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NATL INST OF STANDARDS & TECH R.I.C.

A11 103072886

Conference on Neutro/Proceedlngs of a Co

BC100 .U57 N0.554, 1979 C.2 NBS-PUB-C 19

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NBS SPECIAL PUBLICATION 554

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

ftr

itors

1

Proceedings of a G(

on Neutrons fron

Medical Aci
NATIONAL BUREAU OF STANDARDS

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Proceedings of a Conference ^^jo^

on Neutrons from Electron m :;n

Medical Accelerators

Proceedings of a Conference held at the National Bureau of Standards, C- Gaithersburg, Maryland, April 9-10, 1979

Edited by:

H.T. Heaton, II National Measurement Laboratory

National Bureau of Standards

Washington, DC 20234

R. Jacobs

Bureau of Radiological Health

Food and Drug Administration

Rockville, Maryland 20857

Sponsored in part by:

Bureau of Radiological Health Food and Drug Administration

U.S. Department of Health, Education, and Welfare

With cooperation from:

^ American Association of Physicists in Medicine

U.S. DEPARTMENT OF COMMERCE, Juanita M. Kreps, Secretary

Luther H. Hodges, Jr., Under Secretary

Jordan J. Baruch, Assistant Secretary for Science and Technology

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Director

Issued September 1 979
Library of Congress Catalog Card Number: 79-600133

National Bureau of Standards Special Publication 554

Nat. Bur. Stand. (U.S.), Spec. Publ. 554, 172 pages (Sept. 1979)

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PREFACE

These are the proceedings of the Conference

on Neutrons from Electron Medical Accelerators

held April 9-10, 1979 at the National Bureau of

Standards, Gaithersburg, Maryland. When medical

accelerators such as betatrons, microtrons and

electron linear accelerators are operated at

energies above about 10 MeV, neutrons can be produced by photonuclear interactions with machine

components such as collimators, beam flatteners,

etc. The purpose of this conference was to

acquaint members of the scientific, radiotherapy,

regulatory, and accelerator manufacturing communities with the latest state-of-the-art

information concerning this neutron production.

The program opened with a keynote address

which was followed by five invited paper sessions.

These sessions were concerned with the uses of

high energy beams in radiotherapy; the existing

regulations, their interpretation and application;

the biological effects of neutrons and the

appropriate quality factors to use in evaluating

the hazard associated with a given electron

accelerator; the characteristics of neutron

radiation from existing accelerators, the instrumentation, measurement and calculational techniques

available for determining the neutron spectral

fluence and dose in both the primary as well as

the leakage beams; and finally what shielding and

design techniques can be employed to reduce the

neutron production to a minimxam. The conference

closed with a summary session which was recorded

and is reproduced in full in these proceedings

for those interested in comments and review of

the full conference.

In order to achieve broad and even coverage

of the field, papers were presented by invitation only although attendance at the conference

was open to anyone. In spite of the highly

specific topic of the conference there were 208

registrants with 22 participants from 8 foreign

countries.

The papers are printed in the proceedings as they were received from the authors and in the

order in which they were presented in the sessions. For convenience we have preserved the conference

notation for the sessions. To speed the publication of the proceedings, all papers were submitted

by the authors in camera-ready form. We are greatly

indebted to the authors and all those who assisted

in the preparation of the manuscripts. Their

efforts have made it possible to get the proceedings

in print much more rapidly than would otherwise be

the case.

When commercial equipment, instruments and

materials are mentioned or identified in this

proceedings it is intended only to adequately

specify experimental procedure. In no case does

such identification imply recommendation or endorsement by the National Bureau of Standards,

nor does it imply that the material or equipment

identified is necessarily the best available for

the purpose.

The Editors gratefully acknowledge the

assistance of the National Bureau of Standards

Office of Technical Publications in the preparation of these proceedings and of Mrs. Kathy Stang

of the National Measurement Laboratory and

Miss Jo Ann Lorden of the Public Information

Division for help in the arrangements for the

conference, and the excellent secretarial

assistance of Mrs. Karen Fritz and Mrs. Joan Donahue.

H. Thompson Heaton, II National Bureau of Standards

Robert Jacobs Bureau of Radiological Health

iii
ABSTRACT

These proceedings are the compilation of 18 papers presented at the Conference on Neutrons from Electron Medical Accelerators held in Gaithersburg, MD on April 9 and 10, 1979. The topics addressed include:

properties of high energy beams for radiotherapy, regulations, biological

interpretation, physical measurements and calculations, and neutron reduction techniques.

,

Key words: Dosimetry, high energy radiotherapy beams, leakage, measurement techniques, medical accelerators, neutrons, relative biological effei standards, shielding techniques.

iv
TABLE OF CONTENTS

Preface ,

Abstract . .

Conference Personnel

KEYNOTE ADDRESS Unwanted Neutron Contribution to Megavoltage X-ray and Electron Therapy

J. S. Laughlin, A. Ried, L. Zeitz, and J. Ding

BEAMS FROM HIGH ENERGY ACCELERATORS FOR RADIOTHERAPY

Chairman: H. E. Johns

Advantages and Properties of High

Energy Photon Beams

H. E. Johns

Use of High Energy X rays in Cancer

Cure

W. E. Powers

Scattered Radiation Relative to Leakage

Radiation in High Energy X ray Beams

J. A. Rawlinson

REGULATIONS

Chairman: R. S. Caswell

International Standards for Medical

Accelerators

G. R. Higson •.

State Regulations for Medical Accelerators

C. H. Smith

BIOLOGICAL INTERPRETATION

Co-chairmen: J. A. Devanney

L. M. Bates

The RBE of Neutrons

H. H. Rossi

Risk of Cancer Based on Promotion

from Cytotoxicty

T. D. Jones, D. G. Jacobs, J. A. Auxier, and G. D, Kerr . . . , Clinical Tradeoff: The Maximum

Probability of Uncomplicated Cure

L. Cohen

Dependence of Radiation-Induced

Neoplastic Transformation In Vitro

Upon Radiation Quality and Repair

A. Han and M. M. Elkind
PHYSICAL MEASUREMENTS AND CALCULATIONS

Co-chairmen; G, R. Holeman

W, P. Swanson

Neutron Sources and Their

Characteristics

R. C. McCall and

W. P. Swanson ............ 75

Mixed Photon-Neutron Field Measurements

R, Nath, K. W. Price and

G. R. Holeman 87

Neutron Contamination in the

Primary Beam

P. H. McGinley and

M. Sohrabi 99

Neutron Production from Electron

Accelerators Used for Medical Purposes

E. J. Axton and

A. G. Bardell

Survey of European Measurements on Photoneutron Production from Electron

Accelerators Used for Medical Purposes

E. J. Axton and

A. G. Bardell

Neutron Leakage from Current Machines

P. R. Almond

NEUTRON REDUCTION TECHNIQUES

Chairman: C. S. Nunan

Design Criteria for Photon Beam Exits

on High Energy Medical Accelerators

Accelerators to Minimize Neutron Production

L. Taumann 139

Neutron Leakage Characteristics Related

to Room Shielding

E. Tochilin and

^ P. D. LaRiviere 145 SUMMARY PANEL DISCUSSION 155

List of Attendees 163

Author Index 175

vi
CONFERENCE PERSONNEL

Conference Co-chairmen

J. E. Leiss

National Bureau of Standards

J. C. Villforth

Bureau of Radiological Health

Program Committee

L, M. Bates American Association of Physicists

in Medicine

R. S, Caswell

National Bureau of Standards

J. A. Devanney

Bureau of Radiological Health

E. G. Fuller

National Bureau of Standards

H. T. Heaton, II National Bureau of Standards

G. R. Holeman

Yale University

R. Jacobs Bureau of Radiological Health

J, S. Laughlin

Memorial Sloan-Kettering Cancer Center

R. C. McCall

Stanford Linear Accelerator Center

J. McDonald

Memorial Sloan-Kettering Cancer Center

C. S. Nunan Varian

R. Schneider

Bureau of Radiological Health

R. J. Schultz

Yale University

W. P. Swanson Stanford Linear Accelerator Center

vii

NBS SP 554 (1979)UNWANTED NEUTRON CONTRIBUTION TO MEGAVOLTAGE X-RAY AND ELECTRON THERAPY

J.S. Laughlin, A. Reid, L. Zeitz, & J. Ding

Memorial Sloan-Kettering Cancer Center New York, N.Y. 10021

The significance of megavoltage x-ray and electron therapy and their development

is briefly outlined and illustrative plans presented. Early and recent measurements of photoneutron fluence present in the x-ray beam, and outside the x-ray beam but also incident on a patient, have been reviewed. The variation

of incident neutron production with x-ray energy, as well as that of the dose

in the patient due to internally produced photoneutrons , has been estimated.

(Accelerators; betatrons; electrons; megavoltage; neutrons; therapy; x-rays)

Introduction

I appreciate very much the privilege of

speaking at the initiation of this important con- ference on neutrons from medical accelerators.

This conference has attracted participation by

most, if not all, of the experts in this field

and it is likely that the work to be reported

here will lead to a resolution of this problem.

We are all concerned with the problem which confronts those employing high-energy x-rays and

electrons for treatment: that of the proper

evaluation of the role of the associated neutron

contamination. Before adding my comments on the

inherent neutron additive to megavoltage therapy,

I should like to review something of the frame of reference of this problem.

Not only were the radiations with which we

are concerned originally discovered in physical

laboratories , but their continuing development in

an optimum manner appropriate to medical application has been paced by physical developments. The most useful x-ray energies for diagnostic purposes

were achieved early and by the time of World War I

and thereafter, x-rays with peak energies ranging

from 30 to 150 kiloelectronvolts were readily

available. Although there was recognition that

higher energies were necessary in the megaelectron

voltage range for optimum treatment application,

the appropriate energies were not achieved until

after World War II.

My own involvement was at this point. I had

the opportunity to work with Professor D.W. Kerst,

developer of the betatron, and Dr. Henry Quastler,

therapeutic radiologist at the University of

Illinois. Our group, with the leadership of Kerst

and Quastler, first employed high-energy

x-rays, those from a 20 MeV betatron, for treatment

of cancer in 1948'"'''. The early measurements and

experience with betatron x-rays was rapidly added

to by H.E. Johns and other workers in the immedi-

(2 ,3) ately ensuing years . In 1951, H.E. Johns

initiated Cobalt-60 beam therapy which became the most prevalent methodology for megavoltage ther-

(4) apy . Also m 1951, Dr. Roger Harvey and I initiated high-energy electron beam (6 to 24 MeV)

therapy'^' at the University of Illinois College

of Medicine. This had required me to spend the previous year or so, in addition to getting betatron x-ray therapy started, in developing methods

of collimation, monitoring, dosimetry and treatment

planning for high-energy electrons, as well as on

improvements in beam extraction. Linear accelerators were soon also developed and applied to both

(6 7

electron and x-ray therapy ' Accordingly,

within the decade immediately following the second

World War, megavoltage x-ray and electron beam

therapy were fully launched with their many advantages and a few problems, with one of which we are concerned today.

Early Neutron Measurements

for Megavoltage Therapy

Neutrons are produced by photodisintegration

in the target of an accelerator as well as in the

differential filter and collimator arrangement.

These neutrons will be incident on the patient as

a contaminant in the primary x-ray beam and will

also reach the patient through the shielding. Neutrons are also produced by photodisintegration

within the patient, and much lower energy neutrons

are ambient in the radiation room. My early measurements with respect to this problem were published but I will review a few aspects of them

which are relevant to our problem. At that time I employed two methods of measurement, one a gated

boron trifluoride (BF^) counter within a specially

designed paraffin cave (Figure 1) , and also threshold detectors in the form of rhodium foils with

and without cadmiiam enclosures for thermal neutron

detection. For fast-neutron detection the foils were used inside a moderator. In the latter case,

a cadmiiam envelope around the entire moderator was sometimes employed. A radiiam-beryllium (Ra-Be)

standard neutron source calibrated at the Argonne

Laboratory was used for calibration purposes . The

(8) results were published in 1951 . They were

carried out for three different field sizes in

addition to a 1-centimeter-diameter field, with

these conclusions:

1 . The neutrons incident on the patient

were produced primarily in the target with lesser amounts in the filter and collimator arrangement.

2. The primary neutron energies from the betatron were lower than those from the radiumberyllium source as determined by absorption measurements (Figure 2)

3. The concrete walls of the room moderated

the neutrons giving rise to a slow neutron ambience .

1
"Figure 1. Horizontal cross section of BF^ counter

in paraffin directional geometry. Plot shows

relative number of neutrons approaching survey position horizontally. Ra-Be measurements were

made with standard at same distance from the

( 8

)

counter as the betatron target

4. Fast neutrons in the beam produced a dose in the patient's body. (The values of fast neutron fluence in the x-ray beam are given in

Figure 14.) Outside in the primary x-ray beam the neutron level fell off but not as rapidly as the x-ray beam. The dose due to neutrons incident on

the patient outside the primary x-ray beam was about half the neutron dose in the x-ray beam.

5. The average dose throughout the patient

produced by absorption of photoneutrons generated

in the patient was estimated on the basis of body

constituents, photodisintegration cross sections

as a function of bremsstrahlung energy, and simplifying geometrical assumptions . For the three

different field sizes, the average photoneutron

-4 -4 dose ranged from about 0.05 x 10 to 0.4 x 10 rad, per rad in the x-ray beam. This was less,

by an order of magnitude, than the dose produced

in the body immediately outside the incident x-ray beam by incident fast neutrons , and much less than

the scattered x-ray dose internally in the patient.

Accordingly, the internal neutron dose was considered to be clinically permissible.

6. The slow and thermal energy neutrons

from the shielding and walls were determined to be

negligible

7. With electrons, for which the cross sec- tion for electrodisintegration had been measxired to be less than that for photodisintegration by

(9) a factor of 400 , measurement at a position just

outside an electron beam indicated that the neu- tron flux was only 6% of that present with an x-ray beam at the same dose rate.

Para f f i n (cm)

Figure 2. Absorption of neutrons for mixed Ra-Be

source compared with that of primaries from beta-

(8) tron, and of those reflected from walls

Since 1951 there have been many more measurements, most carried out with improved technology.

It is naturally gratifying that subsequent measurements have essentially confirmed the measure- ments carried out in my laboratory at that time

and also that the conclusions reached then appear

to be still justificible. Before reviewing some of the measurements in the literature I should

like to say something about the science of high- energy x-ray and electron beam treatment planning

against the background of which this neutron problem must be viewed.

Megavoltage X-ray

and Electron Beam Treatment

Specific cases will illustrate advantages of

megavoltage x-rays and electrons . These cases were supplied by Ann Reid of our treatment planning group.

1. Liposarcoma, post-surgery. This patient

received 4000 rad of 10 MV x-rays in 20 fractions

over 30 days using two lateral arcs of uneven weighting and with uncommon centers separated by

7 cms. The resulting dose distribution is shown

in Figure 3. The required target volume was vin- usual both in shape and extent and because the margins were indefinite clinically. This plan

not only offered good coverage of the tentative

target volume, but also had the advantage of a slower fall-off of dose outside this area than

would have been offered by a fixed field technique

For this large volume the moving beams offered the

2
further advantage of spreading the dose and re- sulting in better clinical tolerance of the course of radiation therapy than a plan using

only four fixed fields.

from 3000 to 4000 rad with small hot spots of 6000

rad at the junction while the lung dose was kept

under 2000 rad.

(field size = 22 119 toreacli arc)

Figure 3. Patient treated for liposarcoma using

a waving beam technique with 10 MV X-rays

.

2. Recurrent adenocarcinoma of the right

antrum, post surgery. Extensive recurrent disease necessitated a large treatment volume. The

6 MeV linear accelerator unit was used to take

advantage of its sharp penumbral characteristics

in cutting off the dose to the left eye. However, due to the infiltration of the disease, a high skin dose was required and so bolus was

necessary. The technique used was a heavily

weighted anterior field contributing all the dose

in the anterior region and two wedged fields of

lower weighting coming in behind the eyes to compensate both for the falling anterior field

pattern and boost the dose in the posterior

region of the target volume. The resulting dose

distribution (Figure 4) shows good uniformity

throughout the target volxame and a dose of about

4% to the left eye.

3. Mesothelioma. The problem was to produce the required dose over a large irregularly

shaped volume while still keeping the lung dose

under tolerance limits. This was achieved by

combining 6 MeV x-ray fields with 10 MeV electron beams. The technique is simple, but the

positioning of the blocks critical. The individual dose distributions for the two modalities are

shown. Approximately 3800 rad were delivered

with opposing fields with 6 MV x-rays (Figure 5) with the areas overlaying the lung blocked out. These shielded areas were then treated with 10 MeV electrons to a dose of about 3500 rad (Figure

6) . Both x-ray and electron fields were treated

on the same day. The resulting dose distribution

is shown in Figure 7. The large volume received

6 Mev Linac 100 cm SAD

Figure 4. Patient treated for carcinoma of the right antrum using 6 MV X-rays.

3
Figure 5. 6 MV X-ray distribution for patient tre=''-ed for Kesothellotiia.

Figure 6. 10 MeV electron distribution for same patient as in Figure 5.

4
PM

MESOTHELIOMA COMPOSITE DOSE

DISTRIBUTION (PADS)

Figure 7. Dose distributions of Figures 5 and 6 combined

to show final treatment plan for this mesothelioma.

4. Osteogenic sarcoma. The difficulty in this case was that the required dose for tumor

control was 6000 rad while the tolerance dose to

the cord is 4000 rad, and the tumor was wrapped

around three sides of the cord. The solution lay

in the use of a gravity-oriented block of the

shape shown in Figure 8 and a 360 rotation with

Co-60 gamma rays. The block is small and its

placement critical. Computerized axial tomograph

(CAT) scans were done with markers indicating both

patient alignment for treatment and the outline of the shadow of the block on 'the skin when positioned

5
(Figure 9) . The scan allowed clear confirmation of patient and block alignment as well as the final treatment plan involving inhomogeneity corrections

(Figure 10) and the final treatment plan incorporating phantom measurement corrections for scattering into the shadow of the block (Figure 11)

5. Chest wall. For chest wall recurrences

in post-mastectomy breast carcinoma patients,

electron beam therapy is frequently used. Treat- ment planning incorporating corrections for changing absorption patterns in lung has been done by

hand (Figure 12) . Computer models for electron

beam therapy treatment planning have been developed and incorporated with CAT information to

Figure 9. CAT scan with markers verifying block and spinal cord alignment with

patient in treatment position.

Figure 10. Computerized dose distribution for osteogenic sarcoma(.

6
Figure 11. Final dose distribution incorporating inhomogeneity and scatter contributions for patient with osteogenic sarcoma.

Figure 12. Hand calculated dose distribution for treatment of chest

wall recurrence in breast cancer post mastectomy.

7
correct for inhomogeneities so that such distribution and displays as (Figure 13) shown are available.

Figure 13. Computerized dose distribution utilizing CAT data for same patient as in Figure 12.

Neutron Measurements

in Megavoltage Therapy

In the limited time it is possible to comment

on only a part of the recent literature'"*"*^ Of course, the results to be presented by other

investigators may change, and will certainly add

to our information on pertinent dose levels.

Measurements were reported in 1972 by Axton

and Bardell '"'"'^^ which were carried out with a 35 MeV Brown Hoveri betatron and a 16 MeV Mullard

linear accelerator. Gold foils were employed in polythene spherical moderators, and calibrated at

two different neutron energies with the National

Physical Laboratory Van de Graaff accelerator us7 7 ing the Li(p,n) Be reaction. The measured levels were somewhat high with respect to most other in- vestigators which may be partially explained on

the basis of the large field size (25 cm. by 25

cm.) employed and also possible photonuclear reactions in the moderator. Measurements have been

reported by McGinley, et . al . ^"'''''^ with a 10 MeV

Varian Clinac-18 linear accelerator, a 25 MeV Allis-Chalmers betatron, and a 45 MeV Brown Boveri

betatron. For fast neutrons they employed a bare

indium foil in a cubic water-filled container,

and the fission neutron spectrum of californim-252

was used for calibration. For thermal neutrons

they used the indium foils with and without a cadmium shield. A plutonium-beryllium neutron

source housed in a paraffin moderator was used for calibration purposes. As would be expected, since

the threshold for photodisintegration in copper is

9.9 MeV, the neutron values measured for an x-ray

energy of 10 MeV were very low. In fact, the neutron fluence in the beam was lower than that

outside the beam. These and their other measurements are shown in Figure 14. Measurements with

a 25 MeV Shimadzu betatron were reported by Fox

(12) and McAllister in 1977. They employed the neutron-proton reaction in aluminum for fast neutrons. They determined the primary target to

be the major source of neutrons and that a large

beam flattener contributed approximately 15% of the neutrons. Their measurements are shown in Figure 14. Measurements with a Philips linear

accelerator operated at 18 MeV were reported by

Gur et.al. in 1978. Their fast neutron measurement technology was similar to that employed

by Fox and McAllister. Indium foils were used ' for thermal measurements. The experiments were repeated for three different field sizes. Neu- tron fluence was found in their case to be independent of field size. The slow neutrons ac- counted for only about 1.5% and 5% of the total neutron dose in rads inside and outside the treat- ment fields, respectively.

Measurements with the 25 MeV Sagittaire were

(14) reported by Price, Nath and Holeman in 1978. They employed phosphorus-31 for both thermal and

fast neutrons . In the case of the fast neutrons

31 31 they relied on the reaction P(n,p) S, which

has a half life of 2.62 hours and a beta particle

8
v.7ith a maximiom energy of 1.48 MeV. For thermal

neutron detection they relied on the reaction

31 32 P(n,Y) P with a resulting half life of 14.3 days

and a beta particle with a maximum energy of 1.71

MeV. The beta activity was counted using a liquid

scintillation spectrometer. Phosphorus was used

in the form of ^2^5 P^^'^®^ placed in a small vial. All other activation products were shown to have

either very short half lives or to be stable, ex31 cept for the two desired radionuclides Si and

32

P. They had demonstrated that the photon sensitivity was less than 4%, which makes these threshold detectors particularly attractive.

Measurements were reported by Wilenzick

et.al.'"''^' in 1973 with a 25 MeV Sagittaire Linear

Accelerator and a 19 MeV Siemens betatron. They

employed both fission fragment track detectors and

silicon diode detectors. The fission fragment

detectors consisted of a thin uranium foil in contact with a disk of Lexan polycarbonate plastic.

Calibration was carried out with a califoi;nium-252 source. The sensitivity of this detector appeared

to be linear with a neutron dose between 4 and 300

rad. Because of the appreciable cross section for photofission, neutron measurements could not be

made in the primary x-ray beam with the track detectors. The silicon diode detectors consisted of

a p-i-n structure produced by diffusion of boron

and phosphorus atoms into high ohmic monocrystalline silicon. The change in forward voltage was measured by applying a constant current of 25 mA

before and after exposure to neutrons and observing

the potentials. The increase in voltage was determined to be a non-linear function of the neutron

dose. Neutron flux measurements outside of the beam with the track detectors are somewhat larger

than those obtained with the silicon diodes, possibly due to underestimation of the photofission

correction. Also, it appears that diodes are particularly sensitive to electrons and photons so

that they cannot be reliably employed in an x-ray

beam. In 1978 McCall and Jenkins ^"''^^ carried out

a study of the photon and electron response of

silicon diode neutron detectors. A Varian Clinac35 accelerator was used from which the electron

beam could be extracted as well as the x-ray beam.

They made a comprehensive study of the response of silicon diodes, silicon disks as well as moderated

indium foil detectors. Their measurements quantitatively defined the photosensitivity of silicon

diodes

.

Shown in Figure 14 are measured fast neutron

fluences obtained with threshold detectors plotted

as a function of the accelerator energy. The

numbers are identified with the references in the caption and results are given for both betatrons

and linear accelerators with measurements made of neutron fluence in and outside of the primary x-ray

beam. Considering the differences in field sizes, measurement techniques and correction factors employed, the spread of measurements is not surprising. Also, differences in spectra of the calibration sources can account for some of the spread.

An important conclusion which is suggested by this

distribution is that the magnitude of the neutron

fluence, either in or outside the primary x-ray

ACCELERATOR NEUTRON MEASUREMENTS

1.

•

in begm outatda BETATRON a m LINAC o •

10 20 30

ACCELERATOR ENERGY (M«v)

Figure 14. Fast neutron fluence measurement vs accelerator energies . Measurements numbered

(2,3): reference 10; (7,8,9): 11; (12,13):

12; (14): 13; (15): 14; and CS): 8

beam, does not increase appreciably with x-ray energy. Measurements of the neutron flux in the primary x-ray beam as a function of field size are plotted in Figure 15. As indicated, two investigators found field size dependence for their collimator arrangements whereas two other installations did not have such a dependence

.

• PmCC 25 MV 11978)

^ GUfl 18 m (13781 *LAUGHLm 23 MV (1951) ^FOX 23 MV (1977)

^

I 0

° '°° 300 50 700 900 FIELD SIZE cm^

Figure 15. Fast neutron fluence in x-ray beam as

2 a function of field size. Normalized at 100 cm .

,9
Table 1

Investigator 'V;ar/Machine (ref .)

Ratio of Fast Neutron Fluence in Primary

X-Ray Beam to that Outside

Fast Neutron Fluence

in becun out of beam

Distance Outside

Edge of Field cm

K. Lofgren (1970) Betatron (17) 2.8 0-7 Axton (1972) Betatron (10) 3.6 27.5 Axton (197?) Linac (10) 2.1 27.5 Wilenzick (1973) Betatron (15) 4 3 5 Wilenzick (1973) Linac (li.> 6.4 5 McGinley (1976) Linac (11) 0.7 5 McGinley (1976) Betatron (11) 1.4 5 McGinley (1976) Betatron (11) 1.4 5 Fox (1977) Betatron (12) 2.1 5 Gur (1978) Linac (13) 3.2 5 Price (1978) Linac (14) 2.0 5 Stranden (1977) Betatron (19) 2.1 50 Laughlin (1951) Betatron (8) 1.5 12.5

In Table 1 ratios, as determined on different

machines, of the neutron fluence in the x-ray beam

to the neutron fluence incident on the patient but

outside the x-ray beam are tabulated. The values

determined by threshold detector methods range

from 1.4 to 3.6. This degree of difference can be

readily explained on the basis of different collimator arrangements, differential filter size,

etc. Higher ratios are obtained with the use of

silicon diode detectors but the authors have indicated that photon sensitivity of the detectors was

(22) not appreciated. Chaudri, et.al. in an ab- stract indicated measurements carried out with a betatron at energies of 15, 20, 25 and 30 MeV. Their observed neutron fluence increased rapidly

with energy but appeared comparable to other results reviewed here. Relevant to the consideration

of x-ray dose internally scattered are the measure-

(26) ments of Rawlinson and Johns . They showed by

phantom measurements that the scattered x-ray

energy exceeded the leakage x-ray energy by factors

ranging from 5 to 20 for different situations.

An associated problem to patient exposure is that of personnel exposure. This problem has been

(21) recently discussed by Schulz who emphasizes

that the REE for carcinogenic effects obtained from

analysis of atomic bomb data from Japan ranges be-

(27 28) tween factors of 30 to 60 ' . This consideration increases the significance of the problem of

measuring low level amounts of neutron fluence in the presence of low level x-rays and electrons

including areas outside the treatment room. Schulz

has designed an ionization chamber filled alternately with argon or propane which does appear to provide an improvement in the method of measurement

of low level neutron exposure.

Calculated Neutron Production

Since most of the measurements of neutron

fluence were with accelerators in the 20 MeV range,

it appeared important as well as interesting to inquire into the situation with even higher energy

x-rays. As indicated in Figure 14, there are a few data at higher energies and these do not sug- gest any particular increase in neutron fluence

incident on the patient. A possibly more importeuit

problem is whether or not the neutron dose due to photoneutrons produced within the patient, which

appeared to be at accepteible levels at 20 MeV,

might increase to unacceptctble levels at higher

energies. The important assumptions were that the

target thickness is in all cases equal to the mean

range of the electrons, and that all of the photons

in the bremsstrahlung spectrum could interact with

all of the nuclei present in one target thickness.

The number of x-rays produced in the target was

(23) estimated on the basis of the Kramers equation A niomerical integration over the appropriate energy

(24) region of the product of the cross sections for

photoneutron production eind the number of x-ray photons gave the number of photoneutrons produced

in the target. The values of the thresholds and

cross section dependence on energy were tedten from

the "Atlas of Photoneutron Cross Sections Obtained

(24) with Monoenergetic Photons" by B.L. Bermem

On this basis, the ratio of the total number of

photoneutrons to x-ray energy produced in a tungsten target was obtained and is plotted in Figure

16. The calculations for platinum and gold gave closely similar results. Calculations carried out

for copper indicated a lower yield by a factor of

at least three. The calculated production of neutrons in the target rises from a low level at 10 ^

a

III

10
No. OF TARGET PHOTONEUTRONS/Unit of BrBmstraMung

TUrvJGSTEN TARGET

No. OF PHOTONEUTRONS PRODUCED IN STANQARD MAN

PER RAO OF X-RAYS

10'-

10' 30 40 50

IMPINGING ELECTRONS MeVFigure 16. Photoneutron production in the

accelerator target as a function of energy of

the impinging electrons.

MeV; increases by a factor of 50 in going from 10 MeV to 20 MeV; and by a factor of less than 2 from 20 MeV up to 30 MeV. On this basis we would

not expect the shielding problem for accelerators

to increase substantially above 30 MeV, neglecting

the slightly higher average energy of the photoneutrons produced.

This contrasts with the situation for the

photoneutrons produced within the patient as shown in Figure 17. We estimated the number of

photoneutrons produced and the x-ray energy absorbed per gram of tissue in a defined volume of

a standard man. The average dose produced in

that standard man by the absorption of these

photoneutrons was calculated on the basis of the

following assumptions:

1. The man was a rectangular parallelepiped

12 cm deep in the direction of the x-ray beam

with a length of 152 cm and a width of 38 cm, amounting to a volume of 70,000 cubic cm.

2. The nuclei contributing appreciably to photoneutron production reactions are oxygen

2 2 (7.44 grams/cm ), carbon (2.77 grams/cm ), nitro2 2 gen (0.31 grams/cm ), calcium (0.17 grams/cm ),

2 and additionally hydrogen (1.12 grams/cm ) for the

absorbed dose calculation. The absorption of neutrons from the irradiated region to the most

distal corner of the phantom man was estimated on

(29) tne basis of known cross sections . It was

assumed that this fraction applied to all of the neutron production which overestimates the amount

of neutron energy absorbed in the man. All of

I5 10'

0 10 20 30 40 50 MeV-»

ELECTRON ENERGY INCIDENT ON TUNGSTEN TARGET

Figure 17. Photoneutron production within

the patient.

the difference between the energy of the incident

photon and the threshold energy is transferred to the photoneutrons.

In Figure 17 it is indicated that the production of photoneutrons in the patient increases be- tween 20 MeV and 30 MeV by a factor of 20 and is still increasing more gradually at energies above

that. When converted into the average dose pro- duced within the patient on the assumptions given

above, we have for a 200 square centimeter field

with a 20 MV x-ray beam an average dose throughout

_g the man of about 5 x 10 rad, increasing to

2 x 10~^ at 30 MV, to 5 x 10~^ at 40 MV, and to

~s

8 X 10 rad at 50 MV. Values just outside the

periphery of the x-ray beam are somewhat higher

and dose levels at the extremities of the man are somewhat lower. These calculations come from a

, . , . (in preparation, 31) ^ ^, comprehensive analysis r r

of the distribution of high LET dose produced throughout

the patient's body during megavoltage x-ray treatment. These values are less than those estimated

in my earlier calculations, which, however, as- sumed the total absorption of all of the neutrons. An important conclusion is that despite this dramatic increase in the internally produced neutron

dose as a function of energy, it is still substantially less than that due to the internally scat- tered x-ray dose.

Horsley, Johns and Haslam ^"^'^^ estimated the total energy absorption by human tissue for the major nuclear reactions caused by high energy

x-rays, including that for photoneutrons. They

assumed all of this energy was absorbed in the irradiated region. This is acceptable for the charged particle reactions, but only a small pro- portion of the photoneutrons are locally absorbed.

11
most being absorbed elsewhere in the body or escaping. For this reason, their estimate of 0.3 - 0.6% additional energy absorbed in the irradiated region

is somewhat high. Our estimate for photoneutron

production for 30 MV x-rays, converted to the en- ergy in the fluence of neutrons produced in the irradiated volume per rad delivered, is about a factor of 2 less than that calculated by Horsley,

(29) et.al., for 28 MV. In our calculations, use

was made of the cross section profiles from the

(24) Berman Atlas These differed considerably, at least for the ^^(Yi1)''^^0 reaction, from the pro- files used by Horsley, et.al.. (30) When the cross section profile of Berman

(24) was replaced by that

employed by Horsley, et.al., calculations

showed an increase of 1.5 in the nxjmber of photoneutrons produced by the oxygen reaction. Since

the oxygen present contributes about 70% of the photoneutrons produced in man, this diffelrence in the cross section employed would account for most

of the factor of 2 elevation of the Horsley,

et.al. , '^'^^ estimate.

It is also essential to consider the quality

factor to be employed for neutron equivalent dose

calculations. In order to obtain a biological

neutron equivalent dose it is important to know

the RBE of neutrons for cell killing. The RBE values obtained in our laboratory for lethality

with HeLa cells in a fast neutron field, whose

dose average LET was about 50 KeV/ym, went from 2 (32) at high doses to about 3.5 at low doses . Similar results were obtained by Broerse, et.al.

(34) and by Hall, et.al. This has also been exam-

(24) ined recently in a .paper by Slater and Chu who

have emphasized the consequence of a changing

modifying factor 'DP) in the computation of dose

equivalent in rem It should be kept in mind that

higher values ^£ in going to lower doses or dose rates resi. because the reference radiation

has a shoulder in its dose response as contrasted

with heavy particle radiation, such as neutrons,

in which the extrapolation number is closer to unity

Summary

1. The important physically based advantage

of megavoltage photon and electron treatment has been illustrated, and the historical development

of this technology briefly reviewed.

2. Measurements of neutron fluences from

various accelerators have been reviewed. With

allowance for differences in machine geometries,

accelerator operating parameters, measurement

technologies, the results are in fair agreement.

3. It is recommended that basic measurement

data be reported separately before conversion fac- tors such as quality factor are applied. Such

factors are functions of energy and are not applied

in a uniform manner by different investigators.

Availability of data before folding in such factors facilitates interpretation.

4. Values of fluence or absorbed dose produced in a patient's body by either incident neutrons or internally generated neutrons are substantially less than that due to scattered x-rays.

In the opinion of the author, a leakage rate greater than 0.1% appears reasonable, even if it

is on an equivalent dose basis and includes a reasonable quality factor.

5. Calculations based on a simple model in- dicate that the incident neutron fluence per unit x-ray intensity should not increase markedly

above 30 MeV x-ray energy. Measured values are

not in disagreement with this prediction.

6. Calculations also indicate that the internally produced photo neutron dose will still

increase appreciably above 30 MeV before leveling

off.

In conclusion, with the participation of those present and the data to be presented, I am

sure we can look forward to a stimulating and

useful meeting. The National Bureau of Standards

and the Bureau of Radiological Health should be

congratulated on organizing this timely and important conference.

(1)

(2)

(3)

(4)

(5)

References

Quastler, H., Adams, G.D. , Almy, G.M. , Dancoff, S.M., Hanson, A.O. , Kerst, D.W.

Koch, H.W., Lanzl, L.H. , Laughlin, J.S.,

Reisen, D.E., Robinson, C.S., Austin, V.T.,

Kerley, T.G., Lanzl, E.F. , McClure , G.Y.,

Thompson, E.A., and Skaggs, L.S. Technique, for application of the betatron to medical

therapy. American Journal of Roentgenology , 61: 591-625 (1949)

Johns, H.E., Darby, E.K., Haslam, R.N.H.,

Katz, L. , Harrington, E.L. Depth dose data

and isodose distributions for radiation from

a 22 MeV betatron. American Journal of Roentgenology , 62: 257-268 (1949).

Laughlin, J.S., and Davies , W.D. Procedure

in dose distribution measurement of 25 MeV

x-rays. Science , 11 : 514-516 (1950).

Johns, H.E., Bates, L.M. , and Watson, T.A. 1000 Curie cobalt units for radiation therapy . British Journal of Radiology , 25 : 296-318 (1952).

Laughlin, J.S., Harvey, R.A. , Haas, L.L.,

Lindsay, J.E., and Beattie, J.S. Physical

aspects of rotation therapy with the betatron , Part I . American Journal of Roent- genology , 65: 947-951 (1951)

(6) Miller, C.W. Travelling-wave linear accel- erator for x-ray therapy. Nature, 171 : 297

(1953) .

(7) Greene, D., Nelson, K.A. Performance of a linear accelerator in clinical service.

Instrumental S Technical Notes , 33 : 336-338

(1960)

12
(8) Laughlin, J.S. Physical considerations in

the use of a 23-MeV medical betatron.

Nucleonics , 8^: 5-16 (1951) .

(9) Skaggs, L.S., Laughlin, J.S., Hanson, A.O.,

and Orlin, J.S. Electrodisintegration of ^ 63 , 107 109 . , „ . Cu , Ag , Ag . Physical Review , 73 : 420 (1948).

(10) Axton, E.J., Bardell, A.G. Neutron production from electron accelerators used for

medical purposes. Physics in Medicine and

Biology , 17: 293-298 (1972)

.

(11) McGinley, P.H., Wood, M. , Mills, M.

,

Rodriguez, R. Dose levels due to neutrons

in the vicinity of high-energy medical accelerators. Medical Physics , 3^: 397-402

(1976)

.

(12) Fox, J.G., McAllister, J.D. Fast neutrons

from a 25-MeV betatron. Medical Physics , 4^: 387-396 (1977).

(13) Gur, D. , Rosen, J.C. , Bukovitz, A.G., and

Gill, A.W. Fast and slow neutrons in an 18-MV photon beam from a Philips SL/75-20

linear accelerator. Medical Physics , 5^: 221-222 (1978)

.

(14) Price, K.W. , Nath, R. , and Holeman, G.R. Fast and thermal neutron profiles for a 25-MV x-ray beam. Medical Physics , 5^: 285-289 (1978)

.

(15) Wilenzick, R.M. , Almond, P.R. , Oliver, G.D.,

and DeAlmeida, C.E. Measurement of fast neutrons produced by high-energy x-ray beams

of medical electron accelerators. Physics

in Medicine and Biology , 18: 396-408 (1973)

.

(16) McCall, R.C., and Jenkins, T.M. , and Oliver,

G.D. Photon and electron response of silicon-diode neutron detectors. Medical Physics , 5^: 37-41 (1978) .

(17) Lofgren, K. , and Spring, E. Neutron radiation produced by the 32 MeV roentgen beam

of a medical betatron. Acta Radiologica , 9: 247-256 (1970)

.

(18) Deye, J.A. , Young, F.C. Neutron production

from a 10 MV medical linac. Physics in Medicine and Biology , 22 : 90-94 (1977)

.

(19) Stranden, E. Neutron doses to patients in

high energy x-ray therapy. Physics in Medicine and Biology , 22: 1011-1013 (1977).

(20) Ernst, W. and Ovadia, J. Contaminant dose

from incident neutrons associated with 22.5- MeV x-rays from a betatron. Radiology , 66 : 105-106 (1956) .

(21) Schulz, R. Argon/propane ionization-chamber

dosimetry for mixed x-ray/neutron fields.

Medical Physics , 5^: 525-531 (1978) .

(22) Chaudri, M.A., Allen, P., Millar, M.

,

Muirhead, E., Rouse, J., Spicer, B. Absolute

determination of the neutron flux produced by

the interaction of high energy Bremsstrahlung

with tissue. Abstract, private communication,

(23) Sproull, W.T. X-Rays in Practice , Chapter 3, The continuous x-ray spectrum, pp. 20-44. McGraw-Hill, New York (1946).

(24) Berman, B.L. Atlas of photoneutron cross sections obtained with monoenergetic photons.

Preprint UCRL-78482, Dec. 30, 1976. (Original

references are cited by Berman in Atlas.)

(25) Slater, J.M., Chu, W.T. A comment on the

leakage radiation regulations of high energy radiotherapy machines. Private Communication.

(26) Rawlinson, J.A. , Johns, H.E. Letter to the

Editor. Medical Physics , 4_: 456-457 (1977),

(27) Rossi, H.H., and Mays, C.W. Leukemia risk

from neutrons. Health Physics , 34 ; 353-360

(1978)

.

(28) Ishimaru, T. , Otake, M. , and Ichimaru, M. Dose-response relationship of neutrons and x-rays to leukemia incidence among atomic

bomb survivors in Hiroshima and Nagasaki by

type of leukemia, 1950-1971. Radiation

Research , 77: 377-394 (1979)

.

(29) Stehn, J.R., Goldberg, M.D., Magurno, B.A.,

and Wiener-Chasman, R. Neutron cross sections. Vol. I, Z = 1 Fo 20, BNL 325, Second

Edition, Supplement No. 2 (1964)

.

(30) Horsley, R. J. , Johns, H.E., and Haslam,

R.N.H. Energy absorption in human tissue by

nuclear processes with high-energy x-rays. Nucleonics , 11: 28-31 (1953)

.

(31) Zeitz, L. , and Laughlin, J.S. High LET dose

distribution produced by megavoltage x-ray treatment (in preparation)

.

13
(32) Zeitz, L. , Canada, T.R. , Djordjevic, B., Dymbort, G. , Freeman, R., McDonald, J.C.,

O'Neil, J., and Laughlin, J.S. A biological

determination of the variation of fast-neutron field quality with depth, RBE, and OER. Radiation Research , 63 : 211-225 (1975)

(33) Broerse, J.J. RBE values of fast neutrons

for damage to organized tissues in experimental animals. Iri: Radiation Research Biomedical, Chemical and Physical Perspectives . Eds. O.F. Nygaard, H.J. Adlar, and W.K. Sinclair, Academic Press, New York, p. 1073

(1975i

(34) Hall, E.J., Novak, J.K. , and Marino, S.A. Comparative radiobiological measurements with

two high-energy cyclotron-produced neutron

beams presently used in radiotherapy. British Journal of Radiology , 47 : 882-887 (1974) .

14
I

NBS SP 554 (1979)

ADVANTAGES AND PROPERTIES OF HIGH ENERGY PHOTON BEAMS

H.E. Johns

Ontario Cancer Institute, 500 Sherboume Street,

Toronto, Ontario, Canada MAX 1K9

A brief history of the development of high energy machines - cobalt 60 units,

betatrons and linacs is presented. The properties and advantages of these

machines are discussed. The large increase in cure rate resulting from the use of these machines in the treatment of cancer of the cervix is presented.

Instead of worrying about neutrons, we should be dealing with the properties

of these beams. They should be designed to give as small an entrance dose

as possible and the peak dose should be placed as far below the skin as possible. These parameters are much more important than neutron leakage. Ways

of improving these beams by altering the flattening filter and other components are described. Problems arising from the use of very large half body

fields are discussed. With very large fields, surface doses become large and

the buildup properties of the beam disappear. Ways of improving these fields

are dealt with. It is suggested that regulatory agencies should be more

concerned with beam improvements than with neutron leakage which is not a problem. The hope is expressed that with CT scanning and the possible improvements in beam direction and dose calculations, a further dramatic

increase in cure rate will result.

(betatrons; cobalt units; dose buildup; linacs; patient survival)

Introduction

This is a conference on neutrons from electron medical accelerators and according to the prospectus for the meeting, we have been asked to study "the shielding and design techniques which

can be employed to reduce the neutron production

to a minimum". I object to this statement since

I am only interested in reducing neutrons to a minimum provided that in doing so we do not interfere with some of the other properties of the

beam. We and the regulatory agencies should be

interested in optimizing the beam for radiotherapy

not merely in reducing neutrons to a minimum . In

this session I intend to discuss some of the properties and advantages of high energy photon

beams. This will be followed by Dr. Powers, who

will talk about the use of high energy x-rays in cancer cure and finally, we will hear a discussion

by Alan Rawlinson on scattered radiation relative

to leakage radiation. I hope this session will

show that neutron production is not a problem.

Some 30 years ago when I first became involved in using the betatron for medical purposes,

we were, of course, aware of the fact that with

high energy photons one could produce (Y,n), (y»p)

and (y.ci) reactions in tissue^. After a radiation

treatment patients were (and are) radioactive, and

it is relatively easy to measure this activity

With a geiger counter. The question immediately

arises: Is the energy absorbed in these processes

important in comparison with that from the conventional ionization and excitation of molecules?

! We showed that for 20 MeV bremsstrahlung the energy absorbed by these processes was about . 3% of

the total. Since this type of interaction gives

rise to high LET particles, the biological effect

; could be 10 times as high or contribute 3% of the

\ biological effect. Thirty years ago, we decided

this would not be a problem, and to my knowledge.

it never has been. Similar findings were made by

Laughlln in 1951^, and more recently by Chaudhri

in Melbourne^.

Properties of High Energy Photon Beams

Now, let us return to the main problem - the useful properties of high energy photon beams In cancer therapy. Figure 1 shows three Isodose

patterns from 200 kV radiation (left) cobalt 60 radiation (centre) and betatron radiation (right)

r— —

A -5 ' fl

J \ I-, ) rv/' r V so J

_K3 y so J

V A3 /

-15

200 KV FSOSOcm I .0 j

.20

MVL I5inm Cu 5.7cm

5 V 50 y

. . . 1 . . . .

Co*" SSO eOcm 5.7cm 10 .... 1 ... .

22 MEV

6cm 0 FSO 70cm

10

Fig. 1. Isodose distributions for 200 kV, Co-60,

and 22 MeV betatron radiation.

Today, we have available all three types of radiations and radiations with intermediate properties. It is useful to acquire some perspective

in this matter. At the end of World War II, most cancer centres had 200 kV x-ray equipment, and a few had 400 kV units, and a very few had 1 MeV

15
Van de Graaff generators.

200 kV radiation is really completely unsuitable for most types of radiotherapy. The dose on

the surface is 100% and this falls to about 25%

at a depth of 10 cm. With such radiation it is almost impossible to deliver a tumorcidal dose to

a tumour near the centre of the body without producing intolerable skin reactions. In fact, in those days, radiotherapists did not even attempt

to treat an obese person. Attempts were made in the Scandanavian countries to increase the tumor

dose relative to skin dose through rotation therapy, but with only limited success.

All of this was changed when betatrons came

on the scene. The first betatron to be used in cancer therapy was at Urbana, Illinois, under the direction of Dr. Donald Kerst. He, with a group

of young enthusiastic physicists, including such

people as Laughlln, Adams, Skags, Skaggs, Lanzl,

and others, used the betatron to treat one of their graduate students suffering from a brain

tumorS. Figure 1 is worth examining in detail,

to see what had really happened as a result of

this invention.

The surface dose from a betatron is now nearly zero, and with proper design the maximum

dose is at a depth of 5 cm and the 10 cm depth

dose is some 80%. The radiotherapist now has a beam which can be used to concentrate radiation

in the tumor and save normal tissues surrounding

it. Betatrons were developed rapidly by our group in Saskatchewan^ , and by Laughlln^ »^ »5

working in Chicago.

the normal survival curve. Today, patients are actually cured by radiotherapy. Dr. Bush has done

a cost benefit analysis on this problem, and has shown that if a woman's life is worth $205 a year,

the whole of the high energy radiotherapy program

can be paid for by the extra years of life given

to the many women with cancer of the cervix.

There can be no doubt today then, that high energy radiotherapy has had a major impact on cancer and we as physicists may take some of the credit. I am sure our next speaker. Dr. Powers, will also

discuss this increased cure rate resulting from

the use of high energy rad^5^^^on.

100

Princess Margoref Hospital Ca Cervix Survival

with Radiation

10

Yrs. ofter Dx

14

46 F

Two years later, cobalt 60 came along.

Figure 1 shows that it has many desirable properties, the surface dose is very small, the maximum dose is about 5 mm below the skin and the

10 cm depth dose is about 50%^>S. Now, cobalt

units do not produce as good a radiation distribution as the betatron, but they are, of course, very much simpler to operate. During the next

20 years cobalt units were placed in operation in all parts of the world, and were the main tool of radiotherapy. In recent years, they have been

challenged by linacs, which can give much higher

dose rates and can operate in the range from 3 MeV up to 20 MeV (Co-60 is equivalent to a 3 MeV

linac) and can give very large fields.

Can we, from the vantage point of 1979, look

back and see whether or not these technical developments have increased the cure rate? Dr. Bush, Director of our Institute, has looked at treatment results in two periods of time in

Ontario. The period from 1935-1944, before the Introduction of Cobalt 60 and the periods 1958-

1960 and 1960-1969, after the introduction of

cobalt units and betatrons. Figure 2 shows percentage survival of cancer of the cervix for these two periods of time. You can see that the final slope of the survival curve for low energy radiation is always greater than the slope of the normal population, so that in actual fact, patients were never cured by this type of radiation.

Figure 2 shows that after the introduction of

high energy radiotherapy, the survival is much

greater and the final survival curve parallels

Fig. 2. Cure rate for cancer of the cervix in the era before and after the introduction of high energy radiation (Co-60, betatron or linac radiation) due to Bush^,

Cobalt 60 Versus Betatron Radiation

During the last 30 years, there has been a continual Improvement of cobalt units, linear

accelerators and betatrons in the energy range 4-

30 MeV, and the user today is forced to try and

decide which machine should be purchased. Should

it be a cobalt unit, a linac at 4 MeV, or 6 MeV,

or 10 MeV, or should it be a higher energy linac

or betatron in the 18-25 MeV range? Cobalt units

are gradually being replaced by linacs - often

for the wrong reasons (they are supposed to be a health hazard) , while betatrons with their low

output and small field sizes are gradually being

replaced by linacs. Now, what photon energy is best? To my knowledge, there is only one controlt

ed study which seriously attempts to answer this

question. This is a study performed in our Institute by Dr. W.W. Allt^^, who carried out a random study on patients with advanced cancer of

the cervix. He treated half the patients with

our best technique using cobalt 60 and the other

half using betatron radiation. Figure 3 shows

the distribution obtained by these two machines. You can see that the dose in the region of the tumour can be made uniform and can be made to cover about the same area from both machines , but

the dose in the regions near the skin are much

higher with cobalt 60 than with betatron radiation.

16
Cobalt -60

8xl5cm at 92cm SAD

55°

Betatron 9xl5cm Qtl 20 cm TSD

55

Fig. 3. Dose distributions In the controlled

clinical study by Allt in the treatment of cancer of the cervix. Upper Co-60, lower

betatron.

This random study suffered from a number of

problems since there were differences in the number of fields per day, and the way patients were

placed in the two beams. There were also differences in the absolute dosimetry of the two systems since this was in an era when it was exceedingly difficult to perform an absolute calibration of the betatron beam. In spite of these

reservations, the differences in clinical results

were spectacular, as shown in Figure 4. This

figure is based on the original work of Allt^O,

but updated by Bush^, to cover a longer period of

time following treatment. The figure shows that we get a gain in the 14 year survival from about

23% to 41%. The graph also shows that both beams

in fact give cures since their final slopes are equivalent to the survival curves for normal

populations.

The question arises then, why did we do better with high energy beam than with cobalt 607 This is a question which cannot be answered

unequivocally, but I believe the differences in results are mainly due to the quite different

dose patterns for the same dose to the tumor

volume; the high energy beam deposits much less radiation in the normal tissue in the region between the skin and the tumour.

What Photon Energy Is Best?

We now turn to the vexing problem as to what

photon energy is best. We know from our clinical

experiment that our betatron operating in the photon mode at 22 MeV gave much better results

than Co-60, but we do not know what clinical results would have been achieved had we used a

100 90

80

70

60

50

40t30-

20 -

Normal survival 22 MeV t'r 60 ' ... population Co population

^157-1

MeV

148

Princes, Margaret Hospital Random study Co®°v 22MeV

Stages 2B and 3

I Co

145-145

10 12 Yrs after

Fig. 4. Survival data for cancer of the cervix

for the clinical random study by Alltl^^ up- dated by Bush^ to longer survival times.

10 MeV linac or an 8 MeV linac or a 6 MeV linac.

This question can probably not be answered imequivocally, since It would be unethical to carry

out such a clinical experiment today. In an attempt to answer this question we show Figure 5, which is the dose distribution when two opposing

fields are applied to a patient 25 cm thick. The distribution obtained in our clinical study really

involved two of these distributions nearly at right angles to one another. The graph shows that

with 25 MeV radiation the dose rises from a small

10 15 Depth in water, x , [cm] Podgorsok.Ro wl Inson , Jotins

flmer.JI Roent, 123,189.1975 27166

Fig. 5. Dose distribution for opposing pairs of

fields treating a patient 25 cm thick. Distributions are given for cobalt 60, 10 MeV,

25 MeV, and 32 MeV beams. Data by Podgorsak,

Rawllnson and JohnsH.

value at the skin surface to reach a constant value from a depth of 5 to 20 cm. In contrast,

the cobalt beam contributes considerably more rad- iation to the region immediately under the skin

surface, while 10 MeV is about half way between

.on. 17
the two. This would then suggest that 25 MeV Is better than 10, which in turn is better than

cobalt 60. It should be noted that increasing the photon energy above 25 MeV produces a minimal

effect, and suggests that nothing is to be gained

from a 'ose distribution point of view by using

radiation of energy greater than 25 MeV. When one

dl' -• realizes that the flattening of a high energy t»3m becomes increasingly more difficult as the energy is raised, and since the higher the energy

the greater the neutron yield from the collimating

system, there seems to be a good argument against

using energy much above 25 MeV. We believe then, that a linac operating in the photon mode should

operate in 20-25 MeV and that this is the optimum

energy for the treatment of deep seated tumors. Because of 'his belief we are installing such

machines in the six government operating clinics

in the Ontario Cancer Treatment and Research

Foundation. Lower energy machines such as Co-60

are still very useful for more superficial tumors such as those of the head and neck, and for much

of the paliative work of a clinic.

10 II 12 13 14 15 Deplh(cm)

Fig. 6. Depth dose distribution for a betatron at

25 MeV and a betatron at 16 MeV. Taken from Podgorsak, Rawlinson and JohnsH.

inacs and Betatrons of the Same Energy

Equivalent?

About 10 years ago, when we realized the clinical advantage of using a betatron over a cobalt unit, it was necessary to purchase a new

therapy machine. We therefore looked for an accelerator which would have the distribution of

a betatron but which would be capable of irradiatin- arge fields at high output. We purchased

a 35 M 'arian linac. In the photon mode, this

machine operates at 25 MeV. Much to our surprise, we found that it was equivalent to our betatron

operating at 16 MeV, as illustrated in Figure 6^1. Detailed investigations showed that the inferior

dose distribution was mainly due to the use of a high atomic number flattening filter. We found

that by using a lower atomic number target of aluminum and an aluminum flattening filter, the dose distribution could be made equivalent to that

of the betatron as illustrated in Figure 7, The

main factor responsible for the differences in the

dose distribution is the material used in the flattening filter. The differences are due to the way the absorption coefficient varies with photon energy for materials of high, medium and low

atomic number (see Fig. 8). Because pair production increases with increase in photon energy

above the threshold at 1 MeV, high energy photons

are more easily stopped in a material such as lead than are low energy ones. If a high atomic

number material like lead or tungsten is used as the filter, it will preferentially remove the high energy photons from the beam. It would seem rather foolish to go to the trouble of producing

high energy photons and then remove them by the flattening filter. These Ideas were thoroughly

discussed years ago^y^, when betatrons were first

put into operation. The reason for the use of

high Z filters is the difficulty of getting the large filter of low Z material in the limited

space which is available in the head of a linac.

It is for this reason that manufacturers of linacs

have refused to build flattening filters of the proper material. I am delighted to note, however

that in the last year, Varian has investigated

this problem and has shown how its old Clinac,

-1 r 1 1-

Betatron (AC) 25 MeV

A Linac 25 MeV

Al target and Al flattening filter

10 12 Depth [cm] 20

I475B

Fig. 7. Dose distribution for 25 MeV linac using

an aluminum target and aluminum flattening

filter compared to an Allis Chalmers Betatron

both operating at 25 MeV. Data from reference

11,

5

1.0

0.1

0.01

0.001

\ Lead y

\Copper\

\ai V \

V \ \

. Carborr

Lead—"

Copper— Vs5j^ __Ai—

CcrtSQa

lOkeV lOOkeV I MeV

Ptioton energy

10 MeV lOOM.

Fig. 8. Total attenuation coefficient in metres

squared/kg for carbon, alumlnimi, copper, and lead as a function of photon energy.

18
such as the one we have in Toronto, can be Improved to look like a betatron through the use

of a medium atomic number filter. This is illustrated in Figure 9, taken from the work of

McCall, Mclntyre and Tumbulll3.

A Old Clinac 35 o New Clinac K w'o Electron Shutter

McCall, Mclntyre & Turnbull,

Med. Phys. 5, p. 518, 1978.

"0 2 « 6 8 10 12 14 16 18 Oeplh in H^O (cml The depth dose curves for the old and new designs of the Clinac

35. The energy was 25 MeV, SSD of 1 00 cm, and 1 0 X 1 0-cm fields.

Fig. 9. Depth dose data for the old and new

design of Varian Clinac 3513.

Possible New Design for Linac Head

In designing the head, space is at a premium

and it is very difficult to include in the design

a flattening filter made of aluminum, or aluminum

oxide or some other low atomic number material.

We have studied this problem for the last year

and Figures 10a and 10b show a possible way of

overcoming this problem. Figure 10a shows the

conventional way of putting all the components

which are required into a linear accelerator head. Figure 10b shows how one can place the adjustable

collimator in one plane and in this way save enough space so that a full size flattening filter

of aluminum may be used. We are developing this

collimator for use on our 35 MeV Varian linac.

With it we hope that the depth dose distribution

from the linear accelerator will be equivalent to our betatron. In my opinion, these problems

which I have discussed are more important than

designing a head with minimum leakage for neutrons . We really want a high energy photon beam

which gives practically no dose on the surface of

the skin and gives its maximtjm dose as far below

the surface as possible.

Electron Contamination

Thirty years ago, the main emphasis in radiotherapy was in obtaining very small precisely

determined fields. In recent years, radiotherapists have turned more and more to larger fields. An extreme example of this is illustrated in Fig. 11 and shows a cobalt 60 unit, which we have designed to treat half the body of the patient at

one timel^. This gives a very large field of 60

X 150 cm. In the head is a flattening filter to give a uniform dose in the mldplane of a patient

lying under this machine. This machine was developed for Dr. Fitzpatrick and Dr. Riderl5,16^

who have now treated over 500 patients with half

body irradiation. The patient is treated by two opposing pairs of fields, over half their body to

Open port VlTj^jfi' Conical shaped electron beam

for tungsten target on slide l^E Electron scottering foils Primary '^^on slide collimator

Open section L^'li^^

'

for electrons

Mirror on slide

Fig. 10a. Schematic diagram to show the arrange- ment of the components in a conventional

linac head.

Electron beam

Target on slide Scattering foils —

V... J \

Primary /^l collimoiori

7 /:ollimol»; \

/ and shapecl \ / jm onslide^ ^

Retractable electron filter

j^ond matrix dosem^ter

J

'Single plane 4

'

/ jaw collimator \

1 Fig . 10b . Schematic diagram showing arrangement

of components to take an aluminum oxide flat- tening filter. The space is acquired by designing the collimator so that the four com- ponents move in one place on the surface of

the spheres rather than in two planes in the conventional system.

a dose of about 800 rads in as short a time as possible, to avoid radiation sickness. Patients

are then returned to the hospital about a month

later, and the other half of their body is treated in the same manner. This has proved to be a good way to treat patients with multiple metastases, and some of our patients are alive five years after their first treatment. This is an example of an exceedingly large field, but very large fields are also being used in treating

Hodgkins disease. When large fields are used the beam is often very badly contaminated with

electrons. Figure 12 shows some curves taken

with cobalt 60 and the effec^ of these low energy

19
Problems for the Future

Fig. 11. Photograph of half body irradiation at the Ontario Cancer Institute.

O 0.1 0.2 0.3 01 0.5 Depth in gm/cm'^

Fig. 12. Relative dose as a function of depth for

a large field setting for a cobalt-60 unit. Curves are shown for different electron filtersl7

electrons which completely destroy the build-up

characteristics of the beam. I believe that agencies such as the Bureau of Radiation Health,

and the manufacturers should spend far more effort

in finding ways to remove the electrons from the beam so that pure photon beams may be used to treat patients than to worry unnecessarily about neutrons which may be present in the beam.

1 . Removal of electron contamination

I believe we need to find ways to reduce electron contamination in high energy photon beams,

especially for large fields. I would like to see some imaginative research using perhaps high magnetic fields, which might be switched on to remove

these electrons from the beam.

2 . Higher precision in dose delivery

With the introduction of CT scanning in many

radiotherapy centres, we now have a method of assuring very accurate beam direction. In addition, the possibility of doing very precise dose

calculations now exist. More careful work in this

field is reqired.

Another aspect of the problem of precise dose

delivery involves calibration procedures for high energy (6-25 MeV) machines. For years, these have been performed using a dosemeter calibrated

at Co-60 energies but used at the higher energy. The user measures the "exposure" using the calibrated dosemeter and then multiplies the reading

so obtained by agreed on factors, C;^ for photons

and Cg for electrons (ICRU). He then applies numerous correction factors which are not universally

accepted and are not accurately known to estimate

the dose. This is an unsatisfactory situation

since today we could do much better using a direct

calibration of the machine, using a calorimeter.

Laughlin' ff^ group have long been Involved in recom- mending such a procedure. Calorimeters of two types have been developed - those using tissue

equivalent plastic (TEP) and those using pure

graphite. With pure graphite DomerJ^at NBS has

developed a beautiful system allowing a dose determination in graphite to 0.1%. The TEP calori- meter is of less precision because of the unknown

amount of energy which produces chemical change

in the plastic and so does not appear as heat.

The next step in dose calibration is more diff- icult. We require a transfer instrument (probably

an ion chamber) which can measure the ratio of the dose in graphite (or TEP) to the dose in a water

or tissue phantom at some reference point. Such a transfer instrument would be calibrated at the standardization laboratory in grays/meter reading

in water (or tissue) using the calorimeter and then used in the field. Development of the best

type of transfer instrianent and the appropriate

correction factors is required to enable us to go from the dose in a graphite (or TEP) calorimeter

to the dose at a suitable reference point in the users water or tissue-like phantom.

I believe that if we solve some of the problems which I have summarized, that in ten yearts, we will find that we have made another leap forward in increasing the cure rate. To do this,

however, we must pay a great deal of attention to detail, we must find ways to improve our beams and

improve the precision with which the beams are administered to the patient. We should not become

lost in regulatory arguments concerning allowed neutron fluxes. The important problem is the improvement of the beams for high energy machines.

20
References

[1] Energy absorption In human tissue by nuclear

processes with high energy x-rays. R.J. Horsley, H.E. Johns and R.N.H. Haslam.

Nucleonics 11, No. 2, 28-31 (1953).

[2] Use of a 23 MeV medical betatron. J.S. Laughlin. Nucleonics 8^, No. 4, 5-16,(1951).

[3] Absolute determination of neutron flux pro- duced by the interaction of high energy bremsstrahlung with tissue. M.A. Chaudhri,

M. Pion, E. Miller, E. Muirhead, J. Rouse

and B. Spicer. A. A. P.M. Workshop. Electron

Linear Accelerators in Radiotherapy Workship,

(1978).

[4] Techniques for application of the betatron

to medical therapy. G.D. Adams, G.M. Almy,

S.M. Dancoff, A.O. Hanson, D.W. Kerst, H.W. Koch, E.F. Lanzl, L.H. Lanzl, J.S. Laughlin,

H. Quastler, D.E. Riesen, C.S. Robinson and

L.S. Sksggs. Am. J. Roentgen. 60^, No. 2,

153-157,(1948).

[5] Techniques for ^plication of the betatron

to medical therapy with report of one case.

H. Quastler, G.D. Adams, G.M. Almy, S.M. Dancoff, A.O. Hanson, D.W. Kerst, H.W. Koch,

L.H. Lanzl, J.S. Laughlin, D.E. Riesen, C.S. Robinson, Jr., V.T. Austin, T.G. Kerley, E. F. Lanzl, G.Y. McClure, E.A. Thomson, L.S. Skaggs. Am. J. Roentgen. 61, No. 5, 591-625,

(1949).

[6] Depth dose data and isodose distributions

for radiation from a 22 MeV betatron. H.E. Johns, E.K. Darby, R.N.H. Haslam, L. Katz,

and E.L. Harrington. Am. J. Roentgen 62 , 257,(1949).

[7] 1000 Curie cobalt unit for radiation therapy.

I. The Saskatchewan Cobalt 60 Unit. H.E. Johns, L.M. Bates and T.A. Watson. Br. Jl. Radiol. 25, 296-302,(1952)

[8] II. Depth dose data and diaphragm design for

the Saskatchewan 1000 curie cobalt unit. H.E. Johns, E.R. Epp, D.V. Cormack and S.O. Fedoruk. Br. Jl. Radiol. 25^, 302-308 (1952).

[9] Malignancies of the overy, uterus and cervix. R.S. Bush. Edward Arnold Publishers, London,

England, (1979)

.

[10] Supervoltage radiation treatment in advanced

cancer of the uterine cervix. A preliminary

report. W.E.C, Allt. Can. Med. Ass. J. 100 792-797 (1969).

![11] X-ray depth doses from linear accelerators

in the energy range from 10 to 32 MeV. E.B.

I Podgorsak, J. A. Rawlinson and H.E. Johns.

Am. J. Roentgen. 123 , 182-191, (1975).

[12] Percentage depth dose for high energy x-ray

i^eams in radiotherapy. J.A^ Rawlinson, and

H.E. Johns. Am. J. Roentgen. 118, 4, 919-

922, (1973).

[13] Improvement of linear accelerator depthdose curves. Richard C McCall, Raymond D. Mclntyre, and William G. Tumbull, Med. Phys. _5. No. 6, 518-524 (1978).

[14] Cobalt-60 therapy unit for large field

irradiation. H.E, Johns, W.D. Rider, H. Aget, P.M.K. Leunt, and H.P. Webb. In Press, Medical Physics. (1979).

[15] Half-body radiotherapy. Peter J. Fitzpatrick and W.D. Rider. Int. J. Radiation

Oncology Biol. Phys. 1, 197-207, (1976).

[16] Half-body radiotherapy of advanced cancer.

P.J. Fitzpatrlck, and W.D. Rider. J. Can. Assn. Radiol. 27, No. 2, 75-79, (1976).

[17] Use of electron filters to improve the buildup characteristics of large fields

from Cobalt-60 beams. P.M.K. Leung, and

H.E. Johns, Med. Phys. 4_, No. 5, 441-444,

(1977).

[18] A heat-loss compensated calorimeter: Theory,

Design and Performance. S.R. Domen, and

P.J. Lamperti. J. of Res. of the NBS A - Physics and Chemistry _78A> P' 595, No. 5, (1975).

[19] Calorimetric and ionometric dosimetry for cyclotron produced fast neutrons. J.C. McDonald, M.I. Chang, and J.S. Laughlin.

Proc. 3rd Symposium on Neutron Dosimetry in Biol, and Med. (Ed. Burger, G. , and Ebert,

H.G.)

.

21

USE OF HIGH ENERGY X PAYS IN CANCER CURE

W. E. Powers

Wayne State University School of Medicine

Detroit, MI 48201

The speaker did not submit either an abstract

or written version of his talk for inclusion

in the proceedings of this conference. For further information, contact the speaker

directly.

23

NBS SP 554 (1979) SCATTERED RADIATION RELATIVE TO LEAKAGE RADIATION IN

HIGH ENERGY X-RAY BEAMS

J. A. Rawlinson

Physics Division

Ontario Cancer Institute

Toronto, Ontario CANADA M4X 1K9

It has been common practice recently to limit the neutron leakage reaching the

patient outside the useful beam to 0.1% in rems of the dose on the central

axis. This value appears to be based on earlier recommendations regarding

X-ray leakage. It is shown by comparison to the dose received by the patient

due to photon scatter that this value is unnecessarily low.

(Dose, leakage, medical accelerator, neutrons, scatter. X-rays)

One of the purposes of this conference will

be to consider the risk to the patient as a

result of stray neutron radiation received by

those parts of the patient outside the useful

treatment beam. To reduce this risk to an

acceptable level we must set limits to the

maximum neutron leakage produced by medical

electron accelerators. The purpose of this paper

is to present some ideas which will help provide

a more rational basis than hitherto for setting

such neutron leakage levels.

Up to now the thinking about neutron

leakage levels has been guided largely by earlier

recommendations concerning X-ray leeikage levels

.

For X-ray leakage present day regulatory practice

appears to stem from the recommendations of ICRP

Committee IV^^', recommendations made over 25

yeeirs ago. This committee specified that the

radiation leetkage through the permanent housing

of X-ray therapy equipment should not exceed 0.1%

of the dose rate on the central axis of the useful

beam. At that time the recommendation was

intended to apply to X-rays only. Equipment of

the period could meet the requirement without

undue difficulty. There was no problem.

More recently, however, in the case of high

energy medical electron accelerators, the same

[2] 0.1% leakage requirement has been interpreted

as including neutrons as well as X-rays, with the

leakage being expressed as a dose equivalent (in

rems or sieverts) . It is here that the problem

arises for recent measurements have suggested

that the neutron leeikage in rems from present-day

medical accelerators operating at high energies

(20 MV and above) usually exceeds 0.1% and is

r 3-51 typically 0.2%-0. 5%*^ Equipment manufacturers

claim that considerable changes might be required

in the design of medical accelerators if the new

leeikage requirement was to be met. Fxirthermore

it is possible that such changes would seriously

compromise the usefulness of these accelerators

for their intended purpose, radiation therapy.

We will hear more on these issues during the

remainder of this conference.

In view of this debate it seems timely to

reconsider the criteria on which to base leeJcage

requirements. Does the original 0.1% leakage

value for x-ray leakage really bear close

scrutiny? it appears to us that, in fact, this

value is rather lower than it need be.

To understand this, it must be borne in mind

that a patient undergoing radiation therapy

receives stray radiation not only from machine

leakage but also by scattering from the useful

beaun. This situation is depicted in Figure 1. The scatter arises from the patient himself and

also from the surfaces of the treatment room. The

former component is absolutely unavoidable; it is an accepted consequence of his radiation treatment. The latter component, also, we can in

practice do very little about. The two components

together in effect form a lower limit to the dose

received by the patient outside the useful beam.

It seems unnecessary, therefore, to limit leakage

radiation (whether it be X-rays or neutrons) to doses or dose equivalents much below those that

the patient inevitably receives from scattered

photon radiation.

The question then is, what is the magnitude

of the scattered radiation in normal radiation

therapy practice? Figure 1 illustrates how this can be determined. If a close-fitting insert of

say 3 tenth value layers is fitted into the collimation system, we measure just the leakage

radiation. If this value is subtracted from the

dose measured for an open beam of the same field

size, then we obtain the contribution due to scatter (or more precisely the component of the stray radiation arising from the useful beeun)

.

Our measurements indicate that in normal

radiation therapy practice the stray radiation due

to scatter is much greater than presently allowed

leakage radiation. Figure 2 shows the measured

dose due to photon scatter received at points

along the longitudinal axis of a long unit-density

2 phantom of 20 x 30-cm cross-section irradiated

2 by a 30 X 30-cm beam of 25-MV X-rays and a

30 X 30-cm^ beam of ^^Co y rays . The irradiation

geometry is shown in the inset of the figure.

Each point on the plot is the difference between

the dose obtained with the collimator jaws open

and the dose obtained with the useful beam blocked

off. Measurements at points greater than 5 cm

3 from the geometric edge were taken with a 3-cm ion cheimber in a Presdwood phantom; all other

3 points were obtained with a 0.6-cm ion chamber in

a water phantom. The dashed curve in Figure 2

25
a) Open beam b) Blocked beam

Patient

scatter J

Leakage

Figure 1 - Schematic illustration depicting the sources of stray radiation reaching

a point in a patient outside the useful treatment beam. The stray radiation due

to scatter can be obtained by subtracting the dose measiired in b) from that measured in a)

^ 0.5

0.2

0.1

0.05

0.02

0.01

W Total scatter 25 MV

Total scatter "xo

Maximum leakage level

I

I Edge of

^

I useful beam

|

Phontom scatter 25 MVn

10 20 30 40 50 60

Distance from central axis (cm) 70

II23F

Figure 2 - Dose due to photon scatter along the longitudinal axis at depth 10 cm in a

2 60 phantom irradiated by a 30 x 30-cm beam. Open circles: Co 80 cm SAD; closed

circles t 25-MV X-rays 100 cm SAD. Doses are normalized to 100% on the central

axis of the useful beam at depth 10 cm.

26
ENERGY IMPARTED TO A PHANTOM OF 20 x 30 CM^

CROSS-SECTION AND MASS 70 KG FOR AN ABSORBED

DOSE OF ONE GRAY DELIVERED TO THE CENTRE OF THE PHANTOM

25 Mv x-rays

(Varian Clinac-35)

fin ^"Co X-rays

(AECL Theratron 780)

30 x 30 cm^ beam

100 cm SAD

2 30 x 30 cm beam

80 cm SAD

Energy Imparted

within useful beam

16.2 joule 15. 8 joule

Energy 1 mparted

beyond useful beam

due to scatter

1.39 joule 1. 69 joule

Energy 1 mparted

beyond useful beam

due to 0. 1% leakage

0. 052 joule 0.061 joule

Ratio of scatter

to leakage

27: 1 28: 1

II47F

Table 1

shows the contribution of the scatter arising

from the phantom. It was separated from the

total scatter using the method described by

[ 6

1

Karzmark

It is evident from the figure that the

scatter dose exceeds the allowed leakage dose up

to a considerable distance from the central axis

- approximately 50 cm for the 25-MV beam and 60

' cm for the ^*^Co beam. In any other irradiation

condition, of course, the value will be greater

or lesser depending primarily on the volume of

medium irradiated.

An alternative way of assessing the relative

importance of scatter and leakage is to compare

j the relative energy imparted to the phantom

(integral dose) outside the useful beam. This

calculation has been performed for the

irradiation parameters of Fig. 2 and the results

are shown in Table 1. In this table the useful

beam is considered to be the volume bounded by

the geometric edges of the beam and the edges of

the phantom. Calculations within the useful beam

and to 5 cm beyond the geometric edge were

obtained by summation of dose values calculated

at 1-cm intervals through the phantom. The dose

I values were calculated using the treatmentI [781 planning computer algorithms of Cunningham ' with machine parameters selected to fit the experimental data. Calculations beyond 5 cm from

I the beam edge were made by direct summation of

experimental scatter-dose values measured at

5-cm intervals through the phantom. For the

leakage calculation it was assumed that the

leakage dose was constant and equal to 0.1% of

2 the dose on the central axis of a 10 x 10-cm

field at the depth of dose maximum.

It is seen from the table that the energy imparted outside the useful beam due to scatter

is more than 20 times greater than the energy due

to leakage. This is true for both the 25-MV beam

60 and the Co beam and therefore presumably for energies in between. The ratio will of course be

less for smaller beams. Yet, even for a small

2 beam (10 x 10 cm ) directed at one end of the

phantom, similar calculations show that the scatter energy imparted is more than five times

greater than the allowed leakage. In short, in

conventional radiation therapy practice, patient

dose from scattered radiation is much larger than

that from leakage .

It can be concluded from this that existing

maximum leakage levels could be increased

substantially. If it is indeed true that there

are real difficulties in having present equipment

meet the 0.1% leakage requirements than we would

recommend that the leakage requirements be

relaxed somewhat (by perhaps a factor of 5) . No significant additional risk to the patient would

result.

References

'"^Report of Committee IV (1953-1959) on Protection Against Electromagnetic Radiation

Above 3 MeV and Electrons, Neutrons and

27
Protons. ICRP Publ. No. 4 (Pergamon, Oxford,

1964)

Suggested State Regulations for Control of

Radiation Section F9 - X-Ray and Electron

Therapy Systems (1 MeV and above) , revised

draft (U.S. DHEW, Washington, DC. 1976).

E. J. Axton and A. G. Bardell, Phys. Med.

Biol. 17, 293 (1972)

R. M. Wilenzick, P. R. Almond, G. D. Oliver,

and C. E. DeAlmeida, Phys. Med. Biol. 18 , 396 (1973) .

G. D. Oliver, J. A. Rawlinson, and R. Cox, AAPM Q. Bull., p. 97 (June 1973) (abstract).

C. J. Karzmark and T. Capone, Br. J. Radiol.

41, 222 (1968) .

J. R. Cunningham, Phys. Med. Biol. r7, 42

(1972)

J. R. Cunningham, Phys. Can. 25^, 23 (1969)

(abstract)

28
NBS SP 554 (1979) INTERNATIONAL STANDARDS FOR MEDICAL ACCELERATORS

G R Higson

Department of Health & Social Security

London, Ehglemd

The International Electro—technical Commission is engaged in a major

programme for the development of standards for medical electrical

equipment. A standard for the radiation safety of electron medical

accelerators will soon be published. The main features of this

standard are described and some of the background to them is discussed.

(Accelerators; lEC) interlocks; leakage; radiation; safety; standards)

Introduction

The International Electro—technical Commission

(IBC) is the major international standardisation

I body for electrical equipment. It was formed in

I 1906 and has since become the electrical

I counterpart of the International Organisation

for StEUidardisation (iSO) and 41 countries are

; members

In 1968 a Technical Committee (IBC TC 62) on

Electrical Equipment in Medical Practice was

created "to prepare international recommendations

I concerning the manufacture, installation and

I application of electrical equipment in medical

I practice". This Committee now covers a wide

variety of medical electrical equipment throu^

4 sub-committees and 24 working groups.

Working Group 1 of Sub-Committee 62C ^eam

Teletherapy and Particle Accelerators) started

work in 1972 and was given as its first task the

production of a comprehensive standard covering

all aspects of medical electron accelerators,

their accessories and buildings. Work is in

hand on 4 parts of this comprehensive standard:

General Requirements; Radiation Safety;

Electrical and Mechanical Safety, and Performance.

Parts 1 and 2 are now in the final processes of

publication and I hope will become available

towards the end of this year. It has taken a

long time to produce this standard but it must

be remembered that the protocol of the IBC

depends not only on reaching agreement in

comnittee but on obtaining a concensus of

j agreement from the member countries and the

document containing Parts 1 and 2 has been

circulated to National Committees 3 times: In

(a ) 1974 as a draft for discussion; ' in 1975 as a

(2)

draft for views on its acceptability; and in

1976 for voting under the Six Months Rule^^^ and

in 1977 some modifications to the final docximent wre circulated for voting under the Two Months

Procedure

la discussing the radiation safety requirements

specified in this document it is important to

emphasise that it does not attempt to describe

the opriimum conditions of an accelerator for

giving radiotherapy treatment. This description

is contained in Part 4 - Performance - which is

currently in draft. Rather it describes certain

essential safety provisions and such levels of

performance beyond which the equipment is positively unsafe.

Dose Monitoring

There must be two independent dose monitoring

systems either of which will shut down the

accelerator when a pre-^et value is reached.

Both the dose monitoring systems must work

correctly and consistently at dose rates up to twice the maximum specified for the equipment.

If that level is reached a separate dose rate monitoring system must shut down the equipment.

In the event of a mains power failure information

about the dose given must be held for at least

twenty minutes in such a way that it can be retrieved

.

Some rather crude checking of the dose distribution is required by a system which will stop

the irradiation if the measured distribution

chaoiges by more than 20^ as such a change will

indicate a major failure e g in the beam control

system or in the filtering.

If irradiation is stopped for any reaison other

than the primary dose monitoring system reaching

the preset value, then a warning signal must be given at the control panel. If, as will be expected, the irradiation is stopped by the

primary system working correctly, then the other

dose monitoring system must be checked for

correct functioning before the next treatment

can be started.

A timer may be fitted as an optional back-up to the dose monitoring systems but, if it is, it must comply with requirements analogous to those

for the dose monitoring sjrstems.

Radiation Type

If the accelerator is capable of both electron

and X-ray treatment then the radiation type

must be selected at the control panel before

irradiation is possible. If actions have to be carried out in the treatment room (e g replacing

a target by a scattering foil) in order to change

the radiation type, then radiation must be

29
prevented if the actions at the control panel

and in the treatment room do not agree and

furthermore no display of the selected type

must be given until all these selection

operations have been completed and are in

agreement

Selection of Other Operating Conditions

There are analogous provisions for the selection

of energy, stationary or moving^beam therapy,

field flattening or beam scattering filters,

wedge filters and applicators. Specific

requirements are added according to the particular

operating condition e g systems to check that

filters are properly inserted in the equipment,

that the equipment moves during moving^beam therapy and remains stationary during stationarybeam therapy and that the electron beam energy

does not vary by more than 20^ from the selected

value

In the case of scanned—beam accelerators the

scanning control signals must be monitored to ensure that they remain within the manufacturer's

specified limits.

This combination of checking and interlock systems

has been criticised for its complexity and the

corresponding increased cost of accelerators but,

in fact, almost all of these systems are in use

to varying degrees on accelerators being used

today and the IBC document is regarded by its originators as simply combining the best

features of existing equipment rather than as breaking new ground.

Contamination of the Treatment Beam

Nominal Energy (MeV)

Figure 1 : Limits of absorbed dose at the

surface during X-irradiat ion

Permissible limits of contamination of an X-ray

beam by electrons and an electron beam by X-rays are shown in Figures 1 and 2, These figures

ui m

o! ft nJ I 1 '

I |_

10 20 30 40 50

Nominal Energy (MeV)

Figure 2: Limits of stray X-radiation

during electron irradiation

were originally derived from the German standard

(5) DIN 6847^-^'' and the limits for X-ray contamination of an electron beam agree with that standard.

Those for electron contamination of an X-ray

beam are more severe than DIN 6847 • No limit is set for neutron contamination of an X-ray beam

but the manufacturer is required to mesisure it and to declare the level to the purchaser.

Leakage Radiation

Vfiien we first considered the subject of leakage

radiation we turned naturally to ICRP 15 with

the intention of simply referring to the levels quoted in that document. However, consideration

of ICRP 15 led to the conclusion that that

document was rather inappropriate for modem

accelerators capable of large field sizes.

Eventually a more complex specification was produced which is largely based on the idea of

giving special care to the unwanted radiation

arising over a plane of 2 metres radius in which

the patient mi^t be found and relaxing a little

outside this area.

The permissible leakage through the diaphragms may be 2% of the Central Axis Dose — as permitted

by ICRP 15 — for equipment with a maximum field

size up to 22 x 22 cms. For accelerators with

maximum field sizes larger than this the

permissible leakage through the diaphragms is reduced in accordance with a formula which keeps

the leakage dose-«.rea product constant. The

effect of this is shown in Figure 3 and it will

be seen, for instance, that an accelerator

capable of a maximum field of 40 x 40 cms must

have diaphragms which permit leakage no

more than 0,6^ of the Central Axis Dose.

30
2 f

Area of largest field (crlx 10^)

Figure 3: Leakage radiation throu^

beam limiting devices

The structure outside the diaphragns must not allow leakage radiation over the 2 metre radius

"patient plane" to exceed an average of 0.1^

(ICRP value) of the Central Axis Dose althou^

2 individual areas up to 100 cm may rise to 0,2%,

In the area outside the "patient plane" leakage

radiation must be restricted to 0.5^ of the

Central Axis Dose at 1 metre from the path of

the electron beam, Althou^ this leakage level

represents a considerable relaxation on ICRP 15 the disposition of the protected region introduces

an important new concept.

The logic of this approach is that in the area

outside the patient plane, the main consideration

is the protection of persons outside the treatment

room. This relaxation of the ICRP figure requires

the addition of a few inches of concrete to the

treatment room walls and is considered to be more cost-effective than building more absorption into

the accelerator structure

These concepts, and figures, have been disciissed

with the ICRP Committee 3 during the years of

development of the IBC standard. No objections

have been raised by ICRP althou^ that must not

be taken as implying that the ICRP are in any

way supporting the values given in the IBC standard, ICRP are now preparing to revise ICRP

Publications numbers 15 and 22 and it is our

hope that there will be no clash between ICRP

recommendations and IBC standards.

Neutron Leakage

The setting of standards for neutron leakage

posed particular difficulties both becaxise of

the absence of authoritative recommendations and

the shortage of reliable data about the characteristics of existing accelerators. It was uncertain whether the 0.1^ limit suggested in ICRP 15 was intended to apply to accelerators

and whether it was intended to include a neutron component but paragraph 1.3,5 of ICRP 15 states "in the application of new techniques or

equipment, e g neutron therapy, this requirement

may not be practicable". The Committee were

uiclined to interpret this as less than mandatory

but, in attempting to keep as close as possible

to ICRP recommendat ions, the first draft standard

contained the requirements that neutron leakage

"should not exceed 0,01 ^5 (of the Central Axis Dose at the same distajice ) and shall not exceed

0.02f&', (These figures are in rads ),

When this draft was circulated comments were

received from many National Committees, including

the USA, that the limit of 0,02% was impractical.

The results of Axton and Bardell^^^ and

(7) Wilenzick et al ' showed that neutron leakage

doses could vary from 0,015 - 0,08^ of the

Central Axis Dose, Figures of the same order were later published by other workers.

Consideration of these results led eventually to the present formulation that over the 2 metre

radius patient plane the average dose for neutron leakage shall not exceed 0,02^ of the

Central Axis Dose at the same distance and the

maximum dose in the patient plane shall not exceed 0,05^ thus allowing for the relatively

hi^ values at the edge of the useful field.

This formulation calls for all accelerators to achieve the performance currently achieved by

the best available accelerators but no logical

argument could be found for requesting a lower

value of leakage. An average leakage of 0,02%

implies that a patient receiving a therapy

dose of 6,000 rads would experience a whole body

neutron dose of 1 ,2 rads,' or 12 rems assuming an RBE of 10, This value can hardly be considered

as excessive for a patient undergoing therapy

treatment and consideration of the practical

difficulties in patient treatment that would

result from the addition of sufficient neutron shielding to reduce these values significantly

led the Committee to confira these values which

were accepted in the last circulation of the

document to National Committees.

In our view there is no case for pressing for

more severe standards than these in the present

state of knowledge. If all accelerators to be

produced in future can meet all the requirements

contained in this standard the already very small

risks to patients resulting from treatment by

radiotherapy equipment will be reduced yet

further without making severe demands on new

technology or novelty in desi^. Almost every

featiire called for in this standard has been

shown to be practicable at reasonable cost and

the aim of incorporating all the best features

currently available on individual machines on all accelerators is regarded as a cost-effective

approach to improving the quality of electrical

equipment used in medicine.

31
References

(1)

(2)

(3)

(4)

(5)

(6)

(7)

Medical Electron Accelerators in the Range

1—50 MeV, Section One; General. Section

Two: Safety Requirements for Equipment

(Manufacturer), 620 (Secretariat )3. Inteniational Electro—technical Commission, Geneva (June 1974).

Medical Electron Accelerators in the Range

1—50 MeV. Section One: General. Section

Two: Safety Requirements for Equipment

(Manufacturer). 62G (Secretariat )7. International Electro—technical Commission, Geneva (September 1975).

Medical Electron Accelerators in the Range

1—50 MeV. Section One: General. Section

Two: Safety Requirements for Equipment

(Manufacturer). 62C (Central Office )4. International Electro—technical Commission, Geneva (November 1976).

Amendments to Document 62C (Central Office )4: Medical Electron Accelerators in the Range

1—50 MeV. Section One: General. Section

Two: Safety Requirements for Equipment

(Manufacturer), 62C (Central Office )8.

International Electro—technical Commission, Geneva (May 1978).

Draft Revision of DIN 6847, Medical Electron

Accelerators: Radiation Protection Rules

for Manufacturers. Deutches Institut fur

Normung, Berlin, (1975).

Neutron Protection from Electron Accelerators

vtseA. for Medical Purposes. E J Axfcon and

A G Bardell. Phys Med Biol IJ, 293 (1972).

Measurement of Past Neutrons Produced by

Hi^ Ehergy X-ray beams of Medical Electron

Accelerators. R M Wilenzick, P R Almond, G D Oliver and C E De Almeida. Phys Med

Biol 18, 3, 396-408 (1973).
IBS SP 554 (1979) STATE REGULATIONS FOR MEDICAL ACCELERATORS

C.H. Smith

Oklahoma State Department of Health

Oklahoma City, Oklahoma 73152

The "Suggested State Regulations for the Control of Radiation"

includes a section regarding medical accelerators. The provisions

of that section are reviewed.

(Regulations, SSRCR, Accelerators, Leakage Radiation, State)

Introduction

State radiation control agencies have a need for comprehensive regulations regarding

the use of ionizing radiation. Those regulations need to cover all types of radiation

sources and be compatible with regulations of federal agencies functioning in radiation

control.

These regulations also need to be reasonably uniform from state to state to allow

industry to function on a national scale.

The "Suggested State Regulations for the Control of Radiation" (SSRCR) are written to be

used by the individual state agencies when ra- diation control regulations are being promulgated .

The Suggested State Regulations for the Control of Radiation does contain a section re- garding medical accelerators which will be used

by many state radiation control programs as the basis of constructing regulations regarding

medical accelerators which will be effective in

their jurisdiction.

Background of the SSRCR

State radiation statutory authority is generally written to cover all sources of ionizing radiation. This wide authority is not duplicated at the federal level. To be utilized as a model the SSRCR must cover all sources except those over which the NRC retains

authority even in "agreement states", e.g. nuclear power plants, be compatible with existing regulations of several federal agencies,

and recommendations of nationally or internationally recognized authorities.

The SSRCR can serve only as a reference

when state agencies are writing or revamping

their regulations. Program goals, legal re- quirements, and fiscal restrictions within an individual state make it such that there may

never be an absolutely uniform set of regulations regarding ionizing radiation across

the United States. However, the difficulty of writing regulations is such, that individual

authorities will probably prefer to use or shape the model regulations for their own

state regulations. But no state authority is obligated to adopt any portion of the SSRCR.

Currently the various parts of the SSRCR which regard different types of radiation

sources or particular uses of radiation sources

are written or updated by working groups, ap- pointed by the Conference of Radiation Program

Directors from members of that conference.

Federal representatives also serve on each

working group.

The current 1978 edition has just been

published and was prepared by the Conference

of Radiation Control Program Directors, the FDA/Bureau of Radiological Health, the U.S. Nuclear Regulatory Commission, and the Environmental Protection Agency, and the Public Health Service.

The SSRCR was first published in 1962 by

the Council of State Governments with the assistance of the U.S. Public Health Service

and the U.S. Atomic Energy Commission.

The original version was intended to pro- vide for the control of all sources of ra- diation and did include a- part regarding the use of x-rays in the healing arts.

The SSRCR was updated in 1964 and 1966 to maintain compatibility with the U.S. Atomic

Energy Commission regulations. The 1970

edition among other changes, modified the part

regarding the use of x-ray in the healing arts

to reflect recommendations made by the NCR?

Report #33.

The 1974 edition of the SSRCR included new

parts regarding instruction to workers, analytical x-ray equipment, and particle accelerators. A rationale for changes was added as part of the format. The part regarding the use of x-rays in the healing arts was given a total rewrite.

One of the reasons for that rewrite was

the federal diagnostic x-ray standard which

was published in 1972. Section 360F of PL 90-602, The Radiation Control Health and Safety Act of 1968, requires that if a standard has been established by the federal

government regarding electronic product radiation that "no state or political subdivision

of a state shall have any authority either to establish or to continue in effect any standard which is applicable to the same aspect of performance of such product and which is not identical to the federal standard."

33
The 1978 edition of the SSRCR which

should be generally available shortly, in- cludes a section directed toward medical

accelerators and a rewritten section regarding

lower energy therapy units.

Accelerator Regulations

The original version of the SSRCR con- tained a section which stated requirements for therapeutic x-ray installations. This section was maintained in later editions with few changes. The requirements stated were largely

directed toward the 250 kVp units which were

being replaced by the cobalt teletherapy units

in radiation therapy applications at that time.

With the emergence of the high energy accelerator units it became apparent that the requirements as written in 1962 needed re- vision to provide adequate safety measures for accelerators

When the working group assigned responsibility for writing the portion of the SSRCR

assembled its materials and references, the existence of an excellent document, which was

still in draft form, produced by the Inter- national Electrotechnical Commission was noted

That document, as well as recommendations made

by the National Council on Radiation Pro- tection and Measurements, the American Association of Physicists in Medicine and individual

comments, were used as the basis of the SSRCR

section regarding medical accelerators.

The final version of the SSRCR accelerator requirements differentiates between new

equipment and existing equipment using a date

included in the definitions. The date was

given a special notation indicating to the reviewing agency that the implications of including the SSRCR date within their regulations should be carefully considered.

The SSRCR accelerator regulations, if adopted and placed into effect by an individual state, may require a given amount of backfitting on accelerators. During the construction of the model, consideration was given to the cost and practicality of pro- viding a particular safety feature on an accelerator where that safety feature was not part of the original accelerator design versus

the additional safety which would be provided

by the feature.

Provisions of SSRCR Accelerator Regulations

Leakage Radiation to the Patient Area.

It became apparent very early in the groups discussions that any criteria written

regarding accelerator leakage radiation was going to be controversial. Many conflicting

opinions regarding leakage radiation were

received by the group during the time that the model was being written.

The provision regarding leakage radiation

is stated in terms of new equipment (defined

as equipment manufactured after August 1, 1980)

and existing equipment (defined as equipment

manufactured on or before August 1, 1980.)

The requirement for new equipment is stated as 0.1% of the useful beam in the patient plane outside of the useful beam and includes the neutron component as part of the leakage radiation.

It was recognized that some currently

produced high energy accelerators cannot meet

that standard, but the group was of the opinion that a more lenient standard could not be stated until the ICRP or the NCRP took such

a position. It was concluded that the 0.1%

criteria as stated in ICRP Report Number 15,

1969, should be retained for new equipment.

The requirement for existing equipment is stated as 0,1% of the useful beam but excluded

the neutron component from being considered as part of the leakage radiation.

The criterias are stated in terms of rems rather than rads. Another part of the SSRCR

contains a conversion table for neutron flux

equivalents. That table is identical to that found in lOCFR Part 20 of the Nuclear Regulatory Commission Regulations.

The patient plane concept which is utilized wao taken from the lEC draft.

Leakage Radiation Outside the Patient Area

The criteria stated in this section of 0.1% of the useful beam for x-ray leakage and 0.5% for neutron leakage can be met by cur- rent units and the figures are included to give a base facility shielding design.

A criteria of 2% transmission through the beam limiting device is stated. Previous versions of the SSRCR had stated 5%. The ICRP Committee 3 noted the inadequacy of the 5% criteria.

Beam Monitors . The requirement is written in of new equipment and existing equipment. New equipment is required to have two useful beam monitoring systems. Existing

equipment is required to have one monitoring

system.

Termination of Irradiation by the Dose Monitoring System. The requirement is similar

to lEC recommendations but differs in that a secondary system is required to terminate irradiation when 102% of the pre-set dose monitor units has been reached rather than 115%.

Timer. A timer is required for both new

and existing units. It differs from the lEC

draft where a timer is suggested but not required.

34
Absorbed Dose Rate. Parallels lEC draft

recommendations. A portion of the last lEC draft regarding a device to terminate irradiation if the dose rate exceeds 10 times maximum

was not incorporated.

Surveys

.

A review of all new or significantly revised facilities is required to be

performed by a qualified expert.

Other Requirements. Many parrallel and

are similar in language to the lEC draft re- quirements. These include requirements for filters, beam quality, beam symmetry, selection of display of dose monitor units, termination of radiation, interruption switches,

termination switches, selection of radiation

type, selection of treatment mode, location

of focal spot and beam orientation, and

system checking facilities.

Calibrations. Semi-annual calibrations

are required and are to be performed by or under the direct supervision of a qualified

expert using instrumentation traceable to a national standard.

Spot Checks. The spot check procedures

are required to be developed by a qualified

expert but not necessarily performed by that

person. The procedures must state the time

intervals at which the spot checks must be

performed.

Operating Procedures.

Compliance and Enforcement

The model does specify some conditions

under which some measurements will be made,

but in general, the method of demonstrating

compliance is left to the registrant.

The language used permits the use of manufacturer furnished data to demonstrate compliance with some measurements which may be

difficult to perform at an individual facility,

e.g. neutron leakage to the patient plane.

Administrative procedures will vary from

state to state regarding inspection frequency

and records to be maintained for an individual

facility. The facility may be asked to submit

plans prior to initiating installation or may

be asked to demonstrate compliance to all

parts of that states regulations prior to initiating patient treatment.

Future Accelerator Regulations

The format of the SSRCR and the procedures which have been arranged will allow a frequent updating of the model. Comments which you may have regarding any provision of the SSRCR, particularly the portion regarding

accelerators, may place such comment with the FDA/Bureau of Radiological Health, Rockville,

Maryland' 20857.

If the FDA/Bureau of Radiological Health

does place into effect regulations regarding

accelerators it may require that the current SSRCR provisions be revised, if requirements

regarding the same aspect of performance are stated which are significantly different from

those stated in the SSRCR.

NBS SP 554 (1979) THE RBE OF NEUTRONS

Harald H. Rossi

Radiological Research Laboratory

Department of Radiology Cancer Center/Institute of Cancer Research

Columbia University College of Physicians S Surgeons New York, New York 10032

The RBE of neutrons is not only a function of energy but more importantly a function of absorbed dose. A combination of experimental and theoretical

findings permits at least a fair estimate of the dependence of neutron RBE

on both of these factors. It would appear that in radiotherapy with high

voltage electron accelerators neutrons have a minimal biological effect on

the treatment volume but that they may be an important and perhaps a biologically dominant factor in leakage radiation.

(Neutrons, radiotherapy, RBE)

In efforts to obtain a better insight into

the action of ionizing radiations, biological

systems have very often been irradiated under

varying conditions. Perhaps the most important

physical aspects altered in such experiments have

Jj been the temporal and the spatial distributions

j of radiation energy. Obviously the interpretation

I of the results obtained requires an understanding

of the physical change involved. This is rather

trivial in the case of temporal changes but far

, from simple in the case of the spatial ones. In fact some theories of the biological action of

radiation have been proposed that imply impossible

energy distributions in the irradiated cell.

The principal factor in considerations relating to spatial energy distributions in irradiated

tissues is the difference between the absorbed

dose , D, i.e., the average value of the quotient

of energy and the mass in which it is deposited,

and the specific energy , z, which is the actual

value of this quotient in regions of specified

dimensions. If these regions and/or the absorbed

dose are so large that the dose is imparted by

many charged particles the difference between z and D is small. However, if fewer particles are involved statistical fluctuations become increasingly significant. In the extreme case, when the

I mean number of particles is much less than one, there is usually no energy deposited but in those

cases when z' is not zero it is generally much

larger than D, and its magnitude is independent

of D. In this condition D merely determines the fraction of the regions that receive energy.

For regions of given dimensions the absorbed

doses for which single particle action predominates depend on particle LET or, more generally,

I on radiation quality. Evidently single particle

I action obtains when D is much less than z^, the average specific energy due to one particle (or more accurately due to one event , i.e., the appearance of one particle or of several correlated particles) . Taking the nucleus of the typical

! mammalian cell to have a diameter of 5 ym, it

receives a from the electrons secondary to

!, hard gamma radiation that is about 4 mGy (0.4 rad).

On the other hand z^ for the secondaries of typical

fission neutrons is of the order of 400 mGy

(40 rad) . As will be shown the region in which

energy concentration is critical is smaller than

the nucleus being_only about 1 pm in diameter

.

For such regions is about 100 mGy (10 rad) for gamma rays and 10 Gy (1000 rad) for fission neutrons.

It is difficult (although not impossible) to observe the effects of 100 mGy of gamma radiation

on mammalian systems. On the other hand a neutron dose of 10 Gy produces such extensive and

complex damage that most significant research is carried out at lower doses. Thus mammalian radiobiology generally involves multiple event action

in Y or X irradiations and. single event action in

irradiations by neutrons having energies of the order of 1 MeV.

One reason for this is that in addition to greatly different event frequencies there is a well established difference in biological effectiveness which for somatic effects on higher

organisms is invariably higher for neutrons. This in turn indicates that in general the biological effectiveness of radiation increases

more than proportionately with specific energy. There can be little doubt that the dependence is at least very largely one on the square of the specific energy. This is indicated by the general dependence of RBE on absorbed dose.

As shown in Fig. 1 the relation between the logarithm of RBE and the logarithm of the neutron absorbed dose, D^, is a line of slope -1/2 over

the range from a few gray to a few milligray.

Since this is clearly in the range of single

particle action, e , the fraction of cells showing

any type of effect, must be proportional to the absorbed dose since the latter is proportional to the frequency with which energy of any magnitude

is deposited in cells. Thus,

^ = ^ ^N (1) On the other hand the RBE (=D^/'^n) is according to Fig. 1 related to D by:

37
log(D^/Dj^) = logC - 1/2 log D^^ (2) or V^N =

Combining (1) and (3)

e = KC^D 2 X where C is another constant which for the fission

and 0.43 MeV neutrons in Fig. 1 is about 4.5 with

absorbed doses expressed in Gy. Relation (2) applies to low LET doses that are so large that

specific energies do not differ siibstantially

from them in domains larger than a fraction of a micrometer . Although these considerations only

show a dependence on for low LET radiation it seems reasonable to assume the same dependence

for high LET radiation particularly since observations at intermediate LET are also consistent with

a 7?- dependence. ,

.OOOI .OOl 01 .40 I 10 NEUTRON DOSE (Gray)

combinations of pairs of sublesions . In the site

approximation the latter are assumed to be produced

at a rate that is proportional to the specific energy in the site in which they combine at a rate that does not depend on their separation. If a correction is applied for saturation at high

values of z one obtains a relation between neutron energy and RBE (Fig. 2) that is in general agree- ment with observations (Fig. 3).

NEUTRON ENERGY (MeV)

Fig. 2 - Theoretical relation between y^, the dose

mean lineal energy,_and neutron energy,

y* is the value of y^ corrected for saturation. According to theory the. RBE should be proportional to y*.

100 1 r

Fig. 1 - Logarithmic plot of the RBE of low energy neutrons vs. neutron dose.

la: Shellabarger et al.^

2 lb:

2:

3:

4:

5:

Bateman et al. Sparrow et al.

4 Hall et al. Biola et al.^

, 6 Awa

RBE

10

.01

-> 1 r- 1 1 11 \ r

.1 I 10 NEUTRON ENERGY (M«V) 100

6 : Rossi

The quadratic dependence of biological

effect on specific energy is one of the basic

g tenets of the theory of dual radiation action.

The other tenet derived from a detailed analysis

of the relations shown in Fig. 1 is that the average diameter of the subnuclear volume in

which the specific energy is pertinent is about

1 ym.

The theory postulates that the lesions

responsible for radiation injury are due to the

Fig. 3 - Observed dependence of PBE on neutron

energy.

1: Lens opacification at x-ray dose 40 rad

2 (Bateman et al.)

2,3: 50% growth reduction of Vicia Faba

(anoxic and oxygenated) (Hall et al.-) ^

4: Cellular inactivation (initial part

of the survival cvirves) (Barendsen)^

5: 37% depletion of spermatogonia (Bateman)

38
The curves in Fig. 3 refer to various kinds

of effect at selected levels of effect. At the

same absorbed doses there is frequently not a strong dependence of RBE on effect as indicated

in Fig. 1. Thus for many effects a fair approximation to the RBE of neutrons relative to x-rays

can be derived by determining from Fig. 3 the RBE

for the neutron energy of interest relative to that of 0.5 MeV neutrons, dividing the absorbed

dose of interest by this RBE and noting in Fig. 1 the RBE corresponding to this (equivalent) dose

of 0.5 MeV neutrons

.

Recent research has indicated that the site

model is an approxmation of limited accuracy. In particular the molecular ion experiment"'"''' has

shown that the distance of about 1 ym is the mean

of a highly skewed distribution with about half of

the interactions taking place at separations that

are at least 10 times smaller and the remainder

probably occurring throughout the nucleus. A

12 recently piablished generalization of the theory

takes account of this complication but does not permit an accurate numerical determination of

neutron RBE since this requires the knowledge of

presently unknown parameters

.

A factor that may be related to differences

in the distribution of separations is the fact

that the maximum RBE values at low doses vary

between effects. As shown in Fig. 1 the RBE of neutrons having energies of the order of 0.5 MeV

reaches a maximiim of about 50 for a somatic

mutation while for at least two effects it exceeds

more than 100. On the other hand the maximum RBE

for the killing of cells in tissue culture is unlikely to exceed 30 and there is one in vivo

13 effect for which it differs little from 1.

Thus, the above conclusions on the RBE of neutrons are general but may not fit specific

cases. However it is clear that in the great

majority of instances the neutron RBE increases

with decreasing dose and that it can, at least

for energies of the order of a few hundred KeV,

reach values of the order of 100.

• Hlroahlmi I = 2.2 i l6'+ 8x16*Kh|

oNagaukI 1 = 1.8 x lO^ t S « 10* k|„

Fig. 4 - Annual mortality from all malignant

neoplasms in Hiroshima and Nagasaki

(1950-1974) vs. total kerma. The limits

correspond to ± one standard deviation.

In applying these findings to the radiation

from high energy accelerators it should be kept in mind that the neutrons usually have higher energies

which reduces their RBE compared with moderated

fission neutrons (E eff = 0.5 MeV) as is indicated

in Fig. 2. Furthermore because of the high doses

in the direct beam the effect of these neutrons is probably low compared with that of the x-rays

.

For leakage radiations which impart much lower

doses the contribution of neutrons may well be

dominant. Although the attendant risk may be

small - especially considering the seriousness of

the condition being treated - all reasonable

efforts should be made to reduce the doses due to leakage neutrons and to do this in preference to similar reductions of x-ray doses.

References

This behavior is strongly indicated by the difference in cancer mortality between Hiroshima

(where roughly 10% of the organ doses were due to neutrons) and Nagasaki (where the neutron doses

were much less) . Fig. 4 shows mortality (1950 - 1974) from all malignant neoplasms as a fraction

of the total free-in-air tissue kerma for the two cities indicating that the relatively small neutron dose at Hiroshima had a profound effect.

It can be calculated that for a neutron kerma of

1 rad the cancer mortality was about 3 x 10 ^. Although the risk factor for gamma radiation is difficult to quantitate it evidently depends on

the dose and is probably more than two orders of

magnitude less at doses of a few rad.

'" Shellabarger, C.J., Brown, R.C., Rao, A.R.,

Shanley, J. P.,. Bond, V.P., Kellerer, A.M.,

Rossi, H.H., Goodman, L.J. and R.E. Mills. Rat Mammary Carcinogenesis following Neutron or X-Irradiation. Symp. on the Effects of Neutron

Irradiation upon Cell Function, Munich 1973. In Biological Effects of Neutron Irradiation , lAEAS-179/28, 405-416, 1974.

^ Bateman, J.L., Rossi, H.H., Kellerer, A.M.,

Robinson, e-V. and V.P. Bond. Dose-dependence

of fast Neutron RBE for Lens Opacification in

Mice. Radiat . Research 61: 381-390, 1972.

^ Sparrow, A.H., Underbrink, A.G. and H.H. Rossi.

Mutations Induced in Tradescantia by Millirad

Doses of X-rays and Neutrons: Analysis of dose- response curves. Science 176: 915-918, 1972.

39
4

5

10

11

12

13

Hall, E.J., Rossi, H.H., Kellerer, A.M.,

Goodman, L. and S. Marino. Radiobiological

Studies with Monoenergetic Neutrons. Radiat . Research 54: 431-443, 1973.

Biola, M.T., LeGo, R. , Ducatez, G. , Dacher, J. and Bourguignon, M. Formation de chromosomes

dicentriques dans les lymphocytes humains

soumis in_ vitro a un flux de rayonnement mixte

(gamms, neutrons) . In: Advances in Physical and

Biological Radiation Detectors , pp. 533-645,

IAEA, Vienna, 1971.

Awa, A. A. Chromosome Aberrations in Somatic

Cells. J. Radiat. Res . Suppl . 16:122-131, 1975.

Rossi, H.H. The Effects of Small Doses of

Ionizing Radiation: Fundamental Biophysical

Characteristics. Radiat . Res . 71:1-8, 1977.

Kellerer, A.M. and H.H. Rossi. The 'Theory of

Dual Radiation Action. Current Topics in Radiat. Res . Quarterly 8: 85-158, 1972.

Barendsen, G.W. Cellular Responses Determining

the Effectiveness of Past Neutrons Relative to X-rays for effect on Experimental Tumors. Europ . J. Cancer 7: 181-190, 1971.

Bateman, J.L., Johnson, H.A., Bond, V.P. and

H.H. Rossi. The Dependence of RBE on the Energy of Fast Neutrons for Spermatogonia

Depletion in Mice. Radiat . Res . 35:86-101, 1968.

Rossi, H.H., Bird, R. , Colvett, R.D., Kellerer,

A.M., Rohrig, N. and Y.M.P. Lam. The Molecular

Ion Experiment. Proc. Sixth Int. Symp. on Microdosimetry . Euratom, Eds. J. Booz and

H. Ebert, p. 937-947, Brussels, June 1978.

Kellerer, A.M. and H.H. Rossi. A Generalized

Formulation of Dual Radiation Action. Radiat . Res . 75: 471-488, 1978.

Carsten, A.L., Bond, V.P. and K. Thompson.

The RBE of Different Energy Neutrons as Measured by the Haematopoietic Spleen Colony

Technique. Int. J. Radiat . Biol . 29(l):65-70,

1976.

40
NBS SP 554 (1979) RISK OF CANCER BASED ON PROMOTION FROM CYTOTOXICITY*

T. D. Jones

D. G. Jacobs!

J. A. Auxier^

G. D. Kerr

Health Effects and Epidemiology Group

Health and Safety Research Division

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37830

Cytotoxicity is reviewed as a basis of risk assessment for environmental cancers. Linear relations of several cancers to organ-specific cell-surviving fractions have

led to the practical assumption that cytotoxicity can stimulate cell proliferation

which may permit altered cells to become tumors. Furthermore, an insult which re- tards cell proliferation or selectively blocks or destroys transformed cells may decrease tumor incidence. This simple assessment model for cancer growth seems to be unique because it requires no assumptions about the number of initiating events

per cell, bases the increased demand for proliferation on cytotoxicity estimates by

mathematical analyses of in vivo and in vitro irradiation of cells, and also assumes

that the exact number of neoplastic cells, their death rate, and the repair rate, are of secondary importance. The significant point appears to be that a sufficient

quantity of altered cells exist in order to respond to the proliferation stimulus

from normal or enhanced cell turnover. The use of this model for extrapolation to spontaneous cancer incidence in a population is discussed. Many investigators

suggest that cancers occur linearity with low dose; however, carcinogenic response

based on linearity with cytotoxicity does not seem inconsistent with linearity at low dose, and furthermore this relation seems to hold to saturated cell-killing

levels. Cytotoxicity may also provide a common basis for comparing risk from ionizing insults with risk from chemical insults; relating risk in different organisms

(over a wide insult range) to an experimental determination at a single insult

level; summing risk to different tissues; and estimating risk for partial body or nonuniform irradiation.

(Cell-proliferation; eel 1 -turnover; cytotoxicity; homeostasis; promotion; RBE; risk)

Background and Objective

There is an increasing interest in the extension

of radiobiological concepts in order to establish

protection guidelines for cancers induced by

chemical agents. Committee 17 of the Council of the Environmental Mutagen Society has reviewed

the problem and has suggested a unit of chemical

insult, the rem-equi valent chemical (REC) to correspond to radiation doses [1]. The REC is "that dose or product of concentration multiplied

by the time which produces an amount of genetic

damage equal to that produced by 1 rem of chronic

irradiation". It seems unlikely, however, that a direct equivalence will be adequate for carcinogenesis, and we may have to reconsider basic concepts if we wish to make a more appropriate pre- diction of effects from other carcinogens or to the synergistic or antagonistic effects arising

from exposure to mixtures of carcinogens.

Committee 17 goes on to state [1] "A system is urgently needed for monitoring the cells of individuals, particularly those at greater than average risk." One method, which is explored in this

paper, would be to relate cancer incidence to cytotoxicity (observed in vitro or estimated

in vivo by mathematical simulation) rather than

to radiation or chemical exposure dose. At the present time, absorbed dose serves as the basis

of cell-survival and insult-response models (for

radiation) and, in this paper, will only be used

as an index to relate carcinogenesis to the sur- viving fraction of cells. Protection concepts

based on cytotoxicity to cell cultures in con- junction with other screening techniques [2,3]

should quickly and inexpensively determine most

safe situations and also provide direct knowledge

about the degree of carcinogenicity of the insult. The objective of this paper is to develop a simple

model for risk assessment and not be overly con- cerned about the understanding of cancer or advancement of biology — tasks for which we are

not qualified.

Calculation of Radiation Cytotoxicity

For Marrow Cells

A formula for the surviving fraction of a cellular

population exposed to a mixed radiation field has been expressed by Katz [4]. In dose ranges of in- terest for risk assessments, the precision with

which Katz's analogue can describe the action of ionizing radiation on cell survival is well known

and essentially depends upon the determination of

*Research sponsored by the Office of Health and Environmental Research, U.S. Department of Energy under contract W-7405-eng-26 with the Union Carbide Corporation.

41
three radiosensitive parameters plus the ion-kill

probability of the track segment, P, and the stopping power (LEToo) of the track segment. These

parameters have been determined for hamster, HeLa,

kidney, leukemia, and mouse bone marrow cells for

several radiation environments [4]. This "fitted"

model can also "predict" survival in a different

radiation environment once the parameters have

been evaluated. Katz estimated the radiosensitive

parameters from the survival of colony-forming

units in mouse bone marrow after in vivo irradiation with D-T neutrons, X and gamma radiation

[5.6].

The choice of Katz's survival model was based on the availability of necessary parameters and the fact that it can be "predictive" in an untested

radiation environment. Linear-quadratic cellsurvival "fitted" models based on only two parameters [i.e., Jin (S) = -aD - bD^] are preferred by most investigators, but even though a corresponding theory is attached to all models, they

essentially remain empirically "fitted" equations

with an implied biological significance [7-p52].

For many applications, the initial slope of the

survival curve governs the extrapolation of effects; however, none of the malignancies investigated herein will depend upon this low-dose

region. Once survival has been computed for the various insult-response levels, the cell survival

model is no longer needed in the extrapolation of our model for carcinogenesis to the background

level of effects.

The mean value of the recoil-ion stopping power

in the atomic-bomb survivors is about 56.2 keV cm^ g"i [8] which is nearly identical with Katz's value of 56.1 keV cm^ g"i for mouse marrow cells exposed to 1 MeV neutrons. Katz's parameters of

P and L (LET<„) are selected in pairs and it is essential that the relative magnitudes be correct.

The distribution of dose to the active bone marrow

for the A-bomb produced radiation environments was obtained by the application of geometrical

probability-distribution-functions (CHORD doseattenuation operators) [9] to the depth-dose

curves corresponding to the A-bomb radiations.

Katz's equation, separated into the high-LET

(neutrons) and low-LET components (y-rays) was evaluated as a function of dose for mouse marrow

cells exposed to the A-bomb radiations and the respective curves are shown in Fig. 1. Till and McCulloch [10] estimated the survival of mouse marrow cells for in vivo irradiation. Over a wide dose range, they found a statistically higher

factor for in vitro irradiation and estimated an extrapolation number of 2.5 for in vitro versus

1.5 for in vivo. We used an extrapolation number

based on Broerse et al. [5] and estimated by Katz [4] at 2.5. The curve labeled as low-LET in Fig. 1 appears to be practically identical to the in vivo curves of Chadwick and Leenhouts [11] and McCulloch and Till [12]. The data in Fig. 1 seem

to be also in good agreement with estimates by

Mole [13] for the A-bomb survivors.

1.0 (—

0.1

1

>

trZD

0.01

0.001 I I I I I II

LOW LET

10

-1 10" 10 MARROW DOSE (rods)

Fig. 1. Cell survival for the bone marrow exposed to fission neutron radiations.

42
Regression of Mouse Leukemia to Surviving

Fraction of Marrow Cells

Regression of Human Leukemia to Surviving

Fraction of Marrow Cells

'MyelCid leukemia in mice [14] plotted as a funcItion of cell survival is shown in Fig. 2. Data

above about 400 rads for low LET and 100 rads for

high LET are not shown in this paper because of

statistical uncertainties associated with saturated sterilization processes (cell kill 100%)

and the frequently observed "therapeutic effect"

j of large single doses of ionizing radiations [15,

jl6]. At these high insult-levels, altered cells

jare readily killed and it appears that cellI killing effects outweigh increased carcinogenesis

[17] (personal communication from W. V. Mayenord)

Ullrich (personal communication) suggested that

bone marrow induced thymic lymphoma rates are

higher at low insult levels than myeloid leukemia

and should be investigated relative to cytotoxicity. Cells in the thymus gland do not pro- lliferate but are produced presumably by thymusj dependent (committed) cells from the bone marrow

I [18]. Thymic lymphoma incidence as a function of

I marrow-cell survival, down to dose levels of 4.8

rads of fast neutrons [19], is shown in Fig. 3. At once it becomes apparent that myeloid leukemia

and thymic lymphoma in certain inbred strains of

mice are more closely related to cytotoxicity or cytological alterations than to the robust^ concept of absorbed dose.

We assumed that human-marrow cells exhibit the same degree of radiosensitivity for lethality as mouse-marrow cells. Similar sensitivities are expected from tabulations of Okada [20-p266] and extrapolation techniques based on gene activation

for cell differentiation and total nuclear DNA content [21,22,23]. For the Japanese atomic-bomb

survivors, the mean neutron dose was 0.26 K(n),'* the mean autointegral gamma dose was 0.07 K(n),

and the mean photon dose was 0.55 K(y) [9]. The mean cell-survival values from insult profiles were compared with cell-survival values corresponding to the mean-dose values for the marrow

in adult survivors. These values were very nearly equal, and the mean marrow-dose factors

were applied to chronic and acute leukemia in the A-bomb survivors [24]. Leukemia, plotted as a function of cell survival, is shown in Figs. 4 and 5. The individual points are at the centers

of H used for Hiroshima and N used for Nagasaki. Because of statistical uncertainties in shielding, ^ dose,^ and cell survival, confidence intervals of the individual points should be two- dimensional and the deviation of the points about

the straight lines should be measured along a normal as shown in Fig. 4. The marrow-dose esti- mates were based on in-air doses or kerma. The

RADS OF LOW LET RADIATION

0-10 25 50 100 150

N/No= MARROW CELL SURVIVAL

Fig. 2. Incidence of myeloid leukemia in mice vs marrow cell survival.

43
RFM FEMALES 7 r ULLRICH ^^45rod/mln /

-= 40 h

20

rt

n 'V/^ 5 rod/mi n

_L J I I 1.0 0.6 0,2

RFM FEMALES

t- ULLRICH ? 40

u 201- o

0

_ 25 rod/min .A?

t.o

0.^

-^X rod/day

0.6 0.2

BALB/c FEMALES

3 1_ ULLRICH

2 [— "

1 ''^ " 1 rod/day 0 I

^1

25 1— UPTON

6.7rodA'iin 20 -

RF FEMALES

5 5j^_X__x-X^

60 rod/mln RF MALES

X J I 10 0.6 0.2

1 1.0

1 0.8

1 0.6

1 0.4

1 0.2

1 0

1

<I0 25

1 501 (00

1 2001

I I r

<.1 2 5 20 40

CELL SURVIVAL (N/Nq)

RAOS OF LOW LET

RAOS OF FISSION NEUTRONS

Fig. 3. Incidence of thymic lymphoma in mice vs marrow cell survival

1.0 0.8 0.6 0.4 0.2

N/No= MARROW CELL SURVIVAL

Fig. 4. Incidence of chronic leukemia vs cell survival in A-bomb survivors.

44
Based on Figs. 4 and 5, risk can be estimated as shown in Table I. In Table I, the absorbed dose

to the active marrow is given in column 3 and the neutron or gamma kerma which yields such a dose is given in column 1 or 2, respectively. Risk esti- mates from Table I are in good agreement with Court-Brown and Doll's recommendation^ that if the entire red marrow within the body were treated

with radiation of the same average energy, a dose

o

CO

<

Fig. 5.

0.8 0.6 0.4 0.2 0

N/No= MARROW CELL SURVIVAL

Incidence of acute leukemia vs cell survival in A-bomb survivors.

TABLE I. Risk of Leukemia in the Japanese A-Bomb Survivors Based on Cell Survival

Kernao

(rids)

HirroM

dose

(r»ds) Surviving fraction of marrow cells

Risk of Acute*

10"' (person-yr)"'

Risk of Chronic''''^ 10"' (person-yr)"' Relative Risk Acute Relative Risk Chronic

(n or (n or (n or y) + (n« or yf) or (n or y) + (n or y)

0.4 0.2 0.1 >0.99 >0.99 vO •vO M) 1 1 1 3.9 1.8 1 0.96 >0.99 2.5 •vO 1.5 2 2 1 19 9 S 0.82 >0.99 12 •vO 8.5 5.6 5.5 1 38 18 10 0.68 >0.99 21 M) 17 9.2 9.5 1 77 36 20 0.46 0.97 34 2 27 14.4 1.8 15 1 lis SS 30 0.32 0.94 43 4 - 31 0.5 18 2.6 17 1.3

154 73 40 0.22 0.89 50 7 1 21 3.8 1.5

192 91 50 0.14 0.84 55 11 1.5 23 5.2 1.8

231 109 60 •fO.01 0.78 63 14 2.5 2S 6.8 2.2 —t 182 100 0.55 29 5.5 12 3.8 364 200 0.18 52 10 22 6 S4S 300 <0.01 63 12 26 7

but Incidence from neutrons becomes I

% - D„/0.26. Dy/0.55.

^Spontaneous rat* taken at 2.5 x 10~' 1eukem1as/(p«rson-yaar).

"Spontaneous rate taken at 2.0 x 10-' leukemi as/ (person-year). ''CttlMted according to ^ (D„) x (D.^) • ^ which determines an Incidence I

*Fron Jones et a1. (8), 1.5 < Ky/K^ < 2.5 for survivors at 1000 m < distences < 1500 m; hence, IC, • 2K„ assumed to obtain D., 4.2 0„ 'Curve troai Figure 9 for Nagasaki. 'Decreased Incidence because of saturation.

45
of 30 to 50 R would double the expected incidence

of leukemia.

Other Radiation Diseases vs Cytotoxicity

Puck et al . [26] investigated survival curves of cells from normal human tissues of skin, spleen, ovary, conjunctiva, liver, appendix, and lung.

Except for lung cells, the radiosensitivities were

not greatly different for the other six tissues.

Todd (personal communication) also suggested that,

except for hematopoietic and lymphatic tissues, most other mammalism cells display a similar re- sponse to insults of ionizing radiation [20-p266]. Mean tissue-kerma-in-air estimates^ were modified

by CHORD operators for active bone marrow [9] in order to estimate whole-body dose levels [27], and Katz's cell-survival model was evaluated for human kidney cells [28] exposed to the atomic-bomb

radiation environments. Fatal malignancies^ from

1950-1972, excluding leukemia, were then normalized to the incidence rate in the 0-9 rad exposure group and plotted vs cell survival ini Fig. 6.

As shown in Fig. 7, Albert et al. [29] found

linearity between the number of skin tumors in rats and atrophic follicle incidence which probably is an estimator of tissue destruction. A single-dose-radiation-cell-survival curve for Fischer-rat mammary-gland cells [30] was used in

order to plot the percent of animals with mammary

neoplasms as a function of cytotoxicity. The re- sults are shown in Fig. 8 for Sprague-Dawley rats [31-34], Fischer rats [35], RFM mice [19], and the female atomic-bomb survivors in Nagasaki [36]. The survival of mammary cells in vivo has now been found to have a broader shoulder (Gould-personal

communication) than published data [30]. These newer data are not available, but the net effect

would only be a rotation, through the spontaneous

tumor values, of the relationships in Fig. 8. If error bars of one standard deviation are added to data in Fig. 8, then a linear fit cannot be re- jected on statistical grounds. The survivor data

is normalized to risk for the 0-9 rad group and the data above 200 rads were excluded because of possible neutron contributions. Although the RFM mouse and human mammary tumors do not appear to demonstrate a significant dose-response, none of these plots seem in conflict with linearity be- tween cancer incidence and cytotoxicity.

Radiation induced cancer is such that almost no observation seems universally true. The best methodologies for risk assessments, then, should probably be based on relationships that are com- monly reliable. The only cancer (which we observed) to conflict with linear incidence as a function of cytotoxicity was ovarian tumors in mice [19] plotted against Chinese hamster ovary

(CHO) cell-survival data [37]. Our coupling of

Fig. 6. Fatal malignancies, relative to 0-9 rad groups, vs cell survival

in Hiroshima and Nagasaki A-bomb survivors (leukemias not included).

46
<

cr

ino

UJo

z

llJ 9

o

Fig. 7.

5.0

4.0

3.0

2.0

1.0

0 .0 I— "750 RADS /1.65 mm

1700 RADS /I 65 mm •

9 2000 RADS /1.40 mm

• 4000 RADS /0.75 mm

0 2500 RADS /1.65 mm

• 7500 RADS /O 36 mm

^2000 RADS /0.75mm

5000 RADS /0.36mm

0 5 10 15 20

ATROPHIC FOLLICLE INCIDENCE

(ATROPHIC FOLLICLES / RAT X10"^)

Incidence of skin tumors vs atrophic follicle incidence in rats- reproduced' with permission of R. E. Albert and publisher.

1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 MAMMARY CELL SURVIVAL-FISCHER RATS

Fig. 8. Incidence of mammary neoplasms vs mammary cell survival in Fischer rats.

47
these two studies was also encouraged by relative

DNA contents and similar gene activation. The incidence of ovarian tumors rose rapidly for low cytotoxicity and seemed to plateau at 40-50% for

cell lethality greater than about 20%. Ovarian

and certain other tumors may be justifiable ex- ceptions to usual cell kinetics because the adult ovary is incapable of proper regeneration [38].

A Simple Model for Environmental Cancer

Hyperplastic agents, irritants, viruses,^ and cytotoxic insults have been assessed varying and confusing roles in theories describing the induction of neoplastic diseases. It has been ob- served that many effects, e.g., injury to bloodforming and GI organs, are probably caused by the death of cells [40]. Some studies have revealed

a clear parallelism between the carcinogenicity

of a substance and its cytotoxic effect.

Physicians and oncologists have frequently noted

that accelerated tissue renewal due to surgery,

toxic or hyperplastic stimuli, growth hormones,

immunological treatment, partial hepatectomy,

unilateral nephrectomy, chemotherapy of non- malignant disorders, etc., often results in an increased risk of cancer [41-43] - especially if accompanied by an initiator, e.g., a small dose

of ionizing radiation. The 1977 UNSCEAR report

[44] provides a good review of the carcinogenic

role of cell proliferation and states that,

"Cellular proliferation is implicit in the notion

of promoting action and is clearly necessary for

the tumor progression, and the evidence reviewed

is generally in accordance with these concepts.

However, so far there is no unequivocal evidence

that the induction of proliferation in the target

cells is a key event in radiation carcinogenesis

. . . . In any case, a systematic comparative

analysis of events associated with the induction

of tumours in target cells would be required be- fore any postulated promotional role of cell pro- liferation could be fully elucidated." In addition, continued insults of estrone or testosterone can control proliferation and be used to stimulate or regress specific cancers in primary

and secondary sex organs [45]. A vivid illustration is 100% mammary tumor incidence resulting

from estrone injections in male mice belonging to

a strain in which only the females normally ex- perienced a high incidence rate of mammary tumors

[45].

Biological techniques are entirely inadequate for

quantification of the role of cell proliferation

[44-p577] and the UNSCEAR Committee expresses

doubt about a direct role probably because normal tissue renewal rates sometimes seem unrelated to cancer rates [44-p617] and because induced bone fractures failed to enhance bone tumor rates in mice [47]. Also, a biochemical experiment led Raick et al . [48] and the UNSCEAR to the con- clusion that hyperplastic changes due to very weak

insult levels and promoting action at higher

levels are independent processes [44-p578].

Cancer is expected to derive from stem cells and amplification of cell populations by cells com- mitted to .differentiation obscure any conclusions

based on tissue renewal where both factors are present [46]. Unenhanced bone tumor rates could

be explained in terms of an equilibrium cell popu- lation and the migration of cells as no significant amount of cell killing should accompany

fractures [39]. The conclusions, that hyperplastic changes due to very weak insult levels

and promotion are independent processes, seem un- substantiated because the low-level hyperplasia,

of duration less than the period between the administrations of the chemical, leads to a timeintegrated cell-population that is exactly equal

to the steady-state or zero-dose time-integrated

population. Additional insults to a hyperplastic

state (or a promotion which results in necrosis)

yields a greatly elevated time-integrated cell population and an increased cancer rate.

It has been observed generally that precancerous

lesions occur at a rate that leads to a large

overestimation of the incidence of neoplastic

diseases [49]. For low LET insults, these trans- formations usually occur at a much greater rate than does cell inactivation. Most cancer models

specialize in the production of altered cells and repair processes but it seems unlikely that the

irmiune response or repair enzymes would reduce the populations of altered cells at a rate that is often linear with cytotoxicity yet relatively in- sensitive to the number of transformed cells. It has frequently been observed that a single neo- plastic cell has a very low probability of ever

becoming a cancer, yet cell -marker analysis has demonstrated that many tumors are of a singlecell origin [50,15]. Promoters such as croton

oil, TPA, etc., are often strongly inflammatory,

producing edemas and hyperplasia, but the inductions of all cancers cannot be described simply.

However, it appears thet a few conditions can be noted that will permit the development of general

protection guidelines. These include:

1. Production of altered cell popu- lations by damage events or misrepair processes [51];

2. Cellular abnormalities must be "fixed" by cell division [51,

52];

3. Stimulus to proliferate;

4. Attainment of critical mass [53] or state in an individual of a certain sensitivity [54,55], and

5. Escape from normal systemic regu- lation processes.

We essentially reoffer the practical hypothesis of

Bergonie and Tribondeau [7-p3,56-p261 ,57] in that

normal or accelerated cell turnover may permit

altered-cell populations to become tumors [58-68].

Furthermore, an insult which retards cell pro- liferation, e.g., malnutrition, or an insult which

selectively blocks (if repair processes continue

[51,72-p403] or destroys transformed cells may produce no observable excess of tumors [65-p269].

In fact, every agent, that is known to affect cancer incidence rates, has mitotic or antimitotic

behavior, without any exceptions. It is apparent

I
that some altered cells are never repaired and can

be "scored" later by insults which stimulate pro- liferation [59]. Our hypothesis is illustrated in Fig. 9. Generally it is quite difficult to obtain

quantative information on cell-proliferation

kinetics [7-pl7] and this paper demonstrates, that,

for protection guidelines, cytotoxicity should be considered as a substitute index.

Carcinogenesis is often modeled in terms of timesequential events to a single cell, e.g., probability of the production of sequential initiating transformations multiplied by the probability

of survival of the transformed cell. For protection purposes it appears that because individual

altered cells are usually expected to be present

in the human body, the important factor is the processes which permit altered cells to clone.

Thus, we subscribe to Nordling [60] because

"cancer is caused by mutations multiplied and

accumulated usually through large-scale proliferation of cells" and "the unexpected high incidence

of internal neoplasms in childhood is explained

by the high frequencies of cell division in the

fetal stage. Other high cancer incidence rates

in particular organs may be explained on the

basis of exposure of the tissues to mutagens or

to agents increasing cell proliferation" [73,38].

Several cancer vs dose-response models suggest

that the population of altered cells, decreased

by a cell-killing factor, is related to the incidence of cancer. Indeed, the resulting doseresponse curve has the correct shape, which is very gratifying. However, subcarcinogenic doses

of ionizing radiation accompanied by partial

hepatectomy [74], CCli, [59], uninephrectomy [43],

etc. result in decreased populations of altered

cells and increased cancer rates--a serious con- tradiction to the simplicity of these models.

The observation of a linear relationship between

the cumulative incidence of late-onset cancers in man and the sixth power of age has led to theories

involving kinetic mechanisms [70,60,65]. This

approach seems consistent with time-to-diagnosis

periods depending upon many doublings (>30) of tumor volume after proliferation stimuli have

adjusted to maintain homeostasis.

The lower tumor incidence generally observed for

fractionated and chronic exposures, relative to acute exposure, seems consistent with slower pro- liferation resulting from a lesser degree of cell killing. Repair enzymes should be more reliable

than when a large population of altered cells must

undergo DNA repair [51] during a strong growth

stimulus. Furthermore, according to Brown [61],

"A relationship between the number of virus

particles and the number of sterilized cells is

a possibility which would explain the higher incidence of thymic lymphomas following weekly fractionated exposure as opposed to single doses."

Another model [70] based on "the transforming

agent imagined to be passed from "dead" to viable

cells may be a virus" could possibly help explain

the behavior of ovarian tumors in mice because for high levels of cell killing, viable cells would

receive redundant stimuli. More generally, such

a process would almost certainly produce a convex- shaped response (as was observed for ovarian

Biological Stress

Death of Cells

b.

T 1 Altered Gel Is

.1 Normal Cells

Growth Demand on Progenitor Cells

Surviving Fractions Affect Proliferation Rate

d.

Proliferative Activity Fixes Trans- formations and Controls Growth

c = growth hormones

e = feedback mechanism

(chalones, contact

inhibition, etc.)

Growth Conditions Unfavorable to Neoplastic CellsGrowth Conditions Favorable to Neoplastic Cells

No rma Shortened Lifespan

Because of Increased Cellular Activity?

Fig. 9. A simplified model for environmental cancer as a linear functiorrfof cell survival.

49
tumors) which would conflict with other relationships described in this paper.

Thresholds for Risk Assessments

Curvilinear dose-response relationships present

special problems, since they suggest either that

the mutagenic response may become insignificant

at very low levels, or even that true thresholds

may exist [1]. The question of a threshold is not easily resolved and must be investigated on a "mechanistic" rather than a "dose-response" basis.

On a cellular level Chadwick and Leenhouts [11]

have explained the possible absence of a threshold

based on the difficult process of exact repair of

all nonlethal DNA double-strand breaks. Zasukhina

and Sinelushikova [71] suggest that mutagenesis

and carcinogenesis are connected with the formation of crosslinks between strands of DNA and alkylation of nitrogenous bases; however, other

work [56-p202] indicates only DNA degradation is important in the presence of oxygen because the

ends become peroxides which prevent crossl inking.

These studies and many others will have to be more clearly understood before "threshold" applications

will be accepted.

There appears to be many modes of cell death from

a variety of insults [49,75]. In addition, there

is a small but definite probability that normal cell-turnover in non-insulted or "low-dose" tissue

will "fix" or "lock" a DNA lesion, which in cer- tain individuals, may escape the immune response

[54] and be diagnosed as cancer 30-40 volume

doublings later.

From this paper it appears that incidence rates,

in a population, are straight-line, no-threshold,

relationships from spontaneous or background

cancer rates to cancer rates associated with

saturated cell-killing levels. Terzaghi and Little [51] have suggested that DNA errors may be" inserted by repair processes and fixed by prompt

cell division. If given sufficient time before

the first division, the damage structure and even

the misrepairs could be corrected. This behavior

could be confused with a threshold effect; how- ever, for a population , a threshold for ionizing

radiation generally seems inconsistent with cancer

genesis from a single cell and latency periods ex- pressed in cell lifetimes. Radiation insults are self delivered, but chemical insults depend upon metabolic processes. Thus, it has been suggested

that "very low levels of some agents will be in- activated before they reach the biochemical

element of the cell (DNA) required for the pro- duction of malignancy" [76]. Such behavior could

not be expected for all chemicals because some require metabolic conversion in order to reach a carcinogenic state. For all practical purposes,

the excess cancer incidence may approach zero for

low insult levels (as shown in Fig. 10) and threshold debates become academic.

Partial Body Irradiation and Risk to Spondylitics

The problems of partial body and nonuniform irradiation are quite easily solved for cancers which

occur linearily with proliferation estimated from

cytotoxicity. The risks of these malignancies

need not be based on the absolute number of transformed, misrepaired, or killed cells but rather on \ the surviving fraction which is the integration of

the probability of cell survival over the organ of

interest. According to UNSCEAR [77] an average of 15 30-40% of the bone marrow in the ankylosing

spondylitis patients treated by X-rays was irra- ij diated. The BEIR Committee [24] estimates 42.3%.

The marrow dose levels in the path of the beam j, were greater than 300 rads while outside the

beam, the dose fell quite rapidly.^ Thus, it appears that the surviving fraction was in the range of 0.5-0.7 of the total marrow. From

Fig. 4 and 5, the risk for a first exposure^" can

be estimated, based on the atomic-bomb survivors,

to be 0.7 X 10"^ leukemias/person-year for i chronic leukemia and (22-34) x 10"^ leukemias/

| person-year for acute leukemia. The 52 observed

cases of leukemia in the spondylitics divided by

141,796 person-years of observation yields a risk 1 of 36 X 10'^ which seems to agree better than ex- | pected because of multiple exposures to many of . the spondylitics, the fact 80-90% of the ' spondylitics were men — incidence in males is 1 higher, and many other differences in the popuj lations at risk. This agreement may indicate that patients having a "predisposition" to leukemia were sufficiently insulted by a single exposure

and more resistant patients were not subjected

to significantly greater risk by multiple treat- 1 ments. Leukemia rates in other populations show

j equally good agreement and are discussed

elsewhere [75]. !"

Discussion and Conclusions j

I Thus from our simple model for a population, the I ; dominant condition necessary to estimate the risk ,

' of a particular tumor [assuming non-zero altered- 1 cell populations (initiation)] is an index of the j

I growth stimulus (promotion). It also appears that ii I cancer incidence need not be based on the number I of initiation events per cell, the number of j: altered cells, or the number of cells at risk

[13], but instead on a mechanism that prescribes !

1 the fractional cell-replenishment requirement. i:

' For organ-specific promotion insults not con- nected with initiation insults, the neoplastically

transformed cell-population may be null or low and Ij normal cancer rates may result as illustrated by 'f hepatoma rates in rats stressed only by hepa- i tectomy or CCli* [74,59]. !

Assessments based on cell survival seem to provide

hope for summing risk to different organs, relating effects in mouse to those in man, relating

risk from radiation insults to risk from chemical

insults, describing multicentered and single-cell , origin of cancer, understanding hot-particle

j behavior, and explaining the synergistic effects

| of different insults. Risk analyses based on product-type models can explain synergism which '

has been noted for combinations of severe insults

and also the behavior of weak insults in conjunction with a strong insult where additivity cur- rently seems possible. A generalized cellsurvival model may permit the assessment of over- ; all cancer risk from all hazards.
Randi and Thomas [78] suggest that absorbed dose

is a sterile concept for operational health physics

in that it produces little or nothing. In fact,

it is becoming more apparent that dose-type parameters are insufficient to assess biological re- sponse [79] because in order to put radiations of

different quality on the same basis for the pre- diction of effects, it is necessary to establish

correction factors— physical factors (Quality

Factors) and biological factors (Relative Biological Effectiveness). Such processes seem

unscientific and will always be subject to technical abuse and criticism.

If one prefers to follow the traditional dose- response approach then, curves are shown in Fig. 10 and the ratios of photon dose to neutron

dose for specific levels of cell survival are shown in Table II. The dose-response curves are amazingly similar to acute myelogenous leukemia grown in culture, [80-if the abscissa is made

linear]. The RBE values in Table II are also

quite consistent with RBE's for late effects

[81-83] but "Turning concepts upside down can sometimes straighten them out" [65-p92].

r51
Table II. Cytotoxicity RBE for Fission-Neutron

Radiation Environments

Marrow Cells Kidney Cells

Dn V^n Dn t Survival (rads) (rads) (rads) (rads)

>99 <0.2 <10 8(P >0.3 <20 lOO**

98 0.5 20 40^ 0.5 30 60&

95 1 30 30^' 1 40 40^'

90 3 40 13 2 60 30

80 6 60 10 5 90 18

70 9 70 8 8 110 14

60 13 90 7 11 140 13

50 18 no 6 15 170 n

40 23 130 6 20 200 10

30 30 150 5 26 240 9

''The biological effectiveness of high LET radiation should not dimenish for

fractionated or chrnoic exposure; however, the effectiveness of low LET

radiations may be influenced by repair processes.

^Values of Dy/Dn above 90% survival are in the region where models begin to differ and investigators commence to argue about the effect of slope at

zero dose. These values are based on Katz's model.

Footnotes

1. Office of Environmental Policy Analysis,

presently with Evaluation Research Corp.

2. Industrial Safety and Applied Health Physics

Division

3. Robust is used in the statistical sense to denote a high degree of insensitivity to minor parametric variations.

4. Tissue-keritia-in-air or "air-dose" is con- sidered to apply to infinitesimal amounts of

biological tissue under exact analogue ex- posure conditions, except for the absence of

the specific survivors. The kerma represents

the j(inetic energy released in materials and becomes equal to the absorbed dose if, and only if, all energy released when a particle

interacts is assumed to be locally deposited,

i.e., the pathlengths of the secondary

particles are short in comparison to the dimensions of the critical biological organs.

For the Atomic-Bomb Survivors, these estimates are usually called "T65D doses."

5. S. Jablon, "Atomic Bomb Radiation Dose Estimation at ABCC," Atomic Bomb Casualty Com- mission Technical Report ABCC-TR-23-71

6. W. M. Court-Brown and R. Doll, "Leukemia and Aplastic Anemia in Patients Irradiated for

Ankylosing Spondylitis," Medical Research

Council Special Report Series No. 295, Her Majesty's Stationery Office.

7. S. Jablon and H. Kato, "Mortality Among

Atomic-Bomb Survivors, 1950-1970," Atomic

Bomb Casualty Commission Technical Report,

ABCC-TR-10-71.

8. I. M. Moriyama and H. Kdta, "Mortality Ex- perience of Atomic- Bomb Survivors 1970-72,"

Atomic Bomb Casualty Commission Technical

Report, ABCC-TR-15-73.

9. "Viral agents... are capable of eliciting the whole spectrum of proliferative lesions from

mild hyperplasia and near-neoplastic reactions

to highly malignant neoplasias of the dependent type. These agents. .. provoke neoplastic^

reactions when supplied on a continuing basis"

[39-p231].

52
10. Risk estimates to spondylitics are often

brought into line with Atomic- Bomb Survivors

and other populations by averaging the ab- sorbed dose over irradiated and non- irradiated marrow centers. Averaging helps

to account for "wasted radiation", i.e., the excess above saturated cell-killing levels. Although averaging may produce reasonable

estimates, it is dangerous and misleading

[44-P365].

11. ORNL-TR-4241 , The Problem of Hot Particles.

A report written for the conference, "Medical and Biological Problems of Radiation Exposure," Research Institute, Borstel , May 30, 1975.

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56
NBS SP 554 (1979) CLINICAL TRADEOFF: THE MAXIMUM PROBABILITY OF UNCOMPLICATED CURE

Lionel Cohen

Michael Reese Medical Center

Chicago, Illinois 60615

Any new procedure intended to yield more effective tumor control in clinical radiation therapy

carries the concomitant risk of more severe acute and remote side effects. The effective gain or trade-off can then be evaluated in two ways. One is an instantaneous optimization function

(cure-rate minus complication-rate) determining the probability of uncomplicated cure, which

can presumably be maximized by judicious choice of technique. Consideration of long-term

effects however suggests that a different optimization algorithm, integrating the total expec- tation of disease-free person-years is more realistic. It can be shown that even if higher

energy radiation leads to a small increase in the rate of local control, or in the time to recurrence in uncontrolled cases, these gains would effectively offset the risk of induced

neoplasia or similar remote complications

.

Introduction

There are many precedents in the history of

radiation therapy where the transition to a new modality of treatment has raised questions in re- gard to the merit and safety of the supposedly

more effective procedure. Any change in technique

or departure from well-established procedures in a therapeutic regime could result in a more or less

significant change in the probability of controlling the disease as well as in the risk of serious

side effects. To some extent both the advantages

and hazards can be assessed in advance, and some

estimate made of the probable gain to be expected.

Since the new modality has necessarily not been

studied exhaustively, the information reqiiired to estimate the expected gain is seldom complete

.

More often the new procedure must be explored in

the belief that the inevitable errors, difficulties and coitjjlications that will be encoxintered can be minimized with experience, and that any inproved results which can be achieved will be

maximized by the necessary attention to technical

detail. The sequence of exploration, information

feedback, error correction and optimization provides the conditions under which the clinical gain

can be made to offset the additional risks

.

As an example of the foregoing reasoning, one

recalls the occasion when radiation oncologists

were required to justify the acquisition of cobalt

units at a time when the best available modality

was 250 Kv x-ray therapy machines . The main motivation for acquiring the new equipment was better

depth dose and skin sparing. These were attractive advantages in view of the severe skin reactions which invariably accompanied radical orthovoltage radiation therapy. On the other hand, the

advantages of the high-energy beam could be discounted by many new problems . Not the least of

these was the absence of a visible skin reaction.

Previously the skin reaction had been used as a biological dosimeter, prescribed treatments being

supplemented if reactions were mild or curtailed

if they were excessively severe. In practice the radiation therapist had implicitly corrected calibration or dosimetric errors by standardizing

normal tissue reactions. Without this safeguard

the new modality required a new order of dosimetric precision if severe complications from

excessive dosage in deeper tissues were to be

avoided.

Whether improved cure-rates with cobalt

therapy would be large enough to justify the

additional risks , the major physics effort needed

for calibration and planning, and the extensive

rethinking in regard to dosage and tolerance

limits required on the part of the prescribing

radiotherapist was seriously questioned. Many years of accumulated experience by both radiation

oncologists and medical physicists were required

before the superiority of the new technique

could be proved by hard statistics.

There were to be similar difficulties in the

transition from cobalt therapy to megavoltage

linear accelerators, with further complications

as the radiotherapeutic art evolved to include

high-energy electron beams . A further escalation

in complexity both in the physics and biological

\inderstanding of the beam was entailed when use of heavier particles (nuclei, neutrons or pions) was implemented. In each case the theoretical

rationale for exploring the new modality needed

to be strong enough to justify the professional

effort and financial cost involved in developing

the new modality, and the many years of research

and experience were necessary before clear-cur

statistics on the efficacy of the new modality

could be presented.

Instantaneous Optimization Functions

The possible existence of an optimal dose, or even an optimal fractionation scheme , was recog- nized early in the history of radiation oncology.

Holthusen first noted that the steep sigmoid dose

and normal tissue damage leads to the conclusion

that the probability of uncomplicated cure after

a course of radiation therapy depends critically

upon the dosage delivered (Ho35) . If the probability of tumor control (P^) and the probability

of normal tissue injury (P,^) are stochastically

independent, then the conditional probability of

uncomplicated cure (Pc) is given by: Pc=Pt)l-Pn)-

The shape of this function (Fig. 1) is clearly

related to the relative positions of the curves for tumor control and for normal tissue complications

on the dose axis . The greater the separation of

these two curves the higher the probability of uncomplicated control at the defined optimal dose.

57
KILORAOS

Figure 1 Dose-effect relationships. Sigmoid

response curves for control of mammary carci- noma (left) and damage to normal tissues

(right-hand curve) treated on 10 fractions

over 3 weeks . The probability of uncomplicated

control (lower curve) has a clearly defined

maximum at 45 Gy.

Strandqvist (ST44) demonstrated the dependence

upon treatment time of the median doses required

to produce a standard reaction in radiation

therapy. Later DuSault (Du56) showed that different tissues and tumors have different dosetime relationships and that the separation between

the sigmoid curves of Figure 1 could vary widely

depending upon the fractionation scheme used. Treatment could be optimized in regard to both

the dosage delivered and the particular fractionation scheme chosen, and the optimal procedure

could be computed for various combinations of tumor and normal tissues irradiated. A model for

optimizing the treatment scheme was proposed by Moore and Mendelsohn (No72) from which the most

acceptable procedure could in principle be

determined, given appropriate weighting coefficients for the probability of tumor cure and the risk of complications

A consequence of this analysis is that

cautious radiotherapy (never exceeding safe dos- age limits) is not the best policy if cure-rates

are to be maximized. Aggressive radiotherapy,

in which the incidence of significant complications is at least as high as the incidence of

local recurrent, is more likely to be associated

58
with the highest probability of uncomplicated

cure (Co73) . Since tumor recurrence or failure

to cure is in reality the most serious outcome of

a course of treatment, the total incidence of

complications (which in this instance includes

recurrence) is minimized by the more aggressive

approach. This trend is becoming increasingly

recognized in modern radiation oncology. Current

practice encourages delivery of higher doses to wider fields in order to approach the theoretical

optimum.

There is, however, an inherent limitation to

this approach which should be recognized. In this discussion cure, complications and probabilities were considered as if they described instantaneous events , in which the application of a particular treatment leads to a specified immediate outcome. Under these circumstances the

probability of uncomplicated cure is a meaningful

figure of merit to be maximized. In practice,

the outcome is not instantaneous but a protracted,

time-dependent process spread over many years.

The practicing radiation oncologist recognizes

that a tumor may regress initially and appear to

be cured, and yet recur or metastasize in later

years. If the remission period were sufficiently

prolonged, such a late rec\arrence might not be

considered a failure. Similarly, the most significant side effects may be acute reactions following immediately after the course of radiation,

but more commonly we are concerned with late

effects appearing characteristically between 1 and

3 years after treatment. At this juncture we are considering an even later side effect, namely

that of induced neoplasia which may occur between

5 and 25 years after exposure. Under these cir- cumstances the instantaneous optimization function

based on the response observed at a specific

follow-up date is clearly inappropriate , and a new figure of merit is required for optimization.

This should take into account not only the probability of uncomplicated control, but a timeintegrated response, possibly the number of

disease-free person-years, generated by the particular treatment policy under scrutiny.

INTEGRAL OPTIMIZATION FUNCTIONS

An optimization algorithm based on the total

proportion of disease- free person'-years , that is the integrated sum of uncomplicated tumor control

over survival time, is a complex function of

dosage and treatment technique which can probably

only be evaluated in qualitative terms at the

present time. The variables that will need to be

considered include the probability that the tumor

is completely sterilized, the time to recurrence

if the tumor does recur, the expected incidence

of acute reactions of unacceptable severity, the

anticipated incidence of late radiation damage,

the anticipated incidence of "very late" effects

such as neoplasia, and the actuarial expectation

of life for a normal population corresponding in age and other social factors to the patient population at risk. Some of these rates can un- doubtedly be estimated on the basis of current

therapeutic experience, but others will need to

be evaluated with future experience, paying particular regard to dose-effect relationships. Although increasing dosage may well increase the risk of all side effects, including late neoplasia, higher tumor doses may well lead to improved local control in two respects. Firstly,

since ionizing radiation provides a measure of

tumor control by killing or sterilizing tumor

cells following a deterministic survival function,

the probability that the tumor is completely

ablated (surviving cell population reaches zero

or some critically small number incapable of

maintaining continued growth) is a function of

dosage. Secondly, if a small residual viable

cell population persists, the time to recurrence will depend on the number of surviving cells, and

hence the remission-time is likely to be directly

proportional to tumor dose . Consequently , both

the cure-rate and the number of disease-free

person-years (or any equivalent time-integrated

figure of merit) will increase with increasing

dosage or increasing treatment efficiency, independently of the increasing risk of complications .

The conditional probability of xancomplicated

time-integrated survival is illustrated, albeit

in a qualitative manner, in Figure 2. In this

representation line A indicates the actuarial

expectation of life of the corresponding normal

population. It is assumed that a certain pro- portion of tumors are cured (line B 60%) and the remainder will eventually recur (line C, 40%)

,

with a specified median time to recurrence

(R = 2 years) . Line S = (B + C) is the total

survival curve for the treated population. In this diagram complications are assxmed to com- prise 5% severe acute reactions, 5% severe late

effects arising between 1 and 3 years after treatment and 5% late neoplasia arising between

8 and 12 years after treatment. These complications are super-imposed upon the surviving popu- lation and will consequently resemble the com- plex functions illustrated by Dl , D2 and D3 in

the diagram. The total prevalence of radiation

induced complications in the population at any one time is then given by the summation shown by

line D. The curve for disease- free or uncomplicated survival can then be described by the stepped function shown by line E in the diagram.

It will be noted that the total effect attributable to late neoplasia with an incidence of

5% represents a very small fraction of the number

of disease-free person years as illustrated by

the shaded area (F) in Figure 2. Clearly a very

small increase, on the order of 2%, (raising the origin of line B from 60% to 62%) or a prelongation of the average time to recurrence (R in

Figure 2) from 2 to 3 years would be sufficient

to counterbalance the 5% incidence of neoplasia

occurring at a median survival time of 10 years

.

Naturally, should the real incidence of induced

neoplasia be different from the assumed 5%, the necessary improvement in cure rate or remission

time would have to be adjusted proportionally.

The general rule, however, remains that a relatively hish incidence of late neoplasia appearing

many years after treatment is readily offset by

quite a modest improvement in local control.

59
O 2 4 6 8 10 12 14

Figure 2 Integrated probability of long-term

control. A = actuarial siarvival rate; B = survival in "cured' group; C = survival of

"failed" group; D = incidence among survivors

of early and late side effects; E = resultant

symptom-free survivals; F (shaded area) = net

loss in person-years attributed to induced

late neoplasia.

CONCLUSIONS

Since the state-of-the-art in radiation oncology at the present timeis such that local cure

rates (excluding the problem of metastases) sel- dom approach 100%, methods for improving local

control by optimizing the treatment plan and

fractionation scheme will be pursued vigorously.

All such technical innovations carry risks of side efffects, either increased frequency or

severity of expected complication compared to the

risk with more conventional procedures ; or new

reactions not normally observed. In the few clinical situations where local control rates do

approach 100% (early Hodgkins ' Disease, seminoma,

and small epidermoid carcinoma) the use of a new

modality carrying a significaint risk of additional

late effects, obviously cannot be justified. In all situations where local control rates fall

60
short of this ideal, and where a modest improvement in local control can be expected by virtue

of higher or more effective local dosage being

achieved with a new modality, or better definition of the target volume with the exclusion of

dose-limiting vital organs in the irradiated

field, the expected gains are likely to offset

the additional risks. The risk of late neoplasia,

which is the major concern of neutron contamination in high-energy electron accelerators, would

therefore appear to be a trivial one relative to possible benefits , provided those cases in whom

100% local control can be achieved with low

energies are excluded from treatment by this

modality

.

REFERENCES

Co60 Cohen L., 1960, "The Statistical Prognosis

in Radiation Therapy, Radiology, 84, 741.

Co73 Cohen L. , 1973, "Cell Population Kinetics

in Radiation Therapy: Optimization of Tumor

Dosage", kCancer, 32, 236.

Du56 DuSault L. , 1956, "Time-dose Relationships",

Amer. J. Roentgenol, 75, 597.

F173 Fletcher G.H., 1973, "Clinical Dose- response Curves of Hiunan Malignant Epithelial

Tumors:, Brit. J. Radiol., 46, 1.

Ho36 Holthusen H., 1936, "Erfahrungen uber die

Vertraglichkeitsgrenze fur Ftontgenstrahlen und

deren Nutzanwendung zur Verhutung von Schaden",

Strahlentherapie , 57, 254.

Mo72 Moore D.H. and Mendelsohn, M.L., 1972,

"Optimal Treatment Levels in Cemcer Therapy",

Cancer, 30, 97.

St44 Strandqvist M. , 1944, "Studien uber die

kumulatiewe Wirkung der Rontgenstrahlen bei

Fraktionierung" , Acta Radio., Suppl . 55.

61
i i
NBS SP 554 (1979) DEPENDENCE OF RADIATION-INDUCED NEOPLASTIC TRANSFORMATION

IN VITRO UPON RADIATION QUALITY AND REPAIR

A. Han and M. M. Elkind

Division of Biological and Medical Research

Argonne National Laboratory

Argonne, IL 60439

C3H/10T1/2 mouse embryo cells grown in culture were irradiated with single or fractionated doses of fission-spectrum neutrons from the JANUS reactor or with

50-kV X-rays; survival and frequencies of neoplastic transformation were measured and the relative biological effectiveness (RBE) for each end point

was estimated. As expected, for both survival and transformation, RBE's depend

upon level of effect, being higher for small than for large doses. Changes in

transformation frequencies following X-ray dose fractionation indicate appreciable repair of both sublethal and siibtransformation damage. Results of

fractionation of the fission-spectrum neutron dose suggest repair of the firstdose s\abtransf ormation damage and no repair of sublethal damage. The observed

results for the induction of neoplastic transformation in C3H/10T1/2 cells are qualitatively similar to data for tumor induction in animals.

(Carcinogenesis; neutron; repair; survival; transformation; X-ray)

Introduction appropriately treated hosts

.

[2,3,5]

Radiobiological effects of fast neutrons are currently receiving considerable attention mainly

because of the potential usefulness of fast neutrons in tumor radiotherapy. Interest in fast neutrons also stems from the fact that they are a by-product of nuclear power and there is a need to evaluate the hazards to man. For these reasons,

fast neutrons are of a major importance in considering the influence of radiation quality in various radiobiological endpoints relevant to determination of their relative biological effectiveness (RBE)

.

Studies with experimental animals provide

ample evidence that ionizing radiations induce

tumors. This is true for both low and high LET radiations . In view of the complexity of the whole organism and studies of carcinogenesis in

the whole animal, it is useful to employ simpler

systems such as cultured mammalian cells . Such

systems can provide basic information of doseeffect relationships, proliferation kinetics of

normal and transformed cells, and can help in our understanding of tianor induction in^ vivo . Mammalian cells in culture are widely used to study

mechanisms of radiation action as they provide

reproducible data free of interactions that exist

in whole organism.

Aside from cell survival, one of the most

important responses studied is the potential of surviving cells to produce cancer. In the past

decade a number of mammalian cell systems in culture became available for evaluating the neoplastic transformation of cells iri vitro. It has been

generally established that exposure of normal

[1 2 3 4 5 ] cells in culture to ionizing radiations • < <

• results in neoplastic transformation, i.e., the

conversion of normal cells that exhibit a high

degree of growth control through contact inhibition and oriented growth patterns into cells that

lose the growth control, begin to pile up, and

grow at random . '''' ^

' In contrast to normal

cells, transformed cells induce neoplasms in

To date, two systems have been used for

studies of radiation-induced neoplastic transformation in vitro , primary cultures of Syrian hamster embryo cells '^'^ and mouse embryo derived

C3H/10T1/2, Clone 8, cells. ^ ' ^ In both systems,

cells isolated from transformed foci produced

tumors upon inoculation into appropriately treated

hosts. Despite certain differences in the results

obtained with the two systems, very useful information concerning the cellular aspects of neoplastic

transformation has been provided thus far. Dose- response curves for neoplastic transformation

following X-rays ' ^ and neutrons, have

been determined including the estimates of RBE of

fast neutrons relative to X-rays for cell killing

[4 51 and neoplastic transformation, ' changes in the frequency of neoplastic transformation following

[5 6 7 81 dose fractionation of X-rays and

neutrons [5] have been studied.

We review here comparative observations of the capacity of fission-spectrum neutrons and X- rays to induce neoplastic transformation in

C3H/10T1/2 cells following single and fractionated

exposures. The effect of dose fractionation on neoplastic transformation is examined in greater

detail to determine if repair processes are involved in repair of sublethal damage and neoplastic

transformation

.

Dose Responses for Survival

cmd Neoplastic Transformation

The survival curves of lOTl/2 cells after

fission-spectrum neutrons and X-rays (Fig. 1) have

shoulders indicating that these cells have appre- ciable capacity to accumulate sublethal damage

caused by both radiations. The neutron survival

curve has a relatively large shoulder, as may be estimated from the specific shoulder width of 1.5

(see legend of Fig. 1). Smaller shoulders, that is

almost exponential survival, are generally observed

63
for high LET radiations like fission-spectrvim neutrons . The RBE for survival in these cells

exhibits relatively small dependence on dose. The

estimates of BBE gives a nominal value of 2.9 for small exposures producing about 0.8 surviving

fraction, with progressive decrease with dose to

a nominal value of about 2.2 at 0.001 survival.

DOSE, krad

0 0.2 0.4 0.6 0.8 1.0 1.2 1.4

IO"^l I I I I I— L

Fig. 1. Survival (colony formation) of C3H/10T1/2 cells after exposure to different doses

of 50 kVp X-rays (open symbols) or fission-spectrum neutrons (closed symbols) . Various

symbols represent different experiments. Error bars, steindard errors of individual data

points , are shown where they are larger than the points . The survival curve parcimeters

shown have the following meanings : Dq is the dose required to reduce survival by 1/e

along the terminal straight line portion of the curve; D„, is the quasi-threshold dose; ^ ' n, is the extrapolation number [exp (Dg/Do) ] ; the specific shoulder width is given by ->- Dq/Dg, and F, is the average number of cells in microcolonies at the time of irradiation,

i.e., the average cellular multiplicity. Cells were incubated at 37°C for 10-14 days

following irradiation cind colonies were then counted. (From Han and Elkind, Cancer Res.

39^, 123-130, 1979)

64
2x10-3

10

-3

-5

10

3x10-5

h JANUS NEUTRONS

Do = 70 rods

I 0.2 0.4 0.6 0.8

DOSE , krad

1.0

Fig. 2. Induction of transformed cells by 50 kVp X-rays (O) and fission spectrum neutrons (#

in C3H/10T1/2 cells. Transformation is expressed on a per irradiated cell basis (i.e.,

independent of cell killing) . Error bars represent the standard errors of data pooled from

different experiments. The cultures were inciabated 6 weeks after irradiation and refed weekly

with fresh medi\an (Basal Eagle Medium + 10% fetal calf seriam) . Incubations were carried out

at 37°C in humidified mixture of air (98%) and COj (2%). (From Han and Elkind, Cancer Res. 39_, 123-130, 1979)

The incidence of neoplastic transformation

following a range of single exposures of neutrons

and X-rays is shown in Fig . 2 . The data shown are expressed on a per cell basis, that is, the transformation frequency is expressed as a number of transformed cells as a fraction of the total number of cells at risk and no correction is made for

cell killing. Presentation of the data in this way is relevant in comparing the radiosensitivity

of potentially transformed and nontrans formed

cells, as well as relating the dose response

curves to the observations on tumor induction in vivo . The two curves , although similar in shape

,

exhibit significant quantitative differences . The shapes of these curves, as first noted by Gray

,

reflect the combined effects of transformation

induction (ascending part) and the killing of target cells (descending part) . The Dq's of the

final slopes of the induction curves on Fig. 2 are

equal to those of the survival curves (see Fig. 1) indicating that in the region of large doses^ at

least, potentially transformed cells are killed

at the same rate as untransformed mouse cells. This observation is in contrast to data obtained

[21 with the Syrian hamster embryo cells which

indicate that at doses above 300 rad potentially

transformed cells are preferentially killed by

X-rays. t^^l

As noted , the induction curves in Fig . 2 are calculated on the basis of the total number of

cells at risk, that is, they do not account for cell suirvival . When this correction is applied

to induction curves from Fig. 2, the transformation frequency per surviving cell is obtained.

The dose-response relationship for the induction

of neoplastic transformation per surviving cell

65
DOSE, krad i

i Fig. 3. Transformation frequency of C3H/10T1/2 cells induced by 50 kVp X-rays (O) and

fission-spectriam neutrons ( 9 ) . RBE stands for relative biological effectiveness and at ' the frequency levels shown, was determined from the ratio of the X-ray to neutron doses. i Error bars represent the standard errors of data pooled fron different experiments (2-5 i per point). Hie experimental conditions were as described in Fig. 2. (From Han and

Elkind, Cancer Res. 39_, 123-130, 1979)

is obtained. The dose-response relationship for the induction of neoplastic transformation per

surviving cell by fission-spectrum neutrons and

X-rays is shown in Fig. 3. As can be seen, both

curves appear to be initially exponential, with

the neutron induction curve having a steeper

slope . The maximum frequency induced by neutrons

of about 6 X 10"^ transformants per survivor is reached at a dose of about 300 rads. The X-ray

induction curve rises up to about 600 rads where

a plateau at a frequency of about 3.5 x 10~^ transformants per survivor is reached. The neutron induction curve shows the transformation

frequencies for single doses ranging from 21-568

66

1

rads, and X-ray curve for a range of doses from 80- 1238 rads. As indicated on Fig. 3, the PBE for neoplastic transformation is about 10 in the region

of small doses and decreases to about 2.6 as the

dose increases. It would seem, therefore, that

the RBE of neutrons relative to X-rays, for induction of neoplastic transformation, is somewhat

different from that for cell survival. In the region of the plateaus, RBE cannot be estimated

because the requirement of "equal level of effect"

^. „. , [13] IS not satisfied.

/
Fig. 4. C3ianges in survival and transformation frequency per survivor in C3H/10T1/2 cells

exposed to fractionated doses of 50 kVp X-rays. Cell survival (O); transformation (•).

Error bars represent standard errors of the pooled data from different experiments. The single-dose induction curve (dashed line) comes fran Fig. 3. Inset abscissa, time at 37 °C between the two exposures; the 0 fractionation time is plotted from the total dose of 700 rads. (From Han and Elkind, Cancer Res. 39_, 123-130, 1979)

Effect of Dose Fractionation on Survival

and Neoplastic Transformation

Thus far we have shown a greater ability of neutrons to induce neoplastic transformation and

variation in incidence of neoplastic transformation as a function of a single dose of neutron or

X-rays. However, in view of the possible modes of neutron applications , the likelihood of exposure

to a relatively large single dose of radiation is rather small. As far as the radiation hazard is concerned, it is likely that therapeutic exposure

Will almost certainly be in several fractions

rather than in one single exposure . It is also

likely that our environmental expos>are will be

essentially a fractionation or protraction experience. In this context, damage registration

cannot be dissociated from the possible effect of

repair on damage expression. During the frac- tionation interval, various repair processes may

act upon the induced damage and as the ultimate

result we observe increased survival. Thus, it is important to establish what, if any, is the effect

of repair of sublethal damage or repair of subtransformation damage during the dose fractionation upon the incidence of neoplastic transformation.

To facilitate comparison of our observations

with results from animal experiments , we again

analyze our data expressed both as the trcins formation frequency on the per cell at risk basis, and

the number of transformants per surviving cell. The effect of dose fractionation on survival and

67
10'

10-2

10-

JANUS NEUTRONS

189 rads

SINGLE

, DOSES

// L /

/

/

/

/ I I _i_L

10'

i

SURVIVAL

189 rods + 189 rods

-TRANSFORMATION

J I I I I L J I I l_

20

I 378 rods

I I

HOURS, 37°C

J 1 L

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 DOSE, krad

1.0

Fig. 5. Changes in survival and transformation frequency per survivor in C3H/10T1/2 cells

exposed to fractionated doses of fission-spectrum neutrons. Cell survival (O); transformation (•). Error bars represent standard errors of pooled data from different e..^^_ 'tients. The single-dose curve (dashed line) comes from Fig. 3. Inset abscissa, time at 37 °C between

the two exposures; the 0 fractionation time is plotted from the total dose of 700 rads.

(From Han and Elkind, Cancer Res. 39, 123-130, 1979)

transformation frequency following X-rays and neutrons is shown in Figs. 4 and 5, respectively. We first note that the effects of dose fractionation

on survival shown in these two figures are qualitatively similar to data obtained with other cell

lines in vitro ^^'^''^^ and clonogenic cells in

vivo.''"''^^ As expected, fractionation of the X-ray

dose results in prompt increase in cell survival

due to repair of sublethal damage, whereas for neutrons there is no increase in survival with

fractionation. '^^ It is of interest to note that

in C3H/10T1/2 cells large capacity of these cells

to accumulate sublethal neutron damage (note the specific shoulder width on neutron survival curve-

-Fig. 1) is not accompanied by the repair of sub- lethal damage. The number of trans formants per

surviving cell decreases significantly with time

between X-ray exposures (Fig. 4), but only

slightly with time between the two neutron ex- posures (Fig. 5). The number of transformants per

survivor after X-irradiation declines steadily

with increase in time between fractions up to about 12 hr giving about 5-fold reduction in

transformation frequency. In comparison, fractionation of neutron dose results in about 1.7- fold reduction in transformation per survivor, at

most. Since the fractionation of X-ray dose re- sults in substantial increase in net survival, and

that of neutron dose in essentially no change in

net survival, it is probable that differences in

cell survival are an important contributing factor,

but not necessarily the only factor, in evaluating

the effect of X-ray and neutron dose fractionation

on the incidence of neoplastic transformation.

Support for this view also comes from data in Fig.

6 where the transformation resxilts after fractionated neutron and X-ray irradiation are expressed

68
Fig. 6. Changes in transformation frequencies for C3H/10T1/2 cells exposed to fractionated

doses of X-rays (O) or fission-spectrum neutrons (•) when expressed on a per irradiated

cell basis. Error bars represent standard errors for data pooled from different experiments.

Cells were incubated at 37 °C between exposiires. (From Han and Elkind, Cancer Res. 39, 123-130,

1979) as transformants per exposed cell, and the shape

the fractionation curve is not influenced by the repair of sublethal damage in surviving cells.

This plot suggests a 1.5 to 2-fold increase in

transformation frequency of X-irradiated cells

during the first 9 hr and comparable decrease of neutron irradiated cells up to 24 hr. The shape neutron curve is essentially the same as in Fig.

while the shape of X-ray curve differs appreciably

from the transformation curve in Fig . 4 . From the

data in Fig. 6, one might infer that for small

doses of X-rays that produce little or no cell

killing , dose fractionation with an interval up to about 9 hr between the two fractions might result

in a higher incidence of neoplastic transformation

than the same total dose in a single fraction.

Such inference would be in agreement with observa-

.[6]

of

of

5,

tions in Syrian hamster embryo

.[7,8] cells that fractionation of

and lOTl/2

X-ray doses with

little or essentially no killing, when compared to

a single dose, resulted in about 2-fold increase

in transformation frequency with 2 equal dose

fractions separated by a 5 hr interval. However, our X-ray data suggest that by 16 hr no enhancement

is observed.

Relevance of in Vitro Observations

to Animal Experiments

• The results with cultured C3H/10T1/2 cells on radiation-induced neoplastic transformation are relevant to tumor induction in the whole organism.

The incidence of induced tvnnors depends primarily

upon the number of cells at risk. Since the fraction of cells that survives a particular treatment

is generally not known , our data may be applied

,

as already pointed out, to tumor induction by con- sidering transformation frequencies per cells at risk rather than per surviving cell. Therefore,

69
16

Io

UJo

trUJ

Q.

Iz

<

IJANUS neutrons

50 kVp X-rays

15

0.2 0.4 0.6 0.8

DOSE, krad

1.0 1.2 1.4

Fig. 7. Changes in transformation frequency of C3H/10T1/2 cells exposed to single ( ) or fractionated C 1 doses of X-rays (O) or fission-spectnam neutrons (•). Data from Figs. 2 and 5 are replotted on linear coordinates to simulate the way in which tumor induction data from

animal experiments are usually plotted. Inset abscissas, ti.:.. ?.t 37 °C between fractions. From

Han and Elkind, Cancer Res. 39, 123-130, 1979)

the results from Figs. 2 and 6 are replotted on linear coordinates and shown in Fig. 7 for compar- ison with the data for tumor induction in animals.

Thus plotted induction curves are similar in shape

^ ^v, ^ ^ • J [cf 17,18] , to the curves for tumor induction. As noted earlier in this paper, the shape of the tumor induction cxirve reflects the combined effects of the induction and killing process in target

cells. ^^^^ It is clear that because of the displacement of the curve for induction of neoplastic

transformation by neutrons toward the smaller

doses, compared to the. X-ray curve (see Fig. 7),

the RBE of neutrons is consistently greater than 1

Furthermore, the differences in the shapes of the • initial portions of the two induction curves imply!

that RBE in the region of small doses decreases f with increasing dose. This feature is generally ' [19] ^ observed in animal studies

Experimental data with animals indicate that

tumor incidences following neutron radiation with

relatively low doses are not affected by dose

[17,20] fractionation and decreased dose rate. Since survival following neutron dose fractionation

is not modified to any significant extent, the

minor reduction in the frequency of neoplastic

70
transformation following fractionation of neutron

dose is consistent with only minor contributions

from repair processes on neoplastic transformation.

In contrast to the neutron data , animal ex- periments generally indicate reduced tumor inci-

[17,18,-

dence following X- 21,22] ray dose fractionation

[23] or irradiation at reduced dose rates

.

Mammalian cells in culture and clonogenic cells

exhibit increased survival following X-ray dose

fractionation or irradiation at decreased dose

[24 25] rates. ' This being the case, the questions

arise, what does the increase in transformation in

Fig. 7 represent, and is the increase consistent

with the notion of the repair of transformation

damage? In arriving at an adequate explanation,

three possibilities have to be considered

.

First we consider the possibility that sublethal damage is fully repaired, and subtransformation damage is not repaired at all. If the

sublethal damage in transformed and nontrans formed

cells is the same, the nimiber of transformants per

cell should rise with time by the same factor as does survival. We would expect, therefore, the

transformation frequency to increase to essentially the level at the maximum on the single dose

X-ray curve (i.e., from 1.7 to 10.2 x 10~Vcell) . Such a rise obviously did not occur.

The second possibility is that no repair of

sublethal damage takes place, but subtrans formation damage is fully repaired. This consideration

first of all requires that the transformation due

to the first dose of 350 rad is progressively lost

with time. Consequently, when the transforming

influence of the first dose completely disappears,

the X-ray induction curve from Fig. 2 would be

displaced to the right by 350 rads and downward

for the survivals (Fig. 1) corresponding to 350 rads plus the second dose. When the repair is completed and at a total dose of 700 rads, the net

treuis formation frequency should be about 10-fold

below that for a single 350 rad dose because of

the additional cell killing at 700 rad as compared

to 350 rad (i.e., 1.1 x 10~'*/cell) . Since transformation frequency per cell after fractionated

exposure is greater in the interval up to about 12

hr, not less, than that due to the total dose, it

is unlikely that there is no sublethal damage re- pair with complete repair of siibtrans formation

damage

.

Finally, the third possibility assumes repair

of both sublethal and subtransformation damage.

In order to predict the level of transformation

for full repair of the first dose subtransformation damage concomitant with the repair of siib- lethal damage according to the data in Fig. 4, the survival for 700 rads in the preceding paragraph should be increased by the survival due to

fractionation. As a consequence, the expected

transformation frequency would increase sixfold

if repair of subtransformation damage is completed

by the time fractionation survival reaches maximum

(expected frequency 6 x 1.1 x 10~'*/cell) . Since

the fractionation results in Fig. 7 lie between

the predicted extremes, and are reasonably close

to the expected level considered in this third

possibility, we conclude that at least some repair

of subtransformation damage occurs together with

repair of sublethal damage . The repair of sublethal and subtransformation damage does not have

to be mediated by the same repair process and

could have, therefore, different time dependencies.

In preceding considerations we ignored the

cell cycle dependent changes in survival and in

transformation frequency. It is well established

that response of mammalian cells to radiation, as determined by various end points , depends upon their position in their growth cycle at the time

of radiation. '•^^ The cell cycle dependent

changes in the frequency of neoplastic transformation by chemicals have been reported. ^'^^ It is reasonable to assume, therefore, that there

is a growth cycle specific dependence for radiation-induced neoplastic transformation. The net effect of repair of transformation damage and cell

cycle dependent variations in transformation, different from those for survival, could result in a somewhat lower level of response following dose

fractionation than is predicted above

.

The X-ray induction curve in Fig. 7 is qualitatively similar to data for rat skin tumor inducf 18

1

tion, and for induction of leukemia and hepa-

[17 22] tic tumors in mice . ' Furthermore , our interpretation that both repair of sublethal damage

and the repair of transformation damage are re- sponsible for observed changes after low LET dose

fractionation is in agreement with the interpretar 2^g 1 tion of data for rat skin tumor induction.

Conclusions

The main points that can be extracted from

presented data on radiation induced neoplastic

transformation can be summarized as follows. (1) Fission-spectrum neutrons are more effective than

low LET radiation in inducing neoplastic transformation. In the region of small doses (about 20 rads) RBE for transformation is about 10. (2) Fractionation of the fission-spectrum neutron dose

results in a small decrease in transformation fre- quency, suggesting a repair of first dose subtransformation damage and no repair of sublethal

damage. The change is rather small and further

data are needed to support this interpretation with

confidence. (3) The X-ray data suggest that subtransformation damage due to the first dose is repaired to a great extent. This conclusion rests,

in part, on the assumption that survival properties

of transformed and nontransformed cells following exposure to single and fractionated doses are similar. There is no indication, thus far, that they

are different. Finally, (4) the fact that trans- formation frequency per cell due to the fractionation of the X-ray dose lies below the expected

level predicted by the full sublethal and subtransformation damage repair implies the involvement of other factors , perhaps cell cycle dependent variations in the frequency of neoplastic transformation.

The results discussed here, as well as those

of others, clearly show that iri vitro study of

neoplastic transformation is a valuable approach

in studies of basic mechanisms and concepts of

71
carcinogenesis and that in vitro system is amenble to experiments and approaches that are almost

impossible to explore effectively in vivo.

Acknowledgments

This work was supported by the United States Department of Energy.

[14]

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'sorek, C, and E. Hall. Transformation of

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X-rays. Nature 243 , 450-453, 1973. Terzaghi, M. , and J. B. Little. X-radiationinduced transformation in C3H mouse embryoderived cell line. Cancer Res. 36 , 1367-1374,

1976

Borek, C, E. J. Hall, and H. H. Rossi.

Malignant transformation in cultured hamster

embryo cells produced by X-rays, 430-keV monoenergetic neutrons and heavy ions . Cancer Res

38 , 2997-3005, 1978.

Han, A., and M. M. Elkind. Transformation of mouse C3H/10T1/2 cells by single and fractionated doses of X-rays and fission-spectrum

neutrons. Cancer Res. 39^, 123-130, 1979. Borek, C, and E. Hall. Effect of split doses

of X-rays on neoplastic transformation of

single cells. Nature 252 , 499-501, 1974. Terzaghi, M., and J. B. Little. Oncogenic

transformation after split-dose X-irradiation.

Int. J. Radiat. Biol. 2£, 583-587, 1976. Miller, R. , and E. J. Hall. Effect of X-ray

dose fractionation on the induction of oncogenic transformation in vitro in C3H/10T1/2

mouse embryo cells. Nature 72_, 58-60, 1978. Alper, T. , J. F. Fowler, R. L. Morgan, D. D. Vonberg, F. Ellis, and R. Oliver. The

characterization of type "C" survival cvirve. Brit. J. Radiol. 35^, 722, 1962. Elkind, M. M. Damage and repair processes

relative to neutron (and chcirged particle)

irradiation. Curr. Topics in Rad. Res., Vol. 7, (M. Ebert and A. Howard, eds.), pp.

1-44, N-44, North Holland Publ. Co.,

Amsterdam, 1970. Gray, L. H. Radiation Biology and Cancer. In^ Cellular Radiation Biology, A Collection of

the Papers Presented at the 18th Annual Symposium on Fundamental Cancer Research at the

University of Texas M. D. Anderson Hospital

and Tumor Institute, pp. 7-20, The Williams & Wilkins Co .,, Baltimore , MD, 1965. Borek, C. In vitro cell transformation by

low doses of X-irradiation and neutrons . In Biology of Radiation Carcinogenesis (J. M. Yuhas, R. W. Tennant, and J. D. Regan, eds.). Raven Press, New York, 1976.

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Phys. 9, 357-386, 1963

Elkind, M. M. , cind H. A. Sutton. Radiation

responses of mammalian cells grown in culture.

I. Repair of X-ray damage in surviving Chinese

hamster cells. Radiat. Res. 13^, 556-593, 1960. Withers, H. R. , and M. M. Elkind. Radiosensitivity and fractionation response of crypt

cells of mouse jejunum. Radiat. Res. 3£, 598- 613, 1969. Ngo, F. Q. H., H. Utsumi, A. Han, and M. M. Elkind. Sublethal damage repair: Is it

independent of radiation quality? Int. J. Radiat. Biol., 1979, in press.

Upton, A. C. , M. L. Randolph, and J. W. Conklin. Late effects of fast neutrons and gammarays in mice as influenced by the dose rate of irradiation: Induction of neoplasia. Radiat.

Res. 41, 467-491, 1970. Burns, F. J., eind M. Vanderlacm. Split-dose recovery for radiation induced tumors in rat

skin. Int. J. Radiat. Biol. 32^, 135-145, 1977. Rossi, H. H. , and A. M. Kellerer. Radiation

carcinogenesis at low doses. Science 175 , 200-202, 1972. Ullrich, R. L. , M. C. Jernigan, G. E. Cosgrove,

L. C. Satterfield, N. D. Bowles, and J. B. Storer. The influence of dose and dose-rate

on the incidence of the neoplastic disease in RFM mice after neutron irradiation. Radiat.

Res. 68, 115-131, 1976. Burns, F. J., R. E. Albert, P. I. Sinclair,

and M. Vanderlaan. The effect of a 24-hour

fractionation interval on the induction of rat

skin tumors by electron radiation. Radiat.

Res. 62, 478-487, 1975. Metalli, P., G. Silini, S. Castillo, and V. Covelli. Induction of tumors and leukemia by

split doses of X-rays . In Radiation Induced

Cancer, IAEA-SM-118/14 (Intiernationational

Atomic ESiergy Agency), p. 227, 1969. Shellabarger , C. J., V. P. Bond, G. E. i^onte,

and E. P. Cronkite. Results of fractionation

and protraction of total-body radiation on rat mammary neoplasia. Cancer Res. 26^, 509-513,

1966. Fox, M. , and A. H. W. Nias . The influence of recovery from sublethal damage in the response

of cells to protracted irradiation at low

dose-rate. Curr. Topics in Radiat. Res. I (M. Ebert and A. Howard, eds.) Vol. 7, pp. 71-103, North Holland Publ. Co., Amsterdam,

1970. Withers, H. R. , G. D. Oliver, and D. W. Glenn.

Response of mouse jejunal crypt cells to low

dose rate irradiation with Californium I neutrons or radium gamma rays. Radiology 103 , 741, 1972. Sinclair, W. K. Cyclic X-ray responses in

mammalian cells in vitro. Radiat. Res. 33 , , 620-643, 1968. Sinclair, W. K. Dependence of radiosensitivity upon cell age. In Time eind Dose Relationship in Radiation Biology as implied

to Radiotherapy, pp. 97-116, BNL-50203

(C57), 1970.

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^ Bertram, J. S., and C. Heidelberger . Cell

cycle dependency of oncogenic transformation

induced by N-methyl-N ' -nitrosoguanidine in

culture. Cancer Res. 3£, 526-537, 1974.

[291 ' 'Marguardt, H. Cell cycle dependence of

chemically induced malignant trsuis formation

in vitro. Cancer Res. 3£, 1612-1615, 1974.

^^^^ Jones, P. A., M. S. Baker, J. S. Bertram, and

W. F. Benedict. Cell cycle-specific oncogenic

transformation of C3H/10T1/2 Clone 8 mouse embryo cells by 1-g-D-arabinofuranosylcytosine.

Cancer Res. 37, 2214-2217, 1977.

73

j NBS SP 554 (1979) NEUTRON SOURCES AND THEIR CHARACTERISTICS*

R. C. McCall and W. P. Swanson

Stanford Linear Accelerator Center Stanford University, Stanford, California 94305

The significant sources of photoneutrons within a linear-accelerator treatment

head are identified and absolute estimates of neutron production per treatment dose are given for typical components. It is found that the high-Z materials

within the treatment head do not significantly alter the neutron fluence but do

substantially reduce the average energy of the transmitted spectrum. Reflection

of neutrons from the concrete treatment room contribute to the neutron fluence,

but not substantially to the patient integral dose, because of a further reduction in average energy. The ratio of maximum fluence to the treatment dose at the same distance is given as a function of electron energy. This ratio rises

with energy to an almost constant value of 2.1x10^ neutrons cm~2 rad~l at electron energies above about 25 MeV. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. Reasons for apparent deviations are suggested. Absolute depth-dose and depthdose-equivalent distributions for realistic neutron spectra that occur at therapy installations are calculated, and a rapid falloff with depth is found. The ratio of neutron integral absorbed dose to leakage photon absorbed dose is estimated to be 0.04 and 0.2 for 14 and 25 MeV incident electron energy, respectively. Possible reasons are given for lesser neutron production from

betatrons than from linear accelerators. Possible ways in which neutron pro- duction can be reduced are discussed.

Introduction

I The radiation field around a medical therapy

i electron accelerator is a complicated mixture of photons leaking from the head, scattered photons

from the patient, beam stopper and room walls, and

photons from electrons stopped elsewhere than the

target. If the accelerator energy is high enough

to produce neutrons, i.e., greater than about 10

MeV, there is also a neutron component. This paper

is primarily concerned with the generation and

propagation of this neutron field.

Neutron Production

The primary production of neutrons is through

iy,n) reactions with smaller quantities produced

by (Y,pn) and (Y,2n) if the energy is high enough.

The direct production of neutrons by electrons,

(e,n), is smaller by about two orders of magnitude — the fine structure constant modified by several

other factors — and can be neglected. Medical

electron accelerators at present are limited to

j energies less than 45 MeV and our discussion will

be limited to that energy range. In this energy

domain, neutron production is due to the "giant

photonuclear resonance" , more commonly called the

"giant resonance". The cross section for the

giant resonance is characterized by a threshold

energy, a rapid rise to a prominent peak and a more gradual decrease at higher energies. For the medium and heavy nuclei (A > 40) which are of interest to us, the peak occurs at 13-18 MeV.

' Threshold energies for some materials of interest

I are listed in Table I. A recent compilation of

' threshold energies has been published by

I Howerton. 1^^^ A recent and comprehensive set of

I photoneutron cross sections is by Herman.'-^-'

Work supported by the Department of Energy

under contract number EY-76-C-03-0515.

Table I. Thresholds of Photoneutron Reactions.

Element

Atomic

Number

Abundance

(Percent)

(y.n) Threshold

Energy (MeV)

Pb 206 25.1 8.08

207 21.7 6.74

208 52.3 7.37

Fe 54 5.8 13.4

56 91.7 11.2

W 182 26.4 8.05

183 14.4 6.19

184 30.6 7.41

186 28.4 5.75

The yield of photoneutrons is proportional to the convolution of the (Y,n) cross section and the bremsstrahlung spectrum, which decreases rapidly

with photon energy. The result is a yield curve which increases rapidly with primary electron

energies for constant electron current up to ap- proximately 25 MeV and more slowly thereafter. For constant electron beam power, the neutron yield

is almost constant with primary electron energy above 35 MeV. Swanson*^ has recently published

electron yield calculations applicable to this energy range.

Neutron spectra in the giant resonance contain two components — the evaporation spectrum

and the direct emission spectrum. The evaporation

spectrum is the larger component and can usually

be described adequately by a Maxwelllan distribution

dN ^n / "^n \ di; = :^ "''p \-rl

75
where T is the nuclear "temperature" in MeV for the particular nucleus and is also a function of the excitation energy. One should note that the spectrum peaks at Ej^=T(most probable energy) and has an average energy of Ejj=2T. The evaporation neutrons are emitted almost isotropically

The direct emission neutrons tend to be higher

energy than the evaporation neutrons and may be

emitted nonisotropically . Mutchlerl^^^ found that

for medium to high atomic number materials and

energies near the resonance peak, direct emission

amounted to about 14% of the total neutrons. Figure 1

I- 9 J shows typical photoneutron spectra

with the direct emission neutrons responsible for

the "bump" on the high-energy side of the spectrum. For comparison a fission neutron spectrum is also

shown in Fig . 1

NEUTRON ENERGY (MeV)

7-77 3229A1

Fig. 1. Photoneutron spectra for Ta with peak bremsstrahlung energies of 20 and 30 MeV, compared to a fission

spectrum. From NBS-97 . f 9]

It will be shown later that the spectrum of photoneutrons can be degraded rapidly in heavy

metals so published spectra must be considered as representative of a particular target thickness.

Transport of Neutrons in the Treatment Head

The typical medical accelerator has massive

shielding around the target to provide photon

shielding and produce a collimated beam of x-rays. Neutrons which are produced inside the head are produced approximately isotropically and penetrate

the head shielding in all directions. The photon

shielding is usually of some heavy metal such as tungsten or lead, and there is also a certain

amount of iron and copper from bending magnets in

the head. These materials provide some photon

shielding action since these are all heavy elements. The only significant neutron energy loss

mechanisms in these heavy elements are inelastic

scattering or (n,2n) reactions. Both of these processes are effective in the MeV energy region

but the (n,2n) reaction is most effective at the higher energies. Inelastic scattering can occur only at energies above the lowest excited state of the shielding material. These lowest excited states

are in the neighborhood of 0.6 to 0.8 MeV for lead and iron but about 0.1 MeV for tungsten;

therefore tungsten is more effective in reducing

the energy of neutrons by inelastic scattering.

The energy loss in any inelastic collision cannot

be exactly determined or predicted, but there is

a minimum energy loss equal to the energy of the lowest excited state. Often there is large energy

loss in a single collision resulting in excitation

of higher energy states followed by a cascade of gammas. In the (ji,2n) reaction the minimum energy

loss is equal to the binding energy of a neutron,

and since the energies of two emerging neutrons

tend to be similar, they produce large numbers of quite low energy neutrons. The total of the inelastic plus (n,2n) cross sections is of the order

of 1 or 2 barns for these materials. This means

that the typical neutron penetrating the photon

shielding undergoes several collisions. In addition, a large amount of elastic scattering takes place in these materials at these energies. The

elastic scattering results in negligible energy

loss but does, however, have the effect of lengthening the path length for the neutrons in the shielding material and offering greater opportunity

for the inelastic and Cn,2n) reactions to occur. The attenuation of neutron fluence is small, however, since the capture cross sections of these

materials are small down to thermal energies.

With a spectrum containing high-energy neutrons,

such as a fission or PuBe spectrum, there can, in

fact, be a slight buildup of neutron fluence due

to the Cn,2n) reactions.

In a previous paperl^^] hereafter referred to

as MJS, the Monte Carlo computer program MORSE^^^-^ was used to explore the effects of the head

shielding on several photoneutron spectra. It was shown that while there is very little attenuation

of neutrons by lead and iron, there is some atten- uation by tungsten amounting to about 15% in the typical therapy machine shielding. It was also

shown that the dose equivalent is attenuated by

the shielding because of the reduced energy of the spectrum. In the course of the calculations

in MJS, it was found that the average energy of a neutron spectrum is a very useful parameter for studies around radiotherapy machines. Figure 2, taken from MJS, shows the f luence-to-dose-equivalent conversion factor for a neutron spectrum

with average energy E, plotted against E. It is seen that these data can be fit with a straight

line which is almost parallel to the curve

(labeled ICRP-21 in Fig. 2) drawn through the points listed in ICRP-21 J for the conversion of monoenergetic neutron fluences to dose equivalent.

It should be kept in mind that these conversion I factors from ICRP-21 are calculated at the depths

where the dose equivalent is maximum. For a spectrum, each energy component has it's maximum . I does-equivalent at a different depth, so a simple

addition of dose-equivalent components will re- ' suit in an overestimate. The estimates of Fig. 2

| are slight overestimates for this reason. How- ( ever, it will be shown that the average neutron

j energies around medical accelerators are so low

that this is probably not significant. In Figs.

3 and 4 are shown the average energy of several

76
1 1 1—I I M I T I I

I I

ICRP 21

I I I I I I I I

Fig. 2. Factor for con- verting dose-equivalent to neutron fluence. The

curve is based on recom- mendations of ICRP-2lf^5]

and is a function of monoenergetic neutron energy. The x's are the result of averaging the ICRP-21 con- version factor over spectra

from a large variety of sources, and are presented

as a function of average neutron energy. The

dashed line is a least squares fit which is nearly parallel to the ICRP monoenergetic values.

0.01 E or E (MeV)

spectra calculated after passing through various

thicknesses of tungsten and lead. -I It can be

seen that tungsten reduces the average energy

faster than lead, and also that this reduction

".ontinues put to greater depths. After a certain

TUNGSTEN

1.0 -

0.1

X 15 MeVWPN

o 252cf ^ Pu Be

thickness, of any material, virtually all of the neutrons will have their energy reduced below the lowest excited state, and inelastic scattering can no longer occur. This is well below the (n,2n)

threshold. Thereafter, the neutrons penetrate for very large distances through these heavy metals

with no further attenuation or decrease in dose

equivalent.

LEAD

1.0

10 20

SHIELD THICKNESS

30

(cm)

0.1

X 15 MeVWPN

o252cf

^ Pu Be

Fig. 3. Calculated average neutron energy as

a function of W thickness for 15 MeV (incident

electron energy) photoneutrons, and 252cf and PuBe neutron spectra. Over the range of W

thickness typically used, there is a substantial decrease in average neutron energy in penetrating the head shielding. Data are

from the Monte-Carlo program MORSE.

10 20

SHIELD THICKNESS

30

(cm)

Fig. A. As for Fig. 3 but for Pb . The re- duction in average neutron energy is some- what less for Pb than for W, even when the

two materials are compared on an arealdensity basis.

77
It is frequently stated that a fission spectrum is very similar to the photoneutron spectrum

found in machines in this energy range. This is true for the primary spectra, but not for the spectra after they have penetrated the head

shielding. In Fig. 5, we show the integral photo- neutron spectra for 15 MeV electrons on tungsten

and for ^^2^^ fission neutrons; these are indeed

seen to be quite similar.

100 £ 90

>- 80

70

60

50

40 O

2 30 o

<

1 1 1 1 1 T T

\ V -\ 15 MeVWPN-4"W \\ -

\ \ 15 MeVWPN Bore 1 - \ \ \ 1 \ \ \ ^ ^ \ \ \

\

\

\

\

\

15 MeVWPN-4"W \\

In Concrete Room

1 1 1

\

\

\

\ \ \ \ \ \ \ \ \ \ \ \ \ ^^1 1 , l>l

0.1 E (MeV)

Fig. 5. Comparison of various neutron

integral spectra to illustrate spectral

modification by W shielding and a con- crete room. The median energy can be

reduced from 1.5 to about 0.2 MeV by

the combined effects of the W and concrete. The bare ^^^Cf spectrum is shown for comparison. Data are from

the Monte-Carlo program MORSE.

On the same figure, we show the spectrum from 15 MeV electrons on tungsten after the neutrons have

penetrated 10 cm of tungsten. It is clear that there is a large difference between these spectra.

In particular, one can see that if one measures

the neutrons from a 15 MeV medical accelerator,

using a threshold detector calibrated with a bare

^^^Cf source, the results will be considerably

off. Also shown on this figure is the further

degradation due to the concrete room in which

medical accelerators are commonly placed. This

will be discussed in a later section. In MJS a method was described of measuring accelerator

leakage by measuring neutron fluence at various

points in the room and using MORSE to simulate

the room and the accelerator and to provide fac- tors for converting the fluence measurements to dose equivalent or absorbed dose. Since this

method requires the use of a computer and the availability of the MORSE program, MJS also

developed a "cookbook" method using graphs and tables to provide the same conversion factors

for fluence measurements.

Neutron Transport in a Concrete Room

Nearly every medical accelerator is placed in

a concrete treatment room. In such a room, neutrons from the accelerator scatter in the concrete

and may be absorbed or scattered back into the

room. These neutrons add to the neutron field

coming directly from the accelerator head. MJS

investigated this scattered neutron field in order

to evaluate their fluence measurements. Some time

ago, Patterson and Wallace^ ^^-l discovered that if

a fast neutron source is placed in a concrete room,

a thermal fluence is produced which is approximately uniform over the entire room and is related to the fast neutron source strength by

where

(t> = k(Q/S)

Thermal neutron fluence, Q is the fast neutron source strength, S is the inside surface area of the room, and k is a dimensionless

constant.

Since the thermal neutrons represent a single point

of the wall-scattered neutron spectrum, it was

suspected that a similar relationship might hold

for the entire wall-scattered spectrum. By a series of simulations using MORSE, this was found

to be true; exactly the same relationship was

found as for thermals but with a different con- stant coefficient. Measurements in MJS indicated

that the scattered neutrons were indeed constant throughout the room. These wall-scattered neutrons are of quite low energy; this can be inferred

from the lowest curve on the preceding figure

which shows the effect of the walls of a typical

therapy-sized room on the neutron spectrum. MJS

found that the average energy of the scattered neutron spectrum was proportional to the average energy of the direct spectrum as shown in Fig.

5_[6T pj.jg fluence of the scattered spectrum is given in the following equation:

(f. = kj(Q/S)

From MORSE we found that kj^ is equal to 4.6 for

the tungsten-shielded machines and 5.4 for leadshielded accelerators. The values of k]^ are different since there is no attenuation of the fluence in practical thicknesses of lead, and a transmission of about 0.85 in typical thicknesses

of tungsten. This relationship enables one to correct calculated or measured values of the direct fluence for a particular room. The wallscattered component often makes a significant

contribution (20-30%) to the measured neutron

fluence. However its contribution to the patient's integral dose is a much smaller fraction

of the total, because of the softening of the scattered spectrum discussed above (Fig. 6).

It should be noted that the original paper

by Patterson and Wallace enables one to calculate the fast neutron source strength in a room

of normal concrete simply by measuring the ther- mal neutron fluence at any point in that room. This method works quite well. The variation in

the thermal neutron capture in normal concretes

78
TARGET

33 % {for I R.L.)

0.8

j Fig. 6. Comparison of the average energy of

I source spectra and neutrons scattered from a

I concrete room within which the source is placed. The straight line illustrates the close proportionality of these average

energies.

loes not vary enough to cause a large error. It

should be noted however, that if the room is built

3f heavy concrete such as ilmenite or barite con-

:rete, the thermal neutron capture cross sections

ire considerably larger than in normal concrete,

and the method will then underestimate the source

strength. In such cases, there is enough variation

Ln the various heavy concretes that one would have

to make a calibration of each room by putting, for

example , a PuBe source in the room and measuring

the thermal fluence and redetermining the constant

Ln the equation, then using this constant in

jvaluating the thermal neutron measurements with

the accelerator operating.

Neutron Production in a Medical Accelerator

From Swanson' si- •'5' ^^-^ calculations, one can

alculate the neutron yield from a single target

naterial of finite or infinite thickness ( > 10

i.L.). It is of interest to compare the neutron

l^ield from an accelerator with these models.

We will consider a typical therapy machine

*ith the geometry of Fig. 7.

Case 1. Target, Flattener, Jaws and Shielding

All of the Same Material

This would always be Pb or W and, to a reasonable approximation, the n yield of these can be

onsidered the same. The only photons effective

Ln producing neutrons are those above about 8 MeV.

These are mostly forward directed and will either

Je absorbed in the main collimator or pass through

Fig. 7. Arrangement of neutron-producing parts within a radiation treat- ment unit (not to scale) . All of these components are practically indispensible and are found on all

standard models. Numbers indicate the approximate percentages of neutrons produced at 25 MeV relative to the maximum number possible, assuming all

parts are of high-Z materials (W or Pb)

its opening and through the f-lattener. Those which

penetrate the flattener may or may not strike the jaws, depending on their opening. Those passing

through the jaws' opening may be absorbed in the patient, the beam stopper If there is one, or the room concrete. The neutron yield from concrete can

be considered negligible, especially since selfshielding would be large. We then have the following possibilities:

Case la - Jaws closed - Yield will be that

of an infinite target; Case lb - Beam stop in place - Yield will

be that of an infinite target; Case Ic - No beam stop in place - Yield

will depend on position of jaws.

Cases la and lb provide upper limits to the amount of neutron production. Case Ic is more complicated and we will consider it in more detail. Most medical linacs provide fields up to about

35 X 35 cm2 at 100 cm from the target. The main

collimator covers all forward directions beyond

the extremes of these fields. The half-angle of the main collimator is then about 14 degrees. In order to see what fraction of the high-energy

( > 8 MeV) photons are within this angle we have

used the Monte Carlo program EGS'--^-' for various

target thicknesses and electron energies. We have

used what we believe to be practical linear accelerator target thicknesses. At these thicknesses,

79
about 2% of the incident electron energy is trans- mitted by electrons in the cone within 14°. Since

the average energy of these transmitted electrons

has been reduced sharply, the radiation yield will

be low and there should be no penumbra problems

from bremsstrahlung produced in the flattener or elsewhere. The target thicknesses we have con- sidered are listed in Table II.

Table II. Target Thicknesses Considered

Electron Energy

(MeV)

Target Thickness (R.L.)

Copper Tungsten

14 0.3 0.7 20 0.4 0.85

25 0.74 1.00

30 1.1 35 0.92

45 1.05

From our Monte-Carlo calculation we can construct

curves such as are shown in Fig. 8.

100

90

80

70

60 h

50

40

30

20 -

10

0

I I I I I I 1

1

• 25 MeV I R.L. W

X 20 MeV 0.4 R.L. Cu o 25 MeV 0.74 R.L. Cu

6

6 •

-J I I I I I 1J I I I I 1

10 8 (degrees)100

Fig. 8. Percentage of high energy photons ( > 8 MeV) emitted within

the angle 9 from an accelerator

target, as a function of 9. Three

target material and thickness com- binations are shown. Data are based on Monte-Carlo calculations

using the program EGS.

As a calculational example consider a 25 MeV

accelerator with a 1.0 R.L. W target, W head

shielding and a W flattener. From Fig. 10 of Swanson,L153 we find that the neutron yield from

this target is about 33% of that from an infinite

target. From Fig. 8 we see that the remaining

high-energy photons are 38% inside the main

collimator angle and 62% outside, and these strike

the main collimator which is practically an infinite absorber. The 38% inside the main colli- mator strike only the flattener. A tungsten

flattener for this target would be about 2.3 cm

(6.6 R.L.) thick in the center and approximately

conical. If we weight this conical shape with the high-energy photon distribution, we effectively hav^

2.9 R.L. over this angular range. Using Fig. 10 of Swanson again we find that these photons produce

47% of the infinite target yield (difference bet- ween 1 R.L. and 1 + 2.9=3.9 R.L.). Our total neutron yield relative to the infinite target then

is as follows:

Target

Main Collimator - Flattener

Total -

0.61 X 67%

0.47 X 0.38 X 0.67

= 33%

= 41.5%

= 12% = 86.5% From Fig. 8 of Swanson, the yield for an infinite target of W at 25 MeV would be 1.5 x 10^2 n/sec per kW. In Fig. 9 we have plotted the Monte- Carlo unflattened and flattened photon dose rates

at a typical 1 meter target distance per mA of electron current for the target thicknesses of Table II.

10'

10^

o

I 10"

1 I I

\ 1

' EGS W 0° Unflattened

) EGS Cu Unflattened

i EGS W Flattened

I EGS Cu Flattened Commercial Mochines

9

1

x

10 15 20 25 30 35 40

Electron Energy (MeV) 45

Fig. 9. Photon absorbed dose index

as a function of incident electron

energy, as obtained from the MonteCarlo program EGS. Upper data are from W and Cu targets without

flattener. Lower points show same data but after flattening. Measured values obtained on existing

commercial accelerators are shown

for comparison.

From this figure we have 8.95 x lo^ rads/min per mA or 60 rads/sec per kW for our example. Then

the yield of neutrons relative to the photon out- put would be 1.5 X 10^2 n/sec-kW x 0.865 ^60 rads/

sec-kW = 2.16 X lo^O neutrons/rad . This would be

80
the yield with the jaws fully open. Fully closed

the yield would be 2.16/0.865 x 10^0 = 2.5 x 10^°

n/rad. With intermediate jaw openings the yield

can be calculated by converting the jaw opening to angle and using Fig. 8 to calculate the yield.

Case 2. Target, Flattener, Jaws and Shielding of Different Materials

In all cases, the jaws and shielding will be

jof Pb or W, (the only other possible shielding

[material, depleted uranium, becomes activated due

[to photofission at energies too low to have significant photoneutron production) . There are actually

relatively few choices for the other elements.

The required physical properties of the target

limit the possibilities to copper, tungsten, gold,

tantalum, platinum and silver. Tungsten, gold,

tantalum and platinium are nearly identical in neutron production. The authors know of no example

iof silver being used. Therefore, we can limit the [discussion of target material to copper and tung-

|sten. Flatteners have been made of lead, tungsten

|and iron. Aluminum flatteners have been used in

betatrons and proposed for linacs, ^

'• » but are

too long for rotational therapy machines and produce undesirable changes in beam hardness across

the field [7] In calculating neutron production

then, it is sufficient to consider only iron and

tungsten for the field flattener. We must finally

consider the following cases.

Case 2a - W target and shielding with an

Fe flattener; Case 2b - Cu target and W shielding and

flattener; Case 2c - Cu target, W shielding and Fe

flattener. .

Case 2a is not too different from our previous example. The neutron components from the

target and the main collimator are the same. An

iron flattener for 25 MeV would be about 9.4 cm =

5.3 R.L. long. If we assume that the spectral

difference between the bremsstrahlurg produced in

a Cu and a tungsten target is insignificant we

can proceed as before. The effective thickness

of the flattener is 2.4 R.L. The photons striking

the flattener produce 42% of the infinite target

yield (difference between 1 R.L. and 1.0+2.4 R.L.

from Fig, 10 of Ref. [15]). The flattener neujtron yield is given by 0. 42 x Q, 38 x o. 67 x Fe yield/

|W yield = 1.6% of the infinite W yield. The total

|yield for this machine would be 75.9% of the

jinfinite tungsten target or 1.6 x 10^0 n/rad with

I the jaws open,

j Case 2b. It is easier to consider this case

as relative to an infinite copper target. In the

same manner as before we find the following:

0.74 R,L, Cu Target Yield - 25% of infinite

(Fig. 10 of Ref. [15])

Fraction of High-Energy Photons Passing

through Main Collimator - 47.5% (from

Fig. 8) Effective Flattener Thickness = 3.3 R.L. Infinite Target Cu Yield = 0.36 x 10^2 n/sec-kW

100

90

80

70

60

50

40 -

30 < 20 y 10 -

IEUJ

1 1—I—I

I I I

1

1 r

25 MeV 0.2 R.L. W

rrrrf

I I I I I 1

1

_i I I 10 9 (degrees) 100

Fig. 10. Percentage of high-energy

photons (> 8 MeV) emitted within the angle e, as a function of 6, when

25 MeV electrons are incident on an

0.2 R.L. W target. Data are from

the Monte-Carlo program ECS.

Infinite Target W yield = 1.5 x 10

Target Yield

12 n/sec-kW

25% of Infinite Cu Target

Yield

Flattener Yield = 0.55 x 0.75 x 0.475 x 1.5

X W Yield/Cu Yield

= 82% of Infinite Cu Target Yield

Main Collimator Yield

Total:

0.525 X 0.75 X 1.5

X W Yield/Cu Yield

164% of Infinite

Cu Target Yield

271% of Infinite

Cu Target Yield

9.76 X 10^^ n/sec-kW

The flattened dose rate from the machine would be

6.95 X 10^ rads/min at 1 meter for 1 mA. Then the relative neutron yield wouId be = 2.11 X lolO n/rad.

Case 2c. This case is the same except that

the neutron yield from the flattener has an effective thickness of 3.0 R.L. The infinite target

yield for Fe is 0.23 x 10^^ n/sec-kW.

Flattener Yield

Total Yield

0.50 X 0.475 X 0.75

X Fe Yield/Cu Yield

11.4% of Infinite Cu Yield

200% of Infinite Cu Yield

7.2 X loll n/sec-kW

Since the flattened dose rate would be the same,

the relative neutron yield wou Id be 1 .56 x lolO n/rad.

The results of these calculations are sum- marized in Table III below.

81
Table III. Summary of Neutron Source Calculations C25 MeV electrons)

Target Flattener

Main

Collimator n/rad

Fraction

of Infinite W Yield

(^Percent)

Percentage From

Target Flattener

Main

Collimator

W W W 2.2 X 10^° 86.5 38 14 48

W Fe W 1.6 X 10^° 64 43 2 55

Cu W W

o 1 V 1 n 1

I, I X iU o / 9.2 30 60.5

Cu Fe w 1.6 X 10^° 64 12.5 5.7 82

Note that while the total neutron yield does not change very much the fraction of the neutrons originating in different areas changes. This is useful information for an accelerator designer

trying to minimize neutron production.

It is of some interest to compare the above

results with those calculated for a 25 MeV betatron. The effective target thickness for a betatron is difficult to obtain since it is never known exactly where the electrons strike the target. However, we can get some indication from the angular distribution of the unflattened beam. ATC Betatron Corp. gives a graph of this in their

literature. The target is platinum which is very similar to tungsten. An EGS calculation for angular distribution of bremsstrahlung from a 0.2

R.L. W target matches the ATC graph quite well.

Information kindly supplied by ATC indicates the electrons strike a nearly triangular target with a minimum thickness of 0.08 R.L. and a maximum

thickness of 0.27 R.L. so assuming an effective

target thickness of 0.2 R.L. seems quite reasonable. In this betatron the flattener is aluminum

and che maximum field size is 14 x 14 cm^ at 1 meter. The corner of this field would be at an angle of 5.7 degrees. The main shielding and jaws

are lead. From Swanson'--'-^-' and interpolated results of Seltzer and Berger , the neutron yield

from the target would be about 3% of an infinite

target. The angular distribution of the photons

above 8 MeV is shown in Fig. 3. We can see that

29% of these photons are within 5.7 degrees.

These photons strike only the aluminum flattener.

The Al to W yield ratio is about 3.1% so the yield

from the Al flattener is negligible. The 71% of the high-energy photons striking the main colli- mator give 0.71 x 0.97 = 69% of the infinite target

yield; the total yield will be 72% of the infinite

target yield or 1.08 x 10^2 sec~^ kW^. The photon

dose rate flattened to 5.7 degrees, 1 m from this

target would be about 2.4 x 10^ rads/min-mA and

the relative neutron yield would therefore be

6.8 X 109 n/rad. One might ask what happens to the electrons

transmitted through this target. From the EGS calculation one can find that essentially all of

the electrons are transmitted and carry an average

energy of 19.6 MeV. Presumably, a few of these

lose so little energy that they make another re- volution and strike the target a second time. Most of the electrons, however, must spiral in and

strike the donut. They would produce bremsstrahlung which would make neutrons but would not add

to the useful photon yield. The neutron production

from these electrons is difficult to estimate but

because (1) the photon production is in low Z material, (2) the energy is lower, and (3) the neutron production would be partly in iron and copper rather than lead, it seems that these neutrons would add not more than 10% to the total.

We have made calculations similar to the above for several accelerators for which we have

sufficient data on the geometry and materials. In each case, we made measurements of the total fast- neutron source strength Cdetermlned by the method

described previously of measuring thermal neutrons

in the room) and we have compared our results with

these.* The results of these calculations are

shown in the Table IV. In general, the comparison

is fairly good. One should remember that we are comparing only the neutrons produced by electrons

which produce the usable photon dose. In most

machines, there are other electrons which are accelerated and lost, and these can contribute to ' the neutron leakage dose but not to the useful ! photon dose. It is believed that this is the largest part of the discrepancy for both models of , the Clinac 35. This machine has two bends with \ energy analyzing slits before the beam strikes the I target, and the authors suspect that there is a

| large beam loss at one or both of those bends.

j Both the Clinac 18 and the Mevatron XX would be

expected to have a much smaller beam loss. These | results indicate that, in accelerators where the , beam losses are reasonably well-known and if target thicknesses and geometries are known, one can probably calculate the neutron yield per photon I rad to within ±20%. With these calculations and ; the "cookbook" methods described in MJS, one could i also calculate the head leakage for medical accel- | erators. The total accuracy one would expect | should be no worse than about ±50%. |

I

The authors wish to express their thanks to Siemens Medical Laboratories, Inc., and to Varian Associates for making available measurements and information about their accelerators for these calculations and comparisons.

82
Accelerator

Energy

Relative Neutron Yield (n/rad)

(MeV) (Calculated) (Measured)

ATC 25 MeV Betatron 25 6.8 X 10^ 6.9 X 10^

Siemens 42 MeV Betatron 42 3.8 X 10^ 3.7 X 10^

Varian Cllnac 35 (Old) 25 4.3x10^0 8.1x10^0

Varian Cllnac 35 (New) 25 2.2 X 10^° 6.2 X 10^°

Varian Clinac 18 10 3.9 X 10^ 4.2 X 10^

Siemens Mevatron XX 15 5.8 X 10^ 7.6 X 10^

Table IV. Calculated

and Measured Neutron

Yield per Photon Rad

Leakage Neutron Depth-Dose Curves

There has not been very much attention paid

to the depth-dose distribution of the leakage neutrons in a patient. It has been shown that these neutrons are of very low energy, and one would

expect them to be attenuated quite rapidly in

tissue. We have made an attempt to calculate this

depth-dose distribution using the computer code

MORSE. In the calculation, we have assumed a point

source 1 m from the center of the phantom. The

source spectrum used was that of either a 14 MeV

(incident electron energy) photoneutron spectrum

surrounded by 4 inches of tungsten, or a 25 MeV

photoneutron spectrum surrounded by 4 inches of tungsten. The phantom was a water cylinder one

meter long and thirty centimeters in diameter,

centered at 1 m from the target perpendicularly to the beam axis. In Fig. 11, we show the results of these calculations of absorbed dose for the two

spectra. In Fig. 12, we show the dose-equivalent

calculations for the same two spectra, using the

ICRP-21 conversion factors (see Fig. 2). There

are several things of interest in these two

figures. First, it can be seen that there is a quality factor of about 10 throughout the depth of

the phantom for both spectra. Second, it can be seen that the attenuation is quite sharp, so that

there would be very little dose to the patient

beyond 10 centimeters. Because of the large

difference in penetration between these low-energy neutrons and the high-energy photon leakage, the integral dose-equivalent to the patient would be

much less from the neutrons than from the photons,

for the same dose-equivalent. In the calculations

just presented, the radiation source and phantom were both in vacuum rather than in a concrete room.

If the same calculation were done for a concrete

room, a portion of the leakage dose to which the patient would be subjected would be the scattered

component from the walls which would be much softer

than the direct leakage spectrum. Therefore the attenuation of the scattered component would be

much faster.

Figure 13 shows the calculated neutron fluence

rate at 1 m per mA of incident electron current,

plotted as a function of electron energy (from

Fig. 5 of Ref. [16]). The conditions for which

these curves hold correspond to the case in which

all neutron-producing parts (Fig. 7) are of W or

NEUTRON ABSORBED DOSE

10"

10"

< 10"'^

z 1

1 Illl 1 1 1111 =

- Air 1 HgO

(a) Spectrum

- Air

1 •

(b) Spectrum

•

• 1

1 •

14 MeV Electrons on Tungsten — •

1 • •

25 MeV Electrons on Tungsten ^

•

III 1 •

•

1 1

• Illl 1

1 1 Illl"

0 5 10 15 20 25 0 5 10 15 20 25 DEPTH IN HjO (cm

)

Fig. 11. Absolute depthdose distributions in H2O

for neutron spectra from

therapy targets, modified

by 10 cm of W. The water

phantom is a 30 cm-diameter

cylinder, 2 m long, oriented perpendicularly to the beam axis and centered

at 1 m from target. Units

are rads per primary

(photo-) neutron. The air- water interface is indicated

at 0 cm (the leftmost point

represents the air dose)

.

Data are from the MonteCarlo program MORSE.

(a) 14 MeV incident

electrons

;

(b) 25 MeV incident

electrons

.

83
NEUTRON DOSE EQUIVALENT

Air H,0

(a) Spectrum

14 MeV Electrons on Tungsten

Air

I HgO

(b) Spectrum 25 MeV Electrons on Tungsten

10 15 20 25 0 5 DEPTH IN HjO (cm)

10 15 20 25

Fig. 12. Absolute depthdose-equlvalent distributions in for neutron

spectra from therapy tar- gets modified by 10 cm of

W. Conditions are the same as for Fig. 11,

except units are rem per primary neutron.

(a) 14 MeV incident

electrons;

(b) 25 MeV incident

electrons.

lO'O

Maximum Neutron Fluence Rote ot Im per mA Incident Electron Current

10 15 20 25 30 35 40

Electron Energy (MeV) 45

Fig. 13. Calculated maximum neutron

fluence rate at 1 m per mA of incident electron current, plotted as a function of electron energy. The

conditions for which these curves

hold correspond to the case in which

all neutron-producing parts (Fig. 7) are of W or Pb and the movable jaws are fully closed. Isotropic neutron production and no attenuation of neutrons in shielding materials are implicitly assumed. Lower two curves correspond to cases in which

all neutron-producing parts are made of Fe or Cu and are shown for comparison. Bata are derived from

Fig. 5 of Ref. [16].

Pb and the movable jaws are fully closed or almost

so. Note that the contribution of the movable jaws

would amount to about 13% of the total (at 25 MeV; see examples above) . Isotropic production of photoneutrons is implicitly assumed. For this

graph we also neglect the attenuation of neutron

fluence in high-Z shielding as well as the room- scattered component discussed above. The high-Z

materials would change the fluence by a factor in

the range 0.85 (for W) to 1.0 (Pb) . The roomscattered component would boost the fluence by

about 20% but increase the patient's integral dose

by a relatively smaller amount. As these effects

are not large and tend to cancel, we believe that

the curves of Fig. 13 reliably represent the maxi- mum neutron fluence of significance to the patient,

assuming that the electron beam only strikes the intended target.

The result of dividing the data of Fig. 13 by

those of Fig. 9 gives us the ratio of the maximum

neutron fluence iJimax ^° 'he useful photon dose at

the same distance. This is an absolute prediction

to which comparison with measurement is invited.

This ratio becomes nearly constant above about 25 MeV incident electron energy where its value is about 2.1x10^ neutrons cm~2 rad~l. As discussed

by examples above, measured data that fall significantly below the curve are likely due to cases

in which energy-absorbing accelerator components

are not all of high-Z materials, or measurements

were made with the movable jaws open. Points that

fall significantly above probably represent cases

in which there is substantial loss of beam on "targets" within the transport system before it reaches the intended target.

Using the above information, we are in a position to compare numerically the integral dose

from neutrons to the integral dose from leakage

photons. For the numerator of our comparison we

integrate the curves of Fig. 11 over the range

0-30 cm and multiply by the strength of the neutron

source per photon treatment dose (from Fig. 14). For the denominator we integrate an exponential

84
105

10"

"T 1 r

Bo Be

10^

Maximum Neutron Fluence

Per rod Of Useful Beam

(0°, flottened) o Meosured o Calculated B Betatron

10 15 20 25 30 35 40 45

,-„ Electron Energy (MeV) ,

Fig. 14. Maximum neutron fluence

per rad of useful beam 0°, flattened). Data are from the W curve of Fig. 13,

divided by the photon absorbed dose

index for W (flattened) as calculated

by the Monte-Carlo program EGS (the

triangles of Fig. 9). The conditions

for which this curve holds are the same as for Fig. 13. Calculated

points (circles) and measured points

(squares) are mostly from Table IV.

over the same interval 0-30 cm, using -0.041 and

-0.037 cm~^ as values of the attenuation coefficient for 14 and 25 MeV, respectively. The attenuation of the photon distributions includes the inverse-square reduction with distance, as is the case for the neutron distributions (Fig. 11). The

denominator is multiplied by 0.001, corresponding

to the ratio of leakage radiation to useful beam

found in most standard accelerator models. This

is meant only to be a rough comparison and details

of the phantom geometry are neglected. The results

are:

neutron integral absorbed dose

leakage photon integral absorbed dose

= 0.04 (at 14 MeV), and

= 0.20 (at 25 MeV)

.

Using the quality factor estimated by comparison

of Figs. II and 12, we would multiply by Q= 10 to obtain

neutron integral dose-equivalent

leakage photon dose-equivalent

= 0.4 (at 14 MeV) , and

= 2.0 (at 25 MeV).

We note that most of the difference between these

two energies is directly traceable to the increase

in neutron yield per useful photon dose seen in

Fig. 14.

Conclusions

Based on what we consider quite reasonable

models, we can make the following statements re- garding neutron sources and their characteristics:

(a) We have identified the significant sources of neutrons within the treatment head, and given

absolute estimates of the amount of neutron

fluence per treatment dose for various

choices of target.

(b) MORSE calculations show that the neutron

fluence is nearly unaffected by transport in the high-Z shielding contained in standard

treatment units.

(c) On the other hand, the same calculations show

that the neutron spectrum is softened significantly by transport by the high-Z material.

(d) We have considered the component of neutrons

scattered by the concrete room and found its average energy to be about 0.24 of the primary source average energy. The scattered com- ponent can make a significant contribution to fluence measurements Cabout 20%) , but con- siderably less to the patient's integral dose.

Ce) We have submitted an absolute prediction of the ratio of neutron fluence (at 1 m) to the useful photon dose at the same distance.

This ratio rises to an approximately constant value of 2,1 X 10^ neutrons cm~2 rad~l at incident electron energies above 25 MeV.

(f) Measurements at a variety of linear accelerator installations fall within a factor of three of this curve. We suggest that significant deviations of correct measurements can be explained by (1) use of materials other

than high-Z within the treatment head (if the data fall below), or, (2) partial loss of the electron beam before it strikes the target

(if above)

.

(g) The neutron fluence from betatrons is substantially below that from linear accelerators.

The reasons are twofold: (1) Because most of the energy of the electron beam is expended

in the low-Z material of the donut, there is smaller neutron production per incident electron beam current. (2) Because the smaller treatment fields permit a thinner flattener

there is higher photon output per incident

electron beam current.

(h) By means of MORSE, we have calculated absolute

depth-dose and depth-dose-equivalent distributions for realistic neutron spectra in H2O

phantoms and found a rapid falloff with depth.

(1) Using the neutron depth dose curves presented

in this work, together with the estimate of neutron fluence, we are able to calculate the ratio of neutron integral absorbed dose to leakage photon absorbed dose. These values

are 0.04 and 0.2 for 14 and 25 MeV incident

electron energy, respectively. If a quality

factor Q=10 is applied, these correspond to 0.4 and 2.0 neutron integral dose-equivalent

per leakage photon integral dose-equivalent.

85
(j) As for reducing the neutron fluence per useful

photon dose, the following remarks may be

made: (1) As the effective use of these

machines for photon therapy depends on the use

of hlgh-Z materials to produce intense, wide

fields with sharp edges, the possible reduction In neutron production Is not very great.

Thinner targets would be of some help, and certainly most of the neutron production by

electrons striking parts of the transport

system can be eliminated by improved design

or lower-Z materials at critical points.

Other than these steps, the only alternative

is to treat patients at lower energies where

possible. There is a factor of two reduction

from the maximum in neutron production per treatment dose at about 17 MeV and a steep

drop as the energy is reduced further.

Acknowledgement

The authors wish to express their appreciation to R. A. Shore for his help in doing computer

calculations.

References

[1] ATC Betatron Corporation, Twenty-Five Million

Volt Betatron for Radiation Therapy, P. 0. Box Drawer 14248T, West Allis, WI 5321A

(1974).

[2] B. L. Berman, "Atlas of Photoneutron Cross

Sections Obtained with Monoenergetic Photons,"

Bicentennial Edition, Lawrence Livermore Lab,

Report No. UCRL-78482 (1976). Also see "Atlas

of Photoneutron Cross Sections Obtained with

Monoenergetic Photons," Atom. Data Nucl. Data

Tables 15, 319 (1975).

[3] R. L. Ford and W. R. Nelson, "The EGS Code

System: Computer Programs for the MonteCarlo Simulation of Electromagnetic Cascade

Showers (Version 3)," Stanford Linear Accelerator Center, Stanford, CA, Report No. SLAC210 (1978).

^4] R. J. Howerton, "An Integrated System for Production of Neutronlcs and Photonics Calculatlonal Constants," Lawrence Livermore

Lab, Report No. UCRL-50400, Vol. 9 (1970).

[5] International Commission on Radiological

Protection, "Data for Protection Against

Ionizing Radiation from External Sources:"

Supplement to ICKP Publication 15, ICR? Publication No. 21, Table 9, p. 19 (1973).

(Oxford: Pergamon Press.)

[6] R. C. McCall, T. M. Jenkins and R. A. Shore,

"Transport of Accelerator Produced Neutrons

in a Conoiete Room," (IEEE Transactions on Nuclear Science NS-26 P. 1593 (1979) , Stanford Linear Accelerator Center, Stanford,

CA, Report No. SLAC-PUB-2214 (1978).

L7J R. C, McCall, R. D. Mclntyre and W. G. Turnbull, "Improvement of Linear Accelerator

Depth-Dose Curves," Med. Phys. _5, 518 (1978).

[8] G. S. Mutchler, "The Angular Distributions

and Energy Spectra of Photoneutrons from

Heavy Elements," Ph.D. Thesis, MIT, Report

No. MIT-2098-224 (1966).

[9] National Bureau of Standards, National

Council on Radiation Protection and Measurements, "Shielding for High-Energy Electron

Accelerator Installations j' NBS Handbook No.

97 and NCR? Report No. 31, Washington, D.C.,

Fig. 15, p. 25 (1964).

[10] H. W. Patterson and R. Wallace, "A Method of Calibrating Slow Neutron Detectors," Lawrence

Berkeley Lab, Berkeley, CA, Report No. UCRL8359 (1958).

[11] E. B. Podgorsak, J. A. Rawlinson, M. I. Glavinovlc and H. E. Johns, "Design of X-Ray

Targets for High Energy Linear Accelerators

in Radiotherapy," Am. J. Roentgenol. Rad. Therapy Nucl. Med. 121 , 873 (1974).

[12] E. B. Podgorsak, J. A. Rawlinson and H. E. Johns, "X-Ray Depth-Doses from Linear Accelerators in the Energy Range from 10 to 32 MeV," Am. J. Roentgenol. Rad. Therapy Nucl.

Med. 123 , 182 (1975).

[13] S. M. Seltzer and M. J. Berger, "Photoneutron

Production in Thick Targets," Phys. Rev. CT^, 858 (1973).

[14] E. A. Straker, P. N. Stevens, C. C. Irving

and V. R. Cain, "The MORSE Code - A Multigroup Neutron and Gairana-Ray Monte-Carlo

Transport Code," Oak Ridge National Lab,

Oak Ridge, TN, Report No. ORNL-4585 (1970). j. Also see Radiation Shielding Information

Center, RSIC Computer Code Collection,

"MORSE-CG, General Purpose Monte-Carlo

Multigroup Neutron and Gamma-Ray Transport

Code with Combinatorial Geometry," Radiation

Shielding Information Center, Oak Ridge, TN,

Report No. CCC-203 (1976).

[15] W. P. Swanson, "Calculation of Neutron Yields

Released by Electrons Incident on Selected

Materials," Stanford Linear Accelerator | Center, Stanford, CA, Report No. SLAC-PUB2042; and Health Phys. 35, 353 (1978). I

[16] W. P. Swanson, "Calculation of Neutron Yields

Released by Electrons near the Photoneutron i Threshold," Stanford Linear Accelerator \ Center, Stanford, CA, Report No. SLAC-PUB2211 (Rev. 1979); and Health Phys., to be . published (1979). j,

,1
NBS SP 554 (1979) MIXED PHOTON- NEUTRON FIELD MEASUREMENTS

Ravinder Nath Yale University School of Medicine

Department of Therapeutic Radiology

333 Cedar Street

New Haven, Connecticut 06510

and

K. W. Price and G. R, Holeman

Yale University

Health Physics Division

260 Whitney Avenue

New Haven, Connecticut 06520

High energy x-ray radiotherapy machines produce neutrons by photonuclear

reactions which present a potential radiation hazard to the personnel and

patient. A series of measurements of the neutron flux and spectrum from a 25 MV x-ray linear accelerator, inside and outside the treatment room, have

been performed using a multisphere spectrometer, Nemo dosimeter, and activation

detectors. These results are compared with other mixed photon-neutron field measurements for the same machine using an argon/propane ionization chamber,

moderated gold foil, silicon diode, track-etching detectors, and Monte Carlo

calculations illustrating some of the experimental difficulties which are encountered in an accurate measurement of the neutron contamination of a high energy photon beam.

(High energy x-ray radiotherapy machine; Mixed photon-neutron field dosimetry; Neutron leakage)

Introduction

High-energy x-ray radiotherapy machines in the super-mega voltage region also generate

neutrons by photonuclear reactions in the target,

field- flattening filter, and beam-defining

collimators, resulting in a mixed radiation field

in the beam and peripheral areas. The photonuclear cross sections for most materials commonly

used for targets, filters, and collimators have

photon energy thresholds of about 7-8 MeV and peak

cross sections at about 12-14 MeV (1). Therefore,

significant numbers of neutrons may be produced

in accelerators with photon energies above 10 MeV. These neutrons present additional shielding

problems for radiation protection and may lead to significant neutron dose to patients. Several

recent biological and epidemiological studies

(2,3,4) indicate that the relative biological

effectiveness of fast neutrons may be higher than

the currently accepted values. Some of the Investigators have suggested quality factors in the range of 30-50 for low level neutron

Irradiation. These high values of the neutron

quality factors necessitate accurate measurements

of small amounts of fast neutron dose (hundredth

of a mrad/hour) to personnel in a mixed photonneutron field (mrads/hour). In addition, the

hazard to the radiotherapy patient from the small amount of neutron contamination must be determined.

Accurate neutron dose measurements have

generally been more difficult than x-ray measurements for several reasons. Many of the neutron

detectors, such as activation foils measure

particle fluence which is converted to absorbed

dose using appropriate flux-to-dose conversion

factors. These factors depend strongly on neutron energy and hence, the neutron energy spectrum must be known or approximated. Even for detectors which measure absorbed dose

directly, such as the tissue equivalent

ionization chamber, it is necessary to determine

the neutron energy spectrum in order to evaluate

the dose equivalent, because the neutron quality

factors vary from about 2 to 11 as the neutron energy changes from thermal energies to 1 MeV

(5,6). In many of the measurements reported, the neutron detector is calibrated against a Pu-Be, Am-Be or Cf neutron source. Since Pu-Be and Am-Be sources have a much higher effective

neutron energy, their use can lead to systematic

uncertainties. It is well known that the spectrum of fission neutrons from Cf-252

approximates a typical photoneutron spectrum,

and that the shape of a photoneutron spectrum is rather independent of the incident electron energy (7). However, when heavy metal shielding

Is placed around a photoneutron source, as is the case in the treatment head of a radiotherapy

machine, the photoneutron spectrum may deviate

considerably from a fission spectrum. McCall,

Jenkins, and Shore (8), and McCall (9) have obtained quantitative information about this

effect and have shown that assuming a fission

spectrum may lead to significant errors in evaluating experimental results, especially

outside of the primary photon treatment beam.

87
Considerable attention has been devoted to the problems of mixed-field dosimetry for neutron radiotherapy (10). All neutron radiotherapy

machines have a small, but finite photon

component which must be determined. Although

this problem is the opposite of the case of small neutron contamination of photon beams of high energy x-ray machines, it is still instructive to review the methods employed. In order to separate

absorbed dose due to the photon and neutron

components, two dosimeters with different

sensitivities to neutrons and photons are employed. One instrument is usually constructed to have

equal response to both neutrons and photons, and

the other, a small response to neutrons. Since

photons contribute a small fraction of the total dose, this combination results in one dosimeter

responding predominantly to neutrons and the other to photons. It has been shown (10) that

with this combination even a very large

uncertainty in determining the neutron sensitivity

of the photon-sensitive dosimeter contributes

only a small uncertainty to the overall

uncertainty in the neutron dose. Using the same argument for the present problem, if the neutron detector has a low sensitivity to photons, even

a large uncertainty in the evaluation of its photon sensitivity would contribute a small amount

to the overall uncertainty.

Another problem which arises with mixed field measurements is that the high energy photon beam

results in photoneutron production in the detector. As these neutrons produced in the detector are indistinguishable from the external

neutrons, this leads to a serious problem in measuring the dose from neutrons in the beam. Kushelevsky and Shani (11) have reported order

of magnitude calculations of the photoneutron

yield per gram for Si, 0, Al and carbon

irradiated by bremsstrahlung beams with energy

of up to 24 MeV and have used these to calculate

the photoneutron yield of two neutron detectors,

NE213 scintillation detector and 8" polyethylene

spherical rem-meter. Their calculation shows that

the total number of neutrons produced in an 8" spherical rem-detector by the primary beam of

20-24 MV can be quite comparable to the flux of

neutrons in the primary beam and gross errors in the estimation of neutron flux will result. McCall, Jenkins and Tochlin (12) have reported

similar calculations for a moderated neutron

detector, an In foil in a polyethylene cylinder.

Also, McCall, Jenkins and Oliver (13) have cal- culated the photoneutron production in silicon

diodes. Since the number of neutrons entering

the detector depends on the area of the detector while the photoneutron production

depends on the mass, detectors with large

surface-to-mass ratio, i.e., small detectors,

would have lesser interference from internal

photoneutron production. Also, detectors made

from hydrocarbon material have lesser photo- neutron production within them because the threshold photon energy for photoneutron pro- duction in carbon is 18.7 MeV.

Neutron production resulting from high-energy

x-ray beams in radiotherapy has been Investigated

at various centers for many years. Neutron dose levels outside the treatment room for radiation protection, neutron dose in the photon

beam to the patient, and whole-body neutron dose

to the patient have been extensively reported in the literature. Moderated neutron detectors

have been used by Lauqhlin (15), Pohlit (40), Axton and Bardell (16), McGinley et al (17),

Stranden (18), Holeman, Price, Friedman and Nath

(19) ; fast neutron activation detectors by Ernst and Ovadia (22), Frost and Michel (41),

Adams and Paluch (23), Brenner (24), Fox and McAllister (21), Stranden (18), Deye and Young

(20) , Price, Nath and Holeman (25), Gur, Rosen,

Bukovitz and Gill (47); track detectors by

Lofgren and Spring (26), Kehler and Robinson

(27), Sohrabi, Morgan and McGinley (28),

Wilenzick et al (29); silicon diodes by

Wilenzick, Almond, Oliver and DeAlmeida (29),

Marbach (14); scintillation detectors by

Devanney (30), Daniels and Silberberg (37),

Silberberg, Walchie and Daniels (32); and an ionization chamber by Schulz (33). It is difficult to compare these methods quantitatively

because of differing experimental conditions and

radiation sources. However, a series of measurements of the neutron dose equivalent at various

locations in the vicinity of a 25 MV Sagittaire

x-ray radiotherapy accelerator have been per- formed by Schulz (33), Holeman, Price, Friedman

and Nath (19), Price, Nath and Holeman (25),

McCall (9), Marbach (14), and Wilenzick et al (29). In the sections to follow, these measurements will be described and compared to illustrate

the problems associated with these mixed field measurements.

Methods

A schematic drawing of the treatment room of

the Sagittaire accelerator is shown in figure 1. All of the radiation protection measurements described below were made with the accelerator

producing 25 MV x-rays at a dose rate of 400 rad/

min. The entire installation is made from

ordinary concrete, with steel plate in those parts

of the wall struck by the primary beam. The

sliding door, B, contains 2 Inches of lead and 2 Inches of polyethylene. A second swinging door,

A, containing 10 inches of polyethylene was added to reduce the neutron dose to acceptable

radiation levels.

Multi sphere Spectrometer

The multi sphere spectrometer consisted of a bare detector, a 0.03 inch thick cadmium cap, and

six polyethylene moderating spheres (2-, 3-, 5-,

8-, 10-, and 12- inch diameter). The detector

within the spheres used in this study (19) con- sisted of a set of four lithium fluoride thermoluminescent dosimeter chips, two Harshaw TLD-600

(6LiF, - 95.6% 6li) and two Harshaw TLD-700

(7L1F, - 99.99% 7Li). The TLD-700 and TLD-600

are both sensitive to photons, while the TLD-600

is many times more sensitive to thermal neutrons

than is the TLD-700. Thus, the photon con- tribution to the TLD-600 may be subtracted.

88
Figure 1. Schematic drawing of the treatment room of the Sagittaire linear accelerator.

leaving only a thermal -neutron response. A set of four TLD chips is placed at the center of each

sphere and measures thermal neutrons produced as

a result of moderation of the incident neutron

spectrum.

The photon and thermal -neutron responses of

individual TLD chips within a batch were found to vary significantly. Therefore, each chip was calibrated for photon response using a 137Cs source and for thermal -neutron response using a calibrated pi utonium- beryllium source (PuBe) in

a moderating water tank. Each detector assembly

was exposed to the PuBe source and a calibration

of the system was made. This calibration was performed by forcing the spectrometer to yield

a total number of neutrons equal to the known source output. The resultant neutron energy spectrum peaked at 5.5 MeV, very near the average

of 4.1 to 4.5 MeV, as reported in the literature

(34). The response matrix, which relates the

incident neutron energy in a given detectorsphere geometry to thermal neutrons at the sphere

center, was based on machine calculations (35).

In order to compute absorbed dose and dose

equivalent from a neutron spectrum measurement,

dose and dose-equivalent conversion factors were

applied (35). The present calculations represent

27 neutron energy groups ranging from 0.2 eV up to 35.2 MeV neutron energy.

Nemo Dosimeter

The Nemo dosimeter, based on the design of

Bramblett, Ewing and Bonner (43), and Hankins

(44), (model 9140, NEMO spherical Neutron Dosimeter System, Texas Nuclear, Chicago) was

also employed for neutron measurements outside the treatment room. It consists of a 10" diameter

polyethylene sphere with a ^Lil scintillation

detector at the center of the sphere. This

instrument is designed to have an energy response

closely proportional to the thick tissue RBE dose curve.

Phosphorus Activation Method

The reactions 31p(n,p)31si and 31p(n,Y)32p

were employed (25) to determine fast and thermal neutron fluxes in and just outside of the primary treatment beam. The 3ip(n,p)31si reaction is

sensitive to fast neutrons above "0.7 MeV, as shown in figure 2. The neutron-capture

Figure 2. The 31p(n,p)31si activation cross

section.

reaction 31p(n,Y)32p is significant only for thermal neutrons. The activation products ^Isi and 32p are essentially pure beta emitters and

are counted using a liquid scintillation

spectrometer. Details of the activation analysis

using the liquid scintillation spectrometer have

been presented In a previous publication (36).

The phosphorus is used in the form of P2O5 powder

placed into a small vial. The vial (1.2 cm In diameter by 3.5 cm high) usually contains on the average of 2.5 g P2O5. The irradiated powder is later dissolved In water, then mixed with a scintillation cocktail and counted In a liquid

scintillation spectrometer. This technique leads

to counting efficiencies of 95% and 97% for 31si and 32p respectively. When P2O5 is exposed to a mixed neutron-photon flux, many reactions are possible. It has been shown that all activation

products have either short half lives or are stable except for 31 si and 32p, which are pro- duced by fast and thermal neutrons, respectively

(36). The fast neutron flux (t>f (n/cm^-s ) is obtained from the measured 31si saturation

activity, and the thermal neutron flux from the

32p saturation activity. In order to utilize the

31Si activation data for fast flux determination,

the 31p{n,p)3lsi activation cross section must

be known as well as the normalized differential

89
neutron energy spectrum (45). The thermal flux

is assumed to be constant in evaluating the 32p

activation data (46).

Results

Neutron Spectrum Measurements

Neutron spectrum measurements were only

possible outside of the photon beam, and the multisphere spectrometer was the only instrument

used which could provide neutron energy spectrum

information. As an evaluation of the multisphere

technique, a neutron spectrum measurement was made at the Yale Electron Accelerator Laboratory.

The neutron spectrum from 32 MeV electrons

impinging on a thick lead target was measured

at 90° to the beam line by the multisphere

spectrometer and a neutron time-of-fli ght

technique (37). Figure 3 compares the unfolded

I I I I I I ] I ~f"~r--<

0 2 4 6 8 10 Neutron Energy , Me V

Figure 3. Comparison of neutron spectra

determined by time of flight and Bonner spectrometer, normalized to 1 MeV neutron energy.

the findings of others (33). All measurements were performed without the phantom. The data

for the neutron spectrum just inside the shielding door (location A) were accumulated for 2997 photon rad with the beam directed into the

floor. The data for the neutron spectrum outside

the door (location B) were based on integrated

patient photon doses of 85,479 and 26,770 rad. Some of these measured spectra are shown in figure 4. Although the neutron spectrum measurements suffer from poor energy resolution, it is

10

Kauttcn lAargy, luv

Figure 4. Measured neutron spectra at the inside

and outside shielding door of the treatment room.

apparent that neutron intensities and doses are reduced in passing through the door, and the neutron spectrum hardens in passing through the door (see Table I).

neutron spectrum versus the time-of-fli ght

spectrum above 0.5 MeV neutron energy. The neutron source term calculated from the time-offlight spectrum was 2.53 x lO^^ n/sec, while the multispheres indicated a source term of

3.62 X 10^1 n/sec, comparing favorably with the time-of-fli ght source term. Considering the low resolution of the multisphere technique and possible interference due to the large photon flux, the agreement in the determination of the neutron

source term and spectrum is surprisingly good.

It should be noted that the neutron spectrum

derived from multisphere data represents a general

trend in the neutron energy distribution, and any fine structure generated should not be taken

seriously. The main advantages of the system are determination of total neutron flux and resultant

dose equivalent.

The multispheres were placed at various

locations within and outside the treatment room

of the Sagittaire accelerator (outside the treatment beam, in all cases). It was found that

the measured spectra were the same with or without a phantom in the beam, which agrees with

Tabit I

HtS5ur«d Neutron Doit Equivalent Rates and Inferred Quality Factors

Perpendicular

Distance

(hi) Neutron Dose Rate (renVmln) QF

0.5 1.0 X 10° 7.6

1.0 8.3 X 10-1 7.5

2.0 4.2 X 10-1 6.9

4.8 2.0 X 10-1 6.7

Inside Shielding Door 2.6 X 10-2 5.3

OuUlde Shielding Doer 3.3 X 10-8 6.5

5.9 X 10-5 7.0

PuBe 2.6 X 10-* 7.7

90
Measurements inside the treatment room were

taken at perpendicular distances of 0.5, 1.0, 2.0 and 4.8 m from the beam axis with a field size of

10 X 10 cm2. Some of the resulting neutron

spectra are shown in figure 5. The integrated

Neutron EntrgT, KeV

Figure 5. Measured neutron spectra inside the

I treatment room at a perpendicular distance of Im

i and 4.8m, and at the inside door.

fast-neutron flux (> 0.1 keV) varies with the inverse-square law when the distance from the

target is used rather than the distance from the

beam axis. This would indicate that the major

source of photoneutrons is the target and collimator assembly. Slow-neutron fluxes of

less than 0.1 keV do not vary significantly with

the distance from the beam axis and contribute

only a small fraction to the total neutron dose

equivalent throughout the treatment room.

Figure 6 illustrates a comparison between

the differential neutron energy spectra determined

by the multi sphere spectrometer, neutron transport calculations of McCall (9), and an unfiltered photoneutron spectrum from Ta and Pb (7,39).

I All of these spectra are normalized at 1 MeV

' neutron energy. The calculated spectrum agrees

I reasonably well up to about 5 MeV and is in general, more degraded in energy compared with

the spectrometer results. Considering the low resolution of the spectrometer, the overall

agreement with the neutron transport calculation

is fair. It may be noted here that assuming the neutron energy spectrum to be the same as a fission spectrum or unfiltered photoneutron

spectrum, is reasonable only in the beam. As

one moves away from the beam axis, the neutron

I spectrum is markedly degraded compared to the

I fission spectrum (8,9). This is due to neutron

' interactions in the treatment head shielding

and general scatter throughout the room.

I Measurements in the beam with the multi

sphere were attempted but failed due to a variety of interferences, including: (y.n)

• rravldid by K. C. HcCdl, SUC

I I I I I 1 I I I I 0 123436789 10

Figure 6. Comparison of various determinations

of the neutron spectrum inside the treatment

room.

reactions in the cadmium, unequal photonuclear

cross sections of ^Li and 'Li, and (y.n)

reactions in the polyethylene spheres. As long

as the spheres were located outside the beam

there was no indication of these interferences.

Neutron Dose Equivalent Measurements Far From Treatment Beam

Each neutron spectrum measured using the multisphere spectrometer was converted to absorbed-dose rate and dose-equivalent rate by

utilizing maximum values of flux-to-dose con- version factors (6) and the appropriate quality

factor at each energy interval of the spectrum.

The resultant total dose equivalent rates and

quality factors at various locations

within and outside the treatment room are given

in Table I. As a calibration of the multisphere

system, the neutron dose-equivalent rate, and

quality factor from a PuBe neutron source are also shown in Table I. Previous studies

indicate a calculated Quality factor of 7.1 to 7.5 for PuBe neutrons (34). For PuBe neutrons

a quality factor of 7.3 was inferred from measurements at a source-to-detector distance

of 50 cm. This good agreement between

theoretical and measured PuBe-neutron quality

factors provided confidence in determination

of average quality factors for fast neutrons.

91
Table II shows a comparison of neutron

dose equivalents measured by various methods.

For measurements outside the room, with the doors

Table II

Conparlson of V«r1ous Methods for Neutron Measurements

Far From the Treatment Beam

Dose Equivalent Method QF (mrem/hr)

Outside treatment room with doors closed

Multlsphere Spectrometer 7.0 0.33

NEMO (PuBe Calibration) 0.51

(Cf Calibration) 0.62

Ionization Chamber* 7.1** 0.35'''

Outside treatment room with doors open

Moderated Au Foil** — ISO.

Ionization Chamber* 7.1** 62.

In the Haze

Multlsphere Spectrometer 5,4 1560.

Moderated Au Foil** 1488.

*

Schulz (33)

**McCall (9)

* Corrected for by a factor of 2.64 for difference In location of detectors.

closed, neutron dose measurements using an argon/

propane ionization chamber. Nemo survey meter,

and Bonner spectrometer, are in good agreement

with the Nemo dosimeter indicating the largest

dose equivalent. The Nemo dosimeter was calibrated using PuBe and Cf-252 sources, in order to evaluate its capability to perform under different

energy spectra. It was found that the Nemo results did not differ by more than 25% depending

on the source used for calibration. Also, it may

be pointed out that the Nemo dosimeter is known

to overestimate neutron dose equivalent especially

in the intermediate neutron energy range (44). In the usage of the multi sphere spectrometer,

maximum values of flux-to-dose equivalent con- version factors were employed. The ionization

chamber underestimates the dose by about 15%

because of a lack of proton equilibrium (33).

Considering these factors, the overall agreement

between various methods outside the room is fairly good. Neutron dose rate with the door

open and in the maze was measured by McCall ( 9) using a moderated gold foil and a neutron trans- port calculation using MORSE code for subsequent

determination of dose equivalent. His results compare very well with the results from the multlsphere spectrometer and differ by a factor

of about 2.5 from the Ionization chamber results.

The Nemo dosimeter was not used in the treatment

room because of the pulse pile-up problem in a high photon flux.

Neutron Dose Equivalent Measurements In and Near Treatment Beam

Because of the large flux of high energy

photons in and near the treatment beam, fast neutron measurements in this area were per- formed using the 31p(n,p)31si activation method. Thermal neutron measurements were performed

using the reaction 31p(n,Y)32p. There are no direct photonuclear interactions which will

produce 31 si or 32p. jhe only photon contamination is due to a two-step process:

C Y» (n. pn, or 2n) ] in P2O5 followed by

31p[n, (p or t)3 reaction. In order to determine

this photon sensitivity of the detector, several experiments were performed at the Yale University

Electron Accelerator Laboratory. Capsules of

P2O5 were placed in the intense 32 MeV bremsstrahlung beam which also had a neutron

component. In addition, capsules were placed

just outside of the photon beam where neutrons were dominant. Activities induced by the mixed

n-Y fields with different amounts of photon

component were measured. From these data, the photon sensitivity of the neutron detector was determined (36). The neutron source term was measured using a neutron time-of-flight spectrometer, and the high-energy photon fluxes

(> 7.5 MeV) were determined using gold activation

foils '37Au(Y,n) '9°Au and a GeLi-y spectroscopy

system for determination of 196/\u saturation

activity. The photon differential flux

K (y/cm^-s-MeV) was then calculated from the l36Au saturation activity. By taking the difference between the "in" and "out" of beam

31 Si induced activities and correcting for the known neutron source term, the 31 si saturation

activity due to the photons was determined. It was found that a good estimate for the photon

sensitivity of the P2O5 capsule is 1.12 x 10"^^

(31si dps/target nucleus)/(Y/cm2-s-MeV). Photon

fluxes at the 25 MV Sagittal re were measured

with gold foils and the photon sensitivity for

the Sagittaire beam with various field sizes was found to be less than 4%, as shown in Table III.

Capsules of P2O5 were placed at various

distances from the Sagittaire photon central axis in a plane 105 cm from the target, with

no scattering material near the detectors. The

capsules were given 4000 photon rad per field

size measured. Field sizes utilized in the present work were 5 x 5, 10 x 10, 20 x 20 and 30 x 30 cm2. In a separate experimental

run, a gold foil was placed on the photon beam

axis for a photon- flux determination for each field size. Shown in figure 7 are plots

of the fast-neutron fluxes and dose equivalent

rates obtained from the measurements. The

edges of the field are rounded in all cases, and the neutron flux increases as the field

size increases. The activation data were

92
Tible in

EstliMted Photon Sensitivities of Neutron Detector

in Si9ltt»1re Beams

Field Size

(cm)

5 X 5

10 X 10

20 X 20

30 X X

Estimated

Photon Sensitivity

(»1

3.8

2.7

1.9

1.7

I I , I I I 10 20 30 40 OISTMCE FROM BE«H AXIS, i

50 60

Figure 8. Measured thermal -neutron fluxes and

dose-equivalent rates vs. distance from photon

central axis.

*U$in9 1.12 X 10'^' (^'si dps/target nucleus)/{Y/cm^s MeV).

DIttanct from Btom Ails ,cm

Figure 7. Measured fast-neutron flux and doseequivalent rates vs. distance from photon central

axis.

converted to neutron flux by assuming a neutron

spectrum shape equivalent to a Ta-Pb photoneutron

spectrum (discussed later in this section). For

every energy group in the assumed neutron energy

spectrum (0.5 to 11.0 MeV), maximum dose

equivalent conversion factors were obtained from

ICRP 21 (6). From these conversion factors, doseequivalent rates were computed from the activation

data. The maximum values of dose equivalent at

each neutron energy represent the dose equivalent

at the depth of maximum dose equivalent. This

procedure overestimates the dose because dmax for

the various neutron energies occurs at different

depths. The largest neutron flux measured was

1.0 X 106 n/cm2-s for the 30 x 30 cm field,

representing a dose-equivalent rate of ~ 2.25 rem/

min (using the Ta-Pb spectrum).

Measurements were also made of the thermal

neutron flux for all of the photon beam sizes

given above. Plots for a 5 x 5 and 20 x 20 cm

beam are given in figure 8. Thermal -flux profiles

were essentially constant for all beam sizes.

Thermal neutron fluxes were low and near the threshold of detection for 32p, Thermal fluxes

varied between 1.5 x 10^ and 3.0 x 10^ n/cm2-s.

or, using maximum dose-equivalent fconversion

factors, 0.01 and 0.02 rem/min. Thermalneutron dose equivalent was about 0.7% of

the maximum fast-neutron dose-equivalent

measured.

In addition to the measurement of photon

sensitivity (described earlier), two further

confirmations of the lack of photon sensitivity

of this activation technique were observed.

Plotted in figure 9 are the photon and fast- neutron normalized fluxes for a 5 x 5 cm field

size. It may be seen that the photon flux falls

Figure 9. Comparison of fast-neutron and photon

profile for 5 x 5 cm field.

off much more rapidly than does the neutron flux.

In addition, the neutron profile has a rounded

shoulder as opposed to the photon profile.

Since the measured neutron profile does not follow the same shape as the photon profile, this

provides an indirect check on the insensiti vity

of the neutron detector to photonuclear interferences. Shown in figure 10 is the variation

of the on-axis neutron and photon fluxes as a function of field size, normalized to the 10 x 10 cm field. As the field size increases, the neutron fluxes increase much more rapidly than

do the photon fluxes. This again provides an indirect check on the photon Insensiti vity of

the system.

93
Figure 10. Photon and fast-neutron central -axis

flux vs. field size.

One important assumption in the determination of the fast flux from the 31si activity is the shape of the differential neutron spectrum.

In order to evaluate the uncertainty in the determination of fast neutron flux resulting from

a lack of knowledge of the exact neutron spectrum,

the factor for converting the 31 Si saturation

activity to fast flux was evaluated for neutrons produced by 32 MeV electrons on a thick Pb target,

23-30 MeV electrons in Ta and Pb targets (7,38,

39), neutrons from a 252cf source (42), and

McCall's in-beam Monte Carlo generated neutron

spectra (see figure 6). The values of this

conversion factor are presented in Table IV. It

Table IV

Conversion Factors from Saturation Activity to F1u«

Conversion Factors

32 HeV electrons o25

on Pb target 2.325 X lO'

23-30 MeV electrons , ,„ , ,-25

Pb and Ta targets

252Cf source 2.897 X 10'

Monte Carlo « 3 205 x 10^^ calculation '•^"^ *

McCall (9)

may be noted that maximum deviation for these

different spectra is from 2.3 x 1025 to 3.2 x 10^5. The overall uncertainty in the fast neutron flux in the photon beam arising from a lack of knowledge of the exact neutron spectrum

is significant, but is not as large as one might

expect. However, one would expect the uncertainty to increase as the distance from the beam

axis increases (see figure 6).

Figure 11 and Table V illustrates the

comparison of neutron dose measurements in and

around the treatment beam using the P2O5

activation method, silicon diodes (14,29), track

etching detectors (29) and multisphere spectrometer. Neutron dose equivalent measurements

10 « 10 riau

too lUd/Klii

0 Vileulck, tt.tl., SI Sliidu ud Track acoja <29) A Karbach, SI Diodaa (U)' — PjOj Activation, Koata Carlo Spactrua AcclvAdOD, T« And Pb SpftccruM

C1«UQC« Froa 8«u AxU, Kac«»

Figure 11. Comparison of neutron dose profiles

measured using P2O5 activation method, silicon

diodes, multispheres, and track-etching

detectors.

plotted here are for a 10 x 10 cm^ field at a photon dose rate of 400 rad/min. Two plots are shown for the fast neutron dose equivalent

as determined from the P2O5 activation measurements. The upper curve represents a varying

spectrum shape with distance from the beam axis

as computed by McCall using Monte Carlo calculations. The lower curve represents a constant

spectrum assumption from a combined tantalum

and lead target (7,39). As mentioned earlier,

the choice of spectrum makes a significant

difference in the measured neutron dose,

especially at points outside the x-ray beam. Since the Monte Carlo calculated spectrum in- cludes the details of filteration, collimation

and room scatter, it represents a better choice

than the assumed unfiltered photoneutron spectrum, as one moves out of the treatment beam. A quality factor of 7 has been used in converting

the data of Wilenzick et al from neutron rad/

photon rad to neutron re^/photon rad. The

quality factor used by Marbach was 10. Wilenzick

et al observed that outside the x-ray beam, the silicon diode and track detector agreed very well with each other. However, in the x-ray beam,

their track detector was sensitive to photo94
Tiblt V

Comwrlson of Various Methods for Neutron Heasuremnti

In and Near the Treatment Beam

Neutron Dose Equivalent Rat»

(rem/nrinl

At Isocenter At 50 cm to the side

PjOs Hithod

MuUlsphert

Mont« Carlo*

Silicon Diode**

Silicon Dlode^

Trick Etching**

3.04

2.25

14.8

1.52

1.04

1.07

1.40

1.40

At Im

.82

1.36

McClll (9)

*H11«Mlek tt al (29)

Hirbach (14)

fission and was not used. Their measurements

with Si diode in the beam are an order of

magnitude higher than activation detector

results, and this has been attributed to the photon sensitivity of the silicon diode by

McCall, Jenkins and Oliver (13). Marbach's

diode data outside the x-ray beam are slightly

larger than the activation results, but this

difference can be attributed almost entirely to his choice of a quality factor of 10 which is most likely too large. Also plotted in figure 11 are the multisphere data. It should be pointed

out that the multisphere results represent the

total neutron dose equivalent from thermal to fast neutrons. At distances greater than 50 cm

the multisphere data agree quite well with the

P2O5 activation detector results. It appears

that the P2OK data, when the Monte Carlo spectra

are applied In interpreting the results, yield

a good estimate of the total neutron dose

equivalent. This would indicate that the contribution from neutrons less than 0.5 MeV

are not significant in the dose equivalent

determination.

Discussion

Most of the results from the measurements of neutrons far from the treatment beam using

the multisphere spectrometer, Nemo survey meter,

Ionization chamber of Schulz, moderated gold

foil method of McCall, and P2O5 activation

method, agreed with each other within a factor

of 2. In and near the treatment beam, measurements using the P2O5 activation method, silicon

diodes, track etching detectors, and multisphere

spectrometer were found to agree well only outside the x-ray beam. In the x-ray beam, the

activation detector was the only detector with

sufficiently low photon interference. One of

the requirements for the use of most dosimeters

including the P2O5 activation detectors is that

the neutron energy spectrum must be known. Al- though the neutron spectra in the beam has not

been measured, reliable Monte Carlo calculations

for specific geometries have been made and em- ployed. Thus, a combination of the P2O5 activation detector with the calculated neutron energy

spectra for specific installations offer the best available method among the techniques used

here for determination of neutron contamination

in the x-ray beam.

In this series of measurements to determine

the neutron dose from a 25 MV x-ray machine, it was found that overall agreement between

different methods is in the range of a factor of

2. Chief sources of uncertainty are the photon

sensitivity and photoneutron production in the neutron detector, and the determination of the energy spectrum of neutrons. If the relative

biological effectiveness of low-level neutron irradiation is significantly higher than currently accepted values, more accurate methods

for the measurement of neutron contamination,

with special attention to the problems addressed

in this paper, would need to be developed.

Acknowledgement

The authors gratefully acknowledge the

assistance of Drs. R. J. Schulz and R. C. McCall

in providing their data and helpful discussion.

We would also like to thank Ms. Ann Marie DePetto for her assistance in preparing this

manuscript. The work was partially supported

by NCI grant CA-06519.

References

1. B. L. Berman, University of California

Lawrence Radiation Laboratory, Rep. UCRL- 78482 (1976)

2. Symposium on the Radiobiological Response

Relationships at Low Doses; Radiat. Res.

71, 1 (1977)

3. J. W. Baum, Proceedings of the IV Inter- national Congress of the IRPA, Paris, April,

1977. Paper No. 467

4. G. D. Kerr, T. D. Jones, J. M. L. Hwang,

F. L. Miller and J. A. Auxier, Proceedings

of the IV International Congress of the IRPA,

Paris, April, 1977. Paper No. 335

95
I

5. "Protection Against Neutron Radiation", NCRP Report 38, 1971

6. "Data for Protection Against Radiation from

External Sources", ICRP Report No. 21,

Pergamon Press (1971

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Installations", NCRP Report No. 31, NBS

(1964)

8. R. C. McCall, T. J. Jenkins and R. A. Shore,

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9. R. C. McCall, (private communication)

10. "Neutron Dosimetry for Biology and Medicine",

ICRU Report 26 (1977) n. A. P. Kushelevsky and G. Shani, Nucl. Instr.

Meth. 137, 71 (1976)

12. R. C. McCall, T. M. Jenkins and E. Tochlin,

SLAC Report SLAC- PUB- 1768, June (1976)

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Publication, SLAC-PUB-1782 (1976); Med. Phys. 5, 37 (1978)

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15. J. S. Laugh! in. Nucleonics 8, 5 (1951)

16. E. J. Axton and A. G. Bardell, Phys. Med. Biol. \l> 293 (1972)

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18. E. Stranden, Phys. Med. Biol. 22, 1011 (1977)

19. G. R. Holeman, K. W. Price, L. F. Friedman

and R. Nath, Med. Phys. 6, 508 (1977)

20. J. A. Deye and F. D. Young, Phys. Med. Biol. 22, 90 (1977)

21. J. G. Fox and J. D. McAllister, Med. Phys.

4, 387 (1977)

22. W. Ernst and J. Ovadia, Radiology 66, 105

(1956)

23. G. D. Adams and I. R. Paluch, Phys. Med. Biol. 10, 355 (1965)

24. M. Brenner, Cotimentat. Phys. Math. Soc.

Sci. Fenn. 31,3,1 (1965)

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28. M. Sohrabi, K. Z. Morgan, P. H. McGinley, - HPS Atlanta Meeting, Neutron Dosimetry ' in High Energy X-Ray Beams Produced in i Medical Accelerators (1977) |(

29. R. M. Wilenzick, P. M. Almond, G. D. I Oliver, Jr. and C. E. DeAlmeida, Phys. Med. jf Biol. 18, 398 (1973)

30. J. A. Devanney, Fourth Conference on i Application of Small Accelerators, North

Texas State University (1976)

|

i 31. C. J. Daniels and J. L. Silberberg, i Proceedings of IEEE, Southeastern,

| Williamsburg, Virginia (1977)

'

I 32. J. L. Silberberg, R. L. Walchie and C. J.

j Daniels, IEEE Transactions in Nuclear

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33. R. J. Schulz, Med. Phys. 5, 525 (1978) ,

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' Handbook on Calibration of Radiation

| Protection Monitoring Instruments. IAEA, Vienna, p. 28 (1971) I

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Code for Unfolding Neutron Spectra from

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;

36. K. W. Price, G. R. Holeman and R. Nath, J Health Phys. (1978)

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Nucl. Instr. Meth. 108, 199 (1973) and

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96
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M. D. Goldberg et al , Brookhaven National

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B. G. Holmes, Phys. Med. Biol. 23, 1183

1 (1978)

IBS SP 554 (1979) NEUTRON CONTAMINATION IN THE PRIMARY BEAM

P. H. McGlnley

Emory University Clinic

1365 Clifton Road

Atlanta, GA 30322

and

M. Sohrabl

School of Nuclear Engineering

Georgia Institute of Technology

Atlanta, GA 30322

In this work the absorbed dose due to neutrons was measured at points inside

and outside of a 10 x 10 cm photon beam for several medical treatment units

(Varian Clinac - 18, Allis-Chalmers 25 MeV betatron, and Brown Boveri 45 MeV betatron). The neutron dose was evaluated using a moderated activation detector,

solid state track detectors, and by the fast neutron activation of phosphorous.

For the llnac it was found that the maximum neutron dose at the beam center was

0.004% the photon dose and maximum values of 0,054% and 0.08% were observed for

the Allls-Chalmers betatron and the Brown Boveri betatron.

(Activation; detectors; dose; fast; neutrons; track)

Introduction

High energy photons are generated for use

in radiation therapy by the decelleratlon of electrons in metal targets. The radiation beams

produced by medical electron accelerators are usually contaminated with neutrons resulting

from photon interactions in target, shield,

collimatpr, and flattening filter. The dose

delivered by these neutrons in patients undergoing radiation treatment and personnel operating the accelerator is of concern to the therapist and medical health physicist. It was

for this reason that a study was initiated to evaluate the neutron dose in and near the primary photon beam using a number of neutron

dosimeters.

In Figure 1 is shown the major components

of the accelerator which may act as sources of

1 fast neutrons. All components located above

I the dotted line are found in the llnac and the

\ parts shown below this line are common to both

the llnac and betatron.

j -

[^^

I BENDING MAGNET

I

-i TA RGET

/\ FLATTENING FILTER

COLLIMATOR

Figure 1. Simplified diagram of electron accelerator.

The materials which act as neutron sources are summarized in Table 1 for the accelerators used

in this work.

Table I. Accelerator materials

Accelerator Energy

(MeV)

Collimator

Material

oeam

Fil. Mat.

Target

Mat. TargetThick,

(mm)

Clinac - 18 10 W w Cu 6.3

AC-Betatron 25 Pb Al Pt 1.6

BBC Betatron 45 W & Pb Pb Pt 2.0

Figure 2 shows a typical cross section

curve for the production of neutrons by photodisintegration of nuclei by photons.

( V, n )

PHOTON ENERGY

Figure 2. Photoneutron cross-section curve.

99
High Z materials such as Pb and W have photo- neutron cross section thresholds near 7 MeV; copper has a threshold of approximately 10 MeV. Three to six MeV beyond the threshold the photo- neutron cross section rises to a maximum and then falls off to exhibit a resonance peak.

The majority of neutrons generated by photon

Interactions in high Z materials are produced

by compound nucleus formation. The nucleus may

be pictured as heated by the photon with the result that a neutron is "boiled out" of the nucleus. The resulting energy distribution is Maxwelllan in form and resembles a fission specEl, 2] trum. This process predominates for nuclei

with atomic mass number greater than 40 and the angular distribution of emitted neutrons is

[3 4] approximately isotropic. ' To a lesser de- gree fast neutrons may be produced in high Z materials by a "direct" or "photoionization"

process, whereby the neutron acquires all of the photon's energy minus the binding energy of the neutron to the nucleus. Direct neutrons thus form a high energy tail on the energy distribution. For direct ejection of neutron the angular distribution is expected to be slightly

peaked at 90° to the direction of the incident

photon

ion causing the formation of ''-^n in its

ground state is not observed due to the short

half-life of 13 sec. A single-channel analyzer was used to detect the photon activity induced

in the indium foil by an irradiation. The

equipment consisted on an Ortec counter, an Ortec model 452 amplifier and preamplifier, an Ortec single-channel analyzer model 406, and Ortec model 456 high voltage power supply, a timer,

and a 2-in. X 2-in. Nal (Tl) scintillation

crystal. The pulse height analyzer was calibrated

with a Na source, which has a Y ray of 1.275- MeV energy. Indium-116m decays with the emission

of five photons and three Y rays of which one of the Y photons (1.293 MeV) was used for neutron

detection. After the irradiation of a foil, a delay of 5 min was taken before counting so that short metastable states of indium would not interfere with the count. Counting time (real) for all measurements was 5 min. The net count rate was corrected back to its value at the end of the exposure and also adjusted to its saturation

value as shown In Eq. (1)

(c/m)g3j, = A (total counts) Atd

. (1) e- ^t,i

MATERIAL AND METHODS

Moderated Activation Technique

The measurement of neutron dose levels in and near photon beams used in radiation therapy

requires a dosimeter with the following

properties. First, the dosimeter must be

insensitive to pulsed photon fields. Secondly,

the dosimetry system must have adequate sen- sitivity to neutrons. Any detector used in the Integrating mode should be able to detect fastneutron fluence levels of 1 X 10 n/cm^ or less

in order to keep the irradiation time to a reasonable value. Third, it would be desirable

to have a simple and rugged system.

Based on the above requirements a moderated

thermal activation detector was investigated in this research. The fast neutron component of each accelerator was measured by suspending a bare indium foil (1.69 g in mass) in a cubical

water-filled container in a 10-cm X 10-cm photon

beam. The base of the container was 15-cm X 15-cm square and it was filled with water to a height of 15 cm. Fast neutrons that enter the water moderator are slowed by elastic collisions

and some are captured by the indium foil; The

115 activation of In by neutron capture to form

116tt In was employed to detect photoneutrons. In

this reaction "'"^^In has a neutron-capture cross section of the order of 155 b for thermal neu116 trons and the In product has a metastable

state with a half-line of 54.2 min. The reactwhere

(c/m)g3^ = saturation count rate at end of exDosure (counts/min)

= decay time (5 min)

tj. = counting time (5 min).

t2^ Irradiation time (min)

A = 0.693/54.2 (mln-1)

252 A 5- pg Cf source, whose neutron emission

rate had been determined by the U.S. Atomic Energy

Commission to an accuracy of ±3%, was used to calibrate the dosimetry system. Indium foils were irradiated in the water moderator for a period of 4 h at a distance of 100 cm from the

252

Cf source. After the exposure a 5-min delay

period was observed before counting of the foil was begun. The foil was counted for a period of

5 min employing a single-channel analyzer and the net count observed was used in Eq, (1) to obtain

the saturation count rate at the end of the irradiation period. Equation (2) was then used

to evaluate the flux density of fission neutrons

at the position of the detector for the calibration exposure.

(source strength Ur)(2.4 X 10 n/sec yg)

, 2 4 TT r

(2)

100
where (() Is the neutron flux density at the

detector position (n/ctn2 sec) and r is the dis252

tance from Cf source to detector (cm). The

fadtor of 2.4 X lO^n/sec pg shown in Eq. (2) was obtained from a study performed by Stoddard

I [5] 252

j and Hootman of the Cf neutron spectrum,

j The neutron flux density established for the

j 252Cf calibration and the observed saturation

count rate were used in Eq. (3) to evaluate the

flux density of photoneutrons associated with the various photon beams.

(c/m') sat» (3)

(c/m).

'sat

where is the fast-neutron flux-denaity comiponent of photon beam (n/cm2)^^^ is the saturation count rate for foil exposed in photon beam.

Thermal-neutron flux-denaity levels were

also determined by employing a technique similar

to the one outlined above for fast neutrons. In

this case k bare and a cadmium- cove red foil were

used without the water moderator to detect

thermal neutrons in the photon beams. The bare

and cadmium-covered foils were calibrated by

placing them in a known thermal neutron field

239 produced by a 5-Ci Pu-Be neutron source

housed in a paraffin moderator of 10-cm radius.

The difference in induced activity between the

bare and shielded foils was used to obtain the thermal-neutron component of the photon beams.

The fast- and thermal-neutron flux density

were measured in the center of a 10-cm X 10-cm

photon beam at the target to skin distance

commonly employed for each accelerator (100-cm

Allis-Chalmers betatron and Clinac-18 and 110 cm

for the BBC 45-MeV betatron). Neutron flux

densities were also established for a point 5 cm

outside each photon beam. Activation of the

foils was produced by delivering photon dose

levels of the order of 1000 rad at the point of

maximum dose buildup. Irradiation times of 15-25 min were required for the two betatrons

and approximately 10 min for the Varian ClinacQ 9

' 18. A conversion factor of 3.97 X 10" cm , rad/n

suggested by Stoddard and Hootman for the

252Cf neutron spectrum was used to convert the fast-neutron fluence levels to the absorbed dose

in a small mass of muscle located in the photon

beam and 5 cm outside the beam edge. A similar

procedure was employed to establishi the absorbed

dose due to thermal neutrons based on the

difference between the activation of a bare and cadmium-shielded indium foil. A thermal neutron

fluence to dose conversion factor of 4.57 X lO"-'-^

2 r 61 cm rad/n was used to obtain absorbed dose

values.

Dose values due to neutrons in the primary

beam and in the area occupied by the patient were

also established using a second In moderated

system. The system consisted of a 2.54-cm

diameter indium foil placed at the center of a 6-in. diameter paraffin sphere encased in a cadmium shield. It has been shown by a number of workers that for this moderator the activation of

indium foil per unit flux density of fast neutrons

is fairly insensitive to neutron energy in the

range 20 keV-'20 MeV^''^. A gas-flow proportional

counter was employed to measure the activity of each foil after irradiation. Calibrations were

252 carried out with the same Cf source described

earlier in this article. Equation (3) was used

to evaluate the fast-neutrom flux density at the position of the detector. A decay time of 5 min was observed before counting of the foil was begun.

All foils were counted for 10 min and Eq. (1) was used to calculate the saturation count rate for a given irradiation.

Fast Neutron Activation of Phosphorous

Price and co-workers have developed methods

for measuring fast and thermal neutron flux

[8] density based on the activation of phosphorous.

This technique has low sensitivity to high energy

photons and should be suitable for use in intense

photon beams. Of the many reactions produced by

jieutrons and photons in phosphorous only the

following result in radioactive products with half

lives greater than 2.5 minutes.

% (n, p)^^Si

(n,Y )32p

^1/2 = 2.62 h (4)

Tj^/2 14.28d (5)

The first reaction has a threshold value of 0.7 MeV neutron energy and can be used to establish

the fast neutron fluence for a given irradiation.

On the other hand, the second reaction is caused

32 by thermal neutrons and the P activity can be

determined by counting the sample a few days after

activation.

The phosphorus was irradiated in the form

of phosphorus pentoxide (P2O5). All radioactive

products of photon and neutron interactions in the oxygen portion of the sample have half lives less

than 123 sec. The sample is then prepared for liquid scintillation counting using the method

r 8] described by Price, et al. • ;The neutron flux

density is related to the saturation activity, A (dis/sec/target nucleua) , of the sample by

101
i Polycarbonate Foil Detector

Ag = (\> (E) a (E) dE

'th

(6)

where (j) (E) Is the differential neutron flux

(n/cm2 sec MeV) and a (E) is the activatioi

cross section in (cm^/atop) . The activity

obtained when the sample ^.s counted, A^, ±3 related to A by Eq. (7)

(7)

Np (1 - e"^ '^I) (1-e - ^tc)

v^ere X is the decay constant, for the activation

product (4.408 X 10"^ min for Si), tj is the delay time from the end of the irradiation to the beginning of the count, tc is the counting time,

j ti is the irradiation time and Np is the m^er

of phosphorous atoms. The activity, At, was

evaluated from the count rate Cj- (counts/sec)

observed in a portion of the irradiated sample and

the counting efficiency of the scintilliation

detector (93%)

Based on the value nf Ag obtained from

Eq. (7) the fast neutron flux density was

calculated by

' (E) dE

'^miax

(8)

<!)n (E) a ^ p

(E) dE

where <|) j, is the total integrated flux in

n/cm sec and 4>ij (E) is the normalized neutron

spectrum. A^ value of 2.451 X 10^^ for the

second term In Eq. (8) reported by Price et al.

t®' for 20-30 MeV bremsstrahlung radiation on Ta

and Pb targetp was used in this work to evaluate

The P2O5 powder was exposed in small glass

vials at a nuniber of points inside and outside of

photon beams (10 X 10 cm) produced by the various

accelerators. After several hours time delay eiach sample was beta counted in a liquid scintillation

counter for a period of 5 minutes and (j) p was

calculated from the measured count rate (Ct).

The registration of tracks induced directly

inside poljrmers by particles^ was used as a third

method of detecting photoneutrons. The tracks

observed in polycarbonate suclj as Lexan

^^^16 ^14 ^3)n ^i^s due to recoil carbon, oxygen,

[9] nitrogen, alpha, and hydrogen atoms. Only

recoil particles causing a damage density above a threshold value can be registered. For example,

the polycarbonate foils used in this research have

a threshold value such that neutrons with less than about 0.5 MeV do not induce tracks.

Polycarbonate foils of 250 \i m thickness were

exposed in a 10 X 10 cm beam at each of the medical accelerators to photon dose values ranging

from 20 to 2000 rads. The foils were etched by use

I' [9. 10] of the electrochemical technique for a period of 4 hours at 25° C in a 28% KOH solution

applying 800 V at 2 kHz. The track densities were then obtained and converted to absorbed dose

and dose equivalent values by use of conversions

factors established in the following manner. Foils were irradiated at the various facilities

shown in Table II to neutron dose levels from

1 to 1000 rads.

Table II. Fast Neutron Sources Used in This

Study.

Source Facility Neutron Energy

at Maximum Yield

(MeV)

Fission Source HPRR

ORNL

1

5 Ci Pu-Be GA Tech 4.1

16 MeV d"*" on Be Texas A & M

Cyclotron

7

50 MeV d"^ on Be Texas ASM

Cyclotron

20

16 MeV d"*" on Be

Univ. of Washington7

35 MeV d"*" on Be Naval

Research

Laboratory

15

r

The absorbed dose and the quality factor associated

with the peak energy of each neutron beam were

used to evaluate the dose equivalent produced by

the irradiation. Based on this procedure it was

found that the neutron sensitivity (tracks/neutron)

as a function of neutron energy follows approximately the shape of the ICRU rem curve (rem/

neutron/ cm^). Therefore, the polycarbonate foil may be used to establish the dose equivalent of a fast neutron field without knowledge of the neutron energy distribution. At the indicated

102
etching conditions, a track density-to-rem

conversion factor of 105 - 7 tracks/cm'^ rem

was obtained from 1-20 MeV neutron energy. The

dose equivalent values associated with the neutron component of the photon beams was evaluated

using this factor. On the other hand, the absorbed dose associated with the neutron com- ponent of the photon beams was determined using

the conversion value obtained for the fission

spectrum at ORNL.

The photon sensitivity of the polycarbonate

foil should be very low since the authors of this

paper could find only one photon induced reaction that could lead to track production. Some

tracks might be produced by the betatron beams

as a result of the "'"^0

( y > np)'''^N reaction

which has a threshold at about 22 MeV photon

energy and a cross section of less than 0.5

nib. ^^^^ Work is now underway to establish the photon sensitivity of the polycarbonate foil to high energy photons. RESULTS AND DISCUSSION

Central Axis Dose and Fluence

Table III summarizes the results obtained

for the fast neutron fluence at the center of a

10 X 10 cm beam for the three medical accelerators. All values in Table III have been normalized to unit dose of photons at the point of maximum dose build up in a water phantom. As

can be seen the fluence values varv by as much

as a factor of 1.8 for the Allis-Chalmers betatron

and by a factor of about 2 for the other accelerators depending on the measurement technique.

Table III. Fast neutron fluence per unit

absorbed dose due to x-rays at center of a 10 x 10 cm beam

Accelerator TSD

(cm)

(n/cm^/rad photon)

In + PoO2^5 In +

H2O Mod Paraffin

Mod Activ,

1 Varian

lc-18

ACBetatron

BBC Betatron

100

100

110

1.06 X 0.55 X

10^ 10^

1.36 X 1.22 X 0.76 X

105 10^ 105

1.44 X 1.05 X 1.98 X

105 105 10^

A number of factors may contribute to the variability shown in Table III. For example,

the In plus water moderator system was calibrated

252 by use of a Cf neutron source. If the fission

neutron spectrum is not a good approximation of the photoneutron energy spectrum then corrections

may be required. Another source of error is the fact that no thermal neutron shielding material

was used to encase the water moderator and the

system may over respond in the presence of a thermal neutron field. The production of photo neutrons in the moderator material may also lead

to an over estimation of the neutron fluence.

The paraffin plus indium system should have a very low response to thermal neutrons due to the cadmLum shield. However, the cadmium as well as the moderator will act as a source of photo- neutrons when exposed in the x-ray beam. It has

[12] been estimated that an over response of approximately 22% will occur for 25 MeV photons.

The phosphorous activation method is sensitive

to fast neutrons with energy greater than 0.7 MeV.

It expected that the mean energy of the photoneutron spectrum produced by the Varian Cllnac-18

will be low and little phosphorous activation will

occur. No detectable neutron activation was observed when 20,000 rads was delivered by the Varian Clinac-18 x-ray beam. Therefore, one should not use the phosphorous activation method

for accelerators with energy of 10 MeV or less.

It should also be pointed out that the neutron

energy spectrum employed in Eq. (8) is based on

Ta or Pb targets bombarded with 20 to 30 MV x-ray

beams. As can be seen from Table I none of the accelerators utilized a Ta target and two of the accelerators operated outside of the voltage range indicated above.

In general the major uncertainty in the neutron fluence values obtained with the moderated

indium foil detectors is caused by the photon

sensitivity of the systems. On the other hand,

the phosphorous activations results are uncertain

due to the lack of adequate neutron spectral

information. Table IV list the major features of each detector employed to measure the neutron

fluence at the center of the beams. The minimum

detectable flux levels shown in Table IV are based

on a photon dose rate of 100 rad/min. , 5 minute

count time, 30 minute irradiation and a 10 minute

decay time for the indium detectors and 1 hour

decay time for the phosphorous activation.

Table IV. General features of flux detectors

used In this work

Instrument Photon Neutron Mln. Thermal

Sensitive Energy detec neutron

Depent

.

flux sensitive

(n/ cm / sec )

H2O + In yes yes 57 yes

Paraffin + In yes no 20 no

P2O5

Activation low yes 5 X 10^ no

All of the values shown in Table III were

converted to the absorbed dose due to neutrons per

rad of x-ray by use of the conversion factor for

103
the ^^^Cf fission spectrum (3.97 x 10~^ cm^ rad/

n). As can be seen from Table V a minimum

-4 value of 0.22 x 10 rad neutrons per rad x-ray

-4 and a maximum value of 7.8 x 10 rad neutrons

per rad pPujtona were oBtained.

Table V. Absorbed dose due to neutrons per rad

pc-ray at center of 10 x 10 cm beam

(rad ',n/ rad x)

Accelerator In + In + P2O5

H2O Paraffin

Mod. Moder. Activa.

Varian C-18 0.42 X 0.22 X

10-^ 10-^

AC Betatron 5.40 X 4.84 X 3.0 X

10-^ 10-^ 10-^

BBC Betatron 5.72 X 4.17 X 7.8 X

10-^ 10-4 10-^

The measurements carried out with polycarbonate foils yielded a ratio of neutron dose

equivalent to that of photons of 0.006 and 0.013

at the beam center for the 25 to 45 MeV betatrons

respectively. Unfortunately this dosimeter was insensitive to the neutron field produced by the Varian Clinac-18. This was expected since the mean neutron energy of this beam may be lower

than the threshold of the dosimeter. No correction of the dosimeter response due to photons was

made. However, work is now under way to eval- uate the photon sensitivity of the polycarbonate

foil which is expected to be very low.

The slow neutron fluence and absorbed dose

per rad of x-ray based on the activation of indium are shown in Table VI. As can be seen

the dose and fluence values are a factor of about 100 times smaller than the fast neutron

values.

Table VI. Slow neutron fluence and dose at

center of 10 x 10 cm x-ray beam

Accelerator Slow Neutron

fluence per rad x-ray

(n/cm^/ rad)

Slow Neutron

dose per rad

of x-ray

(rad n/rad x ) Varian

Clinac-18 6.8 X 102 3.1 X

10-7

AC Betatron 9.5 X 10^ 4.3 X 10-6

BBC Betatron 6.3 X 10-^ 2.9 X 10-6

Transverse Fluence Distributions

In Figures 3 and 4 are shown the neutron

fluence profiles for each betatron based on the activation of phosphorous by fast neutrons. The neutron fluence for each experimental run was

calculated using Eq, (8) and then expressed as

a percentage of the maximum fluence.

100-

90-

80-

X

70-

_i 60-

1

1

%

FAST

1

000 J 1

20- -^a 1 1- |-

10-

1 0 '

, 1'

PHOTO N BEAM

T 1 I I

1 1—I—r—

0 2 4 6 8 10 12 14 16 18 20 22 24 POSIT lON(cm)

Figure 3. Percent fast neutron

flux verses position for

10 X 10 cm 45 MV X—ray beam based on phosphorus

activation ( O = In plane

of electron orbit, = perpendicular to electron

orbit).

No corrections were made to account for changes

in the neutron energy spectrum for points not in the photon beam. As a result psf this procedure

the neutron fluence is underestimated for points

outside of the photon beam where the energy of the neutrons have been lowered by interactions

in the shielding of the accelerator.

1 00-

90-

^ 80- X 70- 3 ea

50- 40-

30-

20-

10-0

V)

1 I I

'

I—r—T—I I

' I—I

2 4 6 8 10 12 14 1 6 18 20 22 24 POSITION (cm)

Figure 4. Percent fast neutron

flux verses position for

10 X 10 cm 25 MV X-ray

beam based on phosphorus

( activation ( O " In plane

of electron orbit, = perpendicular to electron

orbit).

104
I

To Illustrate this point the effect of fllteration by 5 cm of Ta on the neutron energy

spectrum produced by 20 to 30 MV x-ray beams

impinging on a Ta target is shown in Figure 5. The neutron energy distribution shown in Figure

[13]

5 were calculated by Ford using the

evaporation model and considering only inelastic

scattering of neutrons in the 5 cm Ta layer.

As would be expected, the average energy of the

I neutron spectrum is decreased. The largest

I Increase in number of neutrons per energy channel

occurs in the range 0.5 to 1,0 MeV, However,

there is only a small increase in the lowest

energy channel (0.0 - 0.5 MeV).

If one uses Eq. (8) and the data snown in Figure 5 to determine the factor by which the

I saturation activity As must be multiplied to

I obtain the neutron flux density for the neutron

I energy spectrum before and after it enters the

I slab it is found that the value increases by a

factor of 1.57 after the beam has been filtered

by the 5 cm Ta layer. In this pase if one did

not allow for changes in the neutron energy

spectrum the neutron flux after penetra,ting the

Ta would be underestimated by 36%. It should

also be noted that due to the relative number

I of neutrons and the value of the cross-section

i for the P(n,j( ) reaction the neutron energy

I interval which is most effective in causing

activation of phosphorous is 3-3.5 MeV.

NEUTRON ENERGYIMavl

Figure 5. Croaa section for fast neutron activation verses neutron

energy and photoneutron energy

spectrum before and after passing

through 5 cm of Ta. ( after

5 cm of Ta, - before passing

through 5 cm Ta).

A well defined neutron beam whose size is approximately the same as the collimator setting

(10 x 10 cm) was found for the 45 MeV betatron.

The neutron fluence is reduced by a factor of

4 or 5 for points greater than 5 cm from the edge

of the photon beam. It should also be pointed

out that the neutron beam profile in the plane

of the electron orbit has a lower neutron fluence

outside of the beam than the profile observed

perpendicular to the orbit. A less well defined neutron fluence profile was found for the AlllsChalmers 25 MeV betatron. The neutron fluence

level outside of the beam was on the order of a factor of 2 less than that found at the beam

center.

In order to access the whole body neutron

fluence experienced by the patient measurements

were made with the paraffin moderator plus indium

foil detector. The neutron fluence was established in the plane of the treatment table with

the moderator positioned at various distances

from the central axis. Table VII presents the neutron fluence In the plane of the electron orbit

(N-S) and perpendicular to the orbital plane

(E-W) for a 10 X 10 cm x-ray beam produced by the Allis-Chalmers betatron.

Table VII. Fast neutron . fluence per rad x-ray

in plane of treatment table for

10 X 10 cm 25 MV x-ray beam. Electron

orbit is in N-S direction.

Distance from

central axis

Cdn)'

N —*S

(n/cm^/j?adT5x)

W —* E

Cn/em^/yadT- X ^

IflO 2.1 X lo:^ 2.4 X 10^

70 3.4 3.5

50 4.6 4.4

30 5.7 6.9

0 12.2 12.2

30 6.4 6.2 50 3.7 4.0

70 2.3 3.0

100 1.3 1.9

Similar measurements (Table VIII) were made

for the Varian CIlnac-18 In a direction transverse

(N-S) and radial (W-E) to the electron path in the bending magnet. The neutron fluence per unit

absorbed dose of x-ray falls by a factor of about

two as one goes from the central axis to a point

30 cm away. However, the absorbed dose due to neutrons will decrease by a larger factor due to the reduction of the neutron energy outside the

primary x-ray beam.

105
Table VIII. Fast neutron fluence per rad x-ray in plane of treatment table

for 10 X 10 cm 10 MV x-ray beam

Distance from N S WE

Central axis

(cm) (n/cm'^/rad-x)

2 (n/cm /rad-x)

100 2.1 X 10^

85 2.5 X 10^

70 2.6 3.1

50 3.3 3.3

30 4.1 4,1

0 5.5 5.5

30 3.4 3.4

50 3.0 3.0

70 2.7 2.7 100 2.1 2.1

Th.e maximum fast neutron dose received by

a 20 cm thick patient treated to 5000 rads tumor dose through parallel opposed 10 x 10 cm ports

(1:1) has been estimated using the paraffin plus

In foil fluence values shown in Table VII and VIII. The percent depth dose due to x-ray for each accelerator was utilized to calculate the given dose for the ports. A fission spectrum

was assumed for the neutron beam and the neutron depth dose data in reference [14] was used to evaluate the maximum neutron dose. Table IX presents the maximum dose at the central axis as well as the neutron dose 30 cm

iiransverse to the central axis.

Table IX. Neutron dose for a typical treatment

of 5000 rads tumor dose

Accelerator Maximum Maximum fast

fast neutron neutron dose

dose at central 30 cm from

axis central axis

(rad) (rad)

Varian

Clinac-18 0.081 0.062

AC-Betatron 1.60 0.90

BBC-Betatron 1.30 0.52

The dose 30 cm from the central axis

should be viewed as an upper limit due to the assumption of a fission spectrum. Inelastic

scattering of neutrons in the shielding material

will lower the average energy of the neutron beam

and a larger fluence to dose conversion factor

will be required. For example, it has been

reported ^^^^ that 15 cm of lead shielding will

252 reduce the average energy of the Cf fission

spectrum from 1.8 to 1.0 MeV.

Acknowledgement - We wish to thank

Mr. Russell Marian who assisted with the experimental measurements and the staff of Atlanta West Hospital for the use of the BBC

45 MeV betatron.

References

[1] Dose Levels Due to Neutrons in the Vicinity

of High-Energy Medical Accelerators,

P. H. McGinley, M. Wood, M. Mills, and

R. Rodriguez, Medical Physics 3, 397 (1976).

[2] Measurement of Fast Neutrons Produced by

High-Energy X-ray Beams of Medical Electron

Accelerators, R. M. Wilenzick, P. R, Almond,

G. D. Oliver, and C. E. DeAlemida, Phys. Med. Biol. 18, 396 (1973),

.3] Energy and Angular Distributions of Photoneutrons Produced by 70-MeV X-rays,

W. R. Dixon, Canadian Journal of Physics

33, 785 (1955).

[4] Photoneutrons from Lead, M. E. Toms and

W. E. Stephens, Physical Review 108, 77 (1957).

[5] 252

CJf SBieldlng Guide, D. H. Stoddard and

H. E. Hootman, U. S. AEC Rep. DP-1246 (1971),

[6] Protection Against Neutron Radiation, NCRP Rep. No. 38 (1971).

[7] Fast Neutron Surveys Using Indium Foils

Activation, L. D. Stephens and A. R. Smith,

U. S. AEC Rep, UCRL-8418 (1958).

[8] A Technique for Determining Fast and Thermal Neutron Flux Densities in Intense HighEnergy (8-30 MeV) Photon Fields, K, W. Price,

G, R. Holeman, and R. Nath, Health Physics

35, 341 (1978).

[9] Electrochemical Etching Amplification of Low LET Recoil Particle Tracks in Polymers

for Fast Neutron Dosimetry, M. Sohrabi, Ph.D,

Thesis, Georgia Institute of Technology

(1975),

[10] The Amplification of Recoil Particle Tracks

in Polymers and its Application in Fast Neutron Personnel Dosimetry, M. Sohrabi,

Health Physics 27, 598, (1974).

106
[11] Atlas of Photoneutron Cross Sections

Obtained with Monoenergetic Photons,

B. L. Berman, Preprint UCRL-78482 (1976)

[12] High Energy Photon Response of Moderated Neutron Detectors, R. C. McCall and

T. M. Jenkins, SLAC-PUB-1768 (1976).

[13] Private Communication, J. Ford, Atlanta,

GA (1977).

[14] Neutron Dose, Dose Equivalent, and Linear

252 Energy Transfer From Cf Sources, T. D. Jones and J. A. Auxier, Health Physics

20, 253 (1971).

[15] Accelerator Health Physics, H. W. Patterson

and R. H. Thomas, Academic Press, New York,

p. 310 (1973).
la

i.

\
NBS SP 554 (1979)

NEUTRON PRODUCTION FROM ELECTRON ACCELERATORS

USED FOR MEDICAL PURPOSES

E J Axton and A G Bardell

National Physical Laboratory

Teddington, Middlesex

Measurements have been made of the photoneutron production in and out of the

photon and electron beams produced by the SLT5-20 linear accelerator installed at Addenbrookes Hospital, Cambridge. The in-beam neutron dose-equivalent to the

patient per treatment rad in water is estimated at k.3Q mrem in the photon mode

and O.2U9 mrem in the electron mode. The present results confirm less accurate measurements published earlier by the same authors, based on a prototype

accelerator of this type. Similar measurements on a 33 MV Brown Boveri betatron

published at the same time should be reduced by 25 - 30^ to obtain the neutron

dose-equivalent to the patient from externally generated neutrons , although the

published values were probably a truer representation of the total patient

dose-equivalent

.

(Dose-equivalent: Electrons; Kerma; Mtiltisphere ; Photon; Photoneutrons

;

Radiotherapy; Spectrometry).

Introduction

?eYeral years ago some preliminary measure-

„

'' ' were made on the Brown Boveri 35 MV

betatron, at St Luke's Hospital, Guildford, to determine whether a significant whole body

neutron dose-equivalent would be received by a patient undergoing radiotherapy with either X-rays

or electrons. Further measurements were made on a prototype M.E.L. Equipment Co. Ltd linear

accelerator operated at I6 MV. The method of

multi-sphere spectrometry was used for the

measurements, which were then at an early stage of

of development at NPL.

Published curves^^^ giving the shapes of the response for each sphere size as a fimction of neutron energy were adopted and these were

normalised by calibrating two spheres (8 inch (a) and 12 inch diameter) with monoenergic neutrons

of 0.511 MeV and 1.22 MeV energy from the NPL

Van de Graaff accelerator using the '^Li(p,n)

I reaction. The same two spheres were then

I irradiated both in and out of the collimated X-ray

or electron besjn, and a crude estimate of the mean

neutron energy was taken as that which predicted

the measured ratio of response for the two sphere

sizes. In-beam dose-equivalent rates of about

0.5 per cent of the central axis X-ray dose rate were obtained with both accelerators, the results

for the betatron in the electron mode being a factor of about 30 lower.

j With more detailed knowledge of the sphere

I responses as a function of neutron energy, and

' the use of more sphere sizes, it was thought to

be worthwhile to repeat these measurements in

^ order to obtain an estimate of the neutron energy

spectrum from which a more realistic estimate of

the neutron dose-equivalent could be obtained.

Unfortunately the Guildford betatron is no longer

' operated in the X-ray mode. It turned out that

! the only machine which was available for measure-

\ ments in time for this -conference was the 16 MV

M.E.L. Equipment Co. Ltd SL 75/20 linear

accelerator installed at the Addenbrookes

Hospital, Cambridge, England, the higher energy

machines being available only in the electron

mode. The results obtained with the Cambridge

accelerator which are described below, tend to confirm the earlier measurements

.

The Detection System

The detectors consist of a set of six polyethylene spheres of density 922 kg.m~3 each

containing a thermal neutron detector in the form

of a gold foil 1 cm^ in area and 100 mg-cm~2 in thickness. Gold foils were chosen rather than a dynamic counting device as they are completely

stable and reproducible, independent of electronic

drifts, and suitable for use in pulsed beams.

Furthermore the response to photons is zero below

the threshold for photoneutron production at

8 MeV. Above this threshold photoneutron

production leads predominantly to ^96au which is easily separated from the ^98au produced by (n,Y)

reactions by its half life. The major drawback

of these detectors for some applications is their

low sensitivity, and poor spatial resolution.

The sphere efficiency is defined as the

saturation disintegration rate per mg of gold

(equal to the neutron capture rate per mg) for

unit neutron fluence rate at the centre of the

sphere in the absence of the sphere. Six sphere

sizes were used with diameters 5-1. 7.62, 12.70,

20.32, 25. i< and 30.1*8 cm (2, 3, 5, 8, 10 and 12 inches ) , *

* As the spheres were constructed to exact inch diameters, reference to these

spheres in this paper will be given in inches

.

109
Sphere efficiencies have been calculated ty a Monte Carlo technique in which the energy scale is divided into logarithmic bins eight to a decade,

and using the hydrogen and carbon neutron cross section data obtained from the European data centre

in T9T2. The efficiencies were computed for energies at logarithmic intervals five to a decade.

The programme also records, for each energy, the

thermal neutron fluence and the Westcott epithermal flux parameter r which enable the relevant efficiencies to be obtained when other thermal

neutron detectors are placed at the centres of the

spheres, provided that they are not 'black'. The Monte Carlo calculations did not extend above

5 MeV because the computing power available was insufficient to cater for the inelastic scattering

reactions in the carbon. There are now several

sets of sphere response computations published in the literature which show general agreement in the

shape above 5 MeV, although there are wide

differences at lower energies. The shapes of

(2) reference were therefore used to extend the

efficiency calculations by a further two bins

(mid points 6.3 MeV and 10 MeV). Thus there are

h^ bins covering the energy range from 0.08 eV to 12.6 MeV, with an additional bin for thermal neutrons at 0.0322 eV, the mean energy of a Maxwellian distribution characterised by a temperature of 20 C. In order to test the

calciilations the five smaller spheres were calibrated in the NPL thermal column, and the four

smaller spheres were also calibrated with a standard Sb-Be neutron source (mean energy 22.8 KeV).

These calibrations were in agreement within a few

percent with the Monte Carlo calculation. However,

the calibrations referred to in the introduction

indicated that the calculated results might be

about ^k% too high at 0.511 MeV and 20^ too high

at 1.22 MeV. Further calibrations of the 8 inch

sphere at 0.116 and O.565 MeV confirmed the

apparent energy-dependent trend. It is very

important that the detector efficiencies should

not be over-estimated since this would eventually

lead to an under-estimation of the neutron radiation levels. Accordingly efficiencies

above 25 keV were adjusted downwards by an energy- dependent correction amounting to 25^ at 10 MeV.

The correction is given by the expression, (i > 28).

e'. . = e. . exp((i - 28) x 0.0221U) (I)

where e. .is the efficiency for sphere in the

.th ^"^ ^.

1 energy bin.

As an example of the effect of this correction, the estimated central axis dose-equivalent

rate per rad * of in-beam dose was increased by

18^ in the X-ray mode and by 20^ in the electron

* The SI unit of absorbed dose is the

gray, equal to 100 rads. However the latter special unit, the rad, has been

retained in this paper for ease of comp- arison with previous references.

Similarly the rem has been retained in place of the SI unit of dose-equivalent,

the sievert, equal to 100 rems.

110

mode by the use of the modified efficiency matrix.

This difference may be regarded as an overall limit

of estimated systematic error in the measurements

of dose-equivalent. The modified efficiency

matrix is shown in appendix 1 , together with the

modified thermal neutron fluence and the Westcott epithermal flux parameter r.

The Measurements

Before planning the measurements it is convenient to consider the type of neutron spectrum

likely to be encountered. From theoretical

considerations the primary emitted spectrum is likely to approximate to a Maxwellian distribution

characterized by a certain 'temperatiare' T(MeV)

shown as a dotted histogram in figure 1 . The primary spectrum will be degraded by scatter in the target, collimator assembly, beam-flattening

filter etc., and the degraded beam will then

interact with the detectors. The degraded beam

will also go on to interact with the walls, floor

and contents of the room to produce a well moderated 'l/E' shaped spectrum culminating in an approximation to a thermal Maxwellian characterised

by another ' temperat\ire ' T^. The room scattered

background will then interact with the detectors

so that the overall spectrum interacting with the

detectors will look something like the solid line

in figure 1

Although most of the contribution to doseequivalent and tissue kerma will come from neutrons

above, say 100 keV energy, a significant pro- portion of the fluence will be in the lower energy region and will contribute to the detector

activation, particularly in the case of the smaller spheres. In order to interpret the measurements it is convenient to divide the spectrum into two parts at the energy of the cadmium 'cut off

just below 1 eV. In other environments this is easily achieved by enclosing the detectors in a cadmium box. In the present situation this would

not be wise because photoneutron production in the cadmium would interfere with the field.

However, the problem can be overcome by

irradiating metallic foils, bare and under cadmium,

to measure the thermal neutron fluence at the point of interest. Furthermore, the cadmium

ratios give a measure of the epithermal fluence

per unit lethargy interval at the energy of the principal capture resonance of the foil.

It is then possible to calculate the responses of the spheres to neutrons below the cadmium

cut-off energy and subtract it from the measured

response before carrying out the fast neutron

analysis. The correction is small except for the 2 and 3 inch spheres (20^ and Q% respectively).'

The procedure for these thermal calculations

is an extension of the method of Westcott ^3) and is described in appendix 2. The treatment - is strictly applicable only in well moderated

environments since activation integrals are evaluated in l/E shaped spectrum, and the

p presence of a preponderance of higher energy i| neutrons will give undue weight to higher energy '[ resonances in the foil. On the other hand, the - resulting errors in the low energy fluence will ' have little effect on the ultimate derived dose- >
3 X 10 -

8 X 10

1 X 10

Tot*l.

Past and thcrml Maxwellians.

Non - thermal scatter.

0.001 0.01 0.1 —I—

10 100 1k

1 r

10k look 1M 10M ev

Figure 1. Maxwellian Spectrun Model.

equivalent or tissue-kerma rates.

For the X-ray measiirements each sphere was

irradiated in turn on the central axis (C)of the

beam at a distance of 103 cm from the target.

The accelerator was operated at a peak energy of

16 MeV and irradiations were of 10 to 20 minute

duration at dose rates of 600 to itOO rad.min"^.

The rad in this context is a rad in water, measured

at a depth of 5 cm in a water phantom the front

surface of which was at 100 cm from the target,

normalised to the maximum of the depth-dose curve

at 3 cm. Simultaneously, two other spheres were

irradiated at distances of kO cm from the central

I sphere on either side (L and R) of it, on a

I horizontal line perpendicular to the beam axis, the

] left (L) spheres being towards the accelerator

gantry, as shown in figure 2. The field size was

30 X 30 cm so that the edge of the largest

(12 inch) out-of-beam sphere was about 10 cm clear

of the beam, and a few millimetres at the edge

of the in-beam 12 inch sphere were not irradiated.

The electron irradiations were performed

with the centre of the in-beam spheres 11.5 cm

from the face of the largest applicator available

(20 X 20 cm) giving a window-to-sphere-centre

distance of 107 cm. The accelerator was operated

at 20 MV, and irradiations were of 30 to 20 minutes duration at dose rates of 300 and UOO rad.min"^. The rad in this case is a rad in water

at 1.5 cm depth in a water phantom with the face

of the phantom placed against the exit aperture of

the applicator, i.e. 10 cm nearer the electron

window than the centre of the spheres. A measurejment of the dose rate at 1.5 cm depth with the

face of the phantom 10 cm from tlie apprlcator

(i.e. at sphere centre position) indicated a reduction of O.836 in the dose rate. No correction

has been applied for this difference in distance.

Again two other spheres were irradiated at the

same time and in the same relative positions as in

the X-ray case. The applicator size of

20 X 20 cm results in parts of the edges of the

10 and 12 inch spheres not being irradiated,

however, since these regions only contribute a small percentage to the gold foil activities

produced, no corrections were made for this effect.

The X-ray and electron irradiations were performed

with the central beam axis parallel to the floor

and at a height of 123 cm. The left (L) spheres were irradiated in 1 mm thick cadmium boxes.

All spheres were supported on expanded polystyrene

foam above the treatment table which was not exposed to the direct beam in either case. A

comparison of the L and R results thus gives a measure of the effectiveness of the thermal

subtraction.

The simultaneous irradiation of three spheres,

which was necessary in order to accomplish the measurement programme in the time available,

introduces some small partially self-compensating

aberrations. When there is a large sphere nearby

it can shield another sphere from part of the scattered radiation, but at the same time it can

act as an additional source of scattered neutrons.

The latter effect would be most important when an out-of-beam small sphere is close to an in-beam

large sphere.

Ill
Electron.

? Inch sphere. I )- (L) in Cd.

2 Inch sphar*.

3 inch sphere.

(L) In Cd.

Figure 2. Radiation Geometry.

198 ... The Au activities of the gold foils were

counted in the NPL anticoincidence low background

counter which has a background of about O.Ok s~^

Foil g-counting rates varied from 2 to UO s~^ in the

X-ray mode from 0.1 to 2 s~1 in the electron mode.

Interference from ^^dau was easily separated out by half life. Other aspects of photoneutron

production in the detectors will be discussed in a later section.

Derivation of the Neutron Spectra

After subtraction of the thermal response to neutrons below the cadmium cut-off energy each

irradiation of the set of six spheres yields a

set of 6 disintegration rates . 36 energy bins

cover the energy range from exp (-(in 10)/10),

(0.8 eV)_to 10° exp ((in 10)/10), (12.6 MeV). The problem is to determine a set of energy bin

fluences which satisfy the Fredholm equations

y ^. e. . 3. ± AB.

0 0 (2)

Where e.. is the sphere efficiency as previously —'— "'"^ th defined for the j sphere at the mean energy of

the i*''^ bin and B. is the saturation count rate

th of the gold foil in the j sphere.

With six measurements and 36 unknown bin

fluences the equations are underdetermined and

therefore have an infinite number of solutions.

The task is to introduce constraints which lead

to particular unique plausible solutions. There

are several methods by which this may be accomplished, some but not all of which require some pre-knowledge of the spectrum. Having thus

determined a set of ii^ the expected disintegration

rate C

(=1 \b . e . . ) can be calculated and the

1 ij

quantity s, given by [I (1 - C

(where n is the number of spheres, in this case

6) is then a measure of the goodness of fit, and

it determines the random uncertainty attributed

to the quantities evaluated from the

Four methods will be mentioned as examples.

1 . Choose a mathematical representation of the

spectriim which is determined by a few un- known parameters which can be determined in a least squares analysis.

2. Minimise ^ (ii. - F. ) where is an initial

guess. i

112
The above methods require pre-knowledge of

the shape. However, for the following two methods it is only necessary to make the assumption that the spectrum is smooth:

3. Minimise I ('i'^+i ~ *

^i-']^

' i minimises the d.eviation from a straight line

of every set of three consecutive points.

1+. Minimise I ^^±+2 ~ * ~

'''i-i^^- "^^^^

minimises the deviation from a quadratic

through every set of four consecutive points.

It generally turns out that different methods

produce different spectrum shapes, but all

shapes which fit the data yield the same total dose-equivalent rate, and 'the same dose-equivalent ratgs integrated over a few

broad energy bands . This implies that

the system is primarily measuring dose or dose equivalent, rather than fluence.

In the present application it was found that

good agreement with the measurements could

readily be obtained by method 1 . The first step

is to establish a set of energy bin fluences

ilmi. from a Maxwellian distribution characterised

1 by an arbitrary temperature T, namely

1 n(E) a exp (-E/T) , where n(E) is the

probability per xinit energy interval of the

emission of a neutron of energy E MeV. The tfim^, which are shown by the dotted line in figure 1

,

are then used to calculate a set of sphere responses Bm..

an. = y ijim. e. . (3)

The t|)m^ are also used to calculate a set

of scattered bin fluences ij^s, by sharing each

i|mi^ equally between all the bins of lower energy.

The ^s^ are then used to compute a set of

sphere responses Bs

.

Bs.

It is assumed that the true bin fluences

are a weighted sum of tjim^ and i|)s^

^1 1 1 2 1 (5)

where and Cg are unknown constants to be

determined by a least squares fitting procedure

from observational equations such as

B. = C, an. + C„ Bs.

J 1 J 2 J (6)

In other words, by minimising

It is assumed that the B. have equal fractional

J uncertainties, so the weights W. are taken as

2 proportional to l/B. • The fit gives the best

J values of and together with their

uncertainties and covariances, determined from

the root mean square residual s given by

^(B.-C^ an.-C^Bs.) /(6-2)

The \li . so derived can be converted to dose

1 equivalent by means of dose fluence curves

to kerma by kerma fluence curves (6).

(7)

(5) and

The temperature T is then varied to obtain

minimum s. Figure 3 shows s as a function of T

for the three X-ray cases and the three electron

cases, for which the B.'s are shown in table 1.

J The dotted curves were obtained with the modified

Monte Carlo matrix. Unambiguous minima are obtained with the three X-ray cases, and the resulting

value of T is the same for the modified and

un^nodified Monte Carlo matrix and for L and R positions. Also included in figure 3 is the

dose-equivalent rate calculated for the in-beam

X-ray case. The dose-equivalent is a slowly

varying function of T. Values of 0.6 and 0.25 are obtained for the in-beam and out-of-beam X-ray

temperatures, and the random uncertainties are of the order of 1.5-3.5^ of the dose-equivalent rate.

The electron cases present a slightly different picture. The uncertainties are higher, partly

because the fluence rates are lower by a factor

of 25, but there are probably other factors

contributing to the poorness of the fit.

Table 2 shows the fluence, dose equivalent

and kerma for the six sets of measurements and the

totals are broken down into five broad energy bands.

In all cases the results are in reasonable

agreement with the values published earlier in reference (l) although these were for different

machines.

In the X-ray case, 30% of the neutron fluence

below 100 keV contributes less than 3% to the dose

equivalent. This demonstrates that fluence is not

a good quantity to use for comparisons between

different authors. In the electron case the

accelerator was operated at the higher energy of

20MeV^(21.5 in vacuo). The neutron energies eire higher, and the contribution of neutrons below

100 keV is even lower. There is some evidence

that the L dose rates eire higher than the R dose

rates, more so in the case of the electrons. The

L position is nearer to the main part of the

accelerator, and it could be that the latter forms

additional sources. A study of the electron energy spectrum of this accelerator shows the presence

of a secondary low energy peak in the energy distribution of electrons entering the 96 bending

113
Figure 3. rms Residuals as a Function of Energy.

TABLE 1

Measxired response (B.) bq. mg ^ normalised to a dose rate of

J

1 rad in water per minute

2" 3" 5" 8" 10" 12"

X-ray C 0.3930 0.7759 1 .3630 1.0999 0.61*05 0.3720

X-ray L 0.3161* O.616I+ 0.9003 0 . 531+5 0.2811 0.1287

X-ray R 0.2602 0.5829 0.7721 0.5213 0.2711+ 0.1222

Electron C 0.00825 0.02389 0.06028 o.oUitio O.0370I+ 0.02293

Electron L 0.01252 0.02509 0.02359 0.01751+ 0.01200 0.00610

Electron R 0.00581* 0.01276 0.02555 0.02139 0.01286 0.00675

114
'magnet. This low energy component will be

deflected in the magnet so as to strike the bending

! chamber walls and produce a secondary source to the

left (using the same convention for the sphere

positions). Neutrons produced in this region

would be scattered predominantly towards the

accelerator waveguide where there is less shielding. Secondary sources could also be produced

where the electron beam impinges on any beam

defining aperatures in this region. Beam wander

I would cause this source to be unstable. These

' soiirce s would be relatively more important in the

electron mode. Furthermore, such sources would

' worsen the aberrations caused by mutual shielding

amongst simultaneously irradiated spheres. These

effects would all worsen the goodness of the fit.

Alternatively, perhaps, the neutron spectrum in

the electron mode has a different shape and could

be fitted better by a different spectrum.

Temperatures of 0.9 MeV and 0.6 MeV for the C and R positions were obtained.

The L position fit did not give an unambiguous minimum possibly due to the secondary

sources mentioned above. The spectra were also

determined using methods 3 and k described above

which require no pre-knowledge of the spectrum

other than that it is smooth. As an example, the

derived spectra are shown in figure k for the

X-ray in-beam case. Methods 3 and k show a broader peak, and a different, possibly less

plausible low energy shape, but the dose-equivalents obtained by these, cruder, fits are within

20% of those obtained by Method 1 , as shown in Table 2.

Photoneutron Production in the Detectors

The reactions to be considered are photo- neutron production in deuterium, gold, and carbon.

Photoneutron reaction rates can be calculated

readily from the information available in the

-IAEA, Hand.bo<^ on Nuclear Act-ivation Cross

.(T) Sections This reference contains a comp500

*oo7 300

200-

100-

Degraded Maxwelllan Temp. - 0.6 MeV Method 1 3 PaLrameter smoothing fit. Method 3. k Parameter smoothing fit. Method U.

• Flux froB Gold resonance

A Flux from Manganese resonance

Method 3 minimises

Method h minimises

Dose - equivalent from : Method 1 - i».38 .mrem.rad

Method 3-3.56 mrem.rad

Figure h. Spectra Produced by Different Fitting

Methods

.

115
TABLE 2

Neutrons per rad of dose in the Treatment Beam

(a) X-rays

Energy eV Thermal

% Cd - lOK% 10K - 100K% 100K - 1M

• %

> 1M

% Total Uncertainty

% Units

IN BEAM

Fluence 5.80 Dose-Equiv. 0.36 Kerma 0.08

16.77

1.05 o.^k

7.25

1 . 10 1.87

1+5.67

1+7.79 50.18

2I+.51

1+9.70

1+7.73

2.1*1 X 105

1+.38 X

10~^

3.53 X 10

1 .20

1 .71+ 1.75

-2 ^-1 cm . rad

rem.rad"^

_ 1 rad. rad

OUT OF BEAM - RIGHT Fluence lO.Ul Dose-Eq\iiv. 1.12 Kerma 0.23

16.55

1.79 0.2h

13.79

1+.03 6.22

55.71 80.99

83.00

3.66

12.06

10.31

1.1+1* 10^

1.1+97 X

10~^

1.32 X 10

2.10

3.25

3.31+

-2 -1 cm .rad

rem. rad ^

1 rad. rad

OUT OF BEAM - LEFT

Fluence 8.96 Dose-Equiv. 1.03 Kerma 0.2020. U7 2.3U

0.30

lU. 13 U.23

6.56

52.97

80. U8 82.73

3.1+6

1 1 .92 10.21

1 .55 X 10^

1 .51+ X

10~^

-1+

1 .35 X 10

0.97

1.57

1.62

-2 -1 cm .rad

rem. rad ^

rad. rad ^

(b) Electrons

IN BEAM

Fluence 5 . 78 Dose-Equiv. 0.k2 Kerma 0.105.1*7 0.25

O.OU

3.96 0.k9

0.82

1+3.77 35.91 37.02

1+1 .03 62.92

62.02

9539

2.1+9 X lO"'* 2.01 X

10~^

5.26

6.1+6

6.1+7

-2 .-1 cm rad

rem. rad ^

rad. rad ^

OUT OF BEAM - RIGHT

Fluence l8.5'+ Dose-Equiv. O.7O Kerma 0 . 1

10.62

0.71 0.10

5.59 0.95

1.61

U2.20

1+7.59

1+9.97

23.05

50.05

U8. lU

508U

8.61+ X

10"^

6. 91+ X

10~^

1.97 2.70

2.71

cm ^.rad ^

rem. rad ^

rad. rad ^

OUT OF BEAM - LEPT Fluence 12.63

Dose-Equiv. 1.09 Kerma 0.25

36.03

3.16

O.hl

9.33

1.65 2.8k

19.90 28.61 29.73

22. 10

65.1+9 66.77

1+596 6.81 X

10"^

5.56 X

10~^

1+.36 7.50 7.6U

-2 -1 cm .rad rem. rad ^

rad .rad~^

ilatioh oT photoneutron cross sections^ as- -w&il- as photon spectra as a function of accelerator

energy. The photon spectra 4i{k.,k ) are proportional to the energy fluence (MeV.cm ) in unit energy intervals at energy k^(MeV) from an accelerator with a maximum photon energy of k^.

The reaction rate per g of material exposed

to this fluence is:

I I fCk-.k^) o. (y, n + np + 2n + ....)/k.

i where N is the Avogradro constant and A is the

atomic weight of the material.

The absorbed dose delivered by this energy

fluence is given by:

1.602 X 10~®

I )(l+en/P^ water

""^^^ ^"^ '^^^^'^ i Where y /p is the mass energy transfer

en

p (8) coefficient (cm .g ) for water and the initial constant is the conversion factor from

MeV.g"^ to rads. The ratio of these two summations gives the reaction rate normalised

to an absorbed dose of 1 rad in water.

The total neutron production is obtained by

multiplying each partial cross section by the number of neutrons produced per reaction and the production rate for any given product nucleus

is obtained by considering just the partic\ilar

partial cross section.

116
I The neutron production in g . rad~ has

; been C£ilculated using this recipe for a n\imber of

! elements and accelerator energies. For the higher

energies it was necessary to extend the cross , > section curves on the basis of the Lorenz shape^

and apply appropriate neutron multiplicity factors.

The resiats which are shown in table 3 therefore

only give a general guide to the manner in which the

j neutron production varies with accelerator

j energy.

j Measurements with a 16 MeV accelerator

The carbon effect in this case is zero. The

2 total H(Yin) neutrons produced in each sphere

is (from table 3) 1+ x 105 x mass of the sphere,

• • 3 which is proportional to R . This may be compared

I with the number of neutrons entering the sphere

from the accelerator sxirroundings which is given

2

I by ffR X fluence (table 2). The number of neutrons

I produced in the sphere as a function of the number

of neutrons striking the sphere is thus proportional to R. In the worst case (12 inch) the ratio is less than 0,0^%.

The gold problem is treated differently. It has already been stated that the ^ Au activity

has been separated from the ^^°Au activity by

half-life. The ^'^°A\x saturation counting rate in the

e-counter was nearly the same for all sphere sizes,

amounting to about O.Ok and 0.0036 .mg for the

X-ray and electron cases respectively, normalised

to a dose rate of 1 rad.min *

. From table 3 the

^^^Au saturation disintegration rate mg ^ (equal to

the reaction rate) normalised to 1 rad.min ^ is 6.8 X 1o'*/60 000 or 1.1 s~'.mg~\ The observed

3-count rate of O.O^t s .mg~ implies an efficiency for the detection of ^9°Aa decay of

0.036, in good agreement with estimates. On this

196 basis the Au saturation disintegration rate in

the electron mode would be 0.1 s ^ .mg~^

.

The Monte Carlo Method has been used to calculate the conventional thermal neutron fluence

rate nv^ created at the centre of the spheres by

a source of 1 MeV neutrons of strength Q s

^ situated at the centre, nv /Q varies from O.OOU

o for the 2 inch sphere to 0.022 for the 12 inch

sphere. The ^^o^u reaction rate due to such a thermal neutron fluence, expressed as a fraction

of the ^98au reaction rate, B:, shown in table 1 J for both the in-beam X-ray and election cases, is less than 0.01^. It can be concluded that the measurements with the I6 MV accelerators are

free of any distortions due to photoneutron

production in the detectors.

Measurements with a 33 MV accelerator operated

in the X-ray mode

Although the neutron production rate for gold,

(see table 3), is about 30% higher than for I6 MV

198 the resulting enhancement of the Au activity

would still be negligible. In the Guildford

experiment^ ^

^ the Au activity was not separated

out by half-life, although it is still possible

to do this. However the amount of '^°Au produced

would be significantly lower, as at the higher

accelerator energy the situation would be

dominated by the ^^'^Au(y, pn) reaction leading to

TABLE 3

Neutron production g~\rad~^ as a function of accelerator energy

MeV 16 22 27 32 36 25

k.O X 10^ 3.5 X 10^ 3.2 X 10^ 2.9 X 10^ 2.7 X 10^ 3.U X 10^

C 0 7.0 X 10^ 5.8 X 10^ 7.8 X

10-^ 8.3 X 10^ 3.9 X 10^

W 7.6 X

10**

1 .2 X 10^ 1 .2 X 10^ 1.1 X 10^ 1.1 X 10^ 1 .2 X 10^

Fe 6.3 X 10^ h 2.9 X 10

h 3.7 X 10 3.8 X lo'* 3.8 X lo'*

1. 3.U X 10

Au 6.8 X 10^ 1 .0 X 10^ 1 .0 X 10^ 1 .0 X 10^ u 9.5 X 10 1 .0 X 10^

Cd 3.h X lo'*

It 7.3 X 10 7.8 X

10** 7.6 X

10'* u 1.2 X 10 7.5 X 10

117
stable 195Pt, aoad the

18' 196

"197, leading to 185 d^^^Au. Au ,(yi 2n) reaction

The ppe'sence of a small amount of '^^Au in the foils would lead to an

over estimation of the neutron dose-rates but

such an over estimation is thought to be fairly

small.

Table 1 indicates also that photoneutron

production from deuterium would be negligible in

the Guildford experiment.

However the same cannot be said of carbon.

. (9) Kushelevsky and Sham calculated the neutron production rate in the carbon in an 8 inch polyethylene sphere of unit density as being

a factor of approximately 2.5 higher than the

number of neutrons impinging on the sphere from

a 2U MV accelerator. -Their calculation was based on the photon spectra of Johns, Katz,

Douglas and Hasham^^ which are not inconsistent

with those of reference (7)« However, their

calculations contained two important errors. The

first of these was a factor of 60 due to a misinterpretation of the statement of fluence

(2960 cm . s for 1 rad.min ) as being the

integrated fluence cm~2.rad~^. The second of

the errors, in the calculation of the total

number of neutrons striking the sphere, was the

multiplication of the fluence by the surface area

of the sphere (UgR^) instead of by the cross section area (ttR ). Thus the fractional neutron

production in the sphere becomes 2. 5/ 15 = 17%. The first of these errors has been corrected in

a letter^^^^ where the fraction was re-estimated

as h.3%. The second error was not detected at that time so the revised figure of h.3% has now

to be increased by a factor of k to ^^%.

(9) But there is more. The calculation was

for a 2U MV accelerator. The carbon photoneutron production for a 33 MV accelerator would

be a factor of 2 higher, and with the cross section data used in table 3 the fraction could

possibly be as high as The conclusion is that it would not be unreasonable to reduce the

in-beam neutron dose-equivalent rates of

reference (1) by 25 - 30% (to 0^% of the

X-ray dose and to 0.015% of the electron dose)

in order to obtain the dose-equivalent to the

patient delivered by neutrons generated in the

Guildford accelerator.

Field size dependance

The measurements which have been described

were carried out with a field size of 30 x 30 cm. In one case the measurements were repeated with

the field size reduced to 20 x 20 cm and

10 X 10 cm. The results shown in figure 5 show

a linear decrease with field area for the 3 inch

in-beam sphere, amounting to 11%. The out-ofbeam (R) 2 inch sphere response remained

approximately constant, whilst the out-of-beam

(L) 8 inch sphere showed a linear increase amounting to about llt%.

Conclusions

The multisphere spectrometry system has been

used to measure neutron dose-equivalent and tissuekerma rates in in-beam and out-of-beam positions

associated with the 16 MV linear accelerator at Cambridge. The systematic \incertainty is estimated at + 20% at the 99% confidence level and

is attributed to uncertainties in the response

matrix. Random uncertainties at the 68% confidence

level are shown in table 2 as derived from the

IM

1.0-

0.8-

0.6-

O.k0.2-

Z - rays.

3 Inch In beam.

8 Inch In Cd out (L).

2 Inch out (R).

1

1 10 X 10 20 X 20 30 X 30 Field size cm^

Figure 5. Effect of Field Size oit Detector

Response.

118
fitting procedures. The results indicate that

the earlier published results obtained with a prototype linear accelerator at 16MV were not seriously in error.

At higher energies, the photoneutron

production in the carbon of the spheres is not insignificant. It is now realised that the earlier

published results for the 33 MV Brown Boveri

betatron should be reduced by perhaps 25 - 30% in order to obtain the true tissue equivalent kerma

rates and dose-equivalents for the neutrons

produced in the accelerator. On the other hand

the carbon and oxygen (y.n) cross sections are not dissimilar. The carbon in the spheres could

therefore be regarded as simulating the oxygen in tissue, and the published results would then be a truer representation of the total dose-equivalent

to the patient

.

1 1 . Photoneutron Production in Neutron Detectors

in High Energy X-ray beams. E. J. AXTON

A. G. BAKDELL. Nuclear Instrument and Methods

153 563 (1978).

12. Absolute Measurement of the Thermal Neutron Flux in the AERE Reactor GLEEP E. J. AXTON. Journal of Nuclear Energy 21» 125 - 135, (1963)

References

1 . Neutron Production from Electron Accelerators

used for Medical Purposes. E. J. AXTON

and A. G. BARDELL, Phys Med Biol, JJ,, 293 - 298 (1972).

2. Use of the Multispheres neutron detector for

Dosimetry of Mixed Radiation Fields.

M. AWSCHALOM. Proceedings of Symposium on Neutron Monitoring. IAEA Vienna (1967).

3. Effective Cross Sections and Cadmium Ratios

for the Neutron spectra of Thermal Reactors.

C. H. WESTCOTT, W. H. WALKER and T. K. ALEXANDER.

Proc. Second Intern .Conf. Peaceful Uses

At. Energy, Geneva, A/Conf. 15/P/202 (1958).

k. Neutron Fields available at the National

Physical Laboratory and the Need for Low

Energy Neutron Standards. E. J. AXTON and

A. G. BARDELL. Proceedings of Advisory

Group Meeting on Reactor Neutron Dosimetry

IAEA Vienna (Nov. 1978).

5. Conversion Factors and Effective Quality

Factors for Neutrons. Recommendations of the

International Commission on Radiation Protection.

ICRP Publication 21. Pergamon Press (1971).

6. Kerma Factors. ICRU Report 26. Appendix A. (1977).

7. Handbook on Nuclear Activation Cross-Sections.

Technical Report Series No I56. IAEA Vienna

(197^^).

. 8. Photon Mass Attenuation and Mass Energy

Absorption Coefficients. J. H. Hubbell,

Radiation Research, JO, 58 - 81 (1977).

9. Photoneutron Production in detectors in high

Energy X-ray beajns. A. P. Kushelevsky and

G. Shani. Nuclear InstrxMients and Methods

J31 71 - 73, (1976).

10. Gamma-Neutron Cross-Sections. H. E. Johns,

L. Katz. R. A. Douglas, and R. N. H. Haslam.

Physical Review 80, IO62 - IO68 (1950).

119
APPENDIX 1 Monte Carlo Ceilculations

The efficiency e, (table l), the conventional

thermal fluence (table 2) and the epithermal

flux parameter r(T /T )2 are related by the

equation: . . mo

e = 0.001 (N/A) nv (Fg + r(T /T )^ Gs ) o moo

where the symbols are as defined in appendix 2.

The energy of bin zero is 0.0322 eV. The energy

of bin i is given by exp((i - 6). In (l0)/5)

Table 1

Sphere diameter in inches.

i 2 3 5 8 10 12

GO 1988 1572 897 304 136 59

01 2674 2067 1131 366 159 67 02 3369 2635 1399 439 184 76 03 3709 3230 1703 490 207 84 OM 4224 3419 1821 563 237 94 05 4377 3894 2071 614 261 103 06 4044 4196 2271 671 277 112 07 4380 4332 2425 723 296 119 08 4359 4389 2743 763 311 130

09 4250 4577 2664 806 336 135

10 5197 4977 2640 864 359 137

11 5046 5129 2859 900 363 146

12 4343 5125 3024 912 385 152

13 4374 5126 3173 986 390 161 in 4125 5113 3374 1008 408 156

15 3932 5162 3320 1069 422 174

16 3783 4880 3549 1090 447 177

17 3435 4735 3686 1097 459 181 18 3156 4696 3706 1168 479 188

19 3035 4673 3761 1207 517 194 20 2751 4489 3763 1314 551 204

21 2609 4394 3780 1352 547 212

22 23O6 4180 3899 1316 538 221 23 2140 4224 3983 1390 570 229 24 1974 3914 4124 1396 606 231 25 1809 3899 3837 1466 624 247

26 1606 3615 4139 1534 668 250 27 1591 3636 4045 1633 695 269

28 1486 3377 4214 1669 708 282

29 1256 3285 4036 1756 777 307

30 1094 3021 3972 1837 856 321

31 919 2685 4054 1971 875 374 32 761 2464 4080 2361 1030 429

33 649 2100 4275 2398 1218 489

34 516 1771 4028 3142 1526 707

35 367 1557 3816 3303 1938 942

36 240 1046 3480 3639 2473 1401 37 141 781 2928 4427 3179 2111 38 90 514 2197 3740 3133 2485

39 55 282 1509 3007 3139 2510

40 30 215 916 2606 2951 2487

41 23 120 736 1850 2071 2124

120

I

(

1

Table 2

X 10

Sphere diameter in inches.

1 2 3 5

Q0 10 12

00 697 552 107 21

01 937 123 DO 24

02 1 1 82 1 CJl

1 Ot) 26

03 1 300 1 AT 0 1073 571 1 72 73 29 04 1 198 C 0 Q o3o 197 03 33 05 1 535 1 3d5 72b 215 92 36 06 im 0 1 *» n 790 235 OT97 39 07 1535 1519 850 254 104 42 08 1528 1539 964 267 109 45 09 1 429 1605 931 283 1 lo 48

10 1398 1565 928 303 124 48

11 1355 1597 985 318 127 51 12 1238 1 5ot5 1033 321 135 53

13 11 55 1570 1090 342 137 57

14 1 086 1558 1 158 353 143 55

15 1032 1569 1 1 34 375 14o 60

16 994 1475 1204 379 157 62

17 953 1426 1260 383 161 64

18 818 1398 1234 407 168 66

19 789 1398 1243 418 181 68 20 711 1342 1251 458 188 71 21 680 1298 1252 457 191 74 22 614 1233 1294 460 189 77 23 565 1242 1317 483 200 80 24 516 1153 1344 480 211 81 25 467 1142 1250 507 219 87 26 436 1063 1343 533 232 88 27 411 1064 1305 553 243 94 28 385 983 1376 570 249 99 29 384 946 1308 599 269 108 30 283 874 1286 631 305 113

31 233 777 1221 672 307 130 32 201 713 1348 807 359 151 33 165 601 1329 817 418 170 34 127 512 1254 1053 523 261 35 97 451 1215 1091 656 319

36 63 299 1130 1200 829 488

37 36 224 917 1408 1056 711 38 25 149 687 1221 1030 821 39 15 80 462 1002 1050 826 40 8 61 282 849 964 806

41 6 34 225 603 663 686
Table 3

r(T /T )^ X

10' m o

Sphere diameter in inches.

i 2 3 5 8 10 12

00 0 0 0 0 0 0 01 80 5 0 0 0 0 02 236 7 0 0 0 0 03 312 44 0 0 0 0 04 395 63 0 0 0 0 05 463 145 0 0 0 0 06 563 150 0 0 0 0 07 549 164 14 0 0 0 08 530 198 19 0 0 0 09 598 243 17 0 0 0 10 636 240 22 0 0 0

1

1

642 247 35 0 0 0 12 661 277 53 0 0 0 13 689 304 57 0 0 0 14 697 315 67 0 0 0 15 706 323 56 0 0 0 16 705 336 103 5 0 0 17 713 345 86 7 0 0 18 742 373 111 14 0 0 19 716 361 127 26 0 0 20 749 367 113 14 53 0 21 724 393 122 37 22 0 22 757 405 119 13 0 0 23 743 404 127 19 0 0 24 748 399 157 48 18 0 25 753 413 158 33 1 0 26 761 403 169 17 15 0 27 763 415 182 21 6 0 28 763 428 148 63 0 0 29 771 1 7? 0 30 748 447 174 38 0 0

31 794 446 186 71 0 0 32 872 445 188 55 31 0 33 791 472 255 60 44 19 34 768 453 232 100 45 88 35 747 441 213 120 75 72

36 724 447 229 130 96 79

37 833 468 250 153 115 87 38 831 471 254 154 140 167 39 783 497 270 132 157 144 40 750 500 290 160 180 170

41 750 500 310 160 200 180

121
Appendix 2 (3)

The Low Energy Component of the Fluence.

Interpretation of Cadmiiun Ratio Measurements.

This procedure is an extension of the

(3) Westcott treatment for the expression of

effective neutron reaction cross sections and

reaction rates in well-moderated environments.

The treatment pre-supposes a Maxwellian

distribution of the neutron energies characterised

by a temperature T^, supplemented by a '1/E'

shaped slowing down spectrum starting at a lower energy yE^T^/T^ and terminating at a source energy E^, where p is a moderator-dependent

constant or cut-off function, = 293. 6°K, and E = 0.0253 eV. o

. (12), Following the Westcott formalism

expressions can be derived for the 'conventional

thermal neutron fluence', n.,v and the cadmium ' th o ratio R_,. n., is defined as the neutron density

Cd th

in the energy range below the cadmium cut-off

energy, E^^ and v^ = 2200 m.s. \ the velocity

of neutrons of energy E^. Note that the true fluence would be n^, v, where v is the average

th

velocity of the neutron density n^^^. The

conventional fluence is simply the foil reaction

rate divided by the product of the number of

nuclei and the effective 2200 m.s.~^ neutron

capture cross-section

If T is unknown the quantity derived from

_ m '

1 the cadium ratio measurement is r(Tmo/T )^. so is the normalised reduced resonance activation

integral I' (integrated in a 1/E shaped flux,

1 /v component subtracted )

s = ik/v)^ I'/c . o o

G is the correction for resonance self shielding, f . represents the attenuation of the

0 resonance neutrons in cadmium of thickness 6, W is

a correction which is necessary when part of the

first resonance extends below E . K, is related Cd 0 to E^^ by the equation

^6 = ^/^o^'/^

K is tabulated in reference (3).

0

(5)

The essential information derived from thf

foil measurements is thus n^, v and r Rt /T 1 ^

th o ^ m

' from which the total conventional fluence nv may

be derived. °

nv = n^^ V R/(R-1

o th o

(6)

Where R is the cadmium ratio of a pure '1/v'

absorber, given by K/r(T /T )^ , The two m o components of conventional fluence are then given

by

th o

- °°Cd ^Cd^ ^

0.001 F N ga 0° o

(1)

nmoV = nv (1 - kr/iuu) ) o

n V = nv Ur(Tru)^

e o o

(7)

(8)

Where Do is the reaction rate (equal to the

saturation disintegration rate) in the bare foil per milligram of foil, ^°q^ is the reaction rate

per milligram of the cadmium covered foil, A is the atomic weight of the foil, is the

Avogadro constant, is the thermal neutron

capture cross section of energy E^, F is a correction for thermal neutron self shielding

in the foil, and g is the Westcott correction

factor for cross section resonances in the

thermal Maxwellian energy region.

The cadmium ratio is given by

R-, = Do/Do = (F/r(T /T )^ + Gs /g) Cd Cd mo o f^G(s^/g-W) 1/K^

(2)

Where r is the Westcott epithermal flux para- meter which is related to the fraction f of the

total neutron density n which comprises the

Maxwellian part of the spectrum:

where nmeand n are the Maxwellian and epithermal neutron densities.

These conventional fluences have to be

converted to_true fluences by multiplying by the

appropriate v/v^.

For the Maxwellian part, v/v^ is given by m o For the epithermal part

UE T /T

o m o

dE/E and

n V

e o

(E /E)^ dE/E

o yE T /T

o m o

(9)

where c is the normalisation constajit. From which

v/v = In(E^) - In (u E T /T^)

o s o m o 2E ((u E T /T 2 o o m o E

"2) (10)

122
Neglecting E ^ and dividing by the lethargy :-

n V (per unit lethargy) = nv (U/Tr)^r(T /T ; e o m o

(11)

Hence the total true fluence is given by

nv = nv^ (l|T^/TrT^)^|jl - Ur(TTp)"^) + r^

u.J (12)

where the u^ are appropriate lethargy intervals

from yE T /T to E^^

.

o m o Cd

The response of the sphere in this

fluence is given by

nv (UT /ttT )M(1 - Ur(^y)"^e^^ ^. o m o |_ thermal, J

+ r y u. e.

.

V 1 ij

_ "e.bare.s ^""^ ^B.Cd.s

^«

efficiencies determined from coincidence measurements following the irradiation of bare

and cadmium covered foils in a standard flux

which produces a cadmium ratio R , then the bare

s foil efficiency e. , appropriate to an e,bare,u '^^ ^ unknown field which produces a cadmium ratio

is given by

^B.bare.u = ^3 .bare.s^ ^""s ^ ^"V^^

where 1 + a = e. -u s 6,Cd,s 6,bare,

s

(15)

In the interpretation of the measurements of

the foils in polyethylene spheres the appropriate

R 's were determined from the r(T /T )^ data of u m o appendix 1 and equation (2).

(13)

Similarly the dose-equivalent or tissue kerma are given by

nv (Ut /ttT ) o m o

(1 - i*r(Try)~^ DFthermal

+ r I u. DF.

i

1 (1U)

where DF^ are the appropriate fluence-dose con- version factors.

B-Particle Counting

Because of the low foil activities achieved,

the foils were counted in a low background

anticoincidence coiinter rather than in a k-n&y

coincidence counter. It is therefore necessary

to ensure that the correct 3 counting efficiency

is used to convert observed counting rates to absolute disentegration rates. Each of the foils

used was activated in a standard thermal neutron

flux and then counted by UirBy coincidence to determine the appropriate B-counting efficiencies

A cadmium covered foil is activated by

resonance neutrons which are rapidly attenuated

in the foil so that the activity is confined to a thin surface layer. The counting efficiency

Eg Cd' ^ foil activated under cadmium is thus

higher than the counting efficiency s^j^gj^ngl*

appropriate to thermal neutron activation whereby

the activity is produced more uniformly throughout the foil thickness. The efficiency e

p , bar e for a foil irradiated bare is thus a cadmium

ratio-dependent weighted combination of

^B, thermal ^B Cd.

123

UBS SP 554 (1979)

I SURVEY OF EUROPEAN MEASUREMENTS ON PHOTONEUTRON PRODUCTION

I FROM ELECTRON ACCELERATORS USED FOR .MEDICAL PURPOSES

E J Axton and A G Bardell

National Physical Laboratory

Teddington, Middlesex

Brief details -are presented of Eviropean work on problems associated with neutron production from electron accelerators used for medical pxirposes. For completeness,

the earlier references which have already received citation in the literat\ire have

been included.

I (Dose-equivalent, Electrons, Kerma, Multispheres , Photons, Photoneutrons

,

Radiotherapy , Spectrometry

)

Compared with current activities in the

United States the European effort in this field

is rather limited. The material described below

is the result of a search which may not have been

completely exhaustive. Earlier references which

have already received citation in the literature

have been included for completeness

.

In what follows , quoted neutron dose or doseequivalent rates have been normalised to a central

axis X-ray dose of one rad unless otherwise

stated.Frost and Michel^ considered the total

incremental dose-equivalent due to photoneutron

production in the accelerator as well as all

photonuclear reactions produced in the patient

including the production of photoprotons . A

total contribution of 0.1 rem was obtained for a

S^+MV bremstrahlung spectrum.

. (2) Pohlit used Indium probes in a paraffin

moderator to obtain the incremental doseequivalent to the patient from neutrons generated

externally. For a 35 MV betatron he derived a central axis dose-eq\iivalent of 2.5 mrem, which

fell to 1.0 mrem at a distance of 50 cm from the

axis

.

(3) Kretschko, Liesen, Pohlit, Rase and Sewkor

carried out a comprehensive siarvey of the X-ray

and neutron radiation levels associated with

eight accelerators ranging in energy from 15 MeV

to 100 MeV. The distribution of stray

X-radiation was found to be dependent on the

accelerator construction. Neutron levels were

surveyed with a boron trifluoride counter

surrounded by a double moderator. The distribution of fast neutrons is determined by the

shape and dimensions of the the therapy room.

The measurements provide a useful guide to the

optimum design of the therapy room.

(U) Lofgren and Spring used a system of

fission track etching in glass, having calibrated

it with an AmBe neutron source, to determine the

in-beam neutron dose-equivalent for a 32 MV

betatron. The very low neutron rates observed are usually discounted in the literature (eg (5) ).

The work of Kushelevsky and Shani^^^ has , ^ already been mentioned in an accompsinying paper

They used the photon spectraJ. data of Johns et al^ with cross section data from the IAEA compilation'9', to calculate the neutron production

per gram for a number of elements as a function of

accelerator energy. The results are confirmed by

those of reference (

) . They went on to calculate

the total number of neutrons produced in two neutron detectors:- a NE 213 liquid scintillator

detector and an 8 inch diameter spherical remmeter. In each case they estimated that about 10^ neutrons would be produced in a nominal exposure of

100 rontgens at 2k MV. Subsequent comparisons

with the number of neutrons striking the instru- ments from the outside are in error(7).

Stranden^ ^'^^ used indium probes surrounded by

a 7.5 cm wax moderator as a fast neutron detector

in order to estimate in-beam and out-of-beam

neutron dose rates from a U5 MV betatron operated

in the X-ray mode. He obtained k.2 mrem in the

begun and 2.0 mrem out of the beam. As an example

of the consequences of these results, he estimated that a patient receiving a dose to the

thorax of 200 rad would experience a gonad dose of

between 0.05 and 0.1 rad from scattered photons,

and a dose-equivalent of 0.i+ rem from neutrons.

In an earlier paper' ^

^

' he considered the whole

body dose to the patient as a result of the

activation of 0,N,P,K and CI in the patient.

These activities are produced partly by reactions

involving photoneutrons and partly by direct

photonuclear reactions with the elements concerned. The dose due to reaction products, for

example photoprotons, was not included. Stranden

concluded that a patient would receive a whole

body dose of l6 urad as a result of the radioactivity produced by a U5 MV X-ray accelerator.

(12) Asard used an Anderson Braun (Studsvick) rem meter, consisting of a boron trifluoride

counter in a hydrogeonous moderator, to measure

125
neutron dose rates in the control room of a Siemens IT MV betatron. The dose-equivalent

rate was foimd to be acceptably low, the highest

level recorded being 0.2 mrem.h"^ for an in-beam

dose rate of 27 rad.min."^. In a later paper

from the same laboratory, (the Karolinska

Institutet, Stockholm) Brahme(^3) considers the

therapeutic merits of mixed photon and neutron

beams from a 50 MV Microtron. Photoneutron

converters of uranium or lead are placed near the

X-ray target to produce a combined X-ray and neutron beam for patient treatment. Depth dose curves are presented for photoneutrons from U and from Pb as well as for the photon beam.

( 1U) Eipper and Manegold measured neutron dose rates from the h2 MV betatron at the KolnMerheim Radiation Clinic. They used two neutron

detectors, a boron trifluoride counter in a double moderator, and an AEC neutron monitor ELM ^^h LMS. Measurements were made in out-ofbeam positions in the treatment room, and through

the entrance maze to the control room door. A

dose-equivalent rate of 0.6 mrem.h"^ was observed at the control room entrance to the maze for an X-ray exposure of 100 rontgen.min."^

It was concluded that no additional shielding

would be required, such as a paraffin inlay to the door, to reduce the neutron levels in the

control room.

Ulso and Christensen^ ^ ^

^ criticised the

'often claimed' statement that a sufficient

shielding for the X-rays from a medium energy electron accelerator should also be sufficient

to reduce the potential neutron contamination in

the control room to an acceptable level. In an experimental study, a neutron dose rate meter

2202D (Studsvick Atomenergy AB Sweden),

consisting of a boron trifluoride counter inside

a moderator, was used to monitor neutron dose

rates at the entrance to the maze leading to a SL75/20 linear accelerator. The monitor was gated to exclude the PRF pulses from the

accelerator. Additional measurements were made

with nuclear track emulsions which have a sensitivity threshold of 0.7 MeV for neutrons. Initially the neutron dose-equivalent rate was

800 mrem.h"^ for a primary X-ray dose rate of

2it0Gy.h~'' (2I4OOO rad.h"''). The neutron rate was reduced to less than 2.5 mrem.h"^ by reducing by

a factor of 2.U the area of the entrance door,

by adding an additional 10 cm thick borated

paraffin door, and by lining parts of the wa]J.s and ceiling of the treatment room with 5 cm of

borated paraffin.

Vialettes, Cariou, Duffar, Mesples, Quechon,

and Rousson ^'^^ surveyed the neutron and photon

radiation levels in out of beam positions in the treatment room of the 25 MeV Saggitaire linear

accelerator installed at the Henri Mondor

Hospital at Creteil, France. Enriched LiF

thermoliiminescent samples were used for the photon measurements. Neutron levels were sur- veyed with indium activation foils in a paraffin

moderator. The '^S(n,p) reaction was also used

for neutrons above the 2 MeV threshold for this

reaction. The results are presented in the form

of isodose curves which show that the neutron

and photon contributions to the total leakage

dose-equivalent are approximately equal. The

contribution of thermal neutrons is very small.

Drouet, Moreau, Quechon, and Tabotd^) of

the Radiation Protection Service at Saclay performed similar measurements on an 18 MeV Saturne accelerator and reached similar conclusions.

GuntherdS) reported the isodose distribution

of neutron radiation from the 17 MV betatron at the Radiation Clinic of the University of Dusseldorf . At the time of writing it has not been possible to obtain copies of the last

reference for comment.

Conclusions . Only one in beam neutron measurement is reported in this survey, and it

is in general agreement with our own measurements. Conclusions drawn from out-of-beam

surveys are sometimes contradictory, see for example references 14 and 15. It is not clear

without further study whether this is due to differences in measurement techniques or to physical differences in the installations

surveyed

126
References

1 Incremental Dose Value Due to Neutrons in Therapy with Fast Electrons and with Ultrahard X-rays. D. Frost and L. Michel

Strahlentherapie _12l4, 321-350 {^96h)

.

2 The Measurement of Neutron Radiation from a

35 MV betatron. W. Pohlit , Strahlentherapie,

113 , 1*69-^73 (i960).

3 Radiation Protection Measurements on Different

European Betatron Installations. J.Kretschko,

H. Liesen, W. Pohlit, S. Rase and A. Sewkor.

Fortschritte auf dam Gebiete der Rontgenstrahlen und der Nuklearmedizin, £5, 565-572

(1961).

k Neutron Radiation Produced by the 32 MV

Rontgen Beam of a Medical Betatron.

K.E. Lofgren and E. Spring. Acta. Radiol.

2, 2I+7-256 (1970).

5 Fast Neutrons from a 25 MV Betatron. J.G. Fox

and J.D. McAllister. Medical Physics, k,

387-396 (1977).

6 Photoneutron Production in Detectors in High

Energy X-ray Beams. A. P. Kushelevsky and

G.Shani. Nuclear Instruments and Methods

137 , 71-73 (1976).

7 Neutron Production from Electron Accelerators

used for Medical Purposes. E.J. Axton and

A.G. Bardell. This Conference. (1979).

8 Gamma-Neutron Cross Sections. H.E. Johns,

L. Katz, R.A. Douglas and R.N. A. Haslam,

Phys. Rev. 80, IO62-IO68 (1950).

9 Handbook on Nuclear Activation Cross Sections.

Tech. Report Series N0.I56. IAEA Vienna

(197^*).

10 Neutron Doses to Patients by High Energy

X-ray Therapy. E. Stranden. Phys. Med. Biol.

22, 1011-1013 (1977).

11 Activity Induced in Patients by High Energy

X-ray Therapy. E. Stranden, Phys. Med. Biol.

22, 3U8-352 (1977).

12 Radiation Protection Measurements for a 17 MV

Betatron. P.E. Asard, Acta. Radiol. 1»

59-70 (1968).

13 Photoneutron Beams for Radiotherapy Produced

by a 50 MV Microtron. A. Brahme, Private

Communication (1979), to be presented at the

Radiation Research Congress in Tokyo (1979)«

1U Neutron Radiation from a Betatron. H.H. Eipper

and K. Manegold. Strahlentherapie, IUO , 286-290 (1970).

15 Neutron Contamination at a 20 MV Linear

Accelerator. N. Ulso and J.M. Christensen.

Arhus Kommunehospital , 80OO Arhus. C. Denmark.

Private Communication (1979)'

16 Measurements Performed on the Sagittalre

Accelerator at the Centre Henri Mondor.

H. Vialettes, J. M. Cariou, M. Duffar,

D. Mesples, H. Quechon and J. Rousson. Report SPR/S/72-853 GRA/121 HV/nw.

20 June 1972.

17 Measurements Performed on the Therac 20 Satume Accelerator. J. Drouet, J. C. Moreau,

H. Quechon, and L. Tabot. Report D.CEN- S/SPR/SRI/77-507 GRA-092 JD/sg. (1977).

18 Neutron Dosimetry with a 17 MV Betatron.

D. Gunther. Strahlentherapie, 151 , 333-337

(1976)

.

127
1
NBS SP 554 (1979) NEUTRON LEAKAGE FROM CURRENT MACHINES

Peter R. Almond

The University of Texas System Cancer Center

M. D. Anderson Hospital and Tumor Institute

Houston, Texas 77030

A survey has been made of published, and unpublished reports during the last

ten years, of neutron leakage measurements around medical linear accelerators.

The review involved seventeen different studies involving fourteen different

types of machines, six betatrons and eight linear accelerators with measurements

reported on twenty-eight different machines. It was found that as the energy

Increases, the neutron leakage increases and is a maximum at around 25 MV and then remains constant. The target appears to be the main source of neutrons. In a treatment room, the neutrons can be identified as fast primary neutrons, fast

scattered neutrons from the walls and a constant background of thermal neutrons. The neutron dose in the photon beam is approximately four times higher than 50cm

from the beam edge. At greater distances, the neutron dose becomes constant. Due

to the large number of variables in the different reports, it is possible to draw

definite conclusions. It is recommended that standards be established for neutron leakage measurements.

(Betatrons; detectors; dosimetry; leakage; linear accelerators; neutrons)

Introduction

This paper will be an attempt to survey both published and unpublished measurements of neutron

leakage for radiotherapy machines operating in the photon, mode that are currently in use, and

to see if any pattern can be found with regard to machine energy, target material, measuring

techniques, etc. Such a review would be impossible without the cooperation of many people

and I am indebted to the various companies and individuals who sent me material to review.

Table I lists the sources of Information and, in general, the data has been restricted to measurements reported after 1970.

Variables

In order to compare the data from the

different sources, some attempt has to be made

to summarize all the variables involved and to reduce the data to a common form. The basic

variables in this problem can be listed as follows:

1. The treatment machines including energy,

target material, flattening filter and collimator material

.

ii. Measuring devices, calibration, and calculation techniques.

iii. Experimental set up and measuring

conditions.

iv. Presentation of the data.

V. Treatment room information.

Table II lists the various machines that

were considered in this report. Included is target and flattening material and collimator

material. It should be pointed out, however,

that in many of the published papers, this

information was not Included.

Table III lists the measuring devices used

and their means of calibration and conversion

factors. McCall has pointed out the problems

that are often encountered with the use of the various detectors. These are:

1. Reported neutron doses are too high

because of unsuspected and unaccounted for high energy photon response of detectors.

11. For fluence detectors errors in the efficlences of the dectors can arise because of Incorrect assumptions regarding the neutron

spectrum.

ill. For dose-equivalent detectors errors

in the efficlences of the dectors can arise

because of Incorrect assumptions regarding the neutron spectrum.

iv. Spectral errors will also cause incorrect results when fluence is converted to dose-equi val ent.

Often the experimental set-up is not adequately described, but when it is described,

it is apparent that no standard experimental

conditions have been followed. Table IV lists

the experimental conditions that have been reported.

Table V gives the various ways in which the data is presented, again no standard form has been used.

And, finally, very few of the reports give

information about the treatment room and the relationship between accelerators, detectors,

and wall. When this information has been included, it is indicated in Table IV.

129
Table I

1. Deye and Young: "Neutron Production from a 10 MV Medical Linac", Phys. Med. Biol., 1977,

Vol. 22, p. go.f^^J

2. 17. 19. McGinley, Wood, Mills, and Rodriguez: "Dose levels due to neutrons in the

vicinity of high-energy medical accelerators".

Medical Physics, 1976, Vol. 3, p. 397.1^^^

3. McCall Associates: Neutron Measurements on

3 Mevatron XX Linear Accelerators, 1977 & 1978,

Supplied by Volker Stieber Manager, Development,

Siemens.

4. Gur, Bukovitz, and Gill: "Fast and Slow neutrons in an 18-MV photon beam from a Philips

SL/75-20 linear accelerator". Medical Physics,

1978, Vol. 5, p. 221.'='*^

5. 6. 7. Gur, Bukovitz, Rosen and Holmes:

"Relative Measurements of fast neutron contamination in 18 MV Photon Beams from two linear accelerators and a betatron". Abstract

Medical Physics, 1978, Vol. 5, p. 350, Text supplied by Bukovitz.

8. 12. Fox and McAllister: "Fast neutrons from

a 25 MeV betatron". Medical Physics, 1977,

Vol. 4, p. 387.

9. Drouet, Moreau, Quechon and Tabot: "Summary

of Measurements Taken around a CGR/MeV Saturne Accelerator at Buc", C.E.A. C.E.N. Saclay,

Department of Radiation Protection Section of

Installations. Accelerator Group, May 1977,

obtained from AECL and CGR-MeV.

10. Grant: "Neutron Leakage for Therac 20" Memorandum to Radiation Safety Officer, M. D. Anderson Hospital, July, 1978.

11. 13. Wilenzick, Almond, Oliver, de Almeida: "Measurement of Fast Neutrons Produced by HighEnergy X-Ray Beams of Medical Electron Accelerators", Phys. Med. Biol., 1973, Vol. 18,

p. 396. C7j

14. Holman, Price, Friedman, and Nath: "Neutron

spectral measurements in an intense photon field

associated with a high-energy x-ray radiotherapy

machine. Medical Physics, 1977, Vol. 4,

p. 508.

15. Marbach: "Neutron Leakage from the Sagitaire Linear Accelerator at the University of

Indiana at Indianapolis", June, 1975, AECL

Report obtained from J. Marbach.

16. Price, Nath and Holeman: "Fast and thermal neutron profits for a 25-MV x-ray beam",

Medical Physics, 1978, Vol. 5, p. 285.

18. Axton and Bardell: "Neutron Production from

Electron Accelerators used for Medical Purposes",

Phys. Med. Biol., 1972, Vol. 17, p. 293. DoJ

20. Lofgren and Spring; "Neutron Radiation by

the 32 MeV Roentgen Beam of a Medical Betatron",

Acta Radiologica, 1970, Vol. 9, p. 247.

21. Stranden: "Neutron Doses to Patients in High Energy X-Ray Therapy", Phys. Med. Biol.,

1977, Vol. 22, 1011.

22. Oliver: "Fast neutron contamination in x-ray beams of medical accelerators from 19 to 45 MV," Abstract Phys. Med. B1ol., Vol. 19 ,_ 1974. Revised 1976 data supplied by Oliverl")

23. LaRiviere: "Summary neutron leakage

numbers for the Varian machines." Data supplied by Philip D. LaRiviere of Varian.

Note on the numbers: The Initial numbers in the table above can be used as a key to correlate all the Tables I through V, e.g. the

number 15 refers to the report by Marbach In

all tables, and the figures.

130
Table II

Machines

Machine Energy MeV Target mm Filter Collimator Photon dose rate

1. Varian Clinac 18 10 Cu U 200 rad mi n1

2. Varian Clinac 18 10 Cu 6.3 W W 100 rad min-i

13. Siemens Mevatron XX 14 300 rad min-i

4- Philips SL/75-20* 18 W W alloy W alloy 280 rad min-i

5. Siemens Mevatron XX 18 Pt Fe W alloy 300 rad min-i

6 Schiniad7u Betatron"^ 18 Pt Pb W alloy 50 rad mi n~ ^

7.

8.

Philips SL/75-70*

Schimadzu Betatron+

18

18

W

Pt

W alloy

Pb

W alloy 300 rad min-i

9. CGR-MeV Saturne 18 W4 400 rad min-^

10. AECL Therac 20 18 U Pb W alloy 400 rad min-i

Hi. 19 75 rad min~^

12. Schimadzu Betatron'*' 23 Pt 2 Pb W alloy

M3. Sagittaire Linear

Acc. 25 400 rad min-i

14. Sagittaire Linear

Acc. 25 400 rad min-i

15. Sagittaire Linear

Acc. 25 400 rad mi n~ 1

16. Sagittaire Linear

Acc. 25 W Pb 450 rad min-i

17. All is Chalmers

Betatron 25 Pt 1.6 Al Pb 100 rad min-^

18. Brown Boveri Betatron 33 35 rad min-i

19. BBC Betatron 45 Pt 2.0 Pb Pb and W 100 rad min-i

20. Asklepitron 32

21.

1

1

Betatron (make

unknown) Not stated Pt

22. Data for these machines included in Table VI.

23. Varian Clinac 18 at 10 MV and Clinac 20 at 15 MV and 10 MV.

*

+ Same Machines

M Target, Filter and Collimator data same as

j (5).

' ^) Target is Pt. Filter and Collimator are Pb.

, M Target is W Filter and Collimator Pb same

for (14) (15) and (16).

131
Table III

List of Detectors

1. 5.1cm diam Rh-act1vat1on counter - calibration (ncm-^ counts-^ by standard ^'^Cf neutron

sources at Naval Research Laboratory. Conversion

from neutron fluence to dose by 3.0 ± .1x10-'

rad n-^cm-^ Q.F. = 10.

2. 17. 19. Fast neutrons, bare indium foil in H20 moderator - calibration with ^^^Cf source. Fluence converted to absorbed dose in muscle by 3.97 X 10-' cm^ rad/n. Thermal neutrons, bare

indium foils and cadmium covered indium foils

with H2O moderator. Calibrated with a moderated

Pu-Be source. Fluence converted to absorbed

dose in muscle by 4.57 x lO-^^cm^ rad/n.

3. Bare and moderated gold foils. Neutron fluence measured and results interpreted as dose

equivalent based on results from Monte Carlo QO

calculation for each particular room, i.e. factor for converting fluence to dose equivalent

individually calculated but averaged about

7.9 X 10^ n/cm^ per rem.

4. Aluminum and indium foils and cadmium covered

indium foils to check 'so called "slow neutrons'"

Same technique as 8 and 12. Fast neutron fluence converted to Fast-neutron dose by 3.1 X 10-' rad cm^/n and slow neutron fluence

by 4.5 X 10-1° j.ajj cm^/n.

5. 6. 7. Aluminum foils same as 4. 8. Aluminum foils Photon sensitivity taken into account. Conversion factor of 3.1 x 10-' rad/per unit fluence for target neutrons and 3.2 X 10-' rad/per unit fluence for lead neutrons.

9. Indium foils in 5 cm paraffin moderator

covered with cadmium. Combination of this

activation foils Cu, Au, Mg, and S in box covered with cadmium and Cu. Equivalent energy

is derived and equivalent dose per neutron

fluence is found for this energy. 7.7 x 10-' rem

n-i cm^ at 1 meter and 5.2 x 10-' rem n-^ cm^

at 2 meter.

10. Silicon diodes calibrated with Cf-252 same

as 11 and 13.

11. 13. Silicon diodes and fission fragment

track detectors - Lexan using depleted U foil.

Diodes calibrated by ORNL Health Physics

Research Reactor and ^^^Cf. Track detectors

calibrated by ^^^Cf. Conversion from fluence to dose by 3.0 + 0.1 x 10"' rad n'Vcm^. Photo

fission correction applied to track detector.

No correction for diode data*. Q.F. = 10.

14. Bonner multisphere spectrometer using 6LiF

and 7LiF thermoluminescent dosimeter. Also bare

and cadmium covered chips. Maximum dose and dose-equivalent conversion factors from ERDA Health and Safety Lab '^D - applied to each energy

group and summed yielding total neutron dose

equivalent. Q.F. varied from 8.2 to 6.6.

is. silicon diodes same as 11 and 13. Q.F. = 10.

'*In a subsequent report by McCall , et al

isilicon diodes were shown to be unsuitable for

Luse in the photon beam.

16. Activation Detectors using P2O5 31P (np) Si for fast neutron and 31P (ny) 32p for slow

neutrons. Neutron spectrum was constructed from

data in literature and for every energy group in the assumed spectrum, max dose equivalent

conversion factors from ICRP 21 were used. Photon sensitivity range was 1.7% to 3.8%.

18. Gold foils in two polythene sphere 20cm and 25cm diam sphere calibrated at 2 neutron energies

using the Li(p.n.) at NPL. Energy of neutrons were estimated. Conversion to rem by 20 n cm-^s-^ for 2.5 m rem h-^from NCRP 38. 03

20. Solid-state track detector - glass with

uranium oxide incorporated into it. Calibrated

again An-Be source. Conversions factor of

7.66 X 10-' rem(n/cm2) Q.F. = 7.4.

21. Indium foils and Indium foils surrounded by

paraffin and cadmium used to measure total flux

of fast neutrons 12c and 64zn and 56Fe to study energy distribution. Dose equivalent obtained

from spectral information and NCRP 38 data. E3

22. Silicon Diode.

23. Moderated Indium foil calibrated against

PuBe Source.

Table IV

Experimental Conditions

1. Room layout. 25cm cube polystyrene phantom

lOOTSD, 25cm x 25cm field, detector 100cm from

edge of phantom away from gantry in patient plan.

Photon dose rate 200 rad min--' Data behind phantom.

2. 17. 19. Room layout. 10cm x 10cm field. Measurement in beam at standard TSD's (100cm for

Clinac 18 and All is Chalmers Betatron, 110cm for

BBC betatron and 5cm from edge of beam (direction

unknown). Photon dose rate approximately

100 rad/min.

3. Room layout. 20cm x 20cm field. Measurements in patient plane at isocenter and 50cm and

100cm from isocenter left and right and toward

and away from gantry. Some data taken with jaws

closed. Measurements in target plane at 100cm. Beam stopper retracted. Photon dose rate

300 rad/min.

4. No room layout. 5cm x 5cm, 10cm x 10cm and 20cm X 20cm field sizes. Detectors inside and outside of (5cm from edge) beam at target to foil distance of 100cm. Photon dose rate of

280 rad/min.

5. 6. 7. No room layout. 10cm x 10cm field.

In center of beam only, at 100cm. Dose rate

300 rad/min for linear accelerators and 50 rad/

min for betatron.

8. 2. No room layout. Different field sizes

10x10, 20x20, mainly no phantom data some phantom data. With and without flattening

filter, in beam and outside beam at 5cm and 20cm, TSD 100cm.

9. No room layout. 15cm x 15cm field made in patient plane 100cm from target around circles

with radii of 1 and 2 meter and on a sphere of

radius 1 meter centered on target. No in beam

data. Photon dose rate of 400 rad/min-^.

132
10. No room layout. 10cm x 10cm field with

phantom. Measurements made in patient plane at

100cm on circles of radii .5m Im 1.5m and 2m. No in beam data.

11. 13. No room layout. For sagittaire

10cm X 10cm and 30cm x 30cm. In beam and out of

beam data out to 60cm, toward and away from

gantry. In patient plane at Im. Photon dose

rate 400 rad min"^. For Siemens 7.6cm x 11.4cm

field. In beam and out of beam data out to 50cm,

toward and away from gantry. In patient plane

at 60cm. Photon dose rate 75 rad min"\

14. Room layout. 10cm x 10cm field. No in beam

data, measurements in plane of patient (105cm) at

0.5, 1.0, 2.0 and 4.8cm from beam axis. Photon

dose rate 400 rad min-^. No phantom.

15. No room layout. 10cm x 10cm field. Diodes

in plane of patient (100cm) out of beam on circle of radii Im and 2m. Photon dose rate

400 rad min-^.

16. No room layout. 5cm x 5cm, 10cm x 10cm, 20cm X 20cm, and 30cm x 30cm. Detectors in patient plane at 105cm. In beam and out of beam

data out to 50cm. No phantom. Photon dose rate

varied approximately 400 to 480 rad min"^.

18. No room layout. Little experiment information. Photon dose rate was 35 rad min-^. In beam and out of beam data.

20. Room layout. 10cm x 10cm field at TSD 80cm. Measurements in beam and out of beam-up to 7cm.

21. No room layout. Experimental conditions not

known. In primary beam and 50cm and 100cm

outside the beam.

22. See Table VI.

23. Clinac 18 measurements at 3 locations 1 meter from target, collimator closed. Clinac 20 measurement in patient plane at periphery of the

beam and at 1 meter from target. Collimators

closed.

Table V

4. Fast neutrons/cm ^/rad of x-ray. Fast-neutron

dose in rad/rad of x-ray Slow neutrons/cm^/rad

of x-ray and Slow-neutron dose in rad/rad of

x-ray. Tabular

5. 6. 7. Relative neutron contamination inside

primary beam only. Overall accuracies, ±40%

listed. Numerical

8. 12. Neutron fluence per rad of x-rays (neutrons cm-^ rad-^). Graphical and Tabular

9. Dose equivalent due to neutrons in mrem/

1000 rads. Diagramatic

10. Neutron rad and percentages of total photon

dose. Diagramatic

11. 13. Neutron dose as percentage of central

axis dose and as Neutron fluence (n cm-^ rad-^)

graphical and tabular. Experimental uncertainties shown for dose percentage data.

14. Neutron dose rates rad/min for slow, fast

total and in rem/min for slow fast and total with derived Q.F. at each point. Tabular and graphical

15. Neutron dose in rad and rem dose as % of

photon dose. Tabular

16. Fast neutron flux n/cm^-sec. Thermalneutron fluxes n/cm^-sec and Dose Equivalent

Rate rem/min for both Fast and Slow. Graphical

18. Neutron flux for 1 rad min-^ of photon. Neutron dose-equivalent (in mrem min-^ and Neutron dose-equivalent divided by x-ray dose. Uncertainties ± 30%. Tabular

20. Neutron fluence (n/cm^)/R and mrem/R.

Diagramatic, numerical.

21. Total fluence ncm"^ per rad of x-ray dose

and dose equivalent (DE) mrem per rad of x-ray

dose. Tabular

22. See Table VI.

23. Data presented as n/cm^ rad and as rem/ rad %.

Data Presentation

1. Flux density (n cm-^ s'M and Fluence per

rad (n cm-^ rad-^). Tabular

2. 17. 19. Fast-neutron fluence per rad photons

(n/cm^ rad). Fast-neutron dose per rad photons.

Slow-neutron fluence per rad photons (n/cm^ rad).

Slow-neutron dose per rad photons accuracy of

±20%. Tabular

3. Fast-neutron mrem/photon rad at isocenter.

Various components given such as "Direct".

"From Walls" all corrected to 1 meter from

isocenter. Leakage at Im (mrem/photon rad).

"Direct" "Fast-Scattered" "Thermal" and "Total".

Uncertainties given 14%. Average energy (MeV)

Including thermals and excluding thermals.

Tabular

133
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134
Review of the Studies

I From Table I it can be seen that 17 different studies were considered. Of these,

twelve had been reported in the literature, ten

as papers and two as abstracts. Four of the

; studies were done for the manufacturers and one

j report was of a survey done for state health

! department requirements. Twelve of the reports

' have appeared in the last three years so that

j most of the data represents the latest work done

! in this area.

Machines

Data has been collected for 14 different

types of machines (8 linear accelerators and 6

j betatrons) with measurements reported on 28

I individual machines. All of the betatrons had

I platinum (Pt) targets and lead (Pb) flattening

' filters except the Allis Chalmers which uses an

I aluminum (Al) flattening filter. All of the linear accelerators used tungsten (W) targets

except the Mevatron XX which uses Pt and the

Clinac 18 which uses copper (Cu). The flattening

filters were W except for the Therac 20 - Saturne and Sagittaire which use Pb. In almost

all cases W was used as a collimator material

! except for the Allis Chalmers which used Pb.

i Photon dose rates for the linear accelerators

1 were high 300-400 rad min-' and low, upto 100

rad min-^, for the betatrons.

Detectors

Of the 17 studies, 11 used activation

techniques with indium foils being the most

conmon, followed by gold and aluminum. The

latest technique is to use phosphorus pentoxide

to measure both the fast and slow neutron

component and it has a low photon sensitivity.

There were four studies using silicon diodes

which, because of their response to high energy

photons, cannot be used directly in the photon

beam.

Two studies used solid-state track

detectors. One set of results gave such low

j results that it must be questioned.u U The

; other set of data agreed well with the silicon

] diodes when photo fission effects were accounted

for. One group used thermoluminescent

dosimeters. It is quite apparent that the best

results can be obtained if a measurement or accurate estimate of the neutron spectrum can

be made and the appropriate conversion factors

to dose or dose equivalence factor is calculated

for each spectrum.

Data Presentation

The information presented in Table V is as it is presented in the papers. The data is pre- sented in many ways using different units, these

include, flux, flux density, fluence per photon

rad, neutron dose rates in rad/min, neutron dose

in rad per photon rad, neutron dose as percent of photon dose, equivalent ratio, dose equivalent

per photon rad and dose equivalent as a percent

of photon dose. Some of the units are used correctly according to ICRU 19 whereas some

are not.

Discussion

Review of all the data in a comprehensive

way is difficult because different detectors were

used with different experiment arrangements, the

data was presented differently and conversions

factors varied. Some factors were obtained by

detailed calculations while others were taken

from the 1 iterature.

However, certain conclusions can be drawn

that are important to this subject.

1) That the target is the main source of

the neutrons. By target is meant the x-ray target and surrounding ncin-moveable collimators.

2) Throughout the treatment room there will

be a constant background of thermal neutrons representing about 3% of the fast neutron dose.

3) McCallH has shown that the fast neutrons can be divided into a direct component

and a component scattered from the walls. The

scattered component accounts for approximately

25% of the total dose.

4) The neutron dose becomes fairly constant at 30 to 50 cm from the beam edge. The

"in-beam" data is 4 to 5 times higher than this value. At 5cm which is often reported, the neutron dose ratio "in-beam" to "out-beam" is

2 to 3.

In Figure 1 an attempt has been made to -ummarize the data for measurements made out of

the beam. Where possible the data at distances

greater than 30cm from the beam edge is plotted

since this tends to represent a uniform dose

levels. However, some data are only available

for 5cm from the beam edge. The data plotted is the neutron rad dose as a percentage of the photon dose in the main beam. In all cases this

number was obtained from the various papers using

data and conversion factor contained in the papers, except for two points where Q.F.'s were

used that were consistent with Q.F. used by

other authors for the same energy range. Also

plotted on this figure are the data from Swanson Oil for the yield of neutrons as a function of incident electron energy for Au & W targets. The curves were normalized to the neutron yield measured for the 10 MV x-rays.

135
"Out-of-Beam" Data

95 cm

100.5 cm0 *

®141cm

1 00.5 cm

5 cm

PI e

10 20 30

Accelerator Energy-MeV

Figure 1 "Out-of-Beam" Data. The data points are keyed to the numbers In Tables 1 through V. By each data point the target detector distance Is given, the distance of the detector

from beam edge, target material and, 1f necessary, assumed or derived Q.F. Also

plotted Is the data from Swanson on neutron yields from Au & W.

136
'In-Beam" Data

.1 P

(UoQ

T3

(0 QCco

o

.001

10 20 30

Accelerator Energy-MeV

40

beam" data, only the W curve 1s shown.

137
Although there is quite a wide spread in the

data some definite trends can be seen.

1) Below 20 MeV the neutron yields drops

off rapidly with decreasing photon energy. This agrees with the measurements that McCall made on the Siemens Mevatron XX where he noted a drop

in neutron dose by 4 in going from 18 MeV to 14 MeV and supplied by LaRiviere from the Clinac 20 operated at 15 MV and 19 MV.

2) There appears to be no difference

between betatrons and linear accelerators.

3) The percentages neutron dose reaches

maximum at around 25 MV x-rays and then remains

relatively constant with energy.

This last fact has great .significance since

25 MV x-rays have been used clinically for 30 years so that there should be a large amount of

clinical data available to determine if the neutrons have had an adverse effect on the patients, and if there is a time delay before

late effects show up it should also be apparent

in the clinical data.

In 1974 Oliver ^sl measured the fast neutron contamination in x-ray beams of medical electron

accelerators from 19 to 45 MV. This was updated

in 1976 and the data is shown in Table VI, excluding the points for the Sagittaire linear

accelerator and Siemens betatron reported above. The data was obtained with diode detectors and

is plotted on Figure 1 for comparison. This data

is consistent with the previous data and shows

the slightly higher values usually obtained

with the diodes.

Figure 2 is the data for in beam measurements.

As expected, it shows exactly the same trend as the out of beam data except the neutron dose

levels are several factors higher.

Conclusions

It is very apparent from doing this study

that some form of standardization is required for neutron leakage measurements. Just as dose

calibration procedures have been standardized

in terms of equipment, technique, experimental

set up and dose calculations, so the same approach must be applied to this problem.

Those standards should include as a minimum:

1) Typo of detector.

2) Calibration procedures and conversion

factor.

3) Experimental set-up.

4) Data presentation

and some form for handling the data. The

'cookbook' approach suggested by McCall might well be a good starting point for this. _ _ Recently recommendations were made to the NCRPl^U

that this subject should receive their consideration and it is hopeful that in the not too distant future guidelines and protocols will be available.

References

1. McCall, R.C., T.M. Jenkins and R.A. Shore:

'Transport of Accelerator Produced Neutrons in a Concrete Room',' .SLAC-PUB-2214, Oct., 1978.

2. McGinley, Wood, Mills, and Rodriguez:

"Dose levels due to neutrons in the vicinity of

high-energy medical accelerators". Medical Physics

1976, Vol. 3, p. 397.

3. McCall Associates: Neutron Measurements on

3 Mevatron XX Linear Accelerators, 1977 & 1978,

Supplied by Volker Stieber Manager, Development,

Siemens.

4. Gur, Bukovitz, and Gill: "Fast and Slow neutrons in an 18-MV photon beam from a Philips

SL/75-20 linear accelerator", Medical Physics,

1978, Vol. 5, p. 221.

5. 6. 7. Gur, Bukovitz, Rosen and Holmes: "Relative Measurements of fast neutron contamination in 18 MV Photon Beams from two linear accelerators and a betatron". Abstract

Medical Physics, 1978, Vol. 5, p. 350, Text supplied by Bukovitz.

8. 12. Fox and McAllister: "Fast neutrons from

a 25 MeV betatron". Medical Physics, 1977, Vol. 4,

p. 387.

9. Drout, Moreau, Quechon and Tabot: "Summary of Measurements Taken around a CGR/MeV Saturne Accelerator at Buc", C.E.A. C.E.N. Saclay,

Department of Radiation Protection Section of

Installations. Accelerator Group, May 1977,

obtained from AECL and CGR-MeV.

11. Wilenzick, Almond, Oliver, de Almeida: "Measurement of Fast Neutrons Produced by HighEnergy X-Ray Beams of Medical Electron Accelerators", Phys. Med. Biol., 1973, Vol. 18, p. 396.

13. Straker, E.A., P.N. Stevens, D.C. Irving, and V.R. Cain: "The MORSE Code - A Multi group Neutron and Gamma Ray Monte Carlo Transport Code",

Oak Ridge National Laboratory, ORNL-CFN-70-2-31

1970.

14. McCall, R.C., T.M. Jenkins, and G.D. Oliver,

Jr.: "Photon Response of Silicon Diode Neutron Detectors", Stanford Linear Accelerator Publ .

Stanford, SLAC-PUB-1783.

15. Weinstein, M.S., K. O'Brien, and F. Hajnal

"Procedures for Estimating the Errors in Neutron Spectra from Bonner Multi sphere Measurements",

HASL, Rep. No. TM-69-4, 1969.

16. Int. Comm. Radiol. Protect. (ICRP) Publ. 21, 1971.

17. National Council on Radiation Protection and Measurements (NCRP), Rep. No. 31, 1964.

18. International Commission on Radiation Units and Measurements (ICRU) Report No. 19, July, 1971.

19. Swanson, W.P.: "Calculation of Neutron

Yields Released by Electrons Incident on Selected Materials", Health Physics, Vol, 35,

1978, p. 353-367.

20. Oliver, G.D., Jr.: "Fast neutron contamination in X-ray beams of medical electron accelerators from 19 to 45 MV, Phys. Med. Biol., Vol. 19, 1974.

21. Almond, P.R., J.R. Cameron, M. Ehrlich, A.C. Lucas, R.H. Thomas, J.E. Turner, B.A. Glass: Report to the National Council on Radiation

Protection and Measurements, Sept. 12, 1978.

138
NBS SP 554 (1979)

DESIGN CRITERIA FOR PHOTON BEAM EXITS ON HIGH ENERGY MEDICAL ACCELERATORS

TO MINIMIZE NEUTRON PRODUCTION

Leonhard Taumann

Siemens Medical Laboratories, Inc. 2404 N. Main Street Walnut Creek, CA 94596

The major components of the photon beam exit of a medical linear accelerator

are discussed from the viewpoint of beam quality and neutron production. It

is pointed out that the optimization of the beam quality also leads to a minimum in the neutron production.

(Photon beam exit, beam quality, neutron production)

Introduction

From the viewpoint of reliability and beam

quality the design of the X-ray exit for

medical linear accelerators should fulfill

several main requirements.

1. The major components affecting the design should be uncomplicated and easy

to manufacture. Long assembly times

caused by complicated alignment proce- dures should be avoided. Parts which

have to be removed for beam testing

should be easily ajustable from outside

the head.

2. The required RF-power for the specified

beam energy and dose rate should be as low as possible.

3. From the users viewpoint the most im- portant aspect is the beam quality which

can essentially be determined by two considerations.

a. The energy distribution over the whole possible radiation volume

should be as uniform as possible

and the variation insignificant for

different field sizes.

b. The ratio between the tumor dose and the total delivered dose

should be as high as possible.

This also demands a low level of

that kind of secondary radiation

which can contribute to the total dose as whole body dose.

For linear accelerators with photon

energies higher than 10 MV the neutron production has become of major concern

as secondary radiation.

The magnitude of the RBE-factor which

has to be applied for this kind of low

level radiation is still in discussion.

Recommended numbers are changing within

a factor of hundred.

It 1s the intention to demonstrate that

the optimization of the beam quality

also leads to a minimum in the neutron

production.

Performance of the major components of the

beam-exit.

Fig. 1 shows the principal arrangement of the major components of the photon beam exit.

These are the target, the beam stopper, the flattening filter, and the fixed and movable

beam collimators.

139
Fig. 1

MEVATRON 20, Upper Defining Head, X-Ray Exit

Window

1. The Target.

Since there was no conclusive answer in

' the literature for the right target

material for a 15 MV photon beam, central

^ axis depth dose curves were measured for

\ the different target materials and

\ target thicknesses. These curves allow

\ sufficient relative comparison of the energy spectrum produced in the target.

No significant difference in the central

axis depth dose curves could be found

between targets with high and low atomic

charge number. No change in the beam

quality could also be found for targets with ticknesses between .015 radiation length and .4 radiation length.

Fig. 2 shows the depth dose curve for a rather thin aluminium- target and a thicker tungsten target. Ftg. 3 shows

three depth dose curves - one written over the other - for three tungsten

targets with thicknesses of .071 RL, .143 RL, and .214 RL.

As can be seen in Fig. 4 the half value

angle of the photon distribution in the forward direction changes only slightly

with the target thickness, approaching

assymptoticly a maximum angle.

The half value angle for aluminum is considerably smaller and also the total output about 20% lower. Comparing the same flattened field size the aluminum

target would require about '50% more beam current than the tungsten target.

7 10 12 15 DEPTH IN WATER tcm)

FIGURE 2

Central Axis Depth Dose Curves for 15 MV X-Rays

and Different Target Materials

SSD: 100 cm, Field Size: 10 x 10 cm^ Iron Absorber behind the targets

Target: Tiingsten .214 RL

Altiminum . 05 RL

10 12 15 DEPTH IN WATER (cm)

FIGURE 3

Central Axis Depth Dose Curves for 15 MV X-Rays

and Different Target Thicknesses

SSD: 100 cm, Field Size: 10 x 10 cm^ Iron Absorber behind the target

Target: Tungsten; thickness .071 RL

.143 RL

.214 RL

140
Thus a thin tungsten target of a thickness of about .2 radiation length appears to be the best choice. It can

be shown later that the neutron production in such a thin target is almost

negligible compared with other major

neutron sources in the defining head.

The beam absorber.

The accelerated electrons lose about one fifth of their energy in the target by

collision. In order to absorb them

completely a beam absorber has to be situated behind the target. This

absorber has four different tasks:

a. Absorb the electrons

b. Harden the beam

c. Produce a minimum of x-rays

d. Produce a minimum of neutrons

It is obvious that low z-material is the best choice. The space which is left in the target slide behind the target is therefore filled with a solid graphite

plug. Since the electrons are not absorbed completely in the carbon, the beam absorber area is extended into the primary collimator. For mechanical

stability an aluminum plug has been

chosen instead of carbon. This aluminum

plug also absorbs low energy radiation

which is produced in the wall of the primary collimator by Compton Scattering.

Due to its low atomic number, the neutron production tn the electron beam absorber

Is minimal

.

The flattening filter.

For low neutron production and sufficient beam hardening material with a low

atomic number has to be considered.

Fig. 5 shows three central axis depth

dose curves:

a. One has been measured for a .214 RL thick tungsten target with a piece

of iron-acting as beam absorber - behind the target.

b. The other curve has been measured

for the same target absorber arrangement. In addition a piece of lead

with the thickness of one inch has been placed behind the iron in order to simulate the beam hardening

effect of a lead flattening filter.

c. The third curve shows a central axis depth dose curve for the same target absorber arrangement but the lead has been replaced by a piece

of carbon of the same thickness.

.5.5

4.5

Tungsten

"Aluminum

.1 .2 .3 RADIATION LENGTH (g/cm^)

FIGURE 4

Half Value Angle of a 15 MV X-Ray Beam as Function

of the Target Thickness

Target: Tungsten; thickness .0714 RL

.107 RL

.143 RL

.178 RL

.214 RL

100

soso

ft' 70

i 60

I 50

" 40-

30

20

10

TUNGSTEN LEAD TUNGSTEN CARBON TUNGSTEN

7 10 12 15 DEPTH IN WATER (cm)

FIGURE 5

Central Axis Depth Curves for 15 MV X-Ray

s

and Different Beam Hardener

SSD: 100 cm, Field Size: 10 x 10 cm^ Target: Tungsten, .214 RL thick For beam hardening: lead block 1 inch thick

carbon block 1 inch thick

141
It can be seen that the lead filter is still hardening the beam spectrum but

the hardening effect of a low z-material

of the same effective thickness is considerably higher.

The length of an aluminum flattening

filter which has to flatten a field size

of 35 cm X 35 cm of a 15 MV photon beam

is about 5.3 inches. There is just not enough space in the defining head for

such a flattening filter. Also the shape

of this filter is so steep that ex- tremely small mechanical movements would

considerably affect the symmetry of the dose distribution for small field sizes.

So the best compromise for the choice of

the material of the flattening filter is stainless steel.

Due to its shape the compensation filter

is hardening the photon beam in the central axis considerably more than outside the central axis. In order to find out how this effect changes the energy distribution over the whole

radiation volume central axis depth dose curves have been measured with and without flattening filter for different

field sizes. (Fig. 6 and Fig. 7).

There is no significant shift of the maximum of the depth dose curves for

both sets of curves. The change of the dose at 10 cm depth as a function of the field size with and without flattening

filter is of the same amount. The

hardening effect of the flattening

filter is compensated by the influence

of the larger effective field sizes. In the area close to the surface the favorable influence of the flattening

filter can clearly be seen for the 5 cm

X 5 cm and 10 cm x 10 cm field.

If the medium photon energy is degraded

by photons which are scattered at the moveable collimators a measurable change

in the central axis depth dose perfor- mance for different special collimator

openings should be seen. A 5 cm x 5 cm depth dose curve has been compared

against a curve of a 5 cm x 35 cm field

with the collimator pair which is close

to the focus is opened to 35 cm and the collimator pair which is at a larger

distance from the focus is closed to

5 cm. Another pair of curves has been measured with the opposite arrangment

of the collimator pairs. (Fig. 8 and

Fig. 9). No measurable difference could

be seen.

3 5 7 10 12 15 DEPTH IN WATER (cm )

FIGURE 6

Central Axis Depth Curves for 15 MV X-Rays

as Function of the Collimator Opening

SSD: 100 cm, Field Sizes: 5x5 cm^ 10 X 10 cm^

20 X 20 cm^

35 X 35 cm^ Target, Electron Absorber and Compensation Filter

Arrangement according to Figure 1

DEPTH IN WATER (cm)

FIGURE 7

Central Axis Depth Dose Curves for 15 MV X-Rays

as Function of the Collimator Opening

SSD: 100 cm. Field Sizes: 5 x 5 cm^ 10 X 10 cm^

20 X 20 cm^

35 X 35 cm^ Without Compensation Filter

Target and Electron Absorber Arrangement

according to Figure 1

142
100 90-

80- ^o- >•S60-

zTso

z

~ 40-

JO20

10

inner jows x outer jows 5 « 35 cm2

5 « 5

"

7 10 12 15 DEPTH IN WATER (cm)

FIGURE 8

Central Axis Depth Dose Curves for 15 MV X-Rays

as Function of Different Collimator Openings

SSD: 100 cm. Field Sizes:

5x5 cm^ (inner jaws x outer jaws)

5 X 35 cm^ (inner jaws x outer jaws)

X-Ray Exit according to Figure 1

100

90

80

If

>. TO

I- 60- u

|50-

40

30

20

10-

ouler joMt > inner jowt 5 « 35 cm^

5 « 5

-

10 12 15 DEPTH IN WATER (em)

FIGURE 9

Central Axis Depth Dose Curves for 15 MV X-Rays

as Function of Different Collimator Openings

SSD: 100 cm, Field Sizes:

5x5 cm^ (outer jaws x inner jaws)

5x35 cm^ (outer jaws x inner jaws)

X-Ray Exit according to Figure 1

The results are summerized in Table I. The carbon-aluminum absorber has the highest influence on the surface dose. This low z-material gives the thin

tungsten target the chance to perform as

a real target. The beam hardening

effect of the stainless steel flattening

filter can be seen over the whole

depth. But the change in the energy distribution is within tolerable limits.

The neutron-production

The overall neutron production for the

15 MV photon beam of the Mevatron 20 is about .25 mrem per photon rad. No significant deviation could be found for all machines installed. The neutron distribution around the defining head is more or

less isotrophic.

In order to find out how much the target contributes to the whole

neutron production, measurements

were done for three different

done for three different target thicknesses. The extrapolation against zero target thickness shows that the neutron production 1n the target is in the order

of only about 5%. How much does the beam absorber and flattening filter contribute to the neutron production? Only 20% of the

total photon intensity produced in the target is spread in the forward direction into the solid angle covered by the

useful 1 beam. Considering the lower

atomic number and the additional dis- tance of the flattening filter from the

focal spot the neutron production is insignificant.

Almost 95% of all the neutrons are produced in the shielding device of the head which is mainly tungsten. The neutron yield for tungsten in the energy

range of 15 MV changes rapidly with

energy.

Fig. 10 shows the relation between the bending magnet current and the electron

energies at the exit window. These

electron energies were determined by

measuring the extrapolated ranges. According to this curve the electron energy applied at the target is 13.9 MeV

for producing a relative central axis

depth dose of 77.5% at 10 cm depth for a field size 10 cm x 10 cm.

An Increase of the electron energy of

only one MEV would cause a higher neutron

flux of a factor of about 1.7. Since

also the medium energy of the neutrons penetrating the head is higher, another

20% has to be added to the dose equivalent.

143
R«l«tlT« Cantral Axla Depth Doat Curv* R«Bb»ra for Dlffarant Target- Beaaabsorber- Conpensatlon niter Arrangeaenta and Colllaator Openings.

Dep

1

th in Water (cn)

10 15

Fit)ld slz

10*10

e ( CD

20»2o

)

35''35

fie Id size

10«10

(cm^)

20«'2C 35»35

ri eld si]

L0«10

se ( cm'

20»20

-)

35*35

Tungsten Target with

Iron Ibeorber 91.3% 7f».6% 59.7%

AlualniuB Target with

Iron Absorber 91.6% 73.8% 59%

Tungsten Target with Carbon-Alumlnlua Absorber

82.1% 85.it% 85.'»% 86% 73.2% 714.8% 75.8? 76.5% 57.1% 59.5% 60.8% 61.6*

Tungsten Target with Iron Absorber »! Inch Lead 88% 75.'t% iO.9%

Tungsten Target with Carbon- Alualnlum Absorber and Iron Conpensatlon Filter

76% 80% 85% 85.7"^- 76% 77.1»% 78.1*% 79.1*% SO. 5% 62.7% 6/».it% 66%

This means that in this photon energy

range one MeV change in the electron

energy hitting the target is changing

the neutron production within a factor

of two. Therefore, an optimal beam

hardening can considerably minimize the

total neutron production.

ELECTRON ENERGY (MeV)

FIGURE 10

Electron Energy as a Function of the Bending Magnet Current

144
NBS SP 554 (1979) NEUTRON LEAKAGE CHARACTERISTICS

RELATED TO ROOM SHIELDING

E. Tochilin

and

P. D. LaRiviere

Varian Associates, Inc. Palo Alto, CA 94303

ABSTRACT

A systematic study of neutron leakage rates has been made over a range of x-ray energies 10-19 MeV in one of Varian's test cells. Leakage was evaluated at one meter

and in the patient plane. Measurements were presented of neutron and gamma ray leak- age in the concrete test cell, together with door shielding calculations. The information allows scaling of neutron leakage to other size rooms and x-ray energies ex- tending to 25 MV. The neutron activation of concrete and its contribution to treat- ment room dose was investigated. Gamma rays from radiative capture in concrete and

j their influence on leakage through the exit door were also examined.

I INTRODUCTION

Within the past few years considerable pro- gress has been made in understanding the modifying effects of x-ray shielding material on the photoneutron spectrum from high energy medical acI celerators. The mean energy of leakage neutrons

I is considerably lower than the values reported in

I some of the earlier literature. This leads to a lower rem dose and less neutron shielding material for the treatment room.

A considerable amount of neutron leakage

data has been compiled at Varian on neutron leakage from a generic class of medical accelerators

operating a x-ray energies from 10 MV to 19 MV. These include experimental measurements of neutron leakage vs. energy, both within and outside

the patient plane, the average energy of leakage

neutrons, radiation leakage at the entrance door

to the treatment room, and room activation.

The broad objectives of the work reported

here were to elucidate the mechanisms of neutron generation, shielding and interactions with the test cell environment so that machine design could

be improved, and the production of capture and

I residual gamna radiation reduced in the most ecoI nomical fashion.

I

1 Two accelerators were used in this study;

namely, the Clinacs ® 18 and 20. The Clinac 18 produces a beam of electrons ranging in energy

from 6 MeV to 18 MeV and a photon beam at 10 MV. The Clinac 20 produces electrons ranging from 6 MeV to 20 MeV and a single photon beam either at

15 MV (standard) or at 18 MV (optional). A sys- tematic study of neutron leakage was also made

with the photon beam energy increased to 19 MV.

I All measurements were made with the beam flatI tened to meet medical treatment specifications.

Unless otherwise specified, the movable collimators were closed in order to eliminate neutron contamination from the primary beam.

The Clinac 20 medical accelerator is an upgraded version of the Clinac 18. One major modification has been the design of a new more ef- ficient standing wave accelerator waveguide to allow the production of higher electron energies.

Accelerator waveguides for both the 15 or 18 MV versions are approximately the same size.

By positioning the Clinacs at exactly the same location in the same cell we were able to determine the relative neutron leakage of 10, 15, 18 and 19 MV x rays under identical geometric

conditions. The neutron leakage accompanying 25 MV X rays produced by the Clinac 35 in another

test cell is also included.

INSTRUMENTATION

Two neutron detectors were used; namely, a neutron flux detector and a rem counter. Fast neutron flux inside the room was determined by

measuring the activation of indium foils placed

in the center of a cylindrical polyethylene shield

6.25 in. in diatieter and 6.25 in. high. A cadmium outer cover surrounded the polyethylene mod- erator and absorbed any incident thermal neutrons. The detector measures neutron flux approximately

independent of energy in the range from 20 keV to 10 MeV (1). The rem detector consisted of an Andersson-Braun counter (2) with a scaler readout

and was used to measure neutron dose equivalent

rate. Use of this instrument was primarily con- fined to leakage measurements at the door and

within the maze leading into the test cell. A com- parison between the neutron flux and rem de- tectors was also made in the room. In this case

the rem counter readings were corrected for counting losses produced by the pulsed beam. Both neutron detectors were calibrated with a Pu-Be neutron source. Thermal neutron fluences were estimated with bare indium foils, but the results are not reported here because the dose

equivalents were generally negligible.

X-ray leakage measurements were determined

145
by multiple dosimeters which included a pancaketype ionization chamber with a window area of

100 cm2 and Keithley electrometer, LiF TLD dosimeters, Landsverk 2R ionization chamber dosimeters

and Kodak Type AA 14x17 in. Ready Pack radiographic film. The Type AA film, ion chambers and

TLD dosimeters were all calibrated against the

100 cm2 ion chamber which was calibrated directly in the primary x-ray beam. Exposures were

made on a one meter arc positioned around the Clinacs. The 14 in. x 17 in. film was sand- wiched between two sheets of plastic each 0.10 in. thick. The outside of each plastic sheet con- tained a 1 in. X 2 in. lead strip 0.030" thick,

duplicating the conditions used to calibrate the

films. Two TLD dosimeters and two 2R ionization

chambers were attached to the back plastic sheet next to the lead strips comprising the film dosimeter. This arrangement allowed 3 independent measurements of leakage radiation at each location, together with information on the possible

existence of any significant gradients over the area of the film pack. The 100 cm^ chamber was routinely affixed to a marked location on the

back of the arc, thereby facilitating comparison

of film and ion chamber readings.

The one meter arc allows simulataneous po- sitioning of the films at a distance of one meter

from the target as shown in Figure 1. The arc swings in azimuth about a vertical axis coincident with the central axis of the x-ray target.

The arc is swung in angular increments of 25° for each exposure, until virtually 47r coverage is obtained. The aximuth angles requisite for complete coverage are shown schematically in Fig- ure 2, along with the vertical layout of the

films.

HORIZONTAL

A. FILM LAYOUT ALONG ARC. VERTICAL PLANE. FILM NUMBERS IN CIRCLES.

C 1^ 0

L I K

B. ARC POSITIONS, HORIZONTAL PLANE THROUGH TARGET - PLAN VIEW

Fig. 2. Arc positions and film locations on the one meter arc.

A horizontal angle of about 40° is excluded

by the gantry, but supplementary measurements can

be made in the patient plane and at the rear of

the stand to close most of this excluded area. Calibration of the three dosimeters with ^^Co gamma rays have yielded consistent results for leakage radiation over an x-ray range of from 2 to 10 MV within + 15%.

Neutron leakage can also be measured on the meter arc by securing the flux detectors at various locations. Because of their large physical

size the flux detectors do not allow a high de- gree of spatial resolution. This requirement is less important for neutrons than for x rays since

the neutron distribution is approximately isotropic.

Fig. 1. The one meter arc mounted on the Clinac 6X.

146
LEAKAGE NEUTRON SPECTRA

i The photoneutron spectrum has frequently

been reported in the literature to closely re- semble fission neutrons. While this approximation generally applies to the primary spectrum,

inelastic scattering in the shielding material

surrounding the target, normally lead and tung1 sten, results in a significant degradation of neu-

! tron energy. McCall et al. (3) have calculated

' the neutron leakage spectrum of photoneutrons

j from 15 MV photons, when shielded with tungsten.

! Figure 3 from their paper shows the integral spectra for a bare source and from one surrounded by

a 4 in. thick tungsten shield placed in the center

of a concrete room. A comparison is made with

the 252cf fission neutron spectrum. The figure

clearly shows that the fission spectrum is a poor

: approximation for leakage neutrons from a medical

I accelerator. The figure further illustrates the

I difficulties involved when threshold detectors

' are used to evaluate neutron leakage dose under

I these conditions, since most of them are com- pletely insensitive to neutrons below 2 MeV.

0,1 I 10

i. ,0-7 E (MeV) >....,.

' Fig. 3. Comparison of various neutron

integral spectra: ^^^Cf, 15 MeVWPN-Bare,

15 MeVWPN-4"W, 15 MeVWPN-4"W in a concrete room. McCall has determined average neutron energies of 0.45 MeV and 0.64 MeV for 15 MV and 25

MV photoneutrons, respectively, degraded by trans- mission through 4 in. of tungsten shielding. He defines an effective neutron energy, Egff, as the energy of a monoenergetic neutron beam which

would have the same conversion factor from fluenoe

I to dose equivalent as the spectrum with an averI age energy, t. It turns out, for calculations

I covering a wide range of typical conditions, that

Thus for 15 MV X rays from the Clinac 20,

Epff =0.27 MeV, and the conversion factor derived from Figure 14 of ICRP Publication 15 and

21, (4) is 8 X 10' n cm"^ rem-l. For 25 MV

X rays, Eg^f = Q.38 MeV and the factor is 6.2 X 10^ n cm" rem"^.

An independent check on the average photo- neutron leakage energy was obtained from absorption curves in polyethylene. For these mea- surements, indium foil detectors were centrally

interposed in a stack of 0.25 in. thick square polyethylene sheets 16 in. on a side. The sur- face of the stack, which was approximately 7 in. deep, was centered on the beam axis at the i so- center, and the jaws were adjusted as desired.

The thermal activation of the foils, once equilibrium conditions are met, is proportional to the fast neutron fluence, thereby providing a convenient method of measuring fast neutron at- tenuation.

Energy calibration was obtained by reference to a family of curves obtained with mono- energetic neutrons from a 2 MeV positive ion ac- celerator utilizing the T (p,n)-^He and D (d,n)3He

nuclear reactions (5). Thermal neutron distributions with indium were obtained in a 15 in. diameter polyethylene cylinder for neutron en- ergies ranging from 0.25 to 5.1 MeV. These curves are shown in Figure 4. Two trends are observed:

first, the depth at which maximum thermal ization occurs increases progressively with neutron en- ergy; and, second, the slope in the descending

region decreases exponentially with increasing

energy. 100 I ,

I I I I I I -X 1 1.—

I

6 5 10 15 20 25 DEPTH (CM) ^ , Fig. 4. Energy calibration of a modified

long counter by indium foil activation.

147
ergy; and, second, the slope in the descending

region decreases exponentially with increasing

energy.

For the closed jaw condition, the average en- ergy of the neutron leakage through the 6 in. of

tungsten of the Clinac 20 was determined to be 0.37 MeV. This value is consistent with the aver- age energy obtained by Monte Carlo calculations

of 15 MV x-ray photoneutrons through 4 in. of

tungsten, which was 0.45 MeV. The experiment was repeated .with 15 MV photons. Within experimental

error, the indium activation curves were identical to those obtained with 10 MV photons, indicating essentially no change in leakage neutron

energy through tungsten for photoneutrons pro- duced at the two energies.

The same measurements made in the primary

10 MV photon beam, for jaws open to a field size

of 20 cm X 20 cm, gave an average neutron pnergy

of 0.73 MeV.

One additional estimate of neutron leakage energy was obtained by shielding the head of a Clinac 20 with a 2.25 in. thick polyethylene box. This experimental structure was suspended over

the accelerator head by means of a crane. The sides of the box extended down to the polyethylene shielding in the Clinac used to minimize neutron leakage in the patient plane. Measurements

were made through the sides of the box by activation of indium in the fluence detectors. The

geometry was essentially broad-beam with the

sides of the box interposed between the target

shielding material (primarily lead) and the de- tectors. For 19 MV photons the measured TVL was

3.2 in. + 9%. NCRP 38 (6) gives a TVL of about

3 inches for 0.5 MeV neutrons normally incident

on polyethylene slabs. It is quite evident that

the average leakage neutron energy for the Clinac

20 does not exceed 0.5 MeV.

NEUTRON and X-RAY LEAKAGE FROM THE CLINAC 20

Leakage in the Patient Plane

Extensive neutron measurements have been made on the Clinac 20 at photon energies of 15,

18 and 19 MV. These included leakage at one meter from the target and leakage in the patient plane.

Particular attention was given to minimizing leakage radiation to the patient. This region is de- fined as a circular plane of two meter radius

(excluding the area covered by the primary beam)

perpendicular to the central beam axis at the

normal treatment distance. Neutron shielding was provided to insure that the combined x ray and neutron dose in the patient plane did not exceed

0.1% outside the useful beam area.

Neutron measurements were made with the flux

detectors previously described. The neutron data were reduced to rem/rad percent using the dose

equivalent conversion factors in ICRP 21 (4) and

an effective energy of 0.27 MeV calculated for a 15 MV x-ray spectrum shielded with 4 in. of tungsten. An effective neutron energy of 0.31 MeV

was used for leakage neutrons from 18 and 19 MV

X rays. A typical set of measurements obtained

with 15 MV photons will be used to illustrate the

procedures used.

Neutron and x-ray measurements were made at

a number of locations with shielding identical to that used on the Clinac 18. This allowed a direct

comparison of neutron fluence per primary x-ray

rad vs. energy. Neutron leakage in the patient

plane was measured at 25 cm intervals from i so- center along two perpendicular axes, with and

without neutron shielding. X-ray leakage was determined with the 100 cm^ ion chamber and Type

AA film packs. The combined neutron plus x-ray leakage is plotted in Figure 5 for 15 MV x rays. The points are the average of the observations

obtained along the two perpendicular axes. Fig- ure 5 shows the combined x ray plus neutron per- centages as peaking at 25 cm, the critical radius

just outside the useful beam. At this location

the neutron leakage averaged 0.030 percent rem/rad

and the x-ray leakage was 0.018%. The maximum

total leakage was, therefore, 0.05 percent

rem/rad.

-TREATMENT AREA

UJo

<<

Ul

u

U

0.01

100 120

DISTANCE FROM ISOCENTER (cm)

Fig. 5. Neutron and x-ray leakage measurements in the patient plane of a Clinac 20: 15 MV x rays, 100cm SSD. made in isted in isoobi stances

i stance

200 cm

The cr it- neutron

The results of similar measurements

the patient plane for 19 MV x rays are 1 Table I for distances 25 to 200 cm from center. Once again, maximum leakage was tained at 25 cm, decreasing rapidly at d beyond 75 cm because of inverse square d from the source. The neutron leakage at

is four times less than that at 25 cm.

ical measurement for determining maximum

148
leakage in the patient plane would be just at the

I circular limit of the useful beam.

It is reiterated that these measurements

were made with the jaws closed since the intent

was to measure leakage . Much higher values would

have been obtained, had the jaws been open, from neutrons scattered by the open collimator system.

Analogously, Rawlinson and Johns have shown a similar increase in x-ray leakage with jaws open (7).

Leakage at One Meter

X-ray leakage for 10 and 15 MV photons was determined with the normal complement of dosimeters positioned on the meter arc. Both accelerators showed very low leakage, averaging just over 0.02%. The overall design objective was not

j to exceed 0.05%. Limited measurements with 18 MV photons gave similar results.

I Neutron leakage measurements on the one meter

arc were obtained with flux detectors at vertex

angles of 0°, 45°, 90° and 135°. Table 2 gives

the neutron leakage obtained at these locations

for 15 MV photons. The neutron dose equivalent

was obtained assuming an effective energy of 0.27

MeV. X-ray leakage at the corresponding locations

was not over 0.04%. It can be seen that the total leakage did not exceed 0.1% at any point. TABLE I

FAST NEUTRON LEAKAGE IN PATIENT PLANE

DOSE RATE: 500 RADS/MINUTE

rem/rad (%) for 19 MV Photons

Location

(cm)

025

5075

100 125

150

175

200

Neutron Shielding

on Clinac

0.10

0.076

0.063

0.055

0.035

0.024

0.027

0.022

0.019

No Neutron Shielding

on Clinac

0.14

0.14

0.14

0.13

0.08

The accuracy of x-ray dosimeter leakage measurements is believed to be within + 15%. The neutron values should probably be assigned a some- what greater uncertainty since the effective en- ergy is not precisely known and may differ from

one location to the next. This uncertainty

arises because of the amount of shielding material and its composition (lead, tungsten and a combination of both) varied with location.

From neutron measurements taken at one meter

we have compiled the average neutron fluence at

TABLE 2 NEUTRON LEAKAGE AT 1 METER FROM TARGET

DOSE RATE: 500 RADS/MINUTE

rem/rad (%) for 15 MV Photons

Angle from Zenith

Setti nq u

/ic;0 HO yu 1 -icO

1 35

A 0.050 0.031 0.028

N 0.052 0.029 0.030

C 0.053 0.054 0.029

L 0.052 0.055 0.030

E 0.035 0.037 0.024

I 0.039 0.043 0.026

G 0.022 0.030 0.028 0.025

one meter for 10, 15, 18 and 19 MV photons taken

with the accelerator located at identical positions in the test cell. The data has been ex- tended to include neutron leakage from 25 MV photons, obtained with the same detectors from

earlier measurements on a Clinac 35 in a some- what different geometry. The neutron yield is plotted in Figure 6 in n-cm-2.rad"' and in percent

rem/rad. Conversion from n/cm^ to neutron rem

was obtained by our best estimate of effective

neutron energy for each point.

ELECTRON ENERGY (MeV)

Fig. 6. Neutron leakage at one meter from

a number of Clinacs with no neutron shielding.

149
All neutron flux measurements made on the Clinac 18 and 20 were accompanied by neutron

leakage measurements made outside the entrance

door with an Andersson-Braun rem meter. Door leakage in mrem/hr was found to be directly pro- portional to the neutron flux at one meter. This

suggested that the neutron energy at the door

after scattering down the maze was relatively

constant.

Neutron Leakage Down the Maze

The engineering test cell partially shown in Figure 7, consists of a 20 ft x 20 ft test enclosure, 15 ft high. The entrance maze is 6.5 ft wide and 15 ft long with a 3.5 ft concrete wall between the maze passageway and the central room. The cell door consisted of 3/8 in. lead laminated

between 4 in. of wood and an outer sheet of cadmium. The cadmium probably has a minimal in- fluence on the neutron dose rate.

Initial measurements with the Clinac 20 in- dicated that the existing door provided insufficient neutron attenuation. Two 2 in. thick

polyethylene doors were added in the maze at the

location shown in Figure 7. The doors were displaced by two feet to allow easy personnel access. The doors were hinged to swing open for installation and removal of large equipment. All measurements were taken with the gantry positioned vertically down (180°). Measurement

points numbered 1 through 11 were placed one meter apart except where noted otherwise.

Measurements made along the line including

points 5A through 11 were made with the flux and

rem counters and thermal neutron foils. The flux

and rem values remained essentially constant at points 5, 5a, 3 and 6, all located within the treatment room. The dose rate at these locations

was several times greater than that calculated by

inverse square. This indicated that the fast neu- tron fluence and dose at these locations was principally from wall reflected neutrons, a finding

in good agreement with McCall's calculations (3).

The thermal neutron fluence was also constant

over this region. At locations 7 and 8 (polyethylene doors open) the dose rate and fluence

dropped markedly, decreasing at point 8 to approximately 0.2 the values obtained in the room. Be- yond that point the dose rate decreased exponentially, decreasing by one-half for each 1.5 meters of maze length. On the other hand, the thermal neutron fluence decreased gradually with the thermal flux at point 6 only 2.5 times greater than

at point 10, three meters away. These measurements indicate that the neutron energy is significantly degraded down the maze. NCRP 51 (8) states that for multi-leg mazes neutrons beyond

the second leg may be considered to be completely

thermal ized.

Table 3 gives the neutron dose equivalent

dose rates measured at the open doorway of the

test cell for a primary dose rate of 500 rads/min

or 3 X 10^ rads/hr. The accompanying dose rates from scattered x rays and from capture ganma rays are also given. A further discussion of capture

4.6

Fig. 7. Concrete test cell used for measurements and calculations.

TABLE 3

NEUTRON AND X+G DOSE EQUIVALENT RATES

AT ENTRANCE TO CELL

(No neutron shielding on machine)

Photon mrem/hr

Energy

(MV) Neutron X + G Total

10 6 3 9

15 54 23 77

19 162 48 210

gamma rays appears in the next section. The

values in Table 3 may be scaled to rooms both

larger and smaller than this cell, including

rooms with shorter mazes.

Based on Clinac 20 measurements made at Varian, the following equations were developed

to allow calculation of the door attenuation.

Here we assume a built-in lead thickness of 3/8

in. in a polyethylene door.

150
XG: I/Ijj = 0.33 X 10

N: I/Iq = 10-*/5-°2

•t/32

Where t is the thickness of polyethylene

in cm

TABLE 4

.COMPOSITION OF CORE SAMPLE AND TYPICAL CONCRETE

Continuing the example above, for a door

thickness of 3 in. (7.62 cm) of polyethylene in the cell and 15 MV x rays, we obtain

XG = 23 X 0.33 X io"7*^2/32 ^ ^^^^^^

N

= 54 X 10-7-62/5.08 ^ 1.7 mrem/hr

6.1 mrem/hr

NEUTRON ACTIVATION AND GAMMA RAY PRODUCTION IN CONCRETE

Induced Radioactivity in Concrete

Upon initial operation of the Clinac 20 it was observed that the warning light on the radiation monitor inside the cell remained on after

the accelerator was turned off. A radiation survey of the room showed the readings to be uniform throughout the room, with slightly higher

readings in the immediate vicinity of the accelerator. This observation was in accord with ex- pectation for an enclosure emitting gamma radiation uniformly from the walls; namely, that the

dose rate is uniform throughout the room (9).

A decay curve of the gamma-ray dose rate was obtained by placing a survey meter in the treat- ment room immediately after a 10 minute irradiation. Readings were followed for a period of 40

minutes at which time the dose rate could no longer be measured. In order to obtain more pre- cise results a small concrete disk was cut from a core sample obtained from a similar cell. This

disk was exposed to accelerator produced neutrons

and counted in a shielded beta counter. The disk

was exposed on top of the accelerator head between

two one-inch thick polyethylene slabs. Placing

the sample 180° behind the direction of the pri- mary beam virtually eliminated any possibility

of photoneutron reactions; and the polyethylene

moderators enhanced the thermal neutron fluence.

Part of the core sample was sent to a commercial laboratory for a chemical analysis of the major elements found in concrete. The analytical

results are reproduced in Table 4 together with

the composition of a typical concrete (10).

A resolution of half lives obtained from the decay measured in the laboratory counter isolated

five components, three of which were gamma emittfers; namely, 28ai (t = 2.3 min),

49ca (T = 8.8 min) and 24Na (T = 1 5 hr).

From the chemical abundance of these elements,

their thermal neutron cross sections, half lives,

and specific gamma-ray constants a gamma-ray

curve was constructed simulating a ten-minute ex- posure to thermal neutrons. Good agreement was obtained between the experimental decay curve obtained in the cell and the calculations based on

10^^ atoms/cm^

Element

1 yp 1 ca 1 Concrete

Co re Sampl

0 4.73 _

H 1.73

Si 1.57 1.477

Ca 0.26 0.261

Al 0.17 0.219

Fe 0.053 0.110

Na 0.028 0.109

K 0.028

Mg 0.013 0.076 core sample measurements.

Gamma Rays from Radiative Capture In Concrete

The energy released upon ca tron is the binding energy of th order of 8 MeV, less the energy radioactive products that may be radiation ceases as soon as the turned off. Its importance lies

this is the neutron-induced gamm

is not only detectable at the ce overwhelms normal x-ray leakage

large as 20, with 19 MV x rays.

pture of a neue neutron, on the expended by any

formed. This

accelerator is in the fact that

a radiation that

11 door, but which

by factors as

By a process similar to that used to determine induced activity dose rates it was possible

to determine a dose weighting factor, or R/capture, in concrete. This factor was derived from

the measured gamma ray spectra for each element

(11), and the specific gamma ray constant for each energy (12). With this information it was possible to calculate the approximate gamma ray spectrum for the core sample. The derived spec- trum is plotted in Figure 8 together with a similar spectrum from "type 01 normal light con- crete" (13). The two spectra agree very well in their major characteristics. The average energy of this spectrum was calculated as 3.6 MeV. The primary dose contribution was from the sil- con and iron present in the concrete, followed by

hydrogen and calcium. The relative dose from

these materials was 30, 23, 22 and 14 percent,

respectively.

Estimated Capture Gamma Radiation at the Cell Doorway

The Clinac 18 operating at 500 rads/ min

typically produces x-ray leakage levels of 3 mR/hr

and 0.8 mR/hr at the doorway of the cell with the door open and closed respectively. Since the pro- duction of neutrons at 10 MV is negligible, we may assume that these levels are due entirely to

151
CELL CONCRETE, CALCULATED

O WALKER AND GROTENHUIS

ANL- 6443 (1961)

>

UJZ

<r

111 0.z

UJo

(E

tiJ a.

PHOTON ENE9f?Y (MeV^

Fig. 8. Raaiat'ive "capture gamma-^ay spec- trum from concrete.

leakage x rays scattered down the maze.

The Clinac 20, positioned at the same location and provided with the same shielding, pro- duced 48 mR/hr and 18 mR/hr under similar con- ditions when operating at 19 MV. Since x-ray scattering conditions were identical, it is ap- parent that the greatly increased doorway leakage

was due primarily to capture gamma rays resulting

from thermal neutron captures in the concrete

walls. The neutron dose equivalent at the open

door was 162 mrem/hr, approximately three times

greater than the ganma-ray dose rate. Neutron to gamma ray ratios obtained at the entrance maze of

other installations operated at higher and lower

energies ranged from values of two to three. While the observations were strictly empirical

they did point out the additional shielding com- plications imposed by this high energy gamma-ray

component.

It should be recognized that prescribing

lead for ganma shielding has severel imitations.

Usually the door becomes so massive that motor as- sistance becomes mandatory and a host of potential

mechanical problems are generated.

In summary, it was estimated that the capture

gamma-dose rate in the cell was about 270 mrem/hr

under typical operating conditions. For comparison, the leakage x-ray dose was -6000 mrem/hr at

1 meter, for the same conditions, and the residual initial radiation level was 1 to 2 mR/hr,

for operation at 18 MV.

Treatment Room Dose due to Activated Concrete

Since 10 mrem per week is the allowable limit for non-radiation workers, initial activation

levels on the order of 1 mrem/hr will produce mea- surable exposures of technologists who might go

in and out of a treatment room some 40 to 50 times

a day. As a preliminary to calculation of doses

incurred inside a room by hospital personnel. Fig- ure 9 was prepared illustrating the dose rates to

be expected following 500 rad/min irradiations

for 1, 5, 10, 50, 100 and 500 min at 18 MV. The

activation products are 28A1 and ^^Ha. The con- tribution from '^^Ca was negligible. These curves clearly show the influence of irradiation time. At 1 min, the initial decay is governed by the

2.3 min half-life of 28ai . As the irradiation

time is increased there is a progressive increase

in the importance of the 15 hr 24|\ja.

I0.O

a:

13V)o

Q.X

UJ<

2

<

(!)

T

1 500 min.

0.01

100 min. ^

50 min.

10 min.

5 min.

I min. ± ±

10 20 30 40 50 60

Tl ME AFTER EXPOSURE (min)

Fig. 9. Calculated residual radiation from

a concrete room after several running times:

18 MV x rays, 500 rads/min.

We have used a simplified calculational model of a single one-minute treatment dose of 300 rad,

followed by a maximum of 9 min room exposure, re- peated 48 times a day for a weekly x-ray workload

of 7 x 10^ rad. The doses from room activation

by a Clinac 20 operating at 18 MV were compared for varying stay times of from five to 9 min as shown

in Table 5. In this calculation, it was assumed

that the technologists would enter the room as soon as possible. It is evident that potential

doses from activated concrete are the order of

10 mR/week. It, therefore, appears desirable to limit concrete activation to about the levels

shown.

152
TABLE 5

WEEKLY TECHNOLOGIST DOSE FOR INDICATED STAY TIMES

FOR 18 MV PHOTONS

mR

Stay Time 28„,

Al

24.,Na Total

(min)

9 8.57 5.82 14.4

8 8.35 5.18 13.5

7 8.07 4.53 12.6

6 7.67 3.88 11.6

5 7.14 3.24 10.4

SUMMARY

The practice of installing and testing different energy Clinacs at the same location in the

engineering test cell has provided an unique op- portunity for systemization of a wide variety of x-ray and neutron leakage data. This practice,

combined with the fact that the Clinacs 18/20 are equipped with identical x-ray shielding, pro- vided experimenters with the equivalent of a single machine with multiple x-ray energies of 10,

15, 18 and 19 MV. The more important findings ob- tained during thp«;e investigations are the fol- lowing:

1. The calculations presented by McCall and his collaborators (3) that we have been able

to check experimentally have proved to be re- liable. Two of these, for example, are a) the effective energy of the photoneutrons after degradation in lead and tungsten x-ray shielding,

and b) the neutron dose equivalent per x-ray rad

in the test cell proper and in the maze.

2. The maximum leakage in the patient

plane for the Clinac 20 at 15 MV was 0.030%

rem/ rad for neutrons and 0.018% for photons. The

total leakage was, therefore, just under 0.05%

rem/rad. Maximum leakage occurred at the periphery of the useful beam. The adjustable collimators were closed for these measurements in order

to insure that only leakage radiation was being

measured. Higher values were obtained at 19 MV,

where the maximum x and neutron leakage just outside the useful beam in the patient plane ap- proached 0.1% rem/rad. For both energies, the neutron leakage in other (unshielded) directions

was higher. than in the patient plane.

3. At 15 MV, activation of ordinary con- crete became easily detectable, producing radiation levels of approximately 1 mR/hr immediately after a few minutes of irradiation at 500 rad/min. The most important sources were tentatively identified as 2.3 min '^^Al and 15 hr 24Na

resulting from thermal neutron capture.

An estimate of technologist dosage from this

residual radiation in the treatment room was ~10

mR/wk for a workload of 7 x 10^ rad/wk for 18 MV

X rays.

4. Gamma radiation from radiative capture

reactions in concrete was identified as respon- sible for a 20-fold increase in the apparent x- ray leakage between 10 MV and 19 MV at the cell doorway. The principal dose contributors were calculated to be elemental silicon (30%), iron

(23%) and hydrogen (22%). The average energy of

the gamma-ray source spectrum was 3.6 MeV, which

is of course degraded by multiple scattering in the concrete.

5. The weakest link in cell shielding is the doorway. The information available indi- cates the major problem is suppressing capture

gamma radiation, rather than neutrons. An al- ternate to applying massive amounts of lead to the door is to exclude neutrons from the maze, thereby increasing the distance between the gamma emitters and the door. This approach has been tested successfully in the Varian engineering

test cell

.

REFERENCES

1. A. R. Smith, "A Cobalt Neutron-Flux Integratoi^" UCRL-9212 (1960), and L. D. Stephens

and A. R. Smith, "Fast Neutron Survey Using Indium-Foil Activation';" UCRL-8418 (1958).

2. I. 0. Andersson and J. Braun, Neutron Dosimetry, Vol 2, 87, IAEA, Vienna (1963).

3. R. C. McCall, T. M. Jenkins and R. A. Shore,

"Transport of Accelerator Produced Neutrons

in a Concrete Room," IEEE Transactions on Nuclear Science, Vol. NS-26, No. 1 (1979).

4. International Commission on Radiological Pro tection "Protection Against Ionizing Radiation From External Sources," ICRP 15 and 21 (1971).

5. E. Tochilin and B. W. Shumway, "Flux and Spectrum of Simulated Fission Neutrons,"

NRDL TR-448 (1960).

6. National Council on Radiation Protection

and Measurements, "Protection Against Neutron Radiation," NCRP 38 (1971).

7. J. A. Rawlinson and H. E. Johns, "Communica,tions". Medical Physics, 4^, 456 (1977)-

8. National Council on Radiation Protection and Measurements, "Radiation Protection Design

Guidelines for 0.1-100 MeV Particle Acceler- ator Facilities," NCRP 51 (1977).

9. W. S. Gilbert, K. Goebel , H. W. Patterson,

and A. R. Smith, "Concrete Activation Ex- periment at the Bevatron," UCRL-19368 (1969)

'nnpubl ished).

153
10. H. W. Patterson and R. H. Thomas, Accelerator Health Physics, Academic Press, New

York (1973).

11. R. E, Maerker and F. J. Muckenthaler, "Gamma

Ray Spectra Arising from Thermal Neutron Capture in Elements Found in Soils, Con- cretes and Structural Materials," ORNL-4382

(1969).

12. M. Barbier, Induced Radioactivity, John

Wiley & Sons, New York (1969).

13. Private communication R.C. McCall 9/13/78

Data from Walker and Grotenhuis, ANL-6443

(1967).

154
SUMMARY PANEL DISCUSSION

Moderator: J. S. Laughlin - Memorial Sloan-Kettering Cancer Center

Members

L. M. Bates, American Association of

Physicists in Medicine

R. S. Caswell, National Bureau of

Standards

J. A. Devanney, Bureau of Radiological

Health

G. R. Holeman, Yale University

This panel discussion was recorded and slightly edited, but retains the flavor and intent of the speakers and questions alike.

H. E. Johns, Ontario Cancer Institute

C. S. Nunan, Varian

W. R. Swanson, Stanford Linear Accelerator

Center

Panel Session:

Accelerators.

Neutrons from Electron Medical

J. Laughlin : At this time we would like to col- lect the views of those on this panel and those in the audience. I suggest that we rotate around and

take comments, first here from the panel members,

and then throw it open to the audience.

To start things off I have put on the board

a few of the points which I suggest for discussionI. Patient protection in radiation room

II. Operator protection outside radiation

room

Machine design

Room Shielding

Quality Factor

target

filter

col 1 i ma tor shielding

we'll direct our remarks either towards patient

protection in the room, or operator protection

outside the room. The role of quality factor is important for operator protection. With regard

to machine design, these (pointing) are various

topics which have already been covered which may deserve additional comment. I've also put on the board (Table 1) some values of dose • all in rads and which are approximate for a 20 MeV accelerator and approximately a 20 x 20 cm field. I've also indicated at the lower part of the board what I understand to be the lEC recommendations in the patient plane (2 meters in radius

normal to the beam axis 1 meter from the target),

and calling for leakage radiation as far as x-rays are concerned 1/10 of 1 percent (indicated

there as .001) and for neutrons two-hundreths of

one-percent, in rads.

I can't help but note that these values, measurement of neutron dose in the x-ray beam

and also outside, have not essentially changed

in the 20 MeV range over the last quarter of a century. The sophistication of measurement has TABLE I Patient Dose Due to Neutrons From Mega voltage X-Ray Treatment

(approx. 20 MV x-rays, 20 cm x 20 cm field size; all in rad)

in irradiated field

X-ray neutron heavy particle

out of target volume

4x1

0" 3x10-3

x-ray

leakage

-3

x-ray scatter

1x10- 5x10-2

4x10"^

2x10"^

( 5 cm)

(20 cm)

(70 cm)

neutron

leakage

2x10"^

Column

Col umn

Col umn

Col umn

Col umn

Col umn Column

neutron

internal

2x10"^ (20MV)

9x10"° (30MV)

2x10"^ (40MV)

3x10"^ (50MV)

dose of 1 rad at depth of maximum dose in irradiated field

dose in irradiated region due to absorption of neutronl contamination in x-ray beam

maximum dose in irradiated field due to photo nuclear reactions2 (assume all of photo neutrons locally absorbed)

permissible dose due to leakage x-rays dose at indicated distance from x-ray field due to internal x-ray scatter-^ average dose produced by neutrons penetrating accelerator shielding4

average dose produced by absorption of those photo neutrons internally produced^

lEC Recommendations

2 m radius plane at patient position

Percentage of maximum x-ray dose

werage

Maximum

X-Rays

0.1%

0.2%

Neutrons

.02% .05%

155
generally increased and for my part, I have found

this meeting very stimulating. With these com- ments I will start at one end of the table and proceed along the entire panel,

G. Holeman : I will make my comments very brief

so we can be sure and get around the table in the time allotted. In the measurements section that we held this morning, several problems were identified. The problems include: identifying the various sources that are generating the neutrons,

the photon interference of the various measurement techniques, calibration problems, calculation problems; and treatment room layout. All of these are problems that we are having to deal with in this area. It seems like we all agree

on some of the desired end results: information

about beam contamination; information about leakage; and, information about occupational exposure. One inconsistency seems to be in the way

data have been presented in the literature up

to this time, and the various units used; fluence,

absorbed dose, dose equivalent and various other

units. The problem of photon interference has not always been addressed completely in the past.

10*

oQ

oa> jDo

in <

I I

o c .c " Q. (M

10= -

if c 10^

10^

Electron Energy (MeV)

Figure 1. Comparison of Measured Neutron

Fluence per rad of Use Beam with the Theoretical Curve of McCall and Swanson (solid line). 0= Nath, Price

and Holeman, M=McGinley and Sohrabi

and A=Axton and Bardell

In the future it will be better addressed. In- formation about spectrum and beam sizes has not always been available in assessing information in the literature. I will not try to summarize

Peter Almond's excellent paper as it will be in the proceedings, but I will attempt to summarize

the percentage of neutron beam contamination that were given this morning by other authors. The

in-beam dose equivalent per photon rad involved

percentages all the way from .02 for the lower energy machines up to 0.8. In the patient plane

the neutron contamination varied from .01 for the lower energy machines up to about 0.4. These are

percent dose equivalent per photon rad. I will

show one slide (Figure 1). Bill Swanson will use

this Figure too, but I will get to use it first. Seven points are from McGinley's work (M) one from Axton's (A) and one from Math's (0) there is good agreement with the theoretical curve.

10^

)5 _

2 10

10^

Bl -

Measured in Beam

Measured out of Beam

Calculated

Betatron

All W and/or Pb

All Cu

I

\ I 20 30 40

ELECTRON ENERGY (MeV)

Figure 2: Neutron fluence per treatment dose

Upper curve in maximum amount for all- w treatment head; lower curve us for- all-c system. See McCall and Swanson paper for explanation of curves.

156

i
Swanson : One of the things that struck me by the conference is that while there is an awfully

lot of data on the physical aspect - that is the neutron production of fluences and doses, and

also a considerable amount of radiobiological

input, there's not been an estimate of the actual risk in an absolute sense to the patient. Since

I've never worked with radiobiology I am bold

enough to attempt to make an assessment of sorts.

This slide (Figure 2) shows that we can be- lieve that the fluence is a constant thing after

about 25 MeV electron energy and this upper curve is undoubtedly an overestimate. I think

Peter Almond's figure which is better than mine

shows that better. My curve is probably an overestimate by a factor of 2 for the average

accelerator.

This is Table 2 - that I showed just before

the break this morning - that shows integral dose

to the patient in gram-rads including 25 MeV treatment rad at 1 meter, divided up into portions

within and without the useful beam. They are ranked in order of importance in terms of integral dose to the patient in gram-rads including

25 MeV per treatment rad at 1 meter, divided up portions within and without the useful beam. They

are ranked in order of importance in terms of

integral absorbed dose in gram-rads. By far, the largest term is the photon scattering within the patient. This was proved by Rawlinson and Johns

in 1977. This is the most important thing: 500- 1400 gram-rads. Second to that is the 0.1% photon

leakage which contributes 40-50 gram-rads. Then we have 8 rad -grams of neutron leakage which I calculated myself. This result should be critically looked at by other people. It is surprisingly small, partly because of the rapid degradation of energy neutrons as they come through the tungsten or lead of the head shielding. Now let's see if I can figure what that means - if you

TABLE II

Integral Dose to Patient (g rad) at 25 MeV (per rad at 100 cm SSD)

With Useful Beam Outside Useful Beam Neutrons Photons Photon Neutron

Photons of Produced Scattered Leakage Leakage

Useful Beam Within Within (0.1 I)

Patient Patient

1800 (100 cm^) 5** 500 41 8.0

16200 (900 cm^)*50** 1390* 52*

* Rawlinson and Johns, 1977.

**Horsely. et al . (03%), 1953

TABLE III

AVERAGE LEAKAGE KEITTRCN DOSE

Averaaed cjver 70-kg patient

For 5000-rad treatirent course

-6 -1 -1

** Times 28 x 10 leukeanra.as yr rad

** Times 5 for all fatal iralignancies

-4 1.1 10 rad average dose per treatnient rad

0.58 rad average dose per treatjnent course

-6

16 10 leukemias/yr following treatment course

-6

80 10 all malignancies/yr following

** Rossi and Mays, 1977,

** 'Statural occurrence

-6

4 3-48 10 leukanias / yr

157
divide by the average size of the patient,

70,000 grams, that gives only about lO"'* rads (averaged over the full body volume) per treat- ment rad (See Table 3). If a typical treatment

of course is 5,000 rads, we only get about half

a rad of neutron dose to the treatment of course. Now refer to a paper by Rossi and Mays (1977) and assuming for the moment that the bone marrow is distributed uniformly within the body, (which it

is not, we just assume that) we multiply by the risk coefficient 28 x 10-6 leukemia's per year,

per neutron rad, and we get this leukemia induction about 16 X 10-6 per year, following the treat- ment course. Still referring to Table 3, if we

follow Rossi and Mays further and multiply by 5 we obtain about 80 x lO-^ fatal malignancies per year following the treatment course. This could

be compared to the natural occurrence for leuke- mia which is given in the same paper as being

in the range of 40-50 x 10"5 leukemia's per year.

You see the increase in leukemia is less than the natural occurrence and the increase of

all types of malignancies is somewhat bigger than

the natural occurrence of leukemia. Now, these

figures are obtained by averaging the dose over

the entire patient, and some people say that

that is not the right thing to do; one should per- haps look at the maximum dose and find out the risk to the patient from that. I can'c argue one way or the other about this point but I think

that it should be addressed by others who are ex- perts and by the agencies who will make up the rules concerning neutron leakage.

J. Laughlin : I have on the board the lEC values proposed for neutron leakage on the lower line.

I wonder if Higson or associate could check me on

the accuracy of those values, which are supposed

to be in rads.

^- Higson : Yes, they are correct. The smal- ler figure in each case should be the average and

the larger one the maximum.

R. Caswell: I have a few observations I'd like

to tnake. 1 think that this is an appropriate time

to review the regulations of the neutron production from medical electron linacs. I think that

these regulations and recommendations should re- flect the best concensus of scientific thinking

as of now, and therefore the historical values

which may have related actually to different situations and different types of machines should

logically be replaced. We should be very grateful

for the leadership of the I EC in working on this

field, and I think that within the United States probably an NCRP type of committee with representation from medical physics is a reasonable way to try to find these scientific opinions and come out with a set of recommendations. These recommendations should include presentations on how to make

this radiation as well as what might be reasonable

levels. One problem that I don't know the answer

to is: these committees usually work rather slowly, carefully and deliberately and if you want

these regulations done before the committee report

then it's very possible that regulation might be made before the committee has reported. Maybe

someone else has a suggestiong on that.

One other comment that might be a little contro- versial-because discussion on quantities and units

always seems to be controversial-it seems to me

that we should try to have a common unit for a quantity for recording and comparing our data even if people have measured in different quantities. The three main quantities that have been reported here are neutron fluence, absorbed dose,

and dose equivalent. It seems to me that fluence

is probably not too good for the final answer, be- cause effects are dependent on neutron energy. Either the dose equivalent or the dose absorbed

per neutron per square centimeter could be off

by a factor of 100. It would seem to me that the quantity should probably be either rads or rems. Of the two, it would be nice if it could be rads because that's really a physical quality without

the committee-selected "quality factor". On the other hand, I can see someone running around

with an Andersson-Braun rem-meter or Si evert

meter, or whatever and saying, "I don't know how

to get from the measurements I made in rems back

to rads". However, I think that ever worse will

be to have someone trying to assign a quality

factor at a later date. The other argument is that the quality factor itself is likely to change with time.

It seems to me that rads would be the best way to report the data. Then you can always re- construct. Therefore, the paper should report

dose equivalent or whatever was measured, and then try to convert the result into rads. One other remark that I would like to make is that

this question of whether the quality factor

needs to be increased by perhaps a factor of 10 seems to me quite crucial to considerations of

what is a reasonable neutron level. I think that

this problem should be resolved.

H. Johns: I'd like to reiterate what I tried

to say yesterday that I think it would be too bad if we create restrictive legistlation involving high energy machines especially if we deal only with neutrons. It seems to me that neutrons are only one of the problems. For ex- ample, if I understood the example of a moment

ago, you can calculate the number of extra leuk- emias becuase of stray neutrons, you could also cause leukemia from the extra radiation within

the radiation fields because you did not use a good beam. I suspect that the latter calculation

would yeild many more leukemias. We still have not found a good way to handle large fields of

radiation, maybe we need to line the inside of

thses collimators with a material which can re- move scattered electrons, as these are certainly

a problem. It seems to me that if you replace

the tungsten target with aluminum or copper we would get a 5-fold reduction of neutrons. Is that about right?

W. Swanson : If all of the internal shielding and

the target were copper the neutrons would go down by about a factor of 5 at 20 MeV and about

3 at 25 MeV.

H. Johns ; I think you have to remember we are dealing with very sick patients, and the leukemia calculations may be irrelevant. I was worried this morning that I would learn that

158 i
j the machines we have been using all these years

] were really neutron machines. They are useful,

they do cure cancer, but they can be made better

and I think that our efforts should be along this

line - to make them better and not worry about

neutrons. By better I mean we should seek beams

with very small entrance dose and beams for which

the maximum dose is as far below the surface of

the skin as possible. Having achieved these ends

! we might then feel there was some advantage in

' looking at neutrons.

! C. Nunan : As far as the manufacturers are con- cerned they are selecting materials, adding per- missible neutron shielding in the radiation head

of their machines making more thorough measurements in an effort to achieve the maximum operating x-ray energy that would be within the sug- gested state regulation of Q.1% combined x-ray

I plus neutrons in rems per primary x-ray rad. The

manufacturers are presently limited to the region

' of something like 15 to 19 MeV x-ray energy by

I that regulation. The radiotherapists say that

patients fare better with high energy x-rays - they want 20 to 25 MeV x-rays. If these suggested state regulations are implemented by the state

then we won't be able to furnish the radiotherapist with the machine that they want unless we go

to very cumbersome radiation heads which the

! radiotherapist wouldn't buy, so the manufacturers

' hope that an NCRP committee will resolve the

I difficulty because we feel that the states are unable to resolve the difficulty. As far as the risk is concerned, the only risk that I've heard

the radiotherapists mention is potential carcenogenesis, nothing else, and if you do the kind of

calculations where the whole body, or at a point

10 centimeters from the field edge in bone marrow,

and then calculate the compensating reduction in cure rate you find some very small fraction of a percent drop in cure rate would be equivalent,

so we're talking about very, very small risks

relative to the potential gains in cure rate by

designing good machinery. The thing is complicated by this possible change in quality factor,

and I understand the NCRP is studying this, the NCRP has convened a committee, I understand to recommend measurements techniques - it isn't

clear whather this committee is to also recommend

j an allowable neutron leakage - I think it should

' be given that mission - I don't know how to get

Li that accomplished but somebody ought to do it.

L. Bates : This is the fun thing of sitting at

! this end of the table is that by the time they

I get here everybody has said anything you've

! thought that you could possibly say - except for

, the panelist on my right.

I Dr. Swanson has been a radiobiologist 10

i seconds and already one wants to argue with him.

j He talked in terms of the carcinogenisis compared

ij to the natural occurence and I don't really think

i! this can be done for radiotherapy patients. I

i think that you have to look at this as an inI crease in late complications from the treatment

il and I think the proper value below the line is

, the number of complications arising from radio-

|l therapy. The other thing: there was a question

l{ earlier on today about integral dose versus maxi- mum dose. If my logic is correct when you're

talking about carcinogenisis you're talking about

a risk or a certain probability of a cell becoming malignant. In this case, the dose you should

be considering is the average or the integral

(which is always related with the constant) to the particular organ in question and, of course,

the radiosensiti vity of the particular organs.

J. Devanney : Before I start, I'd like to thank Tom Heaton and Bob Jacobs for putting together

a very good meeting - I think it is something

that has been needed for a long time. Several

people have asked what we're going to do with

the information that has come out of this meeting. Some people think that we will be able to generate some numbers and come up with neutron

leakage levels and the like. I do not feel that

way. I feel like the meeting has served as a very good data base on which BRH and other

regulatory agencies can use as a jumping off

point with which to develop a risk/benefit type

study of leakage levels. It has been pointed

out that there is no unified method or accepted

method of measuring the neutron contamination or even the depth-dose for high energy therapy

beams. I feel that we need more input from the Bureau of Standards in developing absorbed dose

standards for the high energy beams. In December I had the pleasure of attending and working with one of the working groups of the lEC and I was very delighted because I hadn't

realized how much the manufacturers had the con- cern of the patients at heart and how they

didn't really wear the manufacturer's hat at these meetings.

There is one thing I would like to say con- cerning the suggested state regulations - they're

passing out F-9 - if you'll look at F-9 you'll

find some brackets in the first few paragraphs.

You will have to read the preface of the sug- gested state regulations to understand what

these brackets mean. These bracketted para- graphs address the area of leakage radiation

and the brackets are there to show that this is a very controversial area, and it is not clear

how far the stated values can be accepted, so I suggest that you read the preface of the sug- gested state regulations to see how to interpret

the bracketted paragraphs.

H Wyckoff , I have just a couple or

three statements rather than questions. It always surprises me that when we talk about

deaths from automobile accidents, deaths from

airplane travel we don't say that the automobile

accident death rate is so many time the rate from aircraft accidents. I think that we're

getting into that kind of a thing here when we

compare the presumed leukemia death rate from a given irradiation with that from natural causes.

I think you forget that actual chances of death

in any one year are about 10"^ from all causes. Second point, I think that the quality factor you should worry a little bit about because you must remember that the quality factors that we

now have were developed back about 1950 at the time when you could almost count on your fingers

the amount of data that you had, and primarily

the factor of ten for neutrons were taken from some experimental data that was available then

159
for cataractogenesis. Now I think we have some evidence that for breast cancer the quality factor

may be closer to one, so I think you should think

in terms of 10 as being about right for some

things, but let's not worry about changing it immediately. The third point I think is that the factor of 5 that Rossi gave, and I'm sorry that

he's not here to hear me say this - if you take

the ICRP risk coefficients, that for leukemia is about a sixth of the total rather than a fifth

of the total. Thank you.

A. Jacobson : One number that did surface out of this meeting is the figure 1.6 rads for the neutron dose at the central axis of patients re- ceiving 5,000 rads tumor dose. I want to be sure there is reasonable agreement amongst the

panel on that figure. It seems very low and I'm happy to hear it.

H. Johns : I'm not sure, but I think they're

not at the same place in space. The 5,000 was

for the center of the body, the other figure I believe was 2 cm below the skin - Who gave that figure?

P. McGinley : I gave that figure - and that's

right.

N. Suntharal ingam : Craig, you made a statement

that you think that the manufacturers will not

be able to meet the current leakage specifications - is that right?

C. Nunan : The state proposed regulations at energies above 20 MeV.

N. Suntharal ingam : Is the requirement for leak- age 0.1% of the absorbed dose?

C. Nunan : Yes, 0.1% combined x and neutron

on a rem basis, 0.1% of the primary x-ray rad.

N. Suntharal ingam : Is this now on a rem basis

putting in the quality factor?

C. Nunan : Yes

N. Suntharal ingam : From tne data which has been

presented here it would not be difficult to sat- isfy the 0.1% of absorbed dose.

C. Nunan : No, no you have to read the sug- gested state regulations, they refer to a source

for the quality factor.

N. Suntharal ...vjam : The old NCRP requirement was

0.1% of absorbed dose. The numbers that I have

nere, that is, 0.02% for x-rays and about 0.03%

for neutrons at 20 MeV. The data that has been

presented by many of the speakers here, including

the last speaker, was that the measurements indi- cate 0.02% for x-rays and about 0.08% for neutrons at 19 MeV and other data indicates that

beyond 20 MeV it is reasonable flat, so that one could probably say that all machines in current

use and in manufacture at 25 MeV energy will meet the current NCRP leakage requirements or recommendations which is 0.1% of the absorbed

dose in the primary beam.

C. Nunan ; The NCRP is rad or rem? I'm confused?

H. Wyckoff ; jhe unit of absorbed dose is rad.

J. Laughlin ; To clarify a point that's been made here I did place on the board what I under- stood to be the lEC recommendations which included

this0LO2%, Craig, in rads per neutrons going up to a maximum of 0.05%. As has been observed over

20 MeV, the indicated levels level off - now do you find any difficulty with the proposed lEC recommendations?

C. Nunan : We believe that we will be able to meet the lEC proposed standard.

M. Cohen : When I came to this meeting two

days ago I was under the impression that the main

reason why we were interested in neutrons from

medical accelerators is in order to protect the operators and the general public. As far as the patient is concerned, the philosophy up to now

has been to design the machine so as to optimize

the treatment and not to worry too much about the neutron dose to the patient which was considered

of secondary importance and has not proved in the past to be of overwhelming significance. I've been a little surprised that the emphasis in this

conference, unless I've midjudged it, has been

on the whole body dose to the patient and possible consequences of that dose, with only perhaps a secondary consideration of the doses to the public

and theoperator. I'd like to know where this new emphasis has come from - has it derived from the regulatory agencies, or if not, where has it come

from?

J. Laughlin : I think several will respond to that - I think that is a very perceptive comment - we've gone along quite satisfactorily for a couple

of decades and with the interest recently in the quality factor which would appear to pertain more

to the operators the questions is why have the regulatory agencies suddenly become much more interested inside the radiation room. I don't

know whether John Devanney wants to respond to that,

J. Devanney ; We're interested mainly in reducing

the dose to any one person, whether it be a patient or a operator, or the staff member, and if design of the medical accelerator could reduce the "eakage radiation to the patient then I think that

it should be done. I think that the regulatory

agencies have always been concerned about the expo- sure to anybody.

P. Almond : I'd like to reply about Dr. Cohen's

inquiries as to where the emphasis has come from. The people who are installing medical linear ac- celerators in this energy range are aware of what

the regulations say, and that you have to satisfy

them. I know that is where we are. You have to make these leakage measurement, but the regulations

themselves are not at all clear as to whether it's

total x-ray plus neutron or just x-ray, or whether

it's rad dose or whether it's rem dose and there 'e a lot of confusion in some of the regulations as

to how the measurements are done, whether it is done with the jaws closed or open, in the plane of

160
I the patient or elsewhere. We are having to satisfy the state health departments. That is where

' our emphasis is. It is not just the dose to the operator, it is also leakage around the machine

in the treatment room.

A. Chung-Bin : I would like to make a comment

as to whether those people who work in the regulatory agencies realize that, if you look at the

j trend in radiation therapy there is a tendency

I to make the field size larger and larger, i.e.,

j using mantle field for radical treatment of

I Hodgkin's Disease and large field hemi-body treat-

' ment of leukemia and other metastasis. Many

I radiotherapists have stated one would prefer to

I give patients total body radiation, if possible,

because the disease that will eventually kill the patients is probably the microscopic cancer cell diffusely distributed in the body. Therefore,

j we really should not be overly concerned about a

j neutron dose delivered to the patients.

, D. Rogers : I would like to ask the panel to

' take a stand on the major issues facing us and

. tell the regulatory agencies our views on what

the leakage limit should be while we've got some

of them here. The statement was made that we're

going to have to do cost/benefit analysis, and

I think we've'wasted a good deal of this conI ference by not addressing the real question which

I is what are the elements which must be conj sidered in a proper cost/benefit or risk/benefit

analysis. I would like the people in the audience to try to enumerate what are the costs

and the benefits which we want the regulatory

agencies to take into account when they to the calculations. In my opinion we are currently

being forced to meet a set of essentially unjusttifiable regulations, and I think it's time we

should try to point out why we think these are

dumb-. The only talks which really dealt with

this area were the first few lectures yesterday

which were very suggestive but didn't really

address the specific questions of justifying a proper limit. I would like to have people start enumerating the risks and the benefits that we

want in the calculations.

J. Lauqhlin : Well, does anyone wish to respond

to that particular challenge at this time? I'll have to say that we have only a couple more

1 minutes.

I W. Swanson : Yes, one reason for my standing up

and giving this estimate of risk to the patient

was to kind of start this thing in motion, and

one reason for my doing this was to emphasize the smallness of the risk of these leakage neutronsthe smallness . In fact from this you could probably formulate some cost/benefit ratio in terms

of cancers cured per cancers induced. Now I know

tnly figures could be improved by experts who should

review them, but what I have in my notes here is that the benefit/cost ratio of cancers cures to those induced, within say ten years, is about

1,000 to 1. That's one suggestion of a figure

that should be looked at and reviewed and improved

by others, then it could be an ingredient of an

overall cost/benefit analysis for the regulatory

agencies to use. Furthermore, I think that that

is such a good cost benefit ratio, and the overall

risk is so small that the whole matter could be left between the doctors, their patients and the manufacturers, and the government need not say anything further about it - that's my personal

feeling about it.

J. Laughlin : Are there any other comments or questions?

We want to thank Tom Heaton.and the National Bureau of Standards, Bob Jacobs and those of

BRH very much for a very stimulating and important conference.

161
IG\IfI[If
CONFERENCE ON NEUTRONS FROM ELECTRON MEDICAL ACCELERATORS

National Bureau of Standards

Gaithersburg, Maryland

AprH 9-10, 1979

List of Attendees

Robert B . Adams

Univ. Hospital of Cleveland

2065 Adelbert Road

Cleveland, OH 44106

Suresh K. Agarwal

Univ. of Virginia

Department of Radiology

1604 Cedar Hill Road

Charlottesville, VA 22901

Devendra K. Agrawal

Loyola University Medical Center

2160 S. First Avenue

Maywood, IL 60153

David P. Alberth

School of Hygiene & Public Health

Johns Hopkins Univ.

615 North Wolfe Street

Baltimore, MD 20215

Peter R. Almond

M.D. Anderson Hospital & Tumor Inst. Texas Medical Center

6723 Bertner Ave.

Houston, TX 77030

Melvyn Altman

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Lee S. Anthony

Physics Associates

1229 Forest Lawn Drive

Salem, VA 24153

L. Armstrong

Atomic Energy of Canada Ltd.

P.O. Box 6300

Ottawa, Ontario

CANADA K2A 3W3

L. Atherton

M.E.L.

Manor Royal

Crawley, Sussex

ENGLAND RHIO 9PZ

John A . Auxier

Oak Ridge National Laboratory

Post Office Box X

Oak Ridge, TN 37830

E. J. Axton

National Physical Laboratory

Queens Road

Teddington, Middlesex

ENGLAND

Farideh Bagne

Duke Univ. Medical Center

P.O. Box 3295, Edwin Morris Bldg.

Durham, NC 27710

Morris Bank

Hurley Medical Center, Dept. of

Radiology

6th & Begole

Fhnt, Ml 48502

Lloyd M. Bates

American Assoc. of Physics in Medicine

Suite 307, 6900 Wisconsin Avenue

Chevy Chase, MD 20015

L. B. Beentjes

Kathoheke Universiteit Nymegen

Toernooiveld

6525 ED Nymegen

NETHERLANDS

F . Behmann

McGill Univ. Radiotherapy Centre

1650 Cedar Ave.

Montreal, Quebec

CANADA H3G IA4

James W. Behrens

National Bureau of Standards

Washington, DC 20234

J. P. Bhatnagar

Mercy Hospital

1400 Locust St. Pittsburgh, PA 15219

163
Peter J. Biggs

Massachusetts General Hosp.

Fruit Street

Boston, MA 02114

Joseph S. Blinick

Maine Med. Center

22 BramhaU St. Portland, ME 04102

Frank J. Bova

J. HiUis Miller Health Ctr.

Univ. of Florida

Box J-385

Gainesville, FL 32610

Allen Brodsky

U.S. Nuclear Regulatory Commission

Washington, DC 20555

David Brown

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Joseph M. Brown, Jr. U.S. Nuclear Regulatory Commission

5903 Greentree Road

Bethesda, MD 20034

Roger Budd

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Jan J. Buitendag

South African Dept. of Health

658 Sak Street

Pretoria, SOUTH AFRICA 0181

Dale W. Campbell

Richland Memorial Hosp.

Radiation Therapy

3301 Harden St. Columbia, SC 29203

Allan David Carlson

National Bureau of Standards

Washington, DC 20234

Wayne Cassatt

National Bureau of Standards

Center for Radiation Research

Washington, DC 20234

Victoria G. Castro

Abington Memorial Hospital

1200 Old Fork Road

Abington, PA 19001

RandaU S. CasweU

National Bureau of Standards

Washington, DC 20234

Frank Cerra

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Myung C. Choi

Washington Univ. Medical School

10225 Canter Way

Overland, MO 63U4

Jim Chu

Hosp. of the Univ. of Pennsylvania

Dept. of Radiation Therapy

3400 Spruce Street, Box 522

PhUadelphia, PA 19104

WiUiam T. Chu

Loma Linda Univ. Medical Ctr.

Radiation Oncology

Loma Linda, CA 92350

Anthony Chung-Bin

Rush Medical CoUege

1753 W. Congress Street

Chicago, IL 60612

Lionel Cohen

Michael Reese Medical Ctr.

29th St. & EUis Avenue

Chicago, IL 60616

Montague Cohen

Montreal General Hospital

McGiU Univ. Radiotheraphy Centre

1650 Cedar Ave.

Montreal, Quebec

CANADA H3G 1A4

164
Alan Cohler

j Abington Memorial Hospital

I Laurel Road

I Solebury, PA 18963

J. Joseph Coyne

National Bureau of Standards

Washington, DC 20234

Wilber Cyr

j DHEW/Bureau of Radiological Health

j 5600 Fishers Lane

i RockviUe, MD 20857

I Arne Dahler

Haukeland Hospital

5016 Hankeland Sykehus

Bergen, NORWAY

j John Day

' Hahnemann Hospital

! 230 N. Broad St. PhHadelphia, PA 19102

Wniiam H. Dederichs

The Radiation Center

1425 E. Enderly Place

Fort Worth, TX 76104

Paul M. DeLuca, Jr. Univ. of Wisconsin

3321 Sterling HaU

475 North Charter Street

Madison, WI 53706

John DePangher

Stanford University

I 67 Encina HaU, Health Physics

Stanford, CA 94305

Nicholas A. Detorie

SIU - School of Medicine

Memorial Medical Center

800 North Rutledge St. Springfield, IL 62702

John A. Devanney

Bureau of Radiological Health

12720 Twinbrook Pkwy

RockviUe, MD 20857

Robert L. Dixon

Bowman Gray School of Medicine

Dept. of Radiology

Winston- Salem, NC 27103

Jack Dobson

WSA Inc. 5053 San Joaquin Drive

San Diego, CA 92109

Walter E. Downs

Atomic Energy oc Can. Ltd.

Commericial Products

P.O. Box 6300

Postal Stat. J. Ottawa, Ontario

CANADA K2A 3W3

Kenneth C. DuvaU

National Bureau of Standards

1436 18th Street SE

Washington, DC 20234

Pavel Dvorak

Radiation Protection Bureau

Brookfield Road

Ottawa, CANADA KIA ICl

Paul D. Eastvold

lUinois Dept. of PiibUc Health

535 West Jefferson Street

Springfield, IL 62702

WiUiam Eden

Florida State Dept. of Health

2812 Duffton Loop

TaUahassee, FL 32303

Bradley D. Eichorst

BRH/FDA/HEW

5600 Fishers Lane

RockvUle, MD 20857

Charles Eisenhauer

National Bureau of Standards

Washington, DC 20234

Elmer H. Eisenhower

National Bureau of Standards

Bldg. 245, Room C229

Washington, DC 20234

165
Melvin J. Ely, Jr. National Naval Medical Center

Radiology Service

Bethesda, MD 20014

Harold Eriksen

Liberty Mutual Res. Ctr.

71 Frankland Road

Hopkinton, MA 01748

David Espejo-Villalobos

Hospitol Universitario

Calzada Madreno Pte. Monterrey, N.L.

MEXICO 47-10-10, Ext 120

Fred Faw

Armot-Ogden Memorial Hospital

Radiology Dept.

Elmira, NY 14801

Thomas Fearon

National Council on Radiation

Protection and Measurements

7910 Woodmont Ave . , Suite 1016 Bethesda, MD 20014

Kenneth P. Ferlic

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Lawrence T. Fitzgerald

Univ. of Florida

Div. of Radiation Physics

Box J-385, JHM Health Center

Gainesville, FL 32610

Claude Flory

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

John G. Fox

Carnegie-Mellon Univ. and St. Francis

General Hospital

Schenley Park, 45th & Penn

Pittsburgh, PA 15213

John Frazier

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Elroy J. Friesen

University of Virginia Hosp.

Box 375

CharlottesviUe , VA 22923

Stephanie V. Frost

Methodist Hosp. of Indianapolis

1604 N. Capitol Ave.

Indianapolis, IN 46202

E. G. FuUer

Photonuclear Data Center

National Bureau of Standards

Washington, DC 20234

L. David Gager

Baylor College of Medicine

The Methodist Hospital

1200 Moursund

Houston, TX 77030

Benjamin M. Galkin

Thomas Jefferson Univ.

920 ChanceUor St. PhHadelphia, PA 19107 Dwight W. Glenn

Mid- Atlantic Radiation Physics,

10408 Truxton Road

Adelphi, MD 20783

John C. Goble

NIH

Bldg. 21, Rm. U6

Bethesda, MD 20014

Donald A. Goer

Varian Associates

611 Hansen Way

Palo Alto, CA 94303

Robert Golden

Walter Reed Army Medical Ctr.

Radiation Therapy Service

Washington, DC 20012

Richard C. Granke

Private Practice of Medical

Radiological Physics

13U Downs Drive

Silver Spring, MD 20904

166
Lance Green

Dosimeter Corporation of America

j 11286 Grooms Road

' Cincinnati, OH 45242

i Michael P . Grissom

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MB 20014

' James A . Grolten

j Georgetown Univ . Medical Ctr

.

Kober Cogan Bldg., Rm. 305

I 3750 Reservoir Road, NW

Washington, DC

Zygmunt C. Gromadzki

Georgetown Univ. Hospital

Radiation Oncology Dept.

3800 Reservoir Road, NW

I Washington, DC 20007

I Walter Gundaker

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

David Gur

Graduate School of Pubhc Health

Univ. of Pittsburgh

Pittsburgh, PA 15261

Richard Haas

Univ. of Illinois

840 S. Wood Street

Chicago, IL 60612

j David Hammond

; Baylor Univ. Medical Center

! Charles A. Sammons Cancer Ctr.

3500 Gaston Ave.

DaUas, TX 75246

Antun Han

Argonne National Laboratory

9700 S. Cass Ave.

Argonne, IL 60439

I Lawrence F. Hans, Jr.

I Mercy Hospital

2500 7th Avenue

I Altoona, PA 16603

George H. Harrison

University of MD Medical Sch.

Radiation Research Div.

660 W. Redwood St. Baltimore, MD 21201

H. T. Heaton

National Bureau of Standards

Center for Radiation Research

Washington, DC 20234

William N. Herrington

M.D. Anderson Hosp. and Tumor Inst. 6723 Bertner Ave.

Houston, TX 77025

G. R. Higson

Dept. of Health & Social Security

14 Russell Square

London, ENGLAND WCIB SEP

George R. Holeman

Yale University - Health Physics

314 Wright Nuclear Structure Lab.

West - 260 Whitney Ave.

New Haven, CT 06520

Carolyn Holman

Orlando Regional Medical Center

Radiation Division

50 West Sturtevant

Orlando, FL 32806

John R. Howley

NIH

Bldg. 21, Rm. U6

Bethesda, MD 20014

Jen-Shu Hsieh

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Geoffrey S. Ibbott

Univ. of Colo. Med. Center

4200 E. Ninth Ave. (A-031)

Denver, CO 80262

Barrie J. Jackson

Atomic Energy of Canada Ltd.

P.O. Box 6300, Postal Station "J" Ottawa, Ontario

CANADA K2A 3W3

167
Robert Jacobs

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockvme, MD 20857

Ahren Jacobson

Univ. of Louisville

Radiation Center

500 S. Floyd St. Louisville, KY 40201

John P. Jacobus

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Stan Jaeger

Bishop Clarkson Memorial Hosp.

Dewey Ave. at Forty-Fourth St. Omaha, Nebraska 68105

Michael F. James

Atomic Energy Control Board

270 Albert St. Ottawa, Ontario

CANADA

Mitchell J. Jarosz, Jr.

St. Francis General Hospital

45th St. and Penn Ave.

Pittsburgh, PA 15201

Roger M. Jean

CGR MeV

BP 34

78530 BUC, FRANCE 356 8002

Harold E . Johns

Ontario Cancer Institute

500 Sherbourne Street

Toronto, Ontario

CANADA M4X 1K9

Joel D. Johnson

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Troyce D. Jones

Oak Ridge National Laboratory

P.O. Box X

Oak Ridge, TN 37830

D. M. Joseph

Lutheran General Hospital

1775 W. Dempster Ave.

Park Ridge, IN 60068

Jae Shik Jun

Korea Standards Research Inst. P. O. Box 333

Dae Jeon, KOREA

Christos Kanellitsas

Universsity Hospital

1405 E. Ann Street

Ann Arbor, MI 48109

C. J. Karzmark

Stanford Univ. School of Med.

Dept. of Radiology

300 Pasteur Dr.

Stanford, CA 94305

Frank E. Kearly

Bur. of Radiological Health (HFX-70)

5600 Fishers Lane

Rockville, MD 20857

David R. Kneeland

Nucletronix

Birch Road

Middleton, MA 01949

George Komada

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockeville, MD 20857

Max R. Kubik

Baptist Medical Center

800 Prudential Drive

Jacksonville, FL 32207

Chris K. Kuyatt

National Bureau of Standards

Center for Radiation Research

Bldg. 220, Rm. B206

Washington, DC 20234

Wing-8chee Lam

Johns Hopkins University

601 N. Broadway

Baltimore, MD 21205

168
i Philip D. LaRiviere

j Varian Associates, Inc.

I 611 Hansen Way

Palo Alto, CA 94303

John S. Laughlin

Memorial Sloan-Kettering Cancer Ctr.

1275 York Ave.

New York, NY 10021

Dennis D. Leavitt

Univ. of Utah Medical Center

Dept. of Radiology

50 N. Medical Drive

Salt Lake City, UT 84132

Paul Leggett, Jr. FDA/BRH

9917 Shrewsbury Ct.

Gaithersburg, MD 20760

David Lightfoot

Hahnemann Hospital

230 N. Broad St. PhHadelphia, PA 19102

Shin S. Liu

Rush-Presbyterial-St. Luke Med. Ctr.

1753 W. Congress Parkway

Chicago, IL 60612

Salmen Loksen

Georgetown University Hospital

3800 Reservoir Rd., NW

Washington, DC 20007

Robert Loevinger

National Bureau of Standards

Radiation Phys. C210

Washington, DC 20234

Richard C. McCaU

Stanford University

Stanford Linear Accelerator Center

P.O. Box 4349

Stanford, CA 94305

Joseph McDonald

Memorial- Sloan Kettering Cancer Ctr.

1275 York Avenue

New York, NY 10021

Patton H. McGinley

Emory University Clinic

1365 Clifton Road

Atlanta, GA 30322

Philip McHugh

Atomic Energy of Canada Ltd.

P.O. Box 6300

Ottawa, Canada K2A 3W3

Lou Milavickas

Univ. of Illinois 214 Nuclear Engineering Lab

Urbana, IL 61801

David W. Minnaar

Michigan Dept. of Public Health

Div. of Radiological Health

3500 N. Lorgan St. P.O. Box 30035

Lansing, MI 48909

Robert Morton

Bureau of Radiological Health (HEX- 70)

5600 Fishers Lane

RockviUe, MD 208.57

Ravinder Nath

Yale Medical School

333 Cedar St. New Haven, CT 06510

J. Rao Nibhanupudy

Howard Univ. Hospital

2041 Goergia Ave . , NW

Washington, DC 20060

Brian Noriega

Radiation Services

1404 N. Wheeler St. Plant City, FL 33566

Craig Nunan

Varian

611 Hansen Way

Palo Alto, CA 94303

Michael P. Nunno

Albert Einstein Medical Center

Radiology Associates

York & Tabor Roads

Philadelphia, PA 19144

169
Thomas Ohlhaber

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockville, MD 20857

H. M. Omar

Univ. of Tennessee

Radiation Oncology Dept.

865 Jefferson Ave.

Memphis, TN 38163

F. E. Owens

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Peter Paras

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Mahendra Patel

Capintec, Inc. 540 Alpha Drive

Pittsburgh, PA 15238

John I. Pearce

Roswell Park Memorial Inst. 666 Elm St. Buffalo, NY 14263

Harry Pettengill

U.S.N.R.C.

Nicholson Lane

RockviUe, MD 20852

Jean-Louis Pourre

CGR MeV

420 Babylon Road

Horhsam, PA 19044

William E. Powers

Wayne State Univ. Medical School

3990 John R. St. Detroit, MI 48201

Frederick Prevo

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockville, MD 20857

Kenneth Price

Yale University

Health Physics Div.

Whitney Ave.

New Haven, CT

Vincent L. Rauaschieri

Seimens Corporation

449 Old Orchard Circle

MiUersville, MD 2U08

Alan Rawlinson

Ontario Cancer Institute

500 Sherbourne Street

Toronto, Ontario

CANADA M4X 1K9

Henry Rechen

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockville, MD 20857

Don W. Reid

EMI Therapy Systems

570 Del Rey

Sunnyvale, CA 94086

Robert E. Rice

Hartford Hospital

80 Seymour St. Hartford, CT 06U5

Gordon Riel

NSWC

White Oak, MD 20910

Karim Rimawi

NY State Dept. of Health

15 SneUen Road, Apt. A

Schenectady, NY 12309

David W. O. Rogers

National Res. Council of Canada

Montreal Road., Bldg. M-35

Ottawa, Ontario

CANADA KIA 0R6

Joseph A. Rose

Wilmington Medical Ctr.

Dept. of Radiation Therapy

201 S. Broom St., P.O. Box 2850

Wilmington, DE 19805

170
Carl K. Ross

National Res. Council of Canada

Div. of Physics, Bldg. M-35

Ottawa, Ontario

CANADA KIA 0R6

Harald H. Rossi

Radiological Res. Lab.

Columbia University

630 West 168th Street

New York, NY 10032

Donald R. Ruegsegger

Miami Valley Hospital

Radiation Therapy Dept.

1 Wyoming Street

Dayton, OH 45409

Joseph A. Sayeq

University of Kentucky

Dept. of Radiation Medicine

800 Rose Street

Lexington, KY 40506

Roger Schneider

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Michael C. Schell

GWUMC

Middle Atlantic Neutron Therapy Assoc.

Washington, DC 20375

j Roald A. Schrack

I National Bureau of Standards

i Washington, DC 20234

Robert J. Schulz

Yale University School of Medicine

333 Cedar Street

New Haven, CT 06510

Robert Schwartz

National Bureau of Standards

I Washington, DC 20234

I Ralph Severance

Defense Nuclear Agency

I Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

S. Shalev

Dept. of Biomedical Engrs.

Technion

Haifa, ISRAEL

Jeffrey Silberberg

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Theodore N. Simmons

Sandia Laboratories

Health Physics Div . , 3312

P.O. Box 5800

Albuquerque, NM 87185

Bill Simon

Victoreen Inc.

10101 Woodland Ave.

Cleveland, OH 44104

Lester A. Slaback, Jr. Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

C. Smith

Oklahoma St. Dept. of Health

Oklahoma City, OK

George Snyder

Student Johns Hopkins Univ.

8900 Centerway Road

Gaithersburg, MD 20760

Robert Stanton

Cooper Medical Center

One Cooper Plaza

Camden, NJ 08103

Edward S. Sternick

Tufts-New England Med. Ctr.

Dept. Therapeutic Radiology

171 Harrison Ave.

Boston, MA 02111

N. Suntharalingam

Thomas Jefferson Univ. Hospital

11th & Walnut Sts.

Philadelphia, PA 19107

171
Roy E. Summers, Jr. Allepheny General Hospital

320 E. North Ave.

Pittsburgh, PA 15212 Goran K. Svensson

Harvard Medical School

Joint Ctr. for Radiation Therapy

50 Binney Street

Boston, MA 02115

Robert W. Swain

NIH

6010 Walton Road

Bethesda, MD 20034

WOliam P. Swanson

Stanford Linear Accelerator Ctr.

P.O. Box 4349

Stanford, CA 94305

Leonhard Taumann

Siemens Medical Laboratories, Inc. 2404 N. Main Street

Walnut Creek, CA 94596

Paul Thiess

Catholic University

P. O. Box 951, Cardinal Station

Washington, DC 20064

Jerry A. Thomas

National Naval Medical Center

13122 Pelfrey Lane

Fairfax, VA 22030

Othniel Thompson

MD. State Dept. of Health EHA

201 W. Preston Street

Baltimore, MD 21050

Winston L. Thompson

Howard University

CoUege of ADied Health

Washington, DC 20059

Joseph Y. Ting

Our Lady of Lourdes Hospital

169 Riverside Dr.

Binghamton, NY 13905

Eugene Tochilin

Varian Associates

611 Hansen Way

Palo Alto, CA 94303

Neil VanHooydonk

PhOips Medical Systems, Inc.

710 Bridgeport Avenue

Shelton, CT 06484

David L. Vassy, Jr. Mid- Atlantic Radiation Physics, Inc. 10408 Truxton Road

Adelphi, MD 20783

Donald E. Velkey

Div. of Radiation Oncology

M.S. Hershey Medical Ctr. of the

Pennsylvania State Univ.

Hershey, PA 17033

Lorne Vernon

Defense Nuclear Agency

Armed Forces Radiobiology Res. Inst. Bethesda, MD 20014

Paul Wagner

Bur. of Radiological Health (HFX-70)

5600 Fishers Lane

RockviUe, MD 20857

Ray Walehle

DHEW/Bureau of Radiological Health

5600 Fishers Lane

RockviUe, MD 20857

Joseph Wang

BRH

5600 Fishers Lane

RockviUe, MD 20857

A. Carl Warner

Dept. of Radiation Oncology

Indiana Univ. Sch. of Medicine

UOO West Michigan Street

IndianapoUs, IN 46223

Oren Wasson

National Bureau of Standards

Washington, DC 20234

172
j William R. Webber

I Defense Nuclear Agency

I Armed Forces Radiobiology Res. Inst. 'Sethesda, MD 20014

Robert V. Wheeler

R.S. Landauer, Jr. & Co.

Science Road

Glenwood, IL 60425

I Ralph Worsnop

Medical Physics

1250 Godetia Drive

Woodside, CA 940062

Ann E. Wright

Univ. of Texas Medical Branch

18635 Point Lookout

Houston, TX 77058

John Wright

Geisinger Medical Center

DanviUe, PA 17821

Kenneth A. Wright

Mass. Inst, of Tech.

Bldg. N-10, M.I.T.

Cambridge, MA 02139

Raymond K. Wu

Temple University Hospital

3401 N. Broad St. Philadephia, PA 19140

Harold Wyckoff

4108 Montpelier Road

RockviRe, MD 20853

Harry Youmans

DHEW/Bureau of Radiological Health

5600 Fishers Lane

Rockville, MD 20857

Gary H. Zeman

National Naval Medical Center

Bethesda, MD 20014

I

173
I

I

I

j

j

I

j

1

1

1

I

i
AUTHOR INDEX

P. R. Almond 129, 160 J. S. Laughlin 1, 155, 158, 160,

J. A, Auxier 41 R. C. McCall 75

E, J. Axton 109, 125 P. H. McGinley 99, 160

A. G. Bardell 109, 125 R. Nath 87

L. Bates 159 C. S. Nunan 159, 160

A. Chung-Bin 161 K. W. Price 87

L. Cohen 57 H. E . Powers 23

M, Cohen 160 J. A. Rawlinson 25

J. A. Devanney 159, 160 A. Reid 1 J, Ding 1 D. Rodgers 161

M. M, Elkind 63 H. H. Rossi 37

A. Han 63 C. H. Smith 33

G. R. Higson 27, 158 M. Sohrabi 99

G. R. Holeman 87, 156 N. Suntharalingam 160

D. G. Jacobs 41 W. P . Swanson 75, 151, 158, 161

A. Jacobson 160 L. Taumann 139

T. D. Jones 41 E. Tochilin 145

H. E. Johns 15, 158, 160 H. Wyckoff 159, 160

G. D. Kerr 41 L. Zeitz 1 P. D. LaRiviere 145

175
NBS-114A (REV. 9-781

U.S. DEPT. OF COMM. BIBLIOGRAPHIC DATA

SHEET

1. PUBLICATION OR REPORT NO. SP 554

2. Gov't Accession No,

4. TITLE AND SUBTITLE Proceedings of a Conference on Neutrons from Electron

Medical Accelerators

5. Publication Date September 1979

6. Performing Organization Code

7. AUTHOR(S)

H. T. Heaton, II and R. Jacobs

8. Performing Organ. Report No.

9. PERFORMING ORGANIZATION NAME AND ADDRESS

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE

WASHINGTON, DC 20234

10. Project/Tasl</WorH Unit No.

11. Contract/Grant No.

12.^SP0NS0RING ORGANIZATION NAME AND COMPLETF m^ RFSS (Street. City, si^te. ziP) Sponsored in part by: ,,. , Bureau of Radiological Health ^^^h cooperation from:

Food and Drug Administration American Association of

Physicists in Medicine

Bethesda, Maryland 20014

13. Type of Report & Period Covered N/A

Rockville, Maryland 20857

14. Sponsoring Agency Code

15. SUPPLEMENTARY NOTES Library of Congress Catalog Number 79-600-33

[2j Document describes a computer program; SF-185, FIPS Software Summary, is attached.

16. ABSTRACT (A 200-word or less factual summary oi most significant information,

literature survey, mention it here.)

If document includes a significant bibliography or

These proceedings are the compilation of 18 papers presented at the Conference

on Neutrons from Electron Medical Accelerators held in Gaithersburg , MD on April 9 and 10, 1979. The topics addressed include: properties of high energy

beams for radiotherapy, regulations, biological interpretation, physical

measurements and calculations, and neutron reduction techniques.

17. KEY WORDS fsjx to twelve entries; alphabetical order; capitalize only the first letter of the first key word unless a proper name;

separated by semicolons)

Dosimetry, high energy radiotherapy beams, leakage, measurement techniques,

medical accelerators, neutrons, relative biological effect, standards, shielding

-hniqi.

18. AVAILABILITY [Kl Unlimited

I I For Official Distribution. Do Not Release to NTIS

[X] Order From Sup. of Doc, U.S. Goverqrnent Printing 20402, SD Stock No. SN003-003- 02115-6

Office, Washington, DC

\_J Order From National Technical Information Service (NTIS), Springfield, VA. 22161

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U.S. GOVERNMENT PRINTING OFFICE : 1979 0—281-067 (245)
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related to the Bureau's technical and scientific programs. As

a special service to subscribers each issue contains complete

citations to all recent NBS publications in NBS and non- NBS media. Issued six times a year. Annual subscription: domestic $17.00; foreign $21.25. Single copy, $3.00 domestic;

S3. 75 foreign.

!Note: The Journal was formerly published in two sections: Section A "Physics and Chemistry" and Section B "Mathematical Sciences." DIMENSIONS/NBS

This monthly magazine is published to inform scientists, engineers, businessmen, industry, teachers, students, and

consumers of the latest advances in science and technology,

with primary emphasis on the work at NBS. The magazine

highlights and reviews such issues as energy research, fire protection, building technology, metric conversion, pollution

abatement, health and safety, and consumer product per- formance. In addition, it reports the results of Bureau pro- grams in measurement standards and techniques, properties

of matter and materials, engineering standards and services, mstrumentation, and automatic data processing.

Annual subscription: Domestic, $11.00; Foreign $13.75

NONPERIODICALS

Monographs—Major contributions to the technical liter- ature on various subjects related to the Bureau's scientific and technical activities. Haodbooks—Recommended codes of engineering and industrial practice (including safety codes) developed in cooperation with interested industries, professional organizations, and regulatory bodies.

Special Publications—Include proceedings of conferences sponsored by NBS, NBS annual reports, and other special

publications appropriate to this grouping such as wall charts, pocket cards, and bibliographies.

Applied Mathematics Series—Mathematical tables, manuals, and studies of special interest to physicists, engineers,

chemists, biologists, mathematicians, computer programmers,

and others engaged in scientific and technical work.

National Standard Reference Data Series—Provides quantitative data on the physical and chemical properties of

materials, compiled from the world's literature and critically evaluated. Developed under a world-wide program co- ordinated by NBS. Program under authority of National

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Standard Data Act (Public Law 90-396).

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The following current-awareness and literature-survey bibli- agraphies are issued periodically by (he Bureau:

Cryogenic Data Center Current Awareness Service. A litera- ture survey issued biweekly. Annual subscription: Domestic, $25.00; Foreign, $30.00.

Liquefied Natural Gas. A literature survey issued quarterly.

Annual subscnption: $20.00.

NOTE: At present the principal publication outlet for these data is the Journal of Physical and Chemical Reference Data (JPCUD) published quarterly for NBS by the American Chemical Society (ACS) and the American Institute of Physics (AIP). Subscriptions, reprints, and supplements

available from ACS, 1155 Sixteenth St. N.W., Wash., D.C.

20056.

Building Science Series—Disseminates technical information

developed at the Bureau on building materials, components,

systems, and whole structures. The series presents research

results, test methods, and performance criteria related to the

structural and environmental functions and the durability

and safety characteristics of building elements and systems. Technical Notes—Studies or reports which are complete in themselves but restrictive in their treatment of a subject.

Analogous to monographs but not so comprehensive in scope or definitive in treatment of the subject area. Often

serve as a vehicle for final reports of work performed at NBS under the sponsorship of other government agencies.

Voluntary Product Standards—Developed under procedures

published by the Department of Commerce in Part 10, Title 15, of the Code of Federal Regulations. The purpose

of the standards is to establish nationally recognized require- ments for products, and to provide all concerned interests with a basis for common understanding of the characteristics of the products. NBS administers this program as a supple- ment to the activities of the private sector standardizing

organizations.

Consumer Information Series—Practical information, based on NBS research and experience, covering areas of interest to the consumer. Easily understandable language and

illustrations provide useful background knowledge for shopping in today's technological marketplace.

Order above NBS publications from: Superintendent of Documents, Government Printing Office, Washington, D.C.

20402. Order following NBS publications—NBSlR's and FIPS from

the National Technical Information Services, Springfield, Va. 22161.

Federal Information Processing Standards Publications (FIPS PUB)—Publications in this series collectively consti- tute the Federal Information Processing Standards Register.

Register serves as the official source of information in the Federal Government regarding standards issued by NBS

pursuant to the Federal Property and Administrative Services Act of 1949 as amended. Public Law 89-306 (79 Stat

1127), and as implemented by Executive Order 11717

(38 FR 12315, dated May 11, 1973) and Part 6 of Tide 15 CFR (Code of Federal Regulations).

NBS Interagency Reports (NBSIR)—A special series of

interim or final reports on work performed by NBS for outside sponsors (both government and non-govemment).

In general, initial distribution is handled by the sponsor;

public distribution is by the National Technical Information

Services (Springfield, Va. 22161) in paper copy or microfiche form.

Superconducting Devices and Materials. A literature survey

issued quarterly. Annual subscription: $30.00. Send subscription orders and remittances for the preceding bibliographic

services to National Bureau of Standards, Cryogenic Data

Center (736.00) Boulder, Colorado 80303.

BIBLIOGRAPHIC SUBSCRIPTION SERVICES
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Washington. D C. 20234

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