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PUBLICATIONS

NBS SPECIAL PUBLICATION 633

U.S. DEPARTMENT OF COMMERCE/National Bureau of Standards

Procedures for Calibrating

Neutron Personnel Dosimeters

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lb. 633

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NATIONAL BUREAU OF STANDARDS

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Procedures for Calibrating

Neutron Personnel Dosimeters

ATIOMAL BVKBAV

or arAm>AKS8

LIBRART JUL 7 T982

Ho-L o.c.c .^ O re.

no. (oZZ

R. B. Schwartz and C. M. Eisenhauer

Center for Radiation Research

National Measurement Laboratory

National Bureau of Standards

Washington, DC 20234

Sponsored by:

Division of Operational Safety

U.S. Department of Energy

Washington, DC 20585

U.S. DEPARTMENT OF COMMERCE, Malcolm Baldrige, Secretary

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued May 1982
Library of Congress Catalog Card Number: 82-600543

National Bureau of Standards Special Publication 633

Nat. Bur. Stand. (U.S.), Spec. Publ. 633, 35 pages (May 1982)

CODEN: XNBSAV

Disclaimer: Certain trade names and company products are mentioned in the

text or identified in an Illustration in order to adequately specify the experimental procedure and equipment used. In no case does such

identification Imply recommendation or endorsement by the

National Bureau of Standards, nor does it imply that the products are necessarily the best available for the purpose.

U.S. GOVERNMENT PRINTING OFFICE

WASHINGTON: 1982
Foreword

This report was prepared as part of the U.S. Department of Energy

Neutron Dosimetry Upgrade Program, and is intended as a guide to recommended

procedures for calibrating neutron personnel dosimeters and other neutron

protection instruments. Although intended specifically for use by DOE

laboratories, the procedures should be generally applicable in any

laboratory doing these types of calibrations.

We would like to thank the U.S. Department of Energy, and in

particular, Mr. Edward J. Vallario, Division of Ooerational Safety,

DOE, for supporting this work.

i i i
Abstract

Procedures are given for routine testing and calibration of neutron

dosimeters and remmeters with radioactive neutron sources. The issues

addressed include: the choice of neutron source; phantom construction;

fluence to dose eouivalent conversion; and corrections for air scatter,

room return, and anisotropic-neutron emission. Explicit, semi-empirical

,

analytic expressions are given for the room return correction, and

calculated numerical values are given for air scatter.

Key WDrds: air scatter; calibration; californium; dose equivalent;

dosimeter; neutron; rerrmeter; room return.

iv
CONTENTS

Page

Foreword iii

Abstract iv

I. Introduction 1 II. Neutron Sources 2

III. Phantoms 4

IV. Source-to-Detector Distance 5

V. Fluence to Dose Equivalent Conversion 7

VI. Corrections 9

A. Principles 9

B. Air Scattering 10

C. Room Return 12

D. Anisotropic Emission and Scattering by

Source Encapsulation 17

E. Summary of Correction Procedures 18

VII. Conclusions 19

VIII. Acknov;ledgcments 19

References 20

Appendix 25

Figures 1-3 27-29

V

Procedures for Calibrating Neutron Personnel Dosimeters

R. B. Schwartz and C. n. Eiserihauer

U.S. National Bureau of Standards

Washington, D.C. 20234

I. Introduction

In this report we shall discuss a technique for routine calibration

and testing of neutron dosimeters and remmeters with radioactive neutron

sources. For the sake of simplicity, we will generally refer to dosimeter

calibrations, but most of the procedures apply equally well to active

instruments, such as remmeters or Bonner spheres. In principle, it is a very simple procedure: the dosimeter is placed at a convenient distance

from a neutron source of known emission rate, and irradiated for a known

time. From the emission rate of the source, the distance, and the time,

the neutron fluence at the dosimeter is calculated. Using a conventional

f uerice-to-dose-equi val ent* conversion factor, we calculate the dose

equivalent to which the dosimeter has been exposed. The dosimeter is

then processed, and tne reading corrected for the background from air

and room scattering. The calculated dose equivalent, divided by the

corrected reading, is then the dosimeter calibration factor for that source.

*Note Concerning Units . The special SI unit for dose equivalent is

the si evert (Sv)

.

1 Sv = 1 J kg"^ = 100 rem.

The sievert, however, is not in general use in DOE laboratories, and

the authors are not aware of any instrument calibrated in sieverts.

Hence, to avoid confusion, we shall use rem (or millirem) in this

report.

Similarly, it would seem to be needless pedantry to refer to the wellknown 9-inch spherical remmeter as a "22.86 cm spherical remmeter."

1
In general, tfiere are several possible ways of performing each of these

steps, leading to dosimeter calibrations which are not unique. This report

recommends detailed procedures to be followed in performing the calibrations.

Our goal is to outline "correct" procedures; at least as far as our present

knowledge allows. It is more important, however, that everyone use the same procedure, so that dosimeter calibrations will be a function only of the

dosimeter type and the source energy spectrum, and not depend upon such factors

as the source-detector distance or the room size. This is the only way in

which dosimeters from different laboratories can be fairly compared, and the

only way in which new dosimeter types can be sensibily evaluated.

II. Neutron Sources

The recent ISO TC85 draft standard^ proposes four neutron sources: D^O- QrOOCOO/ll 9/11 moderated Cf, Cf, Am-B, and Am-Be. Other sources which have been

used for calibrations include PuBe, Pu-Li , and Am-Li, although these sources may no longer be generally available from suppliers. • .

Leaving aside, for the moment, the moderated californium source, the

252

clear winner among the other sources is Cf. It has the following advantages

as compared with the other sources: the neutron spectrum, which is similar to

235 2 3 4

that from U fission, has been carefully evaluated and is very well known; '

'

sources are available in any reasonable strength, are physically small '

(approaching a point source) and relatively lightly encapsulated; the neutron

emission is close to isotropic; and the gamma contamination is lower than for

any of the other sources.

252 The principal disadvantage of Cf is its relatively short half-life

(2.6 years), whi ch, requi res that it be replaced periodically. While this can

be a problem in some laboratories, it should not be a serious handicap to DOE

laboratories.

2
252 252

Whether to use bare Cf or moderated Cf is determined by the type of

dosimeter to be calibrated and the neutron spectrum to which it (and the

wearer) are expected to be exposed. (Note that in this report "moderated

252 californium" will always refer to a Cf source in the center of a D^O

sphere of 15 cm radius,^ covered with 0.020" thick cadmium). For calibrating

albedo dosimeters, particularly if they are to be used in the relatively soft

spectra found in reactor environments, moderated californium is clearly preferred.

We have shown^ that for other common dosimeter types (film, CR-39, polycarbonate)

there are not very great differences (less than a factor of 2) in calibration

252 252 factors between bare and moderated Cf. Thus, the moderated Cf is the more versatile of the two sources, since it is appropriate for all of the

commonly used dosimeter types. It is also the only neutron source given in

the July 1981 Draft Standard developed by the Health Physics Society Standards

Committee (HPSSC), "Criteria for Testing Personnel Dosimetry Performance." On

the other hand, if workers are exposed primarily to neutrons with fission

energy (or higher energy), it is simpler and more logical to use a bare

californium source. This report will thus consider calibrations done with

both bare and moderated californium. Discussion will be limited to these two

sources, although most of the arguments can be readily extended to other

sources.

In any case, the source should be accurately calibrated before being put

into use. NBS calibrates sources to an accuracy of + 1.2%, (la) for emission

6 9

rates between ^ 10 n/s and v. 5 x 10 n/s.

For calibrations in a closed room, the source should be placed at, or near, the center of the room. For outdoor calibrations, the source should be

placed as high off the ground as can reasonably be done. Analytic expressions

are given in Section VI of this report for predicting the effects of scattered

neutrons in these configurations.

6,7

3
III. Phantoms

Albedo dosimeters obviously must be mounted on a phantom and, following

the HPSSC Draft Standard, we recommend that all dosimeters be irradiated on a •f

phantom. ' -

'

8 , , , • While Hankins found no difference in albedo dosimeter readings for

irradiations using phantoms of polyethel ene , Lucite (poly (methyl methacrylate) )

,

or water, the HPSSC Draft Standard explicitly recommends methyl methacrylate.

This recommendation was made largely because methyl methacrylate (CH2=C(CH2)C00CH2)

is closer in chemical composition to tissue than is water. (In addition, a Lucite phantom is less likely to become a puddle on the floor.) ''

.

g Hankins found, however, that albedo dosimeter readings did depend on the

9 • shape of the phantom, and Nash and Johnson showed that the readings would oe too low if the dosimeter were placed too close to ihe edge of the phantom face. : To permit simultaneous irradiations of several dosimeters, the HPSSC Draft

Standard therefore specifies that the phantom be a rectangular parallelepiped,

40 cm X 40 cm X 15 cm thick (rather than the earlier specified 30 cm x 30 cm x 1 5 cm)

and that no portion of the dosimeter be closer than 10 cm to the phantom edge.

This is a safe practice, but it may be unduly conservative, since the experi9 10 mental evidence ' indicates that the "edge effect" is determined only by the

position of the sensitive element in the albedo dosimeter (e.g., the TLD chip

itself) . "

*

We recommend that phantoms be built to the specifications of the HPSSC

Draft Standard: 40 cm x 40 cm x 15 cm Lucite or Plexiglas. Temporarily,

existing phantoms of water or polyethelene, no smaller than 30 cm x 30 cm x 15 cm, may be used, but they should be replaced with the recommended phantom

as soon as possible. Cylindrical phantoms should not be used, even on a

"

temporary basis.

I

4
In any case, the phantom should be placed on a low mass stand for minimum

scattering, and should be the same height off the floor as the source.

IV. Source-to-Detector Distance

There are two questions to be considered: first, between what two points

is the distance to be measured; and, second, is there an optimum (or, at

least, a preferred) value for this distance?

The answer to the first question is: the source-detector distance is to

be taken as the distance from the center of the source to the front face of

the phantom directly behind the dosimeter. If several dosimeters are to be

irradiated at the same time on the same phantom, the sensitive elements

should be on the periphery of a circle centered on the face of the phantom, so that the dosimeters are all equidistant from the source. (The face of the

phantom should, of course, be perpendicular to the line joining the source and

phantom centers.) If this is not possible, explicit corrections should be

made for the variation of the distance from the source for each dosimeter.

The fact that the center of the source should be one of the end points

252 seems obvious in the case of small, bare, Cf sources. For moderated

californium sources, Ing and Cross^^ have shown that the center of the sphere

is also the appropriate measuring point, (i.e. the variation of dose rate with

distance follows the inverse square law) for distances greater than 30 cm (one

sphere diameter).

Measuring to the front face of the phantom is simply a convention, which

has the great advantage of being well defined and easily reproduced. Since

the phantom is, however, part of the albedo system, the distance to use for

calculating the fluence is somewhat ambiguous. An "effective" depth is sometimes

1

5
defined as the point in the phantom such that, when distance is measured to

2 12 13

that point, the response follows the 1/r law. While there are data '

which are consistent with an effective depth of 1 to 3 cm for albedo dosimeters

irradiated with a Cf source, we prefer to define the distance to the face of

the phantom. If definitive data were available, the calibration factor could

be corrected to what it would be at great distances; i.e., where neutrons are normally incident on the face of the phantom. In the absence of such data,

there will thus be some ambiguity in the calibrations, although the problem

is minimized if all measurements are made at the same distance.

The answer to the question of the optimum distance for irradiation is a compromise between conflicting requirements. Too great a distance leads to

very large room scatter corrections (especially for albedo dosimeters) as well as, possibly, inconveniently long irradiation times. Too short a distance

leads to non-uniform illumination of the phantom and serious departure from

2 the 1/r law for the moderated source. With all of the above in mind, the

HPSSC Draft Standard recommends a 50 cm source-to-phantom distance, and we suggest that this recommendation be followed. ^

For a spherical detector (e.g., Bonner sphere, spherical remmeter) and a 14 15 point neutron source, Axton and Harrison have shown that one should again

measure to the center of the sphere, but that the effective fluence is

increased by a factor of approximately [1 + (r/d)'^/6], where r is the radius

of the sphere, and d is the distance between the source and the center of the

detector. Thus, for a source-detector spacing equal to the diameter of the

sphere, r/d =0.5 and the correction factor is 1.04. For a 12" Bonner

sphere, a 50 cm spacing would make the correction almost completely negligible

(1 1/2%). For an Andersson-Braun remmeter, we have found that the inverse

square law is obeyed down to d = 35 cm, where the long axis of the remmeter

is perpendicular to the line between the source and the center of the

remmeter, and the distance is measured to the remmeter axis.

6
We are not aware of either calculations or systematic measurements for a spherical source and a spherical detector. In the absence of such data, we would recommend as a rul e-of-thumb that the center- to-center distance be

greater than the sum of the two sphere diameters.

V. Fluence to Dose Equivalent Conversion

The dose equivalent for a particular spectrum is calculated by con- voluting the spectrum with tne recommended energy-dependent f 1 uence-to-doseequivalent conversion factors. .

'

252

For calculations involving the bare Cf spectrum, we recommend the

2 3 Grundl -Ei senhauer spectrum evaluation. ' This evaluation seems to be

generally accepted by the community of californium users, and calculations

based on this spectrum (in areas such as reactor materials dosimetry) nave generally given excellent agreement with measured values. "'^ The evaluated

spectrum is given in the Appendix, along with a 44-group representation

generated from the evaluation.

A recent re-calculation by Ing and Cross''' of the moderated Cf spectrum

gives results which are in substantial agreement with their 19/7 calculation.''''

We should like to thank Dr. Ing for permission to use his latest values

prior to their formal publication.

The existence of several different "recommended" sets of fluence to

dose equivalent conversion factors has been duly noted, and deplored, in

18 19 20

the literature. ' ' The position within the DOE community was given

21

by Hankins in 1 977, when he recommended that the procedure in iNCRP

22

Report No. 38 be followed, including the recommendation tor linear

23 interpolation between neighboring values. Recent calculations by Cross and

by Chilton, however, give values whicn are more consistent with the

ICRP recommendations (which specify log-log interpolation) than those

of the NCRP. We assume that these recent results are more accurate than

the older ICRP or NCRP conversion factors. The agreement between the new calculations, and the ICRP recommendations, is taken as a verification

7
of the I CRP- recommended conversion factors. We therefore recommend that

the ICRP procedure be followed, including log-log interpolation. This

has the further advantages of eliminating the non-physical cusps in the curves of conversion factor as a function of energy (see Fig. 1 of ref. 18)

and of putting us in consonance with our European colleagues. We do not

23 24 recommend use of the Cross (or the Chilton ) values, because we do not

feel that it is appropriate at this time to introduce yet another set of

'recommended" conversion factors.

252 Using the ICRF conversion factors, the Grundl -Ei senhauer Cf

spectrum, and the 1981 Ing and Cross spectrum for moderated californium,

we recommend the following conversion factors:

252

Bare "^^^Cf:

-5 2 Dose Equivalent Conversion = 3.33 x 10 mrem cm ;

_5

Dose Equivalent Rate = ^'^^^l

""^— x 3600

r -3 2

= 9.54 X 10 Q/r mrem/hr.

Moderated ^^^Cf:

-6

Dose Equivalent Conversion = 9.3 x 10 mrem cm";

q o Y in~^ ? Dose Equivalent Rate = — x 3600 x 0.89 Q/r

-3 2 = 2.37 X 10 Q/r mrem/hr,

where Q is the bare source emission rate, in neutrons/second

r is the source-detector distance in centimeters, as defined in §s 4.

(The factor 0.89 in the expression for the dose equivalent rate from the

moderated source takes into account the a, 11% loss of neutrons which are moderated in the D^O to below the cadmium cut-off.)

8
The fluence to dose equivdlent conversion factor for bare Cf is not

a sensitive function of credible variations in the spectrum shape. We have

calculated that the uncertainty in the dose equivalent conversion, due to

the uncertainty in the spectrum shape, is 0.5\- {Ic). Thus, the uncertainty

in the dose equivalent rate v;ill be determined mainly by the uncertainty in

the value of Q.

252 16

While the bare Cf spectrum has been verified experimentally, the

moderated spectrum has not. Although for the moderated californium, the dose

equivalent conversion is not a sensitive function of the spectral shape, the

lack of experimental confirmation of the shape leads us to recommend that an 252 uncertainty of 7', be assigned to the conversion factor for moderated Cf.

VI. Corrections

A. Principles

We have thus far tacitly assumed an ideal irradiation facility in

2 free space, so that dividing the source emission rate by 4-r gives the flux

density at the dosimeter, and multiplying by the appropriate conversion

factor gives the dose equivalent rate. In fact, we refer to these quantities

as the "free field flux density" and the "free field dose equivalent rate,"

respectively, and their time integrals as the "free field fluence" and "free

field dose equivalent." That is, they are the quantities which would exist

if the irradiations were done in free space with no background due to air and room scattering and no source asymmetry. In practice, air scattering

generally amounts to only a few percent, and the source asymmetry may be very

small. However, since the albedo of fast neutrons from concrete and other

2 6 building materials is greater than 0.5, the contribution of room-reflected

neutrons to the response of the dosimeter may be significant, particularly

if the dosimeter is sensitive to the low energy neutrons resulting from room scatter.

9
These scdttered neutrons have a different spectrum and a different

variation with distance from the source. Therefore, they must not be considered

a proper part of the calibration field, but should ratfier be considered a type

of background, and appropriate corrections made. This follows from our fundamental point of view: the calibration factor should be a unique property

of the dosimeter type and the neutron source spectrum, and should not be a function of the characteristics of the calibration facility. Thus, all calibrations should refer to the "free field" quantities, so that calibrations of

the same dosimeter at different laboratories will give the same result, within

the experimental uncertainties.

Several authors have been concerned with these problems for many years.

27

For example, Schraube et al . discussed them in their 1972 paper, and many

laboratories have appropriate computer codes for dealing with them. As an alternative to computer calculations, the following paragraphs suggest a semi-empirical approach to making these corrections. While this approach

makes use of computer calculations, it offers physical insight into the

problem without, in many cases, requiring further significant calculations.

B. Air Scattering

Air transport calculations have very recently been made by

28

R. C. McCall using the Morse Monte Carlo code. This method is more appropriate for the problem than the earlier moments calculation of

29 Simmons and Eisenhauer which we previously used to estimate corrections for

air scattering.

The results show that air in-scattering is approximately twice the outscattering, so that the net effect is always to increase the fluence at the

detector. The inscattered spectrum is, however, shifted to lower energies

due to energy dependence and kinematics of elastic-scattering in nitrogen

and oxygen. The effect of the in-scattered neutrons is represented by an

10
integral over the energy and angular response of the detector. The

recommended corrections, listed in Table I, are derived from McCall's air

27 30

scatter calculations and Hankins energy response measurements/ An

estimated factor of 0.8 has been applied to the in-scattered neutrons, for

the two dosimeter listings in the Table to account for their anisotropic

angular response. This factor is based on an analysis of recent measurements

by T. L. Johnson.'^ The response of the spheres is assumed to be isotropic.

For the purposes of this section, we will treat the quantities "fluence" and

"dose equivalent" as a type of isotropic response.

Table I -- Net Increase in Response Due to Air Scatter

Type of Response

NTA Film, Polycarbonate Track Etch

Dosimeter

Dose Equivalent

9" Spherical Remmeter

Albedo Dosimeter

Fl uence

3" Sphere

Detector readings should be decreased by these percentages to obtain freefield values, although the correction is clearly negligible in many cases.

Increase per Meter for

Bare ^^^Cf Moderated "^^^Cf

0.5% 0.9%

1.0% 1.5%

1.0% 2 . 3%

1.1% 3.0%

1.2% 4.0%

1.7% 4.5%

n
C. Room Return

Scattering from the room walls, ceiling and floor, or "room

return," is not a new problem, and has been investigated at many laboratories

in the past. We can divide the problem into two limiting cases. First,

consider just a single reflecting surface. This would be the case for

calibrations done out of doors, or in a room with, say, a concrete floor but

31 with thin walls and roof. It nas been shown that the response to reflected

neutrons from a source at height h can be predicted by postulating an image source at a distance h below the floor. If the distance from the image source to the detector is denoted by rp the relative response of the detector

to the reflected and the source neutrons is given by

R a

r r

TT- = 2ag — cos6

R ^ a

0 0

( ^ r

^I

2 (1)

where R^/R^ is the relative response of the detector to the reflected

and the source neutrons,

a IS the albedo of the reflecting surface,

g is a factor to account for anisotropic detector response,

and are the spectrum-averaged responses for the reflected

and source neutrons, respectively,

r is the source-to-detector distance,

the specular angle 6 is given by tan 6 = r/2h; and

= (2h)^+r'^.

For r < h, eq (1) simplifies to

R a 2 r a r r /o\

0 oh

(The value of R^/Rq calculated from eq (1) differs from that obtained

from eq (2) by less than 4% for h>3r.)

12
For epi -cacimi urn neutrons, a has the value of 0.54 for concrete and is

2 6

about the same for dry soil, decreasing by 20% for saturated soil. The

value of g for reflection from a single surface will be determined primarily

by the ratio of the response of the detector in the direction perpendicular

to the surface and tne response in the direction of the source.

Calculated values for the quantity ga^/a^ for various instruments are given in Table 2. A value of unity was assumed for g for all responses

except the dosimeters. In this case g was set equal to 0.5, based on measureI 2 ments by T. Johnson of the relative response at 90° compared to the response

in the forward direction.

Table 2 --Calculated Values of the Factor qa /a for Single-Surface Reflection ^ r 0

^

go /a ^ r 0

Type of Response Bare Cf Moderated Cf

NTA Film, Polycarbonate Track .2 .3

Etch Dosimeter

Dose Equivalent .37 .6

9" Spherical Remmeter 0.68 0.75

Albedo Dosimeter 1.0 0.6

Fluence 1.0 1.0

3" Sphere 1.8 1.1

Values in Table 2 can be substituted in egs (1) or (2) to obtain the relative

detector response due to single-surface reflections.

As an example, consider a 3" sphere and moderated californium source,

2 m above a concrete floor. For a 50 cm source-dosimeter spacing, eq (1)

predicts R^^/R^ = 0.U18; i.e., a 1.8% correction. Although eq (1) has not

been explicitly verifed experimentally, the predictions are consistent with

28

Monte Carlo calculations by McCall.

13
Room return in the other limiting case, that of a completely enclosed

concrete room, is a much more serious problem. Not only are there now six

reflecting surfaces rather than one, but, on the average, each neutron makes

32 ^ 2 1/2 traversals of the room before being captured. A theory for this

32 33 34 case has been given by the present authors in a series of publications, '^^'^

and hence only the results will be presented here. (A very similiar develop35

ment had also been given earlier by Savinskii and hilyushkin. ) It has

been shown that the room scattered neutrons are essentially uniformly

distributed throughout the room, and that for bare californium in a concrete

room:

R a n— = 5.5 q

R

0 0 c

2

r_r (3)

with 4-^r ^ = ZA. , (4)

c 1

where A. is the area of the i^*^ surface of the room, and the summation is

taKen over the six room surfaces. The quantity r^ is the radius of a spherical cavity which has the same surface area as the actual calibration

room. The other symbols have the same meaning as in eq (1). (The expression

for moderated californium is identical except that the numerical coefficient

is 4.5 rather than 5.6.)

Since the fluence of room return neutrons is approximately uniform

2 tnroughout the room and the fluence of the source neutrons varies as 1/r , if we first correct for air scattering, we can, in general, write for the

total response, D:

U D-R+R=-4+R, (5)

0 r 2 r r

where is the response at unit distance to the source neutrons alone. The

quantity D is the quantity which we are trying to determine. Eq (5) is

0 35 36

quite general, and similar expressions have Deen given before. '

14
Combining eqs (3), (4) and (5) gives

Dr^ = D^(l+Sr^) (6)

c - c 4^ •

' ' (7) where S = 5.6 g—'^X"

0 i

2 2 Thus, plotting Dr as a function of r should give a straight line whose

intercept is D , trie desir-ed quantity. The slope will be D^S, with S given

by equation 7. The quantity S is, physically, the fractional room return

correction at unit source-detector distance. Typical data are shown in

Figures 1-3 for measurements taken with the Hankins albedo dosimeter.

252

Figures 1 and 2 are data taken with bare Cf in the LLNL and the NBS

calibration rooms, respectively. Figure 3 shows data taken in the NBS room with the moderated source. The ordinate in these figures is the detector

response per unit dose equivalent. Comoarison of Figures 1 and 2 show that

the value of obtained in the two facilities is the same to within 3% (.090

vs. .092) even though the room return corrections at one meter are 30% and

20% in the respective rooms. Comparison of Figures 2 and 3 shows that the room return correction for the albedo dosimeter is almost a factor of 4 less

2 2 for the moderated source than for the bare source (5.4'^3/m vs. 20%/m ).

(Figures 2 and 3 also show that the sensitivity (response per unit dose

equivalent) of this dosimeter is more than twenty times as high for the

moderated source than for the bare source (2.05 vs. .092).)

Room return corrections can thus be made either by calculation, using

eqs (3) and (4), (or, equi valently , eqs (6) and (7)) or by making response

measurements as a function of distance and fitting to eq (6). To calculate

the room return by eq (3) (or (6)), Table 3 lists some values of (qa^/a^)

for bare and for moderated californium in an enclosed concrete room.

15
Table 3 -- Values of (qa /a ) for Enclosed Concrete Room

Bare Cf Moderated Cf

3" Cd. covered sphere

Albedo Dosimeter

2.9

2.1

1.4

.58

hi uence 1 .'0 1.0

9" spherical remmeter

Aridersson-Braun Remmeter

.52 .86

.22

Dose Equivalent .35 .49

The values of (ga^/a^) for dose equivalent is calculated, and that for

fl uence is unity by definition. The other values are determined from measurements of the quantity S and then solving eq (7) for ga^/a^. In general the

measured values of 5 are in satisfactory agreement with calculated values.

While our experience indicates that the use of eqs (3) and (4), with tlie

values tabulated in Table 3, will give the relative response to within 25%,

there are two drawbacks to this approach. First, data are lacking for several

popular dosimeter types; most notably for track etch dosimeters. (We would

estimate that the response ratios for polycarbonate track etch dosimeters, or NTA film, would be approximately 80% of the response ratios for dose equivalent.

This has not been verified experimentally.) Second, particular calibration

facilities may differ in the scattering produced. For example, an iron-lined

32 room gives considerably less room return than bare concrete. In addition,

the simple formulas probably don't apply to irregularly shaped rooms. Therefore

the values of room return calculated from the data given here should be

considered only as a first approximation. Where the calculations indicate

that the correction will be significant (this will usually be ttie case for

albedo dosimeters irradiated with bare californium), each laboratory should do

its own set of measurements and experimentally determine its own room return

correction. The measurements should be taken at several different distances,

from v. 30 cm out to two or three meters. The data should first be corrected

for air scattering (where significant) by subtracting the amount suggested in

Section VLB. The corrected data should then be plotted as indicated above,

and the value of S obtained.

16
D. Anisotropic Emission and Scattering by Source Encapsulation

Neutron emission from californium itself is intrinsically isotropic, with

negligible scattering. The californium must, however, be encapsulated, and

there is a finite amount of scattering, both elastic and inelastic, from the

encapsulation. In the usual case of cylindrical encapsulation, the ratio of

in-to-out scattering will vary as a function of orientation and thus the

emission will no longer be isotropic. The flux density at a distance r from a source whose absolute total emission rate is Q is then usually given as

Q/4TTr •F(6), where F(6) is the anisotropy factor. Since, however, the anisotropy

is caused by both elastic and nonelastic processes, in general the anisotropy

factor will be energy dependent and its effect will vary with detector response.

For a lightly encapsulated source, F(e) is very close to unity. The NBS

37 californium sources, for example, are embedded in an aluminum pellet a- 6 mm

in diameter by 6 mm high, placed in a stainless steel capsule with 0.5 mm

38 thick walls. The californium sources sold by Amersham are slightly larger:

8 mm in diameter by 10 mm high, and the Savannah River SR-Cf-100 series

39 encapsulation is 9.4 mm in diameter and 33 mm high. Recent calculations by one of the authors (CHE) show that in the direction perpendicular to the

cylinder axis, F(0) = 1.01 for the NBS sources; that is, in-scatter predominates

40 over out-scatter to give a 1% increase in flux density. Hunt has reported a measured anisotropy factor of F(0) = 1.012 for his Amersham californium sources,

in good agreement with the calculated value for the similar-sized NBS source. The situation for the SR-Cf-100 series is less clear, since different

combinations of type 304 stainless steel, 90% platinum -10% rhodium alloy,

and zircalloy 2 are used for these doubly encapsulated sources. (Zircalloy

2 is a 98% zirconium, 1.5% tin, alloy.) The most common types of encapsulation

are listed in the table below, in estimated order of increasing scattering

and hence of increasing anistropy.

1) Monsanto Research Corp.

2) Monsanto Research Corp.

3) Savannah River

4) Savannah River

Suppl ier

SR-Cf-100 Encapsulations

Inner Capsule

Zircalloy

Stainless Steel

Outer Capsule

Zi real 1 oy

Stainless Steel

Zi real 1 oy

Stainless Steel

Pt-Rh

Pt-Rh

17
Hunt has recently reported a measured anistropy of F(0) = 1.037 for the

Pt-Rh-Zi real 1 oy capsule (number 3). We do not, at this time, have accurate

measurements or calculations for the anisotropies of the other encapsulations,

but we estimate that numbers 1 and 2 should be somewhat less than 3, and 4

possibly somewhat more. (We would like to thank Dr. Hunt for supplying us with

the result of his anisotropy measurement.) •

For work of the highest accuracy, the anisotropy factor should be applied

as a correction to the readings of any device which measures fluence (e.g.,

long counters) or is sensitive to low energy neutrons (e.g., albedo dosimeters).

Since, however, some of the inscattered fluence consists of neutrons which have

been inelastical ly scattered to lower energies, calculations indicate that the

effects of in-scatter and out-scatter more nearly balance for instruments which

are less responsive to lower energy neutrons. Hence, for all practical purposes,

it is probably not necessary to make any anistropy corrections for devices such

as remmeters or polycarbonate track etch dosimeters.

It must be emphasized that these results only apply to small, lightly

40 encapsulated sources. Hunt, for example, reports a measured anisotropy

factor F(9) = 0.76 for a large Am-Li source, viewed end-on. Hence, for large

sources, calculations or measurements should be made for the specific source,

and the correction factor applied to the detector response.

E. Summary of Correction Procedures.

We reiterate that the only way in which calibrations can be made comparable

between one laboratory and another, and the only way in which new devices can be fairly compared to existing ones, is to reduce the calibrations to "freefield" conditions. This means that the fluence, or dose-equivalent, delivered

to the device is calculated assuming a "free field." The various scattering

contributions are then treated as backgrounds which are subtracted from the

reading of the device under test. There are three sources of scattering

background, each of which usually tends to increase the reading. Two of

these, air scatter and source scatter, are often negligibly small, but room return is often very large. Some laboratories can readily make these corrections

by Monte Carlo calculations. Other laboratories should do the corrections

following the methods discussed earlier in this section, which are here

briefly summarized.

18
1. Source Scatter (Section VI D)

This correction is independent of distance, and amounts to betv/een

% 1% and a. 3% for lightly encapsulated sources and albedo dosimeters,

and is essentially zero for remmeters or track etch dosimeters. For

heavier encapsulations it must be determined by either measurement or calculation.

2. Air Scatter (Section VI B)

The relative contribution from air scatter increases linearly with

source-detector distance. The numbers given in Section VI B may be used

to estimate this correction.

3. Room Return (Section VI C)

The relative contribution of room return varies as the square of the

source-detector distance, and can become very large for an indoor facility

As a first approximation, eqs (3) and (4) together with table 2 may be

used. If this calculation indicates a large correction (say, > 20 /o) , the room return correction should be explicitly measured for that particular

facility, using the method suggested in VI C.

In an outdoor facility, "room" return is much less serious and may be

calculated from eq (2), together with Table 1.

VII. Conclusions • '

'

•

We have described procedures for dosimeter and remmeter calibrations

which will give well-defined results accurate to within the present state-ofthe-art. Equally important, adoption of these procedures will assure different laboratories getting the same results for the same dosimeter.

VIII. Acknowledgments

We would like to thank Dale Hankins of LLNL and Tom Johnson of NRL for

making several important measurements whose results are quoted here. We

would also like to thank Dick McCall of SLAC for making the air scattering

cal cul ati ons

19
REFERENCES

1. J. L. Chartier, et a1 . , "Neutron Reference Radiations for Calibrating

Neutron Measuring Devices Used for Radiation Protection Purposes and for

Determining their Response as a Function of Neutron Energy," Proc. Fourth

Symp. on Neutron Dosimetry, Munich-Neuherberg , June 1-5, 1981. EUR 7448

EN Volume I, p. 387.

2. J. A. Grundl and C. M. Eisenhauer, "Fission Spectrum Neutrons for Cross

Section Validation and Neutron Flux Transfer," Proc. Conf. Nuclear Cross

Sections and Technology, Washington, D.C., March 3-7, 1975. NBS Special

Publication 425, p. 250.

3. J. Grundl and C. Eisenhauer, "Fission-Rate Measurements for Materials

Neutron Dosimetry in Reactor Environments," Proc. First ASTM-EURATOM

Symp. on Reactor Dosimetry, Petten (Holland), Sept. 22-26, 1975.

EUR 5667 e/f Part I, p. 425.

4. M. V. Blmov, "Neutron Energy Spectra of Spontaneous Fission Sources,"

Proc. IAEA Consultants Meeting on Neutron Source Properties, Debrecen,

Hungary, 17-21 March 1980. INDC(NDS)-1 14/GT, p. 79. , ,,

5. R. B. Schwartz and C. M. Eisenhauer, "The Design and Construction of a 252 D^O-Moderated Cf Source for Calibrating Neutron Personnel Dosimeters

Used at Nuclear Power Reactors," NUREG/CR-1 204, January, 1980.

6. G. W. R. Endres, J. M. Aldrich, and L. W. Brackenbush, "Neutron Dosimetry

at Commercial Nuclear Sites - Report on Subtask B," Pacific Northwest

Laboratory, PNL-SA-9201 , January, 1981.

252

7. R. B. Schwartz and C. M. Eisenhauer, "Use of a D^O-Moderated Cf Source

for Dosimeter Testing and Calibrating," Proc. Eighth DOE Workshop on Neutron Dosimetry, Louisville, Kentucky, June 18-19, 1981. PNL-SA-9950,

p. 153.

20
8. D. E. Hankins, "Phantoms for Calibrating Albedo Neutron Dosimeters,"

Health Physics 39_, 580 (1980).

9. A. E. Nash and T. L. Johnson, "Effect of Badge Position on the Response

of an Albedo Neuiron Dosimeter," health Physics 39, 989 (1980).

10. G. W. R. Endres, J. M. Aldrich, L. W. Brackenbush , R. B. Schwartz, and

C. M. Eisenhauer, "Neutron Dosimeter Performance and Associated Calibrations

at Nuclear Power Plants,'^ NUREG/CR-2233, 1981.

11. H. Ing and W. G. Cross, "Spectral and Dosimetric Characteristics of a Moderated-Neutron Calibration Facility," Proc. Eighth DOE Workshop on Neutron Dosimetry, Louisville, Kentucky, June 18-19, 1981, PNL-SA-9950,

p. 192.

12. T. L. Johnson, private communication.

13. D. E. Hankins, private communication (March, 1980).

14. E. J. Axton, "The Effective Centre of a Moderating Sphere When Used

as an Instrument for Fast Neutron Flux Measurement," J. Nucl . Energy 'Z6_,

581 (1972).

15. K. G. Harrison, "The Calibration of Neutron Detectors with Spherical

Moderators," Nucl. Instr. and Meth 184, 595 (1981).

16. W. Mannhart, "Progress in Integral Data and Their Accuracy: Average

Neutron Cross Sections in the Cal ifornium-252 Benchmark Field," Nucl.

Sci. Eng. 77, 40 (1981).

17. H. Ing and W. G. Cross, "Calculated Spectra for the Dosimetry of D^OModerated Neutrons," Health Physics 32, 351 (1977); See also H. Ing

and S. Makra, "Compendium of Neutron Spectra in Critical ity Accident

Dosimetry," Technical Report Series No. 180, International Atomic

Energy Agency, Vienna, 1978.

21
18. C. Eisenhauer and R. Schwartz, "Difficulties in Calculating SpectrumAveraged Values of Neutron Dose Equivalent," Health Physics, 41_, 774

(1981).

19. H. Kluge and W. G. Alberts, "Neutron Fl uence-to-Dose Equivalent Conversion

Factors," Health Physics 40, 414 (1981).

2 20. D. W. 0. Rogers, "Conversion of Neutrons/cm to Rem," Health Physics 34, . 395 (1978).

21. D. E. Hankins, "Neutron Dosimetry Studies at tne Lawrence Livermore

Laboratory," Proc. Sixth ERDA Workshop on Personnel Neutron Dosimetry,

July 11-12, 1977, Oak Ridge, Tennessee. PNL-2449, p. 67.

22. National Council on Radiation Protection and Measurements Report No. 38,

"Protection Against Neutron Radiation," NCRP, Washington, D.C. 1971.

23. W. G. Cross amd H. Ing, "Calculations on Neutron Dosimetry at the Chalk

River Nuclear Laboratory," Proc. Eighth DOE Workshop on Neutron Dosimetry,

Louisville, Kentucky, June 18-19, 1981. PNL-SA-9950, p. 163.

24. A. B. Chilton and Y.-L. Shiue, "Progress Report on Studies Concerning

Doses in the ICRU Sphere Resulting from Neutron Beams," Proc. Fourth

Symp. on Neutron Dosimetry, Munich-Neuherberg , June 1-5, 1981, EUR

7448 EN, Volume I, p. 63.

25. International Commission on Radiation Protection Publication 21, "Data

for Protection Against Ionizing Radiation from External Sources,"

Pergamon Press, Oxford, 1973.

26. F. J. Allen, A. Futterer, and W. Wright, "Dependence of Neutron Albedos

Upon Hydrogen Content of a Shield," Ballistic Research Laboratories

Report BRL-1224, Aberdeen, Maryland (1963).

22
27. H. Schraube, F. Grlinauer, and G. Burger, "Calibration Problems v/ith

Neutron Moderator Detectors," Proc. Symp. on Neutron Monitoring for

Radiation Protection Purposes, IAEA, Vienna, Dec. 1972, Vol II, P. 453.

28. R. C. McCall, private communication (1981).

29. G. L. Simmons and C. M. Eisenhauer, "Neutron Dose and Fluence Distributions

in an Infinite Air Medium," NBS Technical Note 745, National Bureau of

Standards, 1973.

30. D. E. Hankins, "Energy Dependence Measurements of Remmeters and Albedo

Neutron Dosimeters at Neutron Energies of Thermal and Between 2 keV

and 5.6/ MeV.,' Proc. Int. Radiation Protection Assn. IV International

Congress, Paris, France, 24-30 April 1977, p. 553.

31. C. Eisenhauer, "An Image Source Technique for Calculating Reflection

of Gamma Rays or Neutrons," Health Physics 1J_, 1 1 45 (1 955).

32. C. M. Eisenhauer and R. B. Schwartz, "The Effect of Room-Scattered

Nuetrons on the Calibration of Radiation Protection Instruments," Proc.

Fourth Symp. on Neutron Dosimetry, Munich-Neuherberg, June 1-5, 1981,

EUR 7448 EN, Volume II, p. 421

33. C. M. Eisenhauer, R. B. Schwartz, and T. Johnson, "Measurement of

Neutrons Reflected from the Surfaces of a Calibration Room," Health

Physics, to be published.

34. C. M. Eisenhauer and R. B. Schwartz, "Analysis of Neutron Room Return,"

Proc. Eighth DOE Workshop on Neutron Dosimetry, Louisville, Kentucky,

June 18-19, 1981, PNL-SA-9950, p. 171.

23
35. A. K. Savinskii and I. V. Filyushkin, "An Estimate of the Contribution by

Neutrons Scattered in Enclosed Rooms to a Total Radiation Dose," Proc.

Symp. on Neutron Monitoring for Radiation Protection Purposes, IAEA,

Vienna, December 1972, Vol. I, p. 271. (In Russian, Available in translation

as UCRL-Trans-10771 from University of California, Livermore, California).

36. R. C. McCall, T. M. Jenkins, and R. A. Shore, "Transport of Accelerator

Produced Neutrons in a Concrete Room," Stanford Linear Accelerator

Report SLAC-PUB-2214 (1978).

37. J. A. Grundl, V. Spiegel, C. M. Eisenhauer, H. T. Heaton II, D. M.

Gilliam, and J. Bigelow, "Cal i fornium-252 Fission Spectrum Irradiation

Facility for Neutron Reaction Rate Measurements." Nucl. Technol . 32^, 315

(1977).

38. Amersham Technical Bulletin 76/7, "Neutron Sources: Americium-241/

Beryllium and Cal ifornium-252 , " (July 1976).

39. Monsanto Research Corporation, Dayton, Ohio, "Radioactive Sources, Their

Manufacture and Characteristics ," (1981).

40. J. B. Hunt, "The Calibration and Use of Long Counters for the Accurate

Measurement of Neutron Flux Density," National Physical Laboratory Report

RS 5 (April 1976).

41. J. B. Hunt, private communication (1982). v:

24
Appendix

The NBS fission spectrum evaluation of 1975, updated in 1976, is

described up to 20 MeV by means of a reference Maxwell ian, M(E), modified

by four piecewise continuous segments below 6 NIeV plus one exponential

segment above 6 MeV. The reference Maxwell ian is

M(E) = 0.667 /E exp(-l .5E/2.13) , E in MeV,

and the evaluated spectrum is x(E) = ij(E) • M(E). The adjustment functions

y(E) are as follows:

Energy

Interval ^Cf^^^

(MeV)

0.0

0.25

0.8

1.5

6.0

0.25

0.8

1.5

6.0

20

+ 1.20E

- 0.14E

+ 0.024E

- 0.0006E

- 0.237

+ 0.098

- 0.0332

+ 0.0037

.0 exp[-0.03(E-6.0)/l .0]

A 44 group tabulation of this evaluated spectrum is given in Table A-1

.

25
Table A-1

NEUTRON FLUENCE SPECTRUM OF ^^^CF

^ Fluence per

E * Fluence per

Hj Unit Lethargy Hi ' Unit Lethargy

bO keV 3.32 X 10

^ 3.0 MgV 4 29 X IQ-^

100 1 .06 X 10

^ 3.4 4 02 X IQ-^

200 3.16 X 10

^ 3.7 3 67 X 10"^

250 6.27 X 10-2 4.2 3 26 X IQ-^

300 8.34 X 10-2 4.6 2 78 X IQ-^

400 1.12 X IQ-^ 5.0 2 40 X 10"^

500 1 .51 X 10"^ 5.5 1 99 X 10-^

600 1 .88 X 10"^ 6.0 1 61 X 10"^

700 2.23 X 10"^ 6.5 1 27 X lO""" 800 2.53 X 10"^ 7.0 9 90 X 10-2

1 MeV 2.98 X 10"^ 7.5 7 64 X 10-2

1.2 3.52 X 10"^ 8.0 5. 86 X 10-2

1.4 3.94 X IQ-^ 8.5 4 45 X 10-2

1.5 4.20 X 10"^ 9.0 3 38 X 10-2

1.6 4.32 X IQ-^ 9.5 2 53 X 10-2

1.8 4.47 X IQ-^ 10 1 90 X 10-2

2.0 4.59 X 10"^ 11 1 24 X 10-2

2.2 4.64 X IQ-^ 12 6 84 X 10-3

2.3 4.63 X IQ-^ 13 3 72 X 10-3

2.4 4.61 X 10-^ 14 2 01 X 10-3

2.6 4.55 X 16 8 31 X 10-4

2.8 4.43 X IQ-"" 18 2. 30 X 10-4

Normalized to one source neutron

Grundl , J. and Eisenhauer, C. , "Fission Rate Measurements for Materials

Neutron Dosimetry in Reactor Environments," Proceedings First ASTM-EURATOM

Symposium on Reactor Dosimetry, EUR5667 e/f. Commission of the European

Communities, Petten (September 1975).

Grundl, J. A. and Eisenhauer, C. M. , "Fission Spectrum Neutrons for Cross

Section Validation and Neutron Flux Transfer," Proceedings Nuclear Cross

Sections and Technology Conference, NBS Special Publ . 425, U.S. Dept. of

Commerce, Washington, D. C. (March 1975).

*Upper limit of energy interval

26
0.3

Dr^ = .090 ( 1 + .30 r^),..^

0.2 - HANKINS ALBEDO DOSIMETER

M LLNL ROOM

wo

BARE CF SOURCE

0.1

0

0 2 4 6 8 10

Figure 1 - Response of the Hankins Albedo Dosimeter per unit dose

equivalent as a function of distance from the neutron source. The

252

measurements were taken with a bare Cf source in the LLNL calibration

room. The data are fitted to an equation of the form

Dr^ = D (l+Sr^),

0

^

^ where D is the measured response at a source-to-detector distance r. The quantity (in this case equal to 0.090) is the response per unit

dose equivalent at one meter, to the source neutrons alone, and is

thus the reciprocal of the calibration factor. The quantity S (in

this case equal to 0.30) is the fractional room return at one meter.

27
(MwQ

Figure 2 - Response of the Hankins Albedo Dosimeter per unit dose

equivalent as a function of distance from the neutron source. The

252

measurements were taken with a bare Cf source in the NBS calibration

room. The data are fitted to an equation of the form

Dr^ = D (1+Sr^),

0

where the quantities have the same meaning as in Figure 1.

28
3.0

Dr^ = 2.05 (1 + .054 r'^)

—

2.0 — o

o

HANKINS ALBEDO DOSIMETER

NBS ROOM

MODERATED CF SOURCE

—

1.0

\ 1 0 1 2 3 4 5

higure 3 - Response of the Hankins Albedo Dosimeter per unit dose

equivalent as a function of distance from the neutron source. The

252

measurements were taken with a moderated Cf source in the NBS

calibration room. The data are fitted to an equation of the form

Dr^ = (1+Sr^),

where the quantities have the same meaning as in Figure 1.

29
NBS.114A IREV. 9-78)

U.S. DEPT. OF COMM.

BIBLIOGRAPHIC DATA

SHEET

1. PUBLICATION OR REPORT NO. NBS SP 633

2. Gov't. Accession No. Z, Recipient's Accession No,

4. TITLE AND SUBTITLE

PROCEDURES FOR CALIBRATING NEUTRON PERSONNEL DOSIMETERS

5. Publication Date May 1982

6. Performing Organization Code

7. AUTHOR(S) R. B. Schwartz and C. M. Eisenhauer8. Performing Organ. Report No.

9. PERFORMING ORGANIZATION NAME AND ADDRESS

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE

WASHINGTON, DC 20234

10. Project/Task/Work Unit No.

11. Contract/Grant No.

12. SPONSOpiNG ORGANIZATION NAME AND COMPLETE ADDRESS (Street, Citv, State, ziP)

Division of Operational Safety

U.S. Department of Energy

Washinqton, DC 20585

13. Type of Report & Period Covered Final

14. Sponsoring Agency Code

15. SUPPLEMENTARY NOTES

Library of Congress Catalog Card Number: 82-600543

[

! Document describes a computer program; SF-185, FIPS Software Summary, is attached.

16, ABSTRACT (A 200~word or less factual summary of most si gni ficant information. If document includes a significant bibliography or

literature survey, mention it here.) Procedures are given for routine testing and calibration of neutron dosimeters

and remmeters vn'th radioactive neutron sources. The issues addressed include:

the choice of neutron source*, phantom construction; fluence to dose equivalent

conversion; and the corrections for air scatter, room return, and anisotropicneutron emission. Explicit, semi -empi rical , analytic expressions are given for

the room return correction, and calculated numerical values are given for air

scatter.

17. KEY WORDS fs/x to twelve entries; alphabetical order; capitalize only the first letter of the first key word unless a proper nam>s separa ted by semicolons)

airscatter; calibration; cal i fornium; dose equivalent: dosimeter; neutron;

remmeter; room return.

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