470/20 (31 Dec 2012) Page 7 of 12

A. Assigned to the NPP while in operation (1962-1972)

1. What was your job/duty at the NPP? 

2. Provide details of your work and radiation exposure at the NPP (i.e., which building/area you worked in, 

duties requiring radiation work permits, incident, use of personal protective equipment, etc.): 

3. Were you issued film badge dosimeters (radiation dosimeters) during your assignment? 

Yes ___ No ___

If Yes, skip to next sub-question. If No, answer the following sub-question and skip to the next 

numbered question:

o Do you know the reasons why you were not issued film badges? 

o Did you wear the badge during plant operations? Yes ___ No ___

If Yes, skip to next numbered question. If No, answer the following:

At what times did you NOT wear your film badge and what was the reason for NOT wearing 

your film badge? 
NDC 6470/20 (31 Dec 2012) Page 8 of 12

4. Answer the following questions if you were monitored for internal deposition of radionuclides such as a 

whole-body count or bioassay sample after returning to the United States. 

o Was a whole-body count or thyroid count performed on you? Yes ___ No ___

If Yes, when and where did you receive a whole-body count? 

o Was a radiation bioassay such as a urine bioassay for radioactive material ever performed? 

Yes ___ No ___

If Yes, when and where did you receive a bioassay test? 

5. We you assigned to other nuclear power plants while in military service? Yes ___ No ___ 

If No, skip to the next numbered question. If Yes, answer the following sub-question:

o Where and when were you assigned to other nuclear power plants while in military service? 

6. Are there any other duties, actions or locations that you think may have caused you to be exposed to 

radiation from the operation of the McMurdo NPP that would not have been recorded by your film 

badges (if one was issued)? 

7. Where were your living quarters while at McMurdo Station? 

8. Where did you have your meals while at McMurdo Station? 

B. Duties at the NPP during decommissioning (1973-1979)

1. What was your job/duty at the NPP during decommissioning? 
NDC 6470/20 (31 Dec 2012) Page 9 of 12

2. Provide details of your work and radiation exposure at the NPP (i.e., defueling, decontamination, 

dismantling, removal of contaminated soil, use of personal protective equipment, etc.): 

3. Were you issued film badge dosimeters (radiation dosimeters) during your assignment? 

Yes ___ No ___

If Yes, skip to next sub-question. If No, answer the following sub-questions and skip to the next 

numbered question:

o Do you know the reasons why you were not issued film badges? 

o If you were issued film badges, were there times when you did NOT wear them? Please state when 

you did NOT wear the film badges and the reasons for not wearing them:

4. Answer the following questions if you were monitored for internal deposition of radionuclides 

such as a whole-body count or bioassay sample after returning to the United States. 

o Was a whole-body count or thyroid count performed on you? Yes ___ No ___

If Yes, when and where did you receive a whole-body count? 

o Was a radiation bioassay such as a urine bioassay for radioactive material ever performed? 

Yes ___ No ___

If Yes, when and where did you receive a bioassay test? 
NDC 6470/20 (31 Dec 2012) Page 10 of 12

5. We you assigned to other nuclear power plants while in military service? Yes ___ No ___ 

If No, skip to the next numbered question. If Yes, answer the following sub-question:

o Where and when were you assigned to other nuclear power plants while in military service? 

6. Are there any other duties, actions or locations that you think may have caused you to be exposed to 

radiation from the decommissioning of the McMurdo NPP that would not have been recorded by your 

film badge (if one was issued)?

7. Where were your living quarters while at McMurdo Station? 

8. Where did you have your meals while at McMurdo Station? 

C. Historical documentation

1. Do you have any historical records that might be relevant to radiation exposure such as Form DD 1141 

or the results of internal monitoring? Yes ___ No ___ 

If No, skip to the next numbered question. If Yes, answer the following sub-question:

o Please include a copy of such documents with this returned form and identify the documents here: 
NDC 6470/20 (31 Dec 2012) Page 11 of 12

2. Do you know the name of anyone who was at McMurdo Station during the same time period and 

performed the same or similar duties? Yes ___ No ___ 

If No, skip to the next numbered question. If Yes, answer the following sub-question:

o Provide the names of colleagues and coworkers who may have had similar 

radiation exposures: 

D. Additional Comments

1. Add any information you believe was not covered under the questions in this section:

SECTION V: SIGNATURE

I certify under penalty of perjury under the laws of the United States of America that the information provided on this 

form is true and correct. 

Print Name: 

Signature: Date:
NDC 6470/20 (31 Dec 2012) Page 12 of 12

SECTION VI: PRIVACY ACT STATEMENT

AUTHORITY: 38 U.S.C. 1154 (Veterans’ Benefits) delineates United States Navy (USN) and Department of Veterans 

Affairs (VA) responsibilities for collection of information in the adjudication of non-presumptive radiogenic disease 

compensation. 

PRINCIPAL PURPOSES: The information on this form is necessary to facilitate location of record(s) or information, 

provide participation and dose information, support scientific studies or medical follow-up programs, and provide data 

or documentation relevant to the processing of administrative claims or litigation. For use by Agency officials, 

employees, and authorized contractors.

ROUTINE USES: Disclosures are permitted under 5 U.S.C. 552a(b) of the Privacy Act, to USN, VA, Department of 

Defense, Veterans’ Advisory Board on Dose Reconstruction and under the ‘Blanket Routine Uses’ published at the 

beginning of USN’s compilation of systems of records notices.

DISCLOSURE: Voluntary. However, failure to provide the requested information may delay or preclude the USN/VA 

from producing your radiation dose assessment.

SECTION VII: AGENCY DISCLOSURE NOTICE

The public reporting burden for this collection of information is estimated to be less than one hour. If you have any 

questions regarding this form, please write to Officer in Charge, Naval Dosimetry Center, Bldg 4/6, 4th Floor, 8901 

Wisconsin Ave, Bethesda, MD 20889-5614
119

Abbreviations, Acronyms, and Unit Symbols

AEC Atomic Energy Commission

Ar argon

ASN(E,I&E) Assistant Secretary of the Navy for Energy, Installations, and 

Environment 

ASTSWMO Association of State and Territorial Solid Waste Management 

Officials

Bi bismuth

Bq becquerel

Ce cerium

CDR Commander

CFR Code of Federal Regulations 

Ci curie

cm centimeter

Co cobalt

Cs cesium

d day

DD Defense Department

DF Deep Freeze

DoD Department of Defense

DTRA Defense Threat Reduction Agency

ED external dose

EPA Environmental Protection Agency

ERSP Environmental Radiation Surveillance Program

g gram

Gy gray

HPS Health Physics Society

HPSC Health Physics Personal Safety and Chemistry

h hour

I iodine

IAEA International Atomic Energy Agency

ID internal dose

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IPT Integrated Project Team

L liter

LCDR Lieutenant Commander

m meter

μCi microcurie
120

mL milliliter

μm micrometer

Mn manganese

MPBB maximum permissible body burden

MPC maximum permissible concentration

mrem millirem

N number of years of age

NAS National Academy of Science

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NDC Naval Dosimetry Center 

NIOSH National Institute of Occupational Safety and Health 

NNPU Naval Nuclear Power Unit

NNPUD Naval Nuclear Power Unit Detachment

NPP nuclear power plant

NRC National Research Council

NTPR Nuclear Test Personnel Review 

pCi picocurie

PM-1A Portable, Medium type 1A Nuclear Power Plant

PM-3A NPP Portable, Medium type 3A Nuclear Power Plant

R roentgen

RDA radiation dose assessment 

rem roentgen equivalent man

RTG radioisotope thermal generator

Ru ruthenium

SC-1 VBDR Subcommittee 1 on DTRA Dose Reconstruction Procedures

SPARE Scenario of Participation and Exposure

Sr strontium

Sv sievert

Tl thallium

TLD thermoluminescent dosimeter

TRIGA Training, Research, Isotopes, General Atomics

UA uncertainty analysis

UB upper-bound

USDOE United States Department of Energy

USD(P&R) Under Secretary of Defense for Personnel and Readiness

USEPA United States Environmental Protection Agency

USN United States Navy

USNRC United States Nuclear Regulatory Commission

USUHS Uniformed Services University of the Health Sciences
121

VA United States Department of Veterans Affairs

VARO VA Regional Office

VBA Veterans Benefits Administration

VBDR Veterans' Advisory Board on Dose Reconstruction

VHA Veterans Health Administration

wk week

Xe xenon

y year

Zr zirconium
122

Authors’ Biographies

Lieutenant Commander Gregory R. Fairchild, PhD

Lieutenant Commander (LCDR) Fairchild is the Science and Research Advisor aboard 

the NDC. The NDC provides a centralized radiation dosimetry processing and consultation 

service to naval commands worldwide that use dosimetry as part of the Navy's Radiation Health 

Protection Program. It maintains a registry of ionizing radiation exposure data for Navy and 

Marine Corps personnel (military and civilian) exposed to radiation sources in their work 

environment, and responds to requests for exposure history information.

LCDR Fairchild is responsible for the accuracy and credibility of naval personnel 

radiation dose measurements, to include both external and internal assessments. He is directly 

responsible for accreditation of the Navy’s current radiation dosimeter, the DT-702/PD, through 

the National Voluntary Laboratory Accreditation Program. Since 1990, he has worked in 

radiation physics in a variety of settings, including boiling water and pressurized water 

commercial nuclear power plants, environmental radiological cleanup, research facilities, 

universities, hospitals and aboard a U.S. Navy nuclear aircraft carrier.

LCDR Fairchild is a radiation health officer in the Navy Medical Services Corps. Prior to 

his current tour aboard the NDC, he served aboard Naval Medical Center San Diego, USS 

NIMITZ (CVN 68), Naval Hospital Okinawa Japan, and in duty under instruction at the 

University of Minnesota.

LCDR Fairchild is a plenary member of the Health Physics Society and a full member of 

the American Association of Physicists in Medicine. He completed a bachelor’s degree in 

environmental and occupational health and safety at Brigham Young University, a master’s

degree in health physics at Texas A&M University, and a doctorate degree in medical physics at 

University of Minnesota. He has coauthored articles on the influence of therapeutic radiation on 

bone.

He is a qualified Surface Warfare Medical Department Officer. His military decorations 

include two Navy Commendation Medals, the Navy Achievement Medal, a Navy Unit 

Commendation Medal, a Navy E Ribbon, the National Defense Service Medal, the Global War 

on Terrorism Expeditionary Medal, the Global War on Terrorism Service Medal, a Sea Service 

Deployment Ribbon, three Navy Overseas Ribbons, and a Pistol Marksmanship (Sharpshooter) 

Medal.

Paul K. Blake, PhD

Dr. Paul K. Blake is the Program Manager of the Nuclear Test Personnel Review (NTPR) 

Program, Nuclear Technologies Department, Defense Threat Reduction Agency (DTRA), Fort

Belvoir, VA. DTRA safeguards America and its allies from weapons of mass destruction 

(chemical, biological, radiological, nuclear, and high yield explosives) by providing capabilities 

to reduce, eliminate, and counter the threat, and mitigate its effects.
123

He leads the Department of Defense’s efforts to confirm participation and reconstruct 

radiation doses for veterans involved in U.S. atmospheric nuclear weapons testing (1945 to 

1962), and the post-World War II occupation forces of Hiroshima and Nagasaki, Japan.

Dr. Blake retired from the U. S. Navy as a captain, after serving 26 years on active duty. 

He served aboard the USS Fulton (AS 11) and the USS Samuel Gompers (AD 37). He served as 

Head, Radiation Dosimetry Division, Armed Forces Radiobiology Research Institute. Upon 

completing his doctoral studies in 1986, he reported to the NDC as Science Advisor. Blake had 

the honor of co-designing and implementing the Navy’s personnel radiation dosimeter – the 

DT-648/PD. In 1989, he transferred to Naval Hospital Portsmouth as staff physicist and radiation 

safety officer and later served as Head of the Radiology Department. 

In later duty assignments he served as Head, Radiological Controls Branch, on the staff 

of Chief of Naval Operations and as the Surgeon General’s Specialty Leader for Radiation 

Health while assigned to the Bureau of Medicine and Surgery. In 2002, he returned to the Naval 

Dosimetry Center as Officer in Charge and supervised the implementation of the Navy’s next 

generation radiation dosimeter – the DT-702/PD which currently monitors over 50,000 workers. 

In 2004, he briefly served on the faculty of the Uniformed Services University of Health 

Sciences before retiring.

Dr. Blake is a diplomate of the American Board of Health Physics and is a member of the 

U.S. Naval Institute and Health Physics Society. He serves as the DoD representative on the 

Veterans’ Advisory Board on Dose Reconstruction. He earned a bachelor’s degree in chemistry 

and a master’s degree in health physics from Rutgers, the State University of New Jersey. During 

his graduate studies at Rutgers University he served as a Health Physics Fellow at Brookhaven 

National Laboratory in Long Island, N.Y. He later earned a doctorate degree in medical physics 

from the University of Wisconsin-Madison.

William J. Morris, MS

Mr. Morris is a consultant living in Williamsburg, Virginia. He retired after working for 

the Department of the Navy for 39 and one-half years. He initially worked as a guest worker in 

health physics at National Bureau of Standards at the reactor and linear accelerator for six 

months each and a year at Naval Research Laboratory assisting in the decommissioning of the 

NRL.

He worked as a staff health physicist in the Naval Facilities Engineering Command, 

Nuclear Engineering Division reviewing NPP operating reports, conducting safety reviews, and 

reviewing Navy radioactive material license applications for submittal to the Nuclear Regulatory 

Commission. He served on the NPP inspection team in December 1974 at McMurdo Station to 

review plant operations and on the NPP decommissioning crew in Deep Freeze 76 (October 

1975-early February 1976) conducting field surveys for soil contamination.

Mr. Morris was a staff health physicist at the Navy Radiological Affairs Support Office 

(RASO) reviewing and coordinating Navy radioactive material license application submittals to 

the Nuclear Regulatory Commission. He inspected Navy radiation safety programs, conducted 

installation radiation shielding reviews, taught classes to Navy Radiation Safety Officers, and 

conducted radiation incident and exposure investigations.
124

He served as head of the RASO licensing and laboratory branch. His responsibilities 

included coordinating licensing, shielding reviews and laboratory services. In 1987, the Navy 

obtained a USNRC Master Materials License for radioactive material used by the Navy and the 

US Marine Corp. He was responsible for review of permit applications for industrial uses of 

radioactive material, recommending approval to the Navy Radiation Safety Committee, and 

reviewing the impact of and drafting Navy comments on proposed USNRC regulations.

He later became an environmental protection manager with RASO. His responsibilities 

included reviewing historical radiological site assessments, decommissioning plans and surveys, 

conducting special studies and coordinating issues with senior Navy and federal agency 

managers.

Mr. Morris is chairman of the ANSI N43 Accredited Standards Committee on Equipment

for Non-Medical Radiation Applications and served on working subcommittees for two ANSI 

standards. He is a diplomate of the American Board of Health Physics and a member of the 

Health Physics Society. He earned a Bachelor’s degree in chemical engineering from the 

University of Wisconsin and a Master’s degree in nuclear engineering from Catholic University 

in Washington, DC.

Jason D. Dunavant, MS

Mr. Dunavant is a senior health physicist in the NTPR Program at SAIC in McLean, VA. 

He develops radiological dose assessments and dose reconstructions for atomic veterans to 

include technical analyses of historical data, reviewing radiation measurement, and producing 

detailed radiation dose calculations for atomic veterans. He has developed standardized methods 

and quality assurance procedures used for dose reconstructions. He was one of the authors of the 

basis document for a technical approach for expedited processing of doses for atomic veterans. 

In addition, he has worked on the basis document for doses of veterans who were exposed to 

radioactive light sources in night vision adaptometers.

He has also worked on radiation dose assessments for U.S. DoD-affiliated populations 

who were exposed to radioactive material released from the Fukushima Daiichi Nuclear Power 

Plant following the Great Pacific Tsunami in 2011. He has worked on both deterministic and 

probabilistic dose assessment analysis of DoD-affiliated populations in Japan.

Before coming to SAIC, he was the Director of Radiation Safety at Georgetown 

University Hospital. He oversaw the radiation safety program for a broad scope USNRC medical 

license, a District of Columbia radiation permit, and a wide range of machine-produced radiation 

sources. He oversaw the radiation safety aspects of medical treatment and of human-use research 

involving radioactive material and machine-produced radiation. He participated in a Department 

of Homeland Security pilot program on hospital preparation for contaminated patients and 

radiological incident response. 

Mr. Dunavant retired from the Army after 20 years of service as a nuclear medical 

science officer. He served as a health physicist and radiation safety officer at the Army 

Environmental Hygiene Agency (now called the Army Institute for Public Health). He served as 

the deputy radiation safety officer at White Sands Missile Range in New Mexico; and at Walter 

Reed Army Medical Center and the Walter Reed Army Institute of Research in Washington, DC. 

He was the Chief of Health Physics at Tripler Army Medical Center in Honolulu, Hawaii. His 
125

last assignment in the Army was as the Radiological Health Officer and the Radiation Safety 

Staff Officer for Army Materiel Command at Fort Belvoir, VA where he oversaw the radiation 

safety programs for military items containing radioactive material and the radiation safety 

aspects of equipment acquisition programs.

He is a diplomate of the American Board of Health Physics and a member of the Health 

Physics Society and American Society of Test and Measurements. He earned a bachelor’s degree 

in physics from the University of Houston and a master’s degree in health physics from Colorado 

State University.

Mondher Chehata, PhD

Dr. Mondher Chehata is a senior environmental scientist and program manager for NTPR 

Program at SAIC in McLean, VA. He manages the SAIC program to produce radiation dose 

assessments for atomic veterans. His previous work in NTPR included analyzing historical 

records; evaluating radiation data, and producing detailed radiation dose assessment. During his 

time as program manager, he has developed technical methods and authored reports on using 

probabilistic analysis for radiation dose assessment, on expedited processing for atomic veterans, 

and on a variety of specific technical aspects of the NTPR program.

Dr. Chehata has also participated in dose assessments for U.S. DoD-affiliated populations 

who were exposed to radioactive material released from the Fukushima Daiichi Nuclear Power 

Plant following the 2011 Great East Japan Earthquake and Tsunami. He participated in the 

development of dose estimates for U.S. Service members, civilians and dependents who were 

deployed in Japan and those who carried out the humanitarian relief assistance to the government 

of Japan and affected populations. In, addition, he developed probabilistic models to assess 

radiation doses U.S. DoD-affiliated persons present in Japan following the nuclear reactor 

accident. 

Prior to working on the NTPR program, Dr. Chehata’s professional career of more than 

25 years included supporting programs for the Office of Science and Technology, Office of 

Environmental Management of the Department of Energy, the Environmental Protection 

Agency, the National Park Service, the U.S. Army, the U.S. Agency for International 

Development, and other domestic and international programs. His work included, among others, 

hydrological and environmental modeling, GIS-based water security threat assessment, and 

evaluation of environmental information systems.

Dr. Chehata earned a bachelor’s and master’s degree in hydraulics engineering and water 

resources management from the University of Tunis and a doctorate degree in civil engineering 

from Colorado State University.