Title 32: National Defense 

PART 218—GUIDANCE FOR THE DETERMINATION 

AND REPORTING OF NUCLEAR RADIATION DOSE 

FOR DOD PARTICIPANTS IN THE ATMOSPHERIC 

NUCLEAR TEST PROGRAM (1945-1962)

Contents

§218.1 Policies.

§218.2 General procedures.

§218.3 Dose reconstruction methodology.

§218.4 Dose estimate reporting standards.

AUTHORITY: Pub. L. 98-542, 98 Stat. 2725 (38 U.S.C. 354 Note.)

SOURCE: 50 FR 42521, Oct. 21, 1985, unless otherwise noted. 

§218.1 Policies.

(a) Upon request by the Veterans Administration in connection with a claim for compensation, or 

by a veteran or his or her representative, available information shall be provided by the 

applicable Military Service which shall include all material aspects of the radiation environment 

to which the veteran was exposed and shall include inhaled, ingested and neutron doses. In 

determining the veteran's dose, initial neutron, initial gamma, residual gamma, and internal 

(inhaled and ingested) alpha, beta, and gamma shall be considered. However, doses will be 

reported as gamma dose, neutron dose, and internal dose. The minimum standards for reporting 

dose estimates are set forth in §218.4.

(b) The basic means by which to measure dose from exposure to ionizing radiation is the film 

badge. Of the estimated 220,000 Department of Defense participants in atmospheric nuclear 

weapons tests, about 145,000 have film badge dose data available. The information contained in 

the records has been reproduced in a standard format and is being provided to each military 

service, which can use the film badge dose data to obtain a radiation dose for a particular 

individual from that service. This is done upon request from the individual, the individual's 

representative, the Veterans Administration, or others as authorized by the Privacy Act. Upon 

request, the participant or his or her authorized representative will be informed of the specific 

methodologies and assumptions employed in estimating his or her dose. The participant can use 

this information to obtain independent options regarding exposure.
(c) From 1945 through 1954, the DoD and Atomic Energy Commission (AEC) policy was to 

issue badges only to a portion of the personnel in a homogeneous unit such as a platoon of a 

battalion combat team, Naval ship or aircraft crew. Either one person was badged in a group 

performing the same function, or only personnel expected to be exposed to radiation were 

badged. After 1954, the policy was to badge all personnel. But, some badges were unreadable 

and some records were lost or destroyed, as in the fire at the Federal Records Center in St. Louis. 

For these reasons the Nuclear Test Personnel Review (NTPR) Program has focused on 

determining the radiation dose for those personnel (about 75,000) who were not issued film 

badges or for whom film badge records are not available.

(d) In order to determine the radiation dose to individuals for whom film badge data are not 

available, alternative approaches are used as circumstances warrant. All approaches require 

investigation of individual or group activities and their relationship to the radiological 

environment. First, if it is apparent that personnel were not present in the radiological 

environment and had no other potential for exposure, then their dose is zero. Second, if some 

members of a group had film badge readings and others did not—and if all members had a 

common relationship with the radiological environment—then doses for unbadged personnel can 

be calculated. Third, where sufficient badge readings or a common relationship to the 

radiological environment does not exist, dose reconstruction is performed. This involves 

correlating a unit's or an individual's detailed activities with the quantitatively determined 

radiological environment. The three approaches are described as follows:

(1) Activities of an individual or his unit are researched for the period of participation in an 

atmospheric nuclear test. Unit locations and movements are related to areas of radiation. If 

personnel were far distant from the nuclear detonation(s), did not experience fallout or enter a 

fallout area, and did not come in contact with radioactive samples or contaminated objects, they 

were judged to have received no dose.

(2) Film badge data from badged personnel may be used to estimate individual doses for 

unbadged personnel. First, a group of participants must be identified that have certain common 

characteristics and a similar potential for exposure to radiation. Such characteristics are: 

Individuals must be doing the same kind of work, referred to as activity, and all members of the 

group must have a common relationship to the radiological environment in terms of time, 

location or other factors. Identification of these groups is based upon research of historical 

records, technical reports or correspondence. A military unit may consist of several groups or 

several units may comprise a single group. Using proven statistical methods, the badge data for 

each group is examined to determine if it adequately reflects the entire group, is valid for use in 

statistical calculations, or if the badge data indicate the group should be sub-divided into smaller 

groups. For a group that meets the tests described above, the mean dose, variance and confidence 

limits are determined. An estimated dose equal to 95% probability that the actual exposure did 

not exceed the estimate is assigned to unbadged personnel. This procedure is statistically sound 

and will insure that unbadged personnel are assigned doses much higher than the average/mean 

for the group.

(3) Dose reconstruction is performed if film badge data are unavailable for all or part of the 

period or radiation exposure, if film badge data are partially available but cannot be used 
statistically for calculations, special activities are indicated for specific individuals, or if other 

types of radiation exposures are indicated. In dose reconstruction, the conditions of exposure are 

reconstructed analytically to arrive at a radiation dose. Such reconstruction is not a new concept; 

it is standard scientific practice used by health physicists when the circumstances of a radiation 

exposure require investigation. The underlying method is in each case the same. The radiation 

environment is characterized in time and space, as are the activities and geometrical position of 

the individual. Thus, the rate at which radiation is accrued is determined throughout the time of 

exposure, from which the total dose is integrated. An uncertainty analysis of the reconstruction 

provides a calculated mean dose with confidence limits. The specific method used in a dose 

reconstruction depends on what type of data are available to provide the required 

characterizations as well as the nature of the radiation environment. The radiation environment is 

not limited to the gamma radiation that would have been measured by a film badge, but also 

includes neutron radiation for personnel sufficiently close to a nuclear detonation, as well as beta 

and alpha radiation (internally) for personnel whose activities indicate the possibility of 

inhalation or ingestion of radioactive particles.

§218.2 General procedures.

The following procedures govern the approach taken in dose determination: 

(a) Use individual film badge data where available and complete, for determining the external 

gamma dose. 

(b) Identify group activities and locations for period(s) of possible exposure.

(c) Qualitatively assess the radiation environment in order to delineate contaminated areas. If no 

activities occurred in these areas, and if no other potential for exposure exists, a no dose received 

estimate is made.

(d) If partial film badge data are available, define group(s) of personnel with common activities 

and relationships to radiation environment.

(e) Using standard statistical methods, verify from the distribution of film badge readings 

whether the badged sample adequately represents the intended group.

(f) Calculate the mean external gamma dose, with variance and confidence limits, for each 

unbadged population. Assign a dose equal to 95% probability that actual exposure did not exceed 

the assigned dose. 

(g) If badge data is not available for a statistical calculation, conduct a dose reconstruction. 

(h) For dose reconstruction, define radiation environment through use of all available scientific 

data, e.g., measurements of radiation intensity, decay, radioisotopic composition.

(i) Quantitatively relate activities shielding, position, and other factors to radiation environment 

as a function of time. Integrate dose throughout period of exposure.
(j) Where possible, calculate mean dose with confidence limits; otherwise calculate best estimate 

dose or, if data are too sparse, upper limit dose.

(k) Compare calculations with available film badge records to verify the calculated doses. 

Whether or not film badge data is available, calculate initial and internal doses where identified 

as a meaningful contribution to the total dose.

§218.3 Dose reconstruction methodology.

(a) Concept. The specific methodology consists of the characterization of the radiation 

environments to which participants through all relevant activities, were exposed. The 

environments, both initial and residual radiation are corrected with the activities of participants 

to determine accrued doses due to initial radiation, residual radiation and/or inhaled/ingested 

radioactive material, as warranted by the radiation environment and the specific personnel 

activities. Due to the range of activities, times, geometries, shielding, and weapon characteristics, 

as well as the normal spread in the available data pertaining to the radiation environment, an 

uncertainty analysis is performed. This analysis quantifies the uncertainties due to time/space 

variations, group size, and available data. Due to the large amounts of data, an automated 

(computer-assisted) procedure is often used to facilitate the data-handling and the dose 

integration, and to investigate the sensitivity to variations in the parameters used. The results of 

the gamma data calculations are then compared with film badge data as they apply to the specific 

period of the film badges and to the comparable activities of the exposed personnel, in order to 

validate the procedure and to identify personnel activities that could have led to atypical doses. 

Radiation dose from neutrons and dose commitments due to inhaled or ingested radioactive 

material are not detected by film badges. Where required, these values are calculated and 

recorded separately.

(b) Characterization of the radiological environment. (1) This step describes and defines the 

radiological conditions as a function of time for all locations of concern, that is, where personnel 

were positioned or where personnel activities took place. The radiation environment is divided 

into two standard categories—initial radiation and residual radiation.

(2) The initial radiation environment results from several types of gamma and neutron emissions. 

Prompt neutron and gamma radiation are emitted at the time of detonation, while delayed 

neutrons and fission-product gamma, from the decay of radioactive products in the fireball, 

continue to be emitted as the fireball rises. In contrast to these essentially point sources of 

radiation, there is gamma radiation from neutron interactions with air and soil, generated within a 

fraction of a second. Because of the complexity of these radiation sources and their varied 

interaction properties with air and soil, it is necessary to obtain solutions of the Boltzmann 

radiation transport equation. The radiation environment thus derived includes the effects of shotspecific parameters such as weapon type and yield, neutron and gamma output, source and target 

geometry, and atmospheric conditions. The calculated neutron and gamma radiation 

environments are checked for consistency with existing measured data as available. In those few 

cases displaying significant discrepancies that cannot be resolved, an environment based on 

extrapolation of the data is used if it leads to a larger calculated dose.
(3) In determining the residual radiation environment, all possible sources are considered 

including radioactive clouds, radiation that may have been encountered from other tests, and 

radioactive debris that may have been deposited in water during oceanic tests. The residual 

radiation environment is divided into two general components—neutron-activated material that 

subsequently emits, over a period of time, beta and gamma radiation; and radioactive debris from 

the fission reaction or from unfissioned materials that emit alpha, beta, and gamma radiation. 

Because residual radiation decays, the characterization of the residual environment is defined by 

the radiation intensity as a function of type and time. Radiological survey data are used to 

determine specific intensities at times of personnel exposure. Interpolation and extrapolation are 

based on known decay characteristics of the individual materials that comprise the residual 

contamination. In those rare cases where insufficient radiation data exist to adequately define the 

residual environment, source data are obtained from the appropriate weapon design laboratory 

and applied in standard radiation transport codes to determine the initial radiation at specific 

distances from the burst. This radiation, together with material composition and characteristics, 

leads to description of the neutron-activated field for each location and time of interest. In all 

cases observed data, as obtained at the time of the operation, are used to calibrate the 

calculations.

(c) Activities of participants. This step uses all official records, augmented by personnel 

interviews where gaps exist, to depict a scenario of activities for each individual or definable 

group. When a dose reconstruction is performed for a specific individual, information available 

from the individual is accepted unless demonstrably inaccurate. For military units, whose 

operations were closely controlled and further constrained by radiological safety monitors, the 

scenario is usually well defined. The same is true for observers, who were restricted to specific 

locations both during and after the nuclear burst. Ships' locations and activities are usually 

known with a high degree of precision from deck logs. Aircraft tracks and altitudes are also 

usually well defined. Personnel engaged in scientific experiments often kept logs of their 

activities; moreover, the locations of their experiments are usually a matter of record. Where the 

records are insufficiently complete for the degree of precision required to determine radiation 

exposure, participants' comments are used and reasonable judgements are made to further the 

analysis. Possible variations in the activities, as well as possible individual deviations from group 

activities, with respect to both time and location, are considered in the uncertainty analysis of the 

radiation dose calculations.

(d) Calculation of dose. (1) The initial radiation doses to close-in personnel (who were normally 

positioned in trenches at the time of detonation) are calculated from the above-ground 

environment by simulating the radiation transport into the trenches. Various calculational 

approaches, standard in health physics, are employed to relate in-trench to above-trench doses 

for each source of radiation. Detailed modeling of the human body, in appropriate postures in the 

trench, is performed to calculate the gamma dose that would have been recorded on a film badge 

and the maximum neutron dose. The neutron, neutron-generated gamma, and prompt gamma 

doses are accrued during such a short time interval that the posture in a trench could not be 

altered significantly during this exposure. The fission-product gamma dose, however, is 

delivered over a period of many seconds. Therefore, the possibility of individual reorientation

(e.g., standing up) in the trench is considered.
(2) The calculation of the dose from residual radiation follows from the characterized radiation 

environment and personnel activities. Because radiation intensities are calculated for a field (i.e., 

in two spatial dimensions) and in time, the radiation intensity is determinable for each increment 

of personnel activity regardless of direction or at what time. The dose from exposure to a 

radiation field is obtained by summing the contribution (product of intensity and time) to dose at 

each step. The dose calculated from the radiation field does not reflect the shielding of the film 

badge afforded by the human body. This shielding has been determined for pertinent body 

positions by the solution of radiation transport equations as applied to a radiation field. 

Conversion factors are used to arrive at a calculated film badge dose, which not only facilitates 

comparison with film badge data, but serves as a substitute for an unavailable film badge 

reading.

(3) The calculation of the dose from inhaled or ingested radioactivity primarily involves the 

determination of what radioisotopes entered the body in what quantity. Published conversion 

factors are then applied to these data to arrive at the radiation dose and future dose commitments 

to internal organs. Inhalation or ingestion of radioactive material is calculated from the 

radioactive environment and the processes of making these materials inhalable or ingestible. 

Activities and processes that cause material to become airborne (such as wind, decontamination 

or traffic) are used with empirical data on particle lofting to determine airborne concentrations 

under specific circumstances. Volumetric breathing rates and durations of exposure are used to 

calculate the total material intake. Data on time-dependent weapon debris isotopic composition 

and the above-mentioned conversion factors are used to calculate the dose commitment to the 

body and to specific body organs.

(e) Uncertainty analysis. Because of the uncertainties associated with the radiological data or 

calculations used in the absence of data, as well as the uncertainties with respect to personnel 

activities, confidence limits are determined where possible for group dose calculations. The 

uncertainty analysis quantifies the errors in available data or in the model used in the absence of 

data. Confidence limits are based on the uncertainty of all relevant input parameters, and thus 

vary with the quality of the input data. They also consider the possible range of doses due to the 

size of the exposure group being examined. Typical sources of error include orientation of the 

weapons, specific weapon yields, instrument error, fallout intensity data, time(s) at which data 

were obtained, fallout decay rate, route of personnel movements, and arrival/stay times for 

specific activities. 

(f) Comparison with film badge records. (1) Calculations of gamma dose were compared with 

film badge records for two military units at Operation PLUMBBOB to initially validate this 

methodology. Where all parameters relating to radiation exposure were identified, direct 

comparison of gamma dose calculations with actual film badge readings was possible. Resultant 

correlations provided high confidence in the methodology. 

(2) Film badge data may, in some cases, be unrepresentative of the total exposure of a given 

individual or group; nevertheless, they are extremely useful for direct comparison of incremental 

doses for specific periods, e.g., validating the calculations for the remaining, unbadged period of 

exposure. Moreover, a wide distribution of film badge data often leads to more definitive 

personnel grouping for dose calculations and to further investigation of the reason(s) for such 
distribution. In all cases, personnel film badge data are not used in the dose calculations, but 

rather are used solely for comparison with and validation of the calculations. For dose 

reconstructions accomplished to date, comparison has been favorable and within the confidence 

limits of the calculations. 

§218.4 Dose estimate reporting standards.

The following minimum standards for reporting dose estimates shall be uniformly applied by the 

Military Services when preparing information in response to an inquiry by the Veterans 

Administration, in connection with a claim for compensation, or by a veteran or his or her 

representative. The information shall include all material aspects of the radiation environment to 

which the veteran was exposed and shall include inhaled, ingested, and neutron doses, when 

applicable. In determining the veteran's dose, initial neutron, initial gamma, residual gamma, and 

internal (inhaled and ingested) alpha, beta, and gamma shall be considered. However, doses will 

be reported as gamma dose, neutron dose, and internal dose. To the extent to which the 

information is available, the responses will address the following questions: 

(a) Can it be documented that the veteran was a test participant? If so, what tests did he attend 

and what were the specifics of these tests (date, time, yield (unless classified) type, location and 

other relevant details)? 

(b) What unit was the man in? What were the mission and activities of the units at the test?

(c) To the extent to which the available records indicate, what were his duties at the test?

(d) Can you corroborate the specific information relevant to the potential exposure provided by 

the claimant to the Veterans Administration and forwarded to the Department of Defense? What 

is the impact of these specific activities on the claimant's reconstructed dose?

(e) Is there any recorded radiation exposure for the individual? Does this recorded exposure 

cover the full period of test participation? What are the uncertainties associated with the recorded 

film badge dose? 

(f) If recorded dosimetry data is unavailable or incomplete, what is the dose reconstruction for 

the most probable dose, with error limits, if available? 

(g) Is there evidence of a neutron or internal exposure? What is the reconstruction?

Upon request, the participant or his or her authorized representative will be informed of the 

specific methodologies and assumptions employed in estimating his or her dose.