ORAU Team

NIOSH Dose Reconstruction Project

Technical Information Bulletin in Support of INEEL Technical Basis 

Document Section 6: Reanalysis of Hankins MTR Bonner Sphere

Surveys

Document Number:

ORAUT-OTIB-0009 

Effective Date: 03/17/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 9

Subject Expert: Norman Rohrig

Approval: Signature on File Date: 03/16/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 03/17/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 03/17/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 2 of 9

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 11/19/2003 00-A New Technical Information Bulletin in Support of 

INEEL Technical Basis Document Section 6: 

Reanalysis of Hankins MTR Bonner Sphere 

Surveys. Initiated by Norman Rohrig.

Draft 02/17/2004 00-B Changes in response to ORAU and NIOSH 

comments. Initiated by Norman Rohrig.

03/17/2004 03/17/2004 00 First approved issue. Initiated by Norman Rohrig.
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 3 of 9

Reanalysis of Hankins MTR Bonner Ball Surveys

Norman Rohrig, November 2003

Dale Hankins made a series of neutron field measurements (Hankins 1961) with 2-, 3-, and 8-in.

diameter Bonner balls around the Idaho National Engineering and Environmental Laboratory (INEEL) 

Materials Test Reactor (MTR). The balls were covered with a cadmium shield which eliminates 

thermal neutrons below 0.41 eV. These 25 measurements were mostly around the MTR floor, but 

one was at Experimental Breeder Reactor (EBR-I) through 9 feet of concrete and three were at the 

Engineering Test Reactor (ETR). Six other measurements labeled A to F also include a thermal 

neutron component determined from the difference of a bare and a cadmium covered detector.

Since then, significant improvements have been 

made to understanding the responses of these 

detectors to neutrons of all energies and, in 

particular, intermediate neutrons. The Hankins 

data have been reanalyzed using detector 

responses calculated by Hertel and Davidson 

(1985) for 171 energy groups from thermal to 17.3 

MeV, as shown in Figure 1. This figure shows a 

higher sensitivity for the 2” and 3” detectors at low 

energies as compared to Figure 2 in Hankins’ 

(1961) paper. These calculated response curves 

are more complete than the ones available in 1961, 

particularly below 100 keV neutron energy where 

monoenergetic neutron spectra are not available. 

These calculated response matrices are for 4 x 4 

mm lithium Iodide detectors, whereas the Hankins 

data used 2 x 8 mm detectors. Hertel and

Davidson (1985) also calculated responses for 1⁄2- x 

1⁄2-in. detectors, which have larger response by 

about a factor of the detector surface area. 

Because most reactions in the detector are on the 

detector surface, the calculated neutron fluences 

and resulting dose equivalents are reduced by 

roughly a factor of 2 for all detectors and energies 

to account for the different size detectors used by 

Hankins (1961). 

Limited determination of the energy spectrum can be made with only three measurements of neutron 

response and the additional requirement that the number of neutrons in any energy region cannot be 

negative. Following Hankins (1961), the fraction of the summed response in each of the three balls 

(2-, 3-, and 8-in) is calculated for a fission spectrum, and a 1/E slowing down spectrum, which are 

expected in the reactor. The 1/E spectrum was assumed to go to 0.6 MeV and was divided into two 

sections at 10 keV. To fit the Hankins data, the fission spectrum of the form 

TE

e

E

E /

2/1

1 )( − φ =

was used above 0.6 MeV. For a typical fission spectrum, the characteristic temperature T is about 1.3 

MeV, which was used in these calculations. The measured detector responses were expressed as a 

Figure 1: Energy response of Bonner spheres

0.00

0.05

0.10

0.15

0.20

0.25

1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

Energy (eV)

Response 4 x4 mm (count cm^2/neutron)

2.0 INCH 

3.0 INCH 

8.0 INCH 
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 4 of 9

linear combination of the three spectra 

(1/E 0.4 eV - 10 keV, 1/E 10 - 600 keV, 

and fission) and solved.

To determine the dose equivalent, one 

must multiply the fluence at each energy 

by the dose equivalent conversion factor 

(DECF) for that energy and add them up 

(i.,e. do an integral). The official 

tabulations provide conversion factors at 

limited energy values. Ing and Makra 

(1978) provide a parameterization for 

dose equivalent with energy which we 

use here. Figure 2 compares this 

parameterization to Monte Carlo 

calculations reported in NCRP38 and by 

Auxier et al. (1968). 

Table 1 lists measured values from 

Hankins (1961) and the resulting 

reanalyzed dose equivalents from the 

three energy groups and their sum. 

Table 2 provides the same information 

for the six locations where Hankins provided 

thermal measurements. The thermal values are 

as reported by Hankins. Figure 3 shows the 

resulting neutron spectra for locations 3 and 23, 

which have higher doses and nearly the 

maximum low energy and fission components, 

respectively.

Another survey at the MTR measured the fast 

neutron field (Sommers 1959a) using a Rudolf 

counter described as “a dose rate instrument, 

sensitive to neutrons in the range of 0.2 to 

above 10 MeV” (Sommers 1959b). This has 

been remembered as a detector about the size 

of a soda can with the end painted red. The 

Dennis and Loosemore (1961) detector shown 

in Figure 4 is believed to be what was known as 

the Rudolf counter at the INEEL. It detects 

hydrogen recoils and has a threshold energy of 

about 0.2 MeV. Using 0.2 MeV as the division 

line between fast and intermediate neutrons to 

correspond to that for this instrument, Figure 5 

shows the correlation of the recalculated 0.4 eV 

– 10 keV and 10-200 keV neutron dose 

equivalent rates with the recalculated fast 

neutron dose equivalent rates along with the correlation of the Hankins analysis intermediate to fast 

neutron dose equivalents. Also shown are the Hankins thermal dose equivalent rates compared to 

the recalculated fast neutron dose equivalent rate. 

Figure 3: Sample MTR Spectra from Reanalyzing 

Hankins Bonner Measurements

10

100

1000

10000

100000

1.E-01 1.E+01 1.E+03 1.E+05 1.E+07

Energy (eV)

Number Neutrons

Position 23

Position 3

Figure 2: Neutron Dose Equivalent Conversion 

Factor (nrem cm^2)

0

1

10

100

1.E-7 1.E-5 1.E-3 1.E-1 1.E+1 1.E+3

Neutron Energy (MeV)

DECF (nrem cm^2)

IngMakra p11

Auxier

Based on NCRP 38
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 5 of 9

Table 1. Data from Hankins (1961) and reanalyzed doses.

Location* Total CPM

% Total CPM Dose equivalent (mrem/hr)

2" 3" 8" 0.4 eV-10 

keV

0.01-0.6 

MeV Fission >0.4 eV

EBRI 1 2935.6 31.3 49.9 18.8 0.208 0.337 0.300 0.845

2 491.7 30.7 49.5 19.8 0.034 0.064 0.066 0.163

3 2854.1 31 49.4 19.6 0.200 0.304 0.462 0.966

4 2966.2 30.6 49.1 20.3 0.204 0.340 0.556 1.100

5 2315.2 29.8 49.1 21.1 0.152 0.375 0.350 0.876

6 1698.5 29.9 48.9 21.2 0.113 0.247 0.314 0.674

7 611.1 29.6 48.4 22 0.040 0.084 0.149 0.273

8 981.9 29.5 48.3 22.2 0.064 0.136 0.250 0.450

9 5539 28.9 48.2 22.9 0.350 0.933 1.338 2.621

10 662.9 29.1 48.1 22.8 0.042 0.100 0.176 0.319

11 5700.6 29.5 48 22.5 0.376 0.699 1.699 2.773

12 7603.3 28.7 47.9 23.3 0.477 1.257 2.056 3.789

13 6185.1 29.3 47.8 23 0.404 0.763 2.005 3.172

14 2694.6 29.6 47.6 22.8 0.180 0.258 0.973 1.411

15 3116.3 28.9 47.5 23.6 0.200 0.412 1.071 1.683

16 8877.1 28.5 47.4 24.1 0.556 1.337 3.021 4.913

17 2083.4 29 47.3 23.7 0.135 0.241 0.787 1.163

18 2581.4 28.4 47.3 24.3 0.161 0.391 0.904 1.456

19 1627.8 28.1 47.3 24.6 0.099 0.275 0.548 0.923

20 4848.1 28.2 46.8 25 0.302 0.665 2.008 2.975

ETR 21 111.3 27.7 46.5 25.8 0.007 0.017 0.048 0.072

22 1494.8 28.2 46.2 25.6 0.094 0.158 0.750 1.003

23 4427.9 26.5 45.6 28 0.252 0.774 2.295 3.321

ETR 24 199.2 26 44.7 29.3 0.011 0.032 0.124 0.166

ETR 25 210.9 25 44 31.1 0.011 0.038 0.144 0.193

Total 4.672 10.236 22.393 37.300

* At MTR, unless otherwise noted.

Table 2: Data from Hankins (1961) with thermal neutron measurement and reanalyzed doses.

Location

CPM Dose Equivalent (mrem/hr)

2" 3" 8" Thermal 0.4eV-10 keV 0.01-0.6 MeV Fission Neutron

A 877 1473 700 0.11 0.19 0.56 0.67 1.42

B 196 317 153 0.02 0.04 0.07 0.24 0.35

C 778 1293 647 0.09 0.17 0.41 0.87 1.46

D 3518 5973 3108 0.32 0.76 2.33 3.94 7.03

E 3482 5734 3064 0.21 0.77 1.50 5.36 7.64

F 557 925 470 0.06 0.12 0.29 0.67 1.08

The R2

 values shown on the trendlines are the fractions of the variance which is explained by the 

lines. For the reanalyzed data the R2 values are 92% and 86% compared to only 50% for the original 

Hankins analysis, demonstrating that the reanalysis is a better fit. The slope of the trendlines is a 

dose weighted average of the ratios for the various datapoints. Table 3 provides comparisons of the 

different components of the neutron dose equivalent rate.
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 6 of 9

 

 Table 3: Ratio of Neutron Dose Equivalent Rates at the MTR

Fast

Thermal 0.4 eV-10 keV

Fast

10-200 keV

Fast Fast

Total

Trendline 0.046 0.140 0.131

Average 0.071 0.175 0.147 1.393

Minimum 0.033 0.066 0.073 1.226

Maximum 0.108 0.408 0.259 1.738

St Dev 0.025 0.074 0.050 0.121

The Interactive RadioEpidemiology Program (IREP), which calculates the probability of causation, 

uses certain neutron energy groups. The dose equivalent rate and the fraction of the dose equivalent 

in each of these regions are listed in Table 4. For the numbered locations where no thermal neutron 

value is available, the average value for the ratio of thermal to fast of 0.71 from Table 3 is assumed. 

The NTA neutron dosimeters in use when the MTR was operating at the INEEL respond only to 

neutrons above 0.5 to 0.8 MeV. For the MTR spectra, Table 4 lists the fraction of neutron dose 

equivalent above 0.8 MeV, which would be picked up by the NTA film, that varies from 35% to 66% 

depending on the location with a mean of 0.52 and a standard deviation of 0.08. The remainder of 

these low-energy neutron fields would probably not be detected by the NTA film because of its 0.8-

MeV threshold. 

Sommers (1962) reported thermal and fast neutron dose equivalent rates and gamma dose rates 

around the MTR beam lines. The thermal measurements near beams are believed to not be 

representative of the workplace. Figure 6 shows the correlation of fast neutron dose equivalent to the 

gamma dose for these measurements. The fast neutron component was insignificant for several of 

the measurements. These points are shown by the triangles at 0.5 mrem/hr, 1⁄2 of the smallest 

Figure 4: Dennis/Loosemore detector 

known as the Rudolf at the INEEL with its 

energy response curve.

Figure 5: MTR Neutron Field Components

y = 4.633x

R2

 = 0.501

y = 0.046x

R2

 = 0.715 y = 0.140x

R2

 = 0.918

y = 0.131x

R2

 = 0.860

0.01

0.1

1

10

100

0.01 0.1 1 10

Fast Neutron Dose Equivalent Rate (mrem/hr)

Dose Equivalent Rate (mrem/hr)

InterHankins

Inter .5 eV-10 keV

Inter 10-200 keV

ThermalNeut
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 7 of 9

measured value. Including all the data 

in a Shapiro and Wilk W test for 

normality (Gilbert 1987) gives a slightly 

better result for a normal distribution 

rather than a lognormal distribution. The 

fast neutron dose is 0.419 ± 0.347 times 

the gamma dose. Combining these 

results, with those of Table 3, one can 

conclude that the ratio of total neutron 

dose equivalent to gamma dose is 0.58 

± 0.48 at the 1 sigma level. The 

variation on the relative neutron 

components are buried by the fast 

neutron to gamma variation.

The IREP program uses equivalent dose 

as defined in ICRP 60 using the 

radiation weighting factor wR defined in 

Table 1 of ICRP 60. The NIOSH 

External Dose Implementation Guide 

(NIOSH 2002) provides conversion 

factors from ambient dose equivalent 

H*10 to the organ equivalent dose, and 

ICRP 74 (ICRP 1995) provides a 

calculation of the ambient dose 

equivalent so we can construct the 

ambient dose equivalent for these 

spectra. The ratios of the neutron 

ambient dose equivalent (ICRP 74) and 

the neutron dose equivalent (NCRP 38) 

are shown in Table 4. For the 10 to 100 

keV energy group, the ratio of the 

ambient dose equivalent to the neutron 

dose equivalent is 1.08, while for the 

above 2 MeV energy group it is 1.121. 

For the groups where the spectra are 

not simple multiples of each other there 

is some variation with location.

Figure 7: ICRP 74 Ambient Dose Equivalent 

Conversion Factor 

0

1

10

100

1.E-7 1.E-5 1.E-3 1.E-1 1.E+1 1.E+3

Neutron Energy (MeV)

CF (nrem cm^2)

Dose Equiv

ICRP 74

AmbDE

Figure 5: Correlation of Fast Neutron Dose Equivalent to 

Gamma Dose at MTR

0.1

1

10

100

0.1 1 10 100 1000

Gamma DE Rate(mrem/hr)

Fast Neutron DE Rate (mrem/hr)

Fast Neutron

Fast Insig

Fit 0.419 *Gamma
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 8 of 9

Table 4: Distribution of MTR neutron dose equivalent among IREP energy groups, NTA response, 

and ratio of equivalent dose to dose equivalent.

Location*

IREP Energy Interval

Fractional

NTA 

Sensitivity

>0.8 MeV

Dose Equivalent Fraction Dose Equiv

Ambient Dose Equiv

>10keV

10 keV100 keV100 keV2 MeV 2M-20M <10keV100 keV2 MeV Spectrum

EBRI 1 0.28 0.079 0.47 0.17 0.89 1.38 1.14 0.32

2 0.24 0.077 0.49 0.19 0.88 1.37 1.14 0.36

3 0.24 0.062 0.47 0.23 0.87 1.35 1.13 0.42

4 0.22 0.061 0.47 0.24 0.87 1.35 1.13 0.45

5 0.21 0.084 0.51 0.19 0.88 1.37 1.16 0.35

6 0.21 0.072 0.50 0.22 0.87 1.36 1.15 0.41

7 0.19 0.060 0.49 0.26 0.86 1.34 1.14 0.48

8 0.19 0.059 0.49 0.26 0.85 1.34 1.14 0.49

9 0.18 0.070 0.51 0.24 0.86 1.35 1.15 0.45

10 0.18 0.062 0.50 0.26 0.85 1.34 1.14 0.49

11 0.18 0.049 0.48 0.29 0.84 1.33 1.13 0.54

12 0.17 0.065 0.51 0.26 0.85 1.35 1.15 0.48

13 0.17 0.047 0.48 0.30 0.84 1.32 1.13 0.56

14 0.17 0.036 0.46 0.33 0.83 1.31 1.12 0.61

15 0.17 0.048 0.49 0.30 0.83 1.32 1.13 0.56

16 0.16 0.053 0.50 0.29 0.83 1.33 1.14 0.54

17 0.16 0.040 0.48 0.32 0.82 1.31 1.12 0.60

18 0.16 0.052 0.50 0.29 0.83 1.33 1.14 0.55

19 0.15 0.058 0.51 0.28 0.84 1.34 1.14 0.52

20 0.15 0.044 0.49 0.32 0.82 1.32 1.13 0.59

22 0.15 0.031 0.47 0.35 0.80 1.30 1.12 0.66

23 0.13 0.045 0.50 0.33 0.80 1.32 1.14 0.61

A 0.20 0.075 0.51 0.22 0.92 1.36 1.20 0.41

B 0.17 0.039 0.47 0.32 0.92 1.31 1.17 0.59

C 0.17 0.055 0.49 0.28 0.92 1.33 1.19 0.52

D 0.15 0.065 0.52 0.27 0.92 1.34 1.20 0.50

E 0.13 0.039 0.49 0.34 0.91 1.31 1.18 0.64

F 0.16 0.052 0.49 0.29 0.92 1.33 1.19 0.55

ETR 21 0.14 0.046 0.49 0.32 0.81 1.32 1.13 0.59

ETR 24 0.12 0.037 0.49 0.35 0.78 1.30 1.13 0.65

ETR 25 0.11 0.038 0.50 0.35 0.77 1.30 1.13 0.65

Average 0.18 0.056 0.49 0.28 0.86 1.33 1.15 0.52

Minimum 0.13 0.031 0.46 0.19 0.80 1.30 1.12 0.35

Maximum 0.24 0.084 0.52 0.35 0.92 1.37 1.20 0.66

St Dev 0.030 0.013 0.015 0.043 0.037 0.020 0.026 0.08

* At MTR unless otherwise noted.
Effective Date: 03/17/2004 Revision No. 00 Document No. ORAUT-OTIB-0009 Page 9 of 9

REFERENCES

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Chapter 6 in Radiation Dosimetry, Second Edition. Vol. 1 Fundamentals, -F. H. Attix and W. C. 

Roesch Ed. Academic Press, New York

Dennis, J., and W. R. Loosemore, 1961, A Fast-Neutron Counter for Dosimetry, Selected Topics in 

Radiation Dosimetry (Proc. Symp., Vienna, 1960), IAEA Vienna (1961) 443.

Gilbert, R. O., 1987, Statistical Methods for Environmental Pollution Monitoring, Van Nostrand 

Reinhold, New York.

Hankins, D. E., 1961, A Method of Determining the Intermediate Energy Neutron Dose, IDO-16655, 

Idaho Falls, Idaho.

Hertel, N. E. and J. W. Davidson, 1985, “The Response of Bonner Spheres to Neutrons from Thermal 

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Ing, H. and S. Makra, 1978, Compendium of Neutron Spectra in Criticality Accident Dosimetry, IAEA 

Technical Report Series No. 180, Vienna, Austria.

ICRP (International Commission on Radiological Protection), 1995, Conversion Coefficients for use in 

Radiological Protection against External Radiation, ICRP Report 74.

NCRP (National Council on Radiation Protection and Measurements), 1971b, Protection against 

Neutron Radiation, NCRP Report 38.

Sommers, J. F. 1959a, Neutron Activity Levels at the MTR due to ANP-1, letter to R. L. Doan (site 

manager), Som-74-59A, August 27.

Sommers, J. F. 1959b, Monitoring Problems Concerning Criticality Hazards, letter to J. R. Huffman, 

Som-14-59A, February 25.

Sommers, J. F. 1962, Thermal Column Measurements, letter to W. H. Burgus, Som-126-62A, 

December 18.