Leon J . Goodman

Betty A . Torres

James C . Myska

Brian Veit

Mark Oliver

Michael Stanka

Joel Dhermain

Bruno A . Asselineau

AFRRI TR93-4Document Cleared for Public Release ; Distribution Unlimited

Armed Forces Radiobiology Research Institute

8901 Wisconsin Avenue

Bethesda, Maryland 20889-5603

April 1993

AFRRI

93-4 TECHNICAL REPORT

Reference Dosimetry

for the 1992 NATO

Battlefield Dosimetry

Intercomparison at the

Army Pulse Radiation Facility
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Reference Dosimetry for the 1992 NATO Battlefield

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AFRRI Technical Report 93-4

Reference Dosimetry for the

1992 NATO Battlefield Dosimetry Intercomparison

at the Army Pulse Radiation Facility

Leon J. Goodman'", Betty A . Torres"', James C . Myska'",

CPT Brian Veit, USA(2), Mark Oliver(2), Michael Stanka(2),

Joel Dhermain, Ph.D.(3), Bruno A. Asselineau(3)

")Armed Forces Radiobiology Research Institute, Bethesda, Maryland . ;

(2)Army Pulse Radiation Facility, Aberdeen Proving Ground, Maryland . ;

(3)Etablissement Technique Central de l'Armement, Arcueil-Cedex, France

April 1993

Radiation Biophysics Department

ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE

8901 Wisconsin Avenue

Bethesda, Maryland 20889-5603

Cleared for public release ; distribution unlimited
Supported by the Armed Forces Radiobiology Research Institute, Defense Nuclear Agency . Views presented in this paper are those of the authors ; no endorsement by the Defense Nuclear Agency has been

given or should be inferred .
Contents

Introduction	 1

Radiation Fields	 1

Measurements	 6

Results	 7

Uncertainties	 10

Comparison to Previous Data 	11

Conclusion	 12

Acknowledgement	 12

References	 13

Introduction

A NATO Dosimetry Study was performed from 23 March to 3 April 1992 at the Army

Pulse Radiation Facility (APRF), Aberdeen Proving Grounds, Aberdeen, Maryland . This

study was designed to determine the accuracy of battlefield dosimeters in a simulated

tactical environment, and to compare the different NATO dosimeters in simple, welldefined radiation fields . Dosimeters were irradiated in four radiation fields : in free air

(FIA) and on the surface of polyethylene phantoms by the APRF reactor with the core

bare and with the core shielded by a converter shield .

Measurements to provide reference dosimetry for the NATO Dosimetry Study were

made the week before and the week following the above dates and also during the

second week of the battlefield dosimeter comparisons . Reference dosimetry for the

NATO Dosimetry Study reactor irradiations was performed by three dosimetry groups .

1 . Etablissement Technique Central de l'Armement (ETCA), Arcueil-Cedex, France

2 . Armed Forces Radiobiology Research Institute (AFRRI), Bethesda, Maryland, USA

3 . Army Pulse Radiation Facility (APRF), Aberdeen, Maryland, USA

Tissue kerma rates FIA and on the surface of phantoms, normalized to reactor power,

were measured using paired ionization chambers, a neptunium fission chamber, diodes,

rhodium foils, and aluminum oxide and calcium fluoride thermoluminescent dosimeters .

For these measurements, the reactor was operated in the steady-state mode . Monitor

sulfur pellets and calcium fluoride thermoluminescent dosimeters were used to establish

tissue kerma in the pulsed environment used for the irradiation of the battlefield dosimeters .

This report contains a brief description of the radiation facilities and the measurements,

and a comparison to the reference dosimetry data of the 1986 NATO battlefield dosimetry intercomparison .

Radiation Fields

The APRF "fast burst" reactor is a bare critical assembly that may be operated at power

levels up to 10 kW and at super-prompt criticality to produce pulses with microsecond

durations (1-3) . For these studies, the reactor was operated inside the reactor silo with

the center 2 m above the floor. All dosimetry measurements were made 1 .65 m from

1
the core center and 2 m above the floor (see Figures 1 and 2) . The converter shield,

when used, was placed in front of the reactor core to better approximate the neutron

spectrum 1 km down-range of a nuclear blast . The shield, 12 .7 cm thick and 46 cm

high, was constructed of polyethylene loaded with cadmium oxide (5% by weight) .

10 cm

75oARC 102 .7 cm

CONVERTER 12 .7 cm THICK

POLYETHYLENE LOADED WITH

5 W/O CADMIUM AND .3175

cm ALUMINUM WALLS

PHANTOMS- VIRGIN POLYETHYLENE

CONVERTER -POLYETHYLENE LOADED WITH CADMIUM

REACTOR

STANDS- ALUMINUM

EXPOSURE CONFIGURATION WITH CONVERTER AND SLAB PHANTOMS -TOP VIEW

Figure 1 . Top-view diagram of irradiation configuration showing the reactor core, the converter shield,

and the polyethylene slab phantoms .

2

10 .2 cm

I

165 cm
SHROUD

34 .4

FUEL -• 22.6 I

I

1

1

165 Cm

170 cm

f61 cm

PHANTOMS- VIRGIN POLYETHYLENE

CONVERTER - POLYETHYLENE LOADED

W/CADMIUM AND .3175 ALUMINUM

WALL COVER

REACTOR

ALUMINUM STANDS

bORATED CONCRETE FLOOR

200

EXPOSURE CONFIGURATION WITH CONVERTER AND SLAB PHANTOMS- SIDE VIEW

Figure 2 . Side-view diagram of irradiation configuration showing the reactor core with the converter

shield and the polyethylene slab phantoms mounted on aluminum stands .

Spectral data for the unshielded APRF configuration were given in a previous technical

report (4) . Spectral data behind the converter shield are not available, but the qualitative

effects of the shield were to reduce the neutron component of the tissue kerma in free

air from 92% to 43%, and to produce a harder (more energetic) neutron spectrum .

(Neutrons are absorbed by the hydrogenous material, decreasing the fast neutron flux

and increasing the gamma-ray flux from hydrogen and cadmium capture reactions .

Cadmium absorbs thermal and low-energy neutrons .)

3
For the irradiations made in free air, the dosimeters were mounted on a thin aluminum

screen (see Figure 3) placed along a circular arc located at 1 .65 m from the core centerline . When the phantoms were used, the screen was removed and the phantoms

(polyethylene blocks 61 cm high, 30 cm wide, and 10 cm thick) were placed along this

Figure 3 . Ionization chambers mounted on the aluminum screen for measurements in free air .

4
same arc (see Figure 1) . The dosimeters were mounted directly on the front surface of

the phantoms (see Figure 4) . The aluminum screen, the phantoms, and the converter

shield were supported by aluminum stands (see Figure 2) .

Figure 4. Ionization chambers, the fission chambers, and sulfur pellets mounted on the front surface of

the polyethylene slab phantoms .

5
Measurements

The reactor reference dosimetry was performed with the reactor operated in the steadystate mode and was based on measurements made using a variety of dosimeters .

2 . AFRRI

(a)Tissue-equivalent and magnesium-argon paired ionization chambers (TEMg IC)

3 . APRF

(a)Tissue-equivalent and magnesium-argon paired ionization chambers (TEMg IC)

(b) Calcium fluoride thermoluminescent dosimeters (CaF TLD)

The measurement techniques applied for these dosimeters have been described in

other reports (5-12) . The measurement results obtained with these dosimeters were

transferred to the pulsed environment through the use of sulfur monitor pellets and

calcium fluoride TLDs.

6

1 . ETCA

(a)

(b)

(c)

(d)

(e)

(f)

(g)

Tissue-equivalent and magnesium-argon paired ionization chambers (TEMg IC)

Tissue-equivalent and aluminum-argon paired ionization chambers (TE-Al

IC)

Tissue-equivalent and teflon-carbon dioxide paired ionization chambers

(TE-C IC)

Neptunium fission chamber (Np FC)

Silicon diodes

Rhodium foils

Aluminum oxide thermoluminescent dosimeters (Al 2 O3 TLD)
Results

The results of the reference dosimetry measurements made in free air with the bare

core are shown in Table 1 . For this table and all subsequent tables the average total

kerma and the average percent neutron kerma are calculated from the average neutron

and gamma-ray kermas . The total kerma and percent neutron kerma calculated from

the ionization chambers are presented for information . This configuration produced the

radiation field with the largest component of neutron kerma, an average of 92% . It is

notable that the variety of neutron dosimeters listed have a standard deviation (SD) of

only 1% . The gamma-ray kermas, on the other hand, have a 13% SD, due mostly to a

low (14.7) and a high (22 .1) measurement . It is generally acknowledged that it is difficult

to reliably measure a small gamma-ray component in a mixed neutron and gamma-ray

field .

Table 1 . Reference dosimetry measurements made in free air with a bare

reactor core. All tissue kermas in mGy/(kW min) . Note that the average total

kerma and the average percent neutron kerma are calculated from the average neutron and gamma-ray kermas .

i

7

Group Dosimeter

type

Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

ETCA TE-Mg IC 204 17.8 222 92

TE-Al IC 204 17 .3 221 92

TE-C IC 207 14.7 222 93

Np FC 204 -- -- --

Diode 202 -- -- --

Rh foil 203 -- -- --

A12 03 TLD -- 17 .0 -- --

AFRRI TE-Mg IC 205 16 .1 221 93

APRF TE-Mg IC 205 17 .6 223 92

CaF TLD -- 22 .1 -- --

Average 204 17 .5 222 92

+/- 1 .5 (1%) +/- 2 .3 (13%)
Table 2 shows the results obtained when the dosimeters were mounted directly on the

front surface of the phantoms and irradiated with the bare core . The effect of the phantoms is to increase by 5% the neutron kerma due to backscattering, and to increase the

gamma-ray kerma by 2 .6 times due to neutron-energy moderation and neutron capture

by hydrogen in the phantoms . The net effect is to reduce the neutron kerma to 83% of

the total kerma . For the larger component of gamma-rays now present, the measurements gave an SD=5% .

Table 2. Reference dosimetry measurements made on the surface of the

phantoms with a bare reactor core . All tissue kermas in mGy/(kW min) . Note

that the average total kermas and average percent neutron kerma are calculated from the average neutron and gamma-ray kermas .

8

Group Dosimeter

type

Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

ETCA TE-Mg IC 214 44.4 258 83

TE-Al IC 212 46.5 258 82

TE-C IC 210 47.5 258 82

N p FC 217 -- -- --

AFRRI TE-Mg IC 218 41 .4 259 84

APRF TE-Mg IC 215 45.1 260 83

CaF TLD -- 46.8 -- --

Average 214 45.3 259 83

+/-3 .0 (1%) +/- 2 .2 (5%)
The reference dosimetry measurements made in free air with the converter shield

interposed between the core and the dosimeters are shown in Table 3 . The introduction of this shield reduces the neutron kerma to only 18% of its former value

and increases the gamma-ray kerma 2 .7 times . The neutron kerma component is

thus reduced to 43% . Even though fewer neutron dosimeters were used for these

measurements, for the neutron kerma the SD is now increased to 6% . The gamma-ray kermas, however, now have SD=1 % .

Table 3. Reference dosimetry measurements made in free air with the converter shield . All tissue kermas in mGy/(kW min) . Note that the average total

kerma and average percent neutron kerma are calculated from the average

neutron and gamma-ray kermas.

9

Group Dosimeter

type

Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

ETCA TE-Mg IC 35.1 46 .7 81 .8 43

TE-Al IC 34 .6 47.4 82.0 42

TE-C IC 34 .3 47.7 82.0 42

Np FC 33 .9 -- -- --

AFRRI TE-Mg IC 36 .9 47.5 84 .4 44

APRF TE-Mg IC 39 .3 46.0 85 .3 46

CaF TLD -- 46 .8 -- --

Average 35.7 47 .0 82 .7 43

+/-2.1 (6%) +/-0 .6 (1%)
When measurements are made on the surface of the phantoms with the converter

shield in place, the results shown in Table 4 are obtained . Compared to Table 3, there

appears to be little effect due to neutron backscattering . The gamma-ray kerma, however, is 18% larger than the FIA measurements due to neutron-energy moderation and

neutron capture by hydrogen . This configuration produces the lowest component of

neutron kerma, 39% .

Table 4. Reference dosimetry measurements made on the surface of the

phantoms with the converter shield . All tissue kermas in mGy/(kW min) . Note

that the average total kerma and the average percent neutron kerma are calculated from the average neutron and gamma-ray kermas.

Uncertainties

The precision of the ionization chamber measurements is generally better than 1 % . The

absolute accuracy of the paired ionization chamber method is considered to be around

5%-8% . When a component of the radiation field is small relative to the total kerma,

however, the accuracy for determining this component may be considerably poorer .

1 0

Group Dosimeter

type

Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

ETCA TE-Mg IC 35.8 55 .6 91 .4 39

TE-Al IC 34.8 56 .6 91 .4 38

TE-C IC 33 .7 57 .6 91 .3 37

Np FC 33 .3 -- -- --

Diode 33 .0 -- -- --

Al2O3 TLD -- 52 .0 -- --

AFRRI TE-Mg IC 37 .8 56 .5 94.3 40

APRF TE-Mg IC 39 .0 56 .4 95 .4 41

CaF TLD -- 52 .4 -- --

Average 35 .3 55 .3 90 .6 39

+/-2.3 (7%) +/-2 .1 (4%)
The reference dosimetry measurements made in free air with the converter shield

interposed between the core and the dosimeters are shown in Table 3 . The introduction of this shield reduces the neutron kerma to only 18% of its former value

and increases the gamma-ray kerma 2 .7 times. The neutron kerma component is

thus reduced to 43% . Even though fewer neutron dosimeters were used for these

measurements, for the neutron kerma the SD is now increased to 6% . The gamma-ray kermas, however, now have SD=1 % .

Table 3 . Reference dosimetry measurements made in free air with the converter shield . All tissue kermas in mGy/(kW min) . Note that the average total

kerma and average percent neutron kerma are calculated from the average

neutron and gamma-ray kermas.

9

Group Dosimeter

type

Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

ETCA TE-Mg IC 35.1 46 .7 81 .8 43

TE-Al IC 34.6 47 .4 82 .0 42

TE-C IC 34.3 47 .7 82.0 42

Np FC 33.9 -- -- --

AFRRI TE-Mg IC 36.9 47 .5 84.4 44

APRF TE-Mg IC 39.3 46 .0 85.3 46

CaF TLD -- 46 .8 -- --

Average 35.7 47 .0 82.7 43

+/-2 .1 (6%) +/-0.6 (1%)
Manufacturer's specifications for the CaF TLDs gives a precision of +/- 6 .5% for a group

of 5000 chips . Measurements at APRF using 30 chips have verified this value . The CaF

TLDs are calibrated with a Co-60 gamma-ray cell whose calibration is traceable to the

National Institute of Standards and Technology . Groups of six TLDs are irradiated with

doses ranging from 1 to 5000 Gy and the light output measured with a Harshaw 4000

reader is plotted versus dose to establish a calibration curve. Although the calibration of

the gamma-ray cell is accurate to +/- 3%, their accuracy for the reference dosimetry

measurements depends on how well the calibrator spectrum compares to the gammaray spectrum produced by the reactor . Their accuracy for the gamma rays from the

reactor is estimated to be +/- 10% .

Measurements performed at AFRRI by J . H . Musk indicate a reproducibility within +/-

4% for Al 2 O3 TLDs (13) .

Comparison to Previous Data

The results of the 1986 NATO reference dosimetry measurements were presented in a

previous technical report (4) . At that time, no measurements were made with the bare

reactor core. Table 5 compares the 1986 ionization chamber results obtained using the

converter shield with the 1992 (present) results for the configuration when the dosimeters were in free air . The difference between the two data sets is less than 2% .

Table 5 . Comparison of 1986 and 1992 reference dosimetry for

measurements made in free air using the converter shield . All

tissue kermas in mGy/(kW min) .

1 1

Data set Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

1986 35 .1 46 .5 81 .5 43

1992 35.7 47.0 82.7 43

1992/1986 1 .017 1 .011 1 .015 --
When the dosimeters were mounted on the surface of the phantoms and the converter

shield was used, the comparison of the 1986 and 1992 data sets is as shown in Table

6. It can be seen that the present data for neutron kerma are, on average, about 6%

lower than the earlier data . Examination of Table 4 shows that this is due to the low

values of neutron kerma obtained with the five ETCA dosimeters . The two values of

neutron kerma obtained by the AFRRI and APRF ionization chambers are actually a

little higher than the 1986 value. It is worth noting that ETCA did not participate in the

1986 reference dosimetry measurements.

Table 6. Comparison of 1986 and 1992 reference dosimetry

for measurements made on the phantoms using the converter

shield . All tissue kermas in mGy/(kW min) .

Conclusion

The reference dosimetry measurements described in this report serve to provide an

absolute calibration of the tissue kerma rates in the radiation fields used for the 1992

NATO battlefield dosimeter intercomparison . Although a greater variety of dosimeters

were used for these measurements than in the 1986 study, the eight dosimetry systems

gave results which agreed well among themselves as well as with past values . Thus the

kermas derived from these results can be used confidently to evaluate the accuracy of

the NATO battlefield dosimeters intercompared in 1992 .

Acknowledgement

The authors appreciate the help and cooperation provided by the APRF staff and thank

them for helping to make this project a success .

1 2

Data set Neutron

kerma

Gamma-ray

kerma

Total

kerma

% neutron

kerma

1986 37.6 54.9 92 .5 41

1992 35.3 55 .3 90 .6 39

1992/1986 0.939 1 .007 0 .979 --
References

1 . Kazi, A. H . Operation of Army Pulsed Radiation Facility reactor core III . Transactions of American Nuclear Society 14: 313, 1971 .

2 . Kazi, A. H ., Heimbach, C . R ., Harrison, R . C ., and Robitaille, H . A. Comparison of

measured and calculated radiation transport in air-over-ground geometry to 1 .6 km

from a fission source . Nuclear Science and Engineering 85 : 371-386, 1983 .

3 . Robitaille, H . A., and Hoffarth, B . E. A comparison of measured and calculated airtransported radiation from a fast, unshielded reactor. DREO Report No . 835, Defense Nuclear Establishment of Ottawa, Ottawa, 1980 .

4 . Dooley, M ., Eagleson, D . M ., Mohaupt, T. H ., Kazi, A . H ., and Zeman, G . H . Reference dosimetry for 1986 NATO battlefield dosimetry intercomparison at Army

Pulsed Radiation Facility . Technical Report TR86-4, Armed Forces Radiobiology

Research Institute, Bethesda, Maryland, 1986 .

5 . Goodman, L. J . A practical guide to ionization chamber dosimetry at the AFRRI

reactor. Contract Report CR85-1, Armed Forces Radiobiology Research Institute,

Bethesda, Maryland, 1985 .

6 . Neutron Dosimetry for Biology and Medicine . ICRU Report 26, International Commission on Radiation Units and Measurements, Bethesda, Maryland, 1977 .

7 . Broerse, J . J ., Mijnheer, B . J ., and Williams, J . R . European protocol for neutron

dosimetry for external beam therapy . British Journal of Radiology 54 : 882-898,

1981 .

8 . Waterman, F. M ., Kuchnir, F. T., Skaggs, L . S ., Kouzes, R . T., and Moore, W . H .

Energy dependence of the neutron sensitivity of C-CO 2 , MG-Ar, and TE-TE ionization chambers . Physics in Medicine and Biology 24 : 721-733 1979 .

9. Zeman, G . H ., and Ferlic, K . P. Paired ion chamber constants for fission gammaneutron fields . Technical report TR84-8, Armed Forces Radiobiology Research

Institute, Bethesda, Maryland, 1984 .

10. Ing, H ., and Cross, W. G. A criticality neutron dosimeter using the Rh-103(n,n')Rh103m reaction . Health Physics 25: 291-297, 1973 .

11 . Zeman, G . H . Rhodium-103 and indium-115 inelastic scattering reactions for fission

neutron dosimetry. Technical Report TR84-7, Armed Forces Radiobiology Research

Institute, Bethesda, Maryland, 1984 .

1 3
12 . Zeman, G . H ., and Bice, W . S ., Jr. Kerma factors for use in 37-group neutron spectrum calculations . Technical Report TR83-3, Armed Forces Radiobiology Research

Institute, Bethesda, Maryland, 1983 .

13 . Musk, J . H ., and Kearsley, E . Thermoluminescence characteristics of aluminum

oxide . In: Programme and Abstracts, 10 th International Conference on Solid State

Dosimetry, p . 43 . Nuclear Technology Publishing, Ashford, Kent, England, 1992 .

1 4
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ATTN: DIRECTOR

DEFENCE RESEARCH ESTABLISHMENT OTTAWA

ATTN: NUCLEAR EFFECTS SECTION

ETABLISSEMENT TECHNIQUE CENTRAL DE L'ARMEMENT

ATTN: CENTRE D'ETUDES DU BOUCHET

INHALATION TOXICOLOGY RESEARCH INSTITUTE

ATTN : LIBRARY

INSTITUTE OF NAVAL MEDICINE

ATTN: DEFENCE RADIOLOGICAL PROTECTION

SERVICE

INSTITUTE OF RADIOBIOLOGY

ARMED FORCES MEDICAL ACADEMY

ATTN:

DIRECTOR y

KAMAN SCIENCES CORPORATION

ATTN: DASIAC

NBC DEFENSE RESEARCH AND DEVELOPMENT CENTER OF THE

FEDERAL ARMED FORCES

ATTN: WWDBW ABC-SCHUTZ

MEDICAL BIOLOGICAL LABORATORY TNO

NCTR-ASSOCIATED UNIVERSITIES

ATTN:

EXECUTIVE DIRECTOR

NORWEGIAN DEFENCE RESEARCH ESTABLISHMENT

ATTN: DIVISION FOR ENVIRONMENTAL

TOXICOLOGY

RUTGERS UNIVERSITY

ATTN:LIBRARY OF SCIENCE AND MEDICINE

UNIVERSITY OF CALIFORNIA

ATTN : LABORATORY FOR ENERGY-RELATED

HEALTH RESEARCH

ATTN : LAWRENCE BERKELEY LABORATORY

UNIVERSITY OF CINCINNATI

ATTN : UNIVERSITY HOSPITAL, RADIOISOTOPE

LABORATORY

WEHRWISSENSCHAFTLICHE DIENSTELLE

ATTN:NUCLEAR RADIATION EFFECTS

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