V
DATE DUE

1

Demco. Inc. 38-293

Neutron Standards and Applications

'r- 'i of St'-ndnrc's

7'80306

Proceedings of the

International Specialists Symposium on

Neutron Standards and Applications

Held at the National Bureau of Standards

Gaithersburg, MD, March 28-31, 1977

Edited by

CD. Bowman and A.D. Carlson

Center for Radiation Research

Institute for Basic Standards

National Bureau of Standards

Washington, D.C. 20234

H.O. Liskien

Central Bureau of Nuclear Measurements

Commission of European Communities

B-2440 Geel

Steenweg Naar Retie

BELGIUM

and

L. Stewart

Los Alamos Scientific Laboratory

Los Alamos, NM 87545

Sponsored by

National Bureau of Standards

U.S. Department of Commerce

Central Bureau of Nuclear Measurements

Commission of European Communities

U.S. Energy Research and Development Administration

Electric Power Research Institute (U.S.A.)

American Nuclear Society

V 0 SI

ex

American Physical Society

Nuclear Energy Agency (Europe)

International Union of Pure and Applied Physics

with the Cooperation of the

International Atomic Energy Agency

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

Dr. Sidney Harman, Under Secretary

Jordan J. Baruch, Assistant Secretary for Science and Technology

NATIONAL BUREAU OF STANDARDS, Ernest Ambler, Acting Director

Issued October 1977
Library of Congress Cataloging in Publication Data

International Specialists Symposium on Neutron Standards and

Applications, National Bureau of Standards, 1977. Neutron

standards and applications.

(NBS special publication ; 493)

Supt.ofDocs.no.: €13.10:493

1. Neutrons—Standards—Congresses. I. Bowman, Charles D., 1935- II. United States. National Bureau of Standards. III.

Title. IV. Series: United States. National Bureau of Standards.

Special publication ; 493

QC100.U57 no. 493 [QC793.5.N4622] 602'. 1 s 77-14317

[539.7'2 13*021]

National Bureau of Standards Special Publication 493

Nat. Bur. Stand. (U.S.), Spec. Publ. 493. 379 pages (Oct. 1977)

CODEN: XNBSAV

For sale by the Superintendent of Documents, U.S. Government Printing Office Washington, D.C. 20402 - Price $8.50

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PREFACE

These are the proceedings of an International

Specialists' Symposium on Neutron Standards and Applications held at the National Bureau of Standards, Gaithersburg, Md. from March 28-31, 1977. The field of

neutron standards now spreads across a rather broad

range of subjects including absolute neutron source strength, differential and integral standards for

reaction cross sections, and standards for neutron

dosimetry for nuclear reactor core and containment,

personnel protection, cancer therapy, and fission

reactor design. The purpose of the Symposium was to bring together workers from this spectrum of fields

for the purpose of reviewing the present status, tying

the different branches more closely together, and planning for a more coordinated international program in

the future.

In order to achieve the broad and even coverage

of the field, the papers were presented by invitation

only although the conference was open to anyone. In spite of this and its billing as a symposium for specialists, 142 registrants were recorded with 34 participants

from 13 foreign countries. The symposium closed with

a sunmary session which was recorded and is reproduced

in full in these proceedings for those interested in

comments and review of the full conference.

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 con- venience 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. To make the proceedings

more useful we include, as well as a table of contents,

an author index, a list of participants, and a CINDA

index of subject matter. We would like to thank

Dr. Charles Dunford of Brookhaven National Laboratory

for the preparation of the CINDA index.

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.

We wish to express our appreciation for the

financial support from the U. S. Energy Research and

Development Administration, The Central Bureau of

Nuclear Measurements and the National Bureau of

Standards which made the publication possible.

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. Sara Torrence and Mrs. Lynn Boggs of the

Office of Information Activities for help in the

arrangements for the Symposium. We are appreciative

also of the advice and suggestions of Dr. Bryan Patrick,

Harwell and of the excellent secretarial assistance of

Mrs. Julia Marks, Mrs. Gracie Wood, and Mrs. Sarah

Stewart.

C. D. Bowman and A. D. Carlson

NBS

H. 0. Liskien

CBNM

L. Stewart

LASL

III
PROGRAM COMMITTEE

M. R. Bhat, BNL

J. B. Czirr, LLL

F. Gabbard, Univ. of Kentucky

W. W. Havens, Jr., Columbia University

B. R. Leonard, Jr. , BNW

H. Liskien, CBNM

R. W. Peelle, ORNL

A. B. Smith, ANL

P. G. Young, LASL

C. D. Bowman, NBS, Chairman

V

IV
ABSTRACT

These proceedings contain forty-seven papers, which were presented

at the International Specialists Symposium on Neutron Standards and

Applications held at the National Bureau of Standards on March 28-31,

1977. The topics addressed at the Symposium include light-element cross

section standards, capture and fission cross section standards, integral

neutron standards, flux measuring techniques, and medical and personnel

dosimetry.

Key words: Cross section standards; dosimetry; fission; flux; measuring

techniques; neutrons; standards.

V
TABLE OF CONTENTS

Contents Page

Preface Ill

Program Committee IV Abstract V Table of Contents VI Papers: Session A through I 1 -346

Summary Session 347 -359

List of Participants 260 -363

Author Index 354

Citation Index 365 -370

SESSION A: INTRODUCTORY REMARKS

Chairman: C. D. Bowman

A. 1. Welcome by Ernest Ambler 1

•

A. 2. Remarks by H. Liskien 2

SESSION B: THE 6Li(n,a) STANDARD*

Chairman: P. G. Young

B. 1. Survey of Recent Experiments for the ^Li-System by H.-H. Knitter 3-9

B. 2. Angular Anisotropy in the ''Li(n,a)3H Reaction Below 100 keV by J. A. Harvey and I. G. Schroder 10 - 13

B. 3. Experimental Data Base for the Li-7 System by H. Derrien and L. Edvardson 14 - 29

B. 4. R-Matrix Analysis of the ''Li System by G. M. Hale . . 30 - 36

B. 5. Special Problems with &Li Glasses by G. P. Lamaze 37 - 42

B. 6. Instruments for Use of ^Li as a Standard by L. W. Weston 43 - 46

SESSION C: LIGHT ELEMENT STANDARDS OTHER THAN ^Li

Chairman: S. W. Cierjacks

C. 1. Experiments and Theory for Differential n-p Scattering by C. A. Uttley 47 - 53

C. 2. Use of the n,p Scattering Reaction for Neutron Flux Measurements by J. B. Czirr 54 - 60

C. 3. Surface Barrier Spectrometers for Calibration of Fast Neutrons in MeV Range by

0. P. Joneja, R. V. Srikantaiah, M. R. Phiske, J. S. Coachman and M. P. Navalkar 61 - 66

C. 4. Review of i°B(n,a)^Li Cross Section Measurements in the Energy Range From 10 keV

to 1 MeV by E. Wattecamps 67 - 84

C. 5. Instruments for Use of i^B as a Standard by A. D. Carlson 85 - 92

C. 6. Evaluation and Use of Carbon as a Standard by J. C. Lachkar 93 -100

SESSION D: MEDICAL AND PERSONNEL NEUTRON DOSIMETRY

Chairman: J. L. Fowler

D. 1. Need for Improved Standards in Neutron Personnel Dosimetry by John A. Auxier 101 -105

D. 2. Standards in Medical Neutron Dosimetry by J. J. Broerse 106 -114

D. 3. NBS Facilities for Standardization of Neutron Dosimetry From 0.001 to 14 MeV by

0. A. Wasson 115 -120

D. 4. International Neutron Dosimetry Intercomparisons by R. S. Caswell 121 -127

*The session on ^Li was planned for the program committee by the Nuclear Energy Agency Nuclear Data

Committee as a consequence of its strong interest in this standard and its associated uncertainties.

VI
SESSION E: FISSION REACTOR INTEGRAL NEUTRON STANDARDS

Chairman: A. B. Smith

E. 1. Reactor Core Dosimetry Standards by Willem L. Zijp 128 - 136

E. 2. Standards for Dosimetry Beyond the Core by Frank J. Rahn, Karl E. Stahlkopf,

T. U. Marston, Raymond Gold and James H. Roberts 137 - 145

E. 3. Fission Yields: Measurement Techniques and Data Status by W. J. Maeck 146 - 155

E. 4. Fission Reaction Rate Standards and Applications by J. Grundl and C. Eisenhauer .... 156 - 164

SESSION F: FISSION STANDARDS I Chairman: J. A. Harvey

F. 1. Utility and Use of Neutron Capture Cross Section Standards and the Status of the Au(n,Y)

Standard by A. Paulsen 165 - 169

F. 2. Remarks on the 2200 m/s and 20°C Maxwellian Neutron Data for U-233, U-235, Pu-239 and

Pu-241 by H. D. Lemmel . 170 - 173

F. 3. An Assessment of the "Thermal Normalization Technique" for Measurement of Neutron

Cross Section vs Energy by R. W. Peelle and G. de Saussure 174 - 181

252

F. 4. Review of v For Cf and Thermal Neutron Fission by J. W. Boldeman 182 - 193

252

F. 5. Measurement of the Cf Spontaneous Fission Neutron Spectrum by M. V. Blinov,

V. A. Vitenko, and V. T. Touse 194 - 197

F. 6. Prompt Fission Neutron Spectra by Leona Stewart and Charles M. Eisenhauer 198 - 205

F. 7. On Quantitative Sample Preparation of Some Heavy Elements by A. H. Jaffey 206 - 211

SESSION G: CROSS SECTION - INDEPENDENT FLUX MEASUREMENT TECHNIQUES

Chairman: R. C. Block

G. 1. Black and Grey Neutron Detectors by F. Gabbard 212 - 220

G. 2. Associated Particle Methods by Michael M. Meier 221 - 226

G. 3. Associated Gamma-Ray Technique for Neutron Fluence Measurements by J. D. Brandenberger 227 - 233

G. 4. Associated Activity Method by K. K. Sekharan 234 - 236

G. 5. Accuracies and Corrections in Neutron Bath Techniques by E. J. Axton 237 - 243

G. 6. International Comparison of Flux Density Measurements for Monoenergetic Fast Neutrons

by V. D. Huynh 244 - 249

G. 7. Calibration and Use of Filtered Beams by R. B. Schwartz 250 - 254

45 56

G. 8. Much Ado About Nothing: Deep Minima in Sc and Fe Total Neutron Cross Sections

by R. E. Chrien, H. I. Liou, R. C. Block, U. N. Singh, and K. Kobayashi 255 260

SESSION H: FISSION STANDARDS II Chairman: G. Bartholomew

H. 1. The U-235 Neutron Fission Cross Section From 0.1 to 20.0 MeV by W. P. Poenitz 261 - 268

H. 2. Propagation of Uncertainties in Fission Cross Section Standards in the Interpretation

and Utilization of Critical Benchmark Measurements by C. R. Weisbin and R. W. Peelle. . 269 - 277

H. 3. ^^''np and ^^^U as Possible Standards for the MeV Region by S. Cierjacks 278 - 289

H. 4. Standard Integral Measurement Facilities by A. Fabry 290 - 298

H. 5. Integral Measurement Results in Standard Fields by D. M. Gilliam 299 - 303

H. 6. Absolute Fission Cross Section Measurements Using Fixed Energy Neutron Sources

by G. F. Knoll 304 - 309

VII
H. 7. Impact of ENDF Standards on Fast Reactors by Ugo Farinelli 310 - 312

poc

H. 8. Absolute U, U, Np Fast Neutron Fission Cross Section Measurements by

V. M. Adamov, B. M. Alexandrov, I. D. Alkhazov, L. V. Drapchinsky, S. S. Kovalenko,

0. I. Kostochkin, G. Yu. Kudriavzev, L. Z. Malkin, K. A. Petrzhak, L. A. PI eskachevsky

,

A. V. Fomichev, V. I. Shapakov 313 - 318

SESSION I: SPECIAL TOPICS

Chairman: R. F. Taschek

I. 1. Neutron Energy Standards by G. D. James 319 - 328

I. 2. Neutron Transport Calculations for the Intermediate-Energy Standard Neutron Field

(ISNF) at the National Bureau of Standards by C. M. Eisenhauer, J. A. Grundl and

A. Fabry 329 - 334

I. 3. Standardization of Fast Pulse Reactor Dosimetry by A. H. Kazi , E. D. McGarry and

D. M. Gilliam 335 - 341

I. 4. Dosimetry Standards for Neutrons above 10 MeV by H. H. Barschall 342 - 346

SUMMARY SESSION 347 - 359

List of Particioants 35q _ 353

Author Index 354

Citation Index -n^ JbD - J/U

VIII
INTRODUCTORY REMARKS

Session A:

Dr. Ernest Ambler, Acting Director

National Bureau of Standards

It is indeed a pleasure for me to welcome you to the National Bureau of Standards

for this Symposium. I understand that 15 countries are represented and that about

one-third of you are from outside of the United States. I strongly believe that

almost every aspect of basic standards requires close international cooperation and

collaboration and that such interactions will be established and strengthened at

this meeting.

I might add that here at NBS we place emphasis on the international aspect of

all our work. I am proud of the strong line of communication that we maintain around

the world through conference sponsorship and participation, laboratory visits,

corroborative activities and committee work in international organizations in setting

standards. Dr. Charles Bowman's fine service as chairman of this Symposium is just

one example - an example that I am very proud of - of our activity in this area.

The international flavor and significance of improved neutron standards are clearly

reflected in the endorsements which this Symposium has received. I note the endorsements

of four international organizations - The Central Bureau of Nuclear Measurements, The

Nuclear Energy Agency (Europe), The International Union of Pure of Applied Physics, and

the International Atomic Energy Agency. Within the United States, the Symposium has

been endorsed by four divisions of the Energy Research and Development Administration,

the industry-operated Electric Power Research Institute, the American Physical Society,

the American Nuclear Society, and of course by the National Bureau of Standards.

These endorsements reflect the importance and diversity of the field of neutron

standards. For energy production by fission, a knowledge of neutron cross sections to high accuracy is central. In fusion-energy sources, at least in the near term, nearly

all of the energy is carried by neutrons. Fuel for both fission and fusion systems is

bred using neutrons.

Many medical specialists believe that high-energy neutrons provide an effective

modality for cancer therapy. Neutrons are now commonly used in activation analysis and

radiography. The need for safe working conditions, in all of these fields, requires

accurate neutron personnel dosimetry.

The National Bureau of Standards has long been aware of the need for standards in

the field of neutron based technology. Our program goes back more than 20 years to the NBS initiative in developing the radium-beryllium neutron source standard widely

known as NBS-I. Increased efforts in this field during the late fifties and early

sixties led to the first symposium in this country on the subject of neutron standards.

Sponsored by the European-American Nuclear Data Committee, the meeting took place in

1970 at the Argonne National Laboratory. This was followed by the IAEA-sponsored panel

which met in Vienna, Austria in 1972 and further delineated the challenges in this field.

Here at NBS we were increasingly recognizing the importance of this field. While

Director of the Institute of Basic Standards of NBS, I strongly encouraged a growing

program in neutron standards within our Center for Radiation Research.

Our work has broadened into a program in both differential and integral neutron

standards carried out using such major facilities as our 3-MeV positive ion Van de

Graaff accelerator, our 10-MW reactor, our lOO-MeV electron linac, and several other

specialized facilities.

Our program now spans nine decades of neutron energy from subthermal to 20 MeV. We can count eight calibrated neutron fields which are available as flux standards

for outside use. We have on-going, active programs in support of the full spectrum

of neutron applications including fission nuclear energy programs, fusion energy,

medical and personnel neutron dosimetry. I hope that you will be able to get an opportunity to visit our experiments and facilities in spite of the busy schedule of

the Symposium.

I am justifiably proud of our commitment at NBS to neutron standards, but I strongly believe that our program should be a partnership with other neutron standards

activities both within and outside the U.S. Neutron work in general is perhaps the

most technically difficult area of work in ionizing radiation. Progress in neutron

standards often comes slowly and after much intercomparison and corroboration.

Therefore, I believe that we all recognize a unique opportunity here to advance

the field. Let us hope that some problems will be laid to rest, that progress in

international coordination and collaboration can be made in others, and that the new problems in neutron standard will be recognized. You now face a very full four-day

schedule. Let me again welcome you here and wish you much success.

1
Dr. H. Liskien, Chairman

INDC Subcommittee on Standard Reference Data

There are still many scientific/technical fields where essential effort

has to be devoted to the change of units and the harmonization of standards.

This is due to regionally independent approaches in the past with results which

are insufficient for the international interchanges of today. In contrast to this, a modern field like neutron technology - and especially the data aspect

of it - has been an international enterprise nearly from the beginning. Today

we have WRENDA, an international request list for neutron data, we have CINDA,

an international index to literature containing information on microscopic

neutron data, and we have under the EXFOR agreement an international network for

the compilation, exchange and retrieval of experimental neutron data.

It is therefore not at all astonishing that also the task to establish a set of neutron data standards has seen an international approach. In a month's

time it will be ten years since the IAEA convened a panel on "Nuclear Standards

for Neutron Measurements" in Brussels. In 1970, the European-American Nuclear

Data Committee organized a symposium on "Neutron Standards and Flux Normalization",

while the 2nd IAEA panel on "Neutron Standard Reference Data" was held in Vienna

four and a half years ago. In this sense we are here celebrating the opening

of the fourth international meeting on this subject.

In such a meeting, we try to compare critically our established set of

standards with the real needs, to assess the progress in methods and accuracy and,

to identify weaknesses to be eliminated in the future. The progress itself is of course achieved between the meetings in our home laboratories by improving detectors,

performing accurate measurements and evaluating best values. But also this work

should see more international cooperation. In this respect two good examples are taken from the recent past: l)the first round of comparing results on fluence determinations between various laboratories organized by the Bureau International des Poids

et Mesures and 2)the work of an INDC ad-hoc group to establish a set of neutron

energy standards. I think it would be most fruitful if this symposium could initiate

more such common attacks on crucial points in the field of "Neutron Standards and

Applications".

2
SURVEY OF RECENT EXPERIMENTS FOR THE ^Li-SYSTEM

H. -H. Knitter

Central Bureau for Nuclear Measurements

B-2440 Geel, Belgium

Recent experiments on reactions relevant to the Li-system are described. It concerns the

reactions ^Li(n,t)4He, 6Li(n,n)6Li, 4He(t , n)6Li and 4He(t , t)4He , for which differential cross

sections ai(E , 0) , angle integrated cross sections Oi(E) and neutron total cross sections arp(E) were measured. Also polarization experiments yielding the analyzing power A^(E , 0) are described.

(measurements review; polarization; ^H(Q;,^Li)n,

Introduction

6 4 The Iji(n,t) He reaction cross section is used in

many situations as a standard and it is also of importance in tritium breeding calculations for thermo-nuclear fusion reactors.

Since from the theoretical point of view the reaction data from different entrance and/or exit channels

of the same nuclear system are interconnected with

each other, it seems reasonable to use the experimental information contained in all the experiments made

on the ^Li- system for a quantitative description of the

6Li(n,t)4He standard cross section. The most pertinent reactions of the ^Li-system which are of interest

for the ^Li(n,t)4He standard cross section are displayed in table 1. They are put together in two groups of

reactions. The upper part of the table gives the

^L.i(n,t)4He reaction itself and its inverse reaction

4He(t,n)6Li as well as the scattering processes to the

respective entrance channels. The lower part of the

table shows reactions of less importance. Besides the

capture process, which has only a very small cross

section-^ , they are energetically possible only above

that excitation energy region of the 7Li -nucleus which

is of direct interest for the application of the 6Lii(n,t)4He

reaction as a standard. However in a complex analysis

the second group of reactions can provide useful information too, especially on the characterization of

the "^Li-level scheme, at high excitation energies.

Ve(t,t)^He; ^Li 7 neutron cross sections; Li system

Table 1 : Most pertinent reactions leading to the

system

LiReaction Q- Value (MeV) Reaction

^Li(n,t)'^He + 4. 787

4 6 He(t,n)^Li

Li(n,n) Li 0. 000

4 4

He(t,t) He

^Li(n, Y)^Li + 7. 252 Li(Y.n) Li

^Li(n,n'd)'^He - 1. 471

^Li(n,n')^Li* - 2. 184

^Li(n,p)^He - 2.727

^Li(n,d)^He - 2.36

Due to short coming in time, only recent experiments performed on the reactions displayed in the upper part of table 1 will be discussed. Since contributions to this symposium will be given which deal with

the experimental data base of the ^Li- system and also

with their theoretical interpretation only experimental

methods and techniques are presented.

6 4 Measurements of the Li(n,t) He cross section

There are several recent measurements of the integrated ^Li(n,t)4He reaction cross section. The two

experiments performed by Lamaze et al. ^ and by

Gayther'^ cover the neutron energy range from some

keV to several hundreds of keV. The measurements of

Bartle4 and Bartle et al. ^ cover the neutron energy

range from 2 to 10 MeV and 10 to 14 MeV respectively.

100

CHANNEL NUMBER

200

Fig. 1 : Hydrogen proportional counter pulse height

spectrum for 19 keV neutrons^.

3
The experiment of Lamaze et al. ^ uses the NBSlinac neutron time -of- flight facility. In the neutron

energy range from 3 to 600 keVthe 6L,i(n,t)4He events

were detected in a 0. 5 mm thick ^Li loaded glass scintillator of type NE 912 coupled to a photomultiplier tube. The shape of the cross section was measured relative to the energy dependence of the n-p scattering cross section. The recoil protons were detected with

a hydrogen-filled proportional counter of 5. 08 cm diameter and 60. 96 cm length positioned at the end of a 200 m beam line. A collimated beam of 2. 5 cm diameter was incident on the front face of the proportional

counter. The output signals were sorted simultaneously

according to both, time -of- flight and pulse height. A

pulse height spectrum obtained with this proportional

counter is shown in fig. 1^. The normalization of the

relative cross section curve was done in the keV-range with respect to the 6Li(n,t)4He cross section itself,

in a region where it is known pretty well. This cross

section was obtained with an accuracy of better than

+ 3 % in the region of the resonance.

scintillator. The shape of the ^Li(n,t)'^He cross section was measured relative to the ^^^U{n,{) cross section using the numerical values as evaluated by Sowerby7. The n, as in the case of Lamaze et al. ^ the relative cross sections were normalized in the 2-10 ke V

range to the ^Li(n,t)4He cross section itself. An error

of + 3 % was assigned to the evaluated 235u(n, f) cross

section. Across the resonance the two measurements

are compared with each other in fig. 2.

The experimental set-up of an absolute measurement, based on the associated particle method as used

by Bartle"^, is shown in fig. 3 and is discussed in detail in ref. 8. The deuteron beam, produced by the

EN tandem accelerator of the University of Wisconsin,

is well collimated for reaction kinematical reasons. In the energy range from 2 to 10 MeV the neutrons are produced by the 2D(d,n)3He reaction. The neutron producing targets are made of deuterated polyethylene

foils, 0. 5 urn thick. Besides ^He-particles also CouVETO oeTECTOR

• Lamaze et al. (1976) _ Knitter et al. (1976) A Goyther (i977)

"miDETCCTOR

0.1 0.2 0.3 0.4 0.5

Neutron Energy [MeV]

Fig. 2 : ^Li(n,t)4He cross section is plotted versus

incident neutron energy. •, - and A are results of Lamaze et al. ^, Knitter et al. 18

and Gayther ^ respectively.

Another recent and similar measurement of the

integrated ^Li(n,t)4He cross section was made by

Gayther^ in almost the same neutron energy range

from some keVto 800 keV, using the Harwell-Linac.

The detector and sample was a 1 mm thick ^Li glass

DEUTERATED POLYETHYLEfC

T4R0ET

Siran ALLMMIUM

0 10 20 cm

SaNTLLATOR-PHOT0MULTPl.lER COMBMnOH

Fig. 3 : Experimental set-up for the associated particle method of Bartle et al. ^

.

lomb scattered deuterons and a -particles from the

l^C(d,a)^^B reaction are incident on the ^He associated particle detector. This is just thick enough to stop

the 3He-ions and it is followed by a veto detector. Since the scattered deuterons have a longer range than

the ^He particles, they are stopped in the veto detector. An anti-coincidence circuit prevents the deuteron

signals from being present in the ^He - spectrum. The

coUimation of the ^He particles together with the reaction kinematics defines the associated neutron cone and determines also the energy of the neutrons. In the

present experiment the energy spread of the neutron

beam varied with neutron energy from 80 keV at 2 MeV

to 500 keV at 10 MeV.

As a ^Li target and detector served a ^Li I (Eu)

scintillator crystal of 2. 5 cm in diameter and 1. 3 cm

thick viewed via a light guide by a photomultiplier.

This crystal is positioned such that it encompasses the

associated neutron beam. The cross section determination is possible with a typical uncertainty of + 3 %.

The results are shown in fig. 4^. The 6Li I (Eu) crystal

has different responses for a-particles and tritons. In the case of fast neutrons this property permits to obtain information about the angular distribution of the

^Li(n,t)4He reaction, since the centre -of-mass motion

causes the emitted tritons and a-particles to have

1+
a range of energies in the laboratory system. The

energy of each product particle is a known function of

the emission angle of e. g. the triton. Since each product particle gives different response, also the total

response of the two particles is an unambigious funcrange from 2 to 20 MeV can be covered. The measure- ment of the ^Li(n,t)4He cross section was reported up

200

E

o

UJ

CO

(/> inO

(E

ISO

100

50 -

*Li(n.a)t PrtMnt Meotutwntnlt

- Pendltbury Evaluation (1964)

Clainenit and Rickord Mcoturamcntt (1972)

4 5 6 7 8 NEUTRON ENERGY (MeV)

10

Fig. 4

6 ., ,4 Li(n,t) He cross section versus neutron

energy. •, - and o are results of Bartle et al5,

Pendlebury9 and Clements and Rickard^^.

tion of the angle. Fig. 5 shows some of the pulse height

spectra obtained by Bartle'^. Both response functions are needed to unfold the spectra to obtain angular distributions. They were obtained from the total widths

of the distributions. Although in general the measurements are done with polarized neutrons one obtains angular distributions which are the same as for an unpolarized neutron beam, since the experiment makes an

inherent integration over the whole azimuthal angle.

500 -

100 -

CO3O

U

EXPERIMENTAL

PEAK -SHAPES

COMPARED

WITH IDEALIZED PEAK -SHAPES FOR

ISOTROPY IN CM. SYSTEM

80 100 CHANNEL

Fig. 5 : Pulse height spectra of triton plus a-particles as obtained by Bartle et al. 4.

A new and similar system was recently set up by

Bartle et al. ^ at the cyclograaff of the Australian National University in Canberra. Due to the higher deuteron energy available, an associated neutron energy

to 14 MeV incident neutron energy^.

Rather extensive angular distribution measurements

for the ^Li(n,t)4He reaction below 2 MeV were done

already some years ago by Overley et al.ll. Besides

the above mentioned measurements of Bartle there are

angular distribution measurements by Rosario-Garcia

et al. 12 at five incident neutron energies between 4. 71 and 7. 25 MeV. They produced neutrons by the D(d,n)4He

reaction, utilizing the deuteron beam from the dynamitron of the State University of New York in Albany. The

charged particles were detected with a counter telescope consisting of three proportional counters and one

silicon detector. The sum signal of the proportional

counters served as AE signal whereas the silicon detector was the E detector. Particle identification was

accomplished using an Ortec 423 particle identifier

module. This particle identifier uses the method of

Goulding et al. in which E and AE signals are used to produce an empirical identification function^'^ which

has a nearly unique value for each particle type regardless of energy. Absolute differential cross sections

were obtained by comparison with neutron proton elastic scattering from a 9. 7 mg/cm^ thick polyethylene

target placed inside the telescope instead of the ^Li

target. However, the angular distributions measured

by Bartle4 and by Rosario-Ga rcia ^ ^ show evidently

different shapes.

The anisotropy of angular distributions of the

^Li(n,t)4He reaction at 25 keV neutron energy was

measured by Schroeder et al. using an iron-filtered

beam and a solid state detector device. S. Raman et

al. 1^ measured the a to triton ratio at 180° from 0. 5 eV to 25 keV using a diffused junction silicon detector.

This ratio deviates from unity by 0. 005 eO- 54 where

E is the neutron energy in eV. These measurements

will be treated in a separate contribution to this symposium.

Neutron total cross section measurements of Li

Recent neutron total cross section measurements

were performed by Harvey et al. 1^, Knitter et al. ' and are being made by Smithl9.

The transmission measurement of Harvey was done

in the energy range from 10 eV to 10 MeV using the

ORELA as a neutron source and employing time-offlight spectrometry for the energy determination. At

several discrete energies the total cross section was

measured in a separate experiment employing iron

filtered beam technique. The results obtained with the

two techniques agreed. The total cross section for the

sample material (0. 9875 (^Li) + 0. 0135 (^Li)) is given

and reaches a maximum value of ( 1 0. 97 + 0. 1 0)b at an energy of 246 keV. A correction for the ^Li content is

not made, but can reach in certain energy ranges values near to 1 %.

The total cross section measurement of Knitter et

al. 1^ was made at the Van de Graaff accelerator of

the CBNM in the energy range from 0. 08 to 3. 0 MeV

using monoenergetic neutrons. The neutrons were detected with a time-of-flight spectrometer. The cross

sections were corrected for the ^Li content in the samples and for the energy spread of the incident neutron

beam. A peak value of (11. 27 + 0. I2)b was obtained at

5
an incident neutron energy of (247 + 3)keV.

The preliminary total cross section data of Smith

et al. 19 were measured using a 200 keV broad band of

pulsed neutrons centered around the energy of 250 keV

produced by the Argonne National Laboratory dynamitron accelerator. An interesting energy calibration

method was applied. At distances of 6. 5, 7. 0 and 7. 5 meters, between neutron producing target and time-offlight detector the y-rays associated with the subsequent (~ 1000 nsec later) neutron burst fall just below,

about at, and just above the resonances at 250 keV.

These associated y-ray peaks provide sharp energy ca- libration markers. The according neutron energies depend essentially only from the frequency of the crystal

oscillator of the accelerator. The shape of the cross

section curve, however, was measured at a shorter

distance of about 4 m distance, because here the transmission spectra are not disturbed by the peak of the y- flash. The energy scale transfer to the shorter distance was made by the five major iron resonances in the

range 170 to 270 keV. The energetic positions were

measured also with the above mentioned vernier method

The data are preliminary and, therefore, they are not

displayed here.

Neutron scattering experiments

Recent neutron scattering angular distribution mea- surements were made by Knitter et al. ^^in the energy

range from 0. 2 to 3. 0 Me V, by Lane et al. ^'^ from 4. 0 to 7. 5 Me V, and by Bilpuch et al. 21 from 7. 5 to 14 MeV.

The gap between 3. 0 and 4. 0 MeV is being closed by

Smithl9 with a good overlap at the low energy side. All

measurements are done employing conventional nano- second time -of -flight technique using pulsed beams

from electrostatic accelerators. The measurements of

Knitter et al. and Bilpuch et al. are made relative to the n-p scattering cross section, the ones of Smith et

al. are relative to the carbon scattering cross section

and the ones of Lane et al. are relative to the zero degree differential cross section of the T(p,n)3He reaction. Fig. 6 shows the results of Lane et al. in

terms of Legendre polynomial expansion coefficients

and some representative angular distributions.

Fig. 7 was taken from ref. 18 and shows experimental values of the total cross section and the integrated elastic cross section. The two upper full lines are obtained from a simultaneous fit to the experimental total cross sections and integrated elastic cross

sections by a single level Breit-Wigner formula. The

two lowest full lines give the limits for the ^Li(n,t)4He cross section resulting from the fit. Fig. 2 shows the

data of the most recent ^Li(n,t)4He cross section of

Lamaze et al. 2 and of Gayther 3 The two full lines

have the same meaning as in fig. 7.

3 6 The reaction H(a, Li)n

3 6

It is evident that the inverse reaction H{a., Li)n

will have a large weight in the evaluation of the

^Li(n,t)4He cross section. Recently, the excitation

function at zero degree in the laboratory system was

measured by a group of the Los Alamos Van de Graaff

facility and this work will be published soon^'^. Alpha

particles are incident on a tritium target. The threshold a-particle energy is 11. 136 MeV, and, at the laboratory energy of 11. 61 MeV one reaches the 5/2"

level in the ^Li compound system which is visible in the

inverse reaction channel at about 240 keV laboratory

neutron energy. Due to the large negative Q-value, the

^Li- recoils are kinematically coUimated to a cone of

about + 4 degrees. The ^Li particles are separated

spatially from the incident beam and from the Coulomb

scattered particles by a magnetic field. The zero and

180° centre -of -mas s cross sections were obtained

from the two kinematic groups present at zero degree

in the laboratory system. Their difference in energy is

about 1. 6 MeV. The reaction data were normalized by

measuring simultaneously the yield of elastically recoiling tritons. More details can not be given, since

this work is not yet published.

3.U O.U ,[ Q NEUTRON ENERGY ( " E V ) c 0 S ( > ) Fig. 6 : Neutron scattering cross section results of

Lane et al. ^0 in terms of Legrendre polynomial expansion coefficients and some representative angular distributions.

01 02 03 04 05 06 07

Neutron Energy [MeV]

Fig. 7 : Neutron total, integrated elastic and (n,t) cross

sections of ^Li versus incident neutron energy! ^. A and are data of ref. 22 and 23 respectively.

6
4 4 The He(t,t) He scattering experiments

7 This entrance channel to the Li-system consists

of a spin-zero and a spin-l/2 particle and therefore

the interpretation of the data is rather easy. Earlier

studies of the triton - a, scattering process have contributed considerably to the understanding of the ^Lisy stem'^-' ' ^° ' '. Recent measurements did not come

to my knowledge. There are unpublished data in the

triton energy range from 7 to 14 MeV from the Los

Alamos tandem Van de Graaff facility displayed in graphical form in the work of Hale^^ on the R -matrix analysis of light element standards. However, these data

will be published soon24 under ref. 29.

Experiments using polarized projectiles

Studies of polarization effects can provide information of a basic nature, which can be obtained only

indirectly or sometimes not at all using traditional ex- perimental methods^*^. This was demonstrated already in the first nuclear polarization experiment which

was made by Heusinkveld and Freier in 1951 at the

University of Minnesota^'. This ^He-proton double

scattering experiment gave the level sequence for the

P-state doublet of the ^Li*nucleus, an information

which could not be obtained by conventional scattering

experiments-^ .

The polarization power of ^Li or neutrons was

measured at the electron Linac of the Yale University

33,34 Fig. 8 shows the schematics of the experimental set-up. A beam of unpolarized neutrons is produced at the (y ,n)-target of the Linac. The neutrons become partially polarized by the first scattering on carbon. In a second scattering the analysing power A(0)

of ^Li and of other nuclei were measured in the energy range from 2 to 5 MeV. The polarization of the incident neutrons was obtained from a separate scattering experiment in which the polarizer and the analy- zer consisted of the same material of spin zero nuclei,

in which case the analyzing power is equal to the polarization power32_ The solenoid which is positioned

in the flight path served to precess the neutron spin.

scattering cross section data of Lane et al. 22 their

own polarization results. The differential scattering cross section for unpolarized neutrons 0(6) and the polarization can be given in form of expansions

a(G) = E Bl Pl (cos 6) L

a(e) . p(e) =9t^ EClPl(cos 9) L

where Pl(cos 9) and Pi (cos 9) are the Legendre and

associated Legendre polynomials. The Bl's as functions of the collision matrix are derived by Blatt and

Biedenharn^^ with the phase correction of Huby36 and

the Cj_^'s are given as function of the collision matrix

by Simon and Welton^^.

The analysing power for the ^He(t,t)4He reaction

was measured by a team of the Los Alamos tandem

Van de Graaff facility38 , 39 at centre-of-mass angle s

0£

*1 0.0

4 •» 4 He(t.f)^He

7.0 e.O 9.0 10.0 II.O 12.0 13.0 140

Triton Energy (MeV)

Fig. 9 : Excitation function for ^He(t,t)*He analyzing

power-^®.

Gr«^ll« Scstttrtr

Intilcal Skdl

•. X I aa Ikiek)

Osyf** Sc«tt«r«r

(Crlli><rlc*l Sk<ll 7.S ca Ala.x t» tklckl

Pl»lle SclMlllatwa - (It.Sea <l<.xlSc(k Ikl(k)

Fig. 8 : Experimental arrangement for measuring the

analyzing power in neutron scattering experiments at the Yale-University32_

and the analyzing power could be obtained from the

count rate ratios with and without magnetic field and

from the spin preces sion angle. Firk et al. ^ 3 , 34 made

a R-matrix analysis for the '''Li system, using the

TIME PICK-OFF

PROTON BEAM

^Li TARGET

> X 10 AMP

BEAM MONITOR

0 12 3 4

SCALE INCHES PRE-AMP

Fig. 10 : A plan view of the experimental arrangement

to measure the analyzing power of the

^Li(n,t)4He reaction.

T
of 49. 6° and 123° in the energy range from 7 to 14 Me V

In addition angular dependence of the polarization power at 8. 79 MeV, 10. 82 MeV and 12. 25 MeV was obtained. The polarized tritons were produced with the

Los Alamos Lambshift polarizing ion source for tritium40 At the target of the accelerator a beam current of 80 nA with a polarization of 0. 8 has been obtained. Fig. 9 shows a part of the results. The reso- nance near 8. 8 MeV is of particular interest, since it corresponds to the resonance near 240 keV in the

6Li(n,t)4He reaction.

The analyzing powers of the ^Li(n,t)'^He reaction

was measured by Karim and Overley^l with the University of Oregon pulsed Van de Graaff accelerator in

the neutron energy range from 0. 2 to 1.4 MeV. Fig.

10 shows the experimental arrangement. With a thick

metalic lithium target polarized neutrons are produced in the energy range between 0. 2 and 1. 4 MeV at the angles of + 50 degrees with respect to the proton

beam. The a and triton particles were detected in a

solid state detector and two dimensional spectra were

recorded in energy and flight time. This allowed to determine the neutron energy and to make a particle

identification. The results compare also favourably

with previous results42 at two energies near the 240

keV resonance.

Conclusion

Reference s

1. F. C. Barker , Phil. Mag. 2 (1957) 780

2 . G. P. Lamaze , O. A. Wa sson , R. A. Schrack and

A. D. Carlson , Proc. of the Int. Conf. on the Interactions of Neutrons with Nuclei, Lowell, Massachusetts, 6-9 July, 1976, Vol.2, p. 1341

3. D. B. Gayther ,AERE-Harwell, private communication 12. 1. 1977, will be published as report AERE8556 (1977)

4. C. M. Bartle , Proc. of the Conf. on Neutron Cross

Sections and Technology , Wa shington D. C. , 1 975 , Vol. 2, p. 688, NBS Special Publication 425

5. C. M. Bartle, D. W. Gibbie and C. Hollas, Proc. of

the Int. Conf. on the Interactions of Neutrons with

Nuclei, Lowell, Massachusetts, 6-9 July, 1976,

Vol. 2, p. 1342

6. O. A. Wasson.Proc. of the NEANDC/NEACRP Specialists Meeting on Fast Fission Cross Sections,

28-30 June 1976, held at Argonne National Laboratory, ANL-76-90 p. 183

7. M.G.Sowerby, B. H. Patrick and D. S. Mather , Ann.

Nucl. Sci. Eng. ]_ (1974) 409

8. C. M. Bartle and P. A. Quin, Nucl. Instr. Meth. 121

(1974) 119

9. E. D. Pendlebury, AWRE 0-60/64 (1964)

10. P. J. Clements and I. C. Rickard, AERE-R 7075 (1972)

11. J. C. Overley,R. M. Sealock and D. H. Ehlers, Nucl.

Phys. A 221 (1974) 573

12. E. Rosario-Garcia and R. E. Benenson, Nucl. Phys.

A 275 (1977) 453

13. F. S. Goulding, D.A.Landis, J. Cerny III and R. H.

Pehl, IEEE Trans. Nucl. Sci. ]A (1964) 388

14. H. Bichsel and C. Tschalaer, UCRL-17663 (1967)

15. I.G.Schroder, E. D. McGarry, G.de Leeuw-Gierts

and S. de Leeuw, Proc. of the Conf. on Neutron

Cross Sections and Technology, Washington D. C. , 1975 , Vol. 1 , p. 240

S. Raman, N. W. Hill, J. Halperin and J. A. Harvey,

Proc. of the Int. Conf. on Interactions of Neutrons

with Nuclei, Lowell, Massachusetts, 6-9 July, 1976,

Vol. 2, p. 1340

J. A. Harvey and N. W. Hill, Proc. of the Conf. on Neutron Cross Sections and Technology, Washington

D. C. , 1975, Vol. 1 , p. 244

H. -H. Knitter, C. Budtz- J^rgensen,M. Mailly and

R. Vogt, Report EUR-5726e (1977)

A. B.Smith, private communication January 24 , 1977

Argonne National Laboratory

R.O.Lane, R. M. White and H. D. Knox, Proc. of the

Int. Conf. on the Interactions of Neutrons with Nuclei, Lowell, Massachusetts, 6-9 July, 1976, Vol. 2,

p. 1043

E. G. Bilpuch, D.H.Epperson, D.W.Glasgow,

S. G. Glendinning, C.R.Gould, H.H. Hogue, P. W. Lisowski, C.E.Nelson, H. W. Newson, F. O. Purser , L. S. Seagondolla r , T. Tornow and P. von Behren,

Proc. of the Int. Conf. on the Interactions of Neutrons with Nuclei , Lowell , Massachusetts , 6-9 July,

1976, Vol. 2, p. 1309

R.O.Lane, A. S. Langsdorf Jr. , J. E. Monahan and

A.J.Elwyn, Report -ANL - 6 1 72 , and Ann. Phys. VZ (1961) 135

23. H.B.Willard, J. D. Bair, J. D. Kington and H. O. Cohn

Phys. Rev. J_OJ_ (1956) 765

24. N. Jarmie, Los Alamos Scientific Laboratory, private communication, January 10, 1977

25. T. A. Tombrello and L. S. Senhouse , Phys. Rev. 129

(1963) 2252

26. R. J. Spiger and T. A. Tombrello, Phys. Rev. 163

(1967) 964

27. M.Ivanovich, P. G. Young and G. G. Ohlsen, Nucl.

Phys. Alio (1968) 441

Recent measurements of the Li(n,t)4He cross

section were done in the incident neutron energy range

from some keV to several hundreds of keV^'-^ and

from 2 to 14 MeV4,5 Neutron total cross section measurements were made^^' and are being made^*^ in

the neutron energy range from 10 eV to 10 MeV, where

special emphasis has been paid to the region of the

250 keV resonance. Recent neutron scattering cross

section measurements were performed in different ex- periments from 0. 2 to 3. 0 MeV^^^ from 4. 0 to 7. 5 MeV^O and from 7. 5 to 14 MeV21. The gap between 2 19. and 3 MeV is being close dl9 such that the whole energy range from 200 keV to 14 MeV is being covered by £0

recent scattering experiments. Zero and 180° degree

centre-of-mass cross sections of the reaction 3h( a, ^Li)n are measured and will be available soon24 Similar is

the situation with 4He(t,t)4He scattering experiments

24,29 Also analyzing power measurements for the 2 1. ^Li(n,n)6Li,4He(t,t)4He and ^Li(n,t)4He were done in

the laboratory energy ranges from 2 to 5 MeV33,34j

from 7 to 14 MeV38,39 and from 0. 2 to 1.4 MeV41

respectively. All this recent experimental results re- present a large and valuable body of data for understanding of the ^Li-system.

22.

8
28. G.H.Hale, Proc. of the Conf. on Neutron Cross

Sections and Technology, Washington D. C. , 1975,

Vol. 1 , p. 302

29. R. A. Hardekopf, N. F. Jarmie, G. G. Ohlsen, R. V. Poore, R. F. Haglund Jr., R. E. Brown , P. A

.

Schmelzbach , D.B.Anderson, D. M. Stupin and

P. A. Lovoi, to be published LA-6188 (1977)

30. F. W. K. Firk, Proc. of the Int. Conf. on the Interactions of Neutrons with Nuclei, Lowell, Massachusetts, July 6-9, 1976, p. 389

31. M. Heusinkveld and G. Freier, Phys. Rev. 85^ (1952) 80

32. H. Faissner, Ergebnisse der Exakten Naturwissenschaften 22^ (1959) 180

33. R.J.Holt, F.W.K. Firk, G.T.Hickey andR.Nath

Nucl. Phys. A 237 (1975) 111

34. F.W.K. Firk, J.E.Bond, G.T.Hickey, R.J.Holt,

R. Nath and H. L. Schuttz , Proc. of the Conf. on Neutron Cross Sections and Technology, Washington D. C. , 1975, Vol. 2, p. 875, NBS Special publication 425

35. J. M. Blatt and L. C. Biedenharn, Rev. Mod. Phys.

24 (1952) 258

36. R.Huby, Proc. Phys. Soc. 67A (1954) 1103

37. A. Simon and T. A. Welton, Phys. Rev. 90^ (1953)

1036

38. R. A. Hardekopf , N. Jarmie, G. G. Ohlsen and R. V. Poore, Proc. of the Fourth International Symposium on Polarization Phenomena in Nuclear Reactions, Zarich, Switzerland, 25-29 August, 1975

39. R.A. Hardekopf, G. G. Ohlsen, R. V. Poore and

N. Jarmie, Proc. of the Fourth International

Symposium on Polarization Phenomena in Nuclear

Reactions, ZUrich, Switzerland, 25-29 August,

1975

40. R. A. Hardekopf , G.G. Ohlsen, R. V. Poore and

N. Jarmie, Phys. Rev. C13 (1975) 2127

41 . M. Karim and J. C. Overley , Proc. of the Conf. on Neutron Cross Sections and Technology, Washington D. C. , 1975, Vol. 2, p. 788, NBS Special publication 425

42. H. Waffler and G. Walch, Direct Interactions and

Nuclear Reaction Mechanisms, Gordon and Breach,

New York, 1964, p. 633

9
ANGULAR ANISOTROPY IN THE ^Li(n,a)'^H REACTION BELOW 100 keV

J. A. Harvey

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37830

and

I. G. Schroder

National Bureau of Standards

Washington, D.C. 20234

The data on the differential (n,a) cross section of Li has been reviewed below a neutron

energy of 100 keV. Measurements in this region, compared to that above 100 keV, have been

few. Only recently has interest increased as large and unsuspected anisotropies have been

observed in this energy region. Thus, at 25 keV an anisotropy amounting to 67% in the

forward- to-backward direction has been observed; while in the region between 1 eV and 10 keV

measurements indicate that in the forward- to-backward 66° cone the asymmetry has an energy

dependence, in the laboratory system, which can be expressed analytically as 1 + 0.0055

y^ri(eV) . These angular anisotropies seem to arise from interference between the large p- wave resonance at 247 keV and many s-wave resonances which account for the large 1/v (n,a)

thermal cross section. New and detailed measurements of the differential (n,a) cross section are needed not only to be able to account analytically for these interference effects

but also because it is necessary to consider these anisotropies when the 6Li(n,a)3H cross section is used as a standard.

(Angular distribution; fast neutrons; linac; reactor filtered beam; ^Li(n,a)T; standard)

Introduction

The 6Li(n,a)3H reaction has been increasingly

used as a standard in partial cross section measurements, for flux normalization, for neutron detection

and in neutron spectroscopy. The suitability of this

reaction as a standard rests in its Q-value (+ 4.784

MeV) and large thermal cross section (940 barns); in

the smooth 1/v behaviour of its cross section to within 1% below 10 keV, and that above this energy

and up to 100 keV the cross section varies smoothly

though no longer following the 1/v behaviour. This

is due to a large p-wave resonance (5/2") at 247 keV

which dominates the region between 100 and 500 keV.

The 6Li(n,a)3H reaction has been extensively

studied both experimentally! and theoretically.!

Most of these studies, however, have centered in the

region above 100 keV. Below this energy very little

data exists. In particular, differential (n,a) mea- surements in the region below 100 keV have been few

and mainly unpublished. 2>3 jh-js -jg regrettable as large and unsuspected anisotropies have been recently

observed in this energy region. The present review

will try to give a summary of these recent experimental results and some of their implications.

Early Measurements

The first attempt to measure the 6Li(n,a)'^H

differential cross section below 100 keV was made by

G. de Leeuw-Gierts in 19632 using ultra fine grain

nuclear emulsions loaded with ^Li at an incident neutron energy of 20 keV. The angular distribution obtained is shown on Fig. 1 together with a theoretical

fit arising from a single-level S-matrix analysis

made by Mahaux and Robaye.^ This experiment was followed by that of Beets et al.3 who used a thick

tritium target (60 keV) and measured the (n,a) angular distribution in the energy region between 60 and

100 keV. Two different experimental set-ups were

used. The first consisted of a circular multiplate

nuclear emulsion chamber with a central target.

The second experiment used the same type of ^Li

loaded fine grain emulsion as in the 20 keV experiment. The angular distributions obtained from the

two arrangements are shown in Fig. 2 together with

a theoretical fit from the S-matrix analysis of

Fig. 1. Differential cross section at 20 keV from

the data of reference 2. Solid curve is

theoretical fit of Mahaux and Robaye.^

Mahaux and Robaye.^ The discrepancies between the two

distributions can be attributed to the slightly different energy composition of the neutron beams, the non- uniformity of the epicadmium neutrons in the emulsion

experiment and the presence of neutrons > 150 keV that

could not be properly accounted for in the camera ar- rangement. Furthermore, both the 20 and 80 keV experiments using loaded emulsions were subject to rather

large thermal backgrounds of ~ 20%.

Angular Anisotropy at 25 keV and 2 keV

Using Filtered Beams

The installation of both a 25 keV iron-filtered^

and a 2 keV scandium-filtered^ beam facility at the

NBS reactor, the fact that unexplained results were

being obtained with 6Li semiconductor spectrometers in

the region below 100 ke\/,7 and the paucity of data on the behaviour of the differential (n,a) cross section

in this region prompted the study of the angular anisotropy of the 6Li(n,a)3H reaction at 25 keV8 and to later look for similar effects at 2 keV.

10
150

En - 80 keV

0 -1 cos 9 CM

Fig. 2. Differential cross section at 80 keV obtained by Beets et al.3 The circles represent the values obtained with the ^Li loaded

nuclear emulsion, while the dots represent

the values obtained with the emulsion chamber. Solid curve is from fit of Mahaux and

Rob aye.

4

To measure the angular anisotropy in the

^Li(n,a)^H reaction at 25 keV a surface-barrier detector was coated with 80 Mg/cm2 of ^Lip (front face)

and used as a 2Tr detector. This detector was placed

in two different angular positions with respect to the

neutron beam: front face at 90° and back face 90° to the beam. The pulse-height distribution of both a- particles and tritons were recorded for these two positions (Fig. 3). In a subsequent set of measurements a second surface barrier detector was placed

coaxially with the first (at 90° to the beam) and in

such a way as to subtend a 45° cone. Coincidence

measurements that simultaneously recorded the energy

distribution in both detectors allowed a measure of

the angular asymmetry in the backward- to- forward 45° cone (Fig. 4). Lastly, thermal calibration runs were

performed before and after the 25 keV measurements to calibrate the system and to insure that spurious anisotropics were not introduced electronically.

8

Subsequently a similar set of experiments were

performed with the same detectors and electronic system at the 2 keV filtered beam facility. The results

obtained together with those at 25 keV are summarized

in Table I.

TABLE I

Ratio 25 keV 2 keV

R (180°) 1.67 ± 0.11 1.07 ± 0.06

R (45°) 1.96 ± 0.06 1.24 + 0.02

CHANNEL NUMBER

2fi , 200 300 400

\

\ \ \ I r

1000 1500 2000 2500 ENERGY, keV

Fig. 3. Experimental pulse-height distribution obtained in the forward 2ti measurements made

at 25 keV. (Reference 8)

Angular Distribution of the "Li(n,a)-'H

Reaction at 25 keV

The results obtained at 25 keV using the 2it detector were complemented by a set of two other measurements: front face at 45° to the neutron beam and back

face 45° to the neutron beam. From these four 2it mea- surements and from the coincidence experiments a coarse-grided angular distribution was obtained. (The

details of the method are explained in reference 8.)

The results are shown in Fig. 5. This figure shows the

angular distribution in both the laboratory and centerof-mass system. The two dashed lines in these distributions represent the errors arising from the uncer- tainty in the determination of the background (see Fig.

3). The dotted lines in the laboratory distribution

correspond to an isotropic distribution in the centerof-mass system.

Angular Anisotropy in the Li(n,ct) H Reaction in the 1 eV to 10 keV Energy Region

The determination of the energy dependence of the

angular anisotropy of the ^Li(n,a)3H reaction in the

energy region between 1 eV and 10 keV was performed at

ORELA using a thin (101 yg/cm2) 6LiF target. 9 The

measurements were made with a diffused- junction silicon detector which subtended an average angle of 66°

11
ENERGY, keV

2000 2400

300 350 CHANNEL NUMBER

Fig. 4. Composite picture showing the pulse-height

distribution obtained in the coincidence

experiment of Schroder et al.8 performed

at 25 keV.

to the lithium target. The detector resolution was

such that both the triton and alpha groups were well

resolved so that an accurate ratio of their intensities could be obtained as a function of energy. In the

1 eV to 10 keV energy region the ORNL results give an energy dependence for the asymmetry in the forward-tobackward 66° cone (in the laboratory system) of the

form

A = 1 + 0.0055 /E^

where is the neutron energy in electron-volts.

This energy dependence is shown graphically in Fig. 6.

Conclusions

The differential (n,a) cross section of ^Li in

the region above 100 keV has been extensively studied,

most recently by Overley et al.lO No such detailed

work exists yet below 100 keV. Thus all one can say

is that the existence of the angular anisotropy below

100 keV seems to arise from interference between the

large p-wave (5/2") resonance at 247 keV and the many s-wave resonances which account for the large 1/v

(n,a) thermal cross section.

The theoretical analysis of the 6Li(n,a) cross section which ranges from the single-level, singlechannel R-matrix and S-matrix calculations of Mahaux

and Robaye4 to the multilevel, multichannel R-matrix

computations of Hale^l does not have the input which

is available above 100 keV. In order to fill this

gap it is necessary to obtain detailed angular

1.00

0.75

0.50

0.25

45

_L

90

e-\ab

135 180

1.00

0.75

0.50

0.25

45 90

9-c.rn.

135 180

Fig. 5. Angular anisotropy of the Li(n,a) H reaction at 25 keV in the lab and centerof-mass systems (reference 8).

distributions in the energy range from 1 to 100 keV.

Below this energy the differential (n,a) cross section

should be of the form a + b cose and therefore the

ORNL data as it stands seems adequate.

At present only a few experiments are contemplated. The filtered beam work at NBS using solid

state detectors will be completed at 2 keV and a measurement made at 144 keV (silicon filtered beam). 12 Lastly, two groups, one from the University of Michigan! 3 and one from HEDL^^ are planning to perform a series of differential measurements using track-etch

detectors

.

Bibliography

1. See papers on ^Li in this conference and references

therein.

2. G. de Leeuw-Gierts , PhD Thesis, Faculte de Sciences,

Universite Libre de Bruxelles (1968).

12
1 I I r

20 40 60 80 100

Fig. 6. Asymmetry in the forward- to-backward 66°

cone (in the laboratory system) of the

6Li(n,a)3H reaction between 1 eV and

10 keV, Data of Harvey et al.9

3. C. Beets, G. de Leeuw-Gierts , S. de Leeuw, Y. Baudinet-Robinet, G. Robaye, and L. Winand,

Nucl. Phys., 69, 145 (1965).

4. C. Mahaux and G. Robaye, Nucl. Phys., 74, 161 (1965).

5. E. D. McGarry and I. G. Schroder, Proceedings of

a Conference on Nuclear Cross Sections and Technology, Washington, D.C. March 3-7, 1975.

NBS Special Publication 425, Vol. 1, p. 116.

6. I. G. Schroder, R. B. Schwartz, and E. D. McGarry, Proceedings of a Conference on Nuclear

Cross Sections and Technology, Washington, D.C. March 3-7, 1975. NBS Special Publication 425,

Vol. 1. p. 89.

7. G. de Leeuw-Gierts and S. de Leeuw (private com- munication).

8. I. G. Schroder, E. D. McGarry, G. de Leeuw-Gierts,

and S. de Leeuw. Proceedings of a Conference on Nuclear Cross Sections and Technology, Washington,

D.C. March 3-7, 1975. NBS Special Publication

425, Vol. 1, p. 240.

9. J. A. Harvey, J. Halperin, N. W. Hill and S. Raman, ORNL/TM-5450, p. 14, May 1976.

10. J. C. Overley, R. M. Sealock, and D. H. Ehlers,

Nucl. Phys., A221 , 573 (1974).

11. G. M. Hale, Proceedings of a Conference on Nuclear Cross Sections and Technology, Washington,

D.C. March 3-7, 1975. NBS Special Publication

425, Vol. 1, p. 302.

12. R. B. Schwartz, this conference.

13. John C. Engdahl, private communication.

14. Raymond Gold, private communication.

13
EXPERIMENTAL DATA BASE FOR THE Li -7 SYSTEM

H. Derrien and L. Edvardson

Centre de Compilation de Donnees Neutroniques

de I'Agence pour I'Energie Nucleaire (OCDE)

B.P. No.

9

91190 GIF SUR YVETTE

FRANCE

A review of the experimental data available for the Li-7 system is given, including

reactions induced by neutrons and those induced by He-4 on H-3 or bv H-3 on He-4. For

reactions induced by neutrons only, the energy range up to about 5 MeV has been considered.

Some recommendations are given concerning possible future measurements and the validity

of the recent eval uatiois for the Li-^n,a) cross section.

(Elastic scattering; 6Li(n,a); ^L-j total; ^Li system; neutrons; review of measurements).

Introduction

The most recent international meetings on neutron

standards were held in October, 1970 at Argonne (organised by Argonne National Laboratory and sponsored by

the EANDC), and in Vienna in November, 1972, organised

by the IAEA. The problem of the Li-6(n,a) cross section was examined during these meetings. A com- plete review of the Li-6 neutron cross section in the

energy range from Thermal to 1.7 MeV was presented by

C.A. Uttley at the Argonne meeting (1) for all the

data available in 1970, but no such review was presented at Vienna in 1972. Since this date, a great

effort has been made to try to explain the large

discrepancies existing in the Li-6(n,a) cross section,

particularly in the vicinity of the 250 keV resonance

and beyond. The investigations have been made in the

following directions :

1. The energy calibration . The important differences

in the location of the resonance peak do not permit a direct comparison of the cross section from several

measurements. The problem of the energy standard has

been studied in several lai)oratories by the time-offlight method using linacs and cyclotrons (Columbia,

Harwell, Geel , Karlsruhe and Saclay). The results of

these investigations will be presented by G.D. James

at this meeting.

2. The content of Li-6 in the Lithium glasses . The

content given by the manufacturers has been questioned for some glasses used in absolute measurements;

precise measurements have been made, for instance, by

a non-destructive method : measurement of the neutron

transmission of the glasses and determination of the

Li-6 content from the 1/v variation of the total cross section at low energy.

3 . Further measurements of the Li-6(n,a) cross sectiorT The results have not been any more encouraging; for instance, a new method used by Friesenhahn in

1974 led to a cross section which is in disagreement

with all the previous results (61).

4. Analysis of the experimental data using the

nuclear reaction theories^ At the present stage, a choice has to be made amongst the disparate experimental results; a way in which to make this choice

is to try to predict the cross section values by

using a formalism which permits the calculation of

the Li-6(n,a) cross section from other experimental

results considered as accurate or from a set of

experimental results as diverse as possible. This

problem was envisaged as early as 1969 by K.M. Diment

and C.A. Uttley, (2) who calculated the (n,a) cross section at the resonance by using the parameters

obtained from the scattering and total cross sections.

We know that the cross sections obtained at the peak of

the resonance was about 1 barn higher than the

values measured by Schwarz (46). At this time, the

one-level Breit-Wigner formula used by Diment and

Uttley was suspected : for instance, it does not

take into account the interference due to a 5/2

level which lies below the neutron threshold. As a consequence, a solution to the problem which could

be considered with high confidence, should be obtained

by using a complete R-matrix formalism which takes

into account the interferences between all the

levels of same spin in all the open channels. This

kind of analysis has recently been performed by

G.M. Hale et al . (3) for the ENDF/B data files and

the results are presented at this session by the

author.

In fact, in the R-matrix formalism, one must

consider all the reactions leading to Li-7 compound

nucleus and all the possibilities of de-excitation

of this system, i.e., all the possible entrance

and exit channels. A complete compilation of the

Li-7 levels and reaction channels has recently been

performed by F. Ajzenberg-Sel ove (4); the data in

Figure 1 have been extracted from this compilation;

they show the 5/2" levels at 6.68 and 7.47 MeV

excitation energy (the neutron binding energy is

7.2506 MeV). The nuclear reactions energetically

available for our purpose are the following :

Li-6 total neutron

Li-6(n,Y)Li-7

Li-6(n,n)Li-6

Li-6(n,a)t

a-t scattering

t(a,n)Li-6

The reaction Li-6(n,Y)Li-7 has very little

importance; the corresponding cross section is

smaller than 50 mb at thermal energy and is negligible at higher energies and then the neutron absorption cross section is nearly equal to the (n,a) cross section. The a-t scattering is of great importance

because it corresponds to the exit channel of the

Li-6(n,a)t reaction. G.M. Hale et al . have particularly shown that the calculated cross sections for

Li-6(n,a)t are very sensitive to the a-t scattering

angular distributions. The same remark applies to the a(t,n)Li-6 reaction which is the inverse reaction

of Li-6(n,a)t.

In this paper, we will review the experimental

data available for the reactions listed above,

except for the Li-6(n,-Y)Li-7 reaction. The neutron

data are available in the EXFOR international files

on request from the 4 Neutron Data Compilation

Centres (NNCSC, Brookhaven; CCDN, Saclay; NDS, Vienna,

14
and CJD, Obninsk). The references given at the end

of this review are sorted following the corresponding

reactions. For the neutron entrance channel reactions,

some details concerning the experiments are also

given and the references are coded according to the

procedure used in CINDA 76/77. For the other reactions,

the references and some mixed references are given

in the usual manner.

g-t Scattering and t(ct>n)Li-6 Reactions

The experimental data available correspond to the references 8-i3. They include differential cross section, angular distribution and polarization

measurements. Some important theoretical analyses

have been done on a-t scattering and the results have

been published in the references 14-17; we think that

these works should not be ignored when trying to solve the Li-6(n,a) problem by a nuclear reaction

formal ism.

The differential cross sections and angular distributions for a-t scattering have been measured by

M. Ivanovich et al . (9) in the 3-11 MeV a energy

range; by R.J. Spiger et al . (10) for a energies

between 4 and 18 MeV and by L.S. Chuang (11) for

triton bombarding energy in the range 2-3 MeV. Phase

shift analyses have been performed by these authors

and parameters for the Li -7 levels have been obtained.

The polarization of tritons scattered by He-4 have

been measured by P.W. Keaton et al . (8) for incident

triton energy equal to 6.0; 6.9; 9.5; 10.2 and 10.25 MeV

and for several angles; concerning the phase shift

analyses, their results are in agreement with those

of Spiger et al . More recently, R.A. Harde Kopf et al

.

(12), at Los Alamos, have obtained the excitation

function and angular distribution for the He-4(t,t)He-4

analysing power in the triton energy range 7-14 MeV

for 8 incident energies; they have also measured the

angular distribution of the He-4(t,t)He-4 scattering

for 6 energies and 17 angles in the energy range

8.23-12.00 MeV; the results have been used by Hale et

al . for an R-matrix analysis.

Several theoretical analyses were performed on the above-mentioned data by different authors. The

resonating method has been used by R.E. Brown et al

.

(14) to calculate the phase shifts and the results

are in good agreement with the experimental results

of Spiger and those of Ivanovich. An optical potential analysis was performed by V.G. Neudatchin (16),

which happens to be as good as the resonating method

when applied to the results of Spiger or Ivanovich.

The shell treatment of separable interactions made by

L.M. Kuznetsova (15) is also in agreement with the

experimental results of Spiger. An R-matrix analysis

was also performed by R.C. Baker (17) on the Spiger

data. The interesting feature of these theoretical

analyses is that no inconsistencies in the experimental data have been pointed out.

Concerning the H-3( c(,n)Li -6 reaction, only two results are known. Firstly, the measurement of

Spiger et al. (10) for a incident energy 11.0-12.4 MeV

with the dominant effect of the 7.47 MeV Li -7 resonance; the results were normalized to the Schwarz

Li(n,a)t results (inverse reaction). Secondly, the

results of R.E. Brown et al . (13) not yet published : the 0 cross sections have been obtained at 15 a energies from 11.3 to 12.0 MeV (range of En=0.08 to 3.9 MeV for the inverse reaction); 2 to 5% accuracy

is expected and the results could be of great importance for solving the problem of the Li-6(n,a)

reaction.

Concerning the a-t scattering, we have not compared

the 3 most important results, i.e., Spiger et al .

,

Ivanovitch et al . and Harde Kopf et al . A direct com- parison is not easily possible because the measurements were not done at the same angles and energies.

The R-matrix analysis should be applied simultaneously

to the three sets of data to verify their consistency.

The Total Neutron Cross Section

The total cross section is generally obtained

from transmission measurements using samples of

appropriate thicknesses. If the resolution is good

enough, which is always the case in the Li -6 measurements, the transmission is an absolute measurement of

the total cross section : the measurement of the

incident and transmitted neutron flux is done with

the same detector in the same geometry. If the Li -6

content of the sample is well known, the total cross section is generally obtained with an accuracy better

than 2% and can be considered as a basic measurement

for testing the coherence of partial cross sections

which are more difficult to measure with high accuracy.

A precise knowledge of the total cross section is

particularly important at low energies in order to determine the 1/v variation of the absorption cross section, i.e., the (n,a) cross section and the thermal values can also be obtained by extrapolating the

1/v law.

The references 18-26 correspond to the total cross section measurements performed between 1954

and 1976 in the energy range from 70 eV to about

10 MeV. One notes that there is only one measurement

at low energy : that of Diment and Uttley (24) for

neutron energies above 72 eV. However, J. Harvey (25)

announced, at the 1975 Washington conference, a measurement from 10 eV neutron energy; unfortunately,

the lower part of the results have not been published

and the data present in the EXFOR files begin at 24.6 keV. There is no measurement at very low energy,

and the only way in which to obtain the total cross section below 72 eV is to extrapolate the 1/v law

obtained by Diment and Uttley, i.e.:

a.^(barn) = (149.5±0.3)//r + (0.70±0.01);

however, the extrapolation gives a thermal absorption

cross section equal to (940±2) barn, in agreement

with direct measurements at thermal energy.

The experimental total cross sections are com- pared in Figures 2 and 3 over the whole energy

range. The data of Diment and Uttley' are present in

the total energy range and are used in Table I as an element of comparison. In Figure 2, the energies

have been adjusted to obtain the peak of the reson- ance at 246 keV for all sets of data; for a given

set, the corrections are constant in relative values.

However, this method of adjustment could be inexact.

For instance, in the case of a linac time-of-flight

measurement, the part of the error due to a poor

evaluation of the so-called t increases relatively

when the time-of-flight decreases. One example of

this kind of effect is given in Figure 4 which shows

the superposition of the results of Diment and Uttley

(chosen as reference data) and those of Meadows (23);

in Meadows' results the resonance seems to be wider

than in that of Diment's results. This is probably

due to an under-estimation of the energy correction

in the high energy range of Meadow's data (relative

to Diment's data).

Table I shows that, with the exception of the

old values of Johnson (18) and those of Farrell (19),

the measured cross sections agree within Zt over the resonance. However, in the low energy part of the

15
resonance, the values of Harvey and Meadows have a tendency to be higher than those of Diment and Uttley.

Again, if one accepts the values of Johnson, the

agreement is rather good between the different mea- surements in the high energy range (0.5 to 10 MeV);

but there is a tendency for the Diment and Uttley

values to be lower than the others. A careful evaluation of the data should lead to an evaluated total cross section with an accuracy better than 3%; in

particular, one should reach an accuracy of TL at the

peak of the resonance.

Neutron Elastic Scattering

The experimental results available in the experimental data files correspond to references 27-35.

These data contain angular distributions, polarization measurements and integrated cross sections. The

integrated cross sections are plotted in Figure 5 with the Hale et al . R-matrix evaluation. Above about

0.5 MeV no severe discrepancies exist between the

different results. But the 250 keV resonance is not

well defined and it is difficult to evaluate the en- ergy of the maximum o(n,n) cross section. There are only 9 points between 200 and 500 keV in the recent

results of Knitter (35); 14 points between 100 and

500 keV in Lane's (28) data and only 3 points in

Villard's data (27); the values are generally given

with a large error bar in the resonance. At low

energy (from 1 to 100 keV), there is only one mea- surement by Asami and Moxon (32) giving a constant

o(n,n) cross section equal to (0. 752±0.043) b up to about 50 keV. This value is about 0.05 b higher than

the one obtained in the Hale et al . R-matrix analysis

(0.71 b at thermal energy). The thermal value obtained by Uttley et al . (1) by analysing the Diment and

Uttley total cross section is equal to 0.724 barn.

The Li-6(n,a)t Reaction

The experimental data available, including an- gular distribution and polarization measurements cor- respond to the references 36-69. Everyone knows that

the main feature of the data is the important discrepancies existing in the experimental results for all

energy ranges above about 100 keV. Between 1950 and

1960, nine measurements were performed with a discrepancy of about 25% at the peak of the 250 keV resonance; 7 measurements between 1960 and 1970 give

approximately the same discrepancy; 13 measurements

have been performed since 1970 and the inconsistency

is still 25% in this period of time. If we consider

all the measurements between 1950 and 1977, the difference between the lower and higher values in the

peak of the resonance is about 40% relative to the mean value. Apparently no improvement has been

obtained in 27 years, and the evident conclusion

drawn by the unbiased observer would be that we do

not know how to measure the Li-6(n,a) cross section

with high accuracy in the vicinity of the 250 keV resonance and at higher energies. Thus the question

to be answered is : Is it possible and reasonable to use the Li-6(n,a) cross section as a standard above

100 keV neutron energy? Perhaps at the end of this

session we will be able to answer this question.

The thermal absorption cross section has been

measured by Meadows (53) using the pulsed neutron

method. He obtained a value of (936±4) barn. From

the measurement of the total cross section of a Li2S0^ sample, using the BR2 chopper facility at Geel

(Belgium), Becker and Deruytter (51) obtained the

value of (944±19) barn for the absorption. In this

experiment the accuracy was limited by the difficulty

in obtaining the accurate value of the Li-6 content

of the sample. The value obtained by Uttley and

Diment by extrapolating the 1/v law is equal to (939±2)

barn. The value proposed by Hale et al. from the R-matrix evaluation is equal to 935.89 barn. One can see that there is an agreement within less than 1% in the

different values proposed for the (n,a) cross section

at thermal

.

The 1/v law established by Diment and Uttley

from the total cross section measurement has been

confirmed to be valid by Coates' et al . (59) measurement with an approximation of ±1% up to 10 keV.

Unfortunately, there are no other measurements at low

energy to confirm the Harwell results. However, the

Hale et al . evaluation is also in agreement with

Coates' results as shown in Figure 6.

Figure 7 illustrates the large discrepancies

existing at the 250 keV resonance for some selected

results. It is not necessary to comment on this

Fi gure . In Figure 8, we have shown the most recent

measurements and evaluations :

1) the new values of Fort et al . (69) which correspond to the earlier Helsinki values corrected by a factor of 1.117 due to the re-evaluation of the Li-6

content of the glasses (by transmission measurement

at the Saclay Linac (5) )

.

2) the values obtained by Lamaze et al . (65) from a measurement relative to (n,p) scattering;

3) the values obtained by Gayther et al . (67) from

a measurement relative to U-235(n,f) cross section

using the U-235(n,f) standard of the Sowerby evaluation;

4) the R-matrix evaluation by Hale et al .

;

5) the value calculated by Knitter (68) from a fit

of his measured total and scattering cross sections

using a single level Breit-Wigner formula.

The Lamaze data have been included in the Hale

et al . R-matrix analysis and it is worth mentioning

that the discrepancy between the experimental and

calculated values is still 3% at the peak of the

resonance, whereas the evaluation of Knitter is in

agreement with the values of Lamaze.

We have also plotted, in Figure 8, the data from

Poenitz (60) and Coates which are not too far from

the above results at the peak of the resonances. The

integral cross section between 0.100 and 0.500 MeV

corresponding to some sets of data in Figure 8 are given in the following Table :

Area in b-MeV

Lamaze 0.481

Gayther 77 0.492

Knitter 77 0.488

Hale 0.486

Poenitz 74 0.471

One can see that :

1) there is only a 0.4% difference between the area of the resonance calculated from the Knitter and

Hale et al . values (Knitter's area is 0.4% higher

than Hale's, whereas Knitter's values are 2.5% lower

at the peak of the resonance);

2) the area calculated from the Poenitz values is

only 3% smaller than the one calculated from Hale et

al. whereas the peak cross section is about 10%

smaller; perhaps there is an effect of resolution in

the results of Poenitz;

16
3) the resonance in Fort's results seems to be

wider than in the others. It is difficult to know if

this is due to an effect of resolution or to the

correction on the energy scale. However, the Fort

measurement is the only one which is purely absolute

and it is comforting to see that his final value

agrees, at the peak of the resonance, with the recent

measurements and the recent evaluations.

Figure 9 shows the dispersion of the results in

the energy range 0.5 - 5 MeV. A series of six different measurements corresponds to high values of the cross section and one measurement gives very low cross sections (about two times smaller). Between

these extremes, in the energy range 0.5 - 1 MeV, we

find in quite good agreement the values of Poenitz,

the recent values of Lamaze and Gayther and the Hale

et al . evaluation. It is also important to note that

Stephany et al. (63) have tried to resolve the discrepancy in the MeV energy range by an absolute cross section measurement at 964 keV. They found a value

of 0.356 b which falls into the high series of values

mentioned above. However, in the last ERDA Progress

Report (7.b) the authors recognize that this value

is 12 or 15% too high and they are now investigating

a new method for a more accurate absolute measurement

at 964 keV and at other available energies.

Conclusion

In this review we have not tried to examine or criticise the measurement methods and the possible

sources of error in view of eliminating some results

which could be doubtful. This was done by C.A. Uttley

et al . at the 1970 Argonne meeting for the measurements before 1970 and will probably be done at this

meeting by Knitter for the most recent measurements.

We have only tried to gather the maximum of experimental data in the energy range up to about 5 MeV; these data are available to any evaluator who wishes

to make a definitive point on the Li-6(n,a) cross section. However, from a simple examination of the

data, it is possible to make some recommendations for

the future :

1. There is a lack of measurements in some of the

energy ranges for some reactions :

(a) the total neutron cross section has not been

measured at low energy; a recommendation for such a measurement was also made by Uttley at the 1970 ANL meeting; the Harvey measurement has not fulfilled the

recommendation because the results in the low energy

range have not been published. So, the 1/v law is

still established only from the Harwell measurement;

(b) according to Oeruytter (7) more measurements need

to be done at thermal energy; although the few results

available agree within better than 1%, these results

cannot be trusted with more than 1% accuracy (problems

of the Li -6 content in the samples used);

c) a better definition of the resonance in the

elastic scattering needs to be obtained; that was

also a recommendation of Uttley at the 1970 ANL meeting;

only one measurement has been done, by Knitter et al.,

since the ANL meeting.

2. Concerning the (n,a) reaction, a lot of experiments have been performed since 1970, and we have already pointed out that the discrepancies are still

unacceptable. We should not recommend that more

experiments be performed unless the reason for the

discrepancies has been clearly established and the new experiments could be undertaken with a high degree of

confidence. The attempt of Stephany et al. to resolve

the discrepancy in the MeV region is an example of

what could be done : they realized that the cross section value at 964 keV announced at the last Washington

Conference is 12 to 15% too high due to the underestimation of some experimental effect (7.b) and

they are investigating a new method which will be

used only if the preliminary results are sufficiently

encouraging.

3. The problem of energy calibration should also be

examined carefully. According to the calculation of

Hale et al . and Knitter et al . there is a difference

of about 4 keV between the maximum values in the

total and the (n,oi) cross section for the resonance

near 240-250 keV. This difference, which is a function of the resonance parameters and the channel

radii, needs to be verified experimentally. On the

other hand, energy adjustment of the data for a point

per point comparison cannot be done precisely if one ignores how to perform the correction; this means

that each author should give the law of variation of

the error on energy versus the energy, if possible.

4. Nevertheless, we should not be too pessimistic

at the present time. The R-matrix evaluation done by

Hale et al. has shed a supplement of light on to the

Li-6(n,oi) cross section, by using the a-t scattering

data; the parameters of this reaction channel should

be the same as those of the exit channel of the

Li-6(n,oi)t reaction. However, Knitter has also shown

that the parameters obtained from a single level

Breit-Wigner analysis of the total and scattering

cross section give, for the Li-6(n,a)t cross section,

values which are in quite good agreement with the

evaluation of Hale et al . Consequently, it appears

that the interferences between the two 5/2" levels in

Li-7 are not very strong. It is comforting to note

the consistency which now exists between the evaluations and the very recent sets of experimental data

(Lamaze, Gayther, Knitter), including the re-evaluated

data of Fort et al . , and a recommendation should be

made to use those sets of evaluated and experimental

data as a basis for the definition of the Li-6(n,a)

standard cross section.

Acknowledgement

The authors wish to thank Miss Lynette Truran

for her help in correcting the English and the material

realisation of this document.

Mixed References

1. C.A. Uttley et al . , Neutron Standards and Normalization, p. 80, ANL, 21-23 October, 1970.

2. K.M. Diment et al . , AERE-PR/NP 16 (1969), p.

3

3. G.M. Hale et al.

(a) Nuclear Cross Sections & Technology, Washington,

March, 1975, p. 302;

(b) LA-6518-MS;

(c) private communication.

4. F. Ajzenberg-Selove et al . Nucl . Phys. A 227

(1974), 54.

5. H. Derrien, not published.

M. Moxon, private communication.

6. E. Fort et al . , to be published.

7. A.J. Deruytter, private communication.

7.b J.C. Engdahl et al . , BNL-NCS-2151 , p. 143

17
References for g-t Entrance Channel

8. P.W. Keaton, et al . , Phys. Rev. Let. 20 (1968) 1392

9. M. Ivanovich et al . , Nucl . Phys. A 110 (1968) 441

10. R.J. Spiger and T.A. Tombrello, Phys. Rev. 163

(1970) 964

11. L.S. Chuang, Nucl. Phys. A 174 (1971) 399

12. R.A. Harde Kopf et al . , LA-6188 (1977)

13. R.E. Brown et al . , BNL-NCS-21501 (1976)

14. R.E. Brown and Y.C. Tang, Phys. Rev. 176 (1968)

1235

15. L.M. Kuznetsova et al . , Sov. Jour. Nucl. Phys.

13 (1971) 394

16. V.G. Neudatchin et al . , Lettere al Nuovo Cimento

5 (1972) 834

17. F.C. Barker, Aust. Jour, of Phys. 25 (1972) 341

No. Author

References for Total cross section measurements

Reference^ No. of 1 ' Year Energy Range Machine

18 JOHNSON+ PR 96 985 54 33-4150 keV VDG 232

19 FARRELL+ 68 WASH 153 68 44-646 keV VDG 125

20 HIBDON+ 68 WASH 159 68 10-1360 keV VDG 704

21 FOSTER+ PR/C 3 576 71 2-15 MeV VDG 237

22 GOULDING+ GOULDING 72 72 0.7-30 MeV Linac 507

23 MEADOWS+ NSE 48 221 72 0.1-15 MeV VDG 987

24 UTTLEY+ UTTLEY 74 74 0.07-7000 keV Linac 3260

25 HARVEY+ 75 WASH 244 75 0.025-10 MeV Linac 185

26 KNITTER+ EUR-5726e 77 0.084-3 MeV VDG 233

• References for elastic scattering measurements

No. Author

I*.] Reference^ ' Year Energy Range No. of Integratf

27 VILLARD+ PR 101 765 56 210-300 keV Only do/de.

28 LANE+ AP 12 135 61 0.05-2.25 MeV 25

29 BATCHELOR+ NP 47 385 63 3.35-4.64 MeV 6

30 KNITTER+ EUR-3454e 67 1.0-2.3 MeV 14

31 HOPKINS+ NP/A 107 139 68 4.8-7.5 MeV

32 ASAMI+ 70HELS 153 70 1.0-110 keV 1

33 DEMANINS+ INFN-73 2 73 2.0-4.7 MeV 8

34 HOLT+ NP/A 237 111 75 2.0-5.0 MeV Analyzing power

35 KNITTER+ EUR-5726e 77 0.1-3.0 MeV 40

References for Li-6(n,a)t Measurements

(*) No. Reference^ ' Year Author Energy Range Comments No. of points

36 ANL-4515 50 BLAIR+ 142-624 keV relative to U-235(n,f) 13

37 PR 90 1049 53 DARLINGTON+ 200-600 keV only angular distribution 1

18
( *) No. Reference^ ' Year Author

References for Li-6(n^)t Measurements Cont/d

Energy Range Comments No. of points

38 PR 95 117

39 DOK 111 791

40 PR 103 741

41 PR 112 926

42 PR 114 1580

43 PR 114 201

44 PR 115 1707

45 ZN/A 15 200

46 NP 63 593

47 AF 29 45

48 66 WASH 763

49 ONE 21 271

50 ZFK-130 143

51 70 ANL 125

52 JNE 24 323

53 NSE 40 12

54 70 HELS 1 253

55 71 KNOX 2 611

56 AERE-7075

57 EANDC(E)-148

58 NP/A 221 573

59 COATES 74

60 ZP 268 359

61 INT 7011

62 BAP 20 163

63 75 WASH 236

64 75 WASH 240

65 LAMAZE 76

66 76 LOWELL 1340

54 WEDDELL+

56 GORLOV+

56 RIBE.

58 KERN+

59 BAME+

59 GABBARD+

59 MURRAY+

60 BORMAN

.

65 SCHWARZ+

65 CONDE+

66 BARRY.

67 COX+

67 RENDIC+

70 BECKER+

70 SOWERBY+

70 MEADOWS+

70 FORT+

71 PHERSON+

72 CLEMENT+

72 FORT+

74 OVERLEY+

74 COATES+

74 POENITZ.

75 BARTLE.

75 STEPHANY+

75 SHRODER+

76 LAMAZE+

76 RAMAN+

1.5 and 2.0 MeV

9.1-730 keV

0.88-6.52 MeV

12.6- 17.9 MeV

9.0-342 keV

0.025-4.07 MeV

1.20-7.93 MeV

2.5 and 14.1 MeV

2.9-588 keV

100 keV

25-100 keV

10.7- 102 keV

2.7 and 14.4 MeV

0.0253 eV

10 eV - 74 keV

0.0253 eV

82-517 keV

10-287 keV

0.16-3.9 MeV

0.021-1.7 MeV

0.1-1.8 MeV

1.04-327 keV

91 keV - 1.50 MeV

2.16-9.66 MeV

964 keV

25 keV

3.34-945 keV

0.5 eV - 25 keV

74 FRIESENHAHN+ 1.03 keV - 1.7 MeV

Nuc. emulsion, normalized 2 0.42b at 0.6 MeV

absolute 24

absolute 10

absolute 23

Relative to U5(n,f) 29

al so da/de

.

Absolute value determined 123 at 0.255 and 0.600 MeV

relative to U8(n,f) 11

relative to I-127(n,2n) 2

Relative to H(n,p) for E>150 keV

normalized 945 b at thermal

absolute 1

Normalized 950 b at 0.025 eV; 3 standard U5(n,f)=577b at 0.025 eV

absolute (absorption measur.) 7

Diff. sig. int. relative to 2 H(n,p)

absorption = (944±19) barns 1

relative to B-10(n,a) 86

absorption = (936±4) barns 1

absolute 40

Peak normalized at 2.8 b; 31 flux calibration with Pu-Be

Normalized between 0.3-0.5 MeV 68

on previous Uttley, Coates, Fort.

absolute 118

diff. sig. int. 25

Normalized at 149.5//E in the 159

range 1.5-10 keV

Normalized to absolute meas. 67

by the same author

Relative to {n,p) scatt.; 331

normalized 148.9//E in 3.5-4.5 keV

absolute 24

absolute 1

angular anisotropy

relative to (n,p) scattering 88

angular anisotropy

19
References for Li6(n,a)t Measurements Cont/d

No. Reference^ ' Year Author Energy Range Comments No. of points

67 GAYTHER 77 77 GAYTHER+ 3-809 keV Relative to U-235(n,f);

U5(n,f) from Sowerby evaluation

68 EUR-57262 77 KNITTER+ 85-500 keV calculated from o.,. and o (n,n)

69 FORT 77 FORT+ 82-517 keV Helsinki values re-evaluated

*\

' See the codes for references in CINDA 1^/11; the name followed by the year corresponds to

a private communication.

TABLE I

COMPARISON BETWEEN THE Li -6 EXPERIMENTAL TOTAL CROSS SECTIONS

The values given are the ratio to the DIMENT and UTTLEY values

ENERGY RANGE

(MeV)

JOHNSON

0RL,1954

FARREL

DKE,1968

HIBDON

ANL,1968

FOSTER

BNW,1971

MEADOWS

ANL,1972

GOULDING

RPI,1972

KNITTER

GEL, 1974

HARVEY

0RL,1975

0.06-0.08 1.018 1.00 1.022

0.08-0.10 1.047 1.018 1.000 1.059

0.10-0.12 0.968 1.003 1.065 0.974 1.000

0. 12-0.50 0.967 0.979 1.003 1.033 1.020 1.007

0.50-1.0 1.042 1.053 0.989

1.0-1.5 1.143 1.037 1.060 1.033

1.5-2.0 1.031 1.059 1.043

2.0-2.5 1.025 1.041 1.016

2.5-3.0 1.073 1.054 1.060 1.033

3.0-3.5 1.005 1.010 1.024 1.006

3.5-4.0 0.979 0.996 1.017 0.986

4.0-4.5 1.002 1.020 1.013

4.5-5.0 0.995 1.010 0.993

20
11.84

'He+p+n

9978

•He*p

It.isii

'Be+d-a

.1 5.0260

•Li+d-p

r

I Q.993Q

•Li+t -d

1

4668

I r -

98 -

He + t

11.25

10.25.

7 AS7

6 68.

4 633

047761

1-17.348 " *He+*He-p

J'=3/27 T.I/2

30

12.91

« ''Li + 2n

9.61

'He+d

Vi + n 7.2506

2.791 TI °B+n-a

0862

'Be

-II 2027

'Be + p-'He

Figure 1 Energy levels of the Li-7 System (From F. Ajzenberg-Selove (4) ).

21
22
iilT M |iii n rtii|ii MMi M |i Mniiii|tiiilitii|iitt nM i| n [ nn rr| n iii[rii| n riT M ii[iiT n ' M i n T n Tiirr|nnrmT|iiTmTTT|rTTn jr[^

(NMb9) bU3IS

23
t'3' TM003I

24
t-i <C I—

uj e: _i I—

_1 cC _1 •-" <: (/> 1-1 z IC < > i»i

o

_l to

1—

1

LUo LU

s: I— Q-o O CO _l

o

(sujeq) VW9IS

25
26
o CL q: I in

OLlf) > 3 t- _) IE cr a f_) o o

lO U CJ

LD U 2

— IE DL (TI D

O Z d

a U U

(T tn (D LiJ IiD — a

rrniTiTiTt|t n 'Tt m^TTTTTTTTTji I II Ti ti t[t t m il 1 1 [1 mn ri

I J ri m n 1 1 [i M ii mi] t irTriT n |T nnTTTTyTTrnn i n tit h i m t[

i w t

4 f

n » ' t

* **

ft

"' I'l ' I I I I I I IJ I LI I I ll I I II I III I I 111 I I I [ nl L] [ I 1 I LM I I LI I I I [ LI [I L i Ll I 111 111 llM m LI I ImI I I in [I ift ] I] I II ll

(NdbBI UU3IS

27
(NSbSl HU3IS

28

R-MATRIX ANALYSIS OF THE ''lI SYSTEM*

G. M. Hale

Los Alamos Scientific Laboratory, University of California

Theoretical Division

Los Alamos, New Mexico 87545

ABSTRACT

We describe a multichannel, multilevel, R-matrix analysis of reactions in the ^Li system which

was used to provide the ENDF/B-V neutron cross sections for ^Li at low energies. Resonance

parameters obtained from the R-matrix levels are presented. Various features of the data are interpreted in terms of these resonances.

/7 6 , ,

' Li System; Li(n,t); R-Matrix; resonance parameters; standard)

Introduction

Discrepant measurements is a problem one almost

always faces in the evaluation of neutron cross sections, but in the case of a standard cross section,

the problem intensifies the difficulty of obtaining

evaluated cross-section values of the desired accuracy.

However, one knows from general considerations of nu- clear reaction theory that the standard cross section

is linked by unitarity to data for other reactions in

the same compound system, and thus one has the prospect of reducing the uncertainty of the evaluated

standard cross section by analyzing it simultaneously

with these other data in a unitary framework. R-matrix

theory^ is a particularly appropriate unitary framework for such analyses, as was discussed in Ref. 2.

We have used multichannel, multilevel R-matrix

analyses of reactions in the Li system to provide

evaluated cross sections for the neutron- induced reactions on ^Li at low energies for versions IV and V of

the Evaluated Nuclear Data File (ENDF/B) . At the time

of the version IV analysis,^ the data set dominating

the fit was the neutron total cross section measurement from Harwell,'* both because of its quoted precision and number of points. The resulting ^Li(n,t)

cross section (the "standard" in this system) was not

much different from that obtained in Version III, which

had been largely based on the Harwell measurement.

The value of the calculated ^Li(n,t) cross section

at the peak of the 240 keV resonance was 3.5 b, some

16% higher than the value indicated by a contemporary

group of direct measurements.^ ^ However, we noticed,

as had the Harwell group, that unitary constraints

forced the total cross section unacceptably high above

Diment's measurements in the peak when the (n,t) cross section was lowered to better agree with the direct

measurements. It is emphasized that this discrepancy

between the cross section measurements was indicated by

attempts to fit them unitarily. The common practice of

evaluating the total and (n,t) cross sections separately,

then obtaining the elastic cross section by subtraction,

would not necessarily have signaled a problem.

ence on Nuclear Cross Sections and Technology^ two years

ago, indicated the effect of the new data was to lower

the calculated (n,t) cross section in the peak of the resonance (to 3.4 b) and raise the calculated peak total

cross section (to 11.1 b). This trend has continued in

subsequent analyses, including that used recently to provide ENDF/B-V neutron cross sections for Li at en- ergies below 2 MeV.

The Version V analysis will be described briefly in

the following section, indicating the types of data included, and the fits obtained to representative data

sets from each reaction. The third section presents resonance parameters corresponding to R-matrix levels

required for the fit, and the concluding section dis- cusses the interpretation of specific features in the

data for reactions in the 'lI system in terms of these resonances

.

Version V Analysis

The 3 arrangement channels considered in this analysis were t + "He, n + ^Li, and n + ^Li* (2.18). Various quantities specifying the channel configuration are given in Table I.

TABLE I

Channel Configuration for Li Analysis

Arrange- Channel

I Channel

ment Radius max Spins (s)

t + "He 4.02 fm 5 1/2 n + ^Li 4.20 fm 1 3/2, 1/2 n + ^Li* 4.50 fm 1 7/2, 5/2

Data for all possible reactions were considered at triton energies below 14 MeV, and at neutron energies below 2 MeV. Although specific references to all the data

included will not be given, the different types of data

analyzed for each reaction are listed in Table II.

We noticed in performing the Version IV analysis

that in addition to this strong unitary link to the total cross section in the region of the resonance, the

(n,t) cross section was also quite sensitive to the

differential cross section for t + a elastic scattering.

At that time, however, measurements of this cross section^ had not been made with accuracy comparable to that

of the Harwell total cross section measurements. Precise measurements of the t + a differential cross section^ were later made at Los Alamos, and these were incorporated in the analysis, along with new measurements

of the total cross section from Oak Ridge. Preliminary results of this analysis, reported at the ConferTABLE II

Types of Data Included in ''lI Analysis

Reaction

Total

Cross

Section

"He(t,t)"He

^Li(n,t)'*He

^Li(n,n)^Li

a + t n + ^Li

InteReaction grated

Cross

Section

Cross

Section

X

X

Differential

Cross

Section

X

X

X

Polarization

X

X

*Work performed under the auspices of the United States Energy Research and Development Administration.

30
'^He(t,t)^He 0(6)

% = 150-

1 Ivan (1967), Ja (1976)

L '— 1- 1 JI * 9 a 7 a g 10 II i; L

LAB fINETIC ENERGY IN MEV

4 He (t-

'a,

= 49.6°

i ttardekopf (1976 )

CENTER OF MASS ANGLE

He(t,t)\e AO)

= 8.785 MeV

$ Hardekopf (1976) P

s

/

S .y

L*e KINETIC ENERGY IN MEV CENTER OF MASS ANCLE

4 4 Fig. 1. Calculated and measured observables for He(t,t) He, The cross section

data are from Refs. 9 and 12; the analyzing power data are from Ref . 13.

The data were fitted in the usual least-squares sense

by the R-matrix analysis code EDA.'^ Renormalizations

and energy shifts were allowed for some of the data sets;

in these cases, the deviations from the original experimental scales contributed to the overall '2 of the fit.

The resulting fits to some of the t + a elastic

scattering measurements are shown in Fig. 1. Anomalies

occurring in the differential cross section and analyzing power excitation curves (left side of the figure)

at triton energies of approximately 3.9, 7.3, 8.8^ and

12.4 MeV correspond to the first 4 resonances in Li

above the t + a threshold. Parameters for these reso- nances, which have total angular momentum and parity

P _ _ _ _

(J ) assignments of 7/2 , 5/2 , 5/2 , and 7/2 , respectively, are given in the next section. The third resop nance is the J =5/2 level which shows up prominantly

in the neutron cross sections at E = 250 keV. Angular

distributions of the cross section^and analyzing power

are shown on the right side of the figure at triton en- ergies close to this resonance. The low-energy measurements of the cross-section excitation are those of Ivanovich,^^ while those at overlapping energies and higher

are of Jarmie.^ The calculated curve follows the newer, more precise Jarmie data in the region of the overlap.

The differential cross section data shown are also those

The pronounced asymmetry in the

of Jarmie,^ '^^ while the analyzing power excitation and

angular distributions were measured by Hardekopf.'^

Figure 2 compares the R-matrix calculation with

selected data for the ^Li(n,t)'*He reaction. On the

left are shown recent measurements^'* of the ^Li(n,t)

cross section over the 5/2 resonance at center-of-mass

angles of 0 and 180°. These measurements became available after the Version V analysis was completed, so that

the curves represent a prediction, rather than a fit.

The right side of the figure shows the fits to Overley's

E = .1 and 1.0 MeV.

ncross section evident at 100 keV persists down to energies as low as 25 keV.^^ These low-energy asymmetry

effects in the cross section are well reproduced by the

calculations, and can be explained in terms of resonance

interference, as discussed in the concluding section.

Representative fits to the ^Li(n,n)^Li angular distributions measured by Lane^^ and Knitter'^ are shown

in Fig. 3. The Lane angular distributions required

substantial renormalizations at energies near the peak

of the 240 keV resonance, with the result that our calculated integrated elastic cross section lies somewhat

above Lane's points over the peak. This agrees with

recent results of Knitter^' for the elastic cross sec31
A

f\ ^Li(n,t)^He 0(6) _

^CM =

0° trY

i Brown (1976) MY-1-

\ 1

\

^Li(n,t)^He 0(9)

= .100 MeV

I Overley (1974)

UAB KINETIC ENERGY IN MEV CENTER OF MASS ANOLE 4

1.

\

Li(n,t) He 0(9)

\ i Brown (1976)

1 —4

LAB KINETIC ENERGY IN MEV CENTER OF MASS ANCLE

6 4 Fig. 2. Calculated and measured observables for L±(n,t) He. Data for the excitation

curves are from Ref. 14. Data for the angular distributions are from Ref . 15.

tion over the resonance. Calculated values of the polarization (not shown) for elastically scattered neutrons

2 0 also agree well with measurements by Lane.

Figure 4 shows the fits to the total cross sections

of primary interest in this discussion, the neutron total

and ^Li(n,t) integrated cross section. Of course, the

latter cross section is the standard and is, at that,

only recommended for use at energies below the 240 keV

resonance. But because of the strong influence of the resonance on the ^Li(n,t) cross section as it begins to deviate from 1/v behavior (at about 20 keV) , and because

of the close unitary connection (already mentioned) , it

has with the neutron total cross section, an isolated

discussion of the standard cross section would not

suffice

.

We see that the calculated ^Li(n,t) cross section

is generally in very good agreement with recent measurements made at the National Bureau of Standards^ ^ (NBS) up to ~600 keV. In the region of the resonance, the calculated cross section has come down by roughly half the

difference between the Version IV results and the earlier

direct measurements,^ ^ and still exceeds the NBS data

by about 2.5% in the peak of the resonance. The data

shown are those that were used in the analysis. Corrections to the data since that time have raised the experimental values in the minimum around 90 keV to agree

even better with the calculation, but have reduced some- what the experimental values in the peak. The largest

uncertainty of the calculated cross section in the

standards region (below 150 keV) occurs in the vicinity

of the 90 keV minimum. At lower energies, the thermal

value of 935.9 b is in excellent agreement with Meadows'

value, and the cross section is quite consistent with

Sowerby's ratio measurements^^ below 80 keV. when con- sidered in conjunction with the Version V ^ B(n,a) cross section.

The calculated total cross section agrees well with

measurements of Diment'' and Harvey,'" except for a syste

matic tendency to overshoot the experimental values in

the peak of the resonance and undershoot them in the

preceeding minimum. These differences are of the order

of 2-6%, and may be within the bounds of realistic un- certainties on the total cross sections in these regions

if not within the stated errors. The energy scale of

these total cross section measurements'*' ° essentially

determined the position of the 5/2 resonance, with the

calculated peak total cross section occurring at 245

keV, and the calculated peak (n,t) cross section at 240

keV, in agreement with most of the measurements using

time-of-flight neutron energy determination.

32
^Li(n,n)^Li 0(6)

E = .118 MeV

ni Lane (1960)

^Li(n,n)^Li 0(6)

E = .530 MeV

n

i Lane (1960)

CENTER OF MASS ANCLE CENTER OF MASS ANCLE

^L±(n,n)^Ll a (6) E = .2A0 MeV

n

I Lane (1960)

i

1 1

^Li(n,n)^Li a(6)

E =1.0 MeV

ni Knitter (1967)

*- —

1

CENTEH OF MASS ANCLE CENTER OE MASS ANCLE

6 6 Fig. 3. Calculated and measured angular distributions for Li(n,n) Li. are from Refs. 17 and 18.

The data

Resonance Parameters

Since experimental data for nuclear reactions are most directly related to matrices (such as the S-matrix)

of amplitude ratios for asymptotic wavefunctions , it is

appropriate that resonance parameters for asymptotic

measurements be related to the poles and residues of

such matrices. The R-matrix is not, of course, an asymptotic quantity, and its poles and residues depend

on the boundary conditions imposed at the nuclear surface to define the R-matrix, and upon the channel radii

which define the nuclear surface. However, one can always transform the R-matrlx to an asymptotic form

having poles and residues which, although energydependent, no longer depend on boundary conditions at the nuclear surface, or upon radial distances outside

that surface.^**

where the e and E are known, complex (energy-dependent)

y V functions of the parameters Yj^'^^ °^ the original R-matrix. If, for some complex energy ,

E (E ) - E

0'^ 0

0

then E is a pole of R and thus, of the S-matrix as

0 L well. The resonant energy ?.nd total width associated

with the pole are

-2 lm(E^),

We choose the asymptotic form to be the S-matrix

so that our specification of resonance parameters cor- responds with the prescription given by Humblet. The

transformed R-matrix in that case has the form

T

E -E

y

while the partial widths are given by

^yc^^O^I -P (EJ

c 0

=yc(^o)

P (EJ

c 0

33
NEUTRON ENERGY. MEV

6 4 Fig. 4. The Li(n,t).He integrated cross section

and n + Li total cross section for E ben tween .01 and 2 MeV. The solid curves are

the ENDF/B-V results, and the dashed curves

are ENDF/B-IV. Data for the upper curve

are from Refs. 5, 21, and 23; data for the

lower curves are from Refs. 4 and 10.

with

Re(P ) P (E ) = ^

where p is the product of the channel radius with the

channel wave number, and 0 is the channel outgoingspherical-wave function, all evaluated at the complex

energy E^.

Resonance parameters obtained in this way from the

Version V R-matrix parameters are presented in Table

III, along with widths and resonance positions from F. Selove's latest compilation for ^Li.^^ The first 4 Resonances listed are those visible in the excitation

curves shown in Fig. 1. The agreement of our calculated parameters for those resonances with values taken

from the compilation^^ is satisfactory, although not

always within the errors assigned by Selove.

The last two resonances lie above the highest energy at which experimental data were included, and their

parameters are therefore quite uncertain. They do cor- respond roughly with the levels found by Holt et al.^^

in their analysis of n + Li elastic polarizations at neutron energies between 2 and 5 MeV, except that our calculated widths are somewhat narrower.

It should be mentioned in this connection that the

prescription we use can result in resonance parameters

that are quite different from those obtained from the

usual expressions^* relating R-matrix parameters to external widths and resonance positions at real energies. This is particularly true at higher excitation

energies, where the complex pole prescription tends to give smaller widths which appear to correspond more

closely with the widths of the experimentally observed

anomolies. We also note that the complex poles and

partial widths defined above are formally radius-independent in the external region, while those resulting

from the usual R-matrix relations are not.

Conclusions

This analysis indicates that many of the features

observed in the measurements for reactions in the ^Ll

system can be interpreted in terms of the resonances

identified. In addition to the obvious structure in

the cross sections and polarizations due to the presence of the first 4 levels above the t + a threshold,

one can ascribe the behavior of the low-energy ^Li(n,t) cross section to broader, more distant states. NearP +

background levels having J = 1/2 (one of which may be

associated with the broad structure at E =16.8 MeV in

7 +

^ Li), and the broad 3/2 resonance tentatively identified in this analysis seem to be mainly responsible

for the large 1/v integrated cross sections at low

energies. Interference terms, arising from the presence

of the prominant 5/2 resonance at 7.46 MeV excitation

energy and of the 3/2 level higher up, cause most of

the asymmetry in the differential cross sections at low

, +

energies, with interference between the 1/2 and nega2 tive parity levels having non-zero P^ widths contributing to a lesser extent. More recent extensions of

this analysis to somewhat higher energies^^ indicate

that the 3/2 level is responsible for the broad structure seen in both the neutron elastic and total cross sections at E ~ 3.5 MeV, and at the same time, produces

n

a "shoulder" in the ^Li(n,t) integrated cross section

at E ^2.2 MeV.

nRecent measurements for reactions in this system

appear to be approaching unitary consistency. At this

stage, however, the precise experimental determination

of the total cross section below and over the resonance

seems to be most elusive, both in terms of the magnitude

of the cross section near the peak, and the energy of

the peak. New measurements of the total cross section

by Knitter^' and by Smith^" agree well in magnitude with

the Version V calculations, but not in energy scale.

Questions about the ^Li(n,t) cross section persist in

relation to ratio measurements. While Gayther's^^ new

ratio measurement of ^^^U(n,f) relative to ^Li(n,t)

appears quite consistent with the Version V results,

Macklin's'^ ratio of ^''Au(n,Y) relative to ^Li(n,t)

does not. The resolution of these differences will

doubtless require further (but hopefully, small) changes

in the neutron cross sections for Li. We feel, however,

that the utility of this approach, and the importance

of analyzing standard cross sections in a unitarily

consistent way with other data from the same system,

have been demonstrated by the Version V results.

34
TABLE III

7/2"

5/2"

5/2"

7/2"

3/2'

3/2"*

Resonance Parameters for the Li Systefa

_ b

4.672

6.596

7.455

9.568

11.279

4. 633+. 008

6. 675+. 054

7. 467+. 004

9. 61+. 81

9.853 10.25+.1

State Partial Width^ Total Wldth^ Total Width'

^

.093+. 008

.+ .2

t .071 .071

2f

t .944 .945

n

.001

\ .023 .077

n

.054

t .294 .412

n*

.118

2p

t

.182^^ 1.171"^

n

.969^^

2pn

.020'^ \

1.360'^ 2.517*^

1.157*^

.875 -.1

.089+. 007

1.4+.1

a. Values (in MeV) derived from the Version V R-matrix parameters.

b. Values (in MeV) tabulated in Ref . 26.

c. Values given for resonances above the range of the analysis,

which are therefore quite uncertain.

Acknowledgments

This analysis is an outgrowth of work begun by D.

C. Dodder and K. Witte. The effort still benefits from

collaboration with them on many aspects of the problem.

I am grateful to N. Jarmie, R. Brown, L. Stewart, and

p. Young for their help in collecting and reviewing the

experimental data.

References

1. E. P. Wigner and L. Eisenbud, Phys . Rev. 72^, 29

(1947), and A. M. Lane and R. G. Thomas, Rev. Mod.

Phys. 30, 257 (1958).

2. G. M. Hale, "R-Matrix Analysis of the Light Element

Standards," Proceedings of a Conference on Nuclear

Cross Sections and Technology, Vol. 1, 302 (1975).

3. G. M. Hale, L. Stewart, and P. G. Young, Los Alamos Scientific Laboratory report LA-6518-MS (1976).

4. K. M. Diment and C. A. Uttley, Harwell report AEREPR/NP 15, p 12 (1969).

5. W. P. Poenitz, Z. Phys. 268, 359 (1974).

6. W. Fort and J. P. Marquette, "Experimental Methods

Used at Cadarache to Determine the ^Li(n,a)T Cross

Section between 200 keV and 1700 keV," Proceedings

of a Panel on Neutron Standard Reference Data, November 20-24 (1972), IAEA, Vienna. Note: Fort

expects to renormalize his data upward by ~11%.

7. M. S. Coates, G. J. Hunt, and C. A. Uttley, "Measurements of the Relative 6Li(n,a) Cross Sections in

the Energy Range 1 keV to 7500 keV," Neutron Standards Reference Data, IAEA, Vienna, p. 105 (1974).

8. R. J. Spiger and T. A. Tombrello, Phys. Rev. 163 , 964 (1970).

9. N. Jarmie et al . , Bull. Am. Phys. Soc. 20, 596 (1975).

10. J. A. Harvey and N. W. Hill, Proc. Conf . on Nuclear

Cross Sections and Technology, Vol. I, 244 (1975).

11. D. C. Dodder, K. Witte, and G. M. Hale, "The LASL

Energy-Dependent Analysis Code EDA," unpublished.

12. M. Ivanovich, P. G. Young, and G. G. Ohlsen, Nucl.

Phys. Alio , 441 (1968) .

13. R. A. Hardekopf et al . , Los Alamos Scientific Laboratory report LA-6188 (1977).

14. R. E. Brown, G. G. Ohlen, R. F. Haglund, Jr., and N. Jarmie, LA-UR-77-407 (1977) (Submitted to Phys. Rev.

C).

15. J. C. Overley, R. M. Sealock, and D. H. Ehlers, Nucl.

Phys. A221 , 573 (1974) .

16. I. G. Schroder, E. D. McGarry, G. de Leeuw-Gierts

,

and S. de Leeuw, Proc. Conf. on Nuclear Cross Sections and Technology, Vol. I, 240 (1975).

35
IT. R. 0. Lane, Ann. Phys. 12, 135(l96l).

18. H. H. Knitter and A. M. Coppola, EANDC report (E)

5T(U) (1967).

19. H. H. Knitter, C. Budtz-Jorgensen, M. Mailly, and

R. Vogt, CBNM-VG (1976).

20. R. 0. Lane, A. J. Elwyn, and A. Langsdorf, Jr.,

Phys. Rev. 136_, B 1710 (I96I+).

21. G. P. Lamaze, 0. A. Wasson, R. A. Schrack, and

A. D. Carlson, Proc . International Conference on the Interactions of Neutrons with Nuclei, Vol. 2,

13hl (1976).

22. J. W. Meadows, Neutron Standards and Flux Normalization (AEC23), 129 (1971).

23. M. G. Sowerhy, B. H. Patrick, C. A. Uttley, and

K. M. Diment, J. Nucl. Energy 2h_, 323 (1970).

2h. G. M. Hale, Nuclear Theory in Neutron Nuclear

Data Evaluation, Vol. II (IAEA-I90) 1 (1976).

25. J. Humblet and L. Rosenfeld, Nucl. Phys. 26, 529

(1961).

26. F. Selove and T. Lauritsen, Nucl. Phys. A227, 5h

(197»*).

27. R. J. Holt, F. W. K. Firk, G. T. Hickey, and

R. Nath, Nucl. Phys. A237 , 111 (1975).

28. See, for instance, Lane and Thomas (Ref. l), p. 295.

29. G. M. Hale and D. C. Dodder, Proc. International

Conf. on the Interactions of Neutrons with Nuclei,

Vol. 2, 1U59 (1976).

30. A. B. Smith, P. Guenther , D. Havel, and

J. F. Whalen, ANL/NDM-29 (1977).

31. D. B. Gayther, AERE-R 8556 (1977).

32. R. L. Macklin, J. P. Halperin, and R. R. Winters,

Phys. Rev. Cll , 1270 (1975).

36
SPECIAL PROBLEMS WITH GLASSES

G. P. Lamaze

National Bureau of Standards

Washington, D.C. 20234

Properties of Ce activated Li loaded glass scintillators are discussed as well as their

applications as neutron detectors. Three special problems that may arise in their use

for neutron detection are non-uniformity of the ^Li content, multiple scattering, and

after pulsing of the photomultiplier . These problems and their consequences are discussed

as well as some possible solutions.

(glass scintillators; ^Li(n,a)T; Monte Carlo; multiple scattering; neutron detection;

photomultipliers)

Introduction

Before I discuss the problems with the use of ^Li

glass, let me point out some of its advantages. As we

have seen in the earlier talks, the ^Li(n,a)T reaction

has a very large cross section with a value of '^940 b at thermal and a very nice 1/V behavior up to about

10 keV. Figure 1 shows a result of a recent NBS measure

ment.2 The straight line is the ENDF/B-V evaluation

that Gerry Hale talked about previously. ^ One can see

that at 10 keV, the deviation from 1/V is less than 1%

and at 30 keV it is already about 6% and rising rapidly.

Figure 2 compares the NBS results with ENDF/B-V in the

region of 2 to 800 keV. Not only does the ^Li(n,a)

reaction have a large cross section ("^^3. 15b at the peak)

but it also has a very high Q-value ('^^4.79 MeV) . This

combination of features has led many investigators

to search for a fast scintillator to take advantage of

these two attractions.

Early investigations^ centered on ^Lil(Eu), but the

Iodine introduces a great deal of structure in the

response function and ^I,i.J(Eu) is also reported to undergo changes in its scintillation properties as it ages.

In 1958, Ginther and Schulman** reported on Cerium activated glass scintillators. In that same year,

Voitovetskii , Tolmacheva, and Arsaev^ reported the first

'^Ll loaded, Ce activated glass scintillators suitable

for neutron detection. Much work has been done in the

- ensuing years and a wide variety of glasses are now

commercially available.

In fact, to get in the neutron detection business

today is pretty simple; all you need is a photomultiplier,

a piece of ^Li glass and a can of beer. Figure 3 shows

a schematic of the mounting scheme used at NBS with a "typical" pulse height spectrum. The glass is mounted

into an aluminum 7 oz . beer can and viewed edge on by

a photomultiplier about 2 cm away. No light pipe or optical coupling compound is used in order to minimize

the scattering material in the beam. The simplicity

and low cost of this apparatus makes it a very

attractive neutron detector.

37
CO

9

8 cr

7 o

6 ^

2

S FR ES I lNT

ENDF B/\'

2 10 1 2 5 10' 2 5 8 .00

ENERGY. KEV

Fig. 2. Results of a recent NBS measurement of the ^Li(n,a)T cross section compared with ENDF-B/V.

0.05 X 5 X 5 cm

Li^ GLASS

0 20 40

CHANNEL NUMBER

Fig. 3. Method employed at NBS to couple ^Li glass to phototube shown with typical pulse height response to Linac neutrons.

38
I will now describe an interesting series of

measurements by Spowart^'^ in Scotland on the properties

of ^Li glass. Spowart presents an excellent review of

the physics of Lithium glasses and is certainly worthwhile reading for those who might be interested in

developing better glass scintillators. What Spowart has

done is to take the full range of commercially available

glasses and measure the response of the glasses as a function of Li content. Figure 4 shows the pulse height

vs Li content for thermal neutron capture. It is interesting that there is a minimum light output for about

7.5% Li content. The response of glass scintillators

to monoenergetic electrons is shown in Figure 5. This

too shows a minimum light output for 7.5% Li content.

It is not clear that this effect is due to the change

in Li content or perhaps to a difference in the Ce

content of the different glasses. Although the Ce

content of NE907, NE908, and NE912 is nominally the same.

525

500

475

450

425

400

325

300

275

250

225

200

\

i/901

3 4 5 b 7 8 9 10 11 12

WEIGHT I OF LITHIUM IN KELT

Fig. 4. Pulse height vs Lithium content using thermal

neutron beam activation.

Spowart has not measured them accurately as he did with

the Li content. As will be shown later, the glasses

suffer from non-uniformity problems.

Figure 6 gives a measure of the resolution function

obtained by Spowart. As expected, the glasses with the

lowest pulse height also have the poorest resolution.

This indicates that the NE912 and NE908 glasses are producing fewer photons per unit energy deposited than

the other glasses.

Spowart has also investigated the effect of temperature on the pulse height and resolution of Li loaded

glass scintillators. Figure 7 shows a plot of the

variation of the ^Li(n,a) peak pulse height as a function

of temperature. As can be seen, the light output of

the scintillator increases as a function of temperature

from 100 to 450 K. Near room temperature a difference

of 50 K changes the pulse height by about 10%. The

mechanism of this increase in light output is apparently

an increase in the phonon population with teipperature

which assists the energy transfer to the Ce^

» SOSb

PUISE HEIGHT

Fig. 6. Analysis of scintillator performance, activation

by thermal neutrons from Am/Be moderated source.

I 97i ktV PEAK

•912b

48? key PEAK

01134S6JB9I0 111I

WEIGHT I OF LITHIUM IS "lELT

Fig. 5. Peak pulse height vs Li content using ^'^^Bi EC beta activation.

200 2S0

SAMPLE TEMPERATURE (DEGREES KELVIN)

Fig. 7. Variation of ^Li(n,a) peak pulse height with

temperature for medium lithium glass.

39
luminescent centers. This increased efficiency of

energy transfer also improves the resolution as shown

in Figure 8. For most applications, the experimenter

won't bother controlling the temperature; however, one should be aware of this characteristic and take proper

precautions against large temperature variations during

an experiment. This also presents the experimenter with

the interesting option of cooling the photomultiplier

while heating the Li glass to maximize the system

performance.

1.0 1.1 1.1.

(TULSt HEICKT)

Fig. 8. The effect of temperature on pulse height

statistics

.

The next major problem 1 would like to discuss is

the non-uniformity of the Li glasses. This can be a serious problem for many measurements, especially

certain types of Van de Graaff measurements where the

neutron flux might not be uniform over the face of the

scintillator. Figure 9 shows the Li content of a 9 mm

Harwell glass as measured by Moxon.*^ The content was obtained by measuring the transmission of the glass as a function of energy and position. By fitting the

results to a formula of Orj, = a + b//E, the 1/V and

constant components of the cross section could be

determined. The 1/V component (i.e., ^Li) of the

glass seems to be very uniform except at the edges

where some depletion of the Li has occurred. The con- stant component on the other hand does seem to have

some variation in density across the diameter. Moxon

concludes that this glass as well as a 9 mm thick

Cadarache glass were very uniform in Li content, except

at the edges. He speculates that the Li might be

leached out in the cutting and the polishing of

individual pieces. Tests performed by the Analytical

Chemistry Section at Harwell indicate that Li could

easily be leached out if the glass were to come into

contact with slightly acidic water solutions.

Figure 10 shows the result of transmission measurements on a 0.5 mm thick piece of NE 912 using the NBS

Reactor thermal column. As you traverse the glass from

one edge to the other, a 4% variation in ^Li content is

observed

.

GLASS SCINTILLATOR

• 'V COMPONENT

• CONSTANT COMPONENT

--- POSITION AND MAGNITUDE OF CROSS

SECTION COMPONENT GIVEN BY FITTED SHAPE

FITTED SHAPE

_L_

20 30 40 50 60 70 POSITION I mm I (NOTE CHANGE OF SCALE AT 2 0 AND 60) Fig. 9. A least square fit to one of the spatial

distribution measurements for the AERE glass.

Figure 11 shows the transmission of the same glass with

a traversal made at 90° to the first traversal. In this

direction the glass seems to be fairly uniform. The

non-uniformity of a particular glass sample raises

several experimental problems. First, the ^Li content

cannot be determined by chemical analysis of a part of

the glass, or by transmission of part of the glass. In fact, the absolute determination of the ^Li content may

be impossible to obtain with any precision. Even in

relative measurements, the experimenter must be careful

to always illuminate the same portion of the glass and

take care that the neutron flux is uniform over the used

portion.

>

UJI.OO

T T T T T

TRANSMISSION OF Li GLASS RELATIVE

TO TRANSMISSION AT CENTER

-2.0 -1,5 -1.0 -0.5 0.0 +0.5 •1.0 1,5 +2.0 DISTANCE , cm

Fig. 10. Transmission of a 0,5 mm piece of NE912 as measured with thermal neutrons.

<

> I.CO

TRANSMISSION OF Li GLASS RELATIVE

TO TRANSMISSION AT CENTER

i-i+K-|-!-|---i-|-i-i-,-r

J \ \ L

-1,0 -0,5 0,0 +0,5 DISTANCE , cm

J L

M.O + 1,5 +2.0

Fig. 11. Transmission of the same glass as Figure 10

but with a traversal perpendicular to the first traversal .

Let me now turn to the one unavoidable problem

found in any experiment with ^Li glass: multiple

scattering. NE912 which has one of the highest percentages of ^Li contains 21% silicon, 55% oxygen,

22.5% ^Li, 1% ^Li, and 0.5% cerium. Scattering reso- nances in these elements can cause significant errors

in the measure of the neutron flux. Figure 12 shows

the multiple scattering correction factor for a 5.08 cm diameter and 0 . 5 mm thick NE912 scintillator. With the

exception of a very narrow Si resonance at 55 keV, the

correction is very flat from 2 keV to 250 keV. At that

point, the influence of the ^Li resonance is seen. At

440 keV, the effect of the oxygen resonance is very

dramatic and requires a rather large correction for even a very thin glass. As the glass thickness is

increased, the corrections become even larger, as seen

in Figure 13. The correction is shown for three representative energies. For an 8 iran thick glass, a 15%

40
correction must be made at 250 keV and a '\45% correction

must be made at 440 keV. The error bars shown are

only the statistical errors for the MonteCarlo calculation. There are obvious systematic errors involved,

such as the actual composition of the glass and the

accuracies of the scattering cross sections and angular

distributions used in the program. Clearly, high precision measurements in the 400 to 500 keV region with

^Li glass detectors are almost impossible due to the

multiple scattering problem.

.92

"T I I I ! I

I

I

I "T I I

I I I I I 1— "T I Mill

MULTIPLE SCATTERING

CORRECTION FACTOR, PER INCIDENT NEUTRON

J i I I I I I I I

I I

1

I I

' ^

1 10 100 NEUTRON ENERGY, keV

I Mill

Fig. 12. Multiple scattering correction factor for a 5.08 cm diameter by 0.05 cm thick piece of NE912.

S 7

\

'

\

' rMULTIPLE SCATTERING CORRECTION

vs 'Li glass THICKNESS

• 10 keV

o 250 keV A 440 keV

5 1.0 2.0 4.0 bZ

THICKNESS , mm

Fig. 13. Variation of the multiple scattering correction

factor as a function of glass thickness.

Another problem in the use of Li glass in Linac

T-o-F experiments is the yflash. The glasses are very

sensitive to yradiation and this can lead to total

saturation of the photo multiplier and amplifiers as well as after pulsing of the photomultiplier . The

large electron pulse in the photomultiplier ionizes

residual gases in the tube. These ions are accelerated

back to the photocathode and will excite further bursts

of electrons. These signals come from 0.4 y9 to 60 pS

after the Y~flaeh and can look like valid events.

At NBS we solve this problem by offgating the photomultiplier during the yflasb.^ This has the double

feature of keeping the photo multiplier from saturating

as well as preventing the after pulsing. The after

pulse suppression works by pulsing a fine mesh grid

that is placed over the front face of the photomultiplier

tube. During the Y~flash a +300 volt pulse is placed on

the grid, suppressing the photoelectrons and preventing

them from striking the first dynode. Figure 14 shows

the after pulses after a large light source hits the

photocathode. The upper scope trace shows two after

pulse groups at about 0.6 and 1.0 yS after the light

flash. The lower trace shows the effect of off gating;

the light flash has been reduced by a factor of about

15 and the after pulses have been completely eliminated.

Figure 15 shows the after pulsing at long times. The

after pulsing peaks at about 40yS after the flash and

gradually dies away so that no after pulsing is observed

after about 100 yS. The after pulse suppression system

completely eliminates this long term component also.

Although this system can solve the problem of the after

pulsing of the photomultiplier, the glass scintillator

itself seems to after pulse after a strong Y~flash.

Additionally the experimenter is faced with the y-decay

of the neutron producing targets. These effects make

neutron detection at short times after the y-flash very

difficult and seems to indicate that two parameter data

taking (i.e., time-of-f light and pulse height) may be

necessary to separate y-rays from neutron events.

500

mV/div

100

mV/div

LIGHT

FLASH

AFTER

PULSES

200 ns/div

Fig. 14. Observed phototube output. Top trace is with

no after-pulse suppression; bottom trace is with 300

volt pulse applied to wire mesh.

LIGHT

FLASH

AFTER

PULSES

Fig. 15. Observed phototube output for long times

(20 yS/div). When gating is applied, these pulses

are completely suppressed.

41
To summarize, the work of Spowart demonstrates

that both improved scintillation properties and increased efficiency of ^Li glass scintillators is possible. Above 200 keV, uncertainties in the ^Li (n,a)T cross section and multiple scattering corrections

reduce the desirability of ^Li glass as a flux monitor.

Below 200 keV, the cross section is relatively well

known, Y~flash and after pulsing is not a problem

(except at short flight paths), and the multiple

scattering corrections are smooth and relatively small.

If nonuniformity of the scintillator is not a problem

or if a very uniform scintillator Is being used, ^Li

glass makes a very nice flux monitor below 200 keV.

Acknowledgements

I would like to thank R. B. Schwartz for his

assistance in the transmission measurements of the NBS

^Li glasses and M. C. Moxon for advance copies of his

results and figures.

References

1. G. P. Lamaze, R. A. Schrack, 0. A. Wasson, To be

published

.

2. G. Hale, See these proceedings.

3. James Schenck, Nature 171( 1953) 518.

4. R. J. Ginther and J. H. Schulman, I.R.E. Trans.

NS-5(1958) 92.

5. Voitovetskii, Tolmacheva, and Arsaev, Atomnaya

En. 6 (1959) 321.

6. A. R. Spowart, Nucl. Instr. and Meth. 135 (1976)

441.

7. A. R. Spowart, Nucl. Instr. and Meth. 140 (1977)

19.

8. M. C. Moxon, J. D. Downes, and D.A.J. Endacott,

UKAEA Report, AERE-R8409 , Harwell, Oxfordshire (1976).

9. G. P. Lamaze, J. K. Whittaker, R. A. Schrack and

0. A. Wasson, Nucl. Instr. and Meth. 123 (1975) 403.

42
INSTRUMENTS FOR USE OF ^Li AS A STANDARD*

L. W. Weston

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37830

A review is given of the instruments which use the ^Li(n,a)3H reaction as 'a standard in

the measurement of neutron flux. These instruments consist of scintillation detectors,

gaseous ionization detectors, and external particle detectors. The Li glass scintillator is the most versatile because of minimized effects of the angular distribution of

the reaction products and simplicity. The gaseous ionization detector has lower gammaray sensitivity than does the scintillation detector. Surface-barrier diodes may be

used to detect the reaction products when a neutron spectrometer is desired. A choice

of instruments to use this reaction must consider gamma-ray sensitivity, effects of the

angular distribution of the reaction, and the range of the triton in the detector.

(Instruments: ^Li(n,a)3rl, scintillation detectors, gaseous ionization detectors,

external particle detectors)

Introduction

Instruments which use the ^Li(n,a)3H reaction for

the measurement of neutron flux must use the advantages and minimize the disadvantages of this reaction.

The Q value of the reaction is 4.78 MeV, which is high

enough to make the reaction relatively easily detectable for all incoming neutron energies; however, this

high Q value makes the reaction difficult for use as a neutron spectrometer except for very high resolution

detectors such as silicon surface-barrier detectors.

The cross section of the reaction is large with relatively smooth variations except near the large reso- nance at about 244 keV. This p-wave resonance makes

the reaction difficult to use as a standard cross section in the neutron energy range from about 150 to 350 keV and causes the angular distribution to be

anisotropic in the center-of-mass system above a few eV neutron energy. The tritons show forward peaking of

about 5% at 100 eV, 18% at 1000 eV, 55% at 10 keV, and

anisotropics which do not vary monotonical ly at higher

neutron energies. ^"^ Any detector using this reaction

for the measurement of neutron flux must be independent of this angular distribution or must be understood

well enough that corrections can be made. The energy

of the alpha particle from the reaction is 2.05 MeV

for zero incoming neutron energy. The range of the

alpha particle in Argon at atmospheric pressure is 1.1 cm for zero energy neutrons and 2.0 cm for 1-MeV

neutron absorption. These ranges create no problem in most detectors. The energy of the emitted triton is

2.73 MeV for zero incoming neutron energy. The cor- responding ranges of the triton in Argon are 6 cm and

11 cm for zero and 1 MeV neutron absorption and this

rather long range creates problems and limitations in

most detectors using this reaction.

There are three basic methods by which to detect

the ^Li{n,a) reaction: 1. ^Li embedded in a scintillation detector; 2. a gaseous ionization detector; and

3. external charged-particle detectors such as silicon

surface-barrier detectors. There is no convenient

gaseous form of Li such as in the case of ^°BF3.

At the present time the scintillation detector is

the most popular because of high efficiency, accurate

time determination of the event, insensitivity to angular distribution, and low mass in the neutron beam.

The advantages and disadvantages of the three methods

of detection will be discussed.

Scintillation Detectors

There are three practical forms of scintillation

detectors available which may use the ^Li reaction.

These are the ^Li loaded glass scintillator, a LiF

layer sandwiched between two very thin plastic scintillators, and ^Lil scintillators. The glass scintillator is most popular because it is commercially available and easily used. The ^Lil crystal scintillator'*'^

is not practical for the measurement of neutron flux

because the iodine has a high capture cross section

with many large resonances. The plastic sandwich

detector developed by Dabbs et al.^ is an innovation

which is relatively unproven.

The advantages of the Li glass scintillator are that it has a high efficiency, fast timing, and may

perturb the neutron beam to only a small extent. The

principal disadvantage of the Li glass detector is its

sensitivity to gamma-rays. Because of the gamma sensitivity, the optimum thickness for a Li glass detector

is about 0.5 mm if the corresponding efficiency is

acceptable. At this thickness the loss of tritons from

the surface is <10% and the gamma sensitivity is minimized. If the glass is made thinner, a large fraction

of the reaction products is lost from the surface of

the glass because of the rather long range of the triton. If an appreciable fraction of the tritons is

lost from the surface of the glass, one no longer

obtains a peak in the pulse-height spectrum which is

well above most gamma-ray induced pulses. Also if part

of the tritons are lost from the glass, the efficiency

becomes sensitive to the angular distribution of the

reaction products.

Since Li glass can be obtained^ with a concentration of ^Li of up to 7.3%, the efficiency of 1/2-mm

glass can be as high as 93% for thermal neutrons and

0.46% for 1-keV neutrons. For higher efficiency the

glass is available up to 2.54 cm in thickness. Such

thick detectors of Li glass are used for measurements

such as total cross sections; however, one pays a price

in gamma-ray sensitivity. For this reason the volume

of glass scintillitor should always be kept at a minimum for a given application.

The gamma-ray sensitivity of Li glass has been

discussed in detail el sewhere. "^'^ The maximum value of

the peak created by the ^Li(n,a)3H reaction corresponds

to an equivalent electron of 1.6 MeV energy created by

43
a gamma ray. The pulse-height resolution (FWHM) of the

peak is 20-30%. Gamma rays which deposit less than

about 1 MeV energy in the scintillator can be effectively discriminated against by means of pulse-height

selection. With an electron linear accelerator, the

use of Li glass usually requires the equivalent of 2 cm

or more of Pb in the beam to reduce the effects of the

gamma flash for neutron flight times of less than about

10 ysec.

In time-of-f 1 ight experiments the timing of a detector is important. From Li glass a time resolution of about 2 ns can be obtained. This is a good

match to many other detectors such as fission chambers

and capture-gamma-ray detectors and thus is an attractive feature of the Li glass detector.

The mounting of Li glass scintillators so they

can be viewed by photomul tip! iers is not very critical.

In most applications it is preferable to mount the PM tubes outside the neutron beam. Such a mounting minimizes neutron scattering corrections and gamma-flash

problems if a LINAC is used. Tests at Oak Ridge indicated that coupling the Li glass optically at the

edges to two PM tubes gives equivalent results to mounting the Li glass in a light reflecting box with

no direct light coupling. Both methods^'^^ of mounting Li glass are used at ORELA in Oak Ridge.

The plastic sandwich detector has most of the

characteristics of the Li glass detector. This detector, as used by J. W. T. Dabbs,^ consists of a layer

of 96 pgms of LiF per cm^ which is vacuum evaporated

on a 0.01-cm thick NE-110 plastic scintillator. A similar layer of plastic scintillator is placed over

the LiF layer. The LiF layer is 2 cm in diameter and

the plastic 2.5 cm in diameter. The sandwich is

viewed by two photomul ti pi iers in a light-reflecting

box. The principal advantage of this detector is a faster recovery from a "gamma flash" such as is characteristic of a LINAC. The plastic does not exhibit

long-term phosphorescence as does Li glass, which can exhibit a pulse of several microseconds in width under

such conditions. Another advantage is the smooth

cross section of the plastic components. There are no resonances in silicon and oxygen to contend with. Be- cause the reaction products must be allowed to escape

from the LiF layer, it is limited to about 200 pg/cm^.

The maximum efficiency of a single detector is thus

limited to about 0.45% for thermal neutrons and 2.3 x 10"^% for 1-keV neutrons. A disadvantage of this

detector is the light output is lower than Li glass so that electronic noise is more of a problem. Also, the

"gamma flash" response of the detector is greater than

that of Li glass even though the recovery is faster.

This is not a serious problem for beams of the order of

2 cm in diameter, but presents electronic recovery

problems for beams of greater dimensions. The hydrogen in the plastic of the detector scatters about 2.2%

of the neutron beam and a correction must be applied

for neutron scattering in the hydrogen and carbon and

subsequent absorption in the ^Li.

Gaseous Ionization Detectors

The general advantage of gaseous ionization detectors is low gamma-ray sensitivity. In applications

where the gamma-ray intensity accompanying the neutron

flux cannot be effectively reduced to tolerable levels

this can be very important. The ^Li reaction has not

been used as extensively in gaseous ionization chambers

because of the popularity of the use of the ^'^B(n,a)

reaction. Chambers using BF3 gas have been used extensively in the form of cylindrical proportional counters

as well as parallel plate ionization chambers. Since

the ^Li(n,a) reaction is becoming more popular as a standard relative to the ^°B(n,a) reaction, there may

be increased use of the Li reaction in gaseous ionization counters. The ^Li(n,a) reaction has the advantage

of a less complex cross section above about 80 keV and

the absence of two groups of alpha particles as occurs

in the '°B reaction.

The gaseous ionization detector must detect the

alpha and triton from the ^Li reaction. Since the

range of the alpha particle in Argon is from 1.1 cm

for thermal neutrons to 2.0 cm for 1-MeV neutrons, a detector of about 2 cm thickness will transfer all the

energy of the alpha particle to the gas. Unfortunately

the triton range is correspondingly about 6 and 11 cm, which is too long for normal thickness chambers because

of electron drift time and voltage gradient problems.

An example of a gaseous ionization chamber using ^Li

is that of Friesenhahn et al.^^ which was a gridded ion

chamber with a spacing of 2 cm between the grid and the

Li metal coated plate. The pulse-height spectrum was complex as can be seen in Figure 1, because the tritons

which were emitted at 90° with respect to the foil gave

a much larger pulse than those emitted at 0°. This

effect was caused by the fact that the triton range was

long compared to the spacing of the chamber. Examples

RT-06361 IONIZATION ENERGY (MeV)

Figure 1. Calculated versus measured ^Li

gridded ion chamber response. Chamber constructed

by Friesenhan et al . ^

44
of non-gridded parallel plate ionization chambers are

those used by Macklin^^ and Poenitz.i^ with non- gridded chambers the pulse-height spectrum is very

broad because of the dependence on the track orientation of the particle. With these chambers it is important that either the lowest energy triton pulses are counted or that corrections be made for the tritons not

counted; otherwise the efficiency of the chamber be- comes sensitive to changes in the angular distribution

of the ^Li reaction versus neutron energy.

With present day electronics the time resolution

of a gaseous parallel plate ionization chamber is

limited to a minimum of about 10 ns with about 30 ns being achieved with relative ease. This limitation is

imposed by the low energy release of the reaction as compared to fission fragments and the long range of the

reaction products.

The efficiency of a single gaseous ionization

detector is limited by the thickness of the Li layer

in the chamber. Since 200 pg/cm^ of LiF is about the

limit, this efficiency is about 2.3 x 10'^% for 1-keV

neutrons. The effective area of such a chamber may be

many square centimeters. Proportional counters with

amplification of the ionizing events in the gas could

be constructed, but would have no obvious advantages.

External Particle Detectors

There are two forms of external particle detectors which have been used with the ^Li reaction. Both

these systems used commercially available surfacebarrier diodes to detect the reaction products from

the ^Li reaction. One system is when the Li layer is

placed very close to the surface of the silicon

diode^ and the other is when the diode is

placed outside the neutron beam in low solid angle

geometry. 1''"^

5 When the Li layer is placed on or very

close to the diode, a second diode may be placed on the other side to form a sandwich detector.

The use of surface-barrier detectors has two

advantages: very good energy resolution of the reaction products and high stopping power. Surfacebarrier detectors are capable of timing resolution of

about 25 ns. The ^Li sandwich detector has one characteristic which is unique relative to the other discussed detectors. This characteristic is that the

energy resolution of the surface-barrier detectors is

sufficient that the detector can determine neutron

energy as well as the occurrence of an absorption and

thus can be used as a neutron spectrometer. The

uncertainty in neutron energy determination (FWHM) is

about 300 keV.i** The efficiency of the Li sandwich

detectors is comparable to other detectors using a layer of LiF; however, the area is restricted to a few

cm2 in this case because of the size of available

surface-barrier detectors.

The sandwich detector has a disadvantage which

has prohibited its use except in special cases. The

surface-barrier detectors must be very close to the

LiF layer. Thus there is a relatively large mass of

silicon in the neutron beam and high backgrounds are caused by neutron-induced charged-particle reactions

in the silicon and gold of the detectors. This problem is severe for neutrons of greater than about

5 MeV energy. Figure 2 illustrates pulse-height

spectra obtained at neutron energies of 1.99 MeV and

14 MeV. Proposal s20 have been made to construct such a spectrometer using silicon enriched in ^°Si to con- struct the surface-barrier detectors. This would

reduce the charged-particle backgrounds by more than

an order of magnitude; however, this proposal has not

500

400

u 300

o 200

(00

UNCLASSIFIED ORNL-LR-DWG 48797

10=

30 40 50 60 70 80 90 100 1(0 (20 130 (40 (50 PULSE HEIGHT

UNCLASSIFIED ORNL-LR-DWG 5(263R(

10"

<

I

10'

ITH Li^F LAYER —tti M

t i—ft WITHOUT Li^F LiIYER —

14.7 -Mev NEUTRONS

n

11

r PEAKA i

0.45 Mev

i

20 40 60 80 100 120 140 160 PULSE HEIGHT

Figure 2. Pulse-height spectra for Li sandwich

detector^'* using surface-barrier diodes.

been carried out because of the expense. Because of

the high background from charged-particle reactions,

the ^Li sandwich detector will probably be retained

for specialized use.

Surface-barrier detectors outside the neutron

beam to detect the ^Li reaction have been used by

several experimental ists. ^''"i^ jhig method has the

disadvantages of very low efficiency and sensitivity

45
to the angular distribution of the reaction. Because

of these disadvantages this form of detection is suitable only when there is a very high intensity of neu- trons and other means of detection are not suitable.

Summary

The characteristics of the various detectors of

the °Li(n,a)3H reaction have been discussed. Even

though it is far from an ideal detector because of

its gamma-ray sensitivity and the large resonance in

^Li at about 244 keV, the ^Li glass scintillator is

widely applicable and a good choice of a detector for

the measurement of neutron flux. Ionization chambers

may present an attractive choice when the gamma-ray

sensitivity of the glass detector is prohibitive.

The surface-barrier sandwich may be attractive when a determination of the interacting neutron energy is

desirable and the intensity of neutrons above about

5 MeV is low. The plastic sandwich detector requires

further test and evaluation. The other forms of

detectors are desirable only for specialized cases.

References

*Research sponsored by the Energy Research and

Development Administration under contract with

the Union Carbide Corporation.

1. S. Raman, N. W. Hill, J. Halperin and J. A. Harvey, "Angular Anisotropy in the ^Li(n,a)3H

Reaction in the eV and keV Energy Region,"

Proa. Int. Conf. on Interaations of Neutrons loith

iluolei, Lowell, Mass., p. 1340, CONF-76071 5-72

,

Tech. Info. Center, ERDA (1976).

2. I. Schroder et al.. Proa. Conf. Wualear Cross

Sections and Teahnology , p. 240, National Bureau

of Standards Special Publication 425, Vol. I, U.S. Gov. Printing Office, Washington, D.C. (1975).

3. D. I. Garber et al., "Angular Distributions in Neutron-Induced Reactions," BNL 400, Third Edition,

Vol. I, p. 3-6-1, Clearinghouse for Federal

Scientific and Technical Information, NBS, U.S.

Dept. Commerce, Springfield, Va. (1970).

4. F. D. Brooks, "Neutron Time-of-Fl ight Methods,"

p. 389, European Atomic Energy Community, Brussels

(1961).

5. R. B. Murray, IRE Trans, on Nucl . Sci., p. 159,

Vol. NS-5, No. 3 (1958).

6. J. W. T. Dabbs et al., Rhys. Div. Ann. Prog. Rept.,

Dec. 31, 1975, ORNL-5137, p. 3 (1975).

7. Nuclear Enterprises Inc., Scintillator Division,

Scintillator Catalogue, 935 Terminal Way, San

Carlos, Calif. 94070.

8. G. 6. Slaughter, F. W. K. Firk, and R. J. Ginther,

"Neutron Time-of-Fl ight Methods," p. 425, European

Atomic Energy Community, Brussels (1961).

9. R. L. Hacklin, N. W. Hill, and B. J. Allen, Nucl.

Inst, and Methods 96, 509 (1971).

10. R. B. Perez et al., Nucl. Sci. Eng. 55, 203 (1974).

11. S. J. Friesenhahn et al., "The (n,a) Cross Sections

of ^Li and i°B Between 1 and 1500 keV," INTEL-RT

7011-001, Intelcom Rad. Tech., P.O. Box 80817,

San Diego, Calif. (1974); see also GULF-RT-Al 2210,

Gulf Radiation Tech., San Diego, Calif. (1972); see also Proa. Conf. Nualear Cross Sections and

Teahnology, p. 232, NBS Special Publication 425,

Vol. I, U.S. Gov. Printing Office, Washington, D.C. (1975).

12. R. L. Macklin and N. W. Hill, private communication

13. W. P. Poenitz, private communication.

14. T. A. Love et al., Nuclear Electronics I (IAEA

Vienna, 1962) p. 415.

15. I. C. Rickard, Nucl. Inst. Methods 105, 397 (1972).

16. P. J. Clements et al . , Nucl. Inst. Methods 125, 61 (1975).

17. V. V. Verbinski and M. S. Bokhari, Nucl. Inst. Methods 46, 309 (1967).

18. J. R. Lemley, G. A. Keyworth, and B. C. Diven, Nucl

Sci. Eng. 43, 281 (1971).

19. C. Wagemahs and A. J. Deruytter, Annals of Nuclear

Energy, Vol . 3, p. 437 (1976).

20. T. A. Love, private communication.

46
EXPERIMENTS AND THEORY FOR DIFFERENTIAL n-p SCATTERING

C. A. Uttley

Atomic Energy Research Establishment

Harwell

The present status of the n-p differential scattering cross section is presented over the

energy range below 30 MeV. This energy range covers the application of this cross section

for the flux or relative flux spectrum measurements which are used to produce differential

cross section data for fission and fusion reactor systems. Recent neutron-proton scattering experiments between 20 and 30 MeV have improved the isospin-zero phase shifts,

particularly 6('Pj), which largely determine the anisotropy and the asymmetry about Tr/2

in neutron-proton scattering below 20 MeV.

[Nuclear Reactions np scattering E < 30 MeV; experimental 0(6), calculated a(6), phase

shift analyses, model predictions].

Introduction

The differential neutron-proton scattering crosssection is used as a standard relative to which other

elastic cross sections are measured in the MeV region

of neutron energies. It is also the primary cross section for neutron flux or flux spectrum measurements above about 0.5 MeV. In these applications of

the cross section a knowledge of the angular distribution in both hemispheres is required.

The detection of proton recoils from hydrogeneous

radiators to determine, for example, the neutron spect- rum from a linac target, requires the angular distribution at backward angles in the centre-of-mass , and

neutron flux measurements using a proton recoil

telescope usually utilize the cross section near 180 . The angular distribution predominantly in the forward

hemisphere is needed on the other hand for relative

scattering cross section measurements, and for the

determination of the relative response of organic

scintillators to neutron energy by scattering monoenergetic neutrons from hydrogenous samples.

Over the last few years most neutron-proton scattering experiments below 30 MeV have been made in the

energy range 20-30 MeV using Van de Graaff or Tandem

accelerators and the ^H(d,n) He reaction as the neutron

source. Indeed, apart from total cross sections and

some differential scattering data at 14.1 MeV^'^ few

n-p measurements exist below 20 MeV compared with

proton-proton scattering experiments. Thus the

determination of the neutron-proton differential scat- tering at energies below 30 MeV has rplied on

(2) . calculation using the phase shifts obtained from

continuous energy phase shift analyses of all

nucleon-nucleon scattering data up to about 350 MeV. The isospin-one phases obtained from these analyses are accurate and unambiguous due to the accuracy of protonproton experiments and the variety of observables which

have been measured. The relatively inaccurate and

sparse neutron-proton scattering data prior to 1970,

however, resulted in difficulties in achieving a unique set of isospin-zero phases, especially below

80 MeV, which were theoretically acceptable unless

(3) some constraint was imposed on the analysis.

A phase shift analysis of nucleon-nucleon scattering data, which includes several new neutron-proton

scattering experiments, has recently been carried out

in the energy range 20-30 MeV. These new data

comprise neutron-proton polarization measurements at

21.1 MeV^^^ and 21.6 MeV^^^ and neutron-proton angular

distribution measurements at 24 and 27.2 MeV as well

(9) as cross section measurements at two angles at 24 MeV. Between them the measurements have considerably improved

the accuracy of the data set of neutron-proton scattering observables available in this energy range, and the

effect has been to provide a unique set of isospinzero phases for partial waves 5. = 0,1,2 which is

independent of model predictions or other constraints.

Angular Distribution and Differential Scattering Cross

Section Measurements

Two different experimental techniques are necessary

to measure the angular distribution of neutron-proton

scattering over a wide angular range in the centre-ofmass (CM.) system for energies below 50 MeV. Recoil

protons from a thin polyethylene radiator exposed to an incident neutron beam can be detected with a counter

telescope over an angular range 0 to more than 45

before the recoils have insufficient energy to be

distinguished from increasing background events. This

angular range corresponds to the whole of the backward

hemisphere for n-p scattering in the CM. system.

In order to extend the angular distribution to more

forward angles the neutrons scattered by a hydrogenous

target must be detected. This technique enables the

angular distribution to be measured from some limiting

forward angle to beyond 90 . Thus an angular region

of overlap exists between the two methods which enables

the data to be normalized.

The conversion of the relative cross section onto

an absolute scale is usually achieved by fitting a Legendre polynomial expansion to the relative data and

normalizing to the total cross section which has been

measured to an accuracy of better than 1% below 30 MeV. However phase shift analysts prefer their own freedom

to normalise relative data.

Detection of recoil protons

The most recent angular distribution measurements

employ a counter telescope in which one or more thin

solid state transmission detectors are placed between

the hydrogenous radiator and a detector, usually a plastic scintillator or cesium iodide crystal,

sufficiently thick to stop the most energetic recoil

protons. The advantages of fast rise times and fast

coincidence circuits were exploited by Rothenberg (10) to measure the angular distribution of neutron-proton

scattering at 24 MeV to an accuracy of better than 2%

between 164 and 89 in the centre-of-mass. A schematic

diagram of Rothenberg' s counter telescope is shown

in fig. 1 . The hydrogenous radiator is a polyethylene

foil mounted on a thin platinum backing and supported

on a target wheel which also has positions for a

47
TARGET WHEEL recoil telescope and the Hopkins and Breit (2) (YALE)

POLYETHYLENE

FOIL

PHOTO MULTIPLIER

Fig.

Cs I CRYSTAL

Proton recoil detector for angular distribution measurements.

platinum blank, and a carbon target which can be

centred on the telescope axis for background measurements. A modified version employing only one AE detector and with the Csl crystal replaced by an NE102

scintillator as the total energy detector was used by

Burrows to measure the angular distributions at 2A

and 27.2 MeV over the angular range 158 to 71 . Burrows also made use of a particle identification

programe to select the recoil protons which allowed his

measurements to be extended into the forward hemisphere.

The 24 MeV data of Rothenberg and Burrows,

converted into the C. of M. were fitted by least squares

to a two parameter expression A^ + A2P2(Cos 6) and

normalized to the same total cross section of 397 mb.

The values of a(180/a(90) obtained by Rothenberg and

by Burrows were 1.146 + 0.017 and 1.135 + 0.014

respectively.

(11) Montgomery et al . have recently reported an angular distribution measurement at 25.8 MeV in which

backward angle data between 90

0 and 1780.were obtained

using a similar AE - E telescope. The half-angle

subtended by the telescope at the polyethylene foil is

smaller (2.55 ) in this measurement than in the case

of Rothenberg (3.3 ) and of Burrows (6.2 ) so that the

correction to the angle of the telescope axis to obtain

the mean laboratory proton recoil angle is also less.

Few attempts have been made to measure the cross section at or near 180 using counter telescopes due to the difficulty of measuring the neutron flux incident

on the polyethylene target with an accuracy comparable

to that of the angular distribution data. An exception

was the measurement of Shirato and Saitoh(') at 14.1 MeV

in which the flux was determined by the associated

particle method.

(12) Recently Drosg has reported measurements which

give information on the 180 neutron-proton scattering

cross section in the energy range 20 - 30 MeV. In these

measurements, Drosg compared the zero-degree

3H(d,n) 4.He cross sections at 5 deuteron energies between

6 and 11 MeV (E = 23.1 to 28.5 MeV) obtained by time-ofn flight and an NE213 liquid scintillator with those

obtained with a proton recoil counter telescope. The

efficiency of the scintillator over this neutron energy

range was calibrated relative to the differential cross

3 4 section of the H(d,n) He reaction at 13.36 MeV between

neutron emission angles 6(n) of 29 and 90 . The latter

cross section was obtained from charged particle measure4 ments of the D(t, He)n reaction at 20 MeV. The zero3 4 degree H(d,n) He cross sections using the proton

calculation of the 180 neutron-proton cross sections were (5.7 + 3.3)% lower than those obtained with the

calibrated time-of-f light system.

In another experiment using both detectors, Drosg

compared the zero-degree yield of 11.23 MeV neutrons

3 3 from the H(p,n) He reaction with that of 25.3 MeV

3 4 neutrons from the H(d,n) He reaction. Thus only the

relative efficiencies of the detectors are required, and

an independent check of the relative efficiencies of

the scintillator was carried out at these two energies

3 4 4 using the H(d,n) He and D(t,n) Ho reactions. A

comparison of the telescope yields at 11.23 MeV and

25.3 MeV with those for the scintillator at these

energies determines the ratio a(180 , 11.23)/o(180 , 25.3) of the neutron-proton scattering cross sections.

The measured ratio of 2.22 + 0.06 is (6.7 + 2.9)% higher

than the value of 2.08 from the Hopkins and Breit

(YALE) calculation and supports the previous experiment.

Detection of Scattered neutrons

Three recent experiments have been reported at energies below 30 MeV in which the angular distribution

of neutrons scattered from an incident monoenergetic

beam by a hydrogenous target has been measured. In two of these measurements, by Masterson(9) at 24 MeV and by

(13) Cookson et al. at 27.3 MeV, the monoenergetic

3 4 neutrons were produced by the H(d,n) He reaction at zero-degrees, while in the measurement by Montgomery

et al.^''^ at 25.8 MeV using the Davis cyclotron the

incident neutron beam was selected by time-of-f light

.

COLLIMATOR

Sfflin OIAM

' L. ^ 0 0013 mm

Immml Nl WINDOW

Fig. 2 Experimental layout for angular distribution

measurements by detecting scattered neutrons.

The principle of the experiments is illustrated

in Fig. 2 which refers specifically to the measurement

of Cookson et al . Neutrons in the incident beam are scattered by protons in the small target scintillator

and are detected at a laboratory angle ij; by a larger

scintillator placed about Im from the target. A

scattered event is defined by a coincidence between

pulses from the two detectors occurring within a narrow

time interval ~ 100 nsec. A second organic scintillator

at a fixed angle to the incident beam is used as a monitor. Each event is characterized by three

parameters: the scattered neutron time-of-f light and

the pulse amplitudes from the target scintillator

and neutron detector. The first of these parameters

is required to separate scattered neutrons arising

from the primary beam from those of lower energies

produced in the source and from Y~rays . The pulse

amplitude distributions help to identify the type of

48
events producing peaks in the time-of-flight spectra

for the different scattering angles.

Two potentially important sources of error in the

relative cross section measurements by this method

are: (a) scattered events due to fast recoil protons

produced near the edge of the target and reaching

the neutron detector; (b) scattered events due to neutron reactions in carbon in the target scintillator.

The effect of fast recoil protons can be eliminated

by placing a thin counter in anticoincidence between

the target and neutron detectors as in the experiment

of Montgomery et al . Alternatively a thin aluminium

plate shielding the neutron detector will suffice as indicated in Fig. 2. The reactions in carbon which

can simulate a neutron-proton scattered event are

12 12 12 C(n,n')3a and C(n,n'Y) C. For all situations when

the first reaction gives neutrons which could fall

within the time-of-f light limits of the n-p elastically

scattered peak, the light output of the alpha particles

is lower than the discriminator level on the target

scintillator and so the events are not recorded. The

second reaction is significant, however, and must be

investigated experimentally.

400 T

17* SCATTERING ANGLE

I27.3 (n- p) SCATTERING

C(/i,/jO80 90 CHANNEL NUMBER

Fig. 3 Time-of-f light spectrum of scattered events

at a laboratory angle of 1 7 . (Cookson et al.)

A time-of-f light spectrum is shown in fig. 3 for

the scattering of 27.3 MeV neutrons through a laboratory angle of 1 7 in the experiment of Cookson

et al. The small peak to the left (low energy end) of

1 2 the main peak was ascribed to the C(n,n'Y) reaction

in which the 4.44 MeV Y~ray is detected in the target

and the inelastic neutron in the neutron detector.

This assignment was confirmed in a separate experiment

in which the NE102A target scintillator was replaced

by an NE213 cell of similar dimensions used with an

(n-y) pulse shape discrimination circuit . A peak was observed at the same position in the TOF spectrum,

independent of scattering angle, when Y~ray events

in the target scintillator were selected. A correction

for the effect of this reaction is necessary because

the inelastic neutrons are not resolved from the main

elastic peaks at intermediate scattering angles.

Calculated corrections are also necessary for:

(a) the loss of events due to scattering collisions

close to the walls of the target in which the recoil

protons lose too little energy to be detected, (b)

multiple scattering in the target in which the net

contribution is calculated for the attenuation of

neutrons initially scattered at the correct angle and

for those scattered into the detector by subsequent

collisions in carbon or hydrogen. Scattering events

from carbon atoms alone are not detected. The size of

these corrections in the experiment of Cookson et al . is

shown in Table 1

.

Relative efficiency of the neutron detector

The wide energy range of the scattered neutrons

to be detected requires that a measurement of the

efficiency of the detector scintillator be made. In

the experiments of Montgomery et al . and of Cookson

et al . , the efficiency measurements over the required

energy range were carried out using the associated

particle method which gives the absolute efficiency for

neutron detection. Although the absolute efficiency

is not required even for absolute cross section measurements, this method has the essential feature of being

independent of other cross section data.

The associated particle technique adopted by

Cookson et al.'''^^ utilized the \(p,n)\e, ^H(d,n)'^He

3 4 and H(d,n) He reactions in order to measure the

efficiency of a 10 cm. diameter by 2.54 cm thick NE102A.

detector over the energy range 5-25 MeV which

corresponds to the scattering of 27.3 MeV neutrons

over a laboratory angular range from 17 to 57.9 , and

STANTON MODIFIED BY McNAUGHTON •! ol UCD

NE102A ABSOLUTE EFFICIENCY FOR 116 MeV ELECTRON BIAS

15 ENERGY, MtV

Fig. 4 The absolute efficiency of a 1 0 cm diameter by

2.54 cm thick NE102A plastic scintillator for

a bias of 1.16 MeV electron energy.

allowing for an angular spread of scattering events at each detector angle of + 6.6 . The efficiency of their

detector is shown in Fig. 4 for the 1.16 MeV electron

energy bias used in their angular distribution measurement. The curve which is compared with the measured

data is the efficiency predicted by the Monte Carlo code

of Stanton in which some of the input data for the

carbon reaction cross sections and the electron-proton

light output relationship have been modified by

McNaughton et al. The agreement between measured

and predicted efficiencies observed by the Davis group

for their 7.1 cm diameter by 15.2 cm long plastic

scintillators, used in the angular distribution measurement at 25.8 MeV, is similar to that indicated in

Fig. 4. In the latter case, the difference in the

angular distribution between using the relative

efficiency predicted by the Stanton code and that given

by a freehand curve through the experimental efficiency

points is small except for the energy region round

11 MeV, where an additional uncertainty was included to the data of Cookson et al . for the two largest scattering angles in Table 1

.

49
Table 1

Corrections to angular distribution data in which

s'cattered neutrons are detected (Cookson et al.)

Lab

Scattering

angle

(degrees)

proton

bias on target

MeV

Loss of

proton

recoils

'^C(n,n'Y)

Net relative

multiple

scattering

C-of-mass

Relativistic

Correction

17 0.5 1 .001 0.950 1 .000 0.989

26.6 2.3 1.011 0.974 1 .001 0.992

33.2 4.0 1 .023 0.979 1 .002 0.995

39.2 5.4 1 .041 0.985 1 .003 0.998

45.0 6.5 1 .058 0.990 1 .005 0.999

50.8 7.8 1 .08 0.995 1 .006 1 .004

57.9 7.8 1.10 0.996 1 .007 1.01)

Differential neutron-proton cross sections

To measure the absolute cross section by detecting

the scattered neutrons requires the measurement of the

neutron flux incident on the target scintillator using

the same neutron detector. It is achieved by removing

the target scintillator and measuring the incident

neutron flux relative to unit charge collected on the

neutron source. Thus the scattering cross section at laboratory angle \p depends only on the ratio of the

neutron detector efficiencies at incident energy E and

2 o at scattered neutron energy E cos ijj , and not on the

absolute detector efficiency.

The absolute 24 MeV neutron-proton scattering

cross sections at laboratory angles of 19.5 and 25.1

(9) were measured by Masterson to an accuracy of better

than 2% using this procedure. It is applicable only

when a detector is employed capable of discriminating

between neutrons and gamma-rays since, unless a pulsed

accelerator is used, no time-of-f light discrimination

is available for the zero-degree flux measurement.

In principle the accuracy of the differential cross section will be worse than the angular distribution

data due to the additional uncertainty in the absolute

correction for multiple scattering and flux attenuation in the target.

Normalization of relative cross section data

It is common practice for angular distribution

data to be normalized to the total cross section by

fitting a legendre polynomial expansion

n

Z a. P. (cos 6) to the data after a relativistic

oconversion to the centre-of-mass system. However,

little is to be gained by this procedure if the data

are to be included in a phase shift analysis since all

measurements are then separately normalized to the most

recent total cross section data. A comparison can be

made with existing evaluations of n-p differential

scattering based on different phase shift analyses by

normalizing the relative data to unity at 90 . Such

a comparison is shown in fig. 5 of the 27.3 MeV data

of Cookson et al . and the 27.2 MeV proton recoil data

of Burrows with the predicted values of a(9)/a(90) using

the Yale phase shifts and those from the par ametrization

(16) of Binstock. The latter is based on calculations

from phase shifts predicted by the Bryan-Gers ten ( ' ^)

5 1-1

ti

RELATIVE NEUTRON -PROTON SCATTERING AT 27-3 MeV

HOPKINS & BREIT

I YALE PHASES)

BINSTOCK

^ COOKSON etal SCINI TARGET

^ BURROWS. PROTON RECOIL.

Ftg.

8 6 4 2 0 -2 -4 -6 -8 -1 COS 9^

Relative scattering cross section at 27.3 MeV

compared with predictions from a model calculation (Binstock) and from phase shift analyses.

model. The data appear to be in better agreement with

the Yale phase shifts due largely to the datum at the

most backward angle.

DIFFERENTIAL NEUTRON - PROTON SCATTERING

CROSS SECTION AT 273 MeV

^ COOKSON et al I HARWELL)

^ BURROWS I WISCONSIN 1

BOHANNON •tol. PHASE SHIFTS

LRL CONSTRAINED PHASE SHIFTS.

0 COS 9c

Fig. 6

50

Differential scattering cross sections at 27.3

MeV from relative data normalized to total cross section and compared with predictions from phase

shift analyses.
The angular distribution measurement of Burrows

was included in the data set used in the phase shift

analysis of Bohannan et al.^'^^ referred to in the

Introduction. The differential scattering cross section resulting from the normalization of Burrows

data in this analysis is shown in fig. 6 along with

the cross sections predicted by Bohannan at 27.3 MeV

(the effect of 0.1 MeV difference in energy < 5% in cross section) . The Livermore (constrained) prediction

is also shown in fig. 6.

Phase Shift Analyses

The continuous energy phase shift analyses of

(3) MacGregor, Arndt and Wright (Livermore) and of

(4) Seaman et al. (Yale) were made to nucleon-nucleon

scattering data available prior to 1970 and form the

basis for the evaluation of the H(n,n)H scattering

(2) observables up to 30 MeV by Hopkins and Breit.

The simultaneous fitting of all p-p and n-p

scattering observables over a wide energy range gave

the energy dependence of all the significant phase

parameters which improved the determination of the

lower partial wave phases of importance below 30 MeV. The analysis by the Livermore group without any imposed

constraints on the data showed that the isospin-zero

phases were relatively poorly determined due to an inadequate set of n-p scattering observables and

particularly differential scattering measurements at energies below 100 MeV. They observed a strong

3 3 correlation between the S,- D. mixing parameter,

1 Ej, and the singlet p-wave phase shift, 6( Pj), such

that when the former was constrained to be positive at low energies, as suggested by its relationship to the

deuteron quadrupole moment^ the value of 6('Pj)

decreased to a value more in accord with model

predictions. The Livermore group also demonstrated

that the solution obtained by forcing an exact fit to the ratio a(165 )/a(8Q ) at 24 MeV measured by

Rothenberg resulted in a positive £1 at low energies

and an equally acceptable value of 6('Pj). The

solution arising from this experimental constraint is

shown in Table 2.

The Livermore group also reported a single energy

phase shift analysis at 25 MeV which included all p-p

and n-p data in the energy range 20-30 MeV. The phase

parameters obtained from this search are also shown in

Table 2. Evidently the accuracy of the angular

distribution data of Rothenberg at 24 MeV is not

sufficient to outweigh the earlier neutron-proton

angular distribution data included in the analysis,

since the sign of E. is still negative and the value

1 of 6( P ) larger than for the Yale and Livermore

constrained set.

(5) Recently Bohannan, Burt and Signell have

published the results of a phase shift analysis of p-p

and n-p data in the energy range 20-30 MeV which

include the results of several n-p scattering experiments made since the Livermore analysis. The precision

of the new data has meant that many of the earlier

measurements included in the Livermore analysis at

25 MeV could be removed from the data set because they

had no significant effect on the deduced phase

parameters

.

The new high precision data now available and used

by Bohannon et al . consists of neutron-proton polarization measurements at 21.1 MeV by Morris et atl.(6) and

at 21.6 MeV by Jones and Brooks^^^ in addition to the

angular distribution measurements of Burrows at 24 MeV

and 27.2 MeV*-^^

sections at 24 MeV of Masterson.

and the differential scattering cross

(9)

The analysis includes charge splitting of all the

p-p and n-p isospin-cne phase shifts and is not confined

to separate p-p and n-p values of '

. In fact the

charge splitting of 's was included as a separate

parameter. The low orbital angular momentum phase

parameters obtained from this analysis are shown in

Table 2 although calculated higher partial wave phase

shifts were also included.

A comparison of the phase parameters listed in

Table 2 shows that: (a) the isospin-one phases, largely

determined by p-p scattering experiments, are in good

agreement; (b) discrepancies in the isospin-zero phases

are largely confined to 6('Pj) and and arise from

the inadequacy of the earlier n-p differential scattering data. In the case of no neutron-proton

scattering observable has been measured which is particularly sensitive to this parameter, and the range of

the values of noted in Table 2 has a negligible

influence on the neutron-proton differential scattering

cross section below 30 MeV compared with that for 6('p ).

The significance of 6( P^) below 20 MeV

(3) It was suggested by Macgregor et al . that the

neutron-proton scattering observable most sensitive to

6('Pj) which could readily be measured was the differential scattering cross section. This was confirmed in

a sensitivity analysis carried out at 50 MeV by

(19) Binstock and Bryan. The importance of this phase

parameter to n-p differential scattering at low energies

is shown in fig. 7(a). The summed contribution from

the three triplet S-P terms to the coefficient of Pj(cos

6) in the n-p differential scattering cross section

decreases more rapidly with energy than the singlet

S-P term. The calculation was carried out using the

tabulated phase shifts of Seaman et al . down to 10 MeV

and the constrained set of Macgregor et al . below

10 MeV for partial waves I = 0,1,2. It is clear that

most of the coefficient below 20 MeV, and therefore

the asymmetry in scattering, is determined by 6('Pj).

The reason for the suppression of the triplet S-P terms

is indicated in Fig. 7(b). In addition to a cancellation due to the sign of one -^P phase being opposite,

3 3 3 . the individual S- P terms are reduced by S^ passing

through tt/2 near 18 MeV.

The importance of the recent neutron-proton

scattering experiments between 20 and 30 MeV is that,

when included in the data set for a phase shift analysis

by Bohannon et al . , they produce unambiguous isospinzero phases with no external constraints. The

accurate neutron-proton polarization measurements of

Jones and Brooks''^'' at 21.1 MeV over the C.of M. angular range 50 to 170 have determined the spin3 3 3 3 orbit combination of the D phases (9 Dj +5 D2-I4 D^)

51
- 2

-3

TOTAL COEFFICIENT FOR I = 0.1,2

Is . V 0 1

10 20 ENERGY MeV

10 15 20 ENERGY MeV

25 30

Fig. 7 (a) Components of coefficient Aj of (cos 9) in differential n-p scattering below 30 MeV;

(b) energy dependence of individual triplet S-P components of Aj

.

Table 2

Phase shifts in degrees at 25 MeV from analyses of (p,p) and (n,p) scattering data

Phase

Shift

Livermore

(Experimental)

Livermore

(Constrained)

Livermore

(Single Energy)

Yale Bohannon et al

,

T = 1

's (n,p)

o 52.96 + 0.62

8.2! + 0.11

-5.08 + 0.03

2.59 + 0.-04

0.72 + 0.01

52.31 + 0.62

8.21 + 0.11

-5.08 + 0.03

2.59 + 0.04

0.72 + 0.01

48.84 + 1 .75

8.52 + 0.31

-5.04 + 0.15

2.54 + 0.08

0.74 + 0.03

50.51 + 0.01

8.15 + 0.17

-4.96 + 0.06

2.63 + 0.05

0.915 + 0.017

51.0 + 1.5

7.95 + 0.68

-5.06 + 0.16

2.64 + 0.12

0.751 + 0.029

T = 0

' 3^

79.48 + 0.38

-1 .85 + 0.39

-0.68 + 0.42

-2.60 + 0.09

4.16 + 0.05

0.18 + 0.03

79. 18 + 0.37

-4.61 + 0.08

1 .29 + 0.32

-2.42 + 0.08

4.08 + 0.05

0.23 + 0.03

84.49 + 2.70

-4.0 +0.69

-0.34 + 0.731

-3.21 + 0.18

80. 19 + 0.14

-4.90 + 0.48

1 .82 + 0.49

-3.00 + 0.42

3.99 + 0.42

0.14 + 0.20

81.0+ 1 .6

-5.18 + 0.47

1 .03 + 0.58

-2.91 + 0.09

3.89 + 0.10

0.038 + 0.023

from the shape of P(9) with angle. The differential nucleon data near 50 MeV to determine the value of

."'P 3 1 scattering data also involve the separate D phases and 6( P,) at this energy and to compare it with

the two types of measurements together evidently produce

a solution yielding a unique value of 6('Pj),

Model predictions of 6('Pj)

Binstock and co-workers

^ ' ^^^^ ' ^ have carried

out an extensive analysis of experimental nucleonpredictions of various meson-theoretical and

phenomenological models. They find from their

phase shift analysis that the value of 6('Pj) depends

strongly on the n-p differential scattering data

included in the data set, with less significant changes

in the other isospin-zero phase parameters. Bryan and

52
Binstock have compared the values of 6( Pj) from their

analyses with those predicted by models in which the

coupling constants for 77 and heavier meson exchange

have been obtained by fitting nucleon-nucleon data from

0 to 450 MeV. One such model is that of Bryan and

Gersten^'^^ referred to earlier as the model used by

Binstock to parametrize the total, differential

scattering and polarization for (n-p) elastic

scattering above 25 MeV. The predicted values of

6('Pj) at 50 MeV only agree with those deduced from

analysis when selected n-p differential scattering

data are included.

The value of 6('Pj) at 25 MeV predicted by the

Bryan-Gersten model is -6.25° which is more than two standard deviations smaller than the value of

(-5.18 + 0.47) from the analysis of Bohannon in

Table 2. This accounts at least in part for the

increased asymmetry observed in fig. 5 for the

Binstock angular distribution compared with those for

the Yale and Livermore (constrained) phase shifts

which have values of 6('p ) closer to Bohannon's. It

] should be pointed out, however, that the 6( Pj) of

Bohannon et al. at 25 MeV is in agreement with the

value of -5.12 predicted by the phenomenological

(22) potential model of Hamada and Johnston.

The latter model has been recommended by Lomon and

(23) . . Wilson as giving phase parameters which best

represent the n-p scattering observables at low energies.

The phase shifts below 5 MeV vary with energy as

(2£+l

)

k , where £ is the orbital angular momentum and

k is the neutron wave number. This threshold behaviour

was used in the evaluation of Hopkins and Breit. Lomon

and Wilson have discussed the deviation from the

threshold behaviour which occurs with increasing energy

due to the influence of the long range part of the

interaction potential arising from one-pion-exchange

(OPEP) . They have found from a comparison of several

models with experimental data that the Hamada - Johnston potential, which includes OPEP and is

adjusted to reproduce the deuteron data (binding

energy, triplet scattering length and quadrupole moment)

,

gives the best agreement at low energies.

Discuss ion

The phase shift analysis of Bohannon et al . centered

at 25 MeV neutron energy and motivated by recent

accurate n-p polarization data can be updated shortly

by the inclusion of the angular distribution data of

McNaughton et al. at 25.8 MeV and those of Cookson

et al. at 27.3 MeV. A comparison with existing

evaluations at 25 MeV may indicate the need to recalculate the observables 0(6). and P(9) up to 30 MeV in a

convenient format similar to that adopted by Hopkins

and Breit. In this event, the energy dependence of the

phase shifts prescribed by the Hamada-Johns ton potential

could be used and normalized if necessary to those of

the phase shift analyses.

One problem posed by the measurements of Drosg is

the accuracy with which the 180 neutron-proton

scattering cross section can be specified, since this

limits the accuracy of flux measurements above 20 MeV

using proton recoil counter telescopes. The 180 cross sections in mb/sr at 27.3 MeV predicted from available

phase shift analyses are: Yale 31.27, Livermore

(constrained) 31.16, Bohannon 31.68 and Binstock 32.35,

representing a total spread of 3.7%. The value

53

indicated by Drosg's measurements is (5.7 + 3.3)%

lower than the Yale and Livermore (constrained)

predictions and about two standard deviations lower

than the arithmetic mean value of the predicted cross

(24) sections. However Drosg has shown that the Yale

and Livermore predictions of o(180 ) over the energy

range 8 to 16 MeV, when applied to counter telescope

measurements of the 0 excitation function of the

2 3 H(d,n) He reaction of other workers, agree very well

with his measurement using the calibrated time-offlight system discussed earlier.

Acknowl ed gemen t s

Much of the material presented in this paper was produced by my colleagues Drs. J. A. Cookson (Harwell),

J.L. Fowler (ORNL) , M. Hussain (Dacca University) and

R.B. Schwartz (NBS Washington) to whom I am indebted.

Communications with Dr. P. Signell (Michigan State

University) and Dr. M. Drosg (Vienna University) are gratefully acknowledged.

References

(1)

(2)

(3)

(4)

(5)

(6)

(7)

<8)

(9)

(10)

(11)

(12)

(13)

(14)

(15)

(16)

(17)

(18)

(19)

(20)

(21)

(22)

(23)

(24)

Journal of Phys. Soc.

Nucl. Data Tables

Shirato, S. and Saitoh, K. Japan 36, 331 (1974).

Hopkins, J.C. and Breit, G A9, 137 (1971).

Macgregor, M.H., Arndt, R.A. and Wright, R.H. Phys. Rev. 182, 1714 (1969).

Seaman, R.E., Friedman, K.A., Breit, G. R.D., Holt, J.M. and Pratkash, A. Phys.

1579 (1968).

Bohannon, G.E., Burt, T. and Signell, P Rev. C13, 1816 (1976)

.

Morris, C.L., O'Malley, T.K. , Hay, J.W.

Phys. Rev. C9, 924 (1974).

and Brooks, F.D. Nucl. Phys. A222,

, Haracz,

Rev. 165,

. Phys

.

and

Thornton, S.T. Jones, D.T.L.

79 (1974).

Burrows, T.W. Phys. Rec. 07, 1306 (1973).

Masterson, T.G. Phys. Rev. C6, 690 (1972).

Rothenberg, L.N. Phys. Rev. CI, 1226 (1970).

Montgomery, T.C., Bonner, B.E., Brady, F.P.,

Broste, W.B. and McNaughton, M.W. Annual Report,

Crocker Nucl. Lab. UCD-CNL786.

Drosg, M. Proc. of Int. Conf. on the Int. of

Neutrons with Nuclei, P. 1383, Lowell 1976.

Cookson, J. A., Hussain, M. , Fowler, J.L., Schwartz,

R.B. and Uttley, C.A. INDC(UK)-28U also Fowler,

J.L. et al. Bulletin of the Amer. Phys. Soc. 22 , 53 (1977).

Cookson, J. A., Hussain, M. , Uttley, C.A. , Fowler,

J.L. and Schwartz, R.B. Proc. of Conf. on Nuclear Cross Sections and Technology, Washington,

1975, ed. R.A, Schrack and CD. Bowman.

McNaughton, M.W., et al. Nucl. Instr. and Meth.

116, 25 (1974) and ref . 11. Binstock, J. Phys. Rev. C 10, 19 (1974).

Bryan, R. and Gersten, A. Phys. Rev. D 6, 341

(1972).

Signell, P. Advances in Nuclear Physics, ed.

H. Baramger and E. Vogt (Plenum) 1969 vol. 2. Binstock, J. and Bryan, R. Phys. Rev. D9 2528

(1974)

.

Arndt, R.A., Binstock, J. and Bryan, R. Phys. Rev. D8, 1397 (1973).

Bryan, R. and Binstock, J. Phys. Rev. DIO, 72

(1974).

Hamada, T. and Johnston, I.D. Nucl. Phys. 34, 382

(1962)

.

Lomon, E. and Wilson, R. Phys. Rev. C, 1329 (1974),

Drosg, M. Private communication.
USE OF THE n,p SCATTERING REACTION FOR NEUTRON FLUX MEASUREMENTS*

J. B. Czirr

Lawrence Livermore Laboratory

University of California

Livermore, California 94550

Several contemporary proton-recoil detectors are described and compared. These detectors have been used for neutron-spectrum measurements over various portions of

the lO-keV-to-ZO-MeV energy range. Several factors which limit the accuracy of the

results are compared quantitatively. General suggestions are given for setting and

using standard cross sections and for future developments using the n,p scattering

reaction.

(Detectors; flux; neutrons; proton-recoil; scattering; standards)

Introduction

For a neutron-induced reaction to be considered

as a first-class standard, we require that the cross section be known to better than +1%. Even with this

requirement met, experimental difficulties may preclude the use of the reaction in certain energy re- gions of interest. Such considerations have led to the use of the 6Li(n,a) and TOB(n,a) or 10B(n,aY) i^e- actions as standards for neutron energies below 10 or

20 keV. Above these energies, high-precision fluxmeasurement techniques (involving a neutron cross

section) make use of then,p scattering reaction.

Other techniques of comparable accuracy involve totally absorbing detectors or associated-particle systems.

Primary-Cross-Section Genesis

The present paper will describe techniques for

neutron flux measurements in the 10-keV-to-20-MeV en- ergy range which use the n,p scattering reaction.

Figure 1 gives an example of the use of this reaction

to obtain the fission cross section of 239pu. jhis

paper describes methods of obtaining the rate Ri of

Figure 1. Implicit in this figure is the suggestion

that the n,p scattering reaction is not suitable for

everyday use by those interested in off-the-shelf flux

monitors

Contemporary Proton Detectors

A classification scheme is given in Table 1 for

five contemporary proton-recoil detectors and Figure 2 shows the geometrical arrangement for each system. 1-5

Primary-Cross-Section Genesis

Primary-Cross-Section Utilization

Secondary-Cross-Section Genesis

Secondary-Cross-Section Utilization

Tertiary-Cross-Section Genesis

Tertiary-Cross-Section Utilization —

Pu

Fission

Chamber

239r

Fig. 1. Schematic outline of the use of standard cross sections to obtain Of ( Pu)

Work performed under the auspices of the U.S. Energy Research and Development Administration. W-7405-Eng-48

54
FLUX DETECTORS

COLLIMATED ^COLLIMATOR . \

NEUTRON BEAM i \ \ \ \\^

RM. n

RADIATOR

FOILS

KFK Proton-recoil telescope

0.25 mm

SS cap

0.25mm

SS cap .

LASL Proton-recoil detector

Si(L1) semiconductor detector

Pb shield

Collimated neutron beam

6.35-cm diam

-12.7-cm diam

Thin wire

supports

Pt (conical)

particle shield

U foi

Proton fo: Aperture

for recoil protons

100 mm-

-45 cm- Con ima tor

Tij=ir

LLL Proton-recoil detector

radiator foil

Protons

NBS Hydrogen-filled

proportional counter

INCIDENT NEUTRON

SILICON SURFACE BARRIER CHARGED PARTICLE DETECTOI

ORNL Proton-recoil detector

POLYETHYLENI

ALPHA CALIBRAT

Fig. 2. Geometrical arrangements for five contemporary proton-recoi 1 -detector systems.

55
It is interesting that none of these modern detectors

rely on, a precise neutron-energy-measurement capability. This feature permits an important increase in

proton-detection efficiency. It is also to be noted

that none of the five systems match the ideal detector

listed in the last column of Table I.

A quantitative comparison of several detectors is

shown in Table II. Each of these detectors was chosen

to perform with a particular accelerator and they are

not necessarily to be compared with each other for use with a single neutron source. For this reason, the

information must be used judiciously as a starting

point for future developments or for evaluating past

results. Fortunately, most of these systems were used

to obtain the fission cross section of 235u, a circumstance which allows us to make valid comparisons. For this purpose, we show in Figure 3 the total error

quoted for several 235u f i ssion-cross-.sectton measurements as a function of neutron energy. 1-4 Because the error in measuring the fission rate is usually smaller

than that of the flux measurement, these results serve

as a rough figure of merit for the several proton detectors .

Limitations on Accuracy

I will attempt to evaluate those factors which

limit the accuracy of the various detector systems.

Eight such factors have been identified and are listed

in Table III.

1 . Cross Section Errors

The n,p total -cross-section error is less than 1%

below 30 MeV and will be ignored. Only the effect of

uncertainties in the angular distributions will be

compared in Table III. The listed percentages are the

published uncertainty in a^CQ) at the appropriate

angle. 6 jhe numbers in parentheses refer to the neutron energy at which the uncertainty applies.

2. Background

The typical percentage background over the useful

energy range is listed. If it can be assumed that the

fractional error in the background is the same for

each system (±10% of the background, for example), then

a relative uncertainty can be estimated from Table III.

3. Neutron-Energy Resolution

Because of the widely different energy ranges and

accelerators involved, only a crude characterization of

energy resolution will be given. The values of AE/E

are listed near the midrange of the appropriate energy

region for each system. In most cases, the cross- section accuracy illustrated in Figure 3 was achieved

by summing over several high-resolution time-of-f1 i ght

channels. For these cases, the practical resolution is

taken to be the fractional separation between adjacent

published energies. For the NBS proportional counter,

a timing uncertainty of 0.6 ysec was used and for the

ORNL system, a beam width of 50 nsec was the limiting

factor.

4. Counting Rate

I will assume that each investigator was free to place his proton detector at the optimum flight-path

distance. The appropriate counting rates can then be

computed directly from Table II. The resulting rates

are listed in Table III, where only the thick-foil

rates are included. The high rate for the LASL detector is a result of the short flight path (10 cm.)

employed with a monoenergetic neutron source. A

6

4

.E 0

2 -

2 0

6

o 4

KFK

LASL

ABSOLUTE

LLL

NBS

J I 'III J I I I 1 1 1

1

J I 1 1 1 1

1

J—J

10 10^ 10 10 10

Neutron Energy (eV)

Fig. 3. Figure o f merit for four contemporary systems vs. neutron energy,

t

' Neutron energy is measured bv time of flight or is inferred from the source reaction.

56
further efficiency factor of approximately 10-2 must be

applied to this system because of the necessity of obtaining each data point separately. The rate in parentheses reflects this added factor.

5. Pulse-Height Extrapolation

An error is incurred in extrapolating the measured

portion of the pulse-height distribution to zero pulse

height for two of the detector types. The electronic

threshold for the NBS proportional counter is set at

1.1 keV and this results in an extrapolation error of

less than 0.5% for neutron energies above 10 keV.^

This estimate is based on a 5% uncertainty in the

threshold energy and a constant value for energy loss

per ion pair (W) above 1.1 keV. However, variations of

10% in W have been observed in the l-to-lO-keV region

by Bennett and Yule.'' These latter results would imply

the necessity of a small correction to the NBS data

below 20 keV.

The low-energy KFK telescope relies on calculated

efficiencies to obtain the effect of a 300 keV electronic threshold. The calculated efficiency drops

rapidly below 2.5 MeV and limits the accuracy below

this energy. 1 A ±5% uncertainty in the calculated efficiency seems reasonable for energies below 2.5 MeV. Above this energy the efficiency is more slowly varying and less affected by bias uncertainties.

6. Spurious Reactions

A portion of the observed background rates is

due to neutron-induced reactions in the proton detectors. Although an absolute calculation of these rates

would be difficult, it is instructive to compare

rel ati ve rates. For this purpose, we will list the

ratio of the neutron flux to the proton flux impinging

on the various proton detectors. All ratios were calculated at an incident neutron energy of 1 MeV. It is

evident that there is a poor correlation between ^r\/^p

and the observed background rates.

7. and 8. Unwanted Neutron Scattering

The fraction of neutrons which strike the neutron

"radiator" after a prior collision is listed as "Inscatter" in Table III. These estimates are based

upon calculations and indicate a negligible error from

this source.

An estimate of the uncertainties caused by scattering materials in the neutron beam may be obtained

from the row labeled "Outscatter. " The listed quantities are the total atoms/cm^ of material in front of

the proton radiator.

In addition to the eight sources of error discussed above, there are several effects which are not routinely reported on a quantitative basis. These include 1) effective-volume changes for gas-filled

counters and 2) timing errors caused by variations in

pulse height at the input of fixed-level discriminators. In the absence of published data concerning

the magnitude of these effects, a quantitative error estimate is impossible.

General Rules for Setting Standards

I would like to summarize five personal observations concerning the demanding business of setting

standards for world-wide use.

1. Do not push your luck. (Avoid strong statements about accuracy near the ends of your useful

range.

)

2. Never allow your systematic errors to exceed

your random uncertainty. (Statistical precision is

often achieved at high cost by incurring large corrections in the process .

)

3. Do not feel obligated to build "the universal

detector" when measuri ng standards. (Tailor the detector to your source since it need not be used elsewhere. )

4. It is easy to measure the signal --the hard

part is measuring the background. (Think ahead as to what can be changed to obtain the background rate.)

5. Avoid extrapolations into unmeasured regions.

(Keep the signal away from zero pulse height.)

These statements may not be ratified by everyone

involved in standards work, but I feel that they would

go a long way toward preventing "down stream pollution."

Suggestions for Using Standards

What can the measurer of a tertiary cross section

(in the spirit of Fig. 1 ) do to take full advantage of

contemporary standard cross sections?

Include a measurement of a standard cross section

with a reaction similar to the desired quantity. For example, when measuring the fission cross section of

239pu relative to 10B(n,a) below 10 keV, include a measurement of af(235[j)^ to provide a "parallel"

standard in addition to using the ^^B "vertical"

standard as a spectrum monitor. This added information can be very useful even at energies where rapid

fluctuations in the standard cross section (235u)

preclude its use as a spectrum monitor.

Future Developments

I will now propose two new detector systems, one

a direct descendant of the LLL system described above,-^

and the other a more general detector type not cur- rently in use in standards work.

One possible disadvantage of the current LLL system is the presence of a massive lead shield in the

central portion of the neutron beam. Any detector--

such as a fission chamber--which is placed a short

distance in front of this shield would be subject to sizable backscattering corrections. An alternate

geometry with an order of magnitude greater efficiency

is illustrated in Figure 4. A thin proton radiator is

placed a few cm in front of a 1 9-cm-diameter photomultiplier (PM) tube, with the entire assembly con- tained in a vacuum chamber. Protons are detected in a 1-mm-thick 'Li-glass scintillator which covers only an outer ring on the PM tube face. A massive collimator

prevents exposure of the scintillator to the incident

beam but exposes the full radiator area. The spacing

may be adjusted so that the minimum proton energy is

approximately 50% of the incident neutron energy and

the fraction of recoil protons detected is 14%. If (n,a) or (n,p) reactions in the glass scintillator

present background problems, the newly-developed

BaF2(Ce) scintillators^ may be used instead, with an order of magnitude reduction in the troublesome cross

sections. This sytem is listed as number 6 in Table

II and should be useful throughout the l-to-20-MeV

range.

Turning now to entirely new systems, we notice

that the very convenient organic scintillators have never been used--in a manner which takes advantage of

the accurate n,p cross section--for standards-quality

57
spectrum measurements. This unfortunate circumstance

is probably based on the striking non-linearity in the

pulse-height vs proton energy^ and in large surface

losses for thin scintillators. I would like to suggest that the second effect could be eliminated by

backing a thin('^.5 mm)plastic scintillator with

enough ^Li-glass scintillator to stop maximum-energy

protons ('X'l.S mm). The resulting pulse-height spectrum for mono-energetic neutrons would be similar to that observed in proportional counters, for the top

90% of the spectrum. Monte Carlo calculations indicate that the light output from recoil carbon nuclei

and alpha particles [from 12c(n,a) reactions] are confined to less than 10% of that from protons, at a fixed incident-neutron energy. 10 It seems reasonable

that with appropriate effort, the pulse-height spectrum could be extrapolated to zero with an accuracy

approaching 1-2% of the total area. Scattering from

the photomultiplier tube can be eliminated by viewing

the scintillators edge on. Table II, number 7, lists

the characteristics of this detector.

These proposals do not describe finished products,

but may suggest fruitful avenues for investigation.

Undoubtedly, other schemes will arise to permit con- tinued future use of the all -important n,p reaction.

Tube

Fig. 4. Geometrical arrangement for proposed protonrecoil detector.

Appendix

A brief historical survey will be given for

proton-recoil detectors which predate the systems described in Table I. Four basic types have been em- ployed and the geneology of each will be traced back

a generation or two.

1. Gas-filled proportional counters

The NBS counter is similar to the detector described by Bennett and Yule in Ref. 7, a system in use for many years at ANL. The NBS system achieved improved timing resolution by restricting the incident

neutron beam to the central one-fourth of the detector

volume.

2. Multiple-component proton telescopes

The high-energy KFK system followed historically

the early proportional -counter telescopes described in

Ref. 11. Improved timing resolution was achieved by

using gas scintillators.

3. Single-component in-beam detectors

The LASL detector is similar to the system described by Johnson, 11 with solid-state detectors re- placing the inorganic scintillators. This technique

was also suggested by E. Pfletschinger and employed by

Kappeler and Frohner at KFK. 12

4. Single-component out-of-beam detectors

The improved shielding arrangement of the ORNL

system was first employed by Jaszczak and Macklin.5

The LLL detector took advantage of this concept and included an improved proton-detection efficiency.!

References

1. I. Schouky, S. Cierjacks, P. Brotz, D. Grb'schel

,

B. Leugers, Proceedings of the Conference on Nuclear

Cross Sections and Technology, Washington, D.C.,

March 3-7, 1975, p. 277.

2. D. M. Barton, B. C. Diven, G. E. Hansen, G. A. Jarvis, P. G. Koontz, R. K. Smith, Nucl . Sci. Engin.

60, 369 (1976).

3. G. S. Sidhu and J. B. Czirr, Nucl. Instr. and

Methods, 120, 251 (1974).

4. 0. A. Wasson, Proceedings of the NEANDC/NEACRP

Specialists Meeting on Fast Neutron Fission Cross

Sections of U-233, U-235, U-238 and Pu-239, Argonne

National Laboratory, June 28-30, 1976, p. 183.

5. R. J. Jaszczak and R. L. Macklin, Rev. Sci. Instr.,

42, 240 (1971).

6. J. C. Hopkins and G. Breit, Nuclear Data Tables,

A9, 137 (1971).

7. E. F. Bennett and T. J. Yule, Proceedings of the

Conference on Neutron Standards and Flux Normalization, Argonne National Laboratory, October 21-23,

1970, p. 385.

8. J. B. Czirr and E. Catalano, "Cerium-activated BaF2

and CaF2 Scintillators," to be published in Nucl.

Instr. and Methods.

9. J. B. Czirr, D. R. Nygren and C. D. Zafirators,

Nucl. Instr. and Methods, 31, 226 (1964).

10. R. Batchelor, W. B. Gilboy, J. B. Parker and J. H. Towle, Nucl. Instr. and Methods, 13, 70 (1961).

11. C. H. Johnson, Chapter II. C in Fast-Neutron

Physics , Part I, edited by J. B. Marion and J. L. Fowler ( Interscience Publishers, Inc. N.Y., 1960),

p. 247.

12. F. Kappeler and F. H. Frohner, Proceedings of the

Conference on Nuclear Data for Reactors, Helsinki,

June 15-19, 1970, p. 221.

58
Table I

Proton-Detector Classification

1 KFK

2 LASL3 LLL4NBS5 ORNLIdeal

Proton detector in

neutron beam

YES YES 0 YES 0 0

Neutron energy measurement

0 0 0 0 0 YES

txtrapo 1 a L I on to zero proton pulse heightT LO VF<^ T L J P)\J

Solid radiator YES YES YES 0 YES 0

Mul ti ple-element

detector YES 0 0 0 0 0

100% proton efficiency 0 0 0 YES 0 YES

Ref. Detector Code

1 . KFK - Tel escope

2. LASL - One-element telescope

3. LLL - One-element telescope

4. NBS - Proportional counter

5. ORNL - One-element telescope

0= NO

Table II

Detector System Comparison

H Atoms Proton Neutrons Flight Radiator Proton

System

2 per cm Eff

.

^min ^max

Path Area

^min

^max

(XIO^") (%) (MeV) (MeV) (m) (Cm^) (%)

l-KFK(LOW) 0.4 60 2 6 57 75 0

" (HIGH) 0.8 3 5 30 57 75 -vlO

2-LASL 0 41-2.2 1 .0 1 >15 0.10 3.1 99

3-LLL 0 26-2.8 1.7 1 >15 60 95 60

4-NBS 65 100 0.005 0.8 200 4.9 0

5-ORNL 0 16-2.2 0.26 0.2 > 6 40 18 75

6-Proposed 0 3-3.0 14 'X- 1 >15 75 50

7-Proposed 5-10 100 '\' 0.1 >15 100 0

59
Table III

Sources of Error

1 2 3 4 5 KFK LASL LLL NBS ORNL

(LOW) (HIGH)

1) Error in o (e) % 0 0.6% 0.6% 0.4% 0 0.3%

H (all E^) (5MeV) (5MeV) (5MeV) (all E„) (5MeV)

1.1% 1.1% 0.3% 0.1%

(20MeV) (20MeV) {20MeV) (20MeV)

Background

0 0/

1 /b 1 U/o

00/ 0/0

•30/

3) Neutron-energy resolution 3% 2% 3% 4% 2.5% 6%

Energy (3MeV) (ISMeV) (3MeV) (3MeV) (0.1 MeV) (3MeV)

4) Counting rate (relative) 85 8 in5 OA20 1 200 1 .0 (1000)

5) Pul se-height-extrapol ation 5% -- -- — < 0.5% —

error (<2MeV)

D > 1 0^ 105 10^ 2 36 '\' 2

Detector material 85% Ar 85% Ar Si Si 0.8%C Si 15% N2 15% N2 99%H

7) Inscatter <0.1% <0.1% 2% 1% l/.2% ? (Typical

)

8) Outscatter (Atom/cm ) negligible negligible negl i gib!

e

7xl021 8x10^^ 11x10^^ ? Material Al Al 0

60
SURFACE BARRIER SPECTROMETERS FOR CALIBRATION OF

FAST NEUTRONS IN MeV RANGE

O. P. Joneja, R, V. Srikantaiah''', M. R. Phiske, J, S. Coachman

and M. P. Navalkar

Neutron Physics Section

Bhabha Atomic Research Centre

Trombay, Bombay 400 085

India

At present there are very few methods available for calibrating fast neutrons in the MeV

range. In the present paper methods employing Li^ sandwich and proton recoil spectrometers using surface barrier detectors for calibration of neutron energies and fluxes

have been described. The results obtained with mono- energetic neutron sources in the

energy range of 1-4 MeV are given. The accuracies for energy and flux calibrations are also discussed.

(Ejiergy and flux calibration, fast neutrons, isotropic neutrons, Li^ sandwich surface

barrier spectrometer, proton recoil surface barrier spectrometer)

Introduction

At present there are very few methods available for calibration of neutrons from 1 MeV to 4

MeV, both in terms of energy and flux. The standard methods such as time of flight or proton recoil

spectrometers using gas for ionization involve either

elaborate instrumentation or strong neutron sources

or unfolding procedure which have to take into

account wall effects or end corrections. Surface

barrier spectrometers, on the other hand, are free

from some of the above disadvantages and cover a wide range of energy. Moreover sandwich spectrometers such as Li^ or He^, can be used for calibrations with both beam and isotropic neutron sources. This aspect of calibration in isotropic neutron distribution as exist in reactors or moderating assemblies

is very important since the other methods cannot be

used in such a situation. In this paper, the use of

surface barrier spectrometers for calibrating

neutrons in the energy range of 1. 0 to 4 MeV is discussed with some of the experimental results.

Description of the Spectrometers

and the Experiment

A Li^ sandwich spectrometer using two surface

barrier detectors (250 mm^ active area with a depletion depth of about 400 microns) in coincident

mode was assembled J^\J. A block diagram of the

electronics set-up is shown in Fig. (1). A thin layer

of Li^F with thickness of 150 yUgm/cm^ was

deposited on a VYNS film to act as radiator. This

method was prefered to vacuum deposition of Li^F

directly on the surface barrier detector since the

same detecting head could be used for background

corrections,

A similar type of proton recoil spectrometer

using single surface barrier detector and

* Technical Physics Division, BARC

hydrogenous radiator (ZOOytigm/cm^) deposited on VYNS film was assembled J^2.J,

The spectrometers were subjected to monoenergetic neutrons in the energy range of 1, 0 MeV

to 4 MeV obtained from p-t reaction using 5. 5 MeV

Van de Graaff accelerator. The output of the

accelerator was monitored by a calibrated long

counter.

Results and Discussions

a) Li^ Sandwich Spectrometer

A typical response of a sandwich Li^ spectrometer for neutron energy of 2. 23 MeV is shown in Fig. (2). The time integrated counts in a channel *4* corresponding to energy E is given by H *+l

where |^^= Total number of Li atoms

^fjj^- Incident flux of neutrons of energy

E falling on the radiator

j^J"^ = Normalised resolution function

^^g^ = Total efficiency

In order to calculate both energy and flux of the

incident neutrons falling on the radiator, it is necessary to calibrate channel number and efficiency

• H^' which is energy dependent. The method of doing

such a calibration involving both experimental and

theoretical calculations is explained in APPENDIX.

The results of energy calibration which is linear with

channel number is given in Fig. (3), thereby showing

that neutrons of unknown energy can be calibrated.

61
The time ixitegrated incident neutron flux can be obtained using eqn. (1) if i^ (E) can be calculated.

In the present case, the method of calculations for

(E) which is given in APPENDIX was tested using

four neutron energies which gave same time integrated flux within errors of 10%.

b) Proton Recoil Sutface Barrier Spectrometer

A typical response of surface barrier spectrometer for neutron energy of 2. 73 MeV is shown in

Fig. (4).

The proton recoil spectrum is given by

where D^p) is the proton recoil spectrum for

neutrons of energy E , Fig. (5) gives the energy

calibration of the spectrometer with channel number

which is linear, thereby showing that any unknown

neutron energy can be calibrated.

For determining the time integrated incident

flux, the following equation is used

1 <^>nie)Mr '

Where t* '^re the proton counts in Jtt channel.

The time integrated flux for neutron energies in the

energy range of 1 to 4 MeV was obtained and the

results gave a flux value within 10%.

Conclusions

Using the mono- energetic neutrons in the

energy range of I. 0 MeV to 4 MeV and with Li^

sandsrich and proton recoil surface barrier spectrometers, it is shown that it is possible to calibrate the

neutron energy with an accuracy better than 2% and

flux within I 0%. The advantage of Li^ sandwich

spectrometer is that it can be used for calibrations on isotropic neutron distribution.

References

1^1 _7 DAE Symposium on Reactor Physics,

Bombay, 1975, Coachman, J. C. , Joneja,0. P. .and Navalkar, M. P.

1^2_J National Symposium on Radiation Physics,

Mysore, 1976, Srikantiah, R. V.,

Joneja, O, P. , Coachman, J. C. and

Navalkar, M. P.

APPENDIX

.6 Calibration of Li Sandwich and Proton

Recoil Surface Barrier Spectrometers

Absolute calibration of neutron energy as well as flux demands stringent conditions as regards

energy loss mechanisms and overall efficiency of the

system under consideration.

Energy Calibration

The absolute energy calibration can be

effectively determined from the following relation

where Actual energy

Average energy deposited

in the detector system

Average energy lost by

secondary charged particles in

the radiating material

The only unknown factor, £^ can be accurately

calculated by simulating the energy lost for a given

geometry using Monte- Carlo method. In this

method, a fictitious calibrated analyser is con- structed and the events after emerging from the

radiating material are distributed as per their

energy. The difference in peak position from the

actual channel gives the average energy lost in the

radiating material. An average energy loss curve can therefore be calculated for a desired energy

range and the correction can be suitably applied.

The energy calibration without energy loss considerations in the radiating material is obtained with

a thermal neutron source in the case of Li^ sandwich

system and with a double source in the case of

proton recoil system.

Flux Calibration

The flux calibration requires an overall efficiency

of the system which in general can be defined as

»^t(f) Nt^^J - -Cr)

= Total number of events

^0 = Number of events detected

where

IM, Number of events not registered

In the case of a sajidwich system the event is not

registered if the secondary charged particles do

not satisfy the coincidence conditions and under the

circumstances, the unregistered terms consist of

Nt» i Ml -l-NoT-HMsiii - —

W

62
where Events leaked through finite

separation between the detectors

Events in which both the secondary

particles are detected by the same

detector

Either or both the secondary particles do not exceed the experimental

bias level

A Monte Carlo code COINCID is developed to take into account all factors associated for a particular system for a selected energy range. Thus

using this code absolute efficiency which is a function of energy and the system under consideration can be calculated. The calculations can be experimentally verified using various neutron energies.

For proton recoil spectrometer, the total

proton recoils give the absolute incident flux.

63
LU

h-

(/) >

cnuO

cro

UJ

_I

UJ<

cro<o

:^OO

_i

CD

L_ UJ < oO

a_

_ UJ UJ

"I

fp.iuiuuj o z >-

(r

—

X <

CO >

_JUJa <

u. > ^ by

UJz

UJ> u. o u.

too

COO CJ

Q£ Q. O

UJ q: > UJ a." < II.

Q.<

uJ

Q. T

to UJ < 0 CO

.<> cs

CM

64
TYPICAL RESPONSE OF A SANDWICH SYSTEM (Ll^)

t

no

100 - 90-

80 -

\[1 70\-

z3 60h o

^50

AO30 J L

Neutron energy 2.23 MeV

FIG.(2)

J L

AO 60 80 100

CHANNEL No. —

>

FIG. 2

120

ENERGY LINEARITY OF SPECTROMETER (LI ^)

t

1301-

120

110

olOO

90

LU

1 80

2 70

o

60

50 FIG. (3)

±

10 2.0 3.0 40

N. ENERGY (MeV) —

i

FIG. 3

5.0

/

65
S 40

cd

5 30

(A) I-

§ 20

o

o

Q 10

o

z

_l

Ld

140

120

100-

80-

60-

z40|- <

^20

RESPONSE OF A SURFACE BARRIER , PROTON RECOIL SPECTROMETER

NEUTRON ENERGY = 2.73 (MeV)

FIG. (4)

J I 0 10 20 30 40 50 60 70

CHANNEL NO

80 90

ENERGY CALIBRATION FOR SURFACE BARRIER

PROTON RECOIL SPECTROMETER

A ALPHA PARTICLE ENERGIES

FROM ( Pu*AM) DOUBLE SOURCE

o MONO-ENERGETIC NEUTRONS

FROM VAN DE GRAAFF

FIG.(5)

X

3 4

PARTICLE ENERGY (MeV)

66
REVIEW OF '°B(n,a)^Li CROSS-SECTION MEASUREMENTS IN THE ENERGY RANGE FROM 10 keV TO 1 MeV

E. Wattecamps

Central Bureau for Nuclear Measurements

B-2440 Geel, Belgium

Cross-section data of the B(n,a) Li and of the B(n,a,) Li reaction are compiled

together with related cross-section data such as a(n,ao) , o ^, branching ratio

and the ratio of i7(n,a) of ^Li to "^B. For each type of cross-secf ion a characterisation of the measurements and some comments are listed in a table. The data are illustrated in plots together with ENDF/B-IV data. Measurements of different origin

are compared, agreement or disagreement is discussed. Qualitative statements on accuracies that recommendable sets might achieve are made, and motivation for some new measurements is argued.

(Abstract: alpha; boron; branching ratio; compilation; cross-section; elastic;

lithium; neutron; measurements; total).

Introduction

Detectors relying on the '^B(n,a)^Li reaction are widely used for flux determination of thermal, epithermal or fast neutrons. The cross-section underlying

the neutron flux determinations is o(n,a) with a(n,a)=

a(n,ao) +o (n,ai) . The refers to the emission of an a-particle of about

1.8 MeV for a neutron interaction at thermal energy,

leaving the ^Li nucleus in its eround state. The a, refers to the emission of an a-particle of about 1 .5 MeV for

a neutron interaction at thermal energy, leaving the

residual nucleus in its first excited state, which decays by prompt 478 keV gamma ray emission. Some detectors rely on the detection of this gamma ray, and

therefore a(n,a,) data are requested as well as a(n,a)

data.

How large is the uncertainty of present experimental data for a(n,a) and CT(n,a,) and how do measurements of different origin compare with each other?

These questions will be discussed together with related

cross-section measurements such as ''tot'" , branching

ratio R = (n,a,) / ( a (n.Oo) + (J (n.Qi)) and ^''^ the cr (n,a

)

ratio of ^Li to B. The related experimental data

have been compiled also since any subsequent evaluation

and recommendation for a (n, a) or a (n,a]) data must be con

sistent with the present knowledge of some related

cross-sections which might be even more accurately

known, for instance fftot the a{n,a) ratio of Li to

10b.

Some data sets of a(n,a) or <7(n,a,) are deduced

from measurements of R and a(n,ao) or from cr(n,a) ratio

of ^Li to '^B and a (n,a) of °Li. In an attempt to deal

with independent observations related cross-section

data are added to this compilation, but illustrated in

separate plots.

Many reviews and evaluations have already been

made. See for instance [IRVING D.C. ( 1 967) , DERUYTTER

A.J. (1967), GUBERNATOR K. (1968), SOWERBY M.G. ( 1 970)

,

STEWART L. (1972), HALE G.M. (1976) and LISKIEN H. (1976)]

The 2200 m/s values of a(n,a) and R are 3835 ± 5 barn

[DERUYTTER A.J. (1973)] and0.y3692 ± 0 . 00006 [ DERUYTTER A. J (1967)] and therefore c(n,ai) = 3593 ± 5 barn. From thermal energy up to 1 keV the recommended fit of a(n,a ) of SOWERBY M.G. et al (1970) is claimed to be accurate

within ±1%, within ± 2% up to 10 keV and within ± 3% up

to 100 keV. According to SOWERBY M.G. (1965) the branching ratio below 150 keV is constant 0.935+ 0.005. Thus,

below 10 keV the ^^B(,n,a)TLi reaction is a well established and accepted standard. Although the cross-section

above 10 keV is not particularly smooth and large and

although (J (n,n) of hydrogen is an accepted standard,

boron 10 is often used in detectors even in that

higher energy range because of the positive Q-value

and ease of application in various detectors. To get

well overlapping results in broad energy ranges between '0B(n,a)6Li and H(n,n)H normalised data, the

o(n,a) cross-section should be known to within less

or equal 2 per cent up to 1 MeV.

The emphasis of the present review is put on the

energy range from 10 keV to 1 MeV where most of the

recent contributions deal with, and where the requested

accuracy is still not achieved. A typical request for

a(n,a) data in the range of energy from 10 keV to 1 MeV

is 2% and similar requests have been formulated by

eleven laboratories with unanimous priority I in

WRENDA 76/77.

Presentation and Discussion of Cross-Section Data

General Information

The status of experimental data for each crosssection type is drawn in Fig. 1 to 11. The essential

features or particularities of each measurement are briefly summarized in Tables 1 to 7 in Annex 1 . The

compilation deals with about 5000 pairs of energy and

cross-section values. Most of the data have been obtained on magnetic tape from the CCDN, Centre de Compilation de Donnees Neutroniques , Paris; some data were taken from publications and in some very few

cases numerical data have been extracted from published

graphs. All data were put on cards in the same format

in a single set of units [ eV, barn]. Some data sets

with high resolution but poor statistical accuracy were

summed into groups with improved statistical accuracy.

Plots were made at the CBNM computer with the code

ANGELA. The scale of the plots was adapted individually

to get a clear picture even if a drawing comprises

thousand data points. Together with the experimental

data a continuous curve is drawn which is the ENDF/B-IV

evaluated data, see MAGURNO B. A. (1975), made available

by CCDN, and IRVINGS D.C. (1967) evaluated data for the

branching ratio R.

This compilation, together with some qualitative

arguments on accuracies is part of CBNM's contribution

to the INDC and NEANDC meetings on standards and discrepancies. An investigation of assessable accuracies

of fits on the basis of a quantitative analysis is in

progress now for some selected cross-section types and

began with the compilation of an error file.

67
Data of Oj-Qj

Data from 100 eV to 20 MeV are illustrated in

Fig. 2. Detailed plots of small energy regions are given in Fig. 1, 3 and 4. A brief characterisation of

the measurements and some comments are listed in Table 1 of Annex 1

.

The measurements of MOORING F. (1966) and DIMENT K. (1967) agree well and they are the basic data for the

evaluation of this cross-section. Measurements prior to those suffer from poor accuracy of sample composition.

The data of SPENCER R. ( 1 973) were obtained with 0 . 2 ns/m

resolution from 90 keV to 420 keV and do not show a suspected narrow resonance. Spencer's data are shown

in their original high resolution fashion, namely 538

points, in Fig. 3, but in condensed fashion, with improved statistical accuracy, in Fig. 2. The promising

data of AUCHAMPAUGH G. (1976) are claimed to have an uncertainty less than 1 .5%.

Below 10 keV DIMENT 's data and fit are accurate to within 0.5%, and fulfil the requested accuracy. From

10 keV up to 400 keV new data of SPENCER R. (1973) make

it senseful to investigate fits up to 400 keV to get a recommendable curve with properly defined accuracies.

Additional measurements are needed from 400 keV to

1.5 MeV to fulfil the 1% request up to 1 MeV, (see

WRENDA 1977, CASWELL R.S. n° 691016) since the available data sets are scarce, DIMENT K. (1967) and

BOCKELMAN C. (1951), and differ by 10% in that particular energy range. Above 1.5 MeV five resonances are apparent and the ENDF/B-IV evaluation does not fit too well the experimental data as it is illustrated in

Fig. 4.

Data of g(n,a), q(n,ai) and C7(n,ao)

These data are illustrated in Fig. 5, 5 bis, 6 and

7. Characteristics of the measurements with some com- ments are listed in Tables 2, 3 and 4 of Annex 1. Below 10 keV the recommended data of SOWERBY M.G.(1970),

with an accuracy of 2% at 10 keV, are commonly accepted.

Above 10 keV much effort is still devoted to accurate

measurements and many different methods have been and

are still applied. Recent experiments by FRIESENHAHN J

.

(1974) and SEALOCK R. (1976) have been performed with

refined and new techniques but their data did not solve

the problems as such. By detailed consideration of Fig . 5 and in particular Fig. 5 bis, in the range of energy

from 30 keV to 1 MeV, the data can be split into three

groups: "high", "low" and "medium". The "low" group

comprises the results of DAVIS E. (1961), BICHSEL H. (1957) and MACKLIN R. (1968), the "high" group com- prises the data of FRIESENHAHN J. (1974), BOGART D. (1968) , COX S. XI 966), MOORING F. (1966) and BILPUCH E. (1960), and the "medium" group comprises data of

SOWERBY M.G. (1970), SEALOCK R. (1975) and ENDF/B-IV,

HALE G. (1973)

.

Handdrawn curves through the "low" and through the

"high" group, with all data considered of equal weight,

yield two distinct curves which differ by 10, 23, 30

and 25 per cent at 100, 250, 500 and 750 keV respectively. The data of SEALOCK R. (1975) coincide with the

"low" group at 200 and above 600 keV, but with the

"high" group at about 400 keV.

The data of SOWERBY M.G. (1970) are not the result

of an independent observation. The '^B(n,a)^Li cross- section was deduced from a measured ratio of (n, ) of

^Li to "^B and 6Li(n,a) T cross-section which is not

numerically documented. As a matter of test the author

of this paper has used the well documented and accurate

ratio measurements of SOWERBY M.G. (1970) together with

^Li(n,a)T data of the ENDF/B-IV evaluation. It turns

out that this B(n,a) Li cross-section is 2, 3.4, 3.3,

7.5 and 15.4 per cent higher than the recommended data

of Sowerby at 20, 40, 60, 80 and 100 keV respectively.

This and previous arguments illustrate that o(n,a)

of '^B still does not have the status of a standard

cross-section with an accuracy of 2% as requested in

the energy range from 40 keV to 1 MeV.

The peculiar o(n,a) data of ENDF/B-IV above 1 MeV

(see Fig. 5) are to be explained by the evaluation of

o (n,o.t) data in Fig. 6. Above I MeV there are only two measurements, DAVIS E. (1961) and NELLIS D. (1969),

the shape is similar but the amplitude differs by a factor 1.4. Evaluators have preferred the a(n,ai) data

of NELLIS and have renormalized the data of DAVIS.

The latter did also measure a(n,a) but NELLIS did not,

consequently the recommended o(n,a) data float about

40% above the results of DAVIS.

The overall status of a(n,a,) data is much more

favorable than for a(n,a) data. Below 100 keV all

data, except FRIESENHAHN J. (1974), agree reasonably

well and a fit might achieve the ± 2% accuracy level. In the range of energy from 100 keV to 1 MeV the ENDF/

B-IV values, established previously to the measurements

of SCHRACK R. (1976) and SEALOCK R. (1976) are reasonably well confirmed by the latter though a slight de- crease of about 4% might be suggested from 350 to 600 keV. The data of SEALOCK R. (1976) agree with the

data of DAVIS E. (1969), in the region of overlap from

700 to 1000 keV, and this is an argument against the

data of NELLIS. SEALOCK' s data agree also with

SCHRACK' s results in the region of overlap from 300 to 600 keV. Below 300 keV SEALOCK's data. indicate a trend

towards lower values similar to SEALOCK's a(n,a) data

and in both cases in contrast with the other data.

The data of FRIESENHAHN would compare much closer if

a renormalisation by 0.85 might be applied. The data

of COATES M. (1972) seem 10% low at 150 keV and so do

MACKLIN 's data. A recommended data set would hardly

get± 5% accuracy from 100 keV to 1 MeV, even if re- normalisation of some measurements were allowed, and new measurements should be undertaken.

For the sake of completeness, the o(n,ag) have

been compiled in Fig. 7. SEALOCK's data show a bump

at 350 keV. MACKLIN's data, unique results obtained

by the inverse reaction ^Li(a ,n)^0'B , and the cause of

low cr(n,ai) data, unfortunately have no other experiment of comparable quality to compare with.

All data of f^g^' ''('^>°)' ''(n,ai) and o (n,ao) are illustrated in a single plot. Fig. 8. Data of a definite cross-section type but of different origins

carry the same symbol and this might erroneously

suggest equal weight to the different results.

Lines A/v, B/v and C/v were drawn through a (n,a)

,

o(n,ai) and a(n,ao) . It is seen that a(n,a) roughly

has a 1 /v dependence up to 400 keV, whereas o(n,a,)

and o(n,ao) deviate in opposite directions from~l/v

already at 200 keV and 20 keV, respectively. The c(n,a)

line on top of Fig. 8 stresses that a(n,a)at 400 keV

is only 12% of ''tot"

Data of o n,n

Elastic scattering cross-section data from 400 eV

to 20 MeV are illustrated in Fig. 9. Characteristics

and comments to these experiments are listed in Table 5 of Annex 1 . The largest data sets are those of LANE R. (1971), ASAMI A. (1969) and MOORING F. (1966). The

agreement among those is fairly good. The data of

LANE R. and ASAMI A. are taken as input data for the

ENDF/B-IV evaluation but MOORING 's data are not.

68
The recommended ENDF/B-IV data agree fairly well (±10%)

with the measured data, though this accuracy is still

insufficient to use ^ data together with o as a means to obtain o ^-^^ with the help of ''tot~''n n

~ "abs = a„ .At 400 keV a„ „ amounts to 4 b, whereas o n,a n,n ' n,a

is 1 b. Since 2% accuracy is requested for ^ this

would imply a relative error of less or equal 0.5 per

cent on ^ which is not likely to be achieved above

10 keV, neither by experiment, nor by theoretical

analyses

.

The most recent measurement of LANE R. together with

angular distributions and polarisation experiments

have been performed primarily to determine resonance parameters for use in R-matrix analysis.

Data of branching ratio R .

Many different measurements of R have been made,

as illustrated in Fig. 10 and Table 6. Below 100 keV

the branching ratio might be defined to within 1% and

the emphasis of this discussion is put on the less

accurate data from 100 to 1000 keV.

The uncertainty of PETREE B. (1951) data is difficult to estimate, though the energy dependence of R

is smooth and consistent with more recent data. The

data of BICHSEL H. (1952) and DAVIS E. (1961) suffer

from large uncertainties of 1 0 to 20%. Data of SOWERBYM.

(1965) scatter but nevertheless indicate lower values

than any other experiment. The first experiment with

surface barrier semi-conductor detectors , see MACKLIN R. (1967), did separate more clearly ao from Oi than any

former measurement with BFg-proportional counting devices. The absolute error claimed by MACKLIN R. is

0.003 which is extremely small and never achieved so far. MACKLIN 's data should heavily determine recommendable values but a smooth handdrawn curve through the

bulk of the data does hardly fit likewise MACKLIN'

s

value at 500 keV and the LAMAZE G. (1974) value at 790 keV which is claimed to within a standard deviation

of 0.03. The data of FRIESENHAHN J. (1974) have been

deduced from separate measurements of o(n,a,) and

a(n,a). The results scatter and are systematically

high as compared to others. The recent measurements by

SEALOCK R. (1976) are made with a complex but promising

method which essentially relies on surface barrier semiconductor detectors. The experiment was not primarily

designed for R determination. The data suffer from

poor statistical accuracy, though a general trend

towards low values is present.

A critical appraisal of R-values from 100 to lOOOkeV

shows that 3 to 5% accuracy for a recommendable curve might hardly be achieved with existing data.

Data of g(n,a) of ^Li to '^B

As seen in Fig. 11 and Table 7 there are five

ratio measurements and one calculated ratio. The latter

was obtained from CT(n,a) data of ^Li and ' Ob of ENDF/B-IV

.

The measurement of SOWERBY M. (1970) from 1 0 eV to 80 keV illustrates the 1 /v range and the departure from

1 /v from 500 eV on. Particular attention must be drawn

to these measurements since they are the basis of all

recommended '^B(n,a)^Li cross-section sets below

100 keV. See also comments on a(n,a) in previous paragraph . The data of BERGMAN A. (1961) are claimed to be

within 0.4% accuracy and are systematically lower than

SOWERBY's data. The difference amounts to 3% at 30 keV.

The data of FRIESENHAHN S. scatter considerably in the

range of energy from 1 keV to 100 keV. Above 100 keV

there is no other measurement to compare with, and a shift. of 12 keV is observed between their results and

the ENDF/B-IV data. The data of PEREZ (1974) were deduced by the author of this paper from published flux

ratios and more accurate cross-section ratios, taking

into account self absorption and other corrections,

should be acquired. These preliminary data of PEREZ

nevertheless indicate a slight tendency towards low

ratios. Recent measurements of of ^-^^U relative to a(n,a) of ^Li and to l^B, by WAGEMANS C. (1971),

claimed to be accurate within ± 3%, are in favour of

ratios that are 9% lower at 30 keV than the presently

recommended ratios.

Below 1 keV the ratios are well established, from

1 keV to 100 keV the accuracy gradually decreases with

increasing energy and hardly gets ± 3% at 100 keV.

Above 100 keV experimental data are too scarce to draw

significant conclusions.

Conclusions

The compilation of a(n,a) and o(n,ai) cross-section

data shows that data requests from 10 keV to 1 MeV are not fulfilled. To define recommendable data with as- sessable accuracies a new evaluation is recommended.

The evaluation ought to take full account of and be

consistent with related and available accurate cross- section data, such as branching ratios and a(n,a)

ratios of ^Li to 'Ofi.

It is questionable whether new measurements of Oj-q^

from 400 keV to 1 MeV would reduce the error margins

of a (n,a)

.

New measurements of a(n,ai) or a(n,a) are inevitable

if one has to achieve 2% accuracy up to 1 MeV. The

large scattering of present a(n,a) data and the common

use of a(n,a,) for flux determination in cross-section

measurements tends to recommend primarily new a(n,a ) and branching ratio measurements.

Cross-section ratio measurements above 10 keV of

o(n,a) of ^Li to ' ^B with a single detector are re- commended. Evaluations of "-'b and ^Li cross-section

data should yield data sets which are consistent with

the measured ratio of a(n,a) of "Li to B.

Acknowledgement

1 should like to thank Dr. C. BASTIAN, Mr. R. BUYL

and Mr. J. VAN GILS for providing and running the

plotting programs and Mrs. P. DAEMS-LUYPAERTS for her

diligent help in preparing the drawings.

69
REFERENCES

ASAMI A., J. Nucl. Energy 24_ (1970) 85

AUCHAMPAUGH G.F. et al., to be published

BECKER R. et al . , Phys.Rev. J_02, n° 5 (1956) 1384

BERGMAN A. A. etal., Soviet Phy . JETP _12. n° 5 (1 96 1 ) 895

BICHSEL H. et al . , Phys.Rev. \08_, n° 4 (1957) 1025

BILPUCH E., Ann. Phys . _l_0 (1960) 455

BOCKELMAN C, Phys.Rev. 80, n° 6 (1950) 1011

BOGART D. et al . , Nucl. Phys. A _I_25 (1969) 463

COATES M.S. et al . , Conf . Vienna "Neutron Standards and

Reference Data" IAEA-STI/PUB/37 1 , p. 129,(1972)

COOK C.F. et al., Phys.Rev. 94, n° 3 (1954) 651 COOKSON J. A., Nucl. Phys. A _146^ (1970) 417

COON J.H. et al., Phys.Rev. 88, n° 3 (1952) 562

COX S.A. et al., J. Nucl. Energy 2J_ (1967) 271

DAVIS E.A. et al . , Nucl. Phys. 27_ (1961) 448

DE PANGHER J. et al . , Rept. BNWL-260 (1966)

DERUYTTER A. et al .

, J. Nucl. Energy 2J_ (1967)

DERUYTTER A.J. et al. J .Nucl .Energy 27^ (1973) 645

DIMENT K.M., Rept. AERE-R-5224 (1967)

FOSSAN D.B. et al . , Phys.Rev. _122, n° 1 (1961) 209

FRIESENHAHN S.J. et al . , Rept. GULF-RT-A 12210 (1972)

FRIESENHAHN S.J. et al., Rept. INTEL-RT 7011-001 (1974)

GUBERNATOR K. et al . , Rept. EUR 3950 e (1968)

HALE G.M. et al . , see MAGURNO B.A. (1975)

HALE G.M. et al., Rept. LA-65 1 8-MS or NEANDC (US) -200/U

(1976)

HAUSLADEN S.L., Dr. Thesis COO-1717-5, Ohio University

HIBDON C. et al . , Rept.. ANL-4552 (1950) 6

HOPKINS J.L., see MAGURNO B.A., p. 73 IRVING D.C., Rept. ORNL-TM-1872 (1967)

LAMAZE G.P. et al., Nucl.Sci. and Eng. 56^ (1975) 94

LANE R.O. et al . , Phys.Rev. C 4_ n° 2 (1971) 380

see also HAUSLADEN S.L.

LISKIEN H., Conf ."Interactions of Neutrons with Nuclei"

Lowell (1976) 1110

MACKLIN R.L. et al . , Phys.Rev. ^65^ n° 4 (1968) 1147

MAGURNO B.A., Rept. BNL-NCS-50464 or NEANDC (US) - 1 96/L

or INDC(US)-73L or NEACRP-L-148 (1975)

MOORING F.P. et al., Rept. ANL-6877 (1964)

MOORING F.P. et al . , Nucl. Phys. 82 (1966) 16 NELLIS D.O. et al . , Phys.Rev. C J[, n° 3 (1969) 847

NERESON, Rept. LA-1655 (1954)

PEREZ R.B. et al . , Nucl.Sci. and Eng. 55 (1974) 203

PETREE B. et al . , Phys.Rev. 83^, n° 6 (1951) 1148

PORTER D. et al . , Rept. AWRE 0 45/70 (1970)

ROHRER, see BILPUCH E.(1960)

SCHEACK R. et al . , priv. communication and Rept.

ERDA/NDC 3/U (1976) 148

SEALOCK R.M. et al . , Phys.Rev. C 1 3 n° 6 (1976) 2149

SOWERBY M.G., J. Nucl. Energy A/B 20 (1966) 135

SOWERBY M.G. et al . , J .Nucl . Energy U (1970) 323

SPENCER R. et al . , Rept. KFK 1518 and EANDC (E) 1 47"AL"

(1973)

STEWART L., Conf. Neutron Standard Reference Data

IAEA-Vienna (1972) 149

TESCH K., Nucl. Phys. 37^ (1962) 412

TSUKADA K. et al . , J.Phys.Soc. of Jpn JS^ n° 5 (1962) 610

VALKOVIC V. et al . , Phys.Rev. n9 (1965) 331

VAUCHER B. et al ., Helv. Phys .Acta 43 (1970) 237

WAGEMANS C. et al . , Ann. Nucl .Energy _3 (1976) 437

WILLARD H.B., Phys.Rev. 98 n° 3 (1954) 669

WRENDA World Request List for Nuclear Data INDC(SEC)-

55/URSF(1976)

70
T 1 1

- + o L. n 1 1 1 1 1 1360 • SPHM I T T 1960 z

* ^-"^ SRFFORO 1960

Z z , ^--^ PROSDOCIMI 1963 -

/ AAA

^ih^z

3000 -

2000

1 1. .. 1 -J 1 ' -1 ' ' 6 8 10 20 40 60 80 100 ENERGY [meV]

Fig.1 cj^tot ^''OfTi 5meV to lOOmeV.

ENERGY

Fig. 2 (fioi from 100 eV to 20 MeV.

71
9,00

8.00

7.00

6.00

5.00

l5 A.50

4.00

3.50

3.00

—I 1 1 1 1—1" ! -T-T- I I I —1—1—1—I—I 1

r 1 I I—I—I—I—r—III I

1—I I r I I I T1—I—I—I—1—I 1 r

+ BOCKELMflN 1951

ROHRER 1960

* MOORING 1966

DIMENT 1967

X SPENCER 1973

-I I I t . i 1,1 ' ' I I I I I ' ' ' ' -J—I—I—I 1 11. lOkeV 20 O) 60 80 100 keV ENERGY

200 400 600 800 1MeV

Fig.3 o'tot from lOkeV to IMeV.

2.60

2.40

220

2.00

1 r—1

1 1

r

1 1 F1 1 1 p 1 1 1 1 1 1 1 -T~—1

r

1 1 1

. \ *

* —1—1—1

+ BOCKELMflN 1951

\ +•> f COON 1952

\

* V + */\ COOK 195^ ++ \ ( V

+ \ I \* X NERESON 195^

Y BECKER 1956

A < K \ .fa - I

\ *\ * V

X&

FOSSflN

TSUKflOn

1961

1963

A***. * *L* \ * **[ \ A COOKSON 1970

+ +\ +/+ \+++/ W \ + / + « * * * L *V o PORTER 1970

y V

ENDF/B IV 1975

* \ A

\

* \ A X X

Y

I « Y Y

\X Y n sxv Q X at: zz z K k

Z Y yyyV^x

Y Y xiSms^ Z »T«

1 1 1 1 —J 1 1 1—1—1—1 1 1 1 1 1 1 1 1 1 1 1 1 1 j__

X

1 1 1 1

X z

Y X Z

z

Y Y XXI

Y —1 1 1 1

1

' ' 1 V~T^

—1 1 1 1 1 1 1 1

1 1.2 1.4 16 18 2 3 4 5 6 7 8 9 10 12 14 16 18 20

180

1.60

1.40

ENERGY [MeV]

Fig.A (fiot from 1 MeV to 20 MeV.

72
ENERGY

Fig.5 CCn.ao) + CCn.a,) from IkeV to 20MeV.

lOkeV 20 40 60 80 100 keV 200 AOO 600 800 IMeV ENERGY *

Fig.5bis o'(n,ao) + o'(n,ai) from lOkeV to IMeV.

73
0.01

I I I I I

I

\ \ > ' I I I I I I I I I I 1 I l_l I—I—1 J. i, J— . I 1 1 1—I—I—I 1—I—I—i 1 I

I. .—J 1 1—I—I

IkeV 2 A WkeV 20 ^0 lOOkeV 200 AOO IMeV 2 i, lOMeV 20 ENERGY »•

Fig.6 CCn.ai) from IkeV to 20MeV.

20

1 1 1 1

1 1 1—I—1—1

1

f 1 r—r—1—r

1 1—1—1—n 1

r

1 1—I—I—i

r—r—r—i—n—i 1 1 1—i—r-T 1—1—I—1—TT—1 1r r—1—1

10 + DRVIS 1961

X MflCKLIN

^ SEflLOCK

1968

1976

ENDF/B IV 1975

2 O

h

0.4

0.2 "-^ ,

" « • A A /\

0.1

+ + \

0.04

1 1 1 1 1 1 1 1 L_l 1 1 1 l_ 1 1 1 1 1 1 1 1—1 1 1 1 i 1 I 1 1 1 1 1 I—I 1 1 1 1 1 1— 1 +1 1 III

IkeV 2 4 lOkeV 20 40 100 keV 200 400 IMeV 2 4 lOMeV 20 ENERGY

Fig.7 (fin.ao) from IkeV to 20MeV.

74
100 eV 200 ^00 IkeV 2 U lOkeV 20 40 100 keV 200 400 IMeV 2 4 lOMeV 20 ENERGY >

Fig.8 Cross-sectionS: o'tot , o'Caog) + o'(n,a]), o'(n,ai) and o'(n,ao). Continuous curves are ENDF/B IV 1975 evaluation

75
-

- /x

-

X

•

• • • • %* JJ-i-^*'^

z *\x\

z \

« s »- «\

x\ 0

#\

IT

'As

\x

o WILLRRO 1954 ^ \* /

-

1 1 T 1 1 Q D n T n W 1 LLHMU I -H 1 95U \ ^

+

X

TESCH

MOORING

1962

1966 X

\ ^

V

RSflMI PORTER

COOKSON

1970

1970

1970

J 1 1 -J 1 L__J 1 t l_I 1 1 1 1 1 1 1 1 1—1—1—1—1

VflUCHER LANE

ENDF/B IV

1970

1971 1975

J 1 L. 1 1 1 1 1 1 I IkeV lOkeV 20 AO

Fig.9

100 keV ENERGY

Elastic scattering cross-section from 500eV to 20MeV.

200 400 IMeV lOMeV 20

76
77
ANNEX 1 10^ Table 1. Sununary of ° ^^f. cross-section measurements of B in keV and low MeV energy range

Investigator

Name

Date - Laboratory

BOCKELMAN C. 1951 - WIS

COON J. 1952 - LAS

COOK C. 1954 - RIC

NERESON

1954 - LAS

BECKER R. 1956 - WIS

ROHRER

1960 - DKE

FOSSAN A.

1961 - WIS

TSUKADA K. 1963 - JAE

MOORING F. 1966 - ANL

DIMENT K. 1967 - HAR

Data

Energy min.-max.

Number of points

20 keV - 3.4 MeV

197

14 MeV

1

14.1 MeV -18.0 MeV

7

2 . 8 MeV -9.7 MeV

26

4.4 MeV-8.6 MeV

66

3 keV - 82 keV

41

3 . 3 MeV - 1 5 MeV

138

3 . 2 MeV -5.1 MeV

59

10 keV - 500 keV

55

76 eV - 953 keV

83

Experiment

Source - spectrum - resolution

sample

V.d.G. - mono - AE =20 keV

determination isotopic composition might

be in error D(T,n)a - mono - AE = 40 keV

accuracy of sample composition not documented

V.d.G. D(T,n)a - 8 energies - AE < 40 keV

accuracy of sample composition not documented . Reactor - fission spec . - continuous E=!0%

p . recoil spectrom. - sample at least 98% pure

Comments

E = 40 keV,

no indication on accuracy of determination

V.d.G. mono

no indicatic

of sample composition

V.d.G. - mono - AE =30 keV

5% impurities of uncertain composition

V.d.G. - mono - AE<20 keV - 93%

enriched '^B from ORNL no indication on accuracy of composition

V.d.G. - mono - AE =10 keV

same experiment yields also a h ^ ^ i 1 , Bill tot composition of samples known to within 0.01 atom-percent.

Linac - white - 0.5 ns/m

sample ' °B content (93.0 ± 0.2)%

corrections made for impurities

cross-sections about 15%

greater than those of

BARSCHALL-1946

1 .47 + 0.03 barns

accuracy 0.03 barns

Statistical accuracy± 3%

at 3 MeV, ± 8% at 13 MeV

statistical accuracy single

point about 5%.

unpublished, but see BILPUCH E. I960

a = 642 E"'^ + 2.43

tot < 3% statistical accuracy

statistical error < 2%

overall error estimated < 3%

accuracy 5%

0-0 < 100 mb

abs na below 500 keV

fit yields i, = (630.3 ±3.1)e"'^ +

(1 .95 ±0.10) for

E < 10 keV and

tot

a (DIMENT)

yields ~I ' abs

0 (MOORING)

up to 300 keV

COOKSON J. 1969 - ALD

PORTER D. 1970 - ALD

SPENCER R. 1973 - KFK

AUCHAMPAUGH G. 1976 - LAS

9.72 MeV

1

2 . 0 Mev -4.8 MeV

7

90 keV - 420 keV

538

1 . 5 MeV - 1 1 MeV

?

V.d.G. - mono - AE = 40 keV

elast.S inelastic scatt. angular cross section is measured

V.d.G. - mono - AE = 30 keV - elastic

and inelast . scatt . cross section measure- ment relative to H(n,n)H

B-sample composition given to within 0.01 %

V.d.G. - white - 0.2 ns/m.

Up to 5% uncertainty in o by devia-

• • ,. • 1 1 tot ^ tions m chemical analyser.

V.d.G. - white - 25 ps/m - agreement

with ENDF/B-IV better than 1%. Previously unknown resonances in are re- vealed - numerical data not available

yet but expected soon.

o =1597 mb

whereas FOSSAN measured 1430 mb

integration over angle and

known efficiency of calibrated

scintillator yields

tot 0 +0el. non-el.

agreement with BOCKELMAN C. 1951 - MOORING F. 1966 and

DIMENT K. 1967, if chemical

analysis of IMF Frankfurt is

taken. No narrow resonance.

No indication of narrow

resonance structure.

Error < 1 .5%

HALE G. Nov. 73 - BNLEvaluation ENDF/B-IV

0 to 1 MeV R-matrix calculation plus o tot data of DIMENT K. -67.

1 to 20 MeV smooth curve through DIMENT K.-67, BOCKELMAN C.-51, TSUKADA K.-62,

FOSSAN D.-61, COON J. -52, and COOK C.-54.

78
Table 2. Summary of B(n,ao+a,) Li cross-section measurements in keV and low MeV energy range

Investigator

Name

Date - Laboratory

BICHSEL H. 1957 - RIC

BILPUCH E. 1960 - DKE

DAVIS E. 1961 - RIC

MOORING F,

1964 - ANL

MOORING F. 1966 - ANL

COX S. 1966 - ALD

DIMENT K. 1967 - HAR

MACKLIN R. 1968 - ORL

BOGART D. 1968 - LRC

SOWERBY M. 1970 - HAR

FRIESENHAHN J. 1974 - IRT

SEALOCK R. 1975 - ORE

Data

Energy min.-max.

Number of points

20 keV - 5 MeV

98

3 keV - 70 keV

32

200 keV - 8 MeV

100

1 1 keV - 77 keV

8

10 keV - 500 keV

55

1 1 keV - 250 keV

12

100 eV - 10 MeV

83

30 keV - 500 keV

6

30 keV - 800 keV

27

10 eV - 80 keV

33

1 keV -1.5 MeV

152

200 keV -1.2 MeV

21

Experiment

Source - spectrum - detector

flux shape - normalisation

V.d.G. - mono - BF^ tubes - long counter

assiamed flat - norm, at 20 keV on ' /v

from thermal 4010 b.

no experimental determination of o

^ n,a

Comments

Long counter not documented

but see later BOGART D.-68

accuracy: 25%

a of ROHRER R. minus

0*^°*^ of HIBDON C. nn

V.d.G. - mono - BF^ grid ion. Accuracy: 20 - 30%

E< 4.5 MeV long counter, E > 4.5 MeV

known ^Li(p ,n)yield - long counter abs.

by Pu-Be source

not well documented in ANL-6877, but pro- see Mooring F.-66

bably preliminary results of type in

MOORING F.-66

V.d.G. - mono - a and ratio o ovei

fot • nn , ^ measured with transmission sample

an8 scattering detector. No flux shape

no calibration needed.

V.d.G. - mono - spherical shell transmission method - no flux shape - no calib. needed

Linac - white measured

V.d.G. - n,ao determination by inverse

reaction ^Li(a,n)"^B - long counter

"4;t graphite sphere" - additional

'Vi Oo ^ "n a ^stio measurement with SBSC.

V.d.G. - mono - BF^ - "Precision long

counter" see DE PANGHER J.; norm, at 80 keV with ' /v from thermal 3840

Linac - white - ratio measured o of

6Li(n,o) to '^BCn.a)

Linac - white- B ion.ch. but also BF3 - flux shape by CH4 prop, counter, norm,

at 4 keV by ' /v from thermal 3843,8

V.d.G. - white - SBSC-known Li(p,n)

yield gives absolute energy dependent

flux and other quantities absolute

also, therefore absolute cross-sections

Upper limit for ^ ^ '^xci + (^_p < 100 mb, tflus a'^abs

"tot -''abs fits well witTl o- 'of LANE R.-67

n,n

accuracy 10%

data for a^^^ given, but obtained by calculation from

a^^^ of DIMENT K. 1967 minus

o^^T^ of MOORING F. - 1966.

accuracy: 2%

small standard deviation and

particular type of method

used make results attractive.

calibration of BICHSEL H.-57

long counter by "precision

long counter" yields revised

BICHSEL data in good agreement.

The of

10.B is deduced from

measured cr(n,a) ratio of "^B to Li and by assuming o of

6Li to be known. The latter deduced from "tot"*!! n

'^^ numerical data given. Crosssection given by an equation

and uncertainty at 1, 10, 100

and 200 keV respectively is

1,2,3 and, 5%.

cross-checks ion ch. with

BFj-counter results deviate

by 20%.

detailed measurements of n,ao ; n,ai ; n,ao+<ii; o (^) and similar

measurements for ^Li(n,a)t.

Accuracy ± 12%, but at 350 keV is about 1.8 times

value of DAVIS E. or MACKLIN R.

HALE G. Nov. 1973 - BNL

Evaluation ENDF/B-IV

o - below 1 MeV

n,ao

o is sum of CT and a n,a n,ao n,a,

R-matrix calculation and experimental data of MACKLIN-68, DAVIS-61

and VAN DER ZWAN-72

- 1 MeV to 20 MeV based on DAVIS-67, but above 2 MeV renormalised by 1 .

4

a - below 1 MeV R-matrix calculation and experimental data of FRIESENHAHN-72

- 1 MeV to 20 MeV smooth curve through DAVIS-61 and NELLIS-70 , data of DAVIS-61

above 2 MeV renormalised by 1.4

79
Table 3. Summary of "^B(n,a, 7)^Li cross-section measurements in keV and low MeV energy range

Investigator

Name

Date - Laboratory

Data

Energy min.max.

number of points

Experiment

Source - spectrum - detector

flux shape - normalisation

Comments

DAVIS E.

1 961-RIC

200 keV - 2 MeV V.d.G. - mono - AE>25 keV - BF^ grid

82 ion ch. - E<4.5 MeV long counter

E>A.5 MeV known ^Li(p,n) yield

Long counter absolute by Pu Be-source

error: 20 - 30%

MACKLIN R. 1968 - ORL

30 keV - 800 keV V.d.G. - n,i^o determination by inverse

7 reaction ^Li(a,n)"-'B - long counter

"47r graphite sphere" - additionally

"n Oo/^n ai TM^3surement by SBSC. n,ao and n',a, normalised at 30 keV

on '/v from thermal ^= 3843.2 b

and use of own measured ratio n,(io to n,a, equal to 0 . 0689 ± 0 . 0006

Standard deviation estimates

2 to 2.5 % Small error and type of

method used makes results

attractive.

NELLIS D. 1969 - TNC50 keV - 15 MeV V.d.G. - mono - AE > 40 keV

31 Nal and/or GeLi - flux shape and

flux by long counter - 7-ray efficiency

also known!

6%

COATES M. 1972 - HAR

FRIESENHAHN J. 1974 - IRT

SCHRACK R. 1976 - NBS

1 keV - 300 keV

63

1 keV -1.5 MeV

56

5 keV - 600 keV

36

Linac - white - 1 ns/m - B2O2 and Nal - flux shape by spherical B-vaseline long

counter - normalisation to SOWERBY at

1 keV

Linac - white - AE = 3 keV at 250 keV

10b slab Ge(Li) flux shape methane

proport. counter - normalised on '/v

at 4 keV from 3601,7 at thermal.

"n ai '^"^'^ than values of

Sowerby: 2% at 10 keV and

up to 10% lower at 150 keV

error on shape of ^ (E)

is 2.5 to 4.6% from i IceV to 1 MeV. - Data of report

FRIESENHAHN 1972 are superseded

by rept. FRIESENHAHN 1974. In both reports o(n,a, ) data

are identical.

Linac - white - 0.6 ns/m - "^B slab Nal results Nal and Ge(Li) agree

and Ge(Li) - hydrogen proportional counter within 5% - most accurate

relative units normalised at 4 keV on results with Ge(Li) error 3%

3459,8 at thermal. between 8 and 400 keV, 5% between 5 and 700 keV.

SEALOCK R. 1976 - DRU

. 2 MeV -1.2 MeV

21 V.d.G. - white - SBSC - absolute data

by known ^Li(p,n) yield and known geometry - angular distribution also

measured

.

error ranges from 5% at .2 MeV

to 10% at 1.1 MeV.

HALE G. et al. 1975 - BNL

Evaluation ENDF/B - IV below 1 MeV - calculated from R-matrix parameters and experimental data of FRIESENHAHN 1972.

1 to 20 MeV - smooth curve through measurements of DAVIS 1961 and NELLIS 1970 with smooth

extrapolation from 15 to 20 MeV. The data of DAVIS 1961 above approximately

2 MeV were renormalised by a factor of 1.4.

80
Table 4. Summary of "^B(n,ao)^Li cross-section measurements in keV and low MeV energy range

Investigator

Name

Date - Laboratory

DAVIS E. 1961 - RIC

Data

Energy min.-max.

number of points

200 keV - 828 MeV

Experiment

Source - spectrum - detector

flux shape - normalisation

- AE> 25 keV - BF^ grid

E<4.5 MeV long counter

V . d . G . - mono

ion. chamber

E>4.5 MeV known ^Li(p,n) yield - long

counter abs. by Pu-Be source

Comments

error: 20 - 30%

MACKLIN R. 1968 - ORL

30 keV - 800 keV

7

V.d.G. - n,a determination by inverse

reaction ^Li(a,n)"-'B - long counter

"477 graphite sphere" and /o^

ratio measurement by SBSC. '

'

n,ao and n,ai normalised at 30 keV on

' /v from o(n,a) equal to 3843.2 b

and use of own measured ratio o(n,a.o)

toC7(n,a,) equal to 0.0689± 0.006.

standard deviation

estimates 2 a 2.5%.

Small error and type

of method used makes

results attractive.

SEALOCK R. 1976 - ORL

2 MeV - 1.2 MeV V.d.G. - white - SBSC - absolute data

21 by known ^Li(p,n) yield and known geometry - angular distribution also

measured

.

error: 200 keV - 15%,

600 keV - 5%, 1.1 MeV - 9%.

HALE G. 1975 - BNL

Evaluation ENDF/B-IV

below 1 MeV calculated from R-matrix analysis and experimental n,ao data input

for the fit were those of MACKLIN 1968 and DAVIS 1961. In addition,

the angular distributions of VAN DER ZWAN 1972 for the inverse reaction

were included in the analysis.

from 1 to 20 MeV based on DAVIS-61 measurements with smooth extrapolation from 8 to 20 MeV,

DAVIS-61 measurement above approx. 2 MeV was normalised by a factor

of 1.4.

81
Table 5. Summary of measured elastic scattering cross-section data in keV and low MeV energy range

Investigator Data Experiment

Name Energy min.-max. Source - Spectrum - Resolution Comments

Date - Laboratory number of points Detector

WILLARD H.B.

1955 - ORL

0.5, 1 and 1.5 MeV

+ angular distribution measured

from 30° to 120°

at 9 angles

additionally

3 points given for

V.d.G. - mono - hydrogen proportional

counter - E dependent efficiency determined by calib. with long counter

results claimed in absolute units

^ deduced by integration

of' measured da (i?) /dco which

over-all error is ±15%

^ values deviate by 16%

from o^^^ - o^^^; might be

due to non-uniform density

of the samples.

tot abs

TESCH K. 1962 - AAC

14 MeV - 1 point

angular distrib.

from 20 to 155° at

15 angles

H (d,n)a - mono - associated a-particle

technique: NE 102 and Pilot chemicals

scintillator - absolute efficiency

known by previous associated particle

technique in non-scattered beam

^ deduced by integration

of measured da(i?)d<J

"n.n is 900 mb± 50

MOORING F.P. 1966 - ANL10 keV - 500 keV

55 points

"tot V.d.G. - mono ratio over

o measured with transmission sample

no angular distrib. and Att scattering detector

no flux shape needed - no normalisation

needed

deduced from published

values, namely f^gj. - a abs

ASAMI A. 1969 - HAR

1-120 keV,

30 points angular

distrib. measured

at 39°, 65°, 120°

and 144°

Linac - white - 1 ns/m - a„ „ of 'Ob relative to aLi glasses

n,n °f C

no significant difference

between ^ at four angles

total uncertainty ranges

from 1.1% at 120 keV to 3.7% at 1 keV

PORTER D. 1970 - ALD

2-5 MeV, 7 points

ang.distr. measured at 1 0 angles

from 30° to 135°

V.d.G. - mono - AE = 30 keV - elast.

and inelastic angular scatt. crosssection - measurement relative to H(n,n)H. "-'b sample composition given

to within 0.01%

uncertainty ^ 3.8%

angular distributions well

fitted by optical model

calculations

COOKSON J. 1970 - ALD9.72 MeV, 1 point

ang.distr. meas

.

at 9 angles

V.d.G. - mono - AE = 40 keV - 0.5 ns/m

NE 213 liq.scint.

"^B sample composition characterised

to 0.1%- inelast. scatt. cross-section

also measured.

uncertainty 7.3%

angular distribution well

fitted by optical model

calculations

VAUCHER B. 1970 - LAU14.1 MeV, 1 point D(T,n)a - mono - AE = 100 keV - assoang.distr. measured ciated particle technique - NE 213 liq. at 14 angles scint. - angular distrib. of inelast.

scatt. cross section to nine levels

also available

"n.n = 972 ±35 mb

very detailed measurement

in angle and in number of

levels

LANE R.

1 97 1 -ANL

75 keV - 2.2 MeV,

67 points , angular

distrib. measured

at 5 angles and

polarization ex- periments

V.d.G. mono - AE from 30 to 100 keV

BF3 in oil - efficiency known by

calib. with Ra-Be Source - energy

dependent shape known by H(n,n)H

uncertainty < 2 . 5%

a(i?) • P(i3) polynomial

expansion.

ENDFB/IV 0 to 1 MeV calculated from R-matrix analysis - experimental data included in

1975 - BNL fit are those of ASAMI F. et al. 1966 and LANE R. et al. 1971

1 to 7 MeV smooth curve through measurements of LANE R. et al . 1971, PORTER et al .

,

1 970 and HOPKINS, 1969. Constrained to be consistent with a

^i^j. and reactions a^.

7 to 14 MeV smooth curve through measurements of HOPKINS 1969, COOKSON J. et al

.

1970, TESCH K. 1962, VAUCHER B. et al . 1970, and VALKOVIC et al. 1965.

14 to 20 MeV optical model extrapolation from 14 MeV data.

82
Table 6. Summary of branching ratio R = a, /"njOo °n,o-i measurements in keV and low MeV energy range

Investigator Data

Name Energy min.-max.

Date - Laboratory number of points

Experiment

source - spectrum - detector Comments

PETREE B. 1951 - WIS

BICHSEL H. 1952 - BAS

DAVIS E.

1961 - RIC

MACKLIN R. 1965 - ORL

SOWERBY M. 1965 - HAR

MACKLIN R. 1967 - ORL

DERUYTTER A. 1967 - GEL

FRIESENHAHN J. 1974 - IRT

LAMAZE G. 1974 - NBS

SEALOCK R. 1976 - ORU

.3 - 2.6 MeV

18 V.d.G. - mono - BF^ prop, counter

th- 0.5 - 3.5 - 3.9MeV 1 MeV cascade - mono - BF^ ion.

8 chamber

.2 - 8.0 MeV

83

30 - 700 keV

8

23 eV - 0.5 MeV

42

30 - 500 keV

7

,th

1 - 1500 keV

62

790 keV

1

.2 - 1.2 MeV

27

V.d.G. - mono - BF^ grid ion

chamber

V.d.G. - mono - SBSC

10 Linac - white

mult

.

BF^ low gas

V.d.G. - mono - SBSC

reactor - Maxw. - SBSC

Linac - white - Ge(Li) for ai and BF^ for Oo + a,

V.d.G. - Mono - BF^ prop, counter

V.d.G. - white - SBSC

numerical values deduced from

drawing - uncertainty on R

difficult to estimate

AE

neutron spectrum broad ——1%

uncertainty R ranges from

0.6% at E^^ to 10% at 3.9 MeV

R was deduced from published

numerical data of a„ „ and

a_ a. error of which is 20 to 30%

below 200 keV error < 0.005

above 200 keV error < 0.05

abs.error< 0.015 up to 600 keV

abs . error 0.003

R = 0.93692+ 0.00006

high qualityOo ~ Qi resolution

abs. error on E-dependent shape

of R is 2 keV - 0.010;

50 keV- 0.006; 500 keV- 0.031

R = 0.66+ 0.03

R was deduced from published

numerical data of ^ and

o„ „ error of which is 4 to 15% n,ii

IRVING D. 1967 - ORLEvaluation, smooth curve through experimental points of SOWERBY M.-65, MACKLIN R.-65,

DAVIS E.-61 and PETREE B.-51

Table 7. Summary of measured ratios of ^ of ^Li(n,a)T to "^B(n,a)^Li in keV and low MeV energy range

Investigator

Name

Date - Laboratory

BERGI'IAN A. 1961 - LEB

SOWERBY M. 1970 - HAR

PEREZ

1974 - ORL

Data

Energy min.-max.

number of points

50 eV - 40 keV

20

10 eV - 80 keV

86

2-100 keV

17

Experiment

Source - Spectrum - Resolution

Detector Li, detector B

Slowing down time spectrometer - AE

14% around the mean at best - Ar prop,

counter with "^B and ^LiF layer of

.22 and 1.13 mg/cm^ respectively

Linac - white - 2.5 ns/m - 0.3 cm thick

Li glass scint. - BF^ counter below

1 keV boron plug + Nal above 50 eV up

to 74 keV - multiple scattering correct,

for ^Li up to 15%

Linac - white - 0.2 ns/m - 1 mm thick

^Li-glass and BF^ ionization chamber

Comments

corrections small - total error

on ratio about 0.4% - numerical

data were deduced from drawing

on p. 898 of publication

both measurements in relative

units and normalised at 10-20 eV on the ' /v value deduced from

thermal of 3839 ± 20 and 940 ± 4 barns for '^B and ^Li respect.

Uncertainty in ratio<0.0030 i.e. < 1 .2 %.

ratio of neutron flux spectra as published on page 206 were multiplied by ratio of Li to B cross-sections of Sowerby et al

.

to get original ratio of counting

rates. This ratio in turn is

proportional to a of Li to B,

This procedure is crude and more

accurate data are expected from

the authors.

83
Table 7. (continued)

Investigator

Name

Date - Laboratory

FRIESENHAHN S. 1974 - IRT

WAGEMANS C. 1976 - GEL

Data

Energy min.-max.

Number of points

1 - 1500 keV

179

30 keV

Experiment

Source - Spectrum - Resolution

Detector Li, Detector B

Linac - white Li ion chamber

ion chamber and BF. counter

.

2 -proport

.

Linac - white - 0.7 ns/m - SBSC

thin LiF and B layers of 220 and

and 176 i^g/cm^ respectively.

SBSC detectors are located outside the collimated neutron beam

and detect those a-particles

emerging at about 90° on neutron

flight direction.

Comments

Systematic uncertainty in ratio

amounts to 4.6%, 1.2%, 0.2%

and 0.1% at 1 MeV, 0.5 MeV,

0.1 MeV and 50 keV respectively

error about 3% - ratio is

normalised at 1 keV at 0.2479

value of Sowerby et al.

MAGURNO B. Evaluation of ENDF/B-IV of Li(n,a)T and B(n,a) Li cross-section data were used to cal1975 - BNL culate ratios for comparison with experiments.

84
INSTRUMENTS FOR USE OF i^B AS A STANDARD

A. D. Carlson

National Bureau of Standards

Washington, D.C. 20234

The interaction of neutrons with provides two commonly used neutron cross section

standards. The i"B(n ,ai7) ''Li reaction is implemented with detectors such as Nal or Ge^Li)

to detect the 478-keV gamma ray emitted in this reaction. The ^''B(n,ao+ai7)^Li (commonly

referred to as i°B(n,a) ''Li ) reaction is used with proportional counters, ionization

chambers, solid-state detectors and boron scintillators. A discussion of the use of

these detectors for implementing these cross sections will be presented.

(Boron scintillators; Ge(Li); ionization chambers; Nal; neutron flux determination;

proportional counters; solid-state detectors; standard cross section; i°B(n,ai7 )'^Li

;

^OB(n,ao+ai7)^Li)

Introduction

The measurements of neutron cross sections are greatly simplified when the data are obtained relative

to standard neutron cross sections. The neutron flux

is then essentially determined with the standard cross section and the basic flux measuring techniques need

not be employed. The i°B(n,a)''Li reactions are often

used as cross section standards. These reactions and

Q values are outlined in Fig. 1 along with the energy

level diagram for ^^B. There are two basic reactions

of interest. For the i°B(n,ao)^Li reaction, the ^Li

nucleus is left in its ground state and the total

energy shared by the reaction products is 2.792 MeV

plus the kinetic energy of the incident neutron. For

the i°B(n,ai)'^Li* reaction, the '^Li nucleus is left in

its first-excited state and the total energy shared by

the reaction products is 2.314 MeV plus the kinetic

energy of the incident neutron. The ^Li nucleus

promptly decays to its ground state with the emission

of a 0.478-MeV gamma ray. The products of these

reactions can be detected by a number of techniques

0.478

'°B (n,ao)^Li Q = 2.792 MeV

'°B (n.a, ) Vi* Q = 2.314 MeV

+ ^(0.478 MeV)

Figure 1. Energy level diagram for ^^B showing the

decay modes of a level in ^^B to states in ''Li by

emission of an a particle.

which will be the subject of this paper. In Fig. 2 the two cross sections which are used as standards,

i°B(n,ai7)''Li and i°B(n,ao+ai yj'^Li , are shown. At the

higher neutron energies these cross sections are not

considered standards due to the large cross section

uncertainties. It will be assumed that with further

improvements in these cross sections, the upper limit

of their usefulness as standards may approach 1 MeV. Measurements made relative to these standards which

are not well defined at this time can be resurrected

at a later time when the standards have been improved.

However, fundamental problems may arise above "^SOO keV

where the cross sections are falling very rapidly with

neutron energy.

The only detectors which will be considered in

this paper are those for which the efficiency is

directly proportional to a ^°B standard cross section,

neglecting small correction factors. Thus, detectors

such as the vaseline balll used at Harwell for

which the efficiency is almost independent of the i°B

standard will not be discussed.

NEUTRON ENERGY , keV

Figure 2. The i°B(n,ai7) and iOB(n,a) cross sections

from 1 to 1000 keV.

85
n,aiy) ''Li

Cross section measurements relative to the

i°B(n,ai7j ''Li cross section have been made only with

detectors which record the 478-keV gamma ray from this

reaction. This technique is convenient to implement

since the energy of the gamma ray produced is independent of the neutron energy. Thus, response functions

which are dependent on neutron energy are not required

in analyzing these data. In principle it should be

possible to implement this cross section by detecting

the charged particle{s) produced when ''Li is left in its

first excited state. This is not practical except

possibly at low neutron energies due to difficulties

in separating these events from those leaving '^Li in

its ground state.

Gamma-Ray Detectors

Nal and Ge(Li) detectors have been employed in using this cross section. Fig. 3 shows an experimental

layout for making these measurements. Both a Nal and

a Ge(Li) detector are shown but only one of these

detectors would be used for a given experiment. The

neutron beam is collimated to a size larger than the

size of the i°B sample yet small enough so that the

beam does not strike the gamma-ray detector. The

detector-sample geometry shown here is particularly

appropriate for linac experiments. Having the detector

at back angles reduces the effect of the scattered

gamma-flash photons. The background from scattered

neutrons is determined by replacing the i°B sample

with carbon or ^^B. This measurement is important

since scattered neutrons can interact with the detector

materials and produce a gamma-ray background. Also

shown in this Fig. is the detector employed for the

measurement of the cross section to be determined.

The positioning of the two detectors is dependent on

a number of factors which must be evaluated for each

given experiment, i.e. relative counting rates in the

two detectors, materials in the beam for each detector,

relative time resolution of each detector for whitesource measurements, etc. Fig. 4 shows the pulseheight distributions for a Nal detector for 510-keV

neutrons as obtained by Schrack.^ The upper curve are data obtained with the i°B sample in place. The lower curve is obtained with a carbon scatterer replacing the

i°B sample. This Fig. shows the importance of making

scattered-neutron background measurements. In addition

to the prominent 478 keV line, 203-keV and 417-keV

gamma-rays from iodine and 438-ke\/ gamma rays from

sodium can be seen. The resolution of Nal detectors

limits the usefulness of this technique particularly

at high neutron energies where many background gammarays from neutron interactions in sodium and iodine

can produce a significant background near the 478-keV

1 ine.

O . cj f- r I I ^ r I I 20 40 60 80

PULSE HEIGHT CHANNEL

COLLIMATOR

FOR '"b sample

No I DETECTOR

COLLIMATOR

FOR CROSS SECTION MEASUREMENT

DETECTOR

NEUTRONSB SAMPLE

CROSS SECTION MEASUREMENT

DETECTOR

Figure 3. Simplified experimental setup for implementing the i°Bin,aiy) cross section.

Figure 4. Pulse-height distributions for 510-keV neu- trons for a Nal detector system used in implementing

the ^°B(n,aiT) cross section. The upper curve was obtained with a i°B sample. The lower curve was obtained with a carbon scatterer. The arrow labeled

203 keV shows the position of the (n,n'J gamma-ray

from iodine. The arrow labeled 421 keV shows the

position of a peak which is a mixture of 438-keV gammarays from sodium and 417-keV gamma-rays from iodine.

The use of Ge(Li) detectors significantly simplifies this problem as a result of the intrinsically

better resolution compared with Nal and since the gammarays from inelastic scattering in germanium are well

separated from the 478-keV line. Fig. 5 shows measurements by Orphan^ of the pulse-height distribution for

a Ge(Li) detector for neutron energies near 1 MeV. The 596-keV gamma-rays from '^'*Ge and the 695-keV gammarays from ''^Ge are well resolved from the 478-keV

gamma-ray peak. The Ge(Li) detector technique does

suffer from reduced efficiency due to limitations in

the size of Ge(Li) detectors. Both Nal and Ge(Li)

detectors have good time resolution, the Ge(Li) detector

being somewhat better, so they can be used conveniently

for white source neutron time-of-fl ight measurements.

Throughout the useful region for this cross section (up to '^300 keV), the systematic errors are probably -^1%

86
—I

1 1 1 1 1 r

804 S En < 1001 KeV ®

CHANNEL NUMBER

Figure 5. The pulse height distribution for a Ge(Li)

detector used in implementing the ^°B( ^,0.1^) cross section for neutron energies from 804 to 1001 keV.

for the Ge(Li) detector and somewhat larger for Nal detectors. These systematic errors are primarily a re- sult of uncertainties in the background and small un- certainties in the self shielding and multiple scattering in the sample. In addition to these systematic

errors which are appropriate to relative measurements,

a solid angle and detector efficiency uncertainty must

be included for absolute measurements. A convenient

method for determining the product of solid angle and

detector efficiency involves replacing the sample

with a disk of equal area containing a uniform calibrated source of ''Be. ^Be decays by electron capture

to produce the 478-keV excited state of ^Li

.

10B(n,ag + aiYj^Li

This reaction is implemented by employing detectors

which will detect either the or particles (and

also possibly the ''Li nuclei). The energies of these

particles change with neutron energy thus causing some complication in the interpretation of data where the

neutron energy is a significant fraction of the Q value.

Historically this has limited the use of this cross section at higher neutron energies. The cross section

has been implemented with proportional counters, ionization chambers, solid-state detectors and boron scintillators. The use of each of these detectors will be

discussed below.

Proportional Counters

The simplicity of these detectors has led to ex- tensive use of these counters in neutron experiments.

Fig. 6 shows a typical simplified experimental set-up.

Typically the counter gas employed is ^°BF3. Neutrons

incident upon the counter can interact with and

the reaction products, ^Li and an a-particle, will

produce ionization in the counter as they come to rest. Consider an event which occurs near the center of the

counter. Then for low enough neutron energies the re- action products lose all their energy before striking

the walls. The total ionization will be closely proportional to the total energy of the reaction products,

Figure 6. Simplified experimental set-up for the use

of a ^°BF3 gas proportional counter.

(Q + Ep). Thus, aside from resolution effects, a sharp

peak should be observed for the ^°B(n,ao)^Li reaction

and a separate peak for the ^ °B(n ,017) 'Li reaction.

These two peaks will be separated in energy by 478 keV

since the 'Li gamma ray will not generally be detected.

If this entire counter were being irradiated with

neutrons, some of the reaction products would strike

the walls of the counter before losing all their energy

by ionization. This "wall effect" gives rise to a low energy tail in the pulse-height distribution. In Fig. 6 the neutron beam is collimated to a size smaller

than the diameter of the counter. If the beam diameter

is small enough (and the counter dimensions are great

enough), the wall effect can be removed. The reduction

in counting rate is very large for a moderate size

counter if the collimator is designed to entirely

eliminate the wall effect. The collimator also removes

the background resulting from direct beam neutrons

striking the walls of the counter. An added benefit

for linac measurements is that the gamma flash is

significantly reduced by this collimator since much

of this problem is a result of photons which interact

with the walls of the counter.

The time jitter of proportional counters is worse

than that of the other detectors under discussion. This

results from the fact that multiplication in these

counters only occurs very near the central anode wire.

So electrons must drift in a moderate field (i.e. slowelectron velocity) until they are essentially at the

center of the counter. Thus the time between the

reaction (and formation of ionization) and the formation

of the signal at the anode varies from zero for

ionization concentrated near the wire to a maximum for

ionization concentrated near the walls of the counter.

The collimator which was described previously has the

further advantage of reducing the time jitter by re- ducing the effective radius of the counter (note the

effective radius is somewhat larger than the diameter

of the collimator due to the reaction products which

are directed toward the walls of the counter). The

timing of the counter can also be improved by employing

various gas mixtures.

The proportional counter necessarily will have

effects associated with distortion of the field near

the ends of the counter (end effect) however, with

modern counter geometries having reasonable lengths,

this effect is small, aL/L -^1-2% (uncertainty in length/

length). In Fig. 7, the energy spectrum from a 5-cm

87
8000

<

>

SSL LJ^ 6000

>

O

(TUJZ

UJ 4000

cr

LiJ Q.

to2 2000 — Z)o

o

"III

'°B(n,ao)\i,

800 1600 2400

ENERGY DEPOSITED, keV

Figure 7. Pulse height distribution for a ^°BF3 gas

proportional counter for thermal neutrons.

4 diameter proportional counter exposed to thermal

neutrons is shown. The two groups associated with

i°B(n,a,Y)^Li and i°B(n,aQ)'^Li are well resolved. This

resolution (-^^5%) was made possible by operating the

counter with 10% ^°BF3 and 90% argon. The argon was

used to reduce wall effects by increasing the stopping

power of the gas. It also reduced the time jitter of

the counter. The long tail at low energies is a result

CO 8

>-oc

<f 4

m

oc 2

<

liJ 0

<

X 8

C_)

UJ 6

Q_

CO

r>o

O 2

20 cm

8% F.W.HM40 cm

30 cm

14% F.W.H.M -

317. F.W.H.M

60 cm

567. FW.H.M.

0 20 40 6 0 0 20 40 60

PULSE HEIGHT, ARBITRARY UNITS

Figure 8. Pulse height distributions for ^°BF3 gas

proportional counters for fillings of 20, 30, 40 and

60 cm Hg pressure for thermal neutrons.

of wall and end effects. The low ^^B content in this

detector limits its usefulness. Fig. 8 shows the effect

of increasing the pressure of ^"BFs in a gas proportional

counter. 5 The larger the pressure the poorer the

separation of the two groups. This results from electron

attachment to the ^^BFj molecules. However it is not

necessary to resolve the groups. All that is needed

is the total number of counts above a bias which is

appreciably above the noise pulses. Thus useful counters

can be made with one atmosphere of '°BF3. However

the higher pressures lead to poorer time jitter (See

Ref. 6). The rather poor timing is perhaps the worst

limitation in the use of these counters. As the

neutron energy increases the wall effect can become

important. Also at high neutron energies, pulses from

^°B nuclei which have been scattered by neutrons are

large enough so that the discrimination level must be

increased. Thus a possible systematic error in the

spectrum fraction can be introduced. The pulse-height

distribution for a gas proportional counter with

one atmosphere of ^°BF3 for neutron energies of about

500 keV is shown in Fig. 9. The low-energy pulses in

450 - E. ^ 550 keV

1600 2iiOO }200

PULSE HdCMT (VeV)

Figure 9. Pulse height distribution for a one atmosphere

i^BFa gas proportional counter for neutrons from 450 to 550 keV.

this Fig. are i°B recoils. These recoils are present

in all counters which detect charged particles.

The total systematic errors (mainly from room return

neutrons, spectrum fraction and end window scattering)

are in the 1 to 3% range for this type of detector.

This detector is most easily implemented in a cross- section measurement as the counter located further from

the source so that transmission corrections for the

constituents of the counter need not be applied to the

cross-section measuring counter. In white source experiments this reduces the effect of the time jitter

of the counter also.

88
-1800 V (FRISCH GRID) -2700 V ('°B FILM! -1800 V (FRISCH GRID)

Figure 10. Schematic layout of the Frisch gridded ion

chamber used by Friesenhahn.

Ionization Chambers

Ionization chambers containing ^"BFa or solid ^^B

deposits can be fabricated so that scattering from the

chamber is small and the transmission corrections for

the constituents of the chamber are easily calculated.

A ^°BF3 ungridded gas ionization chamber has been used

by Weston7 for low-neutron energies. A 20% ^^BFa + 80%

argon gas mixture is used to reduce the range of the

reaction products and to improve the timing of the

detector. A timing of 35 ns has been obtained. Summing of the reaction products occurs but for ungridded chambers the pulse size depends on the track

orientations of the a particle and ''Li ion. Thus the

pulse height distribution is very broad and the chamber

is limited in usefulness to low-neutron energies where

the pulse height distribution is essentially constant

with neutron energy (i.e., the spectrum fraction is

approximately constant). The use of gridded chambers

removes this problem.

Fig. 10 shows the layout for a detector with

Frisch grids used by Friesenhahn.-^ In this chamber

a self supporting ^°B film was employed which is a colloidal suspension of ^°B in a thin plastic film.

The thickness of the film is 200 pg/cm^. This is

thin enough so that both reaction products can escape

from the film. The, spacing between the film and the

Frisch grid is 2 cm which is slightly greater than

the range of the most energetic reaction products for

the counting gas (10% CO2 + 90% A). The electrons

are drawn through the Frisch grid to the collector,

all moving through the same potential difference to the collector. A fast pulse proportional to the

ionization appears between the grid and the collector.

Thus by summing the pulses at the two collectors shown

here, a pulse proportional to the Q value of the reaction plus the neutron energy is produced. The principal disadvantage of this detector is the low counting

rate due to the thin ^°B deposits. To increase the

counting rate, Friesenhahn employed eleven modules of

this type separated by suppressor grids to prevent

cross talk between adjacent modules. The complete ion

chamber is more than 1 m long and the frames are 25 cm

X 25 cm. It is the largest ion chamber of which I am aware. Fig. 11 shows the sum pulse height distribution

for this detector for neutron energies from 1 to 10 keV.

The greatest difficulty with the use of this detector

is the determination of the spectrum fraction. At low60 80

CHANNEL NUMBER Figure 11. The sum pulse height distribution for the

gridded ion chamber designed by Friesenhahn for neutron

energies from 1-10 keV.

2

1

2

J2

10

0.84 -MeV 'LI -PARTICLES

'°B(/I,oy) 'Li"[478 keV]

(.47- MeV a- PARTICLES

"

'°B(/7,oy ) [478 KeV]

200 300

PULSE HEIGHT CHANNEL

500

Figure 12. The pulse height distribution of the gridded

ion chamber designed by Peelle for neutron energies less

than 2 keV.

neutron energies, as in this Fig., the fraction of the

events below a reasonable pulse-height bias is quite

small. However, at high neutron energies the bias must

be increased to eliminate proton recoils from hydrogen

in the plastic binder. In principal the spectrum

fraction can be calculated however some of the input

data which are needed are not known accurately enough,

such as dE/dx of the film at low ion energies. The

calculation must take into account such effects as the

neutron kinetic energy, the branching ratio and the

angular distribution of the reaction products. The

spectrum fraction correction was 15% at 700 keV. The

systematic uncertainty with this detector is 2%.

89
In Fig. 12 the pulse-height distribution for a gridded ^^B ionization chamber designed by Peelle is

shown. This chamber was designed to be used with

gridded fission chambers for cross section measurements

on a linac. The timing of this detector should be

30 ns. This chamber contains thin deposits of

evaporated onto a mylar backing. Only one of the

reaction products is then observed except at very high

neutron energies where the incoming momentum of the

neutron in a fraction of the events permits both reaction products to go in the forward direction. Four

peaks are observed since there is no summing. These

spectra were obtained for neutron energies below 2 keV.

In normal practice the pulse-height bias would be placed

just above the 'Li particles. The pulse height

resolution is seen to be excellent. At high neutron

energies, however, the spectra are more complicated

than those shown in Fig. 12 as a result of kinematic

effects. Then the separation between the a particles

and ^Li ions is not as clear and corrections must be

made for the a particles lost below the bias. This is

not the most important complication, however. The low

count rate at high-neutron energies due to the thin film

is the real problem.

Solid-state Detectors

The reaction products from the i'^B(n ,a) ''Li reaction can also be detected with any charged-particl

e

detector. Solid-state detectors are convenient and

the demands on these detectors are quite minimal compared with the state of the art. The depletion layer

need not be very thick since the maximum energies of

the reaction products are relatively low. The energy

resolution of the detector can be fairly poor since

the thickness of the ^°B film will dictate the resolution of the system. It is preferable however to have

large area detectors to increase the solid angle for

counting rate considerations.

SOLID STATE DETECTOR

for FISSION FRAGMENTS

U DEPOSIT

n

BACKING MATERIAL

B DEPOSIT

P r-| SOLID STATE DETECTOR

for '°B(n,a) ^Li EVENTS

Figure 13. The detector geometry employed by Wagemans

and Deruytter for a measurement of the ^ssy fission

cross section relative to i°B{n,a)^Li.

Fig. 13 shows the.detector geometry employed by

Wagemans and Deruytter for measurements of the ^^^U

fission cross section relative to the ^°B(n,a)''Li

standard. The solid-state detectors are shielded from

the direct neutron beam by a collimator. Fundamentally

this is the same geometry as that involved in a double

ionization chamber set-up. The nearly equivalent

environments for the two films reduce the systematic

errors in the cross section measurement. This chamber

is evacuated so that the reaction products do not lose

energy when passing from the foil to the detector.

For the geometry of Fig. 13 the reaction products

must pass through a thicker deposit of material than

the areal density before they are detected in the

solid-state detector. This worsens the pulse-height

resolution of the system relative to an ion chamber of

equivalent areal density. Also the solid angle must

be determined with the solid-state detector system

whereas an ion chamber detects all reaction products

which are emitted from the foil which have energies

greater than the bias energy. It is then obvious that

the counting rate will be higher for the ion chamber

than for a solid-state detector system of equivalent

areal density. The limited angular range over which

particles are detected for the solid state detector has one virtue. The kinematic effects which worsen the

separation of the particle groups are reduced if the

range of angles for which particles are detected is

1 i mi ted.

CHANNEL NUMBER

Figure 14. Pulse height distribution for the solidstate detector system employed by Wagemans and Deruytter.

In Fig. 14 the pulse height distribution observed

by Wagemans and Deruytter is shown. This distribution

should be compared with that observed by Peelle (Fig.

12). The resolution of the solid-state detector system

is much worse due to the greater effective thickness of

the i"B film. The measurements of Wagemans and

Deruytter only extend to 30 keV so the shape of the

pulse height distribution should be essentially the same for the full energy region.

The timing with solid-state detectors is very good

(in the ns region). In particular, for this experiment

it was not a limitation to the time resolution. The

solid angle for the solid state detector can be deduced

by calculation or by replacing the ^"B film with a uniform calibrated alpha source having the same area

as the ^°B deposit. In the Wagemans experiment the

systematic errors in the use of the i°B(n,a)^Li reaction

are small , i 1%.

90
Scintillators

In the late 1950's and early 1960's a considerable

effort was expended looking for a satisfactory method

for implementing scintillators containing ^^B. Good

timing should be possible with scintillator-photomultiplier tube combinations. The kinetic energy of

the reaction products would be summed and should produce a convenient response function. The objective

was to obtain high concentrations of i°B so that high

efficiency could be achieved. For some uses high con- centrations of i°B would allow smaller scintillator

thicknesses so that the multiple scattering corrections

(and therefore their uncertainties) could be reduced.

Scintillators which are loaded with i°B compounds

unfortunately poison the scintillator so that the

greater the concentration of i°B the lower the pulse

height from i°B(n,a)^Li reactions. In Fig. 15 the

B CONC, %

Figure 15. Dependence of the light output and light

absorption on boron concentration for a boron-loaded

plastic scintillator.

light output^^ as a function of boron concentration

for a plastic scintillator is shown. To obtain usable

efficiencies the pulses were so small that photomultiplier tube noise was an important factor. Various

techniques were employed to handle these problems.

For example, Bollingerll coupled two photomul ti pi ier

tubes to one scintillator and by requiring a coincidence

between the pulses from these tubes, he significantly

reduced the noise rate. Also cooled photomul ti pi ier

tubes have been employed to reduce the noise rate. Even pulse-shape discrimination has been used; this has

the virtue of reducing noise and discriminating against

Y-ray events in the scintillator.

Unfortunately these methods are complicated and/

or awkward to implement. Thomas^^ fabricated a i°B

loaded liquid scintillator assembly coupled to a single

photomultiplier tube which is usable without any of

the special refinements mentioned above. The pulseheight distribution for that detector is shown in Fig.

16. The pulse-height resolution is about 50% and

photomultiplier tube noise is still a small problem;

but, this represents the best results obtained to date

with i°B scintillators without the special techniques

mentioned above.

Though in the past i°B has been loaded into liquid,

glass and plastic scintillators and a number of different experimental techniques have been tried, very

little work has been done recently on ^^B scintillators.

Some of the problems noted by Bollinger^ were a result

of non-uniform efficiency of photocathodes for the

photomultiplier tubes employed. It seems reasonable

that with the improvements in photomultiplier tubes

and electronics in general in the past 15 years and the

possibility of improved scintillators, more work in

this general area could yield a satisfactory detector.

4000 1—s

1 r

Figure 16. Pulse height distribution of a boron

loaded liquid scintillator for thermal neutrons.

Concl usion

A number of techniques are available at the present

time for implementing the iOB(n,a)7Li cross sections

to an accuracy satisfactory for most cross section

measurements. Further developmental work may improve

the utilization of this standard. Improvement of the

standard cross sections themselves is needed at the

higher neutron energies, however, the rapid fall of

the cross sections above '^-500 keV may introduce

fundamental limitations to the use of this cross section for some experiments.

91
References

1. M. S. Coates, G. J. Hunt, C. A. Uttley and E. R. Rae, Proc. Symp. Neutron Standards and Flux

Normalization, CONF-701002, p. 401, U.S.A.E.C.

(1971).

2. R. A. Schrack, G. P. Lamaze, and 0. A. Wasson, to be published.

3. S. J. Friesenhahn, A. D. Carlson, V. J. Orphan

and M. P. Fricke, "Measurements of the

^°B(n,aw) and i°B{n,a) Cross Sections,"

U.S.A.E.C. Report Gul f-RT-Al 221 0 (1972).

4. G. P. Lamaze, A. D. Carlson and M. M. Meier,

Nucl. Sci. Eng. 56 (1975) 94.

5. I. L. Fowler, Rev. Sci. Instrum. 34 (1963) 731.

6. 0. K. Harling, Nucl. Instrum. Methods 34_ (1965)

141.

7. L. W. Weston, private communication (1977).

8. R. W. Peelle, L. W. Weston, R. W. Ingle, J. H. Todd

and F. E. Gillespie, Neutron Phys. Div. Ann. Progr.

Rep. May 31, 1972, QRNL-4800, p. 10 (1972). and

R. W. Peelle, private communication (1977).

9. C. Wagemans and A. J. Deruytter, Annals of Nucl.

Energy 3 (1976) 437.

10. 6. I. Anisimova, L. S. Danelyan, A. F. Zhigach,

V. R. Lazarenko, V. N. Siryatskaya and P. Z. Sorokin, Pribory i Tekhnika, Eksperimenta 1_ (1969)

49.

11. L. M. Bollinger and G. E, Thomas, Rev. Sci.

Instrum. 28 (1957) 489 and L. M. Bollinger and

G. E. Thomas, Nucl. Instrum. Methods 17_ (1962) 97.

12. G. E. Thomas, Nucl. Instrum. Methods 17_ (1962) 137.

92
EVALUATION AND USE OF CARBON AS A STANDARD . J . C . Lachkar

Service de Physique Nucleaire - Centre d 'Etudes de Bruyeres-le-Chatel

B.P. n° 561 - 92542 Montrouge-Cedex, France.

Available data on the carbon total and elastic scattering cross sections are reviewed.

Major emphasis is placed on two neutron- energy regions below 5 MeV and between 8,5 and 15 MeV,

The overall consistency of the recommended values has been established with the aid of previously performed theoretical analyses. It is concluded that carbon elastic data below 5 MeV can

be adopted as a standard except at the location of resonances. It is also suggested that the

present need for high energy neutron standards could be satisfied by carbon data.

(Standard ; evaluation ; carbon ; total cross section ; optical model)

.

Introduction

Carbon cross-section data are of interest in a number of areas of applied physics. Detailed knowledge

of differential cross sections for neutrons scattered

by carbon is required for calculations of neutron

transport in materials for fusion and fission reactors.

These needs have justified a large number of measurements, the earliest ones being performed 30 years ago.

Since reasonably accurate sets of data were available,

neutron data on carbon have been proposed as standards,

garbon is a good candidate since it is a light nucleus

and then the level density of the compound n + carbon

system is small ; also the threshold for inelastic scat- tering is high. In 19T0, at the Argonne Symposium on Neutron Standards and Flux Normalization, it was suggested that carbon is suitable as a transmission standard up to 1.5 MeV but is marginal in angular distribution applications Since that time new measurements and more complete analyses have been performed in

several laboratories ; consecjuently it has been conjectured that , by including all the presently available

experimental informations on the total cross section,

differential elastic cross sections and polarization in

a R-matrix fit , a complete and consistent set of evaluated data could be generated up to 5 MeV '

'

"* . At higher

energies, where the presence of resonances in the ^ 'C compound nucleus and competing reactions do not allow

carbon to be considered as a standard scatterer, differential scattering cross sections for carbon are

still of interest. In scattering experiments, absolute

normalization is generally made by replacing the sample

by a polyethylene scatterer : moreover, emitted neutrons usually are detected using plastic scintillators.

Thus neutron data for carbon, are needed for those two

parts of the measurements. This need is more and more pronounced as the energy Increases since the relative

importance of the neutron data of carbon to the (n,p)

scattering data increases with energy as shown in

fig.l. In addition it is likely that a measurement of

scattering from a secondary standard such as carbon,

from which good samples are easy to prepare, would

allow easier comparison of data obtained at various

laboratories

.

Also, neutron induced reactions on ^^C produce

mainly one Y~J^8-y corresponding to the de-excitation of

its first excited level at k.hk MeV. This line may be

a good candidate to determine the detector efficiency

at high energy or to be a reference for y-vay production cross sections for neutron induced reactions at high energy ^

.

Finally the neutron cross sections for carbon

exhibit several well defined and sharp resonances and

thus carbon may be considered as an energy-scale standard in both white and monoenergetic source measurement s

^

' *

. Here, we sha±l review the data on the carbon

total and elastic scattering cross sections. We shall

lastic scattering cross section ; R-function analysis ;

EleV)

Fig.l. Energy dependence of the carbon total and elastic cross sections compared to the H(n,p) scattering

cross section.

concentrate on the most recent available data , mainly

in the low-energy range. Contributions from '^C also

will be considered along with radiative capture cross

sections. We shall then discuss the theoretical models

used in the analysis of the data. By comparing theoretical values to the experimental ones we shall give so- me error estimates.

Neutron-induced reactions on carbon

The Q-values and thresholds of the neutron-induced reactions corresponding to the n + ^^C system are derived from ref.^. The ^^C(n,Y) reaction has a Q value

of l4..9^7 MeV. The threshold energy for the inelastic

scattering is U.8l2 MeV.

Natural carbon consists of 98.892/J of ^^C and

1.1 08^ ^'c, so we have to consider the contribution of

the '^C content. Neutron data for ^ ^C have to be taken

into account if a precision of 0,5? or better is required. The (n,Y) reaction for ^'c has a Q value of 8. ITT MeV and the threshold energy for the inelastic scattering is 3.323 MeV.

Analysis of the data

The analysis of the data will be divided into

three parts. In the first one, from the thermal value

up to 2.0-MeV neutron energy, the total cross section

for '^C exhibits no resonance structure. The second

part extends from 2.0 MeV uptol+.8 MeV which corresponds

to the threshold of inelastic scattering for ^^C. The

93
third one is up to 15 MeV. For each part, total and

differential elastic scattering cross sections will be

discussed

.

From 0 to 2 MeV neutron energy. The total cross section for natural carbon has

been measured and evaluated by many groups ; we have

analysed most of these studies and used the data to propose evaluated total cross sections ^

. Our recom- mended values are shown in fig. 2. The most complete

data considered in our analysis are listed in table 1

,

They can be seen in fig. 3 which plots versus neutron

energy the percent difference between each data set and

our recommended values. The data of Cier jacks et al. ^ from KFK are characterized by a good energy resolution

and a precise energy calibration ; however they have

been found to be higher by about 1^. On the other hand,

the data from RPI ^ " seem to be low by about 1 .5^-

Finally nearly all the data from UBS . \ Harwell ' ^

, ORNL and ANL ^

"* lie in a band of total width approximately ^% of the average cross section.

The thermal value of the total cross section has

been evaluated by Leonard et al. and Story et al. ^ These two evaluations were based on all the available

carbon thermal cross section measurements from the year

19^6 to 1970. Their adopted values are given in table 2. More recently, Mughabghab and Garber have proposed

the value of ^+.750 ± 0,02 barn. This value is very

consistent with that (^.756 ± 0.036 b) deduced by

Heaton et al. ^ from their data above 1 keV by as- suming that the total cross section is constant bet- ween 10 keV and thermal energy. Our recommended value,

U.728 + 0.008 b, is consistent with all the previous

ones

.

T 1 1 1 1 1 1 r T r

Tbanul «ilue

io* 10' 10^ io* io' io^ 10 • 10 » lo' To'

E,(.V

Fig. 2, Reaormended aarbon total cross section below

2 MeV.

Between 10 eV and 2.0 MeV, the variation of the

total cross section with neutron energy has been ex- pressed by a polynomial in energy which was fitted to

the averaged data by minimizing x are given by the expression:

a(E) = i;.725 - 3.251 E + 1.316 E - 0.227 E-' (barn

where E is the neutron energy in MeV. The accuracy of

the total cross section is believed to be 0,5^ below

1 keV and less then 1^ up to 2 MeV. Such an analytical

expansion does not take into account resonances in ''*C which will be discussed below.

Various and detailed measurements on differential elastic cross sections have been reported below

2 MeV. Most of them, listed in table 3, have been presented or discussed at the Argonne Symposium on Neutron-Standards and Flux Normalization in 1970. All

these data have been also considered to propose

The proposed values

3

TABLE 1 Available total cross section d ata.

Authors Year Lab Energy-range

(MeV) Remarks Ref

Bockelraan 1951 WIS 1 .25-3.35 exp. data 20

Fossan 1961 WIS 3.30-16 II 23

Uttley 1968 HAR 7. 10~^-1 .5 II 12 Cierjacks 1968 HFK 0.3-30 II 9 Clements 1970 RPI 0.9-30 II 10 Foster 1971 BNW 2.3-15 If 22

Nishimura 1971 JAE 10~^-0.3 II 46

Percy 1972 ORL 0.2-20 II 13 Meadows 1970 ANL 0.5 - 1.5 II

1

4

Holt 1975 ANL 1.5 -5.0 11 18 Heaton 1975 NBS 1 O"-^- 1

5

II 1

1

Auchampaugh 1976 LAS 1.5- 14 preliminary data 35

Francis

Nishimura

1970

197!

KAP

JAE

10-'0-l5

« 2

evaluated data

II

1

46

Perey 197A ORL 10-'0-20 ENDF B/IV 19 Lachkar 1975 BRC 10~'° - 20 evaluated data 8

on

tu

~~1

I I I

1 1—I—1—I—

a NBS ®KFK *HAR «RP| *ORL .ANL

0

© *

' Q B

11 0 *

+

®

» ax'

* • B

e.9'§8**fi ,Sb-b .e.

• • ®S a ^ *

• • S -

' b0»« © • ••

-L X X X X

02 0.<. 0.6 OJB to 12

E (MeV)

1.4 16 1.8 2.0

Fig. 3. Ratios of carbon a total experimental data to recommended curve. NBS data are from ref. ^ , KFK from

ref.^, EAR from ref. RPI from ref. , ORL from

ref. ^ ' and ANL from ref. ^ "

.

TABLE 2

Thermal neutron total cross section for carbon

Authors Year Recommended Oj for thermal energy Ref

Leonard 1970 4. 730 ± 0.008 b 15 Story 1970 4. 723 ± 0.0125 b 16

Mughabghab 1973 4. 750 + 0.02 b 17 Lachkar 1975 4. 728 ± 0.008 b 8

94
TABLE 3 Summary of neutron elastic scattering measurements from 0.05 MeV up to 15 MeV.

Authors Year Lab Ref Neutron-Energy range

(MeV)

Number of measured

energies

Measurement

angles (deg) Remarks

Lane ANL 4 / 0. 05 - 2 . 20 56 25 - 1 45 Ang. Distr.

Lane 1 QAQ AMT A9HZ r\U ,

1

1 nU — /onT U Ang . Distr

.

Pol

.

11 Q7n J / u *4 O U.J z . u 90ZU — 1

1 AnDU Ang . Distr

.

Wills 1 958 ORL 27 1 . 45 - 4 . 1

0

1

2

35 - 1 45 ring . Ul S L r . no i L

1 Q7 A MT 1 R 1.8 -4.0 1 z on 1 An Ang. Distr.

LLC jLiZL T ATT Z

9 AD 2.2 -3.8 QO — 1 43 ti Galati 1972 KTY 24 3.0 - 7.0 32 1 5 - 1 60 ?i Knox 1973 OHO 39 2.63 1 17-118 Ang. Distr

.

Pol.

£* a so 1

X

1 Q7 T 2.1 -4.7 9ZUn _ 1

1 Andu Ang .Distr

.

i O L a i A sect

.

Perey 1969 ORL 28 4.5 - 8.5 13 15 - 140 Ang. Distr

.

5.2 - 8.7 40 15-140

Velkley 1973 ABD 29 7.0- 9.0 5 10 - 150 M

Haouat 1974 BRC 30 8.5 - 11.0 4 30 - 160 II 8.0- 14.5 14 10-160 M Glasgow 1976 DKE 31 9- 15 14 25-160 It

evaluated data in ENDF/B IV

^ ^ . The new results since

that time are from ANL where Holt et ^1. ^ have measured angular distributions at 1.8 and 2.0 MeV. Their

data have been found consistent with the recommended

angular distributions from ENDF/E Vj except at forward

angles. This deviation might suggest that some changes

would he necessary in the existing evaluations at least

above 1 . 8 MeV

.

Above 0.1 eV the integrated elastic scattering

cross section can be taken equal to the total cross

section. This results in a (n.y) cross section that is

too small. The deviations between the integrated elastic cross section, determined from polynomial fits to the measured angular distrihutions , and the recommended

total cross section are less than the experimental un- certainties which are about 3^. Differential cross sec- tion are generally quoted with a precision of less then

5%. From 2- to 4.8-MeV neutron energy.

At the neutron energy of 2.077 ± 0.002 MeV, the

^^C + n reaction reaches the resonance at 6.863 MeV in

^'C ; the total width of this resonance is 6 keV and

the value of the total cross section at the peak is

6.020 b taken from ref. The narrowness of this resonance implies that a good energy resolution is needed in the vicinity of 2.077 MeV. This may be a severe constraint in the extension of the use of carbon as a standard above 2 MeV.

Between 2.1 and 2.7 MeV, the total cross sec- tion show no resonance (fig.l+) and, here , carbon still

could be used as a standard. In this energy range,three

sets of data from NBS , ORNL and the older ones

from the University of Wisconsin ^° have been found

very consistent. The accuracy is of the same order as below 2 MeV, i.e. not greater than \% . Recently reporr

ted data from ANL were also found to be in good

agreement. The major interest in these measurements

lies in a very good energy resolution which was obtained with the aid of monoenergetic source techniques.

At Ejj = 2.8l6 + O.OOU MeV, most of the recent

data exhibit a maximum associated with the resonance

at 7.5^+5 MeV in ^'c. This resonance was ignored in the

evaluation file ENDF/B III but was included in the

ENDF/B rv file. The total width of less than 5 keV reported by Ajzenberg-Selove for this resonance is

consistent with the data of ref. 9,ii,i3,i8_

- evaluated

2.5

Fig. 4. Comparison of neutron total cross sections of

carbon between 2 and S MeV (a = ref. \ b = ref.

0 = re, ref.

ref. the evaluated curve is from

Large discrepancies appear at the 3.05 MeV interference dip due to differences in the resolution

and energy calibration precision. Our adopted values

were based on the data of KFK ' and ORNL ^ ^

. The results of Foster and Glasgow present a significant

shift to lower energies and those from Fossan et al.^^ are subtantially shifted to higher energies. Above

3.05 MeV up to U.8 MeV no other fine structure is ex- pected and only some smooth variations such as the

broad 3.58 MeV resonance and the l+,26l MeV one are en- countered. The proposed values, obtained by smoothing

the composite results of ref. '^-^ j-* ^ in the energy

range 2,8U-U.8 MeV are given with 2% accuracy.

Three recent sets of elastic scattering data are available from the University of Kentucky , from the University of Padoua^ ^ and from ANL ^^.They are higher accuracy

repetitions of older ones from Meier et al.^^ and Wills et

al.^'^. Average relative uncertainties of the measured differential cross sections vary from 5 to 7 ^. The adopted

values are then quoted with an accuracy of less than " %. The angle-integrated cross sections deviate from

.the recommended total cross sections by 3 to 5 %

95
High resolution neutron scattering experiments

are in progress at ORELA from 2 to 3.2 MeV but the data

are not yet available

.

From 4.8 to 15 MeV . Between U.8 and 8.5 MeV neutron energies as many as 12 resonances in '^C are reached, some of them

being good candidates for energy-scale standards. But,

for our purposes, they make this energy range of limited

value. In addition two exit channels are open by the

inelastic scattering and the (n,a) reaction. The data

from KFK ^ , NBS 1

1

ORNL ' and University of Wisconsin's^

which are in good agreement for the resonances energies have been considered. Between the sharp resonances

the recommended total cross section values are very

close to those of Heaton et al . and the uncertainty

is about 2%, The values at the resonances are taken

from Perey et al . and the accuracy is better than

Above U.8 MeV and below 8,5 MeV, elastic angular

distributions a(G) have been extensively measured mainly by Galati et al. up to 7 MeV, by Perey et al.

from U.5 to 8.7 MeV and by Velkley et al.^^ above 7.2

MeV. The data expressed in the center of mass system

were fitted to a Legendre polynomial expansion using the

least sq.uares method. The coefficients are plotted versus energy in fig. 5, they show large variations at the

location of the various structures observed in the total cross section.

Fig. 5. Reoormended Legendre ooeffiaients for elastic saatteving below 15 MeV. The coefficients are defined by :

0.6 04

02

0.0

.; ij.'ii

0.4

00

0-2

0.1

00

008

004

000

F2

0.6

04

02

00 -.i- F3

Fs

004

002

000

Fe

0 4 8 12

En (MeV)

In the region extending from 8.5 to 15 MeV the resonances reached in the ^'C compound system become

broader than in the lower energy range. Moreover they

are generally overlapping so that the cross^ section

variations with neutron energy are rather smooth. The

data from KFK \ NBS ORNL ^\ BNW and from the

University Of Wisconsin are in satisfactory agreement. Nearly all of them lie in a band of total width

approximately Q% of the average cross section. The

recommended values are very close to those from ref.^^

and they are given with h% accuracy.

In this energy range, differential elastic

scattering cross sections are well known. Evaluated

data have been mainly based upon the time of flight

measurements performed at Bruyeres-le-Chatel by Haouat

et al. between 8 to lU.5 MeV. Two independant sets

of data were obtained in two measurements periods and

they were found consistent with each other. The norma- lization used was the n-p scattering cross section near

0°, Glasgow et al. have measured the elastic angular

distributions between 9 and 15 MeV. Their data are in

overall good agreement with those from ref. All

these data compare favorably with the previous measurements below 9 MeV from Perey et al. and Velkley et al. . The integrated elastic cross sections are given

with an accuracy varying from 6 to Q% . The adopted values are then taken within a global uncertainty of

less than 7^. The comparison of these values with those

from ENDF/B III and ENDF/B IV shows significant deviations by about 7 to 15^. The recommended values from

the previous evaluations are too high at about 11.0

MeV and systematically low between 12 and 1^.5 MeV.

The evaluated Legendre coefficients of the elastic angular distributions have been deduced with a rea- sonable accuracy due to the overall consistency of the

avarlable data. The degree of confidence in the evaluated data was increased when they were compared to the

data of Bucher et al . from forward angle elastic

scattering measurements and those of Morgan et al.

from backward angle measurements made at ORELA.

Reactions in 13,

Cotnpeting reactions

Now we have to consider the contribution of the

^ .^% of ^^C present in natural carbon.

The existing data of Cohn et al. from 0.112

to 22.8 MeV and those of Auchampaugh et al . from

1.5 to ^k MeV show that the cross section of this isotope is very close to the corresponding one for ^^C

except in the vicinity of the "'''C resonances observed

at neutron energies of 0.153, 1.751 , 2.U32 and 2.1+51+ MeV 1 7

. The first one at 1 53 keV has been observed m

1

1

the total cross section measured by Heaton et al.

using natural carbon sample. This is mainly due to the

large value of the total cross section (a = 21 ± 2 b)

and to the small width of the resonance (r = 3.7 ± 0.7 keV). This example shows clearly that good '''c data, including differential elastic cross sections

which have never been measured, are needed for a good

knowledge of carbon data.

Radiative capture in carbon

The recommended values for the (n,Y) cross sec- tion for thermal neutrons are for the two isotopes of

natural carbon ; ^^C a(n,Y)th = 3-36 ± 0.3 mb

^^C a(n,Y)th = 0-9 ± 0.2 mb.

These recommended values taken from ref._

have been strongly influenced by the data of Jurney

and Motz ^ ^ . At low energy, i.e. E^ < 100 keV, the

(n,Y) cross section has been assumed to vary with neutron energy according to the 1 /v law.

The radiative capture is the inverse of the

nuclear photo-effect ; by applying the reciprocity

96
theorem, the (n,Y) cross sections can be deduced from

the (y-n) data. The excitation function for the '^C

(Y, n) reaction has been measured by Cook using

these data we have obtained the (n.y) cross sections

from 0.2 to 20 MeV ®. The radiative capture cross sec- tion from thermal energy up to 20 MeV is very low and

does not exceed O.OT^ of the total cross section.

Theoretical analysis

In order to confirm the overall consistency of

the evaluated values deduced from the review of all

the available experimental data a theoretical analysis

is required. Such an analysis would lead to accurate

knowledge of the scattering phase shifts.

The phase-shift analysis of differential cross

' sections can be performed in a rather simple way since

the ground- state spin of ^^C is 0+ and then the chanpel spin has only the value S = 5/2+. In this formalism,

the center of mass differential cross section can be

written as

3 8

where

f =1

/=0

P; (cos

da

dfi

)[(^

(6)

+ 1 )e

' 1

f

.

1 sin e X Pj' (cos 9) I s /=0

sin 5^ + -le "'"'^'^sin ,

|^i(Sj sin 6| ^. e^*^-^ sin ^ijiS| = (S (^, J =^ ± 1/2) is the phase-shift of the partial wave of total angular momentum J = £± 1/2,

P|f(cos 9) and P'^(eos 9). are the ^'^h - order Legendre

polynomial and its derivative, \ is the incident wave

length.

For E below U,8 MeV when only the elastic scattering channel is open, if the radiative capture process is supposed negligible, the phase shift can be

expressed as the sum of two real quantities, the first

one being the contribution from hard-sphere scattering,

the second being due to scattering from compound - nucleus resonances. When nonelastic channels are open

the phase shifts become complex quantities.

Phase-shift analyses have been carried out by

several groups. Some sets of parameters were obtained

by fitting a limited set of measured angular distribution data and the others were obtained by

simultaneously fitting the total cross section, the

differential elastic scattering cross sections and polarization data 1 9, 2't, 2 5, 3 9, "tO Q^gj, large energy

range. In addition to direct fits, R-matrix calculations and coupled channel calculations were used to derive the phase shifts.

R-matrix calculations .

In the' low energy part below I4.8 MeV, extensive

R-matrix calculations have been performed at Ohio

University at Yale University and ANL ^ and at University of Padoua Some other calculations are

in progress at LASL and at ORWL, In this energy range,

the application of R-matrix theory is relatively easy

since the R-matrix reduces to a R function. This function has an explicit energy dependence given by :

R„ -^^^

,2

1

A

E. + R

is the reduced width and E, ^ ^ is the level energy.

The s\im over X denotes the number of levels which are explicity considered, while RJj is the contribution to the R-function from distant levels. This last contribu-

• 3 9/ tion has been treated either as a constant iR^j =

jj-eJ

) Qj. expanded as follows :

R . Rf

All the levels in ^^C up to an excitation energy of

about 10 MeV except the ones at 0 and at 7.5^+5 i^ey

have been considered in the various analyses. The scat- tering in carbon below 2-MeV neutron energy, is

strongly influenced by the -1.86-MeV, 1/2+, state. Two

other bound states have been included specifically •. the -1.09-MeV, 5/2"*", state in the calculation of ref.

and the -1 .27 -MeV, 3/2~, state in the calculations

of ref. and The two strongly interfering 3/2.'*' states at 2.95 and 3.58 MeV influence both the cross sections and polarizations.

The R-function is defined from the boundary

condition for the radial part of the wave function

ug(r) at the channel radius a :

1 duL

dr 1 + B, ) / R . l3 I

J

The B^j quantity is the boundary condition factor chosen to cancel the level shift factor at the resonance energy.

In the different analyses, two distinct values were adopted for the channel radius a. Holt et al.

have taken a equal to the interaction radius which is

h.Gl fm. Then they determined the reduced width of the

Si/2 t'ound state at - 1.86 MeV by fitting the expected

cross section to the data below 2 MeV. They found

Yq 1 /2 - 0.69 MeV. The P3/2 bound state reduced width

was deduced from polarization data. On the other hand.

Lane et al. have assumed that low-energy data 8,re dominated by the contribution from only the above mentioned Si/2 bound state. The s-wave phase shift was obtained from the analysis of the data for a(9) below

0.8 MeV. The deduced parameters were a = 3.72 fm , /2 = - 6.0 MeV, Yq 1/2"^ • '^^^ ^^^'^ large and close to the single particle estimate which

is h.Q'J MeV In this analysis other fits also led

to higher values of a as large as h.9 fm.

The comparison of the adopted R-function parameters derived from these analyses is given in tkble

It. Although they are different, up to 3 MeV they give

^equally good fits to all the data.

The corresponding phase shifts are given by the

relation :

6^j(E) = - 0^+ arctan[p^ R^/ l-(S^-B^j) R^j ] •

where (3^, P^ and S^are hard-sphere phase shift, penetration factor and level-shift factor respectively.

Such a description provides a simple tool for predicting neutron scattering data.

Phase-shift analysis up to 7 MeV ., Phase shift analysis of the elastic scattering

data from 3 to 7 MeV has been carried out by Galati

et al.^"*. The real part of the phase shifts was deduced at 32 energies. In the region of overlap they are

in good agreement with those from ref. . The data

show large variations at the location of the resonances at3.58 MeV (d3/2), ^+.26 MeV (pi/2), 5-37 MeV

(P3/2), 6.29 MeV (f7/2) and 6.56 MeV (sj/a). The

2.95-MeV, resonance which interferes strongly with

that at 3. 58 MeV was not considered. In addition this

analysis led to an ambiguous assigment for the exited

state of ^^C at 9-52 MeV ( it . 936-MeV resonance). Above

lj-.8 MeV the phase shift is a complex quantity and the

imaginary part is needed. Except at few energies,

these parameters could not be determined with a suffix

cient accuracy to show non-zero values.

This analysis was characterized by an excel97
lent agreement with the measured angular distributions

(within the experimental uncertainties) and with the

total cross section between 3 and ^+.35 MeV and from

h.J to 5'^ MeV. Also polarization were calculated from

the derived phase shift and compared to the existing

data. The predicted values agree fairly well with most

of the polarization measurements below h.2 MeV.

Coupled channel calculations at low energy (below

5 MeV) The low-energy resonances in the nucleon -^^C

interaction can be interpreted as single-particle

states and doorway states originated from the coupling

between the incoming nucleon and the low-lying target

levels . Calculations have been made in the frame of a phenomenological collective description of the target.

Coupled channel calculations using a deformed shell

model for the ^^C nucleus were performed by Reynolds

et al. and Leung and Koshel '*^. Emphasis was placed by the former on calculations of neutrons cross sections. The method used there and the results were presented at the Argonne Symposium on Neutron Standards and Flux Normalization and thus they will be

mentioned just for completeness.

From their coupled-channel calculations Reynolds

et al . were able, with suitable potential wells, to

accoxint for all the observed resonances below 5 MeV

except for the 2,82 MeV one. The measured total cross section was well represented at all energies between

0 and 5 MeV, Also the calculated differential elastic cross sections agreed quite well with the data throughout the entire energy region. Finally, the agreement between calculated polarizations and the data

was good up to h.2 MeV but above this energy large

discrepancies appeared.

TABLE 4

R - function parameters

Levels in C ANl'^ a = 4.61 fm OHIO"^^ a = 3.72 fm

E ex keV

^lab

keV

E

cln keV

h

keV keV

R

o fMpV •> keV keV

R

0

s 1/2 3086 - 1861 - 1860 690 0 0.40 (0) 0 - 1860 4000 - 1 .035 0.075

P 1/2

0

8879 4261

- 4947

3932 4200 700 - 0.2 0. 16 0.02 391 1 83.4 - 0.293 0. 1

p3/2

3680

9503

9900

4936

5367

- 1270

4556

4954

- 1270 500 - 0. 18 0.34 0.055

4560

4940

210.3

- 0.247

- 0.247

0.25

0.25

d 3/2

7665

8250

2946

3580

2718

3300

2920

3500

170

870

- 1 .00

- 1.00

0. 15 0.15

0.02

0.02

2734

3537

212

2500

- 1 .368

- 1 .368

0. 107

0. 107

d 5/2

3854

6863 2077

- 1093

1915 2076 14 - 1 .30 0.05 0.0

- 1093

1959

1800

75

-2.213

-2.213

0.0

0.0

f 5/2 7545 2816 2598

TABLE 5 Potential parameters used in coupled channel calculations above 8.5 MeV.

Ref

Real Part Imag;inary part Spin Orbit part Deformation

Parameter

\ ^R R

so

a so V

so

44 1 .25 0.40 51 .50- 0.3 E 1 .25 0.30 0,74 + 0.38 (E - e5.5) 1 .25 0.40 5.0 - 0.60

45 1 .25 0.35 58 - 0.9E 1 .25 0.20 0.28 E 1 .25 0.35 5.0 - 0.64

R and a are in fm, V and E are in MeV.

Coupled-channel calculations at high energy (above

8.5 MeV )At neutron energies above 8.5 MeV, coupledchannel calculations can be used to predict elastic

and inelastic angular distributions. In this energy

range the nuclear structirre effects may be assumed to

be sufficiently averaged and thus described conveniently within the framework of the optical model. Several

attempts have been made to find a set of monotonically

energy-dependent potential parameters that would describe at least the overall features of neutron-carbon

elastic cross sections above 8.5 MeV. Differential

cross sections from ref. 30. and \3 have been analysed

in 1 MeV steps by Delaroche . The potential parameters deduced from this analysis are given in table 5 and compared to those obtained at Duke University . The potential parameters are similar to those found

to describe the nucleon scattering from heavier nuclei

with one exception ; the small diffusenesses , Such

small values were also used by Reynolds et al.'*" at low energy. Further improvements could be made by

adding Breit-Wigner resonance amplitudes to the optical-model scattering amplitude.

98
Conclusion References

In conclusion, it appears that accurate neutron

data for carton can be obtained from comparison of se- veral experimental values. The best accuracy is attained below the threshold of the inelastic scattering

for ^^C except at the location of resonances which are

of little interest for our purposes. For the total

cross section the uncertainties vary from 0.5% below

1 keV up to 2% at U.8 MeV. A large number of elastic

scattering data have been reported and the recommended

values which are derived are given with 1% accuracy.

This accuracy could be improved by considering the

last data from ANL in 1.8 MeV region. Accurate polari-;

zation data also cover the whole energy range.

The consistency of all the data below 5 MeV

has been confirmed by theoretical analyses based on R-matrix calculations or coupled channel calculations.

It has been shown that very accurate phase shifts have

been derived by fitting simultaneously the total, differential, and polarization cross sections. The reason

is that the total cross section gives rather weak con- traints on the parameters while differential elastic

cross sections and polarization data are sensitive to

small phase shifts in addition to the dominant ones because of interference effects.

A source of difficulty for the coupled-channel

calculations in predicting the observed neutron polarization from neutron elastic scattering occurs above

h.h MeV. Good agreement with measured polarization is

found from R-matrix calculations throughout the entire

region and hence these calculations are prefered for

neutron cross section prediction. Values of predicted by the R-function calculations performed concur- rently in two laboratories agree with the recoimnended

values to within less than 2 ^. An overall agreement,

within the experimental uncertainties^, is obtained for

elastic angular distributions by using mainly the ANL

parameters

.

We may definitely conclude that carbon elastic

data below 5 MeV can be adopted as a standard. Evaluated data could be achieved by a R-matrix analysis

including all the observed resonances. It is likely

that no precision improvements could be attained by

further measurements on '^C.As a consequence, the ana- lyzing power '^C is proposed as a calibration standard for neutron polarization studies which is technically more convenient to use than '*He. The discussion of carbon data above 8.5 MeV is

warranted by the need for high energy neutron standards. If the accuracies are not as good as at lower

energy, the overall consistency of the data which

could be confirmed by further theoretical analyses

would suggest that the present lack of adequate neutron standards in this energy range can be fulfilled

by carbon data.

Acknowl edgemen t s

The author is very grateful to A. Michaudon

and S. Cierjacks for their encouragements. He is very

indebted to G. Haouat and M. Gates who contributed to

the preparation of this work. He acknowledges

J. P. Delaroche for providing results in advance of

publication.

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AEC Symp. Series 23 p. l66.

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and J. Sigaud, NEANDC(E) 168 "L", INDC (FR) 7/L

(Nov. 1975).

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and H. Unseld, EANDC (E) "U" (1968) - KFK 1000.

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R.B. Schwartz Nucl. Sci. and Eng. 56 (1975) 27.

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1+1 (1970) 351.

15. B.R. Leonard Jr. Evaluation of the Standard Carbon

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Helsinki - IAEA - CN - 26/110 (1970) 721.

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100
NEED FOR IMPROVED STANDARDS IN NEUTRON PERSONNEL DOSIMETRY*

John A. Auxier

Health Physics Division

Oak Ridge National Laboratory

Oak Ridge, TN 37830

There is a continuing need for standards in neutron monitoring. A discussion of special

problem areas and the benefits of intercomparisons is given. The RBE for leukemia induction

in the survivors of the nuclear bombings of Hiroshima and Nagasaki is greater than ten for

absorbed doses in the bone marrow of less than 100 rads; this may have an important impact

on neutron standards preparation.

(Standards; criteria, performance; dose equivalent; neutrons, low energy; neutrons, high

energy; non-uniform exposures; elements, transuranic; quality factor; glove box;

intercomparison; criteria, accuracy)

More than a quarter century has elapsed since

quantitative measurements of absorbed dose became

feasible, at least on a laboratory scale. 1^2 Several

steps have been taken to establish the bases for

standards on an international level, including

meetings and documents sponsored by the International

Commission on Radiological Protection, the International Commission on Radiological Units and

Measurements, the World Health Organization, and the

International Atomic Energy Agency. Also, there is an approved standard entitled "Personnel Neutron

Dosimeters (neutron energies less than 20 MeV," by

the American National Standards Institute. 3 This

standard gives performance criteria, use factors, and

dosimetry system calibration criteria for neutron

dosimetry systems. However, accuracy criteria are

not included.

Clearly, there remains a definite need for

attainable standards for minimum accuracy in the

determination of neutron energy fluence or dose

equivalent. In addition to standards requirements,

we must remember that there are a number of problems

for which there is still no technologically suitable

answer. Fortunately, the number of persons to be

monitored for neutron exposure in most installations

is small, and special effort may sometimes make

possible a dose determination that is not feasible

with simple monitoring devices.

There are several categories of neutron monitoring, each with special technological problems to be solved, which must be considered in standards

preparation. These include low energy neutrons

(thermal to about 0.5 MeV), high energy neutrons

(greater than 20 MeV), and highly non-uniform exposures (such as those from transuranic elements to be

handled in glove boxes). Each of these has associated difficulties in both measurement and calibration;

though fast-neutrons in the range of 0.5 MeV<E<20MeV

cannot always be measured accurately, there are fewer instrumental and calibration problems than in

the other categories.

Further, there are three generic aspects of the

dosimetry problem of consequence in all the above

categories and which are independent of the instrumentation used: (1) recent heightened concern over the

appropriate quality factor for neutrons, (2) calibration of instruments as a function of energy, and

* Research sponsored by the Energy Research and

Development Administration under contract with

Union Carbide Corporation.

(3) expertise and care in the use of instruments and

interpretation of results.

Technological problems are most apparent in the

energy range below 0.5 MeV. Because of the rapid

variation of neutron non-elastic cross sections and

the low recoil energies resulting from elastic

reactions, no satisfactory solution to measurement

problems, especially for personnel monitoring, is

apparent.

For neutron energies greater than 20 MeV, the

problems depend on the range under study, e.g.,

20 MeV < E < 50 MeV, 50 MeV < E < 300 MeV, etc. However,

by a combination of various track detectors,

activation/spal lation detectors, ionization chambers,

and calculation, the health physicist can generally

assess the dose with confidence that protection is

assured without a huge penalty in conservatism.

Highly non-uniform fields frequently present a combination of problems, including those associated

with low energy neutrons, coexistent soft x-rays, the

need for small detectors for hands and forearms such

that glove box work can be performed, etc.

More research and development is needed in each

of these categories, but no federal agency has accepted the responsibility for solving them. Many who

should be concerned believe that the problems of

dosimetry have been resolved adequately. This does

not appear to be entirely prudent in today's trend

toward "super-safety" in all things.

Let us now examine the more generic aspects of

neutron monitoring. Recent work by Jones^ has

indicated that if one considers the energy absorbed

in bone for the survivors of the nuclear bombings of

Hiroshima and Nagasaki, the leukemia induction RBE of neutrons approaches 30 for low dose levels. For many years, depth dose distributions have been

reported which indicate that neutrons in the fission

range are attenuated in tissue more rapidly than

gamma rays from the fission process or gammas produced by neutron interactions in air. These different

attenuation rates become even more important in the case of the Japanese leukemia data, because RBE is

markedly dose dependent. This dose dependence stems

from the fact that risk from the gamma component

appears to be quadratic with dose while risk from the

high LET recoil ion component seems to vary linearly

with dose. In no case, involving people or large

animals, is there evidence that QF should be greater

than 10 for neutron dose measured either in air or on the surface of the body, so applied health physicists

101
prefer to use QF=10 for protection purposes. In any

event, the magnitude of RBE or renormal i zation of

risk to remove the quadratic behavior of the denominator should be a decision of standards setting

groups

.

The omission of accuracy criteria in the ANSI

N319 standard is primarily because of the lack of a simple neutron dosimeter which responds accurately

over the range of energies which may be encountered

at a given facility. For example, if simple track

counting in nuclear emulsions is used as the basis of

dosimetry, then for any given number of grains used

to identify a track, the average dose per track is

dependent on neutron energy. Consequently, if a Pu-Be

or Pu-B source is used in calibration and the wearer

of the dosimeter is exposed to neutrons of energy less

than 1 MeV, the accuracy can be very poor. In some

cases the measured dose may be an order of magnitude

too low.^ Professional health physicists would, of

course, recognize the circumstances and make corrections; but the accuracy is seldom as good as for

gamma ray measurements, all other parameters constant.

Calibration is an important aspect of neutron

monitoring and one which has an important bearing on standards setting.

Finally, the generic aspect of competence and care must rank high in considerations of standards.

Without a basic understanding of the physical

mechanisms of a detector and its associated apparatus,

results are apt to be unsatisfactory. In addition,

some experience under field conditions is certainly

desirable and probably essential. It is under field

conditions that all generic aspects must be handled

properly, and testing in the field can provide the

impetus to improve the results of a monitoring

program.

One series of field tests that demonstrates the

wide variability of results and the capability to improve systems as a result of testing is the ORNL

Health Physics Division's Nuclear Accident Intercomparison series which. has included at least one major workshop per year for nearly 15 years. Tables

1, 2, and 3 show typical results of these studies;

most of these participants, usually about ten institutions are represented, took part in each of the

sessions shown here. Similar results were obtained

for other groups, i.e., range of reported results was nearly always wide for first-time participants but

improved with experience. These are all high exposures of nuclear accident systems. More recent studies

of routine monitoring systems show similar trends

though there are insufficient data as yet to ascertain

the ultimate improvements to be expected in performance. Table 4 shows clearly that uncertainties of up

to a factor of two must be expected under first test

conditions. For routine monitoring conditions,

especially where detector response may fade or otherwise change during long periods of use, uncertainties

could be much greater. It is disturbing that the

gamma ray dosimetry was subject also to such wide

variation. Table 5 shows marked improvement by the

same group of dosimetrists for the second intercomparison of personnel dosimetry systems. Although the

gamma ray measurements were made in a coexistent

neutron field, the percentile error is much greater

than for the neutrons under these test conditions.

Because of the success of the national and international intercomparisons of nuclear accident systems

initiated at ORNL, the IAEA began sponsoring a series

of intercomparisons of a similar nature. Table 6 lists

studies which have been conducted to date.

It is apparent that standards for neutron

monitoring are needed and, in my opinion, accuracy

criteria are needed as well. However, accuracy

criteria would be of little value without some test

and evaluation procedures pertinent to field applications. It appears that an important aspect of tests or intercomparisons is that of training; using a routine monitoring system without such tests is

analogous to solving problems in textbooks without

knowledge of whether the answers are correct. At

present, accuracy criteria would have to be quite

loose in order to be met by most monitoring groups

for the most simple applications.

References

1. G. S. Hurst, R. H. Ritchie and H. N. Wilson,

Rev. Sci. Inst. 22:981 (1951).

2. G. S. Hurst, J. Radiol. 27:353 (1954).

3. Personnel Neutron Dosimeters (neutron energies

less than 20 MeV), American National Standards

Institute.

4. T. D. Jones, A CHORD Simulation for Insult

Assessment to the Red Bone Marrow, ORNL-5191

(1976).

5. Richard L. Lehman, Energy Response and Physical

Properties of NTA Personnel Neutron Dosimeter

Nuclear Track Film, UCRL-9513 (1961).

102
TABLE 1

INTERCOMPARISON OF AREA DOSIMETERS

IN SIMULATED NUCLEAR ACCIDENTS

HPRR UNSHIELDED

DATE

CALCULATED*

NEUTRON DOSE (rad)

AVERAGE NEUTRON + DOSE REPORTED (rod)

RANGE OF DOSE

REPORTED (rad)

MARCH 1965 324 377 ± 65 300 - 468

JULY 1970 319 310 ± 16 287 - 335

AUGUST 1973 343 338 ± 25 310 - 386

JULY 1974 388 362 ±40''^ 308 - 430

* t . . Based on Absolute Proportional Counter One Standard Deviation

Included Several First-Time Participants

TABLE 2

INTERCOMPARISON OF AREA DOSIMETERS

IN SIMULATED NUCLEAR ACCIDENTS

HPRR — LUCITE MODERATOR

DATE

CALCULATED*

NEUTRON DOSE (rod)

AVERAGE NEUTRON^

DOSE REPORTED (rod)

RANGE OF DOSE

REPORTED (rod)

DECEMBER 1967 26 29 ± 9 18 - 48

JULY 1970 37 42 ± 7 31 - 52

AUGUST 1973 51 56 ± 10 44 - 100

JULY 1974 51 60 ± 11 56 - 77

* Based on Absolute Proportiional Counter ^C)ne Standard Deviation

103
TABLE 3

INTERCOMPARISO N OF AREA DOSIMETERS

IN SIMULATED NUCLEAR ACCIDENTS

HPRR - n-cm STEEL MODERATOR

DATE

CALCULATED*

NEUTRON DOSE (rad)

AVERAGE neutron'^

DOSE REPORTED (rad)

RANGE OF DOSE

REPORTED (rad)

MAY 1967 105 109 ± 36 30 - 154

JULY 1970 104 106 ± 15 73 - 134

AUGUST 1973 138 135 ± 30 105 - 180

JULY 1974 144 133 ± 45 60 - 175

* t Based on Absolute Proportional Counter One Standard Deviation

TABLE 4

PERSONNEL DOSIMETER I N T E R C O MP AR I S O N

DOSAR FACILITY - MAY 1 974

EXPOSURE CONDITION

NEUTRON DOSE

EQUIVALENT (mrem)

GAMMA DOSE

EQUIVALENT (mrem)

UNSHIELDED 453 ±213 24.6 ± 5.9

STEEL-SHIELDED 554 ± 346 18.1 ± 4.3

LUCITE-SHIELDED 675 ± 687 75.1 ± 14.2

14-MEV NEUTRONS 587 ± 501 384 ± 151

104
TABLE 5

PERSONNEL DOSIMETER I N T E R C O M P A R i S O N

DOSAR FACILITY - FEBRUARY 1976

~

NEUTRON DOSE GAMMA DOSE

EXPOSURE CONDITION EQUIVALENT (mrem) EQUIVALENT (mrem)

UNSHIELDED 550 ±217 35 ± 29

STEEL-SHIELDED 753 ± 226 31 ± 30

LUCITE-SHIELDED 532 ± 154 86 ± 46

TABLE 6

IAEA-SPONSORED DOSIMETRY I N T E R C O M P A R I S O N STUDIES

DATE COUNTRY TYPE RADIATION FIELDS USED

JUNE 1970 FRANCE 235u SOLUTION IN CYLINDRICAL TANK - BARE

235u SOLUTION SHIELDED WITH CONCRETE

MAY 1971 USA HPRR UNSHIELDED REACTOR

HPRR REACTOR SHIELDED WITH STEEL

HPRR REACTOR SHIELDED WITH LUCITE

MAY 1973 YUGOSLAVIA RB-REACTOR

APRIL 1975 ENGLAND VIPER REACTOR

105
STANDARDS IN MEDICAL NEUTRON DOSIMETRY

J.J. Broerse

Radiobiological Institute TNO,

Rijswijk, The Netherlands

Clinical and radiobiological experience has shown that differences in absorbed dose of less

than 10 percent can be recognized for tumour eradication probability as well as for normal

tissue damage. These observations determine the degree of precision and overall accuracy

required in medical neutron dosimetry. Standard neutron fields are available only for res- tricted conditions; the feasibility of a number of detectors as standard dosimeters will

be considered. Uncertainties in basic physical parameters will be discussed and recommen- dations for future research in neutron dosimetry for biology and medicine will be included.

(Standards, medical, neutron dosimetry, mixed n-7 fields, basic physical data)

Introduction

To estimate the risk of mixed n-7 radiation

fields and to predict the responses of irradiated biological systems, it will be essential to obtain a quantitative description of the radiation field or of the

energy deposition processes. For purposes of radiation

protection, a rough characterization of the radiation

field in terms of type, energy, direction and number of

particles is in most cases sufficient. For medical and

biological applications generally the absorbed dose and

the radiation quality ^ave to be determined. The absorbed dose is defined as the quotient of the mean

energy imparted by ionizing radiation to the matter in

a volume element and the mass of the matter in that

volume element (the special unit of absorbed dose ^.s the rad and the SI unit is the gray, 1 Gy = 1 Jkg =

100 rad) . The radiation quality can be related to the

neutron energy spectrum, lineal energy (y) spectra or

the linear energy transfer (LET) spectrum and is an essential concept in view of the dependence of the re- lative biological effectiveness (RBE) on these interrelated parameters.

Changes in the quality of initially monoenergetic

neutrons by scattering processes can modify the biological effects to a considerable extent. This has been

illustrated by studies on the survival of colony forming units (CFU) in mouse spleen after irradiation

with d+T neutrons in two different experimental arrangements; in free air and inside a steel collimator . The survival curves presented in fig. 1 show a considerable difference in the radiation response of bone marrow stem cells, resulting in a difference of 25 percent in the RBE for CFU survival. For the irradiations

in the steel collimator 20 percent of the neutron absorbed dose was delivered by scattered neutrons of

lower energy. The contribution of these low energy neutrons increases the RBE according to the general tendency as observed for effects on various biological systems. Even larger differences in RBE values have been

demonstrated for mortality studies in mice after fission

neutron irradiations at two installations where the

fission spectra were produce^ by exposing converter

plates with thermal neutrons . These biological results

clearly indicate the need for quality specification;

at present quality can most unambiguously be characterized in terms of the neutron energy spectrum.

The selection of methods for neutron dosimetry and

the precision (i.e. the reproducibility) and accuracy

to be achieved will depend og the object irradiated

and on the endpoint observed . In studies of effects

of fast neutrons on mammalian systems, doses will range

from 100 to 1000 rad (1 to 10 Gy) ; in fast neutron ra- diotherapy the neutron dose will be administered in

fractions e.g. in daily fractions of 100 rad (1 Gy) at

a dose rate of approximately 10 rad/min (0.1 Gy/min)

.

In radiotherapy applications, differences of 10 percent

in absorbed dose have shown significant differences in

local control of tumours and in ghe incidence of late

radiation sequelae, e.g. fibrosis . For radiobiological

studies, maximum variations over the -biological object

should be restricted to a ratio of 1.10 and preferably

1.06, in order to avoid significant changes in biological responses over the object . Based on these considerations, it is now generally recognized that, for

biological and medical applications, the dose in the

biological object should be determined with a precision

of at least 2 percent and an overall uncertainty of not more than 5-6 percent. Some authors are even more stringent and aim for an overall uncertainty of less

than 2 perceng in the statement of the absolute total

absorbed dose . It must be stressed that the above

mentioned requirements are more severe than generally

encountered in radiation protection, where the dose

effect relations for radiation-induced late effects

are not yet well known. Although, higher accuracy and

Fig. 1. Survival curves for bone marrow stem cells after

irradiation with d+'J neutrons in two different experimental arrangements .

0 200 400 600

_2 neutron dose/rad or 10 Gy

106
precision in dosimetry are required for the medical

applications, the doses and dose rates under these con- ditions are higher than in the protection situation,

which facilitates the measuring techniques.

In the present contribution the principles of

dosimetry in mixed fields will be discussed. The absorbed dose, D, can generally be derived from the response, R, of the dosimeter

r

a R (1)

where a is the response function of the instrument and

the reciprocal of a is the sensitivity. The response

function, a , can be calculated from known constants

and measured parameters or can be determined by calibration in a standard neutron field or by comparison

of the response of the instrument with that of a standard dosimeter. In the context of this presentation a neutron field is regarded as "standard" when

the spatial and temporal distribution of absorbed dose

rate for both neutrons and accompanying photons and

the neutron and photon energy spectrum are known. A dosimeter is regarded as "standard" or "absolute" if it can be constructed and subsequently used to measure the

absorbed dose without the necessity^of calibrating its

response in a known radiation field . The different medical applications include fast

neutron radiotherapy, capture therapy with thermal and

epi thermal neutrons, neutron radiography and in vivo

neutron activation analysis. The availability of standard neutron fields for these different situations will

be discussed and the usefulness of different instruments

as standard dosimeters will be analyzed. The uncertainties in basic physical parameters to be employed for

the conversion of instrument response to neutron and

gamma absorbed dose will be discussed. Some common

values will be recommended, however, future measurements

of some basic parameters are highly necessary for improvement of internationally acceptable standards.

Principles of mixed-field dosimetry

Neutron fields are always accompanied by gamma

rays originating from the neutron producing target,

the shielding and other materials surrounding the target as well as from the biological object or phantom

material irradiated. It is necessary to determine the

neutron absorbed dose, D^^, as well as the gamma ray

absorbed dose, D^, at a certain position in a material

of specified composition, because of the differences

in relative biological effectiveness (RBE) of these two

radiation components.

Generally, two instruments are used for the evaluation of the separate absorbed dose of neutrons and

photons in a mixed field. One device (T) is usually con- structed to have approximately the same sensitivity to neutrons and to photons, whereas the construction of

the second instrument (U) results in a lower sensitivity to neutrons than to photons. For the same mixed

field, the quotients of the responses of the dosimeters

and their sensitivities to gamma rays used for calibration are given by;

R' = k„D^, + h„D^

T T N T G

% = \% ^ ^°G

(2)

(3)

where D^, and are absorbed doses of neutrons and of N G photons m tissue in the mixed field; k^ and k are the

sensitivities of each dosimeter to neutrons re!^ative to its sensitivity to the gamma rays used for calibration

and h,^ and h^ are sensitivities of each dosimeter to

the photons in the mixed field relative to its sensitivity to the gamma rays used for calibration.

Most detectors used in neutron dosimetry have

approximately the same sensitivity to neutrons and photons, e.g. tissue equivalent (TE) ionization chambers

and TE calorimeters. The response of one of the above

mentioned detectors must be combined with that of a detector which is relatively insensitive to neutrons

to determine D and D^. It will be generally advantageous to reduce k. to —minimize overall uncertainties in

-u Y"^ the neutron and photon absorbed doses . A common procedure in neutron dosimtery is the use

of TE, CH or CH^ ionization chambers, as the neutron

sensitive device, in combination with nonhydrogenous

ionization chambers (Al/ftr, Mg/Ar, C/CO^) or GM counters. The principle of the use of ionization chambers

in neutron dosimetry does not differ from their use in

photon dosimetry. It can be shown that:

W (s ) w,g c

w,g N

(K^/K ) t m c

t m N

(4)

W is the average energy expended to create an ion pair;

s is the ratio of the average mass stopping power of

wall relative to the gas and K^/K^ is the ratio of

th^ kerma in tissue to that in the dosimeter material . The subscript c denotes values applicable to the

calibration situation, for which gamma rays are commonly used. Although photon spectra in neutron fields are generally not known, h^ is usually taken to be equal

to one. However, this is not necessarily always the

case; calibration at more than one photon energy is

indicated when there is a possible dependence of h^

and hy on photon energy.

Availability of standard neutron fields

For specific magical applications such as inj^yiy^

activation analysis and neutron capture therapy

it will be sufficient to perform the calibrations in a neutron field of known fluence. Standard neutron

fluence fields are available for radioactive sources

such as Am-Be, S^-^e and Pu-Be, sp(jgtaneous fission

sources such as Cf and reactors . Monoenergetic

neutron beams with energies between 60 eV and 120 keV

can be produced with sets of matched filters and

scatteri^^

f?'"''''^ located in the beam tube of high flux

reactors

Accelerator neutron sources which can serve as primary standards for neutron fluence measurements involve

nuclear interactions where a neutron is emitted in co- incidence with a charged particle. The most important

of the^e standard reactions are H(n,n)H, 0(7, n)H,

D(d,n) He and T(d,n) He. In the first reaction, the

fluence can be derived from the number of recoil protons counted and the scattering cross section cr (n,n).

For the other reactions associated particle counting

can be employed. Considering the reaction kinematics,

the energy and direction of emission of the neutron

can be derived from the measurement of the charged

particle which is emitted in coincidence. In principle

accelerator neutron sources can also be used as fluence standards by measuring the induced activities in

materials with well-known reaction cross sections.

By definition dosimetry includes those measurements (or calculations) which secure information on the

energy deposition processes and the resulting absorbed

dose. For purposes of calibration and intercomparison

of neutron dosimeters, standard neutron fields have to

be established where the absorbed dose (or dose rate)

is known with regard to its spatial and temporal distribution and the neutron energy spectrum. In fast

neutron radiotherapy and radiobiology , the contribution

of thermal and intermediate energy neutrons to the total

dose is generally limited. This can be seen from table

1 , where the kerma due to thermal and intermediate neutrons (neutrons with energies between the Cd cut-off

107
TABLE 1

.

KERMA OF THERMAL AND INTERMEDIATE NEUTRONS RELATIVE TO

TOTAL DOSE FOR A COLLIMATED 15 MeV NEUTRON BEAM

(SSD 45 cm, field size 6 cm x 8 cm)

th K.

depth in phantom

(cm)

0.9

3.0

5.0

10.0

20.0

29.4

""total

(percent)

0.17

0.23

0.23

0.21

0.20

0.18

total

(percent)

0.031

0.031

0.027

0.019

0.018

0.017

energy, Oj^ eV, and 10 keV) is given for a 15 MeV neu- tron beam^^j

The Cf spontaneous fission source has been

suggested as a standard for neutron dosimg^^y delivering

soft tissue kerma rates per microgi^am o^ Cf at 1 ^ met^r in free air of 0.J99 mjad h pg

( J . 99 ^Gy h

pg ) and 0.108 mrad h ]ig (1.08 pGy pg ) for

the neutrons and gamma-rays respectively . In fig. 2 the absorbed dose rate factor is sho^g^^s a function of

distance for a 1.5 cm active length 'Cf source encapsulated by 0.7 mm Pt + 10 percent Ir . It should be

realized that the relativ^^^ose contributions from

beta rays emitted by the Cf equilibrium fission product mixtures can not be neglected near the source when

the encapsulation thickness is less than about 0.5 mm

Pt + 10 percent Ir. It can be seen from fig. 2 that in

a tissue-equivalent phantom the relative contribution

from gamma rays increases with increasing distance from

the source. Provided 'that source encapsulation and

scattering conditions are normalized, californium with

a minimum source mass of 1 mg can be considered to be

a reasonably good source for calibrating and checking

the proper operation of the instruments to be used for

mixed-field dosimetry.

Monoenergetic neutrons with energies ranging between a few tenths of a MeV and 20 MeV can be produced

with accelerators by bombarding targets with beams of

protons and deuterons (examples are the d+T, d+D and

p+T reactions) . A number of factcjrs may influence the

composition of the neutron field . In case of deuteron

bombardment, competing reactions can become important

e.g. with carbon layers originating from organic vapors

in the vacuum system and deposited on the target. For

the d+T reaction, the target will become increasingly

contaminated with deuterium, resulting in a low energy

neutron component from the d+D interaction. Furthermore, gamma rays can result from fast neutron interactions with the target assembly. Extraneous deuterons

striking this assembly or components of the beam transport system will also generate gamma rays and spurious

neutrons. Target thickness determines. the energy spread

of the emitted neutrons and the neutron yield. During

ion bombardment the target deteriorates which results

in lower neutron yields for equal ion currents. These

factors make it difficult to predict the spatial and

temporal distribution of absorbed dose rate for both

neutrons and accompanying photons and the neutron energy spectrum. The d+D reaction, using pressurized gaseous

deuterium targets, should be considered for standardization purposes, in as much as the neutron yield is only

dependent on the reaction cross sec^^on and the number

of target atoms. A recent inventory has shown that

accelerator standard fields with well defined levels of

absorbed dose are not yet available at the national

standards laboratories, although at some institutions

including NBS, NFL and PTE such facilities are in preparation . In three instances (RARAF Brookhaven, GSF Neuherberg, and TNO Rijswijk) accelerator produced neutron

fields have been employed for neutron dosimetry intercomparison projects (the International Neutron Dosimetry Intercomparison, INDI , sponsored by the ICRU and

the European Neutron Dosimetry Intercomparison Project

ENDIP, promoted by the Commission of the European

Communities) and considerable effort has been devoted

to the monitoring of the radiation field, to obtain

a continuous indication of total absorb^^ ^8^1 I 9^'TM^~

ray absorbed dose and radiation quality ' ' . Analysis of the monitoring results indicates that the

total absorbed dose can be delivered to the participants' dosimeters with a reproducibility within 2 percent. Larger variations (with a maximum of 20 percent)

have been observed in the relative dose contribution

of photons, which might be attributed to instabilities

in the ion beam geometry. At one institute two disc

type tissue equivalent transmission ionization chambers

have been employed to monitor total absorbed dose; the

use of this dual monitoring system has proved to be of

great value f^i^ a continuous check on the reliability

of the system . It must be stressed that the radiation

fields for INDI and ENDIP have been established only

for purposes of intercomparison and not for calibration.

Although the objective of these intercomparisons was

to understand differences in dosimetry, much was learned

about how to do an intercomparison and how to do calibrations in a reliable manner . This will be of great

value for the laboratories concerned and perhaps also

for the standards laboratories if they have to offer

calibrations in this field at some time in the future.

Standard neutron dosimeters

For photon dosimetry only three devices are generally accepted as being accurate enough to serve as

Fig. 2. Comparison of the attenuation in a tissue equi-

^l^ent phantom for radiation from a 1.5 cm active length

Cf source (0.7 mm Pt + 10 perce^^^Ir) with the attenuation in water fo:jgghotons froiji Ra (0.5 mm Pt +

10 percent Ir) and Ir sources .

10

^^*Ro (rod h

' cm^ mg"' or 0.01 Cy h"' cm^ mg"')

^Ir (rod h"' cm^ mC!"' or 270.3 pGy h"' cm^ Bq"')

^^^Cf (rod h"' cm^ |jg"' or 0.01 Gy h"' cm^ (jg"')

gomma rays

distance/ cn

108
dosimetry standards: calorimeters, ionization chambers

and Fricke chemical dosimeters . It should be realized

that for the latter two dosimeters, the response

function in a neutron field will be different from that

in the photon field used for calibration, because of

differences in basic physical parameters determining

instrument response for these two types of radiation.

Consequently, it might be more appropriate to qualify

ionization chambers and chemical dosimeters as secondary standards for neutron dosimetry.

The proportional counter lined with polyethylene

and filled with ethylene as designed by Hurst is

still considered to be an absolute device for the

measurement of neutron absorbed dose. However,

measurements with proportional counters have to be performed at beam intensities which are considerably re- duced with respect to the situation under whi^^ biological and clinical irradiations are performed . These

reduced beam intensities involve different operational

conditions of the accelerator and consequently the ra- diation fields can show differences with regard to geometry and radiation quality, including the relative

contribution of photons to the total absorbed dose

.

The calorimeter is a fundamental instrument for

measuring absorbed dose and in principle its response

is independent of neutron energy . The method is too

insensitive and cumbersome for routine use; rather it

should serve as a dosimetry standard for the calibration of secondary instruments. The calorimeter can only measure the total amount of heat produced and so determines the total absorbed dose due to neutrons plus accom- panying photons. A separate measurement, usually of the

gamma ray component, is needed to assess the separate

absorbed doses. The choice of absorbing material is more critical for neutrons than for photons, ideally

the material should have the same atomic composition as tissue. In addition the thermal defect, defined as that

fraction of the energy absorbed from the beam which does

not appear as heat, but is stored in the form of chemical changes or lattice defects, should be zero. The

overall uncertainty in the determination of neutron absorbed dose can be estimated to be about 4.5 percent;

the principal sources of uncertainty being the ratio

of kerma in tissue to that in calorimeter material

(3 percent) , the thermal defect (2 percent) and the

uniformity of absorbed dose in the calorimeter element

( 1 percent) , whereas the random uncertainty can be

taken as 1 percent.

With great care one can construct ionization

chambers capable of measuring exposure with an uncertainty of about 1 percent. lonometric determinations of

the absorbed dose of photons using the Bragg-Gray principle will entail overall uncertainties of^ 2.3 percent

which arise mainly from uncertainties in W (1 percent)

and stopping power ratios (2 percent). The principle of

the use of ionization chambers in neutron dosimetry

does not differ from that in photon dosimetry. In order

to make the ratio of kerma in the dosimeter material to that in soft tissue as close to one as possible, con- ducting tissue equivalent plastic, designated A-150, is

often used for the construction of ionization chambers.

As can be seen from table 2 the principal compromise in

the formulation of A-150 is the substitution of carbog^

for much of the oxygen required to match muscle tissue.

An approximately homogeneous chamber can then be obtained by flushing the chamber with a mixture of gases

with comparable composition (TE gas) . Use of ionization

chambers for the determination of the neutron absorbed

dose involves a major systematic uncertainty since

equation (4) must be evaluated to determine k^. The

systematic uncertainties in W ratio, stopping power

ratio and neutron kerma ratio can be estimated to be

equal to 5, 3 and 3 percent, respectively. Additional

factors such as failure to meet Bragg-Gray conditions

equally for photon and neutron irradiations (with a systematic uncertainty of 2 percent) result in an overTABLE 2.

ELEMENTAL COMPOSITION IN PERCENT BY WEIGHT OF A-150

MUSCLE EQUIVALENT PLASTIC COMPARED TO ICRU MUSCLE TISSUE

element ICRU muscle A-150 plastic

H 10.2 10 2 + . 1

C 12.3 76 8 + .5 O 72 .

9

5 9 + .2 N 3.5 3 6 + .2

Co .007 1 8 + . 1

F not listed 1 7 + . 1

total 98,9 100 + .5

all uncertainty of 7 percent for the determination of

neutron absorbed dose with an ionization chamber . The ferrous sulfate (Fricke) dosimeter has been

extensively studied using photons and electrons. Due

to its good reproducibility, linearity of response and

relatively small energy dependence, the system is suitable for photon, electron and neutron dosimetry. A disadvantage of the system is the rather low sensitivity requiring long irradiation times at sources having

relatively low output e.g. 15 MeV neutron sources. Recent improvements in the optical density measurements

have r^gulted in higher sensitivities of the Fricke

system . For fast neutrons, there are few determinations of G (the number of ferric ions produced by an energy dissipation of 100 eV) . The G value for mixed

neutron and gamma ray fields has a systematic uncertainty of 7 percent. If the low dose technique is used, the

random uncertainties are between 2 and 3 percent, which

implies that with, the Fricke dosimeter, an overall un- certainty of 8 percent can be attained for the determination of the tota^ absorbed dose in mixed neutron

and gamma ray fields .

Uncertainties in basic physical parameters

In order to compare the results obtained by various groups in performing dosimetry in mixed fields with

systems generally encountered in neutron radiotherapy

and neutron radiobiology , two international neutron dosimetry intercomparisons have been performed. The In- ternational Neutron Dosimetry Intercomparison (INDI) was conducted at Brookhaven National Laboratory in 1973,

the analysis of the results completed and will be

published in the near future . A European Neutron Dosimetry Intercomparison Project (ENDIP) was conducted

in 1975 at two locations: the Institut fur Strahlenschutz GSF, Neuherberg and the Radiobiological Institute TNO, Rijswijk. The analysis of the ENDIP results

is near to complet^gn, preliminary results of ENDIP can

be found elsewhere

The quantities to be measured in INDI and ENDIP

were the neutron and gamma tissue kerma in free air for

neutron beams of different energies and the neutron

and gamma tissue dose at various depths in a water

phantom. Analysis of the calculations of the participants showed that, for the same experimental conditions,

the various groups employed divergent basic parameters

characterizing the detector response, such as the

ratio of kerma values in dosimeter material and tissue,

the energy required to produce an ion pair in a gaseous detector and the neutron sensitivity of photon dosimeters .

.109
a. Ratio of kerma in dosimeter material and soft tissue

A considerable number of participants in INDI and

ENDIP employed tissue equivalent ionization chambers,

thus a conversion from kerma in TE dosimeter material

to kerma in soft tissue (ICRU muscle approximation) was necessary. The kerma conversion factors adopted by the

participants in INDI and ENDIP are summarized in table

3. It can be seen that the values employed show differences of up to 8 percent for the same neutron energy.

Information was available from two groups on the set of

parameters employed for the evaluation of the INDI and

the ENDIP results. Apparently, these groups had to change their parameters on the basis of additional ex- perimental or theoretical information.

New kerma factors (products of mass energy transfer

coefficient and neutron energy) have recently been

calculated by Caswellj Coyne and Randolph for neutron

energies up to 30 MeV . In the energy region below 20

MeV, overall uncertainties less than 2 percent and

usually less than 1 percent have been estimated for

hydrogen. For other elements, the estimated overall un- certainties would nearly always be less than 10 percent

and usually less than 5 percent. Overall uncertainties

in kerma factors for tissue and tissue like material

may be estimated to be about 3 percent. For energies

above 20 MeV kerma factors except for hydrogen, have

a much higher uncertainty due to the lack of neutron

cross section measurements, in this energy region,

3g Other groups have also performed kerma calculations ' and it will be very useful to compare the different

calculational procedures and to adopt one common set of

kerma factors.

b. W values

Relativel^y l^arge discrepancies are also observed

in the ratio W^/W^ as employed by the participants in

INDI and ENDIP for TE gas consisting of methane, carbon

dioxide and nitrogen (see table 4) . The main differenTABLE 3.

RATIO OF KERMA IN ICRU MUSCLE TISSUE TO KERMA IN TE DOSIMETER

MATERIAL ADOPTED BY PARTICIPANTS IN TWO INTERNATIONAL

NEUTRON DOSIMETRY IN TERCOMPARI SO N S

Neutron Energy (MeV)

Participant 15.1 5.5 2.1 0.67

AFF 0.926 0.962 0.917 0.962 0.962 ANL 0.9666 0.9835 0.9749 0.9880 0.9847

CHRl 0.925 0.962 0.950 0.950 0.950 CNEN 0.969 0.963 0.956 0.970 0.959 FON 0.988 0.976 0.981 0.982 0.981

GSFF (1

)

0.971 0.990 0.971 0.971

GSFF (2) 0.935 0.980 0.971 0.976 0.971

GSFM 0.969 0.963 0.956 0.970 0.959

IAEA 1 .00 1 .00 1 .00 1 .00 1 .00 MRS 0.99 0.95 0.97 0.97 0.97

RAR 0.956 0.960 0.956 0.960 0.956 TNO (1) 0.972 0.935 0.951 0.95i 0.951 TNO (2) 0.94 0.98 0.96 0.97 WWD 0.948 0.950 0.940 0.940 0.946

(1 ) Values used for INDI

(2) Values used for ENDIP

ce among the various^ groups is due to the fact that

in some cases, the W ratio was taken to increase with

decreasing neutron energy, while other groups employed

a constant W ratio. These discrepancies lead to a maximum difference of 11 percent in the case of the lower

neutron energies. _ For photons and electrons extensive measurements

of W values in air have been performed, resulting in^ a relatively small uncertainty of about 1 percent in W

for electrons in air . W values for electrons in other

gases are less well known, while for protons and other

charged particles only limited information is availab^^

Until recently only th^^sxper^imental results of Larson

and Leonard and Boring on W /W^ were available for

protons in methane-based TE g§s. In fig. 3, l^ges

1 and ^^show the relationships used by Dennis ^nd_

Bewley , respectively, for the calculations of W /W^

for neutrons of different^gnergies . New experiment^!^

results of Chemtob et al. and Roh£ig and Colvett

show that the energy dependence of W for protons is^ small. It can be concluded that the dependence of

on neutron energy needs further experimental veri_fic^ation. Until this is done it is recommended that W^,/W„ N G

for tissue equivalent gas be taken to be 1.05. It is

estimated that the systematic uncertainty in this value

is about 5 percent for neutron energies above 1 MeV and

that the systematic uncertainty increases at lower

energies

.

c. Stopping power ratios

If the gas and the wall of an ionization chamber

have the same composition, it is generally assumed that

they have the sam^gStopping power. There is some experimental evidence that the mass stopping power for

protons and alpha particles in vapour could be higher

than that in the condensed phase. However, in other ex- periments no difference in stopping power has been observed and the measured difference of about 5-10 percent is not much higher than the experimental uncertain

TABLE 4.

RATIO OF W|^/Wg FOR TE GAS EMPLOYED BY PARTICIPANTS IN

INDI AND ENDIP

Neutron Energy (MeV)

Participant 15. i 5.5 2.1 0.67 "2cf

AFF 1 .01 1 .02 1 .03 1 .05 1 .04 ANL 1 .04 1 .04 1 .04 1 .04 1 .04

CHRl 1 .04 1 .05 1 .09 1 .10 1 .10 CNEN 1 .038 1 .058 1 . 106 1 . 158 1.110

EUR 1 .03 1 .098 1 .151 FON 1 .037 I .037 1 .037 1 .037 1 .037

GSFF (1) 1 .042 1 .042 1 .042 1 .042 1 .042

GSFF (2) 1 .053 1 .053 1 .053 1 .075 1 .053

GSFM I .037 1 .059 I . 105 1 . 161 1.110

NIRS 1 .055 I .055 1 .055 I .055 1 .055

RAR 1 .05 1 .05 1 .05 1 .05 1 .05 TNO (1) I .055 1 .055 1 .055 1 .055 1 .055 TNO (2) 1 .05 1 .05 1 .05 1 .05 1 .05 WWD 1 .05 1 .05 1 .05 1 .05 1 .05

(1) Values used for INDI

(2) Values used for ENDIP

110
Fig. 3 Ratio of W for protons and photons in methane1_ and 2 show the relationships used

for neutrons of different

based TE gas. Lines

for calcu

energies.

calculations of W /W

1.4-

0.1 O

15 15

• Leonard and Boring (1973)

A Chemtob el ol. (1976)

Rohrig ond Colvetl (1976)

Lorson (1958)

1

)roton energy/keV

ty. Until further experimental data become available,

no firm conclusions can be drawn. The use^^of a recommended relative mass stopping power of unity may result in an overestimate of the neutron dose in the wall of a hydrogenous ionization chamber b^^about 5 percent.

The good agreement observed , between absorbed

dose in a 14 MeV neutron field measured with a polythene calorimeter and a polythene-ethylene ionization

chamber using a stopping power ratio of one, indicates

that the error introduced by taking equal stopping

power at all neutron energies will probably be less

than about 3 percent. Decreasing the overall uncertainty attainable with ionization chambers requires better

data on the relative mass stopping powers, particularly

for TE plastic and TE gas.

TABLE 5.

RELATIVE NEUTRON SENSITIVITY, k^, EMPLOYED BY PARTICIPANTS

IN INDI AND ENDIP

Neutron Energy (MeV)

Detector Participant 15.1 5.5 2.1 0.67

C/CO2 chamber AERE 0. 105 0.0751 0.0677 0. 106 ANL 0 351 0.097 -- 0.099 CHR 0 356 0.114 0. 089 0.074 0.086 CNEN 0 373 0. 106 0.077 0.073 0.094

EUR 0 39 0.058 --

GSFF (I) 0 33 0.13

GSFF (2) 32 0.10 0.02 0 . 02 0 . 02 GSFM 0 341 0.096 0.070 0 . 065 0 . 086

NIRS 0 365 0. 102 0.090 0 . 083 0 . 090

NRPB 0 361 0 . 104 0.068 0.053 0.090

Al/Ar chamber EUR 0 046 -- 0.020 0.01 1 -- FON 0 127 0.029 0.017 0.017 0.017

IAEA 0 13 0.01 1 0.012 0.014 0.02

RAR 0 132

Ge iger - Mu II er ORNL 0 005 0.005 0.005 0.005 0.005

cou nter RAR 0 015 0.005 0.002 0.002 0.003 TNO (I) 0 0 0 0 0 TNO (2) 0 004 0.002 0.001 0.001 WWD 0 004 0.002 0.001 0.001 0.001

(1 ) Values used for INDI

(2) Values used for ENDIP

d. Relative neutron sensitivity, k

In table 5 the values for the relative neutron sen- sitivity, ky, as employed by the groups participating

in INDI and ENDIP are tabulated for three different

devices, notably C/CO^ chamber, Al/Ar chamber and

Geiger-Muller counter. In two cases, where groups em- ployed either relatively high or relatively low values

of ky the reported gamma ray contributions were found

to be lower or higher respectively than the mean value

observed.

For a number of neutron insensitive devices, the

relative neuJ:ron sensitivity has been derived from

calculations , from measurements made in neutron fields

with minimal gamma ray contamination and by calibration

with a proportional counter. In table 6, relative neu- tron sensitivities are given for relatively smal^^nonhydrogenous ionization chambers and a GM counter

It can be seen that the relative neutron sensitivities

for Al-Ar are considerably less than fog^t^^ C/CO^ com- bination. Using a lead-filter technique ' , k^ values

of 0.32 and 0.28 have been observed for a C/CO^ chamber

for neutrons from the d(35)+Be reaction (mean energy

approximately 15 MeV) and for colli^^ted 14 MeV neutrons, respectively. Kuchnir et al. employed a difference method to determine the relative neutron

sensitivity of ionization chambers as a function of

neutron energy for the wall and gas combinations indicated in fig. 4. The difference technique utilizes

accelerator target reactions for which the neutron fluTABLE 6.

RELATIVE NEUTRON SENSITIVITIES AND THEIR OVERALL UNCERTAINTIES FOR NON-HYDROGENOUS IONIZATION CHAMBERS (CAVITY GAS

AT ATMOSPHERIC PRESSURE) AND GM COUNTER

relative neutron sensitivity, ky

neutron percent C-CO, chamber

energy energy ^

^ ^ MeV spreod computed observed

Al-Ar chamber

observed^

0.7

2

5

15

252

25

5

6

7

0.05

0.07

0.07

0.31

0.07 + 0.006

0.08 + 0.01

0.11 +0.01

0.32 + 0.02''

Cf fission 0.08

spec trum

0.014 + 0.004

0.012+0.002

0.01 I + 0 .003

0.13 +0.01

0.02 +0.015

GM counter

,c measured

0.001 + 0.0005

0.001 + 0.0005

0.002 + 0.0005

0.004 + 0.001

0.001 + 0.0005

a: Goodman, 1974; b: Greene, 1974; c: Goodman and Colveft, 1974; d: Broerse, 1974.

Ill
F^g. 4. Relative neutron sensitivity with respect to Co gamma rays as a function of neutron energy for a magnesium-wall and graphite-wall chamber^i^ith a variety

of filling gases at atmospheric pressure

0.4 H

3

0 -I 1— 1 1 r- 0 5 10 15 20

neutron energy/MeV

ence rate and neutron energy are functions of the angle

with respect to the target but for which the accompanying photon production is isotropic. The data in f^g. 4 have been obtained for a Mg chamber with a 0.5 c^ cylindrical cavity and for a C chamber with a 2 cm spherical cavity. Below 15 MeV the overall uncertainty in

k was estimated to be about 10 percent; above 15 MeV

tSe values of k are less reliable. There is fairly

good agreement between the C/CO and C/air data of

fig. 4 and the corresponding values discussed above.

The values given in table 6 can be used as a guideline

for relative neutron sensitivity, k^. However, these

data should be handled cautiously since ionization

chambers constructed' with the same wall and gas materials, but of markedly different size, configuration or gas pressure, may have different relative ggutron sen- sitivities. Recent studies of Lewis et al. indicate

slightly higher values of k^ for energy-compensated

GM counters than indicated m the table, however, these

results need to be confirmed by complementary experiments .

Conclusions and recommendations

For medical and radiobiological applications there

is certainly a need for standard neutron fields where

the absorbed dose is known with regard to its spatial

and temporal distribution and the neutron energy spectrum. If calibration of absorbed dose with fis^^^''^ spectrum neutrons would prove to be satisfactory, Cf can be considered to be an acceptable standard neutron

source, provided that source encapsulation and scattering conditions are normalized.

Standard neutron fields produced by accelerators

are not yet available at national standards laboratories, however, facilities for calibration of neutron

dosimeters are under construction. At present, the most

preferable standggd fiel^^i^ill be deliverd by a photon source e.g. Co or Cs, with subsequent adjustment of the response function to the neutron field con- cerned. Regarding the preparation of standard fields

for medical neutron dosimetry, the characteristics of

the neutron field should be relevant to the clinical

situation. This implies that the possibility of irradiations with collimated beams at reasonably high dose

rates of e.g. 5-10 rad/min (0.05-0.1 Gy/min) should be

provided.

For standardization purposes, the instruments

having the lowest overall uncertainty for measuring total absorbed dose are the TE calorimeter and the homogeneous TE ionization chamber. In both cases, a separate

measurement with a dosimeter having a relatively low

neutron sensitivity is required to evaluate the partial

absorbed dose of photons. Relatively bulky and complicated equipment is needed for calorimetry. The ionization chamber has many advantages including greater ^ensitivity, lower random uncertainty and smaller size . The two international neutron dosimetry intercomparisons have provided information on the present adequacy of neutron dosimetry and on the advantages,

corrections and ^^stematic errors involved in the vari- ous methods used . The groups participating in these

two intercomparison projects have employed for the same experimental conditions, sets of basic physical data

which showed considerable variations. General agreement on the most accurate parameters will be an essential requirement to accomplish consistency in neutron

dosimetry results obtained at different institutes.

For a specific type of instrument such as the TE ionization chamber, which was employed by the majority of

participants in both INDI and ENDIP an analysis for

uniform parameters was performed. For INDI it was shown

that the differences between reported values could not

be attributed solely to differing factors used to convert^i^nstrument response to neutron kerma or absorbed

dose . For ENDIP, the standard deviations of kerma and

absorbed dose values were in the same order as the standard deviation of the instrument responses reported by

the participants. These analyses clearly indicate that

other factors such as incomplete charged particle

equilibrium, insufficient saturation of the ionization

chambers, and choice of effective measurement point

tend to overshadow inconsistencies in ratios of W,

kerma and mass stopping power applied. Although adoption of uniform basic physical parameters is desirable,

it seems of more importance to agree to use the same dosimeter type, preferabl^^a TE ionization chamber of

standardized construction . The measurement and cor- rection procedures for this common type of dosimeter

should be standardized, e.g., the calibration with

photons, the gas flow rate, the collecting potential,

polarity and the correction for wall thickness. An

alternative would be that a standards laboratory maintain a reference dosimeter which could be used for ca- libration of other dosimeters. At the moment, however,

none of the standards laboratories is prepared to offer

such a service. Different specifications should be

applied for the interpretation of instrument response

at different energies. General agreement_on the neutron

energy dependence of parameters such as W and corrections for wall attenuation will obviously be necessary,

but will only be needed for one type of dosimeter. A thimble type homogeneous TE ionization chamber with a set of TE build-up caps to achieve complete charged

particle equilibrium would be a useful standard dosimeter .It must be realized that for medical applications the quantity of interest is the absorbed dose

in the patient. In general, this value is derived from

in phantom measurements, with separate corrections for

tissue inhomogeneities . For the various centers in

Europe co-operating in the Fast Neutron Therapy Project

Group sponsored by the EORTC (European Organization

for Research on Treatment of Cancer) , a second draft

of i^protocol for medical neutron dosimetry is available

Interest in the use of fast neutrons for biological and medical applications has increased considerably

in the past decade. These applications introduced re- quirements for precision and accuracy in dosimetry

methods which are more demanding than generally encoun112
tered in radiation protection. The following recommendations for future research in neutron dosimetry in

biology and medicine have been formulated :

1 . It is recommended that standards laboratories

develop and establish standard instruments and/or ra- diation fields which will be directly applicable to absorbed dose calibrations for the neutron energies

applied in the life sciences.

2. The results of neutron dosimetry should be

reported by giving the following values:

(a) The absorbed dose of neutrons in the region of

interest and its time dependence or fractionation schedule . (b) The absorbed dose of photons or the magnitude

of this component relative to either the neutron or total absorbed dose in the region.

When the spatial variation of absorbed dose in the

region of interest can affect the biological response,

it is important to describe the variation.

Meaningful comparisons of biological data and

therapeutic results require that radiation quality be

specified, for example, by means of neutron energy

spectra or by lineal energy spectra. These data should

be supplied at more than one point when quality changes

significantly in the region of interest.

3 . Knowledge of neutron energy spectra is important not only for correlation with biological effects

but also for determining the corrections and systematic uncertainties to be applied to measurements with

neutron and photon dosimeters. Neutron spectrometers of

small size which can operate over a broad range of

neutron energies, say from about 1 keV to at least 50

MeV, are needed. The low-energy capability is needed

mainly for radiobiological research; the high-energy

capability is necessary for the spectra encountered in

radiotherapy. The instrument ( s) should be small and

nondirectional to permit spectral studies with good

spatial resolution to be made in phantoms and across collimated beams. The data available on radiation quality need to be expanded to include effects on spectra

caused by variations in the design of collimators,

filters and accelerator target structures.

4. The data on cross sections and related mass energy transfer coefficients for neutron-produced se- condaries in the elements constituting tissue and for

elements employed in dosimeters should be expanded to improve the values presently available above 20 MeV and

to provide necessary information above 30 MeV. These

data are required to compute kerma factors for various

tissues, dosimeter materials and tissue-substituting

materials, particularly for use in radiotherapy.

5. In order to decrease the overall uncertainty of

neutron absorbed doses determined with ionization chambers, it is essential that further measurements of W

for the gases used in these chambers should be made,

especially for protons and heavy charged particles with

energies between 0.01 and 10 MeV. This will permit

effective values of W for neutrons to be calculated

more reliably.

6. Decrease of the overall uncertainty attainable

with ionization chambers also requires better data on

the relative mass stopping powers with regard to the

physical state (solid or gaseous) of the chamber

materials, particularly for low energy secondaries.

7 . In order to decrease the systematic uncertainty

of the measurement of neutron absorbed doses with

calorimeters, further measurements of the thermal defect for the materials used in these instruments should

be made.

8. The overall uncertainty of measurements of photon absorbed doses in mixed fields can be decreased by more reliable determinations of relative neutron sen- sitivities of photon dosimeters as a function of neu- tron energy than has heretofore been available. In particular, this should include neutron energies above

15 MeV. Evaluation of the effects of dosimeter size

and configuration on relative neutron sensitivities

should be made.

In addition to improving data presently available,

many of these recommendations call for a substantial

extension of our knowledge into both higher and lower

regions of neutron and charged particle energies than

have previously been the focus of study.

Acknowledgements

This review contains specific conclusions reached

by a number of scientists preparing the ICRU report 26 on Neutron Dosimetry for Biology and Medicine (1977)

and is also based on the discussions during a workshop

on Basic Physical Data for Neutron Dosimetry (1976).

The stimulating discussions with my colleagues

Drs. B.J. Mijnheer, H. Schraube and J. Zoetelief in the

final preparation of this paper are gratefully acknowledged .

References

1. ICRU report 26, Neutron Dosimetry for Biology and

Medicine ( 1977)

.

2. Broerse, J.J., Engels, A.C., Lelieveld, P . , Putten,

L.M. van, Duncan, W. , Greene, D. , Massey, J.B.,

Gilbert, C.W., Hendry, J.H. and Howard, A., Int. J.Radiat.Biol. 19_, 101 (1971).

3. Rossi, H.H., p. 129 in Proc. Symp. on Neutrons in

Radiobiology, Oak Ridge, CONF-691106 (1969).

4. Broerse, J.J. and Mijnheer, B.J., in Proc. VIII

Int. Congress on Radiation Protection, Saclay

(1976) . 5. Shukovsky, L.J., Amer . J . Roentgenol . 108 , 27 (1970).

6. Hussey, D.H., Fletcher, G.H. and Caderao, J.B.,

Cancer 34, 65 (1974)

.

7. Sinclair, W.K., p. 617 in Radiation Dosimetry,

Vol. Ill, Academic Press, New York (1969).

8. Bichsel, H., Eenmaa, J., Weaver, K. and Wootton,

P.B., p. 71 in Proc. Int. Workshop on Particle

Radiation Therapy, American College of Radiology,

Philadelphia (1976).

9. Roesch, W.C. and Attix, F.H., p. 1 in Radiation

Dosimetry, Vol. I, Academic Press, New York (1968)

10. Proc. of a Panel on In Vivo Neutron Activation

Analysis, IAEA, Vienna (1973).

11. Murray, B.W., Deutsch, O.L., Zamenhof, R.G. , Pettigrew, R.I., Rydin, R.A. and Brownell, G.L., p. 179

in Proc. Symp. Biomedical Dosimetry, IAEA, Vienna

(1975) . 12. Hatanaka, H. and Sweet, W.H. , p. 147 in Proc. Symp

Biomedical Dosimetry, IAEA, Vienna (1975).

13. ICRU report 13, Neutron Fluence, Neutron Spectra

and Kerma (1969)

.

14. Harvey, J.R. and Mill A.J., in Proc. 3rd Symp. on Neutron Dosimetry in Biology and Medicine, Neuherberg/Miinchen , to be published.

15. Alberts, W.G. and Knauf, K. , in Proc. 3rd Symp. on Neutron Dosimetry in Biology and Medicine, Neuherberg/Munchen, to be published.

16. Mijnheer, B.J., Broers-Challiss, J.E. and Broerse,

J.J., p. 423 in Proc. 2nd Symp. on Neutron Dosimetry in Biology and Medicine, Neuherberg/Munchen,

EUR 5273 (1975)

.

17. Colvett, R.D., Rossi, H.H. and Krishnaswamy , V.,

Phys. Med. Biol. 17_, 356 (1972).

18. Goodman, L.J., p. 65 in Basic Physical Data for

Neutron Dosimetry, EUR 5629 (1976).

19. Goodman, L.J., Colvett, R.D. and Caswell, R.S.,

p. 627 in Proc. 2nd Symp. on Neutron Dosimetry in

Biology and Medicine, Neuherberg/Munchen, EUR 5273

(1975)

.

20. Schraube, H., Morhart, A., Schraube, G. and Burger

G., p. 243 in Basic Physical Data for Neutron

Dosimetry, EUR 5629 (1976).

113
21. Zoetelief, J., Bouts, C.J., Engels, A.C. and Broerse, J.J., p. 249 in Basic Physical Data for Neutron

Dosimetry, EUR 5629 (1976).

22. Coppola, M., p. 271 in Basic Physical Data for Neutron Dosimetry, EUR 5629 (1976).

23. Hurst, G.S., Brit. J .Radiol . 27^, 353 (1954).

24. Broerse, J.J., p. 197 in Basic Physical Data for

Neutron Dosimetry, EUR 5629 (1976).

25. Smathers, J.B. , Otte, V.A. , Smith, A.R., Almond,

P.R. , Attix, F.H., Spokas, J.J., Quam, W.M. and

Goodman, L.J., Med.Phys., in press.

26. Law, J., Int . J . Appl . Radiat . and Isotopes 22_, 701

(1971).

27. ICRU, An International Neutron Dosimetry Intercomparison, in press.

28. Broerse, J.J., Burger, G. and Coppola, M. , p. 257

in Basic Physical Data for Neutron Dosimetry,

EUR 5629 (1976)

.

29. Alsmiller, R.G. and Barish, J., Neutron Kerma Factors for H, C, N, O, and Tissue in the Energy

Range of 20 to 70 MeV, ORNL/TM-5702 (1976).

30. Bassel, R.H. and Herling, G.H. , Radiat. Res, in

press.

31. ICRU report 14, Radiation Dosimetry: X-rays and

Gamma Rays with Maximum Photon Energies Between

0.6 and 50 MeV (1969)

.

32. Larson, H.V. , Phys.Rev. 112 , 1927 (1958).

33. Leonard, B.E. and Boring, J.W., Radiat. Res. 55, 1 (1973).

34. Dennis, J. A., Phys. Med. Biol. _18., 379 (1973).

35. Bewley, D.K., p. 5 in Proc. 2nd Symp. on Neutron

Dosimetry in Biology and Medicine, Neuherberg/

Munchen, EUR 5273 (1975).

36. Chemtob, M. , Lavigne, B., Noel, J. P., Chary, J.,

Nguyen, V.D. , Parmentier, N. and Fiche, C. , p. 95

in Basic Physical Data for Neutron Dosimetry,

EUR 5629 (1976)

.

37. Rohrig, N. and Colvett, R.D., p. 99 in Basic

Physical Data for Neutron Dosimetry, EUR 5529

(1976)

.

38. Williamson, J. and Watt, D.E., Phys . Med . Biol . V7_, 486 (1972).

39. Greene, D. , Major, D. and Redpath, T., Phys. Med.

Biol. 20, 244 (I975)

.

40. Goodman, L.J. and Colvett, R.D., p. 24 in Annual

Report on Research Project, USAEC Report COO-3243-

3 (1974) . 41. Attix, F.H., Theus, R.B. and Rogers, C.C., p. 329

in Proc. 2nd Symp. on Neutron Dosimetry in Biology

and Medicine, Neuherberg/Munchen, EUR 5273 (1975).

42. Cleland, M.R. and Wells, N., p. 397 in Proc. 2nd

Symp. on Neutron Dosimetry in Biology and Medicine,

Neuherberg/Munchen, EUR 5273 (1975).

43. Kuchnir, F.T., Vyborny, C.J. and Skaggs, L.S.,

p. 107 in Int. Symp. on Advances in Biomedical

Dosimetry, IAEA, Vienna (1975).

44. Lewis, V.E., Rossiter, M.J., Wood, J.W. and Young,

D.J., p. 133 in Basic Physical Data for Neutron

Dosimetry, EUR 5629 (1976).

45. Broerse, J.J. and Mijnheer, B.J., p. 275 in Basic

Physical Data for Neutron Dosimetry, EUR 5629 (1976).

114
NBS FACILITIES FOR STANDARDIZATION OF

NEUTRON DOSIMETRY FROM 0.001 TO 14 MeV

0. A. Wasson

National Bureau of Standards

Washington, D.C. 20234

The neutron sources available at NBS for use in neutron dosimetry are described in terms

of fluence, beam size, and background contamination. Operational details and neutron

fluence monitoring of the recently installed standardized beam line of the 3 MV Van de

Graaff Laboratory are given for the 200 keV to 1 MeV energy region along with plans for

measurements at 14 MeV. Measurements of the response of typical laboratory dose rate

meters and films to monoenergetic neutrons of accurately known fluence in the 250 keV

to 1 MeV energy region are given. The problem of the neutron fluence to dose equivalent conversion is discussed.

(Calibration; dosimetry; detection; faci li ties ; fluence ; flux; moderation; monitor;

neutron; shielding; sources; standardization)

Introduction

A large variety of calibrated neutron sources and

flux monitors exist at the National Bureau of Standards

which are useful for studies of neutron dosimetry.

These sources and associated energy regions are shown

in Fig. 1. The 140 MeV electron linear accelerator is

used to produce a pulsed neutron source at a rate of

approximately 1 x 10^^ n/sec. The neutron spectrum is

continuous from thermal energies to over 14 MeV and can

be tailored for different experiments. Various evac- uated drift tubes with lengths from 2m to 200m permit

the measurement of neutron energies by the time -of

flight method. Beam sizes are variable, depending upon

the collimation used.

The next group of sources are associated with the

NBS reactor. The filtered beam facility produces mono- energetic beams by transmission of the reactor fission

spectrum through selected filtering materials. The

three beams in operation include a 2 keV beam from a scandium filter, a 24.3 keV beam from an iron-aluminum

filter, and a 144 keV beam from a silicon filter. The

investigation of the response of various detectors and

dosimeters to various energy neutrons is aided by a scanning table and human phantom. A detailed presentation of the neutron dosimetry measurements on this

facility are given in a paper in this symposium by

R. B. Schwartz and will not be given here.

There is also a cavity source located in the

graphite thermal column of the reactor which produces

a fission spectrum with a strength of approximately

3 X lO'''''^ n/sec which is useful for studies of materials

VAN DE GRAAFF 1 1

CALIBRATED SOURCES

REACTOR B B S

LINAC

I

1

1 10-2 10-1 10-° 10' W W W 10= W 10' 10» NEUTRON ENERGY, eV

dosimetry. The details of another source (ISNF) in

this column which produces a neutron spectrum similar

to that of the Liquid Metal Fast Breeder Reactor are given in papers by D. M. Gilliam^ and C. M. Eisenhauer^

at this symposium. The neutron field has a mean energy of 0.8 MeV and flux of ~ 10® n/cn? sec.

The next class of neutron sources are those

produced from radioactive decay. The national standard Ra-Be source (NBS-I) has a source strength of

10 n/sec with an uncertainty of - 17.. This is the

primary standard for all fast neutron source and field

strength measurements. There are also several ^^^Cf

spontaneous fission sources with strengths between 1 and 7 x l(f n/sec which can be calibrated within an uncertainty of - 1.27. based on NBS-I. These sources

emit a broad energy spectrum in the 0.2 to 5 MeV

region.

Other neutron sources are obtained at the NBS

3 MV positive ion Van de Graaff laboratory. Monoenergetic neutron beams in the 0.1 to 2.2 MeV energy

region are produced by means of the Li(p,n)^Be and

^H(p,n)^He reactions. The 14 MeV region is covered

by the ^H(^H,n)*He reaction. Production of 144 keV

neutrons allows a comparison with experiments done

with the silicon filtered beam at the reactor. The

remainder of this paper is devoted to the properties

of the recently installed standard neutron beam line

and to the results of the first measurements of the

response of several neutron personnel monitors to low level neutron exposures.

Van de Graaff Experimental Geometry

The layout of the three beam lines at the Van de

Graaff Laboratory is shown in Fig. 2. The beam to the right is used for other testing and physics

measurements. The center beam line is used for

the calibration of the Black Detector neutron

flux monitor by means of the associated particle

technique. The standard neutron beam line shown

on the left side of the figure became operational

in November of 1976. A more detailed view of this

beam line is shown in Fig. 3.

Fig. 1 Neutron Sources in Standards Program

ns
ASSOCIATED

PARTICLE

w

standard\\

BEAM

TEST AND

FUNDAMENTAL

PHYSICS

POSITIVE ION BEAM

FROM

3 MV VAN DE GRAAFF

Fig. 2 Van de Graaff Experimental Area

MOVABLE SHIELD BEAM

,PIPE

REMOVABLE

INSERTS

BLACK

PRECISION DETECTOR NEUTRON BEAM COLLIMATOR MONITOR

4.6" HALF ANGLE

SECONDARY

MONITORS

LONG COUNTER

Fig. 3 standard Neutron Beam Line

beam direction. Tests have shown that the back angle

detector is the better arrangement. The secondary

monitors are calibrated by means of the Black Detector

using a pulsed proton beam. Continuous beam operation

is used for dosimetry measurements in order to increase

the neutron flux by a factor of 5.

Black Detector Neutron Flux Monitor

Our initial experiments have been limited to the

neutron energy region from 200 keV to 2 MeV. The

neutron sources were the Li(p,n) Be and '^H(p,n)'^He

reactions. The neutron source is surrounded by a large neutron shield in order to reduce room background. The shield is mounted on an air table in order

to produce easy access to the end of the beam pipe for

target charges. The neutron beam is collimated to a cone with a 4.5° half angle by means of a removable

lithium loaded polyethylene insert. This angle provides a 19 cm diameter neutron beam at a distance of

120 cm from the target. The primary neutron flux

monitor is the Black Detector located in a shield

approximately 6 meters from the target. The beam size

incident on this monitor is determined by a precision

machined collimator 30 cm long with a 4.445 - 0.010 cm diameter cylindrical hole. This shield is also mounted

on an air table for easy movement. The proton beam

energy is calibrated by measuring the Li(p,n) Be and

^H(p,n)^He thresholds and checked by measurements of

carbon transmission near the 2077 keV resonance.

The neutron flux at the dosimeter position, r, is

determined from the yield of the Black Detector by the

expression

The Black Detector is a cylindrical shaped

plastic scintillator viewed by a single photomul tiplier

tube as is shown in Fig. 4. The scintillator is 12.5 cm in diameter with a length of 15 cm and a 5 cm diameter reentrant hole 2.5 cm deep. The purpose is to completely absorb the incident neutron in the scintillator so that the light output is proportional to incident neutron energy. Hence the name, Black Detector, as christened by Poenitz* . A detailed presentation of the operation and calibration of this monitor

is given in another contribution to this symposium by

M. M. Meier^

.

The neutron detection efficiency is measured by means of the associated particle method and also calculated by means of a Monte-Carlo type program. The

results for 500 keV and 750 keV neutrons are shown in

Fig. 5. The points represent the measurement while the

solid curves are from the calculation. Since the two methods agree within - 17., the efficiency is taken from

calculation for intermediate energies. The detector

efficiency is greater than 907. for neutron energies

between 200 keV and 1 MeV.

V(r) =

BD

6 T(R - r) A

where ^(r) is the neutron flux at distance r R is the distance from source to exit of

Y collimator

BD is the Black Detector counting rate

£ is the Black Detector efficiency

A is the effective area of the collimator

T(R - r) is the air transmission between the

Black Detector and the position r.

For thick dosimeters which absorb a large fraction of the beam, it is necessary to use secondary

monitors placed outside the direct beam for neutron

flux monitoring. The monitors include a long counter

placed at 90 to the beam direction and a BFa counter

located inside the shield at approximately 180 to the

^ LIGHT PHOTON O RECOIL PROTON O NEUTRON PHOTO CATHODEANODE

NEUTRON BEAM OPLASTIC SCINTILLATOR NE 110 PHOTOMULTIPLIER TUBE RCA 8854

Fig. 4 Black Detector Operation

116
T

1 1 1 1 r

20 40 60 80 100 120 140 160 CHANNEL NUMBER Fig. 6 Black Detector Time of Flight Yield

Fig. 5 Black Detector Response

In addition to detecting neutrons', the monitor is

also an efficient detector of y rays, as is shown in

Fig. 6. Here is displayed a time of flight spectrum

with the Van de Graaff operating in the pulsed mode.

The narrow peak is due to the X-rays from the proton

beam hitting the Ta beam stop, while the broad peak is

from 560 keV neutrons. The width of the neutron peak

is determined by the energy spread of the neutron beam.

This gives us a means of measuring the thickness of the

target as well as the neutron energy. Most of the flat

part of the spectrum^ showed by the dashed line, is the

ambient room background which is present with the beam

off. Because of the strong X-ray peak, the monitor is

used in the pulse mode of operation in order to separate neutrons from y rays and to calibrate the secondary monitors which are relatively insensitive to y rays

.

Table 1

Measured neutron flux at a distance of 120 cm

for a 5nA proton beam incident on a 20 keV

Li target

En, keV cp, n/cm sec DE, millirem/hr'

250

500

1000

600

2100

1200

30

195

155

National Council on Radiation Protection and

Measurements Report No. 38, Table 2 (1971)

After corrections for the effective area of the

collimator beam hole and air transmission the neutron

flux is determined within - 27o at the dosimeter position for neutron energies between 200 keV and 1 MeV.

The uncertanity increases outside of this region.

For operation at 14 MeV, we plan to calibrate the same detector bj using the associated particle technique with the H(^H,n)'*He reaction and by extending

the Monte-Carlo calculations to higher energies. The

detector efficiency will decrease to approximately

307o. The shield around the target will be replaced

by a more massive shield which is under construction.

Neutron Beam Parameters

The dose equivalent in units of millirem per hour is

taken from the NCRP report . The measured y ray background dose equivalent is 1.5 millirem/hr. The low

energy < 1 keV) neutron background flux is less

than 0.87o as measured with a ^He gas proportional

countero The neutron fluxes are low so that a practical upper limit to irradiations is approximately

10 rem.

At 14 MeV neutron energies the expected rates for

a 5nA deuteron beam on a tritium target are a flux of

approximately 6 x 10^ n/cm sec and a dose equivalent

of approximately 1.4 rem/hr.

Most of the dosimeter measurements to be reported

in this paper were done at a distance of 120 cm from

the neutron target. The beam diameter is 19.5 cm with

a uniformity of - 17.. The neutron flux and dose equivalent for a 5pA proton beam on a 20 keV thick Li

target are given in Table 1 for several energies.

The first use of the calibrated neutron fluence

of this facility for dosimetry purposes is to study

the response of typical personnel type neutron monitors to monoenergetic neutron fluences of 250, 500,

and 1000 keV energy at low dose rates.

117
Measurements with Moderating Detecto rs

The first type of detector consisted of a hand

held paraffin moderating cylinder with a BF3 counter

in the center arbitrarily selected from the group of

instruments on hand. The output of the device con- sisted of a current meter calibrated in units of neutrons per cm^ per sec. The results of the measurements are shown in Fig. 7. The cadmium covered cylinder was 12 cm in diameter and 12 cm in length with the

neutron beam incident on the front face. The neutron

flux varied between 10 and 100 n/cn? sec for various

positions and neutron energies. The ratio of the

meter flux reading to the incident flux is shown for

incident neutron energies of 250, 500, and 1000 keV.

The detector was calibrated by means of an Am-Be neutron source. The error bars indicate the statistical

fluctuation of the current meter dial. The response

at 1 MeV is approximately 307, less than that claimed by

the manufacturer who states that the response should

vary less than 107. throughout this energy region.

INDICATED FLUX / INCIDENT FLUX

POLYETHYLENE MODERATOR

1.0

qO.8

UJ>

(r

uso

0.4

1 1 1 1 1

!

—12 cm —

\ ^

V///A

1 1 1

Fig.

0 200 400 600 800 1000

NEUTRON ENERGY, keV

7 Ratio of observed neutron flux to incident

flux for moderating cylinder

The next measurements were done with larger moderating cylinders from two different manufacturers as shown in Fig. 8. These are the Andersson-Braun type

detectors which are supposed to have a uniform dose

equivalent response from thermal to 10 MeV neutron

energies. The detectors are composed of polyethlyene

cylinders 21 cm in diameter and a length of 24 cm with

a small BF3 proportional counter in the middle. The

moderated flux is tailored by means of a boron absorber with holes in order to produce a response which

approximates the neutron dose equivalent in tissue.

The neutron beam was incident on the front, flat end

of the counter as is shown in the figure. Both detectors were calibrated by an Am-Be neutron source at a distance of 1.5 m using a dose equivalent to flux con- stant of 3.49 x 10 rem/(n/cm^) as given by

Nachtigall .

5 10 15 20 DISTANCE, cm

25

Fig. 8 Diagram of larger moderating neutron

rem counter

OBSERVED TO CALCULATED RESPONSE

L50

.25

LOO

O

t-0.75 <

0.50

0.25

T

O DETECTOR I • DETECTOR 2

250 500 750 OOO

NEUTRON ENERGY, keV

Fig. 9 Ratio of observed to calculated

response for two rem counters

The results of the measurements are shown in Fig.

9. The ratio of the observed dose equivalent to that

calculated from the incident neutron fluence, is plotted as a function of neutron energy for the two detectors. The response of both detectors follows the same general shape, although they differ in absolute calibration by 257.0 Two values are shown for each detector

at 250 keV. These result from different extrapolations

of the neutron fluence to dose equivalent relationship

at that energy.

118
Figure 10 shows the dependence of dose equivalent on neutron fluence as a function of neutron

energy. The points connected by the solid lines are from the 1971 report by the National Council on Radiation Protection and Measurements while those con- nected with the dashed line are from the 1957 NBS

Handbook No. 63^ . These recent results differ by

nearly 507. from the earlier results which were in use at the time of the Andersson-Braun detector

invention and the Am-Be source calibration. Since

no values are given for 250 keV, the value at this

energy is obtained by extrapolation from the points

at 100 and 500 keV. Due to the rapid variation in

this region, the value obtained is, of course, dependent on the method of extrapolation and can easily

vary by a factor of two. The circle on the graph

is a result of logarithmic extrapolation which produces a better fit to the measurements as is shown by

the lower datum points at 250 keV in Fig. 9. The

higher values result from a linear extrapolation.

Thus logarithmic extrapolation should be used in the

100 to 500 keV interval for the data in the NCRP

report and not linear extrapolation as recommended

in the report.

FLUENCE TO DOSE EQUIVALENT CONVERSION

10=

10'

T T n—I—I MM.

• TABLE 2 NCRP 38 (1971) -

o EXTRAPOLATION A NBS HANDBOOK 63 (1957)

0.1 1.0 10 NEUTRON ENERGY , MeV

Part of the difference between observed and

incident dose equivalent in Fig. 9 is due to the use

of different values of the neutron fluence to dose

equivalent conversion factor for the detector calibration and the incident fluence. The older 1957

results were used for the former while the 1971

results were used for the latter. Figure 11 shows

the results obtained if the older fluence to dose

equivalent conversion is used for both the calibration

and the fluence measurement. For detector No. 1 the

response agrees with that predicted from the incident

fluence while the response of detector No. 2 is approximately 257. low. Since the error bars indicate

the statistical errors only, the 257. difference between the responses of the two similar detectors is an indication of the systematic errors involved in

neutron dose equivalent measurements with typical

neutron dose meters.

Measurements with Nuclear Test Films

The Nuclear track film used in dosimeters has an energy threshold at approximately 500 keV and an exposure threshold of approximately 100 mrem. As a check of these limits, film dosimeters were exposed to 250 keV, 500 keV, and 1 MeV beams with exposures ranging from 7 to 120 mrem. The films were exposed in

groups of six in both free air and while mounted on

the chest of a plastic phantom. The films were then

treated in the usual manner by the processing laboratory.

The processing laboratory scanned the film visually

and observed no exposure for the 250 keV and 500 keV

beams. This is as expected for a 500 keV energy threshold. Since the results for 1 MeV neutrons were the same, within statistical error, for both the phanton

and free air film backing, only the results observed

with the phantom are shown in Table 2. The first col- umn indicates the incident dose equivalent in units of

millirem determined from the fluence measurement, the

second column lists the total number of observed tracks

in each group of 6 films, while the final column lists

the exposure deduced by the processing laboratory. The

listed error is only that from the finite number of

observed tracks.

Fig. 10 Fluence to Dose-equivalent

conversion factor

1.25

1.00

UJQ^ 0.75 V.

LlJ O

Q 0.50

UJ>

01UJVI

DO 0.25 O

DETECTOR RESPONSE

T T

o DETECTOR I

• DETECTOR 2

NBS HANDBOOK 63 (1957)

250 500 750 1000 NEUTRON ENERGY, KeV

The observed dose agrees, within the - 127. statistical error, with the exposure at the 110 mrem level

but is systematically high for the lower exposures.

These results confirm the observation that the typical

film dosimeters used in personnel monitoring do not

function for neutron energies below approximately

500 keV and for exposure levels less than approximately

100 mrem. Different type monitors are required for

these regions

Table 2. Film irradiation results for 1 MeV neutrons

Incident

Dose Equivalent mrem Observed Tracks

Observed

Dose Equivalent mrem

7.7 36 56 + 9

15.4 28 43 + 8

30.3 43 66 + 10

110 77 119 + 14

Fig. 11 Ratio of observed to calculated

dose equivalent for two rem counters. Conversion factor is

taken from ref. 9

119
Summary

A calibrated neutron fluence beam is available at

the NBS Van de Graaff with an accuracy of - 2°L in the

energy region from 200 keV to 1.0 MeV with low background. This large sized beam is ideal for testing

of instruments at low flux levels which is becoming

more important as permissable personnel dose levels

are lowered. Examples of the use of this beam for

checking the response of typical radiation monitoring

equipment, including the Andersson-Braun type monitor,

were given. This facility along with the other neutron sources available at NBS offers a unique opportunity for standardization of neutron dosimetry for

neutron energies below 15 MeVo

References

1. Ro B. Schwartz, Proceedings of International

Specialists Symposium on Neutron Standards and

Applications, Paper G7 , (1977).

2. D. M. Gilliam, Op. Cit., Paper H4.

3. C. M. Eisenhauer, Op. Cit., Paper 12.

4. W. P. Poenitz, Nucl. Instr. and Meth. 109,

413 (1973)«

5. M. M. Meier, Proceedings of International

Specialists SymDosium on Neutron Standards

and Applications , Paper G2, (1977).

6. "Protection Against Neutron Radiation", National

Council on Radiation Protection and Measurements

Report Number 38, Table 2, (NCRP Publications,

Washington, D. C, 1971).

7. I. 0. Andersson and J. Braun, "A Neutron rem Counter With Uniform Sensitivity from 0.025 eV to

10 MeV", Neutron Dosimetry, Vol 2 , p. 87 (International Atomic Energy Agency, Vienna, 1963),

Proceedings of a Symposium, Harwell, 1962.

8. D. Nachtigall, "Average and Effective Energies,

Fluence-Dose Conversion Factors and Quality

Factors of the Neutron Spectra of Some (a,n)

Sources", Health Physics _13, 213 (1967).

9. "Protection Against Neutron Radiation up to 30

Million Electron Volts", National Bureau of

Standards Handbook 63, (Superintendent of

Documents, Washington, D. C, 1957).

120
INTERNATIONAL NEUTRON DOSIMETRY INTERCOMPARISONS

R. S. Caswell

National Bureau of Standards

Washington, D.C. 20234

Three recent international neutron dosimetry intercomparisons are discussed: the International Neutron Dosimetry Incompari son (INDI)

sponsored by the International Commission on Radiation Units and

Measurements (ICRU); the European Neutron Dosimetry Intercompari son Project (ENDIP) sponsored by EURATOM; and the intercomparison carried

out by the centers doing neutron radiotherapy. Physical dosimetry

to an accuracy of two or three percent is desired in order to achieve

a generally-accepted 5% accuracy in dose to the tumor. In general it

is found that 3% accuracy has not been achieved by the intercomparison

participants; however, the radiotherapy centers agree on an arbitrary

(but not absolutely known) scale within this uncertainty.

(Standards, medical, neutron dosimetry, intercomparisons, radiation

effects)

.

Introduction

Perhaps the first question one might ask is: Why

do intercomparisons at all? Some of the chief reasons

are: (1) to verify the international measurement scale;

(2) to obtain information on systematic errors; (3) to compare measurement methods or instruments; (4) to compare corrections used by different laboratories; (5)

to evaluate the state-of-the-art in measurement; and

(6) to motivate scientists to make better measurements.

In reference to the first reason above, it is very

important that 500 rads of absorbed dose to a tumor be

the same in one hospital as in another so that com- parison of clinical results is on a uniform basis, and

one hospital may benefit from the experience of another

hospital. It is also important to a national standards

laboratory to verify that errors have not been made in the laboratory's own absolute measurements--compari son with one or more other standards laboratories gives some

degree of assurance that no blunder has been made, and

that the international measurement system is uniform,

even though it may not be known absolutely to the ac- curacy to which uniformity can be established. If

national standards laboratories do not offer standards

for a particular kind of measurement, then intercomparisons represent a way of testing the uniformity of

those institutions that constitute the international

measurement system.

Evaluation of systematic errors, unlike r*andom errors for which there is a good theory, is very difficult and is usually done by an estimate or guess.

Intercomparison provides a way for getting at systematic

errors. If all laboratories using independent methods

get the same result we gain confidence that the systematic errors of each method are small. On the other

hand, if one laboratory using a particular method is

20% higher than all other laboratories which are in agreement to, say, 3%, we tend to look for errors in this one method or in this laboratory's procedures.

A scientist participating in an intercomparison

to some extent has his vulnerabilities exposed. He is

therefore likely to try to do the best possible job of

measurement to make sure that his numbers stand up well

in the comparison wi/th his peers.

Three international neutron dosimetry intercomparisons will be discussed here, although there have

been a number of others: (1) The INDI (International

Neutron Dosimetry Intercomparison) comparison was

sponsored by the International Commission on Radiation

Units and Measurements (ICRU) and was carried out by

Leon Goodman and collaborators at the RARAF Accelerator

facility at Brookhaven National Laboratory -.in 1973 with

14 international participating groups.'' ' (2) The

ENDIP (European Neutron Dosimetry Intercomparison

Project) comparison was sponsored by EURATOM and carried

out at TNO Rijswijk, Netherlands for dosimetry for neutron therapy, and at GSF Neuherberg (Munich) for neutron

protection, in 1975.4,5,6 jhe Radiotherapy Centers'

intercomparisons have been carried out on a bilateral

basis with participants from two laboratories performing

measurements at a common radiotherapy facility.''^

INDI Comparison

The physical arrangement for the INDI intercomparison

is shown in FigCire 1. The ion beam from the accelerator

TARGET

5cm DEPTH.

lOcm DEPTH - 20cm DEPTHACCELERATOR

1^ BEAM

0cm REF — 5cm (MONITOR ION CHAMBER)

20 cm FRONT FACE -30cm(KERMA)

"[^ K PHANTOM

PRECISION LONG

COUNTER''

(REMOVABLE)

30cm CUBE

ALIGNMENT'''^ TELESCOPE

(TRANSLATABLE)

-200 cm FRONT

FACE

.ALIGNMENT

TELESCOPE

(FIXED)

INTERCOMPARISON ARRANGEMENT, TOP VIEW

Fig. 1. Arrangement for the INDI intercomparison.

121
was incident upon the target where the neutrons were

produced. At 5 cm downstream from the target was a transmission ionization chamber which monitored the

total radiation (predominantly neutrons) in the

experiments. Air measurements were carried out at 30 cm from the target. A 30 cm x 30 cm acrylic box filled

with distilled water served as a phantom and could be

located with its front face 20 cm from the neutronproducing target. Water was used, rather than tissueequivalent liquid, to avoid problems of change of

composition with time. Multiple monitoring was carried

out, including a Precision Long Counter along the beam

line at 200 cm, and 3 BF^ counters at various depths

In moderators, which were sensitive to changes in neutron spectrum. Positioning was done optically using

an alignment telescope with such accuracy that errors

due to position could be neglected. A photograph of

the experimental area is shown in Figure 2.

Fig. 2. Photograph of the experimental area used for

the INDI intercomparison at the Brookhaven National

Laboratory RARAF facility.

The principal monitor for the experiment is shown in Figure 3 and is a parallel-plate tissue-equivalent-wall

GAS CAVITY

CAVITY WALLS

TISSUE EQUIVALENT

PLASTIC (MUSCLE I

STEM, TISSUE EQUIVALENT PLASTIC (MUSCLE)

CONNECTOR - BLOCK, ALUMINUM

ELECTROMETER

CONNECTOR, ALUMINUM

PHANTOM I I

transmission chamber. The principal gamma-ray monitor

is a GM counter^, also shown in Figure 3, and is located

below the beam line at a horizational distance of 15 cm. A photograph of the transmission ionization chamber is

shown in Figure 4.

10 CENTIMETERS

Fig. 3. Transmission ionization chamber monitor and

location of phantom and GM counter used as a gamma-ray

monitor in the INDI comparison.

Fig. 4. Photograph of transmission ionization chamber

monitor used in INDI experiment.

Measurements were carried out by participating groups

for 11 measurement conditions. Five measurement condi-, ^ tions were in air: 15.1 MeV, 5.5 MeV, 2.1 MeV, 0.67 MeV

and for Cf-252 neutrons. In addition measurements were made at three depths in the phantom for each of the two higher neutron energies, 15.1 and 5.5 MeV. The measurement conditions and reactions are summarized in Table 1. A summary of the INDI results in air is given in Figure

5. Note the larger spread in the results at 0.67 MeV

where the dose was low. Results are grouped tightly at the two higher energies. At the lower energies the

tendency of the CpH. counter and the C2H4 ionization

chamber to read low is evident. This is believed to be dui to the necessity of applying a conducting film, usually

graphite, to the polyethylene to make the chamber wall

conducting. Secondary particles ejected by neutrons may

be absorbed in this conducting coating of the chamber

wall. The tissue-equivalent ionization chambers and

acetylene-polystyrene equivalent chambers (C^H^) are intrinsically conducting, so this problem does not exist.

The silicon diode and the precision long counter are calibrated independently, but do not yield results very

different from those of the tissue-equivalent ionization

chambers.

A more detailed look at the measurements at 15.1 MeV

in air is given in Figure 6. Note that measurements were made over a period of roughly a year, requiring very good

Van de Graaff accelerator radiation field stability. The

stability of the calibration fields used in the experiment

is believed to be good to 2%. Note that RARAF made mea- surements at 4 times during the procedures, and these all

agree within a spread of 1%. One can see that the mea- surements are generally in agreement with the uncer- tainty quoted by the experimentalist. However, one should not be content with this situation, because the

absolute uncertainties are relatively large, typically

122
Table 1. Neutron energies and kerma rates

Reaction Neutron

energy

MeV

Percent

energy

spread

(±)

Nominal maximim total

tissue kerma rate

in air at 30 cm

Gy h'\^00 rad h"^)

Approximate percent

gamma-ray tissue kerma,

relative to total

tissue kerma

in free air

T(d,n)^He 15.1 4 0.40 5

D(d,n)^He 5.5 7 0.80 5

T(p,n)^He 2.1 5 0.20 5

T(p,n)^He 0.67 20 0.05 5

fission

252

Cf spectrum 0.06^ 40

^About 2 mg.

O B

9 4

I I I I

I a67

7'

> I I I I

1—

r

TE V

CzH, a

H, Cir • CzHj 0

SI DIODE o PLC «

ft

1.0 2.0 5.0 NEUTRON ENERGY/ MeV

Fig. 5. Summary of INDI comparison results i n air.

Fig. 6. Summary of INDI measurements at 15.1 MeV from

December 1972 to November 1973. Solid circles, solid

squares, erect hollow triangles and open circles all

refer to tissue-equivalent ionization chambers with

di fferent systems for measuring the gamma-ray component

of the mixed neutron and gamma-ray radiation field.

Inverted open triangle is for C2H2 ionization chamber,

erect and inverted solid triangles are for C2H4 ionization chambers. Open square is for the C2H^ proportional

counter, diamond is for the Precision Long Counter, and

the circle with dot in the center is for the Si diode.

7% or 8%. Since many of the measurers used similar

systems, for example the tissue-equivalent ionization

chamber, one might expect that these methods would agree

to much better than the quoted uncertainty. However,

this is not the result of the. present comparison.

Figure 7 shows the results of the measurement of

gamma-rays in the presence of neutrons, expressed as a percentage of the neutron kerma in air. Several different systems of gamma-ray measurement were used, the

GM counters and the film having generally the lowest

neutron sensitivity. Some negative results are shown

particularly at 15.1 MeV which indicate that the

neutron sensitivity assumed for the gamma-ray detector

was probably too high. Note that for Cf-252 (for which

one should read the right hand scale on the figure) the gamma

rays are a much larger' fraction of the neutron dose than

in the other cases, and the agreement of the various

measurement methods for the gamma rays is in fact much

better. Measurements at 15.1 MeV in air and in phantom

are summarized in Figure 8. Note that this is not a depth dose curve, since the phantom is not present

during the measurements in air. A tendency for the

polyethylene-ethylene dosimeters to read low in the lower

neutron energy spectra associated with depth in the

phantom is evident. Finally, in Figure 9, are shown

measurements in air and phantom at 5.5 MeV. Notice

the spread in results at 20 cm depth, where the dose

rates are rather low, making it difficult to make

accurate measurements with the ionization chambers (which

123
• GM COUNTERS

Mg-a Al-Ar CHAMBERS

* C-CO^ CHAMBERS TLO * V FILM 0 67 MeV

2.1 MeV

i i

5.5 MeV

15.1 MeV —

I

_LL J

I

I I I I I 2 5 NEUTRON ENERGY/ MeV

5.5 MeV NEUTRONS

IN AIR AND PHANTOM

TE 17 CjH, O

CjH, Clr • CjHj 0

Si Diode o PLC

5 10 15 DEPTH IN PHANTOM/cm

Fig. 7. Results of gamma-ray measurements in the mixed Fig. 9. INDI measurements in air and at 5, 10, and 20

radiation field. For Cf-252 read right-hand scale, for cmdepths in the phantom for 5.5 MeV neutrons,

all other energies read the left-hand scale.

are mostly used in this experiment).

15.1 MeV NEUTRONS

IN AIR AND PHANTOM

AIR 5 10 15 DEPTH IN PHANTOM/cm

Some of the conclusions of the experiment were:

(1) the spread in results is greater than the desired

three percent, is roughly 10%, but no methods are in

extreme disagreement with the others. (2) Uniform

corrections for tissue equivalent ionization chambers

were tried in an analysis by a statistical expert,

J. W. MLiller of the Bureau International des Poids et Mesures (International Bureau of Weights and Measures)

in Sevres, France. The essential result of this analysis

was that uniform corrections, for example uniform W /W

ratios, helped very little. This was a surprise. f3)^

Ranking analysis (ranking in the sense of determining

which laboratory tends to read high on all eleven mea- surements conditions, and which laboratories tend to read

low) was tried by Goodman^O and showed groups of 11 laboratories with similar results containing 1,2,6,2

laboratories in the sense from high reading to low

reading. That is, one laboratory tended to read significantly higher than all others, then a group of two

read somewhat lower, etc. This analysis was carried

out on the response of the chamber with the correction

factors removed, and showed that there are systematic

procedure differences between laboratories--that is the

laboratories do not get the same "dial reading" for

similar instruments in a given radiation field.

ENDIP Intercomparison

Fig. INDI measurements in air and at 5, 10, and 20 cm depths in the phantom for 15.1 MeV neutrons. Symbols

are the same as in Figure 5.

Some preliminary results for the ENDIP Intercomparison, are shov;n in Figure 10.^ These measurements were carried out at GSF-Neuherberg in 1975, the example here

being for 15.1 MeV. An analysis of these results is

shown in Table 2. Note in this case the better agreement

in air at the low neutron energies, and also that the

3% spread has not been achieved. Table 3 compares the

results by 3 participants, TNO, GSFM, and CENF who par124
KERMA VALUES (RAD PER 10^ MONITOR UNITS) OBTAINED AT GSF

FOR 15.1 MeV NEUTRONS

ticipated at both radiation facilities. The apparent

difference between these three laboratories, which is consistent throughout the measurements, has not been

resolved satisfactorily to my knowledge.

April

& lolal kerma O neu Iron kermo gommo kerma

My

Radiotherapy Center Comparison

The Radiotherapy Center comparison was carried out in

the Spring of 1974 between three U. S. Radiotherapy

Centers and the Medical Research Council-Hammersmith

Hospital facility in London. The results are given in

Table 4.° Note the high degree of agreement between

the three U. S. therapy centers (University of

Washington, U of W; Naval Research Laboratory, NRL; and

M.D. Anderson Hospital - Texas A&M Variable Energy

Cyclotron, TAMVEC). The lower reading of MRC was later

adjusted upward by 8% in January 1975. The measurements

at depths in phantom are largely consistent with the

measurements in air. Further laboratories were brought

into this intercomparison in the Autumn 1975 and February

1976. (See Table 5). These laboratories were CHHRI

(Christie Hospital-Holt Radium Institute in England);

Louvain, Belgium; and Chiba, Japan. Allowing for uniform

parameters in the case of CHHRI, one can see that these

laboratories are on a very closely uniform measurement

system.

Fig. 10. Results of ENDIP comparison measurements for

15.1 MeV neutrons carried out at GSF-Neuherberg , West

Germany in 1975.

TABLE 2 NUMBER OF GROUPS ENDIP-GSF WITH RELATIVE DIFFERENCE, ax, FROM THE MEAN

Neutron energy 5%<axi 10% Ax> 10%

15.1 MeV ^N 5/11 3/11 3/11

So. 5/11 3/11 3/11

5.25 MeV <N 7/10 2/10 1/10

So. 8/10 2/10 0/10

2.1 MeV ^N 9/11 0/11 2/11

So. 9/11 1/11 1/11

0.57 MeV I^N 9/1

1

0/1 1 2/11

So, 9/11 1/11 1/11

C_T neutrons ^N 7/8 1/8 0/8

So. 8/8 0/8 0/8

TABLE 3 VALUES Of ABSORBED DOSE AND KERMA (RELATIVE TO THE MEAN) AS DETERMINED BY THREE PARTICIPANTS AT GSF AND TNO FOR

15 MeV NEUTRONS

Conditions TNO GSFM CENF

GSF, free In air ^N 0.97 1 .08 1 .17

So. 0.97 1 .08 1 .17

TNO, free in air ^N 0.99 1 .05 1.05

So. 1 .00 1 .06 1 .05

TNO, 5 cm depth °N 0.94 1 .06 1.11 0.94 1 .08 1.10

TNO, 10 cm depth °N - 0.94 . 1 .05 1 .11

°.o. 0.94 1 .07 1 .10

TNO, 20 cm depth ^N 0.91 1 .05 1 .17

0.94 1 .07 1.14

Conclusion

The Radiotherapy Centers, following the intercompari sons

,

are now largely on the same measurement scale which is

well-known relatively, but not absolutely. It is the

author's belief that more efforts in the laboratory are needed to correct possible systematic errors and to understand these consistent differences that appear from

laboratory to laboratory throughout many different mea- surement conditions, before further international com- parisons will be worthwhile. However, following an effort at eliminating systematic errors, international

comparison would be justified to see if the attempted

corrections of systematic errors have in fact worked.

In view of the importance of this problem, it seems appropriate that the National Standards Laboratories

should be developing standards and providing calibration

services for neutron dosimetry. Such calibrations could

follow three largely independent routes to see if the same calibration is achieved following each route. These

routes are: (1) calibration of neutron dosimeters with

known fluences of monoenergetic neutrons using kerma

factors to obtain the dose (a subject very dependent on the work being discussed at this conference); (2) using

a gamma-ray calibration of an ionization chamber and

appropriate ratios of W, stopping power and kerma factors

to determine the dose; and (3) use of a tissue-equivalent

calorimeter with appropriate corrections for calorimetric

defect to determine the dose or kerma in the radiation

field.

125
Table 4. Intercomparison results: ratios of measurements with respect to TAMVEC measurements during the Spring

of 1974.^

Participants MRC U of W NRL U of W NRL TAMVEC TAMVEC TAMVEC TAMVEC TAMVEC

Location MRC U of W NRL TAMVEC TAMVEC

Beam 16 MeV d+Be 21.5 MeV d+Be 35 MeV d+Be 50 MeV d+Be 50 MeV d+Be

Tissue kerma

in air

0.89'' 1.04^ 0.99 0.99 1.00

Dose at depth 0.86^^ Q dmax

0.86^ @ 10 cm

1.01 @ dmax

1.02 (3 10 cm

0.99 & dmax

0.99 010 cm

0.99 (3 2 cm

0.99 @10 cm

0.99 0 2 cm

0.98 (310 cm

Photon

Calibration

1.01

(8 MeV)

1.01

(60co)

1.03

(137cs)

1.00

(60co)

0.99

(60co)

a Data suspicious owing to leakage problems,

b MRC excludes -3% gamma component

c MRC excludes 6% gamma component

d MRC excludes ~ 97„ gamma component

Table 5. Intercomparison results: ratios of measurements with respect to TAMVEC measurements during the Autumn of 1975 and February 1976."

Participants CHHRI MRC NRL Louvain Chiba Chiba

TAMVEC TAMVEC TAMVEC TAMVEC TAMVEC TAMVEC

Location CHHRI MRC NRL Louvain TAMVEC TAMVEC

Beam 15 MeV d+T 16 MeV d+Be 35 MeV d+Be 50 MeV d+Be 30 MeV d+Be 16 MeV d+Be

Tissue

kerma

in air

0.95^ 1.01 1.00 1.01 1.01

Dose at depth

0.97 (3 1 cm

0.96 (3 5 cm

0.97 (310 cm

1. 02 (3 2 cm 1.02 @ 10 cm

0.97 (3 5 cm 0.96 (310 cm 0.97 (315 cm

0.96 (320 cm

Photon

Calibration

1.00

(137cs)

1.00

(60co)

0.99^

(60co)

a Choice of parameters account for 4% of the 5% difference (see text).

b Ratio of calbration made in Japan to calibration made at M.D. Anderson Hospital.

126
References

1. ICRL) Report 29, An International Neutron Dosimetry

Intercomparison , International Commission of Radiation Units and Measurements, Washington, D.C. 20014,

U.S.A. , to be published (1977).

2. R. S. Caswell, L. J. Goodman, and R. D. Colvett,

International Intercomparison of Neutron Dosimetry,

in Radiation Research, Biomedical, Chemical, and

Physical Perspectives , 0. F. Nygaard, H. I. Adler,

and W. K. Sinclair, eds.. Academic Press, New York

(1975), p. 532.

3. L. J. Goodman, R. D. Colvett, and R. S. Caswell,

An International Neutron Dosimetry Intercomparison,

Proceedings of the Second Symposium on Neutron

Dosimetry in Biology and Medicine , Commission of

the European Communities, Luxembourg (1975), p. 627.

4. J. J. Broerse, G. Burger, and M. Coppola, Basic

Physical Data for Neutron Dosimetry , J. J. Broerse,

ed. , Commission of the European Communities,

Luxembourg (1976), EUR 5629e, p. 257.

5. Basic Physical Data for Neutron Dosi metry, (op.

cit.), p. 24X

6. Basic Physical Data for Neutron Dosimet ry, (op.

cit.), p. 249.

7. A. R. Smith, P. R. Almond, J. B. Smathers, V. A. Otte, F. H. Attix, R. B. Theus, P. Wootton, H. Bichsel, J. Eenmaa, D. Williams, D. K. Bewley, and

C. J. Parnell, Medical Physics, 2, 195 (1975).

8- Basic Physical Data for Neutron Dosimet ry, (op.

cit.), p. 267:

9. E. B. Wagner and G. S. Hurst, Health Physics 5, 20

(1961).

10. Basic Physical Data for Neutron Dosimetry, (op.

cit.), p. 21^

127
REACTOR CORE DOSIMETRY STANDARDS

Willem L. Zijp

Netherlands Energy Research Foundation, ECN

1755-ZG Petten (NH) , The Netherlands

Reactor neutron metrology serves to determine directly flux densities, f luences , spectra, and

indirectly effects like burn-up, depletion and displacements. There are tendencies to require

an accuracy of 2 to 51. This gives requirements for the accuracy of nuclear data, of which the cross section data are most important.

Average fission neutron cross sections for many reactions of interest are at present not accurate enough, owing to inadequacy of the spectral cross section data and to inadequacy of the

knowledge of the fission neutron spectrum of ^^^U above about 8 MeV.

More experiments in benchmark fields, performed in interlaboratory experiments and in international collaboration are necessary to arrive at accuracies specified in reactor development

programs

.

(Comparative evaluation; cross sections; fission spectra; integrals; neutrons; radioactivation; resonance integrals; spectral functions)

Purpose of in-core reactor metrology

Reactor neutron metrology performed inside re- search and power reactors provides numerical information on flux density values, fluence values and spectrum data for many purposes. This information is of

the primary type. Information which can be derived

from these source data are e.g. burn-up values, displaced atoms, helium production, and transmutation

rates

.

For providing these secondary data one often needs

full information of the primary type, and a theoretical model for the way of interaction under consideration. Since often the interaction of interest (e.g.

displacements) is a spectrum dependent function, one

then needs spectrum information. The interest in

in-core metrology data can be categorized with respect

to the materials.

1. Fuel material: Of importance is here the fission

rate density (i.e. the number of fissions per unit

volume) and its distribution axially (vertically)

along the fuel rods (or plates), and radially (horizontally) within these rods (or plates) , and its distribution over a cluster of fuel rods (or a fuel as- sembly) , and from cluster to cluster (or assembly to assembly)

.

Flux density mapping over the fuel core is needed to:

- check the assumed flux density pattern macroscopically over the fuel region;

- check the assumed flux density pattern microscopically in a restricted region;

- locate the positions with peak values of the flux

density (detection of possible hot spots);

- calculate the heat dissipation in specified fuel

regions

.

Localized flux density measurements are required to provide information on local flux density values (e.g.

for radionuclide production, in irradiation facilities,

or any other accessible location).

Apart from these interests, from the point of view of

smooth and optimized reactor operation, there are the

interests from the point of view of research and development of new types of fuels.

Here measurements are desired to determine the changes

in the characteristics of fuel samples (due to irradiation under severe test conditions).

2. ^Fuel cladding Toaterial : Of main importance is here

the determination of radiation damage effects in fuel

cladding after long irradiations, i.e. after large incident neutron f luences.

Results of irradiation of test samples are used to predict the radiation damage occurring when the sample

material is irradiated at an other location (often a

longer irradiation in a smaller flux density) in a com- parable neutron spectrum.

3. Structural materials: This category comprises re- flector materials (such as graphite in a HTGR) and materials for reactor tanks and reactor pressure vessels.

Here one needs to know how long the materials can be

positioned at a certain location in a reactor of given

type, before deleterious radiation induced property

changes prohibit further use. Nowadays much money and effort is spent to reactor pres- sure vessel surveillance programs, since the useful

lifetime of such a vessel is an economic issue of first

importance

.

The damage effects in graphite and steels are mainly

induced by fast neutrons, i.e. neutrons with energies

larger than say 10 keV.

Since thermal nuclear reactors are being sold by many

vendors to utility companies, there is a need for a good surveillance metrology and for national and international standards for execution of such a dosimetry at an internationally accepted level.

Accuracies required

In practical applications of reactor neutron metrology the neutron spectrum is not a goal in itself;

it serves as a tool to calculate integral effects (reaction rates, burn-up, radiation damage effects) which

in general are not directly measurable.

The accuracies in neutron metrology are set by the ac- curacies for the integral quantities of final interest.

The 1973 Consultants Meeting'- mentioned that in special

cases like fuel irradiations and graphite irradiations

in high temperature gas cooled reactors, accuracies to 5% or better may be needed.

The following remarks are based on the conclusions of

the IAEA Consultants Meeting in 1976^. For most applications at present the reached accuracies are in the

range from 5 to 10%. In some cases, required accuracies

have recently been re-evaluated to more stringent specifications, partly also as consequence of improved

understanding of the damage functions. These requirements are reflected in target accuracies to be set for

neutron field determination for the three categories of

benchmark fields discussed further on. At the 1975 Petten Symposium these accuracy requirements were stated^ to be in the range from 2 to 5% for

fast breeder reactor, and somewhat less stringent for

light water reactors and controlled thermonuclear reactors . Present state-of-the art accuracies are estimated to be

in the range of 2 to 30%. The 2 to 5% goal objective

may be considered ambitious for some applications, it

is nevertheless reasonable.

128
At least on the long term most reactor fuel and materials development programs will not accept an uncertainty larger than 5%. In order to achieve such an ac- curacy routinely, however, it is necessary to work

towards a better level of accuracy, namely 2 to 5%.

Role of nuclear data

Threshold reactions commonly used to determine

fast neutron fluences are ^^Ni(n,p), ^'*Fe(n,p),

'*^Ti(n,p), ^^Cu(n,a). If the neutron spectrum is unknown, one cannot define effective cross sections to arrive at fluences of neutrons with energies above 0.1 or 1 MeV (or any other energy bound). One can only define an equivalent fission neutron fluence, by using

a cross section averaged over the fission neutron spectrum. The accuracy of the average cross section values determine directly the accuracy of the equivalent fission neutron fluences. Activation detectors with cross section curves approximating a step function with the

step at 1 or 0. 1 MeV are not readily available.

The most interesting reaction in this respect is

^^Nb(n,n'), but there are still some practical problems

with this reaction"*. It has been recognized that for

radiation damage studies reporting of equivalent fission neutron fluences only is often not sufficient.

More information ia required, either in the form of

spectrum information, or in the form of displacements

per atom, calculated using an assumed spectrum and an accepted damage model.

Neutron spectrum measurements using activation techniques and unfolding programs have been reviewed elsewhere^. For this purpose accurate energy dependent

cross section data for a series of detectors must be

available.

Displacements can easily be calculated when so-called

damage cross sections are available, derived from such

an accepted displacement model .

Damage cross sections

At an IAEA Specialists Meeting on Irradiation Damage

Units, held in Harwell, November 2-3, 1976 the present

status of damage cross sections was discussed. Comparison has shown that two sets of damage energy cross sections for Fe, Cr and Ni (and hence steels) based on

the UKNDF file and the ENDF/B-IV file agree within

adequate accuracy, when applied to fission reactor

spectra.

The agreement for Ni and Fe is within a few per cent;

the agreement for Cr is somewhat poorer, but the discrepancy is negligible in applications to stainless

steels

.

Direct integral measurements of damage effects in gra-g

phite are possible with the so-called GAMIN detectors . These detectors consist essentially of a small graphite cylinder between two nickel activation detectors.

By determining after irradiation the change in electrical resistivity in the graphite cylinder, and the incident neutron fluence on the nickel foil, one can derive an index ^q/ (=graphite damage f luence/nickel

fluence). Using the GAMIN technique the damage-toactivation ratio has been determined experimentally in

many research reactors in the Euratom Community.

Available experience show that within roughly 10% these

measured integral effects agree with calculated effects

using damage cross sections and available spectrum information.

Thermal and intermediate energy regions

Much attention has been paid in the past few years to develop neutron metrology methods related to fast reactor neutron spectra.

Apart from this development work there remain still proplems related to improvement of techniques for measurements of thermal and intermediate neutrons with capture

reactions (self shielding effects; flux perturbation

effects; resonance structure studies} gamma sensitivity

of self powered neutron detectors; influence of cadmium

or boron carbide covers of activation foils).

Also in these studies the cross section data play an important role.

Fast energy region

The main uncertainties in nuclear data are in general

related to the fast neutron cross section data. At the 1975 Petten Symposium Paulsen and Magurno^ con- cluded that the situation in the field of spectral

cross section data for reactor radiation measurements

is still unsatisfactory with respect to accuracy and

completeness. As reasons for this unsatisfactory situation the 1973 Consultants Meeting^ recognized the following reasons:

a. The lack of agreement for a limited set of reactions

on which all measuring efforts should be concentrated;

b. The failure to concentrate the differential measurements on the most sensitive energy region for dosimetry purposes;

c. The lack of a sufficient number of laboratories

equipped with accelerators which can produce monoenergetic neutrons in the 6 to 12 MeV region, and

which are used for neutron measurements

.

Since then much attention was paid to the development

of a consistent set of cross section data and to the

set-up of benchmark experiments.

Reference set of cross sections

The ENDF/B-IV dosimetry file''»^ is now generally available and has been adopted for international comparisons

as a reference cross section set. All users are requested to communicate their experience with this data file,

so that future improvements aid to the aim of arriving

at one generally accepted, internally consistent and

extended dosimetry data file.

The dosimetry reactions have been classified in two ca- tegories^. Category I reactions are defined as reactions : a. for which the energy dependent cross sections are well known over their response ranges in standard

neutron fields;

b. for which calculated reaction rates in the standard

neutron fields are consistent with the measured

reaction rates.

The following reactions belong to category I: l97Au(n,Y)^^^Au, 239pu(n f), 237i^p(n,f), ^^^U(n,f),

56Fe(n,p)5%n, 27Ai(n,a)^'+Na, S3cu(n,2n) ^^^y

^^Ni(n,2n) ^'Ni. (Remark: for the (n,2n) reactions with

very high threshold energies, accuracies of about 10% are presently acceptable)

.

A number of other reactions are considered category I candidates: 5^Co(n,Y)^°Co, ^^^\J(n,y)^^^lS

,

ll^In(n,n')115inm^ SS^i (n,p) 58co , 32s(n,p)32p^

5'*Fe(n,p)5'tMn and 59co(n,a) 5%n.

All other reactions used for neutron metrology are category II reactions.

Decay data

Apart from neutron cross section data, also other nu- clear data play an important role in neutron detection:

half-lives, decay schemes, gamma abundances, fission

product yields etc.

129
Role of benchmark fields

The 1976 Consultants Meeting^ identified three

types of benchmark neutron fields for reactor dosimetry:

1. Standard field: A permanent and reproducible neutron flux intensity, energy spectra and angular flux

density distributions characterized to state-of-the

art accuracy (Examples: thermal Maxwellian spectrum;

epithermal 1/E spectrum; ^^^Cf spontaneous fission

neutron spectrum).

2. Reference field: A permanent and reproducible neutron field, less well characterized than a standard

field, but accepted as a measurement reference by a community of users. (Examples: ^^^U thermal fission

neutron spectrum; the spectra in facilities like IS,

ISNF, Big Ten, Tapiro, CFRMF; shielding benchmarks

using a Fe-block or a Na-block)

,

3. Controlled environment field: A neutron field, physically well-defined and with some spectrum definition,

employed for a restricted set of validation experiments

(Examples: HFIR; BSR; BR-2 Cd loops; HFR; reactor pressure vessel mock-ups; cores of ECEL, EBR-II),

In reactor neutron metrology benchmark fields serve

three general obiectives^:

- validation and/or calibration of experimental techniques ; - validation and/or improvement of cross section data

and other nuclear data for proper application of

experimental techniques;

- validation and/or improvement of analytical methods

needed to extrapolate dosimetry data from a monitoring or surveillance position to the location of

interest.

The highest priority is given to the second purpose,

which includes more specifically^:

- establishment of high accuracy, consistent cross section data for basic integral detectors;

- spectrum dependence of fission yields;

- validation of resonance and threshold detectors for

use in neutron fields where differential microscopic

data may be inadequate;

- activation detectors for which energy dependent cross section data are uncertain.

Fission spectrum representation

For the calculation of equivalent fission neutron

fluences one needs the values for the cross sections

of the activation (or fission) detectors used, and in

particular the cross section averaged over a fission

neutron spectrum.

For the fission neutron spectrum of ^^^U several analytical representations have been used in the past

years

.

In the following expressions, which have been normalized to a value of unity, E denotes the neutron energy, expressed in MeV:

- the formula proposed by Watt-'*' Xi(E) = 0.48395 exp (-E) . sinh/2F

- the formula proposed by Cranberg, Frye et a l . ^ X2(E) = 0.4527A exp (-E/0 . 965) . sinh/2 , 29E"

- the formula of Maxwellian type proposed by Leachmai^^

X3(E) = 0.76985 exp (-E/ 1 . 29)

.

- the modified Watt-Cranberg formula proposed by Wood''" Xit(E) = 0.5827 exp(-0.992E).sinh(1.27v^

The IAEA Consultants Meeting on prompt fission neutron

spectra' held in 1971, concluded that a simple

Maxwellian form does not satisfactorily fit all observed fission spectra.

It was felt then that for the present a purely numerical representation of experimental results woiild be

best. Magurno and Ozer^ tested the data on the ENDF/B-IV

dosimetry file also by calculating spectrum averaged

cross section using the Maxwellian spectrum function:

X5(E) = 0.770.v^'.exp(-E/T)

using

T = 1.29 MeV and T = 1.32 MeV.

Recently Grundl and Eisenhauer' ^ ' ' ^ from the National

Bureau of Standards made a new evaluation, based on 16 documented differential spectrometry measurements of

the thermal neutron induced ^-^^U fission neutron spectrum, and of the ^^^Cf spontaneous fission neutron spectrum. Their results can be described in three forms:

- A reference Maxwellian representation, obtained from

a weighted least squares fit in the energy range from

0.25 MeV to 8 MeV.

For 235u . M(E) = 0.7501 ./E.exp(-1 .50E/1 .97)

.

For 252cf. M(E) = 0.6672.v^.exp(-l .50E/2. 13)

.

- A seven-group spectrum of adjusted Maxwellian segments, which fit the data over all energies. Estimated

uncertainties are 1% to 4% for both spectra between

0,25 and 8 MeV and between 5 and 15% outside this

energy range

.

- A continuous line segment correction to the reference

Maxwellian, which establishes a final fit to the ex- perimental data:

X(E) = y(E).M(E)

Below 6 MeV the correction function ij(E) is linear,

above 6 MeV it is exponential. The correction functions for the two spectra are as follows:

energy

interval

(in MeV)

u(E) for 235u U(E) for 252cf

0 to 0.25

0.25 to 0.8

0.8 to 1.5

1.5 to 6.0

6.0 to =>

1+0.800E-0. 153

1-0. 140E+0.082

1+0.040E-0.062

l+O.OlOE-0.027

1 .043{exp-0.06x

(E-6.0)/l .043}

1+1 .200E-0.237

1-0. 140E+0.098

1+0. 024E-0. 0332

1+0. 0006E+0. 0037

1 .0 exp{-0.03x -

(E-0.60)/l .0}

Similar representations have been tried for the spends 9 taneous fission neutron spectrum of Cf

:

- A formula proposed by Knitter et al-' , using a Maxwellian function with an average energy <E>=2.13 MeV;

- A more complicated function used by Green' ^, based on

a detailed evaporation model, and yielding an average

energy <E>=2. 105 MeV.

The choice of the representation of the fission spectrum may not be so important for activation reactions

with low thresholds, but it becomes important when re- actions with very high threshold (say about 10 MeV) are considered.

As can be seen from table 1 the different representations of the ^3^U fission neutron spectrum give clearly

different results for reactions with very high threshold energy. The Euratom Working Group on Reactor Dosimetry noted that the fission neutron spectrum in the

energy region above 8 MeV is only known with an accuracy

of the order of 25%.

For the application of reactions with high thresholds

and for the prediction of helium production by (n,a)

reactions the knowledge of the fission neutron spectrum

should be improved.

Quality of integral cross section data

Integral experiments to determine (the 2200 m/s

value) , I (the resonance integral) and <a£> (the cross section averaged over the fission neutron spectrum) have

been performed by experienced people in recognized laboratories ,

130
The values for Oo , I and <af> which can be derived

from the ENDF/B-IV dosimetry file^ have been compared

in tables 2, 3, A and 5 with evaluated or experimental

data in recent compilations.

For the fission neutron spectra of ^^^U and ^^^Cf the

NBS evaluations of Grundl and Eisenhauer ^ ' -"^ ^ were

used. A rough estimation on the quality can be based on three

aspects: the uncertainty in the measurement, and

the accuracy (or bias) and the consistency of the evaluated values. As a measure for the experimental uncertainty serves the quoted fractional error, v. As a measure of the accuracy serves the absolute value of

the fractional difference A between (evaluated) experimental error and the calculated value from the cross section file. As a measure of the consistency between

(evaluated) experimental value and the calculated value serves the ratio of the fractional difference and

its stated fractional error v. The following indications are used in the tables:

category uncertainty accuracy consistency

++ 0 < V < 2% 0 < A < 2% 0 < A/v < 1 + 2% < V < 4% 2% < A < 4% 1 < A/v < 2 0 4% < V < 6% 4% < A < 6% 2 < A/v < 3 6% < V < 8% 6% < A < 8% 3 < A/v < 4 10% < V 10% < A 5 < A/v

The data for the fission neutron spectra are taken

from recent reviews by Fabry et al^^^".

The normalization adopted involves a so-called flux

transfer, using the ^^^Pu(n,f) reaction and the NBS

^^^Cf source. This californium source was chosen because of its availability and its well known source strength (error 1.1%). The ^^^Pu(n,f) reaction was chosen because of its relatively flat shape in the

energy range of interest and its well known cross section.

It has been concluded by Fabry, McElroy et al^'' that

integral cross section data for dosimetry reactions as measured in standard and reference benchmark neutron

fields depart from computed ones, not only because of

differential energy cross section inadequacies, but

also because the spectral shapes characteristic of

these benchmarks are usually inaccurate in the energy

ranges not covered or poorly covered by differential

neutron spectrum techniques, e.g.:

- below ='250 keV and above -10 MeV for the fission

neutron spectra of ^^^U and ^^^Cf;

- below »10 keV and above ~2 MeV for 11, CFRMF, BIGTEN. The same authors conclude that even in the well covered

energy ranges, the reliability remains questionable,

as is presently the case for the ^^^U fission neutron

spectrum between 3 and 6 MeV, and for LFRMF between

100 and 400 keV.

The only benchmark whose spectral shape appears to be

accurately established between =0.25 and =:10 MeV is

the ^^^Cf neutron spectrum. The inconsistencies observed for some facilities like LFRMF and BIG- 10 are mostly attributed to inaccurate spectral computations

resulting from the inadequate ^^^U fission spectrum,

and the inelastic scattering cross section data in

ENDF/B-IV. This effect is less pronounced for 11, because the spectral characterization from =10 keV up to

~2 MeV mostly relies on differential spectrometry measurements and not on computations^*^.

The 1976 Consultants Meeting recommended that efforts

should be made to remove inconsistencies between integral measurements and differential evaluations at least as concerns the ^^^U fission spectrum, the 11 type facilities and the ISNF, and the cross sections

for 58Ni(n,n) 235u(j^^f)^ ^^^Zoixi,-^) , llS^Cn.n'),

^'*Fe(n,p), ^°^Rh(n,n'), so as to qualify them as standard spectra and category I reactions respectively.

Improvement of consistencies

The consistency between measured and calculated

reaction rates may be influenced by various effects or procedures

:

- The group structure chosen for presentation of the

final spectrum;

- The discontinuity or extrapolation at the lower and

upper energy bounds of the spectrum;

- The group structure of the cross section libraries,

and especially the detailed structure in the reson- nance region (e.g. important for (n,Y) reactions);

- The adjustments sometimes made to a spectrum from

reactor physics calculations to obtain a better fit

for experimental data;

- The accuracy and precision of the experimentally determined reaction rates;

- The perturbation of the neutron field by the presence

of one or more activation and fission detectors, or their encapsulations;

- The uncertainty in the self shielding effect in the

activation and fission detector applied.

The present situation of the cross sections for reactor

radiation measurements can be improved by a series of

actions

:

- Application of recommended evaluated cross section

libraries, both in reactor physics calculations and

in spectrum unfolding;

- Application of a reference group structure and application of recommended procedures for arriving at other group structures or a series of point values;

- Recalculation of neutron spectra using well established reactor physics computer programs under well

defined and improved conditions;

- Selection of a few well known spectra (serving as benchmark spectra) with different shapes;

- Careful definition of the neutron spectra in the

benchmark facilities;

- Accurate determination of experimental activities,

using enlarged series of detectors;

- Adoption of agreed procedures for adjustment and ex- trapolation of neutron spectra based on reactor physics calculations;

- Intercalibration of counting equipment, based on distribution of calibrated radionuclide samples;

- Adoption of agreed nuclear data (half-lives, decay

schemes, gamma abundances, fission product yields,

etc.) .

It will be clear that several parallel actions are ne- cessary to arrive at the desired accuracies within a few years.

Current international efforts

The actions mentioned above are being discussed in

several international meetings (the IAEA Consultants

Meetings in 1973 and 1976; the regular meetings of the

Euratom Working Group on Reactor Dosimetry; the annual

meetings of the IAEA Working Group on Reactor Radiation

Measurements; the IAEA Specialists Meeting on Radiation

Damage Units in 1976; the ASTM-Euratom sjrmposium on Reactor Dosimetry, held in 1975 at Petten, and the next

one scheduled for October 1977 in the USA).

Owing to intensive international contacts the following

progress results can be mentioned;

- the acceptance and the availability of the ENDF/B-IV

dosimetry file as a reference data set for reporting

and comparing obtained results;

- the acceptance of a list of dosimetry reactions in

categories I and II;

- the acceptance of the IAEA program on Benchmark Neutron Fields Applications for Reactor Dosimetry;

- the recommendation for international interlaboratory

studies like the Interlaboratory Reaction Rate Pro*

131
. gram in the USA;

- the definition of a few selected standard neutron

fields and revision of the list of category I reactions ; - the acceptance of a few radiation damage models for

the calculation of the number of displaced atoms;

- the recommendation by the Euratom Working Group to apply the BURLIB 100 groups structure (very similar

to the DLC 100 groups structure) as reference both

for shielding and reactor metrology work;

- the exchange of technical information at the open

biannual ASTM-Euratom symposia on Reactor Dosimetry;

- the quick exchange of items of interest through the

Euratom Newsletter on Reactor Radiation Metrology;

- the intercomparison of some promising neutron spectrum unfolding codes;

- the IAEA intercomparison of computer programs for

analyzing Ge(Li) gamma ray spectra;

- the international intercomparison of gamma emission

rates of ^^Eu sources.

Conclusions

1 . Different reactor development programs require ac- curacies in integral data (fission rates, burn-up,

fluences, damage effects) of 2 to 5%.

2. An internationally accepted starting point for

data treatment in neutron metrology work is the ENDF/B

IV dosimetry file, which has been made available in a world-wide scale through the four nuclear data centres

(Brookhaven, Saclay, Vienna, Obninsk).

3. Appreciable discrepancies exist between measured

and calculated average cross section values in the

^3^U fission neutron spectrum. The representation of

the high energy tail (above ;8 MeV) of this spectrum

needs further study.

4. Dosimetry reactions in categories I and II, and

classes of benchmark neutron fields have recently been

reviewed by the IAEA 1976 Consultants Meeting. The ac- cepted tables constitute the basis for a critical analysis of integral measurement data which is or becomes

available.

The benchmark field approach can serve to detect discrepancies and inconsistencies, to arrive at a better

quality of cross section data, and to improve spectrum

determinations by the unfolding technique.

Many integral experiments in several benchmark fields,

based on interlaboratory and international cooperation

are needed to arrive at the requested accuracies of

2 to 5%.

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:

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Technical Reports Series No 156 (IAEA, Vienna, 1974)

22. MUGHABGHAB, S.F.; GARBER, D.I.; "Neutron cross sections. Vol. I, resonance parameters"

BNL-325, Third edition. Vol, I (NNCS-BNL, June,

1973) .

23. BIGHAM, C.B.: "The slow-neutron fission cross sections of the common fissile nuclides (revised 1975)"

Nucl. Sci. Eng. 59 (1976), 50-52.

24. ZIJP, W.L.; NOLTHENIUS, H.J.: "Comparison of integral cross section values of several cross section

libraries in the SAND-II format"

Report ECN-2 (Netherlands Energy Research Foundation, Petten, 1976).

25. ALBINSSON, H. : "Infinite-dilution resonance integrals"

Contribution to "Handbook on Nuclear Activation

Cross Sections"

Technical Reports Series No 156 (IAEA, Vienna,

1974) .

26. SMITH, D.L.: "Evaluation of the ^ ^ ' )

1 ^ ^In'" r'eaction for the ENDF/B-IV dosimetry file"

Report ANL/NDM-26 (December 1976).

27. PHILIS, C; BERSILLON, 0.; SMITH, D. ; SMITH, A.: "Evaluated (n,p) cross sections of '*^Ti, '*^Ti and

Report ANL/NDM-27 (January 1977).

28. CANCE, M.; GENTHON, J. P.; MICAUD , G.; SALON, L.

:

"Le detecteur neutronique au graphite G. A. M.I.N.

Etudes et applications a la dosimetrie des dommages radio-induits"

Report CEA-N-1823 (Saclay, January 1975).

29. GENTHON, J. P.; HASENCLEVER, B.W. ; SCHNEIDER, W.

;

MAS, P.; WRIGHT, S.B.; ZIJP, W.L.: "Recommendations on the measurement of irradiation received

by the structural materials of reactors"

EUR-5274 e,n,d,f (C.E.C., Luxembourg, 1975).

133
Table 1: Influence of representation of fission neutron spectrum of ^^^U

on average cross sections

Based on cross sections from the ENDF/B-IV dosimetry file, and data reported

by Fabry et al |l9l,|20|. Cross section values are given in fm^(=10-3''m^) .

effective

threshold

energy

(in MeV)

TTi fi a e n T* PmP n t*

rT« r 1 Tl fTTl ^ ^

ratio Oc/om

Maxwellian

<E>=1 .97 MeV

NBS-eval.

<E>=1 .98 MeV

WAtt

<E>=2.00 MeV

237Np(n,f )FP 0.6 131.2 1 .006 1 ,006 1 .019

5SNi(n,p)58co 2,8 10.85 0.926 0.936 0,947

5'*Fe(n,p)5'*Mn 3,1 7.97 0,965 0.975 0,984

27Al(n,p)27Mg 4.4 0.386 1 ,078 1 .067 1 ,062

5^Fe(n,p)56Mn 6,0 0. 1035 1 ,081 1 .017 1 ,000

59co(n,ct)5%n 6.8 0.0143 1 . 140 1 .035 1 .021

27Al(n,a)2'*Na 7.2 0.0705 1 . 106 0.983 0.970

127i(n,2n)12Sl 10.5 0. 105 1 .499 1 . 130 1 .094

^5Mn(n,2n)5'tMn 1 1.6 0.0244 1 ,426 1 .004 0.951

^SNi(n,2n)57Ni »13,5 5,77x10"'* 0.776 0.489 0.440

Table 2: Comparison of 2200 m/s cross section values

Cross section values are expressed in units of 100 fm (in 10" °m)

reaction

compilation value, Cm uncertainty

calculated

value, Oq

Oc/ "m accuracy

consistency

IAEA-TR-156

|21 1

BNL-325

f22|

BIGHAM

|23| (in %)

ENDF/B-III

|24|

^Li(n,a)3H 940±4 0.4 ++ 923 0.982 ++ l°B(n,a)^Li 3837±9 0.2 ++ 3770 0.983 ++ 2%a(n,Y)^'*Na 0.528±0.005 0.53010.005 0.9 ++ 0.524 0.989 ++ +

'*5sc(n,Y)'*^Sc 25±2 26.5 ±1.0 3.7 + 25.9 0.977 + ++ 58Fe(n,Y)55Fe 1 . 14 ±0.05 1 . 15±0.02 1.7 ++ 1 .16 1 .009 ++ ++ 59co(n,Y)60co 37.5 +0.2 37,2 ±0,2 0.5 ++ 36.6 0.984 ++

^3cu(n,Y)^'*Cu 4,4 ±0,2 4,5 ±0,1 2.2 + 4.42 0.982 + ++ 115in(n,Y)ll^Inni 161±5 157 ±15 9.6 164 1 .045 0 ++ l"Aa(n,Y)158^^, 98,8 ±0.3 98.8 ±0,3 98,7 0,2 0.2 ++ 97.

1

0.983 ++ 232Th(n,Y)2^^Th 7.4 ±0.1 7,40±0,08 1 , 1 ++ 7.26 0.981 ++ + 235u(n,f)FP 580 ±2 582,2±1 ,3 576,9 3,4 0,6 ++ 573 0.993 ++ + 2"Np(n,f)FP 0,019±0,003 15.8 0,0163 0.857 ++ 238u(n,Y)23% 2,720±0.025 2,70 ±0,02 0.7 ++ 2.65 0.981 ++ o 239pu(n,f)FP 742 ±3 742,5±3,0 742.8 4.4 0.6 ++ 730 0.983 ++ o

' Table 3: Comparison of cross sections, averaged over a 1 /E neutron spectrum

Values of resonance integrals refer to a cadmium cut-off equal to 0.5 eV, and are expressed in units of 100 fm^ (10~28m2).

reaction

— — - compilation value, uncer calculated

value, Oc ENDF/B-lV

|8|,|22|

Oc/om

accuracy

consis- tency IAEA-TR-156

25 1

BNL-325

|22|

taint y (in %J

^Li(n, total He) 425.87

l°B(n, total He) 1722 0.3 ++ 1722. 17 1 .000 ++ ++

2 3Na(n,Y)2'*Na 0.31 0.31 1 3.2 + 0.346 1.113

'*5se(n,Y)'*^Sc 1 1 1 1.3 8.8 1 1 ,29 0.999 ++ ++ 58Fe(n,Y)^^Fe 1 .2 1 . 19 5.9 0 1 ,58 1 .328

59Co(n,Y)^^Co 75.0 75.5 2.0 ++ 76.67 1 .015 ++ ++ 63cu(n,Y)^'*Cu 5.0 4.9 6.2 5.55 1 . 133 o ^^5in(n,Y)^^^Irfn 2600 3300 3.0 + 3242.74 0.983 ++ + 197Au(n,Y)l^^Au 1550 1560 2.6 + 1564.70 1 .003 ++ ++ 232xh(n,Y)^"Th 82 85 3.5 + 85.58 1 .007 ++ ++ 235u(n,f )FP 275 275 1 .8 ++ 282.00 1 .025 + + 238u(n,Y)239u 280 275 1.8 ++ 277.53 1 .009 ++ ++ 239pu(n,f)FP 310 301 3.3 + 303.90 1 .010 ++ ++

134
Table 4; Comparison of cross sections, averaged over the fission neutron spectrum of ^^^U

Cross section values are expressed in units of fm^ (= 1 ''m^) , Calculated values refer to the ENDF/B-IV dosimetry file and the NBS spectrum evaluation. Table is based on data

reported by Fabry et al |l9|,|20|.

effective integral calculated

reaction threshold measurement

uncertainty value oc accu- consistency

(,in MeV; (in %) \in rm ; racy

115in(n,Y)116inm - 13.45 0 13.59 1 .010 ++ ++ 197Au(n,Y)l^SAu - 8,35 0 . U 0 8.46 1 .013 ++ ++ "cu(n,Y)^'*Cu - 0.930 I C A 1 .099 1 . 182 — + 235u(n,f)FP - 120.3 2 ,

5

+ 124.1 1 .032 + + 235pu(n,f)FP - 181 .

1

+ 178.

1

0.983 ++ ++ "7Np(n,f) 0.6 131.2 J . o + 132.0 1 .006 ++ ++ l°3Rh(n,n')l°3Rhm 0.8 73.3 0 rl6.68^'' b8.90°>

- 115in(n,n')115lnm 1.2 18.9 0 0.883

1 .000 —

++

o

0 232Th(n,f)FP 1.4 8.1 D . / 6.90 0,852 — o 238u(n,f)FP 1.5 30.5 J • J 29.58 0,970 + ++

'*7Ti(n,p)'*^Sc 2.2 1 .90 1 , 126 — +

\ 2,138c; 1 , 1 25 + 31p(n,p)31si 2.4 3.55 7 ft 3.245'^-' 0,914 - + 5%i(n,p)58co 2.8 10.85 s n 0 10.16 0,936 - + ^'*Zn(n,p)6'*Cu 2.8 2.99 JS » LH 0 - - 32s(n,p)32p 2.9 6.68 J.J 0 6.41 0,960 + ++

5'+Fe(n,p)5'*Mn 3.1 7,97 6,1 7.77 0,975 + ++

Ti(n,x)'*^Sc 3.9 1 ,18 6.4 r 0.999

1 1.088^=^ 0.847 0 0,922 0 27Al(n,p)27Mg 4.4 0,386 6,5 0.412 1 ,067 0 56Fe(n,p)5%n 6.0 0.1035 7,2 0,1053 1 ,017 ++ ++ 5^Co(n,a)^%n 6.8 0.0143 7,0 0,0148 1 ,035 + ++ S3cu(n,a)S°Co 6.8 0.0500 11.2 0.0352 0,704 0 2'*Mg(n,p)2'*Na 6.8 0. 148 5,5 0 1 0. 1518^1) 1 ,026 + ++

27Al(n,a)2'*Na 7.2 0.0705 5,7 0 0.0693 0,983 ++ ++

'*STi(n,p)'*8sc 7.6 0.0300 6.0 0 f 0.0173 0,577

I- 0.0303'=) 1,010 ++ ++ 53Nb(n,2n)92Nbm 10.2 0.0475 6.7

127i(n,2n)126i 10.5 0.105 6.2 0. 1 186 1 , 130 0 '55Mn(n,2n)5'+Mn 11 .6 0,0244 6.1 0.0245 1 ,004 ++ ++

^3cu(n,2n)62(;^ 12,4 0,0122 9.8 0.00915 0,750 o 90zr(n,2n)89zr = 13 0,0247 6.9 0.008714 0,353

^^Ni(n,2n)57Ni rl3.5 5,77x10-'* 5.4 0 2.82x10"^ 0,489

a) Magurno|8| reports a calculated value of 16,68 fm2;our calculations yield 16,72 fm . These two values are clearly different from. the value of 18.22 reported by Fabry | 20

b) Based on a recent evaluation by D.L. Smith |26j using a Maxwellian spectrum with

<E> = 1 .98 MeV.

c) Based on a recent evaluation by C. Philis et al.|27|.

d) Cross section data not present in ENDF/B-IV dosimetry file; listed value has been

taken from SAND-II cross section file.

135
Table 5; Comparison of cross sections, averaged over the fission neutron spectrum of ^^^C£

Cross section values are expressed in units of fm^ (= 1 '^m^) . Calculated values refer

to the ENDF/B-IV dosimetry file and the NBS spectrum evaluation.

Table is based on data reported by Fabry et alk | 1 9 | , | 20

[

reaction

effective integral uncertainty calculated

value 0(, accu- consisthreshold measurement (in %) tency

( in MeV ) ^) Om Un fm ; (,in rm ; racy

115ln(n,Y)^^^InTM 12.53 3.4 + 13,03 1 .040 + + 197Au(n,Y)15SAu 7.99 3.6 + 7.99 1 .000 ++ ++ 235u(n,f)FP 120.3 2.5 + 124. 1 1 .033 + + 239pu(n,f)FP 180.4 2.5 + 178.9 0.992 ++ ++

23''Np(n,f)FP 0.6 133.2 2.8 + 135.

1

1,014 ++ ++ l"Rh(n,n')l"Rhm 0.8 75.7 7.0

115in(n,n')115lnm 1.2 19.8 2.5 + 17.55 0,886^)

238ttc„ flirp I . 5 32.0 2.8 + 31 .54 0. 986 ++ ++

'*7Ti(n,p)'*7sc 2.2 1 .89 2.1 + , 2.384 1,261 .

1 .281'^^ SSNi(n,p)58co 2.8 1 1,8 2.5 + 1 1 .50 0,975 + ++

5'+Fe(n,p)5'tMn 3.1 8.46 2.4 + 8.91 1 .053 0 o '*6Ti(n,p)'+6Sc 3.9 1 .38 2.2 + ; >.252 .

^ 1,381

0.907 .

1 .OOl'^'^ ++ ++ 27Al(n,p)27Mg 4.4 0.51 9.8 0,514 1 .008 ++ ++ 56Fe(n,p) ^^Hn 6.0 0. 145 2.4 + 0, 1475 1 .017 + + ++ 27Al(n,a)2'*Na 7.2 0. 1006 2.2 + 0, 1059 1 .053 0 '*8Ti(n,p)'+8sc 7.6 0.042 2.4 + r 0.265 ,

^ 0.0446*^^

0.630 ,

1 .062'^-^ 0 5%n(n,2n) 5'*Mn 11.6 0.058 10.3 0.0528 0.910 ++ 59Co(n,2n)58co 0.057 10.5 0.0379 0.665

63Cu(n,2n)62Cu 12.4 0.030 10.0 0.02415 0.715

a) Threshold values are valid for a 235^ figgign neutron spectrum.

b) Based on a recent evaluation by D.L. Smith | 26 |

.

c) Based on a recent evaluation by C. Philis et al | 27 |

.

136
STMDARDS FOR DOSIMETRY BEYOND THE CORE

Frank J. Rahn, Karl E. Stahlkopf and T. U. Marston

Electric Power Research Institute

Palo Alto, CA 94304

and

Raymond Gold and James H. Roberts*

Hanford Engineering Development Laboratory

Richland, WA 99352

Introduction

The need for well understood, standardized neutron

dosimetry techniques is increasingly evident for

several applications in thermal and fast reactors

beyond the core region. This need for dosimetry

comes mainly from separate but related interests,

namely: pressure vessel surveillance, materials

testing, design and shielding requirements. The needs

of the controlled thermo-nuclear program are addressed

in another session of this symposium.

Objectives of Dosimetry

The problem of dosimetry for use in reactors (damage

studies, irradiation experiments, shield assessment,

reactor studies) is a very critical one that may

influence the development of nuclear power. The selfconsistency of calculated and dosimetrically derived

flux and spectrum is of prime importance to the

development of validated techniques for design,

testing and licensing purposes.^

Dosimetry measurements serve the following objectives:

1) Surveillance of the safety and integrity of the

pressure vessel of a nuclear steam supply (NSS)

system.

2) Knowledge of the neutron fluence and spectrum for

correlation with changes in properties in a materials test program.

3) Validation and/or calibration of experimental

techniques

.

4) Verification of analytical methods for predicting

fluxes far from the core (in terms of a neutron's mean free path) for shielding purposes,

sensitivity analyses,^ channel theory,^ etc.

There are many additional uses of dosimetry results.^

Among these are:

1) choosing between two or more data sets,

2) identifying discrepancies and

3) improving (i.e., adjusting) cross sections.

These latter uses are generally controversial, and

split the engineering and physics community. One

hopes to identify discrepancies between integral and

differential data and, if necessary, make adjustments

in a self-consistent way. It is of prime importance

to obtain the sensitivity of the integral results to uncertainties in the nuclear data.

To reach these objectives it is important that the cross section data for dosimetry reactions be

integrally consistent and well-known.^ Benchmark

neutron fields are now classified according to

guidelines first established

Symposium (1975):

at the Petten

Standard : a permanent and reproducible neutron

field with neutron flux intensity, energy spectra

and angular flux distributions characterized to state-of-the-art accuracy. The main

characterizations must be verified by

interlaboratory measurements and calculations.

Reference : a permanent and reproducible neutron

field, less well-characterized than a standard,

but accepted as a measurement reference by a community of users.

Controlled Environment : a neutron field,

physically well-defined and with some spectrum

definition, employed for a restricted set of

validation experiments.

Dosimetry and Benchmark Experiments

A conclusion of the recent IAEA Consultants' Meeting

of Integral Cross Section Measurements in Standard

Neutron Fields''' is that spectrum averaged cross section data as measured in standard and reference

benchmark neutron fields differ from computed ones, not only because of absolute flux and cross section

inadequancies , but also because of uncertainties in

observed spectra in most of these benchmarks.

Uncertainties in spectral measurements stem from

limitations in the energy range coverage in

differential neutron spectra techniques.

In controlled environments, a combination of

calculations, neutron differential spectrometry and

integral measurements will be required for the

determination of the spectra. The development of

improved unfolding codes to handle the results of

these techniques is desired.^ The success of such

techniques, however, will depend on the availability

of data and analytical methods for handling errors and

their correlation, i.e., not only cross section data

files but also for spectrometry and reaction rate

data. The total flux and spectral shape uncertainties

for the important energy regions are currently

estimated in the +5 to 15% (l<s-) and +5 to 30% (Icr)

9 — • —

range. The current best estimates of the accuracy

required for support of experimental out-of-core flux

and spectral measurements are given in Table 1,

There is no current experimental technique which

allows the determination of the entire spectrum to accuracy of +].0°4. (!«•) or better. Proton recoil, ^Li

and "^He results, taken together, can approach the

required accuracy in the energy range of a few keV to

> 5 MeV. The state-of-the-art determination of

spectrum requires the combined use of calculations,

good unfolding techniques, and data from differential

spectrometry and integral measurements.

*0n leave from Macalester College, St. Paul, MN 55105.

137
Dosimetry Requirements for Design and Shielding

Dosimetry requirements for reactor applications can be

quite stringent. The nature of the problems can be

quite diverse and complex and is often related to specific designs. Situations can arise in which the

uncertainties in calculations can lead to unacceptable

constraints. If in an LWR, for example, the diameter

of the pressure vessel is increased to mitigate

radiation embr ittlement of the steel, then the effects

of dosimetry uncertainties can have a rather dramatic

effect on the cost of the plant. If, in an existing

plant, the dosimetry has unduly large uncertainties,

then licensing and safety considerations may result in

premature plant life and/or operating constraints.

Many other examples exist. For instance, possible

radiation effects on support plates and grids, fuel

assemblies, and other core internals, especially in

the design of fast reactors, can impact on the size of

the pressure vessel required, pressure drop across the core, and in-vessel configuration.^'^ Good dosimetry

can also lead to the optimum location of low and

intermediate power flux monitors and instrumentation.

There is presently a fairly disparate range of

opinions as to the accuracies required for the

dosimetry measurements. In part, this is due to the

lack of consensus on the quantification of the

design/experimental criteria and the diversity of the

problems encountered. In addition to knowledge of the

magnitude of the flux in some region outside the core, spectral information is usually required. This is

true, for instance, of activation rates in control

rod, drive mechanisms or for pressure vessel

irradiation where both the flux and the spectrum (as

related to either the activation cross section or some pseudo-damage cross sections) are required. Some

typical target accuracies (2<y) required in power and

experimental reactors "'^ "'^ are:

1) Radiation Heating +20%

a) thermal shields

b) support plate

c) control rods

2) Displacement Rates +15%

a) reflector region

b) support plate

c) internal shields

3) Instrument & Control Tubes +15%

4) Activation +_3 0%

a) structural members

b) coolant and coolant circuit

c) components

5) Radiation Streaming +30%

Dosimetry Requirements in Operating Reactors

The reference experimental measurements planned on operating reactors generally fall into the category of

controlled envirormient fields. An integral part of a dosimetry program is the comparison of the data with

results from computer codes so as to produce a validated analytic method of estimating the flux and

spectrum for streaming, damage, or activation

purposes. There are also a series of computational

benchmarks available which are intended to check the

methodology, data and adequacy of the models used to calculate quantities of interest.

The purpose of reactor dosimetry programs outside the core is to obtain high accuracy and precision

quantities which can be related to physical quantities

of interest and which can be used to validate the

1 2 calculational methodology. This mandates that the

neutron flux and spectrum be measured as the

intermediary between the radiation exiting the core

and the various quantities to be determined. One of

the greatest needs for good dosimetry is for use in

correlating material effects occurring in the pressure

vessel of LWR's under irradiation.

The requirements for dosimetric measurements on operating and experimental reactors are:

(a) A relatively complete flux and spectrum mapping

starting at the edge of the core through the

outside edge of the pressure vessel. This should

include detectors in the coolant gap radially

outward from the core, and in a sufficient number

of axial and azimuthal positions within the

coolant region in order to assure complete

knowledge of the neutron field at the pressure

vessel. In addition, detectors should be

positioned on both sides of the pressure vessel

to measure the transmission and the change in

spectrum in the steel. Such complete knowledge

is required to validate the computational methods

and to establish a self-consistent set of

experimental points,

(b) The principal energy range of interest for

dosimetry purposes is the range 0,1 < E < 10 MeV,

It is the neutrons in this energy range which are

the greatest contributors to radiation

embrittlement in steels. However, information on

the spectra in the energy region 1 < E < 100 keV

is important for shielding problems due to the

high transmission of neutrons through steel and

sodium for these energies. The most important

location for spectral measurements in this energy

range is at the exit surface of the pressure

vessel. At this point the usual calculational

approach is to interface the 2D finite difference

transport calculations used inside the pressure

vessel with the 3D Monte Carlo and albedo

scattering techniques used in the reactor cavity

region. Flux and spectral measurements, then, in

the 1 to 100 keV range are necessary to verify

the deep penetration predictions and establish

the source for the Monte Carlo simulations.

Thermal flux measurements are usually required

for activation considerations, and for

corrections of the various detector responses to the fast flux.

Using an LWR as an example for fast neutrons, the

water gap is approximately 7 mean free

paths (mfp) wide. For neutrons much below 1 MeV,

the concept of spatial distances in terms of mean

free paths is somewhat misleading, because the

neutron density of these energies is due to high

energy neutrons, undergoing collision and rapidly

thermalizing . The pressure vessel is about

1.5 mfp wide for BWR's and 2.0 mfp wide for

PWR's, At the inside edge of the pressure

vessel, the expected flux (< 0,1 Mev) is

approximately 3,6 x 10^ n/cm^/sec for a BWR and

1,4 X 10-^ n/cm^/sec for a PWR, for plants

producing 3400 MWth and 3600 MWth, respectively,

(c) Practical considerations which determine the

proper balance between desired accuracy and what

is actually obtainable on commercial operating

reactors. On such plants, operational

considerations, accessibility and instrument

locatability require very precise planning to

138
Table 1

Energy Range Standard**

Radiation Field

Accuracy Required* (2cr) Reference**

Neutron Fields

Controlled

Radiation

Environment

thermal

0.414 eV - 1.0 keV

1.0 keV - 100 keV

0.1 MeV - 1.0 MeV

1.0 MeV - 4.5 MeV

4.5 MeV - 10 MeV

>4% (4%)

>10% (10%)

>_10% (6%)

>10% (>20%)

>6% (>_20%)

>10% (>20%)

M0% (10%)

>20% (30%)

>10% (20%)

>_67, (20%)

>_6% (20%)

>10% (30%)

*the first number is what is ultimately needed, the second is what is

. believed to be currently achievable with state-of-the-art techniques

for measuring integral flux over the energy interval.

**for standard and reference fields, the accuracy is specified

for LWR, LFMBR and CTR requirements, reflecting the overall

requirements for all of these systems.

Table 2

COMPARISON OF SELECTED DIFFERENTIAL REACTOR NEUTRON SPECTROSCOPY METHODS

Method

1. (n,p) Emulsions

° Collimated Source

Non-Collimated

Source

3x10rl

3x10,-1

20

10

Resolution

Fair

Accuracy % (Ic)

5-10%

10-20%

10-15%

15-25%

0.3<E<10 Mev

10<E<20 MeV

0.3<E< 3 Mev

3<E<10 Mev

(n,p) Proportional Counters 1 X 10" 2.5 Good

10-50% 0.001<E<0.03 Mev

5-10% 0.03<E<1.0 MeV

10-25% 1 .0<E<2.5 MeV

3. ^Li(n,t)^He 1 X 10 2 6.5 Fair

15-40% 0.01<E<0.5 MeV

15%c 0.5<E<2.0 MeV

20-30% 2.0<E<6.5 MeV

4. Time-of-Flight (TOF)

° ^(d,n)'^e Source

" LINAC Source

5 xlO 5 0.2

5 X 10

Good

10-15% O.OOOOKE

<0.001 Mev

15-20% 0.001<E<0.02 Mev

20-30% 0.02<E<0.5 MeV

10-15% O.OOOOKE

<0.001 Mev

15-20% 0.001<E<0.02 Mev

10% 0.02<E<0.5 Mev

15% 0.5<E<1,0 MeV

20% 1 .0<E<5.0 Mev

a) Approximate lower energy limit of applicability, MeV.

b) Approximate upper energy limit of applicability, MeV.

c) The current accuracy of ^Li(n,t)'^He spectroscopy is mainly

dominated by the uncertainty in the angular and total reaction cross

sect ions

.

139
bring off a successful experimental program.

These various considerations, and especially the

problem of equipment installation, almost in

themselves require that the experiment be

performed during a reactor start up. It is

essential that a set of precalculations exist to bracket the fluxes and fluences expected.

The detectors should be chosen with some care,

and should have overlapping energy ranges.

Redundancy is important in achieving a selfconsistent set of measurements. There are several different classes of detectors, each of

which has particular virtues and disadvantages

for use in operating and experimental reactors.

They are:

(1) Activation Foils - provide good coverage of

the thermal range and energies between 0.5

and 8.0 MeV, Since these are passive

detectors and have a long history of use they are ideally suited to in-vessel

measurements. When using such detectors care must be given to the quality assurance

of the foil materials, and mult ilaboratory

intercompar ison of the activation counting

is recommended . '^ One must also

exercise care that flux perturbations due to the detectors themselves do not influence

the results.

(2) Proton Recoil Proportional Counters - give

good resolution spectra from a few keV to a few MeV and are capable of covering the

range from 0.1 to 1 MeV where activation

foils have difficulties. The proton recoil

detectors are usually H2 or Ch^ filled at various pressures. Experimental devices are being tested using He. Proton recoil

detectors are not operable in neutron fluxes

higher than about 10^ n/cm^sec. When used

in an operating reactor environment, they are not suitable for use inside the pressure

vessel and can be used outside the pressure

vessel only during low or zero power

operation. At low power, the fission

distribution within the core may be

appreciably different and influence the

axial distribution total flux in the cavity.

(3) Special and Experimental Detectors'''^ - In pressure vessel surveillance (PVS)

applications, highly specialized miniature

proportional counters will be required.

Consequently, accurate treatment of

proportional counter finite size effects

will be mandatory.

A proton recoil method which introduces less

perturbation and possesses considerably

reduced finite size effects is emulsionphotographic track plate proton recoil

spectrometry. The emulsion technique

possesses the advantages of passive

monitoring; however, just as for active

spectrometry, low power clean environments

are mandatory. Hence, it is appropriate to combine the emulsion method with active

neutron spectrometry techniques wherever

possible

.

For actual experiments in operating

reactors, it is only possible to utilize

passive multiple foil flux-fluence

spectrometry. In such irradiations it is

highly advantageous to use time integrative

passive monitors. While the passive

multiple foil spectrometry has utilized

radiometric dosimeters in the main, in PVS applications the length of irradiation is

often ill-defined and the power-time history

of the reactor can also introduce

considerable uncertainty. Hence, for more

accurate data from such experiments, it will

be necessary to exploit time integrative

passive monitors, such as long-lived

radiometric dosimeters (encapsulated fission

foils providing long half-life fission

products such as Cs) and, in particular,

solid state track recorders (SSTR), and

helium accumulation fluence monitors (HAFM)

.

It is noted that the latter two passive

methods possess higher sensitivity than

radiometric dosimeters. Moreover, these two methods compliment each other in terms of

the neutron sensitivity provided for the

broad energy range of interest, 0,1 to 10.0 Mev. HAFM utilitizng ^Li or will

provide good sensitivity in the low energy

region, whereas SSTR incorporating fission

threshold nuclides such as ^^^Th, ^^^U, or

237 • • Np, will provide good complimentary

sensitivity at higher energy.

For longer term irradiations where the fast

18 2 fluence exceeds 10 neutrons/cm , another

time integrative passive neutron dosimeter

that can be used is crystalline quartz.

Physical changes in the crystalline quartz can be examined such as changes in density

and in the refractive index of the medium.

These properties can provide sensitive

indicators of fast fluence that accrues during irradiations.

Differential Neutron Spectroscopy

A "state-of-the-art" comparison of differential

reactor neutron spectroscopy that is available for

dosimetry measurements is presented in Table 2. This

comparison is not exhaustive, but has been confined to those selected techniques which have generally found

world-wide acceptance .' ^'^^ All of these

techniques possess limitations, in that no single

spectroscopy method provides an ideal solution, i.e.,

possesses good sensitivity, accuracy, and resolution

over the entire neutron energy region of interest in

reactor applications, namely, from 4 x 10~^ MeV up to 15 MeV for fission reactors. In this regard, Table 2 contains the approximate lower and upper energy limits

of applicability, Ej^ and E^, for each of these

selected methods. Consequently, many different

methods of differential spectroscopy must often be

used to cover the energy region of interest for a given benchmark neutron field.

Perhaps an even more unfortunate shortcoming of all

current differential spectroscopy methods is the

general inability to conduct measurements in high

power reactor benchmark environs. The inapplicability

of differential reactor neutron spectroscopy

techniques in high power environs can be traced to inherent limitations, such as:

1) count rate limitations,

2) radiation damage (of detectors and/or electronic

components)

,

3) sensitivity to background radiation components

(principally the gamma-ray component), and

140
4) temperature sensitivity (of detectors and/or

electronic components).

Although limitations frequently arise which are uniquely associated with a specific technique,

different spectroscopy methods often share common

shortcomings. As has been stressed in a recent

review, one such shortcoming is the need for

unfolding which is apparently a universal necessity in

all differential reactor neutron spectrometry methods.

This universal requirement may well stem from one of

the chief intrinsic characteristics of the neutron,

namely, neutral charge. In neutral particle

detection, one customarily uses interactions that

produce charged particle reaction products which are,

in turn, readily detected. Hence, the response of

such a detector need not generally be in one-to-one

correspondence with the energy of the incident

neutrons. Any departure from this one-to-one

correspondence automatically necessitates unfolding.

The discovery and validation of a differential neutron

spectroscopy method which could be applied in high

power reactor benchmark environs must necessarily be

classified as a major breakthrough in research reactor

technology. However, in assessing possible future

directions and developments in differential reactor

neutron spectroscopy less ambitious goals are obviously more prudent and realistic. Perhaps the

major emphasis will occur in the higher region

extending to 15 MeV. Here one can anticipate a re- emergence of the proton-recoil photographic emulsion

technique, which possesses fundamental advantages for

the high energy range for both fission and fusion

reactor applications, such as small inexpensive

passive detector packages of good energy sensitivity

and negligible spectral perturbation.

In these higher energy neutron environs, increasing

emphasis will also be placed on reactions which

produce helium, i.e., alpha particles. The helium

yield of many relevant reactions is vastly increased

at high energy. Mass spectrometry techniques have

already been perfected for measuring helium production

in high power irradiations. He-recoil proportional

counter spectroscopy has also been advanced as a means of extending the current energy domain of

applicability of proportional counter techniques (see

Table 2). Perhaps the most appealing aspect of this

latter technique is the adaptation of already existing

hardware and instrumentation. By simply filling with

helium instead of hydrogen or methane, proportional

counters (already employed in the proton-recoil

method) can be directly applied for c<v -recoil

spectrometry. Estimates of E^^ ~15 MeV have been

forecasted, provided that proportional counters of

high quality and helium of very high purity are

employed. This estimate implies that just as with

proton-recoil spectroscopy, accurate unfolding of

finite-size effects in c(-recoil proportional counters

must be performed.

Two particularly important spectral characteristics,

which have generally been neglected in differential

reactor neutron spectroscopy, namely, absolute flux

measurements and observation of angular flux

anisotropy, will quite likely be singled out for

emphasis in future work. Only very limited absolute

spectral measurements have been carried out usin^

proton recoil emulsions^^ and proportional counters.

On the other hand, additional information that can be

obtained by means of nuclear emulsions is knowledge of

the anisotropy of the neutron flux,^ If, in th|

emulsion measurement of the proton spectrum ' (which can be unfolded to give the neutron spectrum).

the direction of the proton track is recorded,

anisotropies in the neutron flux can be detected. For

example, in a neutron spectrum in which the neutron

flux is decreasing rapidly with increasing energy in

the vicinity of 2 MeV, a significant fraction of the

recoil protons with energies 2 MeV will be produced

by neutrons of about this energy. For these events,

the proton recoil direction is always forward. For

protons > 2 MeV, the direction of the proton velocity

can be determined with close to 100% certainty. This

confidence factor decreases as proton tracks get

shorter

,

For neutrons of energy 250 keV, nuclear emulsions

loaded with glass specks^ and containing ^Li can provide some knowledge of the anisotropy in neutron

flux. For example, in the Li(n,rt) t reaction the sum

of the ranges of the alpha particle and triton is

significantly longer if the triton goes in a forward

instead of a backward cone. Thus, there will be a strong correlation between the direction of the long

tracks and the neutron direction. The Li(n,o<) t cross section resonance at 250 keV significantly

increases the number of triton-alpha pairs and thereby

provides improved statistical accuracies of angular

observations in the vicinity of the resonance.

In hazarding one final projection for possible future

directions of differential reactor neutron

spectroscopy, one cannot ignore the virtual explosion

of activity and developments which has occurred in the

field of Solid State Track Recorders (SSTR) over the

last decade or so. In particular, the ability to extract relevant physical data such as mass, charge

and energy from the shape of tracks formed in SSTR is

a striking advance. This recent development, which is

commonly referred to as track profile analysis, can

only be considered to be in the very formative stages.

This ability coupled with the vast improvements of

=< -particle sensitive SSTR, such as £ellulose Acetate

Buterate (CAB) and £ellulose tJitrate (CN), augurs for

the evolution of SSTR differential neutron

spectroscopy methods. Advanced SSTR techniques for

observing angular flux anisotropy should also become

possible. Such methods would, of course, be direct

descendants of emulsion techniques and thereby

automatically possess many advantages. However, in

contrast with emulsions, the insensitivity of SSTR to electrons would provide an enormous advantage for

reactor applications. Moreover, the reduced

<^-particle range also implies improved high energy

sensitivity with less attendant stress on the need for

finite-size corrections.

The Importance of Good Dosimetry - A Specific Example

To put the value of good dosimetry into perspective,

let us consider a specific example, relating to the

materials test program and radiation embrittlement in

LWR ' s

,

There are presently 53 commercial nuclear reactors

operating in the United States^ which account for 10%

of our nation's electricity requirements. Commercial

nuclear power has been generating electricity in this

country since 1959 with a safety record unmatched by

any other heavy industry. Since the commissioning of

Dresden I, the first commercial nuclear power plant, no fatalities and no substantial personnel

contamination have resulted from the operation of the

nuclear steam supply systems for electrical

generation. One of the primary reasons for this

excellent operating record is the very conservative

design utilized in the building and operation of these

plants.

141
A conservative design is necessary to account for

unknowns in materials properties, operating stresses,

and operating environments. As these factors are better defined through development of refined

analytical techniques and measurements made on operating systems, it is possible to reduce

overconservatism and still maintain safety,

reliability, and availability.

Overly conservative design or regulatory action can

result in substantial losses to the economy. If we assume a 7 mill/kWh difference in base loaded coal and

nuclear electricity generation rates, and further

assume that all electricity from nuclear outages is

replaced by base loaded coal generation capacity, each

percent of lost nuclear capacity represents a $16,6

million increase in cost of electricity per year. If

each lost percent in nuclear capacity were replaced by

oil the cost to the economy would be about $70 million

per year.

The preceding remarks were presented to emphasize the

economic consequences of overconservative definition

of design and operating margins. The nuclear industry

is presently faced with potential premature retirement

or long forced outage for thermal anneal of a number

of PWR plants due to irradiation embrittlement of

reactor pressure vessel steels. The remainder of this

paper will discuss the present regulatory and code

requirements for irradiated materials properties and

suggest research necessary to accurately define the

proper irradiated materials properties requirements.

The Regulatory Position on Radiation Embrittlement

The prevention of brittle fracture of the pressure

boundary of nuclear systems is assured by compliance

with toughness requirements set forth in 10 CFR

Part 50, Appendices G and H. Recently the Nuclear

Regulatory Commission has issued Regulatory

Guide 1.99, which describes procedures for

predicting the effect of copper and phosphorous on transition temperature shift and upper shelf Charpy

energy decrease of. irradiated pressure vessel

materials. The minimum upper shelf energy during the

life of the plant is set at 50 ft-lb to insure

sufficient toughness to prevent low energy ductile

tearing. If the upper shelf level falls below the

50 ft-lb limit, the following actions may be required

for continued licensability:

1. Extensive testing and analysis of existing

surveillance specimens, where available.

2. Generation of irradiated fracture toughness data

on relevant steels,

3. Extensive fracture mechanics analysis of the

reactor vessel.

4. Extended outage to comply with nondestructive

examination requirements.

5. In extreme cases, an in situ anneal with

consequent prolonged outage or premature plant

retirement could result.

It should be noted that such consequences could also

result from inadequate data, even though the vessel

may in reality be within safe limits. When credible

surveillance data from a reactor are not available,

prediction of radiation damage must be done in

accordance with procedures given in Regulatory

Guide 1.99.

Unfortunately, many installations do not have adequate

surveillance data and thus must rely on Regulatory

Guide 1.99 predictive methodology. The reasons for

the lack of proper data include:

1. The materials incorporated in some vessel

surveillance programs were selected on the basis

of unirradiated materials properties and not on sensitivity to radiation embrittlement. This

situation occurred because the deleterious

effects of copper and phosphorous were not known

at the time of the materials selection. Thus,

these reactors have not included those radiationsensitive materials, which will be controlling,

in their surveillance programs

2. The surveillance capsules in some reactors were

removed because of fretting fatigue damage

resulting from flow-induced vibration. These

plants presently have no active surveillance

dosimetry programs.

Predictive Methodology of Regulatory Guide-1.99

The curves for prediction of transition temperature

shift and shift in Charpy shelf energy are shown in

Figure 1. These curves are based in large part, upon

the data compiled by Bush'^^ for all reported reactor

surveillance capsules tested through 1973.

Serious questions have been raised about the

appropriateness of the utilization of these curves for

prediction of irradiated properties. Specific

objections include:

1. The data base is incomplete and sometimes

inapplicable. The inclusion of low exposure

temperature Yankee Rowe surveillance data

(irradiated at 450-525''F) results in prediction

of greater than normal embrittlement,

2. The trend curves should be statistically

developed, rather than simply bounding the

available data,

3. Weld metal, base metal, and heat affected zone

are unique materials (with different responses to irradiation), but the Guide requires that a single curve be used to predict property shifts

for all materials.

4. The data base used to develop the shift curves

includes data of doubtful accuracy. Source of

possible errors include:

a) Accuracy of dosimetry, particularly of older

data, is in question. Estimates of possible

error range as high as +60%.

b) Calibration information for test machines is not available from much of the data; thus,

testing errors could be significant.

5. The data base includes information from A302-B

steel irradiations. This data may not be

applicable for prediction of property shift for

A533-B, Class 1 steel used in present-day

vessels

.

6. Hawthorne has recently finished a preliminary

assessment of the degree of conservatism in NRC

Regulatory Guide 1.99. His data for A533-B plate

and weld show the Guide to be extremely

142
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<uo

c

b3

c

o

a.

o

c

o

'+-< c3

nj(U3

a

E

cuo

c

c

<uE*-*

<

O)

01c

LU

00

0) QC

00

I- CM ^2

0) U O CJ

CQ

tf^ ^

ro o n

^ OJcoOlOOlflO

m 2 o o o o o

X

2 >

A

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Eo

c

u

c

3

O o o

CM O 00 to CM X

CM

c

(U

Q

CM

.2

iZ

0) O

FLUENCE n/cm2 IE>1MeV)

Figure 3. Data Used to Develop NRC Reg. Guide 1.99

143
conservative in the prediction of upper shelf

Charpy energy drop at low fluence. A summary of

these data is shown in Figure 2.

Because of the aforementioned problems and

inaccuracies inherent in the present trend curves of

Regulatory Guide 1,99, a new trend curve analysis

should be formulated.

To demonstrate the impact of less than optimum

dosimetry data, refer to Figure 3. This figure shows

the correlation of a fracture toughness property as measured by the Charpy shift of various pressure

vessel steels versus fluence. The large scatter in

the data is evident. This data base impacts directly

on the licensing of nuclear reactors in the U.S. The

fluence limits for pressure vessel steels with various

Cu and P content as given in Reg. Guide 1.99 is shown

superimposed. The benefit of more precise and

accurate dosimetry as well as the correlation of the

data with spectral indices is evident.

Conclusion

Dosimetry beyond the core is an essential part of our reactor technology. In the materials test program, it

is the link relating changes in the micro- and

macroscopic material properties to neutron fluence and

spectra. For pressure vessel surveillance programs,

dosimetry is needed for the interpretation of actual

or projected damage effects throughout reactor life

and extrapolation from the surveillance testing

positions to the pressure vessel wall itself. The

dosimetry for shielding is a result of various

problems receiving general attention for all types of

reactors. These shielding requirements have given

rise to interrelated programs of evaluation and

benchmark experiments which coordinate closely with

dosimetry benchmark programs.

References

1. M. Vlasov, "Proceedings of a Consultants' Meeting

on Integral Cross Section Measurements in

Standard Neutron Fields", Vienna, November 15-19,

1976, Report INDC (NDS )-81 /L , IAEA (1977).

2. R. Odette, et al . , "Application of Advanced

Irradiation Analysis Methods to Light Water

Reactor Pressure Vessel Test and Surveillance

Programs", Proc, of the ASTM 8th International

Symposium on the Effects of Radiation on Structural Materials, St. Louis, Mo., May 1976.

3. E. M. Oblow, "Reactor Cross Section Sensitivity

Studies Using Transport Theory", ORNL-TM-4437

,

April (1974).

4. M. L. Williams and W. W. Engle, Jr., N.S,E 62_, 92

(1977).

5. U. Farinelli, "Proc. of the Specialists' Meeting

on Sensitivity Studies and Shielding Benchmarks",

page 25, OECD, Paris, 7-10 October (1975).

6. J. Grundle and C. Eisenhauer, "Proceedings of a Consultants' Meeting on Integral Cross Section

Measurements in Standard Neutron Fields, IAEA,

Vienna, November 15-19, 1976,

7. Proceedings of the First International

ASTM-EURATOM Symposium on Reactor Dosimetry,

Petten, The Netherlands, September 22-26, 1975.

8. W. N. McElroy, et al . , "Proceedings of a

Consultants' Meeting on Integral Cross Sections

Measurements in Standard Neutron Fields", Vienna,

November 15-19, 1976.

9. M, Vlasov, op. cit.

10. A. F. Avery and S. D. Lympany , "Proc. of the

Specialists' Meeting on Sensitivity and Shielding

Benchmarks", OECD, Paris, October 7-10, 1975.

11. V, Herrnberger, et al . , op. cit,

12. G. Hehn and P. Stiller, "Prediction of Radiation

Damage in Structural Materials Outside the

Reactor Core", Paper 32, Proc. of 1st

ASTM-EURATOM Symposium, Petten, The Netherlands,

September 22-25, 1975.

13. Nuclear Technology, 25^, 1975.

14. Proc. of a Consultants' Meeting on Nuclear Data

for Reactor Neutron Dosimetry, IAEA, Vienna,

September 10-12, 1973.

15. Proc. of the ASTM-EURATOM Symposium on Reactor

Dosimetry, Petten, The Netherlands, September

1975.

16. Private Communication, W. McElroy.

17. W. N. McElroy, et al . , Special Series, Nucl.

Techn. 25^, 177-422 (1975).

18. G, DeLeeuw, "In-Pile Neutron Spectroscopy:

Status", CEN/SCK Report (1976),

19. R. Gold, "Neutron Spectrometry for Reactor

Applications: Status, Limitations and Future

Directions", HEDL-SA 901 (1972^) and in Ref. 2.

20. H. Farrar, E. P. Lippincott, and W. N. McElroy,

"Dosimetry for Fluence Applications", HEDL-SA 939

(1975), and in Ref. 2 .

21. J, H. Roberts, "Absolute Flux Measurements of

Anisotropic Neutron Spectra with Proton Recoil

Tracks in Nuclear Emulsions", Rev. Sci, Instr,

28^, 677 (1957),

22. R. Gold, "Energy Spectrum of Fast Cosmic Ray

Neutrons Near Sea Level", Phys. Rev. 165 , 1406

(1968).

23. T. lijima and S. Momoto, "Measurements of

Anisotropic Fast Neutron Spectrum with Nuclear

Emulsions", Nucl. Sci, and Eng. 22, 102 (1965).

24. L. Rosen, "Nuclear Emulsion Techniques for the

Measurement of Neutron Energy Spectra",

Nucleonics JJ, 7, 32 (1953).

25. J. H. Roberts and A. N, Behkami, "Measurements of

Anisotropic Fast-Neutron Spectra with Nuclear

Emulsion Techniques", Nuclear Applications 4, 182

(1967), ~

26. J. H. Roberts and F. E. Kinney, "Measurements of

Neutron Spectra by Use of Nuclear Emulsions

Loaded with ^Li Glass Specks", Rev. Scien. Instr.

28^, 510 (1957).

27. R. L. Fleischer, P. B. Price, and R. M. Walker,

Nuclear Tracks in Solids, University of

California Press, Berkeley, California (1975).

144
28. "Operating Unit Status Report." U.S. Nuclear

Regulatory Commission, NUREG-0020-1 , August 1976.

29. Title 10 - Chapter 1, Part 50, "Licensing of

Production and Utilization Facilities", Federal

Register, Vol. 38, No. 130, July 1973.

30. Regulatory Guide 1.99, "Effect of Residual

Elements on Predicted Radiation Damage to Reactor

Vessel Materials", U.S. Nuclear Regulatory

Commission, July 1975.

31. S. H. Bush, "Structural Materials for Nuclear

Power Plants", ASTM Journal of Testing and

Evaluation, Vol. II, No. 6, November 1974.

32. R. Odette, N. Dudey, W. McElroy, R. Wullaert, A. Fabry, "Application of Advanced Radiation

Analysis Methods to LWR Pressure Vessel and

Surveillance Programs", ASTM 8th International

Symposium on Effects of Radiation on Structural

Materials, St. Louis, Mo., May 1976.

33. J. R. Hawthorne, "Radiation Effects on Vessel

Steels and Welds with Varied Residual Element

Content and Embrittlement Relief by

Postirradiation Annealing", Fourth Water Reactor

Safety Research Information Meeting, Nuclear

Regulatory Commission, September 1976.

145
FISSION YIELDS: MEASUREMENT TECHNIQUES AND DATA STATUS

W. J. Maeck

Idaho National Engineering Laboratory

Allied Chemical Corporation

Idaho Falls, Idaho 83A01

Techniques for the measurement of absolute and relative fission yields are reviewed.

The preferred techniques are isotope dilution mass spectrometric measurement of the

individual fission products, and the heavy element mass difference or the fission product

summation method to establish the number of fissions. The accuracy of most thermal fission

yields appears adequate, and the status of the various yield compilations is reviewed.

For fast fission yields, the data are not nearly as well established. For many heavy

nuclides, fast fission yield data are nearly nonexistent. Because fast fission yields

change with neutron energy, it is imperative that fast yield data be evaluated as a function

of neutron energy to generate the most complete and accurate compilation.

(Fission yield measurement techniques, counting, mass spectrometry; fission yields, thermal, fast, absolute, relative;

data compilations; fission yields versus neutron energy)

INTRODUCTION AND HISTORY

Fission product yield data constitute one of the

most important sets of basic information in the nuclear industry. Fission yield data are important in

the measurement of nuclear fuel burnup; safeguards;

neutron dosimetry and flux measurements; nuclear

physics calculations; reactor design and operation

(both from the physics and engineering standpoint)

;

decay heat studies; shielding calculations; fuel

handling and reprocessing; and environmental studies.

The measurement of fission yields is as old as the discovery of fission itself. Hahn and Strassman"*" performed the first yield measurements in their

initial discovery of the fission process. Fission

yield studies in the early 1940 's established the

asymetric mode of thermal fission; however, the shape

of the mass yield curve was not well defined because

of the rather large uncertainties associated with the

data. In the 1950' s, the shape of the mass yield

235 curves for thermal fission, especially for U,

became reasonably well defined and the existence of

"fine structure" was established. Similar data for

233 239

U and Pu thermal fission became available in

the early 1960's.

Up until the early 1960 's the majority of the

fission yield data had been obtained using radiochemical techniques. In the mid 1960 's the demand

for improved and new yield data increased. Radiochemical data lacked the desired accuracy because of

uncertainties in the decay schemes, half-lives, and

in-pile and out-pile decay corrections. These factors

also tended to limit the applicability of the technique. A more accurate technique for the measurement of fission yields is mass spectrometry because

concentration measurements for the stable isotopes

using the isotope dilution technique are more accurate than counting of radioactive isotopes, and

several isotopes of the same element can be measured

simultaneously. Thus, several laboratories became

involved in fission yield measurements using mass spectrometric techniques with the most extensive work

2 being done at Idaho Falls . The stable and long-lived

isotopes of Rb, Kr, Sr, Zr, Mo, Ru, Cd, Xe, Cs, Ba,

La, Ce, Nd, Sm, Eu, and Gd can now be routinely determined using the isotope dilution mass spectrometric technique.

Up to 1970, most nuclear energy laboratories

had emphasized thermal fission yield measurements

and over 500 literature references are available.

The shift from thermal yield measurement to fast

yield measurement started about 1970. Present fast

fission yield measurements being produced are based

on radiochemical measurements for samples irradiated

in low-flux fast reactor critical assemblies, radiochemical measurements for samples irradiated with

beams of monoenergetic neutrons, and mass spectrometric measurements on samples irradiated in high flux

fast reactors such as EBR-II. In this latter phase,

the most extensive efforts are again being conducted

at the Idaho National Engineering Laboratory (INEL)

3 4 near Idaho Falls '

.

TYPfeS OF FISSION YIELD MEASUREMENTS

Fission yield data basically can be divided into

two types, relative fission yields and absolute

fission yields. Depending upon the specific application of the data, each have their merits; however,

the most fundamental needs are for absolute data.

Unfortunately, these are the most time consuming and

difficult to obtain. In the following discussion,

fission yields refer to cumulative or end-member

chain yields after delayed neutron emission.

Absolute Fission Yields

Fission yield is defined by the following re- lationship :

% FY^ = —X 100

where FY^ is the fission yield of a given nuclide,

X, N , is the number of atoms of the given nuclide

X

formed, and F is the number of fissions that occurred.

In the measurement of absolute yields, the most

difficult measurement is the number of fissions.

Generally, the number of the selected fission product

atoms can be determined with smaller uncertainties.

Measurement of number of fissions - At least

four techniques have been used to establish the number

of fission events. These are: 1) heavy element mass difference, 2) fission product summation, 3) fission

counting, and 4) calculational methods. A discussion

146
of each of these techniques and their associated un- certainties follows

:

1. The number of fissions, F, based on the

heavy element mass technique is derived from

the relationship,

F = U° - (U+U^p)

where U° is the accurately determined number

of heavy element atoms being irradiated, U

is the number of atoms of the heavy element

remaining after the irradiation, and U^p is

the number of atoms of the heavy element

which has been converted to another element

by a capture or decay process. In this

method, an amount of the well characterized

(chemically and isotopically) heavy element

is accurately weighed into the irradiation

capsule and the capsule is sealed. Following irradiation, the capsule and its contents

are dissolved and the amount of the heavy

element and its isotopic composition are accurately measured, usually by mass spectrometry. Also measured, is the amount

of other elements formed by capture and/or

decay processes.

Heavy element mass difference is potentially

the most accurate method for establishing

the number of fissions, provided that the

loss (ie, burnup) of the target element is

10% or greater. Using isotope dilution mass

spectrometry, the pre- and post-irradiation

content of the sample can be determined to at least 0.1%; however, even with a 0.1%

absolute uncertainty, the error in the

number of fissions is '^1.5% relative for a burnup of 10% [(100±0.1) - (90±0.1) = 10±0.14]. Therefore, irradiation to 20% - 40% burnup is preferred, such that the number

of fission can be determined to better than

1%.

The sources of error in this method are:

a) the characterization (absolute number

of atoms) of the heavy element target,

b) the accuracy of the mass spectrometric

analysis, including the spike isotope calibration, c) incomplete dissolution of the

sample, and d) contamination. To minimize

errors which might arise from changes in the

response of the mass spectrometer or from un- known changes in the isotope dilution spike

isotope concentration, it has been strongly

recommended^ that an archive sample of the

heavy element be retained and analyzed when

the post-irradiation heavy element analysis

is performed.

2. A major advantage of the fission product sum2 ation technique , compared to the heavy element mass difference method for determining

the number of fission is that it is especially applicable to samples having low burnup

(1-10%). The summation technique is based

on the fact that the sum of all of the

fission products in one of the peaks in the mass yield curve is equal to the number of

fissions. The success of this technique is

directly dependent upon the fraction of the

selected mass peak which is measured. The

preferred peak for measurement is the heavy

mass peak because the elements in this peak

are more easily measured. The preferred

measurement technique for the individual

fission product atoms is isotope dilution

mass spectrometry.

In our laboratory, approximately 90% of the

4 fission product atoms in the heavy mass peak (isotopes of Xe, Cs, Ba, La, Ce, Nd,

and Sm) are measured to an uncertainty of

better than 1% relative. The number of atoms

of the unmeasured isotopes is obtained by

interpolation and extrapolation, and an un- certainty of 10-20% relative is assigned to these values. This procedure only slightly

increases the uncertainty in the total

number of fissions. This technique is capable of establishing the number of fissions

to an uncertainty of 1.0-1.5%.

A requirement of this method is that the

number of fission events should approach

18

10 such that sufficient quantities of

fission products are available for measurement and the amount of natural contamination

is minimized.

The principal sources of error in this technique are: a) measurement of the individual

fission product atoms, including the uncertainties in the concentration of the various

spike isotopes, b) incomplete dissolution of

the target nuclide and all fission product

nuclides, c) introduction of fission product

element natural contamination from reagents

and equipment, d) leakage of fission gases

from the target assembly during and after

irradiation, and e) incomplete collection of

fission gases.

3. Two methods are used to measure the number

of fission events directly. These involve

use of fission counting chambers and solidstate track recorders (SSTR's). However,

these methods do not measure the number of

fissions which have occurred in the sample

which is measured for fission products. In both of these methods two target foils are

used. Both targets are adjacent and

irradiated in the same neutron flux. One

of the targets is a bare thin foil from

which the fission fragments escape and are counted, and the other is a heavy target,

usually wrapped in aluminum foil to contain

the fission fragments. The heavy target

is analyzed for the desired fission products.

The number of fissions occurring in the thin

foil is counted, and the number of fissions

occurring in the heavy target is calculated

from the relative amount of the target nu- clide in each foil.

When fission chambers (ionization chambers)

are used, the number of fissions occurring

in the thin foil is established by direct

counting of the fission fragments escaping

from the foil. In the SSTR technique the

thin foil is irradiated in direct contact

with a material such as mica. The recoil

fission fragments produce structural damage

in the mica such that when the mica is

etched with hydrofluoric acid, the damage

is revealed in the form of individual tracks.

The tracks are counted under a microscope,

and the number of fissions is calculated by

multiplying the number of tracks by an optical efficiency factor. An excellent

detailed discussion of these techniques and

their associated sources of error has been

147
presented by Grundel, Gilliam, Dudey, and

, 6 Popek . The use of these techniques is limited to low fluence and hence, a low number of

fission events in the heavy foil. The heavy

foil is dissolved and the desired fission

products are separated and analyzed by counting, or the heavy foil is counted directly,

preferably using a well calibrated high

resolution Ge(Li) detector system. Accuracy

in the measurement of certain nuclides is

improved by counting at different time

periods following the irradiation. Because

the number of fission events is low, the use

of mass spectrometry to determine the individual fission product abundances for fission

yield measurements is not practical.

The two major sources of error in both techniques are: the measurement of the number of

target atoms in the thin foil and the re- lationship between the number of recorded

events and the number of fissions. Other common sources of error are: a) production

of fission events from other nuclides in the

target material, b) contamination in the

foil backing or in the materials of con- struction, c) fission fragment adsorption

in the thin foil deposit) , and d) fission

fragment scattering.

Sources of error distinct to the fission

chamber technique are: a) efficiency and

calibration factors for the chamber, b)

correction for dead time losses in the

fission rate measurement, c) background

counting corrections, and d) scattering

corrections. Based on very well controlled

experiments, the number of fission events

determined by use of a fission counting

chamber can be established with an uncertainty of '^'1.5%^. The major distinct source of error for SSTR's

is the optical efficiency factor for con- verting the number of tracks to fission.

Other distinct sources of error are the

counting of the number of fission tracks,

the etching of the mica plate, and the

identification of the tracks which result

from fission fragments. Currently, the

estimated uncertainty for the determination

of the number of fissions by this technique

is -^3%^. The number of fissions also can be calculated using some measured quantity and an assumed value for a given factor, such as

a cross section or the capture-to-fission

ratio, a. For example, the number of

235

U fissions can be calculated from the

measured pre- and post-irradiation isotopic

235 composition of '""''V and an assumed value

235

for a for U. Because a varies with

neutron energy and temperature, its true

value is always in question. Other calculational methods based on neutron fluxes

and cross sections also are subject to significant uncertainties. This is the

least accurate of the various method discussed.

Measurement of fission product atoms - The oldest

method for measuring fission product concentrations

is based on counting of a given radioactive nuclide.

Briefly, the irradiated target is dissolved, the desired nuclide is isolated by chemical means, and the

activity of the separated product is measured. All

of the early yield data were based on beta counting,

because the gamma/disintegration values were unknown.

In some cases, especially where the activity level

is very low, beta counting is still used today.

Precise gamma ray spectrometry has minimized the

need for exhaustive chemical separations because

discreet gamma rays can be associated with a given

nuclide. In some instances, no chemical separations

are performed and the irradiated target is counted

directly for selected radionuclides. The preferred

counting technique is high resolution gamma ray spectrometry. Counting techniques are especially useful

for the measurement of short-lived nuclides, or where

the number of fissions is low.

Sources of error in converting gamma ray

spectrometric data to atoms of the particular fission product are: a) the method for

subtraction of the baseline or background activity,

b) detector efficiency, c) the accuracy of the standards if a comparison counting technique is used,

d) the nuclear constants, specifically the photon-todisintegration ratios and the half-life values,

e) the mass adsorption of the gamma rays by the

material being counted, f) random coincidence summing,

and g) interferences from other gamma rays.

In a limited number of cases, an uncertainty of

1-2% can be attained for the determination of the

number of fission product atoms using high resolution

gamma ray spectrometry. Uncertainties in the range

of 3-10% are more common.

The most accurate method for the measurement of

the concentration of stable and long-lived nuclides

is isotope dilution mass spectrometry. In this technique, a known number of spike isotope atoms, preferably of an isotope not formed in the fission process, is added to a sample of the dissolved target

and chemical identity of the spike isotope and sample

is effected by chemical means. That element is separated chemically and the isotopic composition is

measured using a mass spectrometer. The ratio of the

spike isotope peak to the other mass peaks of that element is measured and the number of atoms of the other

isotopes is determined relative to the number of spike

isotope atoms added. A particular advantage of this

technique is that quantitative recovery in the separation procedure is not required. It is imperative,

however, that isotopic exchange be achieved.

The irradiated sample should have experienced

18 about 10 fissions so that a reasonable amount of

material is produced for measurement. Because the

ionization efficiency in the mass spectrometer varies

from element to element, the minimum amount of a fission product required for a reliable mass spectrometric measurement also varies. For example, 100

nanograms of Rb or Cs on the mass spectrometer filament is more than adequate, while several micrograms

of Ru are ideal for analysis.

148
Because the accuracy of any isotope dilution

technique is controlled by the accuracy of the spike

isotopes, great care must be taken in their preparation and standardization. In many cases, the occurrence of natural contamination can be corrected

because one or more of the naturally occurring isotopes is not formed in fission. For those cases where all of the naturally occurring isotopes of the

element also are fission product isotopes (eg, Rb,

Cs) significant errors can occur. Spectral interference from an isotope of another element can also

occur; however, this problem can usually be minimized

by improved chemical separation techniques. Other

factors which must be considered are mass fractionation, especially for the light elements, from the mass spectrometer filament, and mass discrimination

corrections.

For most elements, the isotopic composition

and atom abundances can be determined to an uncertainty of '\'0.25%.

Other techniques, although limited in scope,

which have been used to measure selected fission pro99 2 duct concentrations are spectrophotometry for Tc , X-ray analysis for one or more rare earth elements^,

2 vapor phase chromatography for the fission gases ,Q and volumetric determination of the fission gases .

Summary - Absolute Fission Yields - The preferred methods for the measurement of absolute fission

yields is the isotope dilution mass spectrometric

measurements of the individual fission products

coupled with the heavy element mass difference or the

fission product summation technique to establish the

number of fissions. Based on state-of-the-art, more

yield data with smaller uncertainties can be obtained

from a well designed irradiation using this technique

than from the others which have been discussed. A

particular item of concern, however, is that while

irradiations to high burnup can result in a more

accurate value for the number of fissions, significant corrections for neutron capture on some of the

fission product nuclides are required. This is

especially true for irradiations conducted in thermal

reactors. Although correction factors can be applied

when sufficient information concerning the irradiation is available (for example, see Ref. 9) the

existence of significant unknown cross sections can

lead to biased fission yield data'^'^. Because of this,

it is believed that the more correct approach may be

to irradiate a larger target to a low burnup, 1% or

less, and to determine the number of fissions by the

summation technique using isotope dilution mass

spectrometry"'"^. Yields with uncertainties in the

1-2% range are attainable using this approach.

generally displaced one or two Z units from the endmember of the chain. Thus, the yield may have a lower value than that determined on the stable endmember of the chain because of the independent yield

contribution from the other non-measured members of

the chain. This difference may be as large as 5%

relative.

Relative Fission Yields

Relative fission yields, which constitute a major fraction of the reported yield data, are yields

which have been measured relative to some other

nuclide or normalized to some value. The basic assum- ption is that once the yield of one fission product

is accurately known, from absolute yield measurements,

is can be used as a monitor of the total number of

fissions, or other yields can be measured relative to

it. Nuclides which have been used as reference stanj J 95^ 97, 99,, 137^ 140„ ^ 148„^ dards are Zr, Zr, Mo, Cs, Ba, and Nd.

The bulk of the relative yield data is associated

with radiochemical measurement.

Several variations of treating the data on a relative basis have been proposed over the years, but

the one receiving the most attention is the R-value

method. This method is based on the assumption that

if the yield values for a given fissionable isotope,

235

usually U, are well known, that uncertainties in

the radiochemical measurement, especially those

associated with decay schemes and counting geometry,

can be minimized. In practice, samples of two different fissionable isotopes are irradiated simultaneously, and each is chemically treated to obtain the same nuclides from each source. The reference

monitor nuclides, for example Ba, from each target

isotope and the unknown from each source are then

counted as close in time as possible. Assuming that

the yield of the reference nuclide is known, the

yield of the unknown can be determined. Various ex- perimental techniques and prior references are given

12

in a review by von Gunten

The major obvious systematic source of error in

any type of relative yield measurement is the accuracy

of the reference standard. Because absolute thermal

yield values are now reasonably well known, relative

yield measurements for thermal yields is a viable option. It is especially useful in the measurement of

the valley nuclides and those on the extreme wings of

the mass yield curve where absolute mass spectrometric

yields have not been determined. The application of

this technique to fast yield measurements suffers

from the fact that the yields are a function of the

energy of the incident neutrons; hence, relative fast

fission yield data must be carefully evaluated with

respect to the assumed yield of the reference nuclide.

Recent improvements in fission chamber counting

coupled with direct gamma-ray spectrometry measurements have resulted in a limited number of absolute

yields being reported with uncertainties of 'v2. 5%"'""'". This approach should be most valuable in the measurement of yields where irradiations are conducted in

low flux critical assemblies or when mono energetic

neutron beams are used.

A feature of any counting technique compared to the mass spectrometric techniques, is that in most

cases, the yield of the nuclide being determined is

FISSION YIELDS - STATUS AND NEEDS

To date, approximately 1000 literature references

relative to all types of fission yield measurements

are in existence. Recognizing that it would be impossible for each user to evaluate and select the

best value from this mass of data, several individuals

have taken up the task of evaluating and compiling

these data. As could probably be predicted, differences occur in these evaluated data sets which exceed

the error assignment attached by the various evaluators. In attempting to clarify the situation, the

149
most used compilations of thermal and fast fission

13 14

yields were reviewed by Walker and Cuninghame , respectively, for the 1973 IAEA Panel Meeting on Fission Product Nuclear Data''"^. The current status

of the yield data compilations and data requirements

for thermal and fast yields are discussed below.

Thermal Fission Yields

The review, evaluation, and compilation of data

from nearly a thousand individual literature references is a monumental task. Several individuals have

undertaken this task with varying degrees of vigor and

continuity. Of these, the most well documented and

referenced are the works of Meek and Rider"*"^,

E.A.C. Crouch"*"^, W. H. Walker^, Lammer and Eden"'"^,

and the U.S. Evaluated Nuclear Data File (ENDF/B-IV

task force group.

19

All of the compilations are based on some version

of pooling and weighting techniques, plus the personal opinion and preferences of the individual

evaluators. Briefly, Meek and Rider''"^ attempt to include all reported data in a computerized data library

and then assign various weighting functions to the

reported data based upon the method of measurement.

Crouch''"'^, whose work is sponsored by the U.K. Chemical

Nuclear Data Committee, also uses a computer based

library but tends to limit the input to well documented data. The errors are assigned by Crouch.

9 Walker's compilation is based primarily on mass spectrometric yield data, except where mass spectrometric data are not available and then radiochemical

18

data are used. Lammer 's and Eden's compilation is

not as extensive as the others and is primarily based

on mass spectrometric data and the evaluation tech9 nique is similar to that of Walker . Version IV of

of the ENDF/B file is basically the same as the Meek

and Rider compilation.

For a more detailed discussion of these data

sets and a comparison study of the data sets, it is

strongly recommended that the reader consult Walker's

13 excellent review presented at the 1973 IAEA Bologna

Fission Product Nuclear Data Conference.

Since that time, the primary emphasis in the U.S.

has been an intensified effort to upgrade and update

the ENDF/B data file. To this end, a task force group

which includes, among others, B. F. Rider and

W. H. Walker, has been formed under the auspices of

U.S. ERDA to develop a preferred fission yield data

file. This file is to be continually updated and

Version V should be available for presentation at the

Second IAEA Advisory Group Meeting on Fission Product

Nuclear Data to be held in Petten, Netherlands,

September 1977. Concurrent with the U.S. effort, the

UK fission yield data file is in a continual upgrade

under Crouch's direction. Rider and Crouch are con- tinually exchanging information with respect to literature references. The update status of the Lammer

and Eden compilation is uncertain at this time. A

review of the current status of the various thermal

fission yield compilations is being prepared by

J. G. Cuninghame, UK, for presentation at the Petten

Conference mentioned above.

At the Bologna IAEA meeting , it was concluded

that the thermal yield data, except for certain isolated cases, generally satisfied the user's required

accuracies, and no new extensive remeasurement program was justified. The exceptions are detailed in

Volume 11 of Reference 15.

In 1975, we had the opportunity to remeasure a 235 limited number of thermal fission yields for U and

239

Pu. Although to date, only relative yields have

20

been measured, the initial results of this study

show for several isotopes that there are large differences compared to previous measurement made in this

2 laboratory and compared to data in current fission

yield compilations"'"^. Particularly significant are

138

14% higher values for Ba and 8.5% higher values

239

for the Xe isotopes for Pu thermal fission. The

239 major impact is that all of the Pu fission yields

will have to be adjusted to preserve mass balance.

235

Only minor differences were observed for U thermal

fission. It is expected that new absolute thermal

235

yield data for about 40 different nuclides for U

239 and Pu will be produced when this measurement program is completed in 1978.

Fast Fission Yields

Fission yield data for fast neutrons are not

nearly as well developed as they are for thermal

neutrons. Based on the fast fission yield compilations produced through 1973 and Cuninghame' s review

the data are fragmentary and usually carry large

uncertainties. The most complete fast yield compil235 239

ations are for U and Pu which were produced by

16 21

Meek and Rider and Crouch . The fast yield data

232^^ 233,, 237„ 238,, , 241. . , for Th, U, Np, U, and Am are limited

and generally have large errors. No reliable fast

yield data are given for the higher isotopes of

plutonium.

This lack of data stems from several sources. Primary, is the lack of available irradiation facilities. Because fast reactor fission cross sections

are low, prolonged high flux irradiations are re- quired to generate a sufficient number of fission

product atoms for mass spectrometric measurements.

If this is not possible, irradiations can be ma'de in

experimental low power critical assemblies; however,

normally only a limited number of radiochemical

measurements can be obtained.

Another consideration is that accurate fast

yields are more difficult to measure than thermal

yields, because all of the heavy nuclides undergo

fast fission, while thermal fission is only signifi233 235 239 241

cant for ''u, U, Pu, and Pu. This is

especially important in the measurement of fast

fission yields for the even-even and odd-odd heavy

^232^,. 238,, 237,^, 240„ 242^ , nuclides ( Th, U, Np, Pu, Pu, and

241

Am) , because all have a high neutron energy

threshold cross section for fission, a relatively

high capture-to-fission ratio, and in some cases,

a long-lived capture product which fissions with any

energy neutron. For example, in a long-term fast

14

150
neutron irradiation of ^^^U or '^'*^Pu, a large fraction

239 241

of the fissions will result from Pu and Pu,

respectively. While corrections for the fission from

the capture product can be made if the fission yields

for the fissioning capture product are well known,

this correction can result in significant errors in

the fission yields for the principal target nuclide.

Before going further, the term "fast fission

yield" should be addressed. It is at best a relative

term because it is well known that fast fission

yields can vary with neutron energy. This one item

has made it most difficult for the evaluators and

compilers of "fast yields" to produce accurate com- pilations, because it prohibits use of the pooling

techniques which have been applied to thermal yield

data. Unfortunately, because most reported experimental fast yield data can not be associated with a known neutron spectrum, the compilers to date have

had no choice other than to pool the data. Thus, it

is not unreasonable that fast yield compilations show

significant discontinuities for adjacent yields

because the data were probably obtained from irradiations conducted in significantly different neutron

spectra. It is appreciated that the early worker

probably had little regard for the neutron spectrum

effect on yields and that just obtaining fast yield

data was of primary concern; however, it is strongly

emphasized that all new measurements be associated

with a known neutron spectrum.

240, The effect of neutron energy on fission yields is

shown in Figure 1. In this presentation, the ratios

22 235 of the fast to thermal yields for U are plotted

as a function of mass number. The general result of

fissioning with increasing energy neutrons is an increase in the yields on the wings of the mass yield curve, an increase in the valley yields, and a small depression in the peak yields. This effect

appears to change systematically as the energy of the

fissioning neutron increases. The discontinuities

in the plotted data may be due to certain fine

structure effects or the result of inaccuracies in

22

data source

An extensive fast yield measurement program is

currently being conducted in our laboratory at the

Idaho National Engineering Laboratory. Mass spectrometric yields are being measured for over 40 nuclides

, 233,, 235,, 238,, 237^, 239„ 241„ 242„ for U, U, U, Np, Pu, Pu, Pu, and

241 2 4 Am which were irradiated in EBR-II '

. Yield data

233 235 238 239 for U, U, U, and Pu and neutron spectral

4 240 data were published in 1975 , and data for Pu,

2A1 2A2 237

Pu, Pu, and Np should be available by the

end of 1977.

Of primary concern in this program are the fast

fission yields for the isotopes of Nd which are used

for monitors for the determination of burnup in fast

<

(TX

I-

>-

<

u.

iO

• •••

1.0

0.8 -L

• •• •

75 80 90 100 I 10 120 130 140 150 160

MASS NUMBER

Fig. 1 Uranium-235 Fast to Thermal Fission Yield Ratios.

151
TABLE I. RELATIVE Nd ISOTOPIC COMPOSITIONS AND ISOTOPIC RATIOS FOR ^^^U FISSIONING

(if l(ff Spectral Index,

8 ^5

Reactor , and Reference

SI df ICff ^8 ^8

Isotopes

Thermal-'-^ 0.0

EBR-Il5

Axial Bl.

0.003

EBR-II^

KOW—

O

0.02

23

0.035

EBR-Il5

Row—

4

0.06

UolKlb

0.089

EBR-I^

Core

0.12

0.3923 0.3898 0.3891 0.3870 0.3840 0.3836 0.3826

0.2583 0. 2589 0.2563 0.2544 0.2545 0.2546 0.2544

0.1964 0. 1975 0.1969 0.1981 0.1985 0.1962 0.1973

148Nd 0.1101 0. 1110 0.1126 0.1133 0.1149 0.1167 0.1161

150Nd 0.0429 0. 0429 0.0450 0.0472 0.0481 0.0489 0.0496

-L. Ow L- ^—' L/ ^ ^ Ratios

150/143 0.1094 0. 1101 0.1157 0.1220 0.1253 0.1275 0.1296

150/145 0.1661 0. 1657 0.1756 0.1855 0.1890 0.1920 0.1950

148/143 0.2807 0. 2848 0.2894 0.2928 0.2992 0.3042 0.3034

For ^"^^U, the change in the "'"^%d/

breeder reactor fuels. In the process of evaluating

our Nd fast yield data and comparing it to data re- ported by other workers, the noted differences were

considered to be too large to be attributed to measurement error. In lieu of comparing absolute yield

values, we evaluated the relative isotopic data

because systematic errors in the measured number of

fissions used to calculate the absolute yields could

be eliminated. A comparison of the relative Nd

isotopic data is given in the top half of Table I. When the data are ordered by spectrum hardness, systematic changes in the Nd isotopic composition are apparent. To amplify these changes, selected isotopic

ratios were calculated and are given in the bottom

half of Table I. 143

Nd ratio between the thermal neutron values and the

very hard EBR-I spectrum values is '^'20%. As a result of the trends shown in Table I, several different forms of neutron energy indices were considered to provide a means for the possible

correlation of yields with energy. Included were mean neutron energy, median neutron energy, mean and

median neutron energy for fission of a given heavy

isotope, fraction of neutrons in a given energy range,

and the ratio of cross section values for two selected

238 fissioning isotopes. Of these the ratio U(n,f)/

235U(n,f) was selected because more yield data could

be associated with this index than any of the others.

This ratio gives a good indication of spectral hardness

because '^95% of the neutron energy .response for

238

U(n,f) is above '^'1.4 MeV and '^'95% of the response

0.25

for

235

U(n,f) is below 2 to 3 MeV.

.04 .06

SPECTRAL

.08 0.10

INDEX

The change in the isotopic ratio. 15°Nd/l^\d,

235 239 for those U and Pu data which could be associ238 235 ated with the spectral index, U(n,f)/ U(n,f),

is shown in Figure 2. The correlation of the isotopic composition of Nd with the selected spectral

index is clearly demonstrated.

Fig. 2 Change in Isotopic Ratio of ^ SOj^jj/li+Sig^j with Neutron Energy for ^^^U and ^^^Pu

Fast Fission.

152
To extend the correlation of isotope ratios with

neutron energy to other fission products, the exten235 sive U fission yield data obtained in the author's

2 10 laboratory for thermal fission '

, fast fission in

row-8 of EBR-II , and fast fission in the core of

2 EBR-I were evaluated. The respective spectral index

values for these irradiations are zero, 0.02, and

0.124. Because little data are available between

index values of 0.02 and 0.124, the correlations

are not as accurate as those for the Nd isotopes.

Isotopic ratios were calculated for the various

fission product elements using the isotope whose

fission yield changed least with neutron energy as the reference isotope. For example, for the Kr

86 isotopes, Kr showed little change with neutron

energy and was selected as the reference isotope.

Fig. 3 Change in Fission Product Isotopic Ratios with Neutron Energy for the Isotopes of Krypton, Rubidium,

Strontium, Zirconium, Molybdenum, and Ruthenium.

153
+60r

UJi +18

<X

o

+ 10

-8

1 1

^..^-^150/146

NEODYMIUM

148/ 146

145/146 ~

.^Z^::;^-- 144/146

143/146

1

1 0.04

SPECTRAL

0.08

INDEX

0.12

Fig. 4 Change in Fission Product Isotopic Ratios

with Neutron Energy for the Isotopes of

Xenon, Cesium, Neodymium, and Samarium.

The percent changes in the ratios, relative to the

thermal values are presented in Figures 3 and 4. The isotopic ratio changes in the light mass peak

are reflected by counter isotopic ratio changes in

the heavy mass peak. For example, the change of

15 to 20% for Kr isotopes 83 to 86 on light wing

of the light mass peak is reflected by a like change

in the mass region 148 to 151 on the heavy wing of

the heavy mass peak. A second example is that the

86 to 88 mass region and its reflected 146 to 148 mass region, both change little. Another example,

is similar changes for the light Xe isotopes and

the light Ru isotopes. Little significant changes

are seen for those isotopes on the peaks of each

wing of the mass yield curve.

This study, of the changes in yields with

neutron energy will be continued with the goal

being the ability to predict a given yield for

any neutron energy. To aid in the effort to quantify the changes in yields with neutron energy,

it is imperative that every effort be made to define

the neutron spectrum associated with fast fission

+ 50

+40

o

+30

z

<

3: +20

o

+ 10

}

1 1 y

— / 154/149

— /

^..--^T52/I49

SAMARIUM

^I5m49^,^

|T4Z^749

0.04 0.08

SPECTRAL INDEX

0.12

yields being determined. For experiments currently

in progress, the workers should attempt to obtain

this information from the reactor physicist, or preferably, irradiate specific samples to define

the neutron spectrum. For new experiments, the

worker should make every effort to include a series

of spectrum monitors in the irradiation assembly.

The compilers of fast fission yield data must

use only well documented data and evaluate each

set of yield data with respect to the neutron

energy

.

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2. Lisman, F. L., Abernathey, R. M. , Maeck, W. J.,

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3. Maeck, W. J., Larsen, R. P., Rein, J. E.

,

USAEC Report, TID-26209 (1973).

4. Maeck, W. J., Editor, U.S. ERDA Report,

ICP-1050-1 (1975).

154
5. Larsen, R. P., Argonne National Laboratory,

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6. Grundal, J. A., Gilliam, D. M. , Dudey, N. D,,

Popek, R. J., Nucl. Tech. 25^ (2), 237 (1975).

7. Larsen, R. P., Schablaske, R. V., Oldham, R. D.

,

Meyer, R. J., Homa, M. I., U.S. AEC Report,

LA-4407, (1967).

8. Arrol, W. J., Chackett, K. F. , Epstein, S.

,

Can. J. Res, 27B , 757 (1949).

9. Walker, W. H. , AECL Rept. 3037-11, (1973).

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,

Helmer, R. G. , Kellogg, L. S., Zimmer, W. H.

,

Nucl. Tech. 25 (2), 294 (1975).

12. von Gunten, H. R. , BNES Proceeding of the

International Conference on Chemical Nuclear

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13. Walker, W. H. , Review Paper 11a, IAEA-169

(1973).

14. Cuninghame, J. G., Review Paper lib, IAEA-169

(1973) .

15. IAEA Proceedings of a Panel on Fission Product

Nuclear Data, Bologna, Italy, IAEA-169 (1973).

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(1974) .

17. Crouch, E.A.C., Paper IAEA/SM-170/94 Symposium

on Applications of Nuclear Data, Paris,

(March 1973).

18. Lammer, M. , Eder, 0. J., Paper IAEA/Sm-170/13,

Symposium on Applications of Nuclear Data,

Paris (March 1973)

.

19. ENDF/B-IV Fission Yield Data File, National

Nuclear Data Center, Brookhaven National

Laboratory, USA.

20. Maeck, W. J., Emel, W. A., Delmore, J. E.

,

Duce, F. A., Dickerson, L. L., Keller, J. H.,

Tromp, R. L. , U.S. ERDA Rept. ICP-1092 (1976).

21. Crouch, E.A.C., UKAEA Rept., AERE-R 7394 (1973).

22. Rider, B. F. , General Electric Co., Private

Communication (1977).

23. Sinclair, V. M. , Davies, W. , International

Conference on Chemical Nuclear Data, BNES,

Univ. of Kent, Canterbury, England (1971).

24. Robin, M. , Bouchard, J., Darrouzet, M.

,

Frejaville, G. , Lucas, M. , Prost Marechal, F.

,

ibid.
FISSION REACTION RATE STANDARDS AND APPLICATIONS

J. Grundl and C. Eisenhauer

National Bureau of Standards

Washington, D.C. 20234

Fission rate measurements in and around prototype and power reactors of all kinds,

as well as in the criticals of reactor physics are vital elements in understanding

nuclear energy generation rates, neutron transport, and the integrity of materials

exposed to reactor radiation fields. Standardization and interlaboratory referencing for this historic measurement activity have improved significantly since

the last neutron standards symposium in 1970. This advancement will be summarized

along with a general orientation and description of fission detector response

characteristics and interpretation. For the last, necessary analytic formulations

and a brief treatment of error propagation are included. Also included is an updated look at observed versus predicted fission cross sections for fission spectrum

neutrons and the related fast criticals of reactor physics.

(Cross sections; fission; neutron reactions; neutron spectrum; reactor fuels;

reactor materials)

Introduction and Summary

The requirement to measure isotopic fission rates

appears in a wide variety of technical-engineering

problems associated with radiation effects in materials

and reactor design and management. Looking only at physical changes in materials exposed to neutrons the

array is bewildering. Some examples are swelling of

reactor fuel claddings, fuel matrix rearrangements,

integrity of pressure vessel weldments, activation of

reactor service equipment, vaporization of liquid

rocket propellant, and gain changes in transistors.

In the management of these radiation effects problems,

fission rate measurements are often used indirectly to derive neutron fluence and spectrum information. Direct application of fission rate measurements are also

important and they arise most often, and with a longer

history, in the area reactor design and nuclear fuels

management. Examples are verification and guidance of

reactor physics calculations; optimization of fission

power gradients; relative fission rates among fissionable isotopes competing for neutrons in a reactor core; determination of fuel bum-up and bum-in rates in all

types of power reactors including nuclear weapons.

Techniques of fission rate measurements are also

diverse. Among active fission detectors there exist

fission ionization chambers as small as a pencil tip

and as large as a bread box; passive fission activation detectors can be as thin as a needle or as large

as a sheet of paper. Among destructive analysis techniques, mass spectrometry for example, will measure

total fissions in almost anything that can be dissolved.

Of primary concem perhaps for a neutron standards

symposium, are fission rates that give rise to fundamental integral cross sections. These are the observables that stand as constraints for the evaluation of

differential cross section data. A single review of

integral results at the previous Neutron Standards

Symposium in 1970 is matched at this meeting with a half dozen papers on integral measurement techniques

,

facilities, and results. The remarks in 19 70 regarding

differential and integral measurement as "areas of

effort that do not always communicate so well," is now almost outdated. ' The need by a much harassed nuclear

energy industry for consistent design and operation

parameters overwhelms the vestiges of mutual disrespect. Today, on one hand, an observed fissionspectrum-averaged fission cross section contributes to the setting of the ^^^U fission cross section scale

for ENDF/B-V, and on the other, reactor dosimetry spectrum characterizations routinely use the new ENDF/B

dosimetry cross section file with only mild threats to "adjust."

Measurement Standards

Accuracy requirements for many of the direct

applications of fission rates can be severe: as lew as

± (2-5)%(la) for the determination of reactor fuel bum

up or in checks of reactor physics calculations. Neutron flux and fluence estimates for managing radiation

effects problems typically are less severe and dependence upon neutron transport calculations with less

experimental verification is common. The latter is

true also, because in situ neutron flux measurements

can be difficult for radiation effects surveillance.

However, when optimum material performance is sought,

and high-investment engineering or safety decisions

are involved, stiff accuracy requirements appear. In this case, a measurement program in support of calculation assumes recognized importance. This is particularly true when critically evaluated spectrum characterization is sought and fission detectors coupled with

other types of neutron energy sensitive activation detectors are employed (i.e., the familiar multiple foil

spectrum unfolding techniques).

Standards and intermeasurement reference for all

of this varied and decades-long effort of fission rate

measurement are of two kinds: 1) artifacts, such as standard solutions or fissionable deposits; and 2) well

characterized neutron fields, such as thermal equilibrium, monoenergetic , and fission neutron spectra.

The latter are most important when fission rates are used to derive neutron field characteristics.

Not all fission rates are standardized in the manner just stated, most in fact, are not. Measurements

which depend upon extraneous instrument calibrations

and various types of nuclear data are common (e.g.,

fission yields and alpha decay constants). Until re- cently, there existed no central repository of mass assayed fissionable deposits generally available for

fission rate measurement calibration or referencing.

Considering the vital role fission rates play in all

branches of radiation effects and nuclear energy development, the deficiencies in standardization and

interlaboratory comparisons is surprising. The introductory paper by R. Taschek at the 1970 Symposium re- ferred to this odd circumstance in relation to the use

of fission rate measurements as a standard for neutron

flux determinations with accelerators.^

156
Improvements Since 1970

After 1970 notable concern and some improvements

in referencing fission rate measurements have taken

place. Recall first the operative distinctions set

forth in References 1 and 3.

Mascroscopic Data . Experimental results from

arrangements in which a dominating feature is neutron

transport, i.e., multiple encounters of neutrons with

nuclei which produce a series of velocity changes , and

in important instances, new sources of neutrons.

Microscopic Data . Experimental results , generally

energy dependent cross sections , from arrangements in

which single encounters of neutrons with nuclei predominate.

Integral Measurements . Measurements that require

for interpretation an integration of pointwise microscopic data over energy intervals and/or the neutron

flux over space intervals that are not small compared

to the range of interest.

Differential Measurements . Measurements for which

energy or space integrations are not necessary for interpretation, only for corrections to data.

(1) In the area of macroscopic data from integral

measurements, organized efforts are underway in the

U.S. to achieve interlaboratory consistency in reactor

physics and in reactor materials dosimetry.'*"^ Internationally, the IAEA has initiated a program of identifying benchmark neutron fields and evaluating their

application. A Consultant's Meeting on Integral Cross

Section Measurements in Standard Neutron Fields for

Reactor Dosimetry was convened in November, 1975.^ To

prepare for this meeting a compendium of benchmark neutron fields was prepared within the context of the

following characteristics:^

1. Simple and well-defined geometry;

2. Adequate neutron fluence and stable flux

density;

3. Reproducible and accurately characterized

neutron spectra based on spectrum measurements and/or reliable calculations;

4. Sustained availability for measurements.

(2) Both integral and differential measurements

.have benefitted from the development of a permanently

available set of reference and working fissionable deposits at the National Bureau of Standards.^ Isotopes

presently represented in the set include ^^^U, ^^'Pu,

^'°U, and ^"Np, with 2'*o,2'»ip^ added within the

next year. Traceability of fission rate measurements

to these deposits may be established by direct alpha

or fission comparison counting of deposit pairs, or by

exposure in a relevant neutron field of a complete

fission detector to be calibrated along with an NBS

double fission ionization chamber operating with an NBS mass assayed deposit.

8

(3) The situation regarding observed versus predicted fission-spectrum-averaged fission cross sections

is better now than it was in 1970.^ New absolute cross section measurements have been performed and the discrepancies with the differential microscopic data are no longer in excess of ten percent. ^''-^^ And most

important for fast neutron standardization there exists now an evaluation of the fission neutron spectrum including uncertainty estimates based on all documented

differential spectrometry .9 > ^ 3 xhe stubborn discrepancies surrounding fission rate ratios in simple macroscopic systems like the Big Ten Fast Metal Critical

Assembly also have changed drastically following the

inclusion of long-awaited changes in inelastic scattering cross sections into neutron transport calculations.

This report will review briefly and largely by

reference these various improvements in fission rate

standardization and the associated advances in applied

measurement. Preceding this and comprising the major

portion of the paper will be an orientation regarding

fission detector measurements and their use for neutron

spectrum characterizations. This is appropriate, hopefully, and no more controversial than is beneficial on the occasion of a symposium devoted to neutron standards and applications.

Neutron Spectrum Response

The neutron energy dependence of fission detection

falls neatly into two categories : 1) full-energy-range

detectors with generally flat fission cross sections in

the MeV range and increasing fission cross sections

down to thermal equilibrium energies; and 2) threshold

detectors with smooth cross sections that rise rather

sharply in the energy range 0.5 to 2 MeV. There is not

much distinction within these two categories and a small

array of fission detectors are sufficient for illustration. Four are chosen for this review: ^'^U(n,f),

^^'Pu(n,f), ^"Np(n,f), and ^''u(n,f). Spectrum re- sponses of these detectors will be summarized as an introduction to quantitative methods of interpretation.

Spectrum Characteristics

Five neutron spectra representative of much of

nuclear technology are displayed in Figure 1. Included

are two spectra associated with differential cross section verification and detector calibrations (the ^^^U

fission spectrum and the one-dimensional IntermediateEnergy Standard Neutron Field (ISNF)), two fast-breeder

related spectra (the FTR and EBR-II core) , and a typical light-water related spectrum (LWR pressure vessel

environment)

.

Two ordinate scales are employed for this display:

log [E(f>(E)] from 1 MeV down to the cadmium cut-off and

log [(()(£) //E] versus a linear energy scale above 1 MeV. They were chosen in order to show the full spectrum

energy range below 1 MeV, along with fission spectrum

components as linear slopes proportional to average

spectrum energy above 1 MeV. The neutron transport

calculations represented in Figure 1 are from multigroup computations normalized to/ct)(E)dE = 1 above 0.4

eV. As a first order approximation of the true shape

of the spectra for this plot, the histogram spectra

from the calculations were transformed to discontinuous

linear segments which approximate slopes and preserve

group fluxes.

Fission neutrons represent the top of the energy

spectrum for fission reactors and their environments.

Interestingly enough, the fission neutron energy distribution remains discernible in a number of these fission reactor environments. Measurements with threshold

detectors have established that the fission spectrum

component is well preserved above '^1.5 MeV in fast

metal criticals of ^'^U and ^'^Pu, and also in a zonedcore critical assembly with a spectrum typical of a fast

breeder.^' Reactor calculations appear to verify this

spectrum characteristic showing it to be maintained in

the cores of the U.S. Fast Test Reactor (FTR) and in

EBR-II. Even for the LWR core shown in Figure 1 the

fission spectrum component persists down to below 2 MeV.

If transport calculations at LWR pressure vessels are

to be believed, the spectrum there also resemble a fission spectrum shape over a significant part of the MeV

157
energy range. In fact, spectrum-averaged cross sections for U(n,f) and Np(n,f), truncated at threshold, for all of these spectra are within 10% of the

fission spectrum value - see Eq. (1) below.

Below the fission spectrum range a transition

energy region exists between 10 keV and 1 MeV for the

reactor spectra shown in Figure 1. This is the energy

region where breeding occurs in fast reactors and re- laxation to a 1/E equilibrium spectrum occurs in the

light-water power reactor systems. This is a difficult energy range for neutron flux measurements , and

for calculations as well, particularly in the decade

above 10 keV. Below 10 keV the four energy decades

down to thermal are characterized by a neutron slowing

down spectrum, E(|)(E) constant.

E, MeV

Fig. 1. Representative neutron spectra for materials dosimetry applications. For energies less than 1 lleV the quantity log [Et(E)] is plotted against log E. For energies greater than 1 MeV, the quantity log [*(E)//E] is plotted against E.

Fission Detector Response

Fission detection for the purpose of estimating

neutron flux intensity and spectrum belongs to a class

of integral measurements which in principle are simple

to perform and interpret. As with all integral mea- surements, however, accuracy of measurement and the

best possible understanding of detector response is

essential if quantitative results are to be obtained.

For two of the neutron fields shown in Figure 1, fission detector responses will be shown in cumulative

spectrum response plots. Figures 2 and 3. In such

plots the integral spectrum a(E)(j)(E)dE, appear to- gether. Energy-dependent cross sections for the display are from ENDF/B-IV.

For the ^'^U fission spectrum in Figure 2, the

ordinate gives the fraction of the response which is

above the corresponding abscissa energy. The threshold

fission reactions ^'^Np and ^^^U are seen to provide

effective complementary coverage for the bulk of the

spectrum. Near the 2 MeV average energy of the fission

spectrum, the ^'^U response fraction is '^' 0.8 while for

^^'Np it has reached only 'v 0.5. Both threshold re- actions cut off sharply providing a well-defined re- sponse range. The wide-energy-response fission detectors, ^^'Pu, and ^^^U, follow closely the fission

spectrum shape. Thus, they are good total flux monitors

for fission spectra, and in fact, ^^'Pu is a good total

flux monitor for all fast neutron spectra which do not

have a significant spectrum component below 10 keV.

1.00

0.95

0.9

UJ>

o

CO 0.8

2 0.6 HO

<

q:^ 0.4

0.2

10 10' 10 E. NEUTRON ENERGY

MeV

Fig. 2. Fraction of neutron flux and fission detector response above energy E in

a ^^^U fission neutron spectrum.

1.00 ++H4f__ I I I 1 1 [

10' 10^ E, NEUTRON ENERGY

Fig. 3, Fraction of neutron flux and fission detector

response above energy E in the FTR breeder core neutron spectrum.

Fission detector responses for the typical breeder

reactor core spectrum of FTR are plotted in Figure 3. The ^^^U response cut-off remains sharp, a result which

remains in need of complete verification. Some recent

measurements are helpful in this regard. The complementary coverage of ^'^Np and ^'*U is even better than

for fission spectra. The cumulative fraction 0.9 for

^'*U corresponds to a cumulative fraction of less than

158
0.4 for ^^^Np. A notable subthreshold tail for ^'^Np

extends to energies well below the 0.95 response fraction energy. This is a significant issue for spectrum

characterizations and especially for checks of reactor

physics calculations. The difficulty is that no satisfactory agreement exists regarding subthreshold fission

for ^"Np.l6 The response functions for ^^^U and ^'^Pu

in Figure 3 fall below the spectrum because of the rising cross sections which shift the detector responses

to lower energies. Almost 10% of their response lies

below 2 keV where transport calculation places less

than 2% of the spectrum. Concerns for neutron selfshielding begin for activation fission foils when significant response below 10 keV occurs. In spite of

this and similar measurement problems, the ^''Pu(n,f)

detector is still an attractive total flux monitor.

The maximum departure of the flux curve from ^^'Pu

response curve in Figure 3 is 15% and the average

^'^Pu(n,f) cross section for this particular breeder

core spectrum differs by less than 1% from the fission

spectrum value. The possibility of neutron flux transfer which takes advantage of these small cross section

differences is discussed in the next section.

The graphical displays just given indicate the

general features of measured fission detector responses

and their significance. In order to initiate quantitative interpretation of these observed fission rates,

calculated spectrum-averaged fission cross sections

must be available. Obtaining such computed cross sections is not always a straight-forward procedure because of the coarseness of multigroup spectra from

reactor physics computations and because of rapidly

changing cross sections in some energy regions. Nevertheless, a consistent set of calculated cross sections

is essential for detector interpretation if arbitrary

computational biases are not to render careful measurement worthless.

Some Principles of Neutron

Field Characterization

The characterization of neutron fields in reactor

physics criticals, and in and around the great variety

of materials testing, prototype, and power reactors,

depends very largely on passive integral detectors. Of

these, fission detectors are the most essential. Among

passive fission detectors, fission product activation

is the dominant technique assisted in special situations

by fission track recorders. Fission ionization chambers

when applicable are important active detector complements. Neutron flux spectrum or intensity information

may be derived from Integral detectors based on absolute

detection efficiencies and absolute fission cross sections, or upon a calibration carried out by exposure in

a benchmark neutron field. A benchmark for calibration

is a well-characterized neutron field which provides an adequate fluence of neutrons for detector calibration

and which exhibits a spectrum that is relevant and

better known than the spectra under study.

The choice between absolute detection methods and

neutron field calibration is often one of convenience

and custom more than an explicit evaluation of the

relative effort and the errors involved. Absolute

detection methods have taken precedence historically,

a fact attributable in part to the minimal development

and poor availability of well-characterized neutron

fields for purposes of calibration. The latter situation has improved, although the recognition of it is

delayed. For this reason the outline below of neutron

field characterization with integral detectors will

emphasize benchmark field calibration. The recognition

that absolute measurements are also important is not

set aside. Such complements are proper and "keep half

an eye" on the claims of proponents of particular benchmark fields.

Response ratios among a set of integral detectors

exposed to a benchmark field provide first and foremost

a test of detector reaction cross sections over the

energy range of the benchmark spectrum. If observed to predicted detector response ratios agree, field characterization may proceed with confidence and minimized

uncertainties. If they disagree by more than the absolute detection errors and the assigned benchmark spectrum uncertainties, allowed adjustment of the cross sections may be justified. Alternatively, for detectors

with reliable cross sections, allowed adjustments of some benchmark spectra may be undertaken. All in all,

consistency of benchmark detector response should be

established (forced somewhat, if necessary) in order

to achieve unambiguous spectrum characterization in the

neutron field under investigation.

Similarly, if the detection techniques employed in

the benchmark exposure are the same as —or calibrated

relative to—the techniques used in the field characterization exposure, observed and predicted ratios for

the benchmark may be brought into agreement by ad hoc

adjustment of the overall detection efficiency. Establishing detection efficiencies in this way removes a number of systematic errors associated with the detection scheme. Examples are errors of absolute cross section scales, activation counter calibrations, and

nuclear parameters including branching ratios and fission yields. This error correlation in turn allows for

a wider choice of integral detector types and of activation detection methods.

Beyond the matter of adjusting reaction cross sections, benchmark field spectra, or detection efficiencies—an issue for which agreement on systematics

does not yet exist among experimenters—the benchmark

exposure provides a basis for setting the accuracy and

confidence for the entire spectrum characterization

procedure. Often enough, it will point to inadequacies

in cross sections or weaknesses in the experimental detection scheme.

Brief Formulations

Some analytic expressions for observed and derived

integral detector responses are needed to move from the

general assertions just given to specific applications.

As noted, activation detectors are most typical for

spectrum investigations, and therefore the formulations

below are in terminology applicable for them. Modifications required for other types of integral detectors

involve for the most part changes in time integrating

quantities, e.g., f lux-fluence , decay constants, etc.,

and do not affect the principles of integral detector

calibration.

Spectrum and Cross Section Definitions

(nv) ; (nvt) = total energy- integrated flux and

fluence, respectively.

(^(E) = neutron spectrum normalized to unity.

*K>E^) = fraction of spectrum above neutron

energy E^.

:

a(E) = detector reaction rate cross section

vs. energy.

o(E) = • s(E), where Oq is the absolute

cross section scale factor, and s(E)

the cross section shape normalized

to unity over a relevant benchmark

spectrum, ^, (E) :

159
a(>E^)

[0(E)i|;(E)]

o

s(E)i|;^(E)dE

a = spectrum-averaged cross section:

oo a = J

s(E)iJj(E)dE

spectrum-averaged cross section

truncated at E^:

00 00 0(>Ej.) = 0^ J

s(E)<Jj(E)dE/y lp(E)dE (1)

detector response function

= G(A,t)

P 1|J(>E^)

• a (>E ) p

[iij(>E^)(nvt)^] (5)

This formulation has the advantage that it contains

the cross section a (>Ep) truncated at an energy Ep

appropriate for the reaction, and the fluence truncated

at another energy appropriate for dosimetry. The factor

can be rewritten as

<J^(>Ep) J F >(E)dE

f iJ^(E)dE

(6)

E = truncation energy for defining a

^ detector energy response range. For

percentile P, the truncation energy is

defined by

00

p where E (P=0.5) = median energy, and

for thii paper E (P=0) = 20 MeV,

E (P=l) =0.4 eV?

P Observed Reaction Rate

R = • U(X, N, Br, Y, I, . . .) • D (3)

R = observed activation detector disintegration

rate in disintegrations per second (dps) at end of irradiation.

D = observed gamma counting rate after neutron

field exposure in counts per second (c/s).

= gamma counting efficiency.

y = composite factor for converting gamma counting rate of a detector to disintegration

rate: decay constant (A), effective number

of detector atoms (N) , branching ratio (Br),

fission yield (Y) , y-ctg losses and activation interference (I).

Derived Reaction Rate

R^ = G(X,t) • N • a • (nvt)^ (4)

R = derived reaction rate (dps) from fluence

(nvt) received in a neutron field with

spectrum 41(E) . G(X,t) = activation decay rate factor. At the end

of an irradiation at constant flux and

duration t, G(X,t) = [1 - exp (-At)]/t.

The departure of this quantity from unity is determined

by the fraction of the spectrum between E and E , the

energy region where the detector does not respond.

Thus, for two different spectra with similar shapes

above E but dissimilar between E and E , the truncated

cross sictions a (>E ) would be aEout eqSal and the

dissimilarities woulS be expressed by the spectrum

component ratio in eq. (6).

Uncertainty in Spectrum Average Cross Section

Spectrum and cross section errors are estimated in

multigroup formats. The spectrum-average cross section

as a discrete summation,

0 = 0 E s. iw. Ae.; a. = 0 s. , (7) o . 1 i 1' 1 01 ' 1

is subject to an error propagation which must account

for the normalization of the neutron spectrum:

Spectral Index Calibrations

The physical datum for spectrum characterization

with integral detectors is a set of measured reaction

rate ratios. Consideration of a single ratio suffices

for an outline of calibration principles. For two

detectors a and 3, the observed disintegration rate

ratio in a neutron field under study, R^/Rg, obtained

from a count rate ratio, D /Dp,, according to eq. (3), can be set equal to the reaction rate ratio according

to eq. (4) in order to obtain an observed cross section

ratio or spectral index:

The neutron fluence greater than a given energy

[(nvt) • ijj(>EQ)] is often used as a dosimetry parameter. For example, <Jj(>Eq = 1 MeV) in U.S. guides for

surveillance of reactor pressure vessel fluence. Since

detector reactions have various minimum response

energies, it is useful to express a derived reaction

rate in terms of a neutron fluence greater than Eq, and

a detector cross section truncated at percentile P. This can be done by expressing 0 in eq. (4) in terms of

P using Eqs. (1 and 2):

R^ = G(A,t) • N

Thus,

p

• O (>Ep) ((;(>Ep) (nvt).

[G • N]^

D

6 [G • N],

(9)

[cf /a„] a 3 s spectrum

observed spectral index for the study

The spectral index expected on the basis

some presumed spectrum for the neutron field under

study, (e.g., measured or computed, "apriori input",

derived from other integral results) , is

160

of
f 3/ ) f Sa(E)>|; (E)dE

(10)

s, exp

"^exp ~ presumed spectrum of field under study,

• s(E) = energy dependent detector cross section.

The double ratio of observed to expected spectral

indexes (eq. 9 and 10) is a measure of the adequacy of to describe the study spectrum. The departure

froS unity of the double ratio provides a basis for

adjusting 4'exp energy groups corresponding to the

response ranges of the two detectors. With or without

the help of unfolding codes (useful when more than a few detectors are employed) , the uncertainty in the

spectrum adjustment will include errors from all of the

factors in eqs. (9) and (10):

6d, &e^, &\i(y, N, Br, Y, I ...), 6N, 6G, 6a^, 6^ .

The propagation of these errors, further emphasized in

the common case of detector pairs with large over- lapping response ranges, can easily reduce a spectrum

adjustment to one of qualitative significance.

Calibration of a spectral index in a benchmark

neutron field can change the uncertainty dramatically.

The similarity of the benchmark and study spectrum is

not of primary importance, a substantial overlap of

detector response ranges in the two spectra is sufficient to achieve most of the accuracy improvement. An

observed spectral index obtained in a benchmark field

irradiation may be set in ratio to the observed spectral

index from the study field. Using eq. (9), the double

ratio dosimetry to benchmark is,

[a /a„] [D /D„]

The spectral index can be specified for the benchmark

from its known spectrum, Ui, , and the expression,

b

CO

[Da/D«]

0 s Wa /-'o

a. / tV^3^b

{%l I

Se(E)^^(E)dE

(11)

This is the observed spectral index for the study field

calibrated by means of a benchmark field exposure.

When this observed index is compared to the derived spectral index for the study spectrum, the un- certainty will involve only the gamma counting rates

(c/s), the benchmark spectrum, and partially, the

shapes of the detector cross sections:

6D, 6s(E) partially, 6iJj^(E) .

For all situations, the detector calibration by exposure

to neutrons in a benchmark field removes the errors due

to detector efficiency 6Ey, the dps conversion factor

6y, and absolute cross sections scale Sa^. If the

detector response functions for the study and benchmark

spectra, [s(E)i(jg (E) ] and [s (E)i)j|^ (E) ] are poorly matched,

the full uncertainty of s(E) remains in the error propagation. Fully consistent cross section shapes

from interlaboratory measurements among benchmarks are required to reduce the latter error. This is generally

seen as a two-step process, with a restricted set of

well-known detectors (category I reactions) made con- sistent by fine adjustment of evaluated cross sections

followed by more serious adjustment based on integral

measurement results of less well understood dosimetry

detectors (category II reactions)

.

Neutron Flux Transfer

When the total neutron flux and fluence, (nv) and

(nvt) , can be specified for the benchmark irradiation

of an integral detector, it is sometimes possible to perform a direct neutron flux transfer to the neutron

field under study. This is most successful when the

integral detector cross section is largely energy independent over the study spectrum energy response range,

or when the detector response function [s(E)(J;(E) ] , for

the benchmark and study fields are well matched. An

example of the first circumstance is the ^^'Pu(n,f) detector applied to fast reactor spectra, an example of

the second is ^^^U(n,f) applied to fields where the

fission spectrum dominates the energy distribution

above 1.5 MeV.

Observed reaction rates, eq. (3), obtained with

experimental techniques matched in the benchmark and

study field are set equal to the derived reaction rate,

eq. (4), involving the computed average cross section

and the neutron fluence in the two fields. Using the

notation of eqs. (3) and (4),

study field: e -y-D =G«N*a • (nvt) •' Y s s OS

benchmark field: CY • y • D, = G • N • a, • (nvt) , b b Ob

Dividing, the study field fluence (nvt)^^ is obtained

in terms of the benchmark field fluence.

(nvt) = • — • (nvt) OS D, - Ob

b a

s

(12)

and the only experimental quantities involved are the

gamma counting rates.

To examine the influence of the cross section ratio

on the fluence transfer, the alternative expression for

the derived reaction rate given in eqs. (5) and (6) is

appropriate. The flux transfer then in terms of cross sections truncated at the lower limit of their energy

response range,

given energy E , becomes

0(>Ep), and neutron fluences above a

[*(>E^)

/ l(j^(E)dE^

/ o

^^(>E^)dE

(>E )dE

s o

a (>E

s p

- • h(.>-E) • (nvt) 1 (13)

) L ° °Jb

where

[<J;(>E^) (nvt)^] neutron fluence greater than E

i);(>E ) = fraction of normalized spectrum

° above E , o a(>Ep) computed spectrum-averaged

cross section truncated at percentile energy E near the

lower limit of the detector

response range - see Eq. (1).

161
The integrals in the brackets of eq. (13) give the

spectrum fraction between Eq and Ep, an energy region

where the detector does not respond. This spectrum

fraction for the benchmark field will be known; for the

dosimetry study spectrum, however, this fraction contributes an irreducible uncertainty which is common to any

fluence measurement technique employing integral

detectors. In multiple foil techniques, using unfolding

codes for example, the corresponding problem is referred

to as the energy range of poor detector coverage.

The cross section ratio in eq. (13) for appropriate

detectors will be near unity. The absolute cross section scale cancels as with spectral index calibrations, and the remaining uncertainty in flux transfer

due to cross section shape errors, 6s(E), propagates more nearly on the departure from unity of the cross section ratio rather than on the ratio itself. Thus,

a ± 10% cross section shape error would affect a flux

transfer involving a cross section ratio of 1.1 by

about ± 1%. An exception to this occurs when the

detector response ranges for the benchmark and dosimetry

fields are very different. In this case spectrum

uncertainties, presumably dominated by the study field,

will propagate into the flux transfer according to the

last term of eq. (8). This term, as applied to Og(>Ep)

in eq. (13), will not involve a sum over energies below

E where 0. =0.

P 1 Fission Rate Measurement Standards

Fissionable Deposits

It is not difficult to achieve electronic pulse

registration of a fission fragment as it emerges from a thin deposit of fissionable material with an absolute

detection efficiency of better than 99.8%. Devices are simple to construct: ionization chambers and surface

barrier detectors are examples. Thus, accuracy of

fission rate measurements is primarily concerned with

fissionable deposit mass determinations and the absorption of fission fragments in the deposit. The latter

problem can be made small in many cases by obtaining

effective masses of thicker working deposits (e.g.,

0.4 to ~ 1 mg/cm^) relative to thin deposits (e.g., < 0.05 mg/cm2) by means of low-geometry alpha counting

or back-to-back fission counting. Well-studied and

permanently maintained fissionable deposits then become

the basic artifact standard for fission rate measurements.

A set of permanently maintained reference fissionable deposits generally available for interlaboratory

comparisons has been established at NBS.^ Characteristics of the deposits including interim errors for

principle isotopic masses are given in Table I. The

TABLE I. REFERENCE FISSIONABLE DEPOSITS AT NBS

PRINCIPLE ISOTOPE; "'Pu 235„ 238„

Np NATURAL DEPLETED

DEPOSIT THICKNESS (MG/ai^) 0.083 0.20 0.18 0.20 0.093

ISOTOPIC PURITY (AT %) 99.11 99. 75 99.275 99.899 100 (NOMINAL) INTERIM ERROR ASSIGNMENT FOR PRINCIPLE ISOTOPE MASS (la) il.2% ±1.2% ±1.5% ±1.5% ±1.7%

STANDARD DEVIATION OF INTER- LABORATORY COMPARISONS

^EUROPE (1974-76) ±1.02 ±0.8% ±0.8%

°U.S. (1972-75) ±0.6?. ±0.7%

ALL DEPOSITS ARE 1.27 CM DIAMETER OXIDE DEPOSITS ON POLISHED PLATINUM DISKS (19.0 MM DIA. X 0.13 MM THK)

^NORMALIZED FISSION RATE MEASUREMENTS AT THE SIGMA SIGMA FACILITY (MOL. BELGIUM) INVOLVING

NBS (U.S.), GFK (GERMANY) AND RON (NETHERLANDS) EACH USING THEIR OWN FISSION IONIZATION

CHAMBERS TO RECORD FISSION RATES.

^INFORMAL FISSIONABLE DEPOSIT COMPARISONS INVOLVING LASL, ANL, OBNL, AND BCNM (GEEL, BELGIUM).

THE FIRST TWO LABORATORIES PROVIDED RESULTS FOR BOTH 239pu AND 235u.

imposed requirement for a systematic program of redundant mass assay, not yet complete, results in relatively

high error assignments for these reference deposits.

Some interlaboratory comparisons with NBS deposits have

been performed. The last two lines of Table I indicate

the standard deviations associated with some of these

efforts: (1) a formal fission rate intercomparison

involving two laboratories besides NBS; and (2) an informal grouping of results from four laboratories that

have participated in joint measurements with NBS. The

interlaboratory results suggest that the interim mass

errors for the NBS reference deposits are conservative.

Natural Neutron Fields

Turning from artifact standards to well-characterized neutron fields as a reference for fission rate

measurements, a natural focus is the upper and lower

bounds of fission-driven reactor systems, namely fission

neutron spectra and thermal neutron distributions. The

latter are commonly employed for precise intercompar isons

of fissile deposits or of fission rate detectors con- taining fissile materials. In addition, very useful

cross checks among isotope species may be carried out

at thermal based on well-studied nuclear parameters.

1. Mass ratios among fissile isotopes, e.g. ^^^Pu:

235y. 233y^ based on thermal fission cross sections.

For high confidence levels, the comparison should be

carried out also with monoenergetic neutrons near

0.025 eV in order to assure that the required non-l/v

correction (e.g., Westcott g-factor) is correctly

known for the thermal Maxwellian employed.

2. Mass ratio between ^'^U and ^^®U based on abundance

of ^^^U in natural uranium. A certified natural uranium

base material should be used (e. g .

,

' NBS-SRM 950a,

Colorado Plateau ore, 0.719 + 0.0007 at.% ^^^U) because

the nominal natural abundance may differ slightly

depending upon origin of the uranium ore.

High thermal fission cross sections for fissile isotopes

along with the general availability of intense thermal

flux intensities also makes these fields convenient for

many interlaboratory comparisons and detector performance investigations, and for the mass assay of very

light fissionable deposits, in the nanogram range for

example.

Fission spectrum neutrons are the complement of

thermal neutrons in the cycle of nuclear energy generation. Fission rate measurements in fission spectra are important for validating and normalizing differential

fission cross sections, for measuring nuclear parameters

needed for certain types of absolute fission rate

detectors, and for direct neutron calibration of

detectors used for reactor spectrum characterizations.

The latter includes the so-called flux transfer technique outlined in the previous sections.

Fission-spectrum-averaged, fission cross section

ratios have been particularly important for the issue

of standardization and were the subject of the

integral measurement paper given at the last Neutron

Standards Symposium in 1970.^ Measurement results

obtained since 1970 and a review of the situation

regarding observation vs. prediction are summarized in

Table II.

Considerable progress in measurement has occurred

with the advent of clean ^"cf fission spectrum fluxes.

Also, there are new

^

' fission spectrum measurements

which employ large cavities and a traverse technique

for establishing the contribution of wall-return^

neutrons^^. Measured ratios among U, U, Np

(and for ^ ^ ^Pu as well but not reported here) are accurate now to about + 2% {la), and the fission rate

162
TABLE II TABLE III

FISSION CROSS SECTION RATIOS FOR ( 235y AND -52cf FISSiaN SEJTRONS

Oj(-^*U)/o^(-"u) Oj("\)/Cj(Np)

^23S„, ,252^„

Yl Ul Yf Ctl

Observed

0.254 t 2.2% 0.266 t 1.6V 0.270 t 4.6t

0.260 t 6%

0.265 t 4. 8%

0.230 ± 3.0* 0.240 ± 2.2*

0.238 ± JJ5*

Fission chbrs.

197S (Refs. 10.11) 1972 (Ref. 3) Activation

1968 (Ref. I) Track recorders

1967 (Ref. 1)

Total spread ±3.2* ±1.8*

Obs. /Predicted

1.065 ± 2.5* 1.047 ± 2.0*

1.13

1.20

1.026 ± 3.0* 1.031 ± 2.2%

1.11

•1976

1972 1967-68

•Predicted values are obtained by convolution of ENDF/B-IV cross sections and evaluated fission spectra from Ref. 9. Spectrum uncertainties are propagated according

to Eq. 8.

Cf FISSION-SPECTRUM - AVERAGED FISSION

CROSS SECTIONS

Isotope Observed *Obs./Pred.

U235 1205 ± 2.2% 0.971 ± 2.2%

U238 321 ± 2.4% 1.018 ± 2.6%

Pu239 1808 ± 2.2% 1.010 + 2.2%

Np237 1332 ± 2.6% 0.986 ± 2.7%

*Predicted values are obtained by convolution of

ENDF/B-IV cross sections and the evaluated 252cf

fission spectrum from Ref. 9. Spectrum uncertainties

are propagated according to Eq. 8.

component of the measurement has been subject to interlaboratory verification — see Table I. The observedto-predicted ratios in Table II carry uncertainties for

the first time based on an evaluation of fission neutron

spectrum measurements that includes uncertainty assian13 ments. The spectrum uncertainty, adds little to the

total uncertainty. The departures from unity for 1976

values vary between 1.026 to 1.065 with uncertainties

between + 2.0% to + 3.0%. These discrepancies with

the ENDF/B-IV evaluation are statistically significant

but still in much better agreement than has been the case in previous years when they were the cause of much

concern and speculation. The relative consistency of

the integral measurement since 1967 indicates that the

changes have occurred primarily in the differential

microscopic cross section data.

Should integral ratio measurements be accepted

uncritically without the associated absolute determinations as a systematic validation? Ratio data has been

a feature of integral reaction rate measurements almost

without exception; and, as has been noted on occasion,

it is a feature much less common for differential

microscopic data. Attention must be called, therefore,

to the absolute fission-spectrum-averaged, fission

cross sections that now exist. Published results for

^^^Cf spontaneous fission neutrons provide integral

microscopic data, namely fission cross sections, which

are independent of any other nuclear cross section and

to first order are independent of any other nuclear

parameter as well.'" Results are given in Table III. The comparison with ENDF/B-IV predicted values include

a propagation of fission spectrum uncertainties so that

departures from unity are a direct check of the differential-microscopic fission cross sections chosen for

ENDF/B-IV. The agreement is generally good.

Critical Metal Spheres

Closely related to fission spectra are the

unreflected critical spheres of ^ U and ^^'Pu metal.

Lady Godiva (^^^U) and Jezebel (^^'Pu), constructed more

than two decades ago, sustain a strong fission spectrum

component, some 60% for Lady Godiva and 80% for Jezebel.

The remainder of the spectrum is largely determined by

inelastic scattering. Fission cross section ratios for

these systems provide vital checks of inelastic parameters presently on the evaluated files. They are also

an example of integral macroscopic measurements nearly

equivalent to their differential microscopic counterpart

since inelastically scattered and fission neutrons are indistinguishable components in both types of experiments .

Fission cross section ratios for the ^

^ and ^^'Pu

critical metal spheres are presented in Table IV. Ratio

measurement results listed in the upper section of the

Table were obtained separately with fission chambers

and with activation detectors, both calibrated by means

of exposures to monoenergetic neutrons with energies

close to 2.5 MeV.'"* Uncertainties for the 1967 measurements on Godiva and Jezebel were +2.5% (fiss. chbr.)

and + 3% (activation) exclusive of fission cross section

errors at the calibration energies. This is an early

example neutron field calibration of fission rates

undertaken in order to reduce errors. The errors given

for average values, +2.5% for all four ratios, do not

include fission cross section uncertainties at the

calibration energy. The actual values in the Table have

been adjusted to be consistent ENDF/B-IV cross section

ratios at the calibration energies. The 1976 absolute

values for BIGTEN were obtained with high resolution

double fission chambers and include, as do the Godiva

and Jezebel results, small corrections for cavity flux

pertubations . Observed-to-ENDF/B-IV predicted ratios

are relatively good for the uranium systems. Lady Godiva

and BIGTEN. The average deviation from unity for the

four obs/pred. ratios, 1.003, 1.043 (Godiva), and

0.976, 1.031 (BIGTEN),

TABLE IV. FISSION CROSS SECTION RATIOS FOR CRITICAL METAL SPHERES

Of("*U)/Oj aj("%/0 flNp)

235„

(GODIVA) Pu

(JEZEBEL) I0Z235„_,„238„

(BIG TEN)* (GODIVA) "'Pu

CJE2EBEL) (BIG TEN)*

MONOENERGETIC

CALIBRATION (1967)«* ABSOLUTE

(1976) MONOENERGETICCALIBRATION (1967)«* ABSOLUTE

(1976)

FISSION CHAMBERS 0.166^ 0.215j 0.0372

1.0006 0.196 0.2190 0.1172

±.0026

ACTIVATION 0.164 0.218j 0. 199 0.229

AV. 0.165

tl.5% 0.217

±2.5!: 0.0372

!l. 72

0. 198 ±2.52 0.224

±2.52 0.1172

£2.22

OBSERVED

PRED. (ENDF/

B-IV) 1.003 1.12 0.976 1.043 1.084 1.031

FISSION SPECTRUM CALIBRATION

FISSION CHAMBERS

(1976) 0.147+1.42 0.510

11.22

ACTIVATION

(1967) 0.617

±3.2% 0.803

±3.22 0.816

12.02 0.907

±1.52

OBSERVED

PRED. (ENDF/

B-IV) 0.887 1.03j 0.918 0.964 0.993 1.007

*A LARGE CYLINDRICAL CRITICAL ASSEMBLY WITH A HOMOGENEOUS METAL CENTRAL REGION. THE CENTRAL SPECTRUM IS LITTLE EFFECTED BY A DISTANT REFLECTOR AND CYLINDER-SHAPED BOUNDARIES.

••DETECTORS CALIBRATED WITH HONENERGETIC NEUTRONS: 2.43 MeV (FISS. CHBR.), 2.75 MeV

(ACTIVATION). (14) VALUES REPORTED IN 1967 ADJUSTED TO CORRESPOND TO ENDF/B-IV CROSS SECTION RATIOS AT THESE ENERGIES. 163
is 2.5% with a standard deviation of + 1.7. This

compares with an observed-to-predicted average deviation

from unity of 6.0% for Lady Godiva reported in 1970 when

calculations were based on the earliest version of

ENDF/B. ^ The departures from unity for the ^^^Pu

critical sphere are large, 1.12 and 1.084, far beyond

the uncertainties of the measurements. This is indicative of inadequate ^'^Pu inelastic transfer cross sections in the calculations if the detector fission

cross sections and the ^^^Pu fission spectrum are correctly known.

A more proper neutron field for calibration of cross section ratio measurements intended to check

neutron transport calculations is the fission spectrum.

It is the strong fission spectrum component in the

critical metal spheres that computation transforms into

the respective spectra of these systems. A fission

spectrum calibration, therefore, circumvents fission

cross section scale errors and fission spectrum

uncertainties.

The lower section of Table IV shows results of

ratios based on experimental calibration with fission

neutrons. For Godiva and Jezebel, this was carried out

by the activation method alone and the ^'^U and ^'^Pu

fission spectra experiments were performed in difficult

geometry.^"* However, the same results, derived in an alternative manner by taking the values based on monoenergetic calibration (upper section of Table IV) and

dividing by the 1975 ^^^U fission spectrum ratios

(Table II), do not differ by more than 4%. The BIGTEN

measurements carried out in 1976 were performed with

fissionable deposits related by alpha and fission

2 3 5 comparison counting to those used in the 1975 U cavity fission spectrum measurements.^ The resulting

spectral indexes for BIGTEN relative to the fission

spectrum driving source (0.147 + 1.4% and 0.510 + 1.2%) are particularly accurate checks of neutron energy

transfer for reactor physics.

The observed-to-predicted ratios now show a very

different pattern from those in the upper section of

Table IV. The ^^^Pu system now is in agreement with

calculation and the uranium systems are not. The

obs/pred. ratios all are uniformly lower, a result

consistent with but not fully accounted for by the

latest obs . /predicted ratios for fission spectra - see 1976 values in Table II.

It is not the purpose of this paper to evaluate

the implications of the results displayed in Table IV. It is a purpose however to illustrate how reference

neutron field calibrations of experiments designed to check reactor spectrum computations may reveal features

not otherwise apparent. The critical metal spheres are particularly apt for this purpose because they involve

little or no complexities of geometry. This

ambiguity for many reactor physics criticals can, and

probably has, obscured the desirability of reference

neutron field calibrations.

References

1. Grundl, J. A. "Fission-Neutron Spectra: Macroscopic

and Integral Results," Proc. Symp. on Neutron

Standards and Flux Normalization, P. 417, Conf701002, ANL (August, 1971).

2.

3.

4. Pinter, M. , Scholtyssek, W. , Fehsenfeld, P.,

Van der Camp, H. A. J., Quaadvliet, W. H. J.,

Fabry, A., DeLeeuw, G. and S., Cops, F. , Grundl, J. Gilliam, D. , and Eisenhauer, C, "Inter laboratory

Comparison of Absolute Fission Rate and Uranium-238

Capture Rate Measurements in the M0L-2Z Secondary

Intermediate-Energy Standard Neutron Field," Conf. on Nucl. Cross Sections and Technology, Washington,

D.C. (March 1975).

5. McElroy, W. N. and Kellogg, L. S., "Fuels and

Materials Fast-Reactor Dosimetry Data Development

and Testing," Nuclear Technology 25, 180-223 (1975)

6. McElroy, W. N. , ed. ILRR Progress Reports, Nos.

9-11, HEDL-TME 75-130 (1973-1977).

7. Vlasov, M. , "IAEA Program on Benchmark Neutron

Fields Applications for Reactor Dosimetry,"

INDC(SEC)-54/L+DOS. (July, 1976).

8. Grundl, J. A., Gilliam, D. M. , Dudey, N. D. and

Popek, R. J., "Measurement of Absolute Fission

Rates," Nuclear Technology 25, 237-257 (1975).

9. Grundl, J., Eisenhauer, C. , "Benchmark Neutron

Fields for Reactor Dosimetry," Consultants Meeting

on Integral Cross Section Measurements in Standard

Neutron Fields for Reactor Dosimetry, IAEA (Nov.

1976). Manuscript copies available from authors

at National Bureau of Standards.

10. Heaton II, H. T. , Grundl, J. A., Spiegel Jr., V. Gilliam, D. M. , and Eisenhauer, C, "Absolute ^

Fission Cross Section for ^^^Cf Spontaneous Fission

Neutrons," Conf. on Nucl. Cross Sections and

Technology, Washington, D.C. (1975).

11. Gilliam, D. M., Eisenhauer, C, Heaton II, H. T.,

and Grundl, J. A., "Fission Cross SEction Ratios in

the "^Cf Neutron Spectrum ("^U: ^^^U: ^^^Pu:

^"Np)," Conf. on Nucl. Cross Sections and Technology, Washington, D.C. (1975).

12. Fabry, A., Grundl, J., and Eisenhauer, C, "Fundamental Integral Cross Section Ratio Measurements

in the Thermal-Neutron-Induced Uranium-235 Fission

Neutron Spectrum," Conf. on Nucl. Cross Sections

and Technology, Washington, D.C. (March, 1975).

13. Grundl, J. A., and Eisenhauer, C. M. , "Fission

Spectrum Neutrons for Cross Section Validation and

Neutron Flux Transfer," Conf. on Nucl. Cross

Sections and Technology, Washington, D.C. (March

1975)

.

14. Grundl, J. A., and Hansen, G. E. , "Measurement of

Average Cross Section Ratios in Fundamental FastNeutron Spectra," in Nuclear Data for Reactors,

Vol. I, pp. 321-336, IAEA, Vienna (1967). Also

see Nucl. Sci. and Eng., 8^, 598-607 (1960).

15. Behrens, J. W. , Carlson, G. W. , and Bauer, R. W.

,

"Neutron-Induced Fission Cross Sections of ^^^U,

"•u, ^^^U and ^^®U with Respect to ^^^U," Conf. on Nucl. Cross Sections and Technology, Washington,

D.C. (1975). See also M, S. Coates, D. B. Gayther

and N. J. Pattenden, "A Measurement of the ^'®U/

^^^U Fission Cross Section Ratio," same conference.

Taschek, R. F. , ibid. 1. 16. Private communication from ENDF/B-IV cross section

Grundl, J. A. "Brief Review of Integral Measure- evaluators (1975).

ments with Fission Spectrum Neutrons," and "Measurement of Av. Fiss. Cross-Section Ratio, ^'^U/^^°U

For ^^^U and ^'^Pu Fission Neutrons;" Prompt Fission

Neutron Spectra, Panel Proceedings Series, IAEA

(1972)

.

164
UTILITY AND USE OF NEUTRON CAPTURE CROSS SECTION STANDARDS

AND THE STATUS OF THE Au(n, v) STANDARD

A. Paulsen

Central Bureau for Nuclear Measurements

B-2440 Geel, Belgium

The main application of a neutron capture cross section standard should be found in ratio measurements using the prompt y-ray detection method. A review of neutron capture cross section

measurements in the last six years shows that the Au(n, y) standard is increasingly used for this

purpose. But the majority of all measurements is still based on other normalization methods

than ratio measurements, although the accuracy established for the Au(n,Y) cross section below 3 MeV competes now well with that of other normalization methods. On the other hand this

accuracy has scarcely reached the accuracy of necessary corrections associated with prompt

y-ray detection measurements below 3 MeV neutron energy and is completely insufficient above

3 MeV. Below 200 keV the cross section fluctuations due to level statistics in the compound system ^98^|j g^j.g seriously disturbing measurements aiming at high accuracy and high resolution.

(Review ; neutron capture cross section; measurement; normalization; reference standard; accuracy; Au(n,Y) cross section; cross section fluctuations)

Introduction The Measurement of Cross Section Ratios

A standard cross section for neutron capture measurements is of quite different importance in the three

neutron energy regions : -thermal and resonance region

-unresolved resonance region

-high energy region.

Thermal and resonance region

In the thermal and resonance region many thermal

cross sections and resonance parameters (e. g. for

55 59^ 103„, 107,109, 113,115, 197, Mn, Co, Rh, Ag, In, Au) have been determined to a rather high degree of accuracy! . These data have been often used to normalize neutron capture cross section measurements in the thermal or resonance region. Therefore, with good rea- sons these thermal capture cross sections and reso- nance parameters could be named "capture standards".

Something similar is valid for certain capture reso- nance integrals (e. g. for Mn and A u) which have been

used as reference standards for the measurement of resonance integrals. The relative large amount of

existing data is responsible for today's relative small

interest in thermal and resonance capture standards.

Unresolved resonance region

A standard cross section in its unresolved reso- nance region is in principle of no use unless the measurement averages over a sufficiently broad energy

interval to arrive at a smooth cross section shape.

This implies that one capture reaction chosen as a re- ference standard can never be suited for the whole

neutron energy range.

High energy region

Due to the required smoothness of the cross section curve a capture standard cross section is of sufficient versatility in application only in the high energy

region. Therefore the main region of interest in the

Au(n,y) reaction, which is up to now the only one re- cognized as capture standard 2, can be only in the

energy region above about 100 keV.

In the thin sample approximation the reaction cross

section ct can be calculated from the number of counts

C of a capture detector with efficiency e, the number

of sample atoms N and the neutron fluence density $

according to Fig. 1. In a similar way the neutron flu- ence density can be deduced from a known standard cross section Oo •

C : number of counts N : number of atoms

£ : detection efficiency 't> -. neutron fluence density

<(,=

e N -4)

1

Co index 'o' corresponds

Eo ' No • Oo ^° standard reaction

\

C • Eq • No standard used to Co E • N ° measure neutron fluence

Ob

if E = Eo

CNo

Co - N

standard used to perform

ratio measurement

(e is unknownJ<(> can not

be evaluated)

Fig. 1 : Explanation of the term "ratio measurement'

165
If the standard reaction requires another detector

than the capture reaction to be measured, then the

efficiencies will not cancel from the equation for a . In this situation the measurement of the neutron fluence

density could also be done with any other suitable stan- dard which has not at all to be a capture cross section.

In this case all detector efficiencies have to be known

and the neutron fluence density can be calculated.

However, if the standard reaction can be detected

in completely the same way than the capture reaction

under study, then due to e= e„ the standard is used for

a ratio measurement for which no efficiency has to be

known and $ can not be evaluated. A neutron capture

standard is therefore characterized by the fact that it permits capture cross section determinations by ratio

measurements.

The Prompt Gamma -Ray Detection Method

Neglecting here measurement techniques based on neutron absorption (shell transmission and pile reactivity measurements) only the detection of the prompt

capture y-rays fulfils approximately the condition for

ratio measurements, whereas the activation method is

faced with characteristic radiations changing completely from radioisotope to radioisotope. For the prompt

y-ray detection method the condition e= is only approximately fulfilled because the capture v-ray spectra

(and possibly also the angular distributions) are changing from isotope to isotope. This has to be corrected

for.

It is a principal question whether the prompt y-ray

detection method measures the full capture cross section also at higher neutron energies. To discriminate

against y-rays from inelastic scattering one is obliged

to use only the upper part (Ej^< Ey) of the capture

spectra. Therefore cascade passages through unbound

states could be missed. For this reason it is interesting to compare prompt y-ray detection results CJganTrna

with activation results Oactiv. • This is done for Au(n,Y)

in the 1 to 3 MeV energy range-^ • ^ ' ^ in Fig. 2 and for

several reactions at 14 MeV in Fig. 3. For the latter

only activation results were taken into account which

passed a critical examination in view of corrections

for scattered neutrons of degraded energy. Within today's experimental accuracy there is no indication

that capture y-ray cascades are passing partially

through unbound states.

SOSOS 70- E

,5 60- o

50

^0

30-

20-

10-

'3^Au(n.Y)TMAu

• Ponitz^ A Lindner et

^

o Paulsen el al^

1.0 2.0 3.0 En (MeV)

Oacti\

^amma

2.0-

1.0-

En = K MeV

• Wagner and Warhanek^ o Wagner and Warhanek^

and Budnar et al.'

197,

181

23 Na

31p

51

133,Cs

127|(i

"At

20

I 50 100 —I

150

1 200

Fig. 3 : Ratio of activation results Oactiv prompt

y-ray detection results CTgamma I 4 MeV.

Actually used Normalization Methods of Neutron

Capture Measurements

A review was made concerning the used standards

and normalization methods for capture cross section

measurements in the neutron energy range above 20

keV by the prompt v-ray detection technique. This re- view covers all measurements for which numerical results have been compiled at one of the four cooperating

Neutron Data Centers between 1971 and 1976. The result is shown in Table 1 where the measurements are

counted per studied isotope or element (Au excluded).

The number of ratio measurements relative to gold (at

least at one neutron energy) is increasing from the

71/73 to the 74/76 period, indicating that the gold standard is now indeed increasingly used. All the ratio

measurements relative to gold were performed at electrostatic accelerators, whereas at linear accelerators

a normalization in the eV range by saturated resonances is preferred. However, an additional normalization

at the high energy end (Ej^ > 100 keV) by a gold ratio

measurement could probably improve the accuracy of

the results produced at linear accelerators. It has to be kept in mind that these ratio measurements can be

renormalized at any time.

Fig. 2 ; Comparison of prompt y- detection (full symbols)

and activation (open symbols) results for Au(n,Y).

166
Table 1: Normalization Methods of Neutron Capture Measurements*

Separ

Satur Therm.

Ratio Measurements

Period Determ.

Reson. Cross Sect. Total

of

c and 4> •rel •rel Au Ir^ Ag

1971-1973 30 15 U 10 10 3 72

197A-1976 7 23 7 23 60

• per isotope or element prompt Y-ray detection method only

for neutron energies above 20 keV only

Comparison of Normalization Methods

Most probably the preference given to other normalization methods than the gold ratio measurement

is due to the fact that experimentalists believe to assess better accuracies that way. Therefore a comparison

is made between the estimated accuracy of the Au(n,Y) cross section and that one of other normalization methods used for measurements above 20 keV neutron

energy. This comparison is shown in Table 2. The un- certainties are estimated at the confidence level of one standard deviation and contain the principal uncertainty of the neutron fluence and the y^ay detector efficiency. The total error was obtained from the individual uncertainty components by quadratic summation.

The table contains further the upper energy limit for

the validity of the quoted error and the reference to the corresponding publication.

Table 2: Uncertainties of Normalization Methods

Method

Eob! or le 1 >QbS <^obs or ijlrei Capture Standard

Total

Uncertainty Method Acc.Ref Energy

Limit Method Acc Ret Energy

Limit

Cross

Section Acc Ret Energy

Limit Separ

Determ.

of

e and

Radiooc

Sources

lp,Y)Re- actions

2%(27)

miia

IMeV

l5MeV

Asspart

Long C

H (n,nl

"\j(n,t)

2% 19) 27. IW) 37. 19) A7. 112)

15 MeV

1 MeV

15 MeV

6 MeV 3-m

Determ

of U*)

Satur

Reso- nances

1-2%

(11,221 100 eV

'Li ln,t)

"b (n,al

47. 113)

37.03)

01 MeV

01 MeV

3 - 57.

Normal it

at

Therm. Energy

"b (n,ci)

. H(n,nl

°^J(n,f)

'47.(131

37. (K)

1 MeV

6 MeV

Therm

Cross

Section

1-27. ID D025tV 3- 57.

Ratio Measurem

with Gold

Au (n,Y) Cross

Section

4% 1291 5% (18)

02-35

MeV 01-3

MeV

A - 5%

The comparison of these uncertainties with today's

accuracy of the gold capture cross section (last line of

Table 2) is not at all unfavourable for the latter. But it is also not yet convincing to give preference to ratio

measurements with gold. However, the accuracy of

the gold capture cross section could finally surpass

that one of the competing normalization methods, be- cause many measurements and all experimental measuring techniques for neutron capture can principally

contribute to it.

Up to now measurements carried out with high effort and using a normalization method according to the

first two lines of Table 2 can be still superior to gold

ratio measurements with respect to resulting accuracy. Here the influence of the simultaneously progressing standards 6Li(n,t), 10B(n,a) and 235u(n,f) has to

be mentioned. Of course, this comparison is only valid

for measurements in or ranging up to the indicated

energy range for the Au(n, y) cross section. Comparable

measurements belonging to the third line of Table 2 are nearly not existing : there is no experimental technique to cover neutron energies from thermal energy

up to several hundreds of keV (slowing-down time measurements just reach 100 keV). The relative simplicity

of ratio measurements and the readjustibility of their

results justify further efforts for an amelioration of

the accuracy of the gold capture cross section.

Uncertainties of Ratio Measurements

There are certain sources of error inherent to all capture cross section measurements performed by the

prompt y-ray detection technique. It is interesting to compare the accuracy of the gold capture cross section with these 'inherent' uncertainties. This is done in

Table 3 by com.paring uncertainties which have been

quoted in literature as average values.

Table 3 Uncertainties of Ratio Measurements.

Source

of Uncertainty

Pulse Height

Weighting Techn.

Moxon - Roe

Detectors

Liquid Scint. TonksSpectrometers

NallTD.GelLi), Pair

Variation

of Y-ray Spectrum

±1%

Ref 8, 15 ±3%

Ref 19, 20

±57.

Ref 21,22 ±m

Ref. 16, 17

Multiple Scattering t 2-3V. Ref 22.23

Neutron Scott

into Y

- detector for-5!--l0' ±0.5-37. Ref. 24,25 "y

Background

Subtraction ± 1-3% Ref 23,24

Au ln,Y) Cross Section t A -57. R«< 18,29

There is a detector-dependent error due to variations of the capture y-ray spectra. This error is minimized for detectors having the least spectrometric

response and is a maximum for the true spectrometers.

The pulse height weighting technique in conjunction

withC^F^or C^D^ y-detectors results ina remarkable independence from the shape ot tne y-spectrum. It is the

clear predominance of the Compton y- scattering process in these detectors with its smooth and relative

weak energy dependence which is responsible for this

independence as well as for a good reliability of efficiency calculations. This type of detector should be al- so applicable in the MeV neutron energy range because

the spectrum fraction above threshold can be well calculated. For the pulse height weighting technique and

for Moxon-Rae detectors there are difficulties in the

data evaluation if the sample is a compound of elements or a mixture of isotopes with unknown capture cross

sections and strongly differing neutron binding energies. Uncertainties originating from these difficulties

have not been considered in Table 3. For tank measure- ments the spectrum fraction uncertainty is astonishingly high due to the relative high thresholds which have

to be used for background reasons with such detectors.

Cross section measurements performed with spectrometers are mainly done in the high MeV range where

the accuracy of the Au(n , y) cross section is anyway

167
not yet defined. Variations of the y-ray angular distributions with increasing neutron energy seem to be negligibly small^^.

Uncertainties due to corrections for multiple scat- tering, response of v-detector to scattered neutrons

and general background can vary considerably and are depending on several experimental conditions. The ave- rage uncertainties of these corrections quoted in Table

3 are smaller than the normalization and y-detection

uncertainties.

In conclusion Table 3 demonstrates that further

efforts for an improvement of the accuracy of the gold

standard seem to be justified.

Status of Au(n,v) Standard

Below 0. 2 MeV

11 ? 7 Recent mea surements ' ' • ^ ' have confirmed that

considerable cross section fluctuations are extending

10 20 50 100 200

En (keV) >

Fig. 4 : The relative neutron energy resolution neces - sary for the observation of + 3 and + 5 % relative cross section fluctuation for the neutron

capture in Au and Ta (acc. to ref. 28).

up to 200 keV. These fluctuations are the high energy

extension of the unresolved resonance region and are caused by the statistical distribution of the capturing

levels in the ^'^'^Au + n system. The amount of observable fluctuations does not only depend on the neutron

energy but also on the energy resolution. Fig. 4 shows

the relative neutron energy resolution which would

permit the observation of + 3 and + 5 % deviations for

capture in gold and tantalum as a function of neutron

energy. These data have been produced ^8 by Monte

Carlo calculation assuming level distances and reduced level widths according to a Wigner and a PorterThomas distribution, respectively. Fig. 4 demonstrates that due to the higher level density Ta(n,Y) would

be in this respect a much better standard in the 50 to 200 keV energy range. As these fluctuations could

strongly influence high resolution ratio measurements

ENDF/B-v29 will no longer support the gold standard

below 200 keV. Furtheron one can read from Fig. 4

that gold activation measurements with Sb-Be neutron sources at 22. 8 4^ 1 . 3 keV can deviate up to 5 % from

the smooth average cross section curve.

B etween 0. 2 and 3 MeV

In this energy region the accuracy of the Au(n . y) cross section is estimated to be + 5 for the

ENDF/B-IV data or + 4 %29 for the ENDF/B-V data

in preparation.

Above 3 MeV

In the energy region above 3 MeV the available

Au(n,y) cross sections are very scarce. Fig. 5 shows

all available cross section data together with the

endf/b-iv curve and the proposed ENDF/B-V data.

It is in this energy region from 3 to 14 MeV where

new Au(n, y) cross section measurements would be very valuable in view of checking the validity of interpolations by model calculations.

30-

20-

10-

& Pbnitz Lindner et qI

'

V Paulsen et ol' o Miskel et at

X Johnsrud et ol^

32

—

r13 —r15 En (MeV)

Fig. 5 ; Cross section data for the neutron capture

in Au above 3 MeV.

Conclusions

Since the introduction of a capture standard cross

section is aiming at the performance of ratio measurements, more measurements of that type should be

done, especiallv at the high energy limit of linear accelerator facilities. Results of ratio measurements

can be easily renormalized afterwards. But already

168
today the accuracy of the gold standard can well compete with other normalization methods in the 0. 2 to

3 MeV energy range. Also the problem of the influen- ce of cross section fluctuations seems to be clarified.

With respect to these cross section fluctuations the

Ta(n,Y) reaction is a much more suited standard in

the 50 to 200 keV energy range. Above 3. 5 MeV neutron energy the existing cross section data are completely insufficient for any reference purpose. This

reflects of course the data needs for fission reactors.

Further measurements in this energy range could

help to extend considerably the applicability of this

standard for developments in fusion technology and

nuclear safeguards''^.

References

1. S. F. Mughabghab and D. I. Garber , Neutron Cross

Sections, Volume I, Resonance Parameters , Report BNL 325 Third Edition (1973)

2. Proc. Neutron Standard Reference Data, Vienna

1972, Recommendations p. 360

3. W. P. POnitz , Nucl. Sci. Eng. 57 (1975) 300

4. M. Lindner, R. J. Nagle and J. H. Landrum, Nucl.

Sci. Eng. 59 (1976) 381

5. A. Paulsen, R. Widera and H. Liskien, Atomkernenergie 26 (1975) 80

6. M. Wagner and H. Warhanek, to be published

7. M. Budnar, F. Cvelbar, A.Kikar, R.Martincic,

M. Potokar and V. Ivkovic , Measurements of 14

MeV Neutron Radiative Capture y-Ra.y Spectra and

Integrated C ross Sections in Sc , Y, Pr, and Ho,

Report INDC(YUG)-5/L (1977)

8. C. Le Rigoleur , A.Arnaud, Proc. NEACRP/NEANDC

Specialists Meeting Karlsruhe May 1973, Report

KFK 2046/NEANDC-U-98, p. 33

9. A.Paulsen, H. Liskien and M. Cosack , Nucl. Instr. Meth. 105^ (1972) 103

10. J. L. Leroy, J. L. Huet and J. Gentil , Nucl. Instr. Meth. 88 (1970) 1

11. R. L.Macklin, J. Halperin and R. R. Winters , Phys.

Rev. Cn (1975) 1270

12. M.Lindner, R. J. Nagle and J. H. Landrum , Report

UCRL-75838 (1974)

13. H. Liskien, Neutron Standards and their Application, Proc. Intern. Conf. on the Interaction of Neutrons with Nuclei , Lowell 1976, p. 1110

14. V.A.Konshin, Review and Evaluation of the 235u

Fission Cross Section in the 0. 1 keV to 15 MeV

Energy Range, Report INDC(CCP) -94/U (1976)

15. J. B. Czirr and M. L. Stelts, Nucl. Sci. Eng. 52

(1973) 299

16. I. Bergqvist, D. M. Drake and D. K, McDaniels , Nucl.

Phys. A191 (1972) 641

17. M. Stamatelatos , B. Lawergren and L, J. Lidofsky,

Nucl. Sci. Eng. 51 (1973) 1 13

197

18. A. B.Smith, The Contemporary Status of the Au

(n,Y) Cross Section, status report to the 8th INDC

Meeting, Vienna October 1975

19. M. C.Moxon, thesis London University 1968

20. R. L. Macklin, J. H. Gibbons and T. Inada , Nucl. Phys.

43 (1963) 353

21. D. Kompe, Nucl. Phys. A133 (1969) 513

22. R. E. Chrien, Neutron Capture Cross SectionMeasurement T echnique s, Proc. Nucl. Cross Sections

and Technology. Washington D. C. 1 975 , NBS Spec.

Publ. 425, Vol. I (1975) 139

23. R. C. Block and R. W, Hockenbury, The Precision

of Tank Capture Cross Section Measurements,

Proc. Neutron Standards and Flux Normalization

Argonne 1970, AEC Symposium Series 23^ (1971)

355

24. M. C. Moxon, D. B. Gayther and M. G. Sowerby,

NEANDC Topical Conf. on Capture Cross Section

Measurements, Report AERE-R 8082, Harwell

1975, p. 12

25. R. L. Macklin and B. J. Allen, Nucl. Instr. Meth. 9i

(1971) 565

26. D. M. Drake , E. D. A rthur and I. Halpern , Proc.

Sec. Intern. Symp. on Neutron Capture Gamma-Ray

Spectroscopy, Petten 1974, p. 227

27. C. Le Rigoleur, A. Arnaud and J. Taste , Report

CEA-N-1662 (1973)

28. H. Liskien and H. Weigmann, Annals Nucl. Energy

in press

29. S. F. Mughabghab, Informal Report BNL -NCS -2 1 774

(1976)

30. J. A. Miskel, K. V. Marsh, M. Lindner and R. J. Nagle, Phys. Rev. j_28 (1962) 2717

31. A. E. Johnsrud, M. G. Silbert and H. G. Barshall,

Phys. Rev. JJ_6 (1959) 927

32. D. Drake, I. Bergqvist and D. K. McDaniels , Phys.

Lett. 36B (1971) 557

33. O. Schwerer , M. Winkler-Rohatsch , H. Warhanek

and G. Winkler , Nucl. Phys. A264 (1976) 105

34. R. M. Lessler, WRENDA 76/77 World Request List

for Nuclear Data, Report INDC(SEC) -55/URSF

(1976)

169
RiUARKS ON THE 2200 m/s and 20° C MAXWELLIAN NEUTRON DATA FOR

U-233, U-235, Pu-239 and Pu-241

H.D, Lemmel

Nuclear Data Section

International Atomic Energy Agency

A-1011 Vienna, Austria

Abstract: Attention is drawn to the still existing systematic discrepancy between experimental

cross-sections for 2200 m/s neutrons and those for a 20° C Maxwellian neutron spectrum for U-235

and U-233.

(Keywords: Neutron nuclear data evaluation. Fission standards, U-233, U-235» Pu-239j Pu-241,

thermal neutron cross-sections, fission-neutron yields. Pu-239 half-life.)

Introduction

In 1965 [1], 1969 [2] and 1975 [3] the IAEA Nuclear

Data Section, in cooperation with external specialists,

published consistent sets of recommended best values of

the thermal neutron data of the main fissile nuclides.

The method of evaluation was a multi-parameter

least-squares fit of all available experimental data,

after reviewing and, where feasible, reassessing the

authors' quoted values and errors, usually in consultation with the authors or other experts.

In the present paper I would like to demonstrate

which discrepancies still exist for these important

data.

existed in two distinct groups around 3.7 and around

3,8 respectively. We then concluded that , considering

the Chalk River a value, a high value of p (Cf-252)

should be correct rather than a lower one, - Soon

after, this conclusion turned out to be wrong.

This wrong prediction of 1969 was a good

illustration of the limited value of a least-squares

fit, when systematic discrepancies are present, and

in particular when several different sources of

discrepancies interfere within the ssime fit.

Today, this limitation is no longer as serious

as previously, but it seems advisable to remember

that the results of a least-squares fit are subject

to certain limitations which I shall discuss briefly.

Discrepancies

Although the thermal neutron data of the fissile

nuclides are basic parameters and reference standards

for the data at higher energies as well as for data of

other nuclides, it has not yet been possible to issue

a final set of recommended values, due to the existence

of systematic uncertainties.

In 1965 the situation looked still fine. The

accuracy of experimental data was not yet as good as today, and possible discrepancies did not significantly exceed the experimental errors.

In 1969 it became obvious that there were, among

the experimental data, some unexplained discrepancies.

For example, experimental U-235 fission cross-section

data were discrepant; data for the neutron-yield per

fission, V, were discrepant; the mean energy values

of the fission neutron spectra, which influence the

5 values, were uncertain; the Westcott g-factors and

their temperature dependence were uncertain; halflife values, which influence the fission cross-section

values, were discrepant; and some of them, for example

the experimental values of the Pu-239 half-life, were

consistent but turned out later to be all together

wrong. In particular the neutron yield data, v and , appeared to be inconsistent with the Chalk River

irradiation experiments which jrielded a capture-tofission cross-section ratio a with high accuracy.

The equation

1 + S = v/^

(where the sign^'denotes the 20° C Majcwellian spectrum

average) was not fulfilled within the error limits of

the experimental data. At that time, the weakest

parameter within this equation seemed to be v, since

all u values had been measured relative to the

spontaneous v of Cf-252, of which experimental values

Treatment of correlated data

The thermal cross-sections and neutron yield

data for one fissile isotope are all interrelated, and

the data of different fissile isotopes are also interrelated due to ratio measurements. Thus there are

about 200 interrelated experimental input data, thereof about 50 independent variables, which are simultaneously treated in the least-squares fit.

The main problem in such a fit is the treatment

of experimental input data which have correlated error

sources. It would lead to wrong results if such cor- related data would be treated in the fit in the same way as uncorrelated data. Several methods were used

to take care of such correlations.

In 1965 a mathematical procedure [1] was used by

which certain types of data correlations can be re- presented in the fit by appropriate down-weighting of

the correlated input data and their ratios. This

method applies when, e.g., the fission cross-section

is measured in one experiment for three nuclides; the

three results are then correlated due to the uncertainties of all those corrections which they have in common.

For a number of input data this method was still

applied, but in 1975 an additional, more flexible

method was introduced, by which error sources common

to several input data are formulated in the fit as independent parameters.

For example, most measurements of the fissionneutron yield v depend on assumptions about the energyspectuxm of fission neutrons. Therefore, the mean

energy values 'S of the fission neutrons of each of the

nuclides considered, were treated as independent parameters in the fit, thus adjusting automatically the

experimental v data for changes in the values of the mean fission neutron energies "S.

Since the author was unable to attend, this paper was summarized by B. E. Leonard, Jr. at the Symposium.

170
Another example are the fission cross-section

measurements which mostly depend on half-life values,

when the sample assay was done by alpha-counting.

Therefore, the half-lives for the nuclides considered

were entered as independent parameters in the fit,

thus adjusting automatically the experimental fission

cross-section data for changes in the half-life values,

A special problem is the treatment of the Westcott

g-factors, which relate the cross-sections for mono- energetic neutrons at the reference velocity of 2200

m/s to the cross-sections for a thermal Maxwellian

neutron spectrum. The g-factors are strongly dependent

on the exact cross-section cuirve shape o (E) in the

thermal neutron energy range, in particular in the not

very well known range below 0,0253 eV [3].

In the fit, experimental values of 2200 m/s cross- sections, of Maxwellian average cross-sections corrected to a spectrum temperature of 20 C, and g-factor

values as obtained from curve shape studies, are all

entered as parameters to be fitted. However, these

data show a correlation which is difficult to formulate.

Firstly, the absorption cross-section curve a (E)

is derived from measurements of the total cross—section.

Thus, the g-factor computed from the curve 0 (E) is

dependent on the scattering cross-section assumed.

Secondly, if in the course of the least-squares fitting

procedure, the value of the absorption cross-section

(3 (0.0253 eV) is adjusted, then this adjustment will

af'fect the cross-section curve-shape a (E) and, as a

consequence, will change the g-factor g . a

Consequently, the absorption cross-section, the

scattering cross-section, and the g-factor for absorption are correlated in the fit. Whereas in the earlier

IAEA evaluations this correlation had been ignored, it

had been considered, at least in a good approximation,

in the 1975 evaluation.

These examples are mentioned here, in order to

illustrate that the more important complications inherent in the least-squares fitting method have been

taken care of, and that the results of the least-squares

fit should therefore be reliable.

The improved fitting procedure by B. R. Leonard Jr. et al

Nevertheless, this way of a least-squares analysis

leaves the evaluator in a difficult situation. If the

g-factors are adjusted in the fit, as a result of

simultaneous fitting of monoenergetic and spectrum

average cross-sections, the evaluator is faced with

the problem , that he has to construct, subsequently,

a best cross-section curve which exactly reproduces the

g-factor resulting from the least-squares fit. Obviously, this is a tedious, if not impossible job.

BoR, Leonard Jr. et al, therefore, achieved a con- siderable progress in the evaluation method of the

thermal cross-sections, by considering in the least

squares fit not only the monoenergetic and thermal

Maxwellian experimental cross-sections but also the

entire low-energy cross-section curves o(E) in a foi^

mulation using multilevel resonance parameters [4].

The results for U-235 were published in February 1976,

and I am not informed at the moment when writing this,

whether this work continues for the other fissile

nuclides. But it would certainly be most useful, if

this work by B. R. Leonard et al could be continued as a simultaneous fit for the four main fissile nuclides.

The recommended U-235 data obtained by B.R. Leonard

et al deviate partly by more than a staindard deviation

from the 1975 IAEA recommended values. However, this

deviation is only an expression of the still existing

disturbing discrepancies among the existing experimental data.

The last (?) discrepancy to be solved

Despite of the more powerful fitting method

developed by BoR. Leonard, the less powerful IAEA fit

still serves a good purpose, because it may give us

a hint about the source of the, perhaps last, disturbing discrepancy which remains to be solved for the • thermal fission data.

In 1969 the results of the least-squares fit were

somewhat unsatisfactory, since a number of different

sources of discrepancies were interfering, thus making

the interpretation of the results difficult. It was obvious that there were disturbing discrepancies among

the experimental data, but it was not evident which of

the many input data may be responsible for the discrepancies.

In the 1975 evaluation, the situation had so much

improved, that one could well localize the origin of

the discrepancies encountered. The one of the discrepancies has meanwhile found its solution, after a new

more reliable value of the Pu-239 half-life has been

established in a number of parallel experiments [5].

Hence it seems that there is only one discrepancy

left which will hopefully be the last one: the systematic discrepancy between 2200 m/s cross-sections and

thermal Maxwellian cross-sections for the uranium

isotopes.

This is illustrated in Table 1.

The data given in Table 1 were obtained from

least-squares fits of exactly the same input data as in 1975 [3] except for two items: a lower value of

the Pu-239 half-life (2413O + 50 years) was tentatively adopted, which seems to be the consensus of several

independent experiments being performed at present [5]»

and the final values of J.R. Smith's r) experiment [6] were used instead of the somewhat lower values reported

in 1974 [7].

When trying to localize the origin of the discrepancies encountered, several least-squares fits were made with different subsets of the experimental

data, and the internal consistency of each subset was

studied. As a result it became obvious that one can devide the experimental data into tv.'o subsets, where

each subset has a very good internal consistency, but

both subsets are inconsistent with each other.

The one subset comprises all monoenergetic 2200

m/s data together with the v data, see column 1 Table 1. The other subset comprises all data measured in a thermal Maxwellian neutron spectrum, see column 3 Table

1, Both of these subsets show an internal consistency

which is striking and by far better than can be statistically expected, (Within each of both subsets

only 4% of the input data deviate from the fitted value

by more than the quoted, or re-assessed, experimental

error,

)

Column 2 in Table 1 shows the thermal Maxwellian

data as deduced from the 2200 m/s data as given in

column 1 using g-factor values as determined from

curve-shape studies (see Table 2 column I).

171
1. 2. 3. 4.

Fit of

experimental

2200 m/s data

incl. V data

20°C Maxwellian

data deduced

from col 1. vri.th 20°C g-factors

of Table 2 col.1.

Pit of

experimental

20°C Maxwellian

data

Difference

between

columns

2. £ind 3.

U-233 0^

0Ya7

573.8 + I08

533.2 + 3.0

40.6 +2.5

0.076 + 0.005

2.294 ± 0.009

2.469 + 0.008

573.1 + 2.1

531.4 + 3.1

41.7 + 2.7

0.079 + 0.005

2.239 J- 0.010

574.3 + 3.7

526.9 + 3.4

47.4 + 0.4

0.090 + 0.001

2.296 i 0.019

+0.8

-4.5 (0.9/0)

(l3/»)

+0.012 (^'^'^o) +0.007

U-235 0,3,

d

Ya

7

€80.0 1.8

583.1 J; 1.9

91.9 + 2.3

0. 156 + 0.004

2.079 ± 0.008

2.404 + 0.006

665.1 ± 2.0

574.9 + 2.0

90.3 i 2.3

0.157 + 0.004

2.077 ± 0.008

663.6 J; 4.5

566.0 i 3.8

97.5 i 0.8

0.172 i 0.001

2.090 i 0.016

-1.5

::8^ (1.5/.)

il^ (8/0) +0.015 flO/o)

+0.013

Pu-239 0^

'^Y a

7

1014.1 +4.3

748.1 i 2.8

265.9 + 4.1

0.355 + 0.006

2.110+ O.OOQ

2.860 j; 0.009

1091.7 ± 7.0

787.8 J; 3.7

304.0 + 7.1

0.386 + 0.010

2.064 ± 0.014

1095.5 + 7.5

787.8 i 5.3

307.7 i 2.6

0.391 ± 0.002

2.060 i 0.019

+3.8

0.

+3.7

+0.005

-0.004

Pu-241 0 a

• 0Ya

?

1377. + 13.

1023. ± 11.

355. + 8.

0.347 ± 0.009

2.165 + 0.013

2.915 + 0.010

1431. + 14.

1069. i 13.

362. i 11.

0.339 Jt 0.012

2.177 i 0.019

1432. Jt 13.

1059. ± 10.

37 3. i 8.

0.352 i 0.008

2.192 ± 0.032

+ 1. -10.

+ 11. tO,oi3 (4/0) +0.015

Cf-252 3.740 + 0.009

Table 1. Discrepancies between experimental 2200 m/s cross-sections and

experimental 20 C Maxwellian cross-sections.

When comparing the directly measured 20 C Maxwellian data with those derived from monoenergetic data

and g-factors, one finds:

1. Discrepancies exist for the uranium isotopes and, to a much lesser extent, for Pu-241. The data for

Pu-239 are consistent. (Some authors, e.g. [4]>

quote that the data for Phi-239 are also inconsisteni;

However, this is correct only for the Pu-239/U-235

cross-section ratios, where however the discrepancies seem to come rather from U-235 than from

Pu-239.)

2. For the uranium isotopes, systematic discrepancies

exist for the fission and capture cross-sections

and their ratio ot, but not for the neutron yield ,

Consistent are also the absorption cross-sections,

where the discrepancy contributions from capture

resp. fission seem to cancel,

3. The order of magnitude of the discrepancies is

1-2^ for the uranium fission cross-sections and

8 - 155^ for the uranium capture cross-sections and

the ratio a.

These discrepancies are well known since long.

But their origin is still unknown.

Since our 1969 evaluation all important experimental data have been carefully remeasured or re- analyzed: the fission cross-sections, the half-lives

involved, the scattering cross-sections, the Chalk

172
River thermal irradiation data, the fission-neutron

yields 9 and v, the fission-neutron spectra, etc. Although one can never be sure of unexpected surprises,

the striking consistency of the experimental data

vri.thin the two subsets suggests that all these values

are reliable within their quoted errors.

The only area which has not been reinvestigated

with comparable scrutiny, is the lowest energy region

around and below 0,01 eV, which significantly contributes to the 20 C Maxwellian cross—sections.

In 1975 [3], I thought that the lowest-energy

cross-section curve shapes and thus the Westcott

g-factors may be responsible for the discrepancies,

but the analysis of B.R, Leonard et al [4] showed that

the g-factors as obtained from the curve shapes, are rather accurately known, at least for U-235» not so much for U-233. However, a re-examination or re- measurement of the few existing data in the lowest

energy region still seems to be advisable.

1. 2.

„ — / rc/E* Mamj dE/o 'S'

U-233 absorption

fission

capture

0.999 + 0,003 [8]

0.9965 ± 0.002 [8]

(or 1.000 for

different extrapolation of a(E)

to zero energy)

(1.03 i 0,02)

1.001

0.988

1. 17

U-235 absorption

fission

capture

(0.978 + 0.002)

0.9775 + 0.0015 [4]

0.982 i 0.002 [4]

0.976

0.963

1.06

Pu-239 absorption

fission

capture

1.077 ± 0.005 [10]

1.053 J; 0.003 [9]

(1.14 + 0.02)

1.03l

1.053

1, 16

Pu-241 absorption

fission

capture

1.039 + 0.003]r[lO]

1,045 + 0.006j[[ll]

(1.02 i 0.02)

1.039

1.035

1.05

Table 2. 20°C Westcott g-factors

col. 1; g-factors calculated from the curve shape

o(E) according to

g= Jd(E) /EMaxw(E) dE/o^/F

The values in () were deduced from the other

two g-factors quoted for the same isotope

using cross-sections of col. 1 in Table 1,

col. 2; g-factors calculated from 20°C Maxwellian and

2200 m/s experimental cross-sections according to g = 3 / 0^

using the values from columns 3, resp. 1, in Table 1.

It is however evident, that the lowest-energy

curve shapes and thus the g-factors derived from them,

cannot be responsible for the full amount of the discrepancies. The g-factors

g = 3/0^

obtained from comparison of experimental data 3 in a

20 C Maxwellian neutron spectrum with a for 2200 m/s

neutrons, cannot be brought into agreement with the

g-factors obtained from curve shapes

S = j

o(E)-^ Maxwellian (E) dE/o^ yF

This is illustrated in Table 2, One must suspect that

there exists a still unknown physical effect in the

lowest energy range.

Scientists at NBS are considering whether this

unknown effect may be related to phonon transitions

between cold neutrons and the sample [l2]. Not knowing any details of these considerations, I believe that

they may go into the right direction, although I cannot

judge whether such effects may give a quantitative ex- planation of the discrepancies encountered. I can only

conclude that the energy range around 0,01 eV will re- quire further investigations before the thermal cross- sections of the fissile nuclides can be established as reliable standards.

References

[1] Westcott C,H. , Ekberg K. , Hanna G,C.,

Pattenden N, S,, Sanatani S,, Attree P,M,

,

Atomic Energy Review J 2 (1965) 3,

[2] Hanna G.C„, Westcott C,H., Lemmel H,D,,

Leonard Jr, B,R, , Story J.S,, Attree P.M,

,

Atomic Energy Review 2 (1969) No, 4 p, 3,

[3] Lemmel H„D, , Proceedings of the Conference on Nuclear Cross-Sections and Technology, Washington

D.C., 3-7 March 1975, NBS Special Publication 425

(Oct, 1975) Vol, 1 p, 236,

Note that the more detailed report announced as Lemmel H,D, , Axton E,J,, Deruytter A, J,,

Leonard Jr, B,R., Story J,S,, INDC(KDS)-64 has

not yet been issued,

[4] Leonard Jr, B.R,, Kottwitz D,A, , Thompson J,K,

,

EPRI/NP-I67 (Pebo 1976),

[5] Vaninbroukx R. , ANL/m)-77-1 (Nov. 1976) p, 29,

[6] Smith J.R, , Proceedings of the Conference on Nuclear Cross-Sections and Technology, Washington

D,C,, 3-7 March 1975, NBS Special Publication 425

(Oct, 1975) Vol, 1 p, 262,

[7] Smith J,R„, USNDC-11 (June 1974) 12.

[8] Primarily based on Steen N,M, , WAPB-'m-1052

(Sept. 1973),

[9] Deruytter A, J,, Becker W. , Annals of Nucl. Sci,,

and Eng. 1 (1974) 311, and Deruytter A.J,,

Wagemans C., J, of Nucl, En, 26 (l972) 293,

[10] Westcott C,H, AECL-3255 (April 1969), value increased due to revised scattering cross-section.

[11] Lemmel H,D,, Westcott C.H., J, of Nucl. En, 2A

(1967) 417, corrected for revised scattering

cross-section,

[12] Bowman C,D, , private communication 1977*

173
AN ASSESSMENT OF THE "THERMAL NORMALIZATION TECHNIQUE" FOR MEASUREMENT OF NEUTRON CROSS SECTIONS Vs ENERGY*

R. W. Peelle and G. de Saussure

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37830

Refined knowledge of the thermal neutron cross sections of the fissile nuclides and of the

(n,a) reaction standards, together with the reasonably well-known energy dependence of the

latter, have permitted resonance-region and low-keV fissile nuclide cross sections to be

based on these standards together with count-rate ratios observed as a function of energy

using a pulsed "white" source. As one evaluates cross sections for energies above 20 keV,

optimum results require combination of cross section shape measurements with all available

absolute measurements. The assumptions of the "thermal normalization method" are reviewed

and an opinion is given of the status of some of the standards required for its use. The

complications which may limit the accuracy of results using the method are listed and

examples are given.

For the ^^^U(n,f) cross section, the option is discussed of defining resonance-region fission integrals as standards. The area of the '^9 eV resonances in this nuclide may be known

to one percent accuracy, but at present the fission integral from 0.1 to 1.0 keV is known

to no better than about two percent. This uncertainty is based on the scatter among independent results, and has not been reduced by the most recent measurements. This uncertainty

now limits the accuracy attainable for the 235u(jj^f) cross section below about 50 keV.

Suggestions are given to indicate how future detailed work might overcome past sources of error.

['"BCnju); cross section; ^Li(n,a); neutron; normalization; resonance; shape; standards;

thermal; 235u(n,f)]

Introduction

Since the earliest days of neutron cross section

measurements using pulsed "white" neutron sources,

workers have utilized the simple (1/v) shape of the

(n,a) cross sections of ^Li and •'"b to exploit the

relatively well known absorption cross sections of the

thermal energy range. ^ Through the last decades our knowledge of the requisite data has become successively

more refined, and the existence of more powerful

sources has allowed this "thermal normalization method"

to be extended over an extremely broad energy range of

'^^8 decades. In this talk the authors explore the definition of the method,- identify some practical problems

which have prevented this method from dominating all

others, show some measure of the success so far obtained, and suggest where the cross section community

might labor in its quest to exploit this technique to the extent many have dreamed should be possible. This

talk is pertinent to the present meeting because the

thermal normalization technique provides a prime re- quirement for the shapes of the standard (n,a) reactions and for the 2.2 km/s cross section standards.

For ^^^U(n,f), the technique is used to aid the definition of a secondary standard of great utility, so many of the examples below are from work on that

nuclide. The recent papers of Leonard^ and of Bhat^

give forceful examples of the effort required to apply . the method to this nuclide in a more correct way than

earlier was possible for a present author.** Carlson

and Czirr^ have discussed the status of the thermal

normalization for this cross section from a somewhat

different point of view.

One reason for the ascendancy of the technique is

that with it a linac or other white-source laboratory

can provide an in-house cross section result which is

complete over a very large energy range. All this can

be done without direct measurements of sample mass which would otherwise generate a minimum uncertainty

of 1-2 percent. Therefore, many cross section sets are normalized at thermal energies, although segments

of the results could as well have been normalized in

another way; therefore, more of the experimental

literature seems to be based on this method than in

fact is the case.

1

From a more global point of view, which requires

that all existing data be combined, one sees at once

that the use of a normalization based on thermal cross sections is conceptually separate from the use of relative measurements of count ratios between a counter

based on the unknown and one based on a cross section

of known shape. (This distinction is nicely emphasized

by evaluation techniques such as those of Poenitz^ in

which shape and normalization decisions are separated.)

In fact, nearly all sets of cross sections vs energy,

in whatever energy range, are based on count ratios to the output of a detector with smooth and presumably

known energy response, and it is a matter of detail

whether or not this response follows the values of a known cross section. One can expect a typical synthesized result from a linac lab to consist of one or more

segments of relative o(E) data, the lowest segment

reaching the thermal energy region and providing one normalization. This normalization has peculiar status

only because it may be tied directly to applications

through thermal critical experiments and when it provides a normalization with smaller uncertainty than

others available to an evaluator. Other means of absolute normalization are available, especially in the MeV

range, so that for sufficiently high energy some absolute measure other than the thermal normalization is

likely to dominate. Whether this crossover of relative

importance occurs at 20 keV or at 1 MeV or above will

depend on the nuclide and the opportunities provided to the experimenters at the several laboratories. However,

in every case a thermal normalization can be applied,

it is important and could not be ignored below '^'10 keV

even if ±1% absolute normalization data were available

at scattered energy points above 20 keV.

In the special case of 2^^U(n,f), one may be

tempted in evaluating the standard cross section to ignore data based on normalization at thermal energies

because the presence of fluctuations prevents this fission standard from being recommended for use at energies below 200 keV, and because there is such a wealth

of experimental absolute data at the higher energies.

Such a reaction would be hasty for three reasons:

1) the weight of measurements based largely on normalization at thermal energies is far from null, and independent methods must always be compared for mutual
consistency if we are ever to achieve the correct cross

sections, 2) average cross sections of "standards"

quality are needed in the region of fluctuations if one

is to take advantage of clean integral fission-rate

experiments in which the lower-energy neutrons play a significant role, and 3) when the approximate shape of

the flux is known and is smooth, one may make use of

the ^^^U fission rate to fix the normalization of the

flux even if one cannot be sure enough of energy scale

alignment to use directly the ratio observed to ^^^U(n,f) within each experimentally defined energy

cell. The question of the importance of ^^^U{n,f)

below 0.2 MeV is actually moot because this cross section must anyhow be known well because of its direct

practical applications.

What the Thermal Normalization Method Requires

For now, assume a restricted definition of the

thermal normalization method. The method requires use

of a standard cross section known well as a function of

energy and at thermal energy (2.2 km/s), and it may be

applied to an unknown whose cross section is also well

known in the thermal energy region. Figure 1 exhibits

the standard relation between unknown, known, and observed quantities, and quotes the restrictions on the

method's applicability. One must previously have subtracted any backgrounds from the observed rates.

Understood departures from the required energy independence of flux ratio or detector efficiencies may be

corrected. The beauty of the technique is that neither

neutron fluxes nor sample masses or areal densities

appear in the equation of Figure 1.

The Thermal Normalization Method Requires No Mass

Measurements and Only Relative Count Rates vs^ Energy

If R , (E) = Reaction Rate Ratio x sample— x/s standard sample

ratio of flux between two samples is not a function

of energy.

Count-rate ratios may be used for I^/gC^) whenever

efficiency for detecting reactions is also not a function of energy.

Figure 1.

To sense whether the necessary standards data are available to apply this method, one may look at an example consisting of the thermal cross sections of the

fissile materials and the evaluated results for the

light-element (n,a) reactions. Table 1 compares

several evaluations of the 2.2 km/s fission cross sections of some important materials. Evaluations will

differ until data inconsistencies are removed. The

sometimes-quoted uncertainties of '^'0.2 percent may be

too ambitious, but one senses that half-percent uncertainties suffice for fission. Achieving such an accuracy in a single absolute measurement involving sample masses involves very great care.^ The 2.2 km/s cross

sections of ^Li and ^'^B are typically quoted to 0.4 and

0,2 percent,^ and Table 2 exhibits the stability of the

evaluated energy dependence of these cross sections

through the last three ENDF/B evaluations. Since measured data for these cross sections at any one energy

Table 1. Evaluations of 2.2 km/s Fission Cross Sections for

Uranium Have Varied By More Than the Uncertainties Quoted

"='U 2'=U 2 39p^ Reference

530.6 + 1.9 580.2 + 1. 8 741 .6 + 3. 1 Hanna (IAEA) 1969'

585.7 + 1. 8 742 .5 + 3. 1 De Volpi, 1971^

526.3 + 0.8 577.5 + 1. 1 Steen, 1972^

533.7 + 2.7 585.7 + 2. 3 742 .0 + 4. 2 ENDF/B-IV, 1973^^

529.9 + 1.4 583.5 + 1. 3 744 .0 + 2. 5 Lemmel, 1975^

583.5 + 1. 7 Leonard, 1976^

G. C. Hanna et al . , Atomic Energy Review ]_, No. 4, p. 3 (1969).

^A. de Volpi, Conference on Neutron Cross Sections and

Technology, CONF-710301 (Vol. 2), p. 564 (1971).

'^N. M. Steen, WAPD-TM-1052 (Sept. 1972).

'^J. R. Stehn, BNL, private communication to Cross Section Evaluation Working Group members on final fit Q4

(Dec. 13, 1973).

H. D. Lemmel, Conference on Nuclear Cross Sections and

Technology, NBS Special Publication 425, p. 286 (1975)

^B. R. Leonard, D. A. Kottwitz, and J. K. Thompson,

"Evaluation of the Neutron Cross Sections of ^^^U in

the Thermal Energy Region," Report EPRI-NP-167-Proj ect

512 (1976).

Table 2. Evaluated ENDF/B Cross Sections for the

6 Li and ^ B(n,a) Reactions a

(Values of a^eV/(cr /0.0253) are

0 Tabulated)

^Li(n,a) 1 0 B(n,a)

E(kev) III IV V III IV V

0.1 0.999 1.001 0.999 0.995 0.994 0.994

1.0 0.987 0.998 0.987 0.985 0.982 0.983

10.0 1.000 1.009 1.004 0.963 0.965 0.963

30.0 1.027 1.054 1.042 0.958 0.970 0.963

50.0 1.078 1.121 1.099 0.971 0.984 0.972

100.0 1.379 1.460 1.386 1.035 1.016 1.000

a in

0 barns

940.3 940.0 935.9 3836.5 3836.5 3836.6

Values for Versions III and IV were obtained from the

files of the National Neutron Cross Section Center at BNL. The ENDF/B-TII standards are also described in BNL 17188 (ENDF-179), M. K. Drake, editor (1972). Version IV was documented by Hale, Stewart, and Young in

LA-6518-MS (1976). The preliminary evaluation for Version V was obtained from G. M. Hale et al., LASL (1976)

and also included in the Minutes of the CSEWG Normalization and Standards Subcommittee, May 17-20, 1976.

in the tens of keV region spread widely (10 percent)

,

the stability shown could be illusory if surprises

develop in proper representation of the data using

reaction theory.^ Though complete evaluated uncertainty information for the energy dependence of these cross sections has not been given, one can see why confidence

in the cross sections through 10 keV to 1 percent has

been gained from the stability of the result and the

closeness to a (1/v) response. (These authors are con- cerned whether one can at present have confidence in

175
either reaction cross section at 50 keV to 2 percent or

at 100 keV to 3 percent, but the doubt arises from

casual rather than detailed studies.)

These (n,a) reactions and the ^'^BCn.ay) reaction

have been used as standards to much higher energies, so

it seems important that work continue to establish the

underlying cross sections. Fortunately, other flux

detectors based on the hydrogen cross section become

available above about 10 keV, but counters based on

'^Li and ^'^B can have much better time response than the

hydrogen proportional counter which becomes useful at this energy. As additional smooth-response detector

systems, not applicable at thermal energies, become

available at energies above 10 keV, the simple thermalnormalization technique discussed above gradually loses

its pre-eminence. In weighing the data or planning an experiment, one must balance the niceness of a fresh

detector system against the need to internormalize

results if the new system can no longer "see" the

thermal-energy guidepost.

Of course, the thermal normalization method need

not depend on an (n,a) reaction, and need not depend on

a cross-section shape standard at all if a detector of

calculable response with adequate time resolution can

be used over a broad energy range down to thermal energies.

Experience with the Thermal Normalization

Technique for ^^^U(n,f)

While thermal normalization may be employed by a single group to give results covering a broad energy

range, its most important application is in cross sec- tion evaluation. An evaluator quickly finds that there

are few individual measurements which span the region

of interest without significant changes of apparatus,

so one must combine the results from a number of individual experiments covering the energy range in patchwork fashion. Moreover, each reported data set may be

referred to a different value of the 2.2 km/s cross section using a different technique for establishing

this normalization, and be based on a different shape

for the standard (n,a) reaction used to determine the

energy dependence of the flux. With great effort, and

the help of the data center system, it is possible to unravel the variations of data treatment.

Leonard has recently

renormalize on a common b

covering portions of the

of the examples in this s Through such consistency

ble to uncover systematic

the original authors. Th

malization requirements r cent, not counting the ef

km/s cross section used.

described his efforts to

as is a number of measurements

energy range below 1 eV; most

ection depend on his work.^

studies it is sometimes possidifficulties not suspected by

e tables of Ref 2 imply renoranging from 0.3 to 1.5 perfect of differences in the 2.2

For reasons given by Deruytter and Wagemans'^'' and

suggested in the next section, and because some of the more careful measurements do not in any case reach to energies above the resonance region, it is convenient

to define the fission integral of the resonances near

9 eV as an intermediate standard. Table 3 gives the

data for this integral using two popular choices for

the energy bounds. Most of the listed data follow the

renormalizations given by Leonard^ and the uncertainties on these values reported by Bhat,^ but the first

line came directly from Ref 10 renormalized to a° =

583.5 b, and the Gwin data on the last line came from

newly analyzed data sets.-^-^ The uncertainties on the

Deruytter and Wagemans data were taken from Ref 10, and

the uncertainties given on the last two data sets were

assigned by this author without much analysis. The

weighted average integrals at the bottom of the table

indicate that the decision, whether to include Leonard's

re-evaluation of Ref 10 or the original results, alters

the output average by as much as the uncertainty assigned by expanding the propagated output uncertainty to make x^/df =1- In the presence of the observed inconsistency and conflict, it is difficult to accept the

small 0.6 percent output uncertainty given. Pending

more detailed uncertainty analysis, and including an uncertainty for the 2.2 km/s value, the overall uncertainty on this integral is judged to be about 1 percent.

Based on more study, Leonard has estimated a 2.8% uncertainty in the 7.8-11.0 eV integral.^

Table 3.

DATA ARE INCONSISTENT FOR THE FISSION AREA"

OF THE ^ 9 eV RESONANCES IN ^'hj.

rlO. .11. dE

J 7.8 dE

(barn-eV) (barn-eV) Reference

221 + 2 238 ± 2 Deruytter & Wagemans (1971)

.

(226 + (243 ± 2) Deruytter, per Leonard (1976).

227 2b Czlrr & Sldhu (1976). Private conmunication

(Feb. 1977). Leonard (June 1976) gives

224 b-eV for this value and 241 b-eV for Czirr's 7.8-11.0 eV Integral.

219 + 3b 236 ± 3b Gwin et^ al . (1976), per Leonard (1976).

225 + 4" 241 ± 5" de Saussure et al^. (1966), per Leonard (1976). 234 7b 252 ± 7'' Bowman (1966), per Leonard (1976).

216 6 Shore & Sailor (1958), per Leonard (1976). 226 + 3 245 ± 3 Gwin (1977). Private communication. These

two integrals come from separate new measurements

.

a(7 4 -10.0) = 223.7 ± 1.5(225.2 + 1.4) b-eV, = 11.8(9.5)<=

a. Normalized to = 583.5 b. Table foil ows that of Leonard (1976)

.

b. Uncertainties as reported by Bhat (1976).

c. Values in parentheses use the Deruytter data renormalized by Leonard using only the data above 0.21 eV,

The energy intervals used in Table 3 which are commonly chosen for study are plausible ones, but if

normalization or the relative behavior of various data

sets are to be considered seriously, it is highly desirable to compare the areas of two or more resonances.

In this way shape differences can be sensed which might

place an evaluation procedure at an appropriate level

of doubt.

If the fission integral of the '^9 eV resonances is

considered given, then the chosen value can be used to normalize experiments which did not reach to thermal

energies or for some reason should not be trusted in

that regime. Table 4 shows most of the independent

data for the 0.1-1.0 keV ^^^(n,f) fission integral which

can be normalized at thermal energy or in this resonance.

Table 4.

THE MOST MODERN VALUES OF THE 0.1-1.0 keV ^'^U FISSION

INTEGRAL DIFFER BY 7 PERCENT. Details depend

somewhat on evaluation technique.

Listed by Bhat Listed by Wagemans Original Author

12.4 11.8

11.4 12.2

11.8

I =

12.3 11.8

11.5

12.3

(11.9 ± .2) b-keV

de Saussure (1967)

Gwin (1976)

Czirr and Sidhu (1976)

Wagemans (1976) Wasson (1976)

^. R. Bhat, ANL-76-90, p. 307 (1976). Normalized to = 583.5 b and using the ENDF/B-V (n,a) shapes. Wagemans

and Wasson data were normalized to 1(7.8-11) = 241.2 b-eV.

''c. Wagemans and A. J. Deruytter, Annals of Nucl. Energy 3^, 437 (1976). This table is based entirely on data and standard shapes utilized by the respective authors.

176
The data sets of Wagemans^^ and of Wasson-'^ were nor- malized in the resonance while the other three sets

retain the thermal normalization. The left column

lists the results as renormalized by Bhat,^ while the

right column lists the values given in the original

papers. Bhat's reworked values are to be preferred

because they were obtained in a consistent way, but in

this particular case the distinction is not great com- pared to the large scatter of results. The scatter

will appear little less than disastrous for anyone having strong hope for early use of thermal normalizations

for precision work; the range of values is 6 to 8 percent depending on whether one includes only the newer measurements which were optimized for the goal of

obtaining precise fission cross sections. In fact

there seems to be no reason to downweight the older

results, since the uncertainty based on the scatter of

the last three values is just as large as that based

on all five. Simply put, the data base to date does

not define the 0.1 to 1 keV fission integral, relative

to the thermal cross section, to better than about 2 percent.

The particular energy interval displayed in Table

4 was chosen for convenience, but experience shows

that such comparisons for normalization purposes are needlessly confused if comparison intervals much narrower than two lethargy units are used. As noted by

Carlson and Czirr,^ it is important that a normalization interval be chosen above '^0.2 keV because below

this energy the shapes of the various measurements

seem more discrepant. Also, as treated in the evaluation of Bhat , ^ many additional measurements are available above about 0.1 keV. It is puzzling that prior

to the experiments of Refs 12-14 so few measurements

spanned the range from 7 eV to 1 keV.

From the scatter indicated in Tables 3 and 4, and

from the additional similar difficulties encountered

at higher energies,^''* one must conclude that the

application of the thermal normalization technique to ^"U(n,f) through the resonance and on into the keV

rej-ion is more difficult than has been assumed. More

care will be required than has already been expended.

Much concern is relieved by recognizing that many of

the studies were not optimized for the fission cross section alone, but as seen above there is important

scatter among the most recent measurements performed

with only this goal in mind. Considering the dispersion among results and uncertainties in the shapes of

the (n,a) standards above 30-50 keV, it is remarkable

that an evaluation task force was able to determine

that an alteration as small as 1 percent to the results

obtained by this method for the 0.01 to 0.2 MeV region

would be all that would be necessary to bring the

results into better agreement with experiments performed using monoenergetic sources and other normalization techniques.^

Complexities in Application of the

Thermal Normalization Method

Given the apparent adequacy at least through 10

keV of the data which underlie the use of the thermal

normalization method, why might various investigators

obtain conflicting results? Figure 2 lists some possibilities

.

The ratio data proving the consistency of the ^Li

and '''B(n,a) standards are quite limited, the work of

Sowerby et dl.^'' being most quoted though some indications are given in papers by Perez et al.^^ and by

Wagemans and Deruy tter . -"^ ^ Since the ratio of these

cross sections is presumably easier to measure than

the absolute value of either one, and since any inconsistency between these standards is causing serious

confusion, a significant effort would be worthwhile to establish beyond doubt the ratios of these cross sections

.

Why Don' t All Authors Get the Same Answers ?

• The (n,a) standards may be inconsistent.

• Efficiencies of detectors for flux and unknown may not

be proportional to reaction rates.

• Conflicting requirements on sample thickness, flightpath length, source pulse rate, and run time inhibit

optimization for accuracy.

• Assessment of counting backgrounds and resolution

functions is beset with hazards.

• Time pressure often curtails complete evaluation and

intercomparison of results.

Figure 2.

Detectors for the light element (n,a) reactions

are usually designed to have uniform efficiency for the

reaction products, independent of neutron energy at least to first order. Not often, however, does one see

a detailed demonstration that this assumption is valid

for a particular detector. The reaction Q-values are sufficiently large that one would not expect significant effects on detector performance from reaction

kinematics for energies up to at least 10 keV, but at higher energies the energy dependence should be worked

out for each detector design. Nonisotropy of reaction

cross sections in the cm. system also requires consideration; we have seen from the work of Raman et at. that a measurable fore-aft asjmimetry is confirmed even

for 10 eV neutrons when the detector subtends tt solid

angle I The observed energy dependence of this asymmetry to 10 keV had the form 1 + O.OOSx E(eV)°"^'^, so the effect is strong. This finding again suggests that

for precision work the energy dependence of flux detectors based on the (n,a) reactions should be analyzed

with considerable care.

The reason that a thermal normalization must so often be carried out through successive renormalization

of partially overlapping data sets is that the experimenter using a pulsed white source cannot simultaneously optimize all the experimental variables which

affect data accuracy; it is mandatory to compromise

other values as one .extends the energy range of an individual data run to minimize the number of successive normalizations. Gwin et al. in their paper

describing very broad-range experiments, discuss some

of the problems to be handled in such measurements.

One set of compromises involves sample thickness:

to obtain adequate counting rate one may use a sample

so thick that self-absorption is serious in the unknown

or in the flux detector, and in some cases a thick

sample in a fission ion chamber may complicate timing

and make less sure a constant efficiency for the detection of fission products. The sample-thickness con- flict might be resolved by establishing precision average cross sections using a thin sample, while filling

in the short-range detailed energy dependence using a detector with more material; a disadvantage of this

solution would be that backgrounds are more difficult

to assess using the black resonance technique when the

counting rates are low.

177
Assuming that the repetition rate and the thickness of anti-overlap filters may be changed at will

when using a pulsed white neutron source, another set

of compromises involves choice of flight-path length

and neutron pulse repetition rate. Short flight paths

(5-20 m) minimize overlap problems but may not permit

satisfactory time resolution for flux detectors and may

place any effect of "gamma flash", or of the electrical

noise associated with neutron pulse production, too close in time to the interesting data. Assuming the

flight-path length has been chosen. Figures 3 and 4 illustrate the shape of the count rate from a (1/v)

flux detector with two different arrangements to reduce

the low-energy sensitivity enough to avoid time-frame

overlap. In Figure 3 the experiment could cover energies to about one-half electron volt, but there is need

CADMIUM RESONANCES

NEUTRON FLUX -COBALT

FILTER

2 5 10 20 NEUTRON ENERGY (eV)50 100 200

Figure 3. The time spectrum in 10 B-based

flux monitor when a cadmium filter was used to eliminate time-frame overlap. The cobalt filter

was used for background estimation. From Ref 19.

to work hard to correct for the effect of the cadmium

resonances which may be resolved to different extents

in detectors for the flux and for the unknown. For

Figure 4 a boron filter was used to cut the flux off at

1 1 • 1

1—1

T 1 —— 1 i r

5(£) = 7,2f°-'e .-5-^/^-2,7 Al 5,8 keV

No 2.85 *<ev

AI 35-2

5'(£"1=0.6''5 (f)

. 96 AND < HOheV

5 (if) SPECTRUM WITH '°B ONLY ^i^^ ^^0^^^^^

I J

/ / S'{E) SPECTRUM WITH RESONANCE

; j

/ / FILTERS AND'°B

1

if<

'\'l eV, but the counting rate at about 4 eV is only one percent of what it would have been if a clean cutoffmechanism had been available to eliminate all neutrons

with energies below 1 eV. If more boron is used, a faster repetition rate is possible and vice versa ; in

the case of Figure 4 the net counting rate was optimized at about 3 keV, and counting rates could have

been improved by decreasing boron thickness and repetition rate had the experimenters wished to improve performance for energies below 3 keV. A point of this

discussion is that lowering repetition rates to increase the energy range of an experiment may increase

counting rates through a large share of the dynamic

range

.

Backgrounds in pulsed neutron experiments are usually determined by inserting into the beam "filters"

designed to remove all the neutrons at the resonance

energies of these absorbers. If the filter can be thin, one assumes that the filter does not affect the shape

of the overall spectrum of higher-energy neutrons which

normally cause all the troublesome backgrounds. Between

the well-spaced energies of the filter resonances, the

experimenter must interpolate a shape based on auxilliary data or an assumption of smoothness. Rarely has

it seemed possible to understand the shape of the background based on calculation of the transport of beam

neutrons through the regions surrounding the neutron

source and detector. Figures 3 and 4 illustrated this

background estimation method for boron-based flux detectors. Note that the backgrounds seem very low. In Figure 4, where a boron filter was used to roll off the

flux, one senses that the background ratio was not so favorable at lower energy since the boron filter attenuated relatively less the higher-energy neutrons inducing the background. Figure 5 from an earlier experiment^^ shows that measured relative backgrounds have

not always been so low; one can see that the average

cross sections and the valleys between resonances would

have been markedly affected by an error in the assumed

:• kIi!

'I

: I j ;

ASSUMED BflCKGROUNO

50 400 ENERGY (eV)

Figure 5. An example of background determination in a fission cross section measurement, from

Ref 19. Coincidences were recorded between a fission chamber and a liquid scintillator tank which

surrounded it.

ENERGY (eV) Neutron Energy Spectrum with ond wittioul ttie Resononce FiltersFigure 4. Time spectra in a '^B-based flux

monitor when a boron filter was used to avoid

overlap. The lowei; curve illustrates the use of resonance filters to help estimate backgrounds.

From Ref 20.

background shape. Figure 6 is a more favorable example

at higher energy where the background level appears to be about 1 percent.

When the background level estimated by the above

technique is not small, more detailed analysis is

required. Gayther et al.^^ have suggested one way to

178
proceed using data obtained using several filter thicknesses. The difficulties arise partly because one struggles to differentiate sharply between a neutroninduced background and a "tail" on the experimental

The small discrepancy on the low energy side of the

peak was subsequently explained by taking into account

neutron diffusion in the thick ^Li-glass detector and

the skew shape expected for the time distribution of

neutrons leaking from the target moderator.

1000

±. 100

10

S 1

<

°^ 0.1

Al 5.96 keV

8 0.01 h

0.001

10 20

NEUTRON ENERGY (keV)

100

Figure 6. A background determination in the

keV energy range using the black resonance technique, from Ref 16.

resolution function. The only difference is one of

degree. Figure 7, though from a transmission experiment, shows the effect of a close-range asjrmmetry in a resonance chosen by Olsen^^ to facilitate analysis.

1.0

0.8

^ 0.6

<

cr.

0.4

0.2

1

1 1 1 1 1 NEED FOR COMPLEX DETECTOR

RESOLUTION FUNCTION

1 1

ii. '1,

P3/2 RESONANCE

^°®Pb, 0.021 AT/A _ NORMAL RESOLUTION

FUNCTION ASSUMED

1 1 1 1 1 1 1 1 3340 3350 3360

NEUTRON ENERGY (eV)

3370

Figure 7. Discrepancy between observed and

expected shapes of a transmission dip when the

skew distribution of neutrons from a moderator

and detector distortions were neglected.

As the thickness of a resonance filter is increased, one typically obtains evidence that the flux

at a given energy is affected a little by that at near- by energies far beyond the nominal system resolution

width. Figure 8 illustrates this effect. The thirdthinnest filter would be computed "black", with less

than 0.001 transmission, yet increasing the sample

1.00

o 0.10 —

0.01

NOMINAL SAMPLE

THICKNESS

-^^U METAL (in.l * 0,0015

• 0.005

o 0.010 ° 0.030

0.100 ^ 0.425

DIFFICULTIES OF

DETERMINING BACKGROUND LEVEL

761 771 781 791 801 811 TIME CHANNEL

821 831

Figure 8. Neutron intensity near 6.7 eV observed through a broad range of ^^^U absorbers.

From the work of Ref 24.

thickness continued to reduce the counting rate near

the minimum as the energy region of low transmission

broadened. At any time after the neutron burst there

are neutrons present at low intensity from a very broad

energy range. The intensity and spectral distribution

of these neutrons depends on the source, collimator,

and detector construction and their environs. Improving accuracy of measurement through background reduction depends on identifying the sources of these offenergy neutrons, reducing their intensity, and learning

to correct more precisely for the presence of those

which cannot be removed. Given the difficulty of knowing the precise content of neutron beams, it seems preferable that the neutron flux shape be measured at the same time and place as the count spectrum from the

sample under study.

Figure 9 is adapted from a very old comparison

among preliminary sets of ^^^U(n,f) data in the reso- nance region. Much of the discrepancy was ascribed

to detection of neutrons scattered from aluminum and

concrete structures in the beam near the detector, and we believe that subsequent final data did not suffer

so much from this difficulty.

Some of the error sources discussed above relate

to counter design, some to beam production and collimation, and some may be inherent to pulsed-source methods.

179
Most have not been investigated and minimized with all

the tenacity and ingenuity which would be possible,

partly because practical experiments are beset by re- current crises originating in electrical noise, counter

decay, electronic malfunction, and sometimes the need

to suffer suboptimal beam conditions to meet the needs

of other experimenters.

technical work has really been completed. The requirements on final reporting are substantial if evaluators

are to have all the needed information.^^

The main standard cross section efforts called for

in this paper are listed in Figure 11; substantial

progress toward these goals has already been made in

this decade.

ORNL-OWG 7r-7l07 ICQ c [/\ 1 1

1 1 1 - 1 -

\\

— "^N / V.

W / I "~ l/\ 1 i —

V/ 'A

1 \

1 \

\\

1 1 1 I i 1 2 3 NEUTROM ENERGY (eV)

Figure 9, Two sets of preliminary observations of the ^^^U(n,f) cross section, illustrating effects ascribed to return of neutrons

scattered from structures in the beam.

How Precise Extrapolation from Thermal Cross Sections

May Yet be Achieved

Experience shows that accuracy improvements are iterative, and depend on improved sources and instruments as well as the broad shoulders of those who have

labored before. Figure 10 lists some major areas of

effort which require attention. In the second point

WE KNOW SOME DIRECTIONS TO LOOK FOR IMPROVING

ACCURACY USING THE THERMAL NORMALIZATION METHOD

• Improved analysis of backgrounds and resolution

functions will be needed for most detectors.

• To cover 7-8 energy decades, important overlap of

several data sets will often be required. Full

evaluation of results through this range is required

to establish credibility of normalization.

• Hasty exhibition of final results must be resisted,

if haste will prevent full analysis and exhibition

of data uncertainties and correlations.

Figure 10.

we mean by evaluation not only the intercomparison of

results on the same cross section but thorough joint

evaluation including the total cross section and other

partial cross sections. It is customary to advise

against the release and use of preliminary results,

but the third point indicates that the present authors

are more concerned about the apparently increasing

pressures to issue final results before the necessary

MORE WORK ON STANDARDS IS REQUIRED TO ENABLE THE FULL POTENTIAL OF THE THERMAL NORMALIZATION METHOD

• Perform measurements to prove compatibility of the 'Li(n,a),

'°B(n,a), and '°B(n,0CY) reactions. • Press hard to reach 100 keV with (n,a) cross sections accurate to one percent. • For ^'^U(n,f), try harder to establish two resonance- region fission integrals to 0.5 percent accuracy. A

standard integral over a broad region between 0.1 and

1 keV is probably also needed. 9 Continue work to achieve thermal-neutron cross section

standards with credible 0.3 percent uncertainties.

(Data inconsistencies now confuse uncertainty analysis.)

Figure 11.

With improved techniques and underlying standards,

the goal of precise extrapolation to high energies

based on thermal-region normalization can yet be

reached. This goal, qualified by the recognition that

absolute measurements at energies above 1^20 keV will

also be important, indeed must be achieved if there is

to be a 2^^U fission standard "for all seasons" and a knowledge of other cross sections sufficiently precise

to satisfy all needs of developing technologies.

Acknowledgements

The authors gratefully acknowledge discussions on

this topic with colleagues the world over, and particularly with R. Gwin, L. W. Weston, R. B. Perez, and

D. K. Olsen. We are indebted to J. Gentry and

E. Plemons for expedited production of the typescript.

References

*Research sponsored by the Energy Research and

Development Administration under contract with

the Union Carbide Corporation.

1. For example, J. F. Raffle and B. T. Price, Proceed- ings of the International Conference on the Peaceful Uses of Atomic Energy , paper P/422, Vol 4 p 187 (1955). Equally, papers by B. R. Leonard,

ibid, paper P/589, p 193 (1955); V. L. Sailor,

ibid , paper P/586, p 199 (1955).

2. B. R. Leonard, Jr., in ANL-76-90, Proceedings of

the NEANDC/NEACRP Specialists Meeting on Fast

Neutron Fission Cross Sections of ^^'u, ^^^U, ^^^U,

and ^^'^Pu, p 281 (1976).

3. M. R. Bhat, in ANL-76-90, o£. ext., p 307 (1976).

Also, related private communications (1976).

4. R. W. Peelle, ORNL-4955, An Evaluation for ENDF/BIV of the Neutron Cross Sections for "^^^U from 82 eV to 25 keV , (1976).

180
5. C. W. Carlson and J. B. Czirr, in ANL-76-90,

op.cit. p 258 (1976).

6. W. P. Poenitz and P. Guenther, in ANL-76-90,

0£.cit. p 15A (1976).

7. A. J. Deruytter, in CONF-701002, Neutron Standards

and Flux Normalization , p 221 (1971).

8. S. F. Mughabghab and D. I. Garber, BNL 325 Third

Edition, Vol 1 (1973).

9. G. M. Hale, NBS Special Publication 425, Proceedings of a Conference on Nuclear Cross Sections and

Technology , p 302 (1975). See also G. M. Hale, in

proceedings of December 1975 Trieste IAEA con- sultant's meeting on Use of Nuclear Theory in

Neutron Nuclear Data Evaluation . Lane, Hausladen,

and Monahan, in CONF-701002, op.cit. p 107 (1971).

Uttley, Sowerby, Patrick, and Rae, in CONF-701002,

op . cit . p 80 and p 151 (1971).

10. A. J. Deruytter and C. Wagemans , J. Nucl. Energy

25_, 263 (1971).

11. R. Gwin, ORNL, private communication (1977).

12. C. Wagemans and A. J. Deruytter, Annals of Nucl.

Energy 3, 437 (1976)

.

13. 0. A. Wasson, in ANL-76-90, o£.£it. p 183 (1976).

14. J. B. Czirr and G. S. Sidhu, Nucl. Sci. Engr. 6£,

383 (1976). See also Ref 5.

15. Sowerby, Patrick, Uttley, and Diment, in

CONF-701002, op.cit. p 151 (1971).

16. Perez, de Saussure, Silver, Ingle, and Weaver,

Nucl. Sci. Eng. 55, 206 (1974).

17. Raman, Hill, Halperin, and Harvey, CONF-760715-P2

,

International Conference on the Interactions of

Neutrons with Nuclei , p 1340 (1976).

18. Gwin, Silver, Ingle, and Weaver, Nucl. Sci. Eng.

59 , 79 (1976). See particularly sections IV. A. and IV. D.

19. G. de Saussure et al . , ORNL-TM-1804 (1967). Also,

with A. Lottin, Proc. Conf. Nucl. Data for

Reactors , Paris, 2, 233 (1967). (The values

given in the ORNL report are to be preferred.)

20. de Saussure, Silver, Perez, Ingle, and Weaver,

Nucl. Sci. Eng. 51, 385 (1973).

21. D. B. Gayther et al . , Proc. of a Panel on Neutron

Standard Reference Data , p 201 (1974). IAEA,

Vienna. Also, D. B. Gayther, AERE Harwell,

private communication (1972)

.

22. D. K. Olsen, ORNL, private communication (1977).

23. A. Michaudon, J. Nucl Energy Parts A/B, r7, 165

(1963)

.

24. D. K. Olsen et al . , Nucl. Sci. Eng. 62, 479 (1977).

25. F. D. Brooks, EANDC(UK)28 (1964).

26. R. W. Peelle, in ANL-76-90, op.cit., p 421 (1976).

181
REVIEW OF V FOR "^^"^Cf AND THERMAL NEUTRON FISSION

J. W. Boldeman

Australian Atomic Energy Commission Research Establishment

Private Mail Bag, Sutherland, NSW 2232, Australia

A review is presented of absolute measurements of v for the spontaneous fission of ^^^Cf and

of relative measurements for thermal neutron induced fission of 233, 235^ 239,2'4lpjj_ rpj^g discussion includes the consideration of a number of sources of revision that have been

suggested for some of the measurements. No evidence is found in the revised data of any

experiment-dependent systematic error. A set of recommended values is given.

(Neutron standards, V, '^^'^Cf , ^ 33 , 2 3 5u , f ) , 239 , 24 1

^ ^ ^ j

1. Introduction

Absolute values of v, the average number of

neutrons emitted per fission*, for the thermal neutron

fission of 233u, 235^^ 239py ^j^^ 241py j^^^g been

requested by reactor designers to accuracies approaching ^/h to For experimental reasons, the most

convenient way of performing such measurements is relative to a spontaneously fissioning standard and, for a number of years now, v for the spontaneous fission of

2 52cf j^gg been the accepted standard. Consequently, an equivalent or better level of accuracy is required for

this standard.

In view of the precision which has been requested,

the presence of a number of discrepancies in the

measurements over the last 10 years or so has been the cause of considerable concern. Historically, the

difficulties began following the publication of v for

2^2cf measured by the Boron Pile^). The value

obtained [3.713±0.015] differed from the average of the

two liquid scintillator determinations2 / ^) [3.79310.023]

by an amount uncomfortably larger than the comparative

error. Subsequently, preliminary values of v for 2 52(-.f obtained using manganese sulphate baths'* '^^ for the

neutron counting also supported a value lower than the

liquid scintillator average. Alternatively, indirect

values of v for 2 52^^ calculated using the 2200 ms"-^ constants (==3.78) were in agreement with the liquid

scintillator values. The 1969 IAEA review^' of the

2200 ms ^ constants ,• therefore , recommended compromise

figures in which both v and ri values were shifted

slightly from their experimental averages. For example,

the average experimental value of v for 2 52(-.£ q£

3.743±0.016 compared with the recommended value of

3.765±0.012.

However, as the precision of the MnSOi^ bath determinations of V for 252q£ improved, the support for the

lower values increased. A third liquid scintillator

measurement''^ obtained an intermediate value. Furthermore, it became apparent that some minor corrections

were required for the two early liquid scintillator

measurements^' . Thus the 1972 Neutron Standards

Reference Panel was able to conclude that within the

direct measurements the spread of the different values

was statistically acceptable. The 1972 Panel

recommended a value of 3.73310.008 for v for ^~'^Cf

.

The implications of this recommendation were either that

there existed some error in the other 2200 ms"'' parameters or, alternatively, the error assignments were

too optimistic. In particular, the n values used to provide indirect values of v were questioned. Subsequent revisions^ ' "^ ' ^ of two n measurements'' ^

' for 2 3 3y 2 3 5u have failed to attribute the

discrepancy to any specific factor. Although Leonard'^'

and Lemmel^^' have questioned the a measurements, the

entire explanation does not seem to lie here either.

*In this review, the symbol will be used for the

average prompt neutron emission, while v will be used

for the total neutron emission, including the delayed

neutrons

.

The unresolved nature of this discrepancy has

stimulated considerable activity in the revision of

previous v measurements . At least two new absolute

measurements are in progress^ ^

z' ^

^

, in addition to a re-examination^^' of the \! ratios, all of which aim to achieve high accuracy. In the present report, all

documented v measurements of reasonable accuracy have

been reviewed, including subsequent revisions to these

measurements and consideration is given to further

corrections that are now appropriate. In section 2,

absolute measurements of v for the spontaneous fission

of 2 52(-.£ gj-g examined, while section 3 is devoted to the V ratios. A set of recommended values is presented

in section 4.

2. Absolute V Measurements

Absolute V measurements may conveniently be subdivided into two categories

:

(a) Delayed coincidence experiments (e.g. liquid

scintillators and Boron Pile) in which a neutron counting gate is opened for a finite

time after each fission event. The technique

to a first approximation is independent of

the absolute fission rate.

(b) Direct measurements (e.g. manganese sulphate

baths) in which the absolute fission and

neutron emission rates are compared from

separate determinations.

Liquid Scintillator Measurements

The large liquid scintillator technique was first

developed for the measurement of v by Diven et al.20'

The neutron detector consists of a large liquid scintillator [lOO-1000 £ in volume] which is loaded with a high

neutron capture cross section material such as gadolinium or cadmium. A fission counter containing the

appropriate target is placed at the centre of a tube

which runs axially through the scintillator tank and

allows entry and exit of a neutron beam. Neutrons

produced by fission in this counter enter the scintillator, are moderated there and, after a mean lifetime

generally of the order of 10 ys, are captured by the

gadolinium or cadmium. The capture gamma rays so produced cause scintillations which may be observed by

photomultiplier tiibes mounted on the outside of the

scintillator tank. By this method a multiplicity of

neutrons produced in the fission event may be counted

individually. Excellent discrimination against background radiation can be obtained by gating the output of

the photomultiplier tubes with the fission pulse and

only counting scintillation pulses for several neutron

lifetimes

.

The neutron detection efficiency of the liquid

scintillator is proportional to the probability of

neutron capture in the hydrogen and gadolinium or

cadmixim loading (and the structural materials) , e^, and

to the probability of the subsequent detection of the

capture gamma rays, Ey The efficiency is a function

182
of the original neutron source energy (mostly because

of the variation in leakage) and, because of the

asymmetry introduced by the axial tube , it is also a function of the neutron emission angle with respect to the scintillator axis 9 . To determine the neutron

detection efficiency of the scintillator for ^^^Cf

spontaneous fission neutrons requires a knowledge of

e^(Ej^,e) and £^(£^^,6) for = 0-15 MeV. These functions cannot be determined entirely experimentally.

The normal procedure of calibration in the past has

been as follows:

(i) The probability of neutron capture in the

scintillator as a function of neutron energy

and emission angle e^(Ej^,e) has been calculated with a Monte Carlo code

.

(ii) It was assumed that is independent of Ej^ and e.

(iii) By measuring the probability of neutron

detection of neutrons of specific energy and

with a particular emission angle into the

scintillator, a value of was obtained.

(iv) The calibration of the scintillator efficiency; was checked by repeating the efficiency

measurement for a number of other values of

and e

.

Four sources of revision have been considered in

the present review.

(a) Fission Neutron Spectra

Because of the neutron energy dependence of the

scintillator tanks, it is necessary to make regular

revision of all measurements to take advantage of the

improvement in the precision of fission neutron spectra.

Unfortunately, recommended values for the fission

neutron spectra from Johansson^ ^

' were not available for

this review, which used the data listed in Table 1. For neutron fission of ^^^U and 2 3 9py ^j^g data were taken from the evaluation of Adams^^'. The spectrum

for neutron fission of ^^^U was taken to be Maxwellian

with an average energy 1.025 times that for ^^^U from

the review of Smith^^'. The neutron spectrum for

neutron fission of ^^^Pu was related to that for ^^^Pu

using the expression of Terrell^**' and v values from

the present paper. Because the experimental data for

^^^Cf show a wide distribution of values^^' , the

spectrum normally assumed, namely Maxwellian with

E = 2.15 MeV, has been retained.

TABLE 1 FISSION NEUTRON SPECTRA

(b) Delayed Gamma Rays from Fission

Because of the gamma ray sensitivity of the liquid

scintillators, the delayed gamma rays which arise from

the decay of isomeric states among the fission products

contribute to the apparent neutron count rate . Table 2 lists the delayed gamma ray cascades which can make a significant contribution. Some gamma ray yields have

not been obtained experimentally. Fortunately, in all cases (bracketed yields in Table 2) , it has been possible to estimate the missing yield using relative

fission fragment yields of the identified parent from

the experimental data of Unik et al.^^' Attention is

drawn, in particular, to the isomer with a half life of

162 ns , where the yield in the spontaneous fission of

^^^Cf is fairly large and the estimated yields for ^BSy

and '^•^^Pu show that it can make a serious contribution

to V measurements. This contribution can be minimised

in future measurements by introducing a delay of at least 500 ns before initiating neutron counting.

No experimental data are available for neutron

fission of either ^33^ 241py_ 233(j^ j have used

the 235u yield data and for ^'tlpu, the 239py data.

TABLE 2 DELAYED GAMMA RAY DATA

Tj^ Energy of Parent % Yield per Fission

Cascade

(ys) (MeV) 252cf(sp) 235u 239pu

0 020 0 614 A=100 0 34 (0 93) (0 85)

0 100 1 723 A=134 0 22 (0 51) (0 49)

0 162 1 692 1 15 (2 7) (2 5)

0 62 1 505 A=135 0 28 (0 51) (0 51)

3 1 1 891 A=137

or ^3^Xe

0 60 0 63 1 30

26 7 1 710 0 38 0 45 0 73

54 0 1 110 93Rb 0 50 0 85 0 72

80 0 1 710 0 64 0 32 0 46

Data taken from refs. 7,25-28)

(c) Improved Monte Carlo Calculations

All liquid scintillation measurements of v are dependent on the accuracy of the calculation of the

neutron leakage. Improved input data, and particularly

the ability to specify more exactly the geometry of the

detector systems, have afforded some recent improvement

Reaction Shape T

(MeV)

A

(MeV)

B

(MeV)

233u(n,f) Maxwellian 1.37

235u(n,f) Watt 0.9878 2.1893

239pu(n,f) Watt 0.9723 2.7005

2'*lpu(n,f

)

Maxwellian 1.40

252cf (s,p) Maxwellian 1.43

Watt spectrum: P (E) = C exp (-E/A) sinh/BE

Maxwellian: P (E) = Ci^E exp (-E/T)

(d) Dependence of the Capture Gamma Detection

Efficiency on E^^ and 9

It has nojrmally been assumed that the capture

detection efficiency is independent of the original

properties of the neutron. This assumption was questioned at the 1972 Panel Meeting. Since then,

Poitou and Signarbieux^'') have investigated the validity

of this assumption by also following the history of the

capture gamma rays in a Monte Carlo calculation. They

find a small variation in neutron capture detection

efficiency with neutron energy which depends on the

size of the scintillator. More specific calculations

have been made by Ullo^^' with particular reference to the liquid scintillator measurement^'

.

Particular application of these four sources to individual liquid scintillator measurements follows.

183
S£e -18)

A new absolute determination is in progress at Oak

Ridge National Laboratory. No details or data are yet

available

.

Boldeman^

The liquid scintillator in this measurement was

spherical, 76 cm in diameter, and contained a loading

of 0.5% by weight gadolinium. The axial hole through

the scintillator had a diameter of 7.62 cm. The

absolute calibration of the system was made as follows.

(a) The relative neutron capture efficiency of

the scintillator was calculated as a function

of neutron energy and angle of emission with

respect to the beam tube, using a Monte Carlo

method

.

(b) The calculated efficiency values were normalised at the low energy end of the

fission neutron spectrum in a measurement in

which 2 MeV neutrons were scattered from a hydrogen gas target. Essentially, the

normalisation was based on the detection

efficiency for 0-1 MeV neutrons emitted at angles of 90-45 respectively, with respect

to the scintillator axis.

(c) It was assumed that there was no energy

dependence of the neutron capture gamma ray

detection efficiency.

(d) The shape and absolute calibration of the

energy dependence of the efficiency curve were checked in a second experiment in which

16 MeV neutrons were scattered from a polythene target. The detection efficiency of

the system was verified within 1.5% to

8.73 MeV.

In the original work, the fission neutron spectrum

assumed for 2 52Qf same as the present and

therefore no correction is necessary. The effects of

delayed gamma rays from fission were also adequately

taken into account.

An extensive re-analysis of items (c) and (d) has

been carried out by Ullo^'^). Most of the structural

details of the scintillator were included in a Monte

Carlo calculation of the neutron energy dependence of

the scintillator, in which the histories of the capture

gamma rays were also followed. The variation in leakage with neutron energy and angle from the original

calculation was confirmed. However, an energy

dependence for the probability of neutron capture

detection was determined.

The relative probability of neutron capture

detection for isotropic ^^^Cf spontaneous fission

neutrons to that for isotropic 0-1 MeV neutrons was computed to be 0.9972. Furthermore, the relative

probability of capture detection of 0-1 MeV isotropic

neutrons to those of the calibration, 0-1 MeV between

90° and 45°, was estimated to be 0.9997. Thus this

work estimates a total difference of 0.31% between the

calibration neutrons and those for ^^^Cf , compared with

a +0.10% correction in the original paper. For the

present paper, the analysis of Ullo^^ has been

accepted and a further correction of +0.21% has been

applied to the measured value.

A further modification suggested by Ullo concerns

the reliability of the estimate of the effect on neutron leakage caused by the axial hole. His somewhat

pessimistic contribution of 0.3% to the experimental

error compares with a value of 0.1% used in the

original paper. A compromise figure of 0.2% has been

used in this review.

Table 3 lists the revised corrections for this

measurement and all identified sources of error. The

final value obtained for the prompt neutron emission

is 3.746±0.016 and 3.755±0.016 for the total neutron

emission after the addition of the delayed neutron

fraction.

TABLE 3 CORRECTIONS TO EXPERIMENTAL DATA

AND SOURCES OF ERROR FOR REF. 7)

Effect Correction

(%)

Contribution

to Accuracy

{%)

Statistical accuracy

Dead-time correction:

(a) 252cf

0.24

+1.107

(b) Low energy proton +0.279

recoil

(c) Relative +0.828

Delayed gamma rays -0.28

Fission neutron spectra:

(a) Accuracy of E

(b) Accuracy of energy

calibration

French Effect -0.10

Effect of hole through

scintillator on neutron

leakage

Background error in -0.33

proton recoil counter

0.10

0.07

0. 12

0.17

0.10

0.20 (0.10)'

0.10

Variation in neutron

capture detector

efficiency for ^^^Cf

fission neutron and

calibration neutrons

+0.31 0.10

+0.05 0.05

b)

a)

b)

Value used in original paper

Two separate effects considered in original paper

Asplund-Nilsson et al 2)

In this experiment the liquid scintillator was

60 cm in diameter with a 6 cm diameter axial hole.

The tank contained 110 SL of liquid scintillator with a Cd/H atom ratio of 0.002. The absolute calibration of

the neutron detector efficiency was carried out as follows

:

(i) For four incident energies (3.0, 4.3, 4.5 and

14.9 MeV) neutrons were scattered from an anthracene crystal at the centre of the

scintillator and the neutron detection

probability was determined for 28 values of

(Ejj,e) . The highest energy for which the

efficiency was determined was 10.75 MeV.

(ii) It was assumed that there was no angular

dependence for the neutron detection

efficiency. Thus the detection efficiency

for a ^^^Cf fission neutron spectrum

(T = 1.40 MeV assumed) was obtained by

integrating over the measured energy

dependence

.

184
(iii) The efficiency figure so obtained was corrected for the effect of the axial hole

using a Monte Carlo calculation.

This measurement was down-weighted in the

review by Axton^' partly because assumption (ii) was shown to be inappropriate. However, because the

neutron detection efficiency was measured for such a large number of neutron energies and emission angles,

it seemed reasonable to try to correct this deficiency

in the original analysis.

For the exact scintillator geometry, the probability of neutron capture in both hydrogen and cadmium

was calculated using the code of ref. 7) to be 0.9035

for an isotropic source of ^^^Cf spontaneous fission

neutrons [e = 2.15 Mev] . For each of the 28 efficiency

measurements at a specific neutron energy and emission

angle the probability of capture was also determined.

Comparison of the experimental data with the calculated

leakage, yields 28 values for the probability of neutron

capture detection. The relevant data are listed in

Table 4.

TABLE 4 CALIBRATION DATA FROM ASPLUND-NILSSON ET AL 2)

Scattered Scattered Detection Calculated Effective

Neutron

Energy

(MeV)

Angle

(degrees)

Probability

(%)

Capture

(%)

e Y

(%)

0 . 24 1

6

. b

"7 O bi J. c yy . ZD 73 .

1

0 . 55 c o DZ . i 1

A

/4 "7-1- T Z y o . y z 75 .

5

1 . 07 53 . 3 74 8+1 1 97 . 75 76 .

5

0 . 15 79 .5 70 9±1 5 99 .46 71. 3 0.50 70 .5 72 9±1 4 98 .97 73.7

0.97 62 .3 73 1±1 5 98 .34 74. 3 1.40 56 .1 73 4±1 5 96 .79 75.8

1.86 50 0 71 5+1 4 94 .01 76.0

2.43 42 .7 72 0±1 4 90 .18 79.8

2.75 38 6 67 3±1 2 87 .30 76.9

2.93 36 2 67 7±1 4 88 .15 76.8

0.21 11 2 73 2±1 6 99 .27 73.7

0.58 68 5 73 9±1 6 99 .00 74.6

1.00 61 2 72 7±1 7 98 19 74.0

1.82 49 4 72 1±1 6 94 79 76.1

3.17 30 8 65 4±2 0 83 .75 78.1

0.48 79 7 71 6±1 7 99 .21 72.2

1.65 70 6 72 3±1 7 96 .02 75.3

1.79 69 7 72 8±1 8 95 48 76.2

2.18 67 5 70 2±1 6 93 .22 75.3

2.49 65 .9 72 4+1 8 91 49 79.1

3.01 63 .3 69 9±1 7 88 . 86 78.7

3.46 61 2 66 8±2 1 88 . 47 75.5

4.27 57 6 62 9±1 8 84 .51 74.4

5.42 52 .8 57 5±1 7 74 .47 77.2

6.69 47 .9 51 5±1 8 68 .04 75.7

7.94 43 .1 51 3±1 7 66 .67 76.9

10.75 31 .9 40 1±1 6 48 .83 82.1

It will be noted that the capture detection

probability is fairly constant, although there is a suggestion of a small drop at low neutron energies,

while the value at 10.75 MeV seems slightly large. An

average value for the capture detection efficiency was obtained by weighting the numbers in Table 4 according

to a fission neutron spectrum. The value obtained,

0.7595, when combined with the probability of neutron

capture gives a value of 0.6862 for the overall

neutron detection efficiency.

Since the publication of the original paper, it

has been reported^^' that a correction of -0.6±0.3% is

necessary in this experiment to account for the French

Effect^^). This is the bias that can be introduced

into V measurements by the use of a coincidence between

the fission counter pulse and a scintillator pulse from

fission gamma ray detection and neutron induced proton

recoils, to initiate the neutron counting gate. There

are many experimental advantages in utilising this

coincidence in liquid scintillator measurements. It was assumed that those fissions without a coincident

pulse had the same average neutron emission as those

with a coincident pulse. In the original identification of this problem, the data suggested a possible

correction of the order of 1% or so. Subsequent work

has shown the French Effect to be very small and it is

only in this experiment that the effect is really

significant.

No correction was applied in the original paper

for the contribution to the neutron count rate of the

delayed gamma rays from fission. This was estimated

for the 1972 Panel Meeting to be -0 . 20±0 . 20% ' . This

correction has been re-evaluated. The threshold on the

neutron detector was set at the equivalent of approximately 600 keV from gamma ray sources. Furthermore,

the neutron counting gate began 160 ns after fission.

With these parameters, the revised correction is

-0.43±0.2%.

The final revised value from this experiment for

the prompt neutron emission is 3.783±0.040 and

3.792±0.040 for the total neutron emission. A full

list of the experimental errors is given in Table 5.

TABLE 5 SOURCES OF ERROR ( ASPLUND-NILSSON ET AL.2) )

Cause Error

Statistical accuracy (^^^Cf) 0 15

Statistical accuracy calibration 0 30

Corrections for proton recoil 0 20

Dead time correction 0 30 Delayed gamma rays 0 20

Fission neutron spectra (+0.03 in E) 0 15

Accuracy of energy calibration 0 75

French Effect 0 30

Effect of hole through scintillator 0 30

Variation in neutron capture 0 20

detection efficiency

Hopkins and Diven ^^

This experiment used a cylindrical liquid

scintillator, 1 m long and 1 m diameter. The 1000 £ of

liquid scintillator had a cadmium to hydrogen atom

ratio of 0.002. For the absolute calibration of this

detector, 3.9 MeV neutrons were scattered in a plastic

185
scintillator located at the centre of the 7 cm diameter

axial tube. Neutrons within the energy range 0-1.3 MeV

at an average emission angle of 72° were selected for

the actual calibration by the bias on the plastic

scintillator. The energy dependence of the neutron

leakage was based, as before, on a Monte Carlo calculation. In a second experiment to confirm the calculated

energy dependence, 14.5 MeV neutrons were scattered in

the plastic scintillator. The efficiency scale was confirmed for two energy groups, 0-2 MeV and 6-8 MeV.

A number of queries regarding the accuracy of the

calculated leakage probabilities were raised by Axton^^

at the 1972 Panel Meeting. The authors subsequently

confirmed the relative accuracy of their leakage

calculation. Prior to the present paper, the leakage

calculations were checked using the Monte Carlo code of

ref . 7) . Our calculations would require the v value

from this experiment to be adjusted downwards by 0.15%.

This adjustment has not been made, as our calculations

included none of the structural details of the

scintillator.

The original value of 3.771±0.031 was obtained

using Bonner's^^' measurement of the ^^^Cf fission

neutron spectrum. An adjustment of +0.11% is required

to correct the efficiency for the assumed fission

neutron spectrum of Table 1.

No correction was applied in the original work for

the effect of delayed gamma rays from fission. The

threshold was set at approximately 1 MeV and it would

appear that the neutron gate began approximately 300 ns after fission. Thus, relative to experiment 7), the

efficiency for detection of delayed gamma rays compared

to that for neutron capture gamma rays will be smaller.

On the other hand, the earlier neutron gate increased

the contribution. A reasonable estimate would appear

to be 0.2±0.1%.

idealised to a homogeneous sphere of graphite, •''^B and

aluminium, with aluminium being used to simulate the

copper bodies of the BF3 counters. The calculated

energy dependence was normalised using four measurements in which the absolute detection efficiency was obtained for four different essentially monoenergetic

sources. For the Boron Pile the reaction used to produce the neutrons was the photo-disintegration of the

deuteron. Four different product neutron energies were obtained using four different gamma ray energies. The

data are listed in Table 6. It will be noted that the

calculated efficiency curve does not reproduce particularly well the variation of the calibration values.

There are unfortunately no data on which to base

a correction for the variation in neutron capture

detection between ^^^Cf. spontaneous fission neutrons

and those used in the calibration (0-1.3 MeV neutrons).

Poitou and Signarbieux^'" indicate that the variation

of gamma ray detection with neutron energy is flatter

for a 50 cm radius scintillator than for the 38 cm radius scintillator in experiment 7). Therefore, the

effect should be smaller than the total effect of

+0.31% calculated by Ullo^-^' for experiment 7). Of

course, the higher threshold in experiment 3) increases

the effect relative to 7) , although this will be

counter-balanced by the higher gamma ray energy emitted

in cadmium capture versus that in gadolinium. In fact,

if all the gamma ray energy is deposited in the

scintillator, it is possible to conceive of a negative

magnitude for this effect. Because of these considerations, it has been decided to make no correction at all.

However, a contribution of 0.1% has been included in

the experimental error of v for the spontaneous

fission of ^^^Cf

.

The final value obtained from the revision of this

experiment is 3.768±0.031 or 3.777+0.031 for the total

neutron emission.

The Boron Pile (Colvin and Sowerby^')

In this experiment, the neutron detector was a 220 cm cube of graphite surrounded by a 35 cm thick

reflector of graphite. The pile contained a lattice

of 240 BF3 counters to detect the thermalised neutrons.

The energy dependence of the leakage from the pile

and therefore the relative efficiency of the system, was calculated by Pendlebury ' using the Carlson Sj^ technique (Fig. 1) . For the calculation, the pile was

If

01 23456789 10

NEUTRON ENERGY (MeV)

Fig. 1: Efficiency calculations for the Boron Pile

TABLE 6 CALIBRATION MEASUREMENTS FOR BORON PILE^'

Neutron EfficMeasurement Energy

(MeV)

iency Error

(%) D(Y,n)p with Th C" Y-rays 0.19 0 6440 0.60

D(Y,n)p with ^'*Na Y~rays 0.265 0 5457 0.21

D(Y,n)p with '^F (p,a,Y) Y-rays 2.0 0 6433 0.25

D(Y,n)p with ^'^Al (p,Y) Y -rays 4.9 0 6500 1.24

Ra-Y-Be' E0.25 0 6475 0.74

*Source calibrated by N.P.L. MnSOi^ bath^^^

Because the v value obtained, 3.713±0.015, was so different from existing measurements, the Boron Pile was subjected to close scrutiny. Colvin et al.^''' have

answered a number of criticisms that have been made.

Furthermore, a number of independent checks of the

efficiency of the Boron Pile were made. For example, an independent measurement was made using a Ra-Y-Be source calibrated by N.P.L. using a MnSOi^ bath^^) . This value,

effectively for neutrons at 250 keV, has been included

in Table 6. Some recent objections have been voiced by
Leonard^ ^'

.

(a) The error assigned to the correction for the

proportion of neutrons detected outside the

gate period should be increased from the

original assignment of 0.1%. The proportion

of neutrons outside the 4 ms gate period is

given by Colvin et al.^^' as 4.3±0.3%. From

the data in Table V of ref . 37) , the correc- tion would appear to be at least as accurate

as this. Thus, for the comparison of the

Ra-y-Be source measurement of the neutron

detection efficiency with that from the

D(Y/n) reaction, this correction must be

applied and the full error in the correction

included in the list of errors. However, for

the comparison of ^''^Cf fission neutrons and

D(Y,n) reaction neutrons, it is the relative

difference in the correction which is

important. It is difficult to see how this

could be any larger than the error of 0.1%

assigned for this effect. It should be

noted that a similar effect exists in all

liquid scintillator measurements (1-2% loss

outside the gate). In these cases, no contribution to the experimental error has

been included.

(b) Loss of neutrons in the Boron Pile caused by

Cu(n,p), Cu{n,a) and C(n,a) reactions was neglected in the efficiency calculations.

Certainly the first two reactions were

ignored; however Sowerby^^' estimates their

effect as less than 0.1%. The effect of the

third reaction has also been estimated by

Sowerby to be 0.15±0.05%. It is thought

that the effect of this reaction was included

in the original calculations; however it has

not been possible to verify this.

(c) Because the measured values of the efficiency

calculations do not fit the calculated energy

dependence, some subjective judgement is

involved in choosing the best method of

normalisation of the curve. In fact, it is

fairly obvious that a recalculation of the

energy dependence of the neutron detection

efficiency of the Boron Pile is required,

especially when the age of the original calculation is considered. A new calculation

would also eliminate any objection on account

of (b) above.

For the 1972 Panel Meeting, Axton^' recalculated

the energy dependence of the Boron Pile using a Monte

Carlo calculation. His calculated efficiency curve is

shown in Fig. 1, together with the original calculation.

Use of this efficiency curve, rather than the Pendlebury calculation, leads to a value for the detection

efficiency of the Boron Pile for ^^^Cf spontaneous

fission neutrons of 0.64145 versus the original value

of 0.6428. Also shown in Fig. 1 is a more recent Monte

Carlo calculation of the relative neutron energy

dependence of the Boron Pile neutron detection efficiency by Ullo'*"' . This work has not yet been published,

but some details are available. For neutron transport

the energy range from 5 kev to 10 MeV was spanned by

3000 energy points and inelastic scattering was handled

implicitly. The angular distribution for elastic

scattering in carbon was described with four Legendre

components in the centre of mass. The cross sections

for Cu(n,p), Cu(n,a) and C(n,a) were included in the

calculation. Finally, some structural details of the

Boron Pile were included. The calculated efficiency

curve is also shown in Fig. 1.

There is a most unfortunate lack of agreement

between the three different calculations. The two

Monte Carlo calculations give the most similar

dependence. In fact, it is likely that the difference

between them can be partly attributed to slightly

incomplete data sets employed in the calculation of

Axton ' . The difference between the original efficiency

curve and that of Ullo is quite marked. The latter

calculation does not show the rapid rise in efficiency

at low energies , but shows a much more dramatic fall in

efficiency at high neutron energies than does the

Pendlebury calculation. The overall effect of the use

of the Ullo efficiency curve is to increase the measured

value of V. For this review the Ullo efficiency curve has been used for two reasons. Firstly, the input data

set was considerably more comprehensive. Secondly, in

the calculation of the energy dependence of the liquid

scintillator of ref. 7) the Ullo code produced values

in substantial agreement with two other Monte Carlo

calculations'^ ' and in agreement with the experimental

data. However, it is important to resolve this question

of the energy dependence of the Boron Pile as soon as possible

.

For the normalisation of the calculated efficiency

curve, a weighted value has been obtained from the

experimental values at 0.190, 0.265, 2.0 and 4.9 MeV

from the D(y,n) measurements, plus the value at 0.250

MeV from the Ra-y-Be source measurement. The weighted

efficiency obtained for a ^^^Cf fission neutron spectrum

is 0 . 6379±0 . 0010 . The error here is the direct experimental error. Because of the failure of the experimental data to reproduce satisfactorily the calculated

energy dependence of the efficiency curve, some contribution to the experimental error is necessary.

Unfortunately, this contribution is somewhat subjective.

A value of ±0.3% has been used which compares with an effective value of 0.25% used by Colvin and Sowerby^'.

Colvin and Sowerby have indicated that they are prepared to accept, provisionally, the recalculated

efficiency curve for the Boron Pile, pending an examination of the details of the calculation. The revised

(provisional) value of Vp for ^^^Cf from the Boron Pile

is therefore 3.73210.016 and 3.741±0.016 for the total

neutron emission. All contributions to the experimental

error are listed in Table 7.

TABLE 7

SOURCES OF ERROR (BORON PILE^^ )

Contribution

Cause to Experimental Error

(%)

Statistical accuracy of calibration^^ 0.15

Statistical accuracy^' 2 52cf g.ll

Anisotropy in pile efficiency 0.20

Beam modulation 0.05

Electrical pick-up 0.03

Neutrons after gate 0.10

Fission spectrum (error in E) 0.05

Lack of agreement of calculated

and experiment shape of energy 0.30

dependence of efficiency

Includes contribution from error in correction

for loss of neutrons caused by chamber materials

187
One objection that might be levelled at any

revision of the Boron Pile v value for ^^^Cf is that

the old value was in agreement with values obtained

using calibrated neutron sources to determine the

efficiency of the pile. Colvin et al.^^' lists these

measurements in Table II of that reference. Three

different neutron sources were employed, namely, the

AERE and AWRE ^'tOp^ sources and a Ra-y-Be source. The

latter source, which was calibrated using the N.P.L.

MnSOtf bath, has already been included in the absolute

normalisation of the efficiency scale.

2kO

For the calibration of the Boron Pile using the

Pu sources, it is necessary to make a correction

for the energy difference between ^-'^Cf spontaneous

fission neutrons and those for spontaneous fission of

^'*''Pu if the efficiency curve of Ullo is appropriate.

A correction of 0.25% is required in this case. For

these source calibrations three independent efficiency

values can be derived for the Boron Pile based on three

independent methods of source calibration, namely,

(a) the revised original calibration"*^' of the

AERE 21*0 Pu source

;

(b) the AWRE oil bath'*2) calibration of the AWRE

^'*''pu source;

(c) the calibration of the AWRE

using the N.P.L. MnSOi, bath.

2hO Pu source

After applying the 0.25% correction referred to above, the efficiency of the Boron Pile for ^^^Cf

fission neutrons derived using the source calibrations

(a) and (b) are

°-^41^^n'mn! and 0.6396-°-°°°^

±0.0104 systematic ±0.0087 systematic

The average of these two values leads to a value of v for 252cf 3.726±0.039. The error was derived assuming the original errors were independent. This is

probably not the case, and the actual error may be

fractionally larger. It is evident that the value of

V derived using these two source calibrations is con- sistent with the revised v value from the Boron Pile

experiment itself.

Finally, the value of v for 252 Cf derived using

the N.P.L. calibrated value for the AWRE 2'*0pu source

is 3.697±0.034. The difference between this value and

the revised Boron Pile value (slightly larger than one standard deviation) is not sufficient to raise any

objections to the present revision. This third source value is more appropriately incorporated with the

N.P.L. band measurements.

Manganese Sulphate Bath Measurements

The absolute measurement of neutron source strengths with manganese sulphate baths will be discussed in considerable detail at this conference by

Axton43) Consequently, few of the details of the

actual neutron counting will be given here and more attention will be devoted to the fission counting.

Bozorqmanesh 17)

A new MnSOii bath determination of v for the spont- aneous fission of ^^^Cf has apparently been completed

at the University of Michigan. No details have been

published as yet. The reported value, 3.744±0.023 for

the total neutron emission, has been accepted for this

review, together with the nominated error.

Me asurements based on N.P.L. MnSOit bath ^ , 8

, it , 4 5

)

in the N.P.L. MnSOit bath, but the method of fission

counting varied. Each is discussed separately below.

Axton et al . ^

, 8

)

In this experiment a series of thin ^^^Cf spontaneous fission sources were aliquotted from a stoclc solution of californium chloride and the absolute fission

rates were counted in a pill box type gas flow proportional counter. The method employed to provide a satisfactory correction for fragment losses in the foils was discussed in detail in ref . 5) . The neutron

emission rate of the remaining californium chloride

solution was then determined in the N.P.L. MnSOi, bath.

This process was repeated a nuinber of times for four

separate californium samples. A total of 20 separate

neutron sources and 200 fission sources were prepared.

The final value for the total neutron emission obtained

in this experiment was 3.725±0.019 where the experimental error includes only the error on the fission

counting

.

White and Axton**^'

For this determination, the fission and neutron

emission rates were obtained for the same source. The

neutron emission rate was measured in the N.P.L. MnSOi^

bath and the fission rate was determined using a low

geometry fission counter at Harwell. At the 1972 Panel

Meeting, Axton^' revised the accuracy of the fission

counting. For the present review the revised value is

retained. The value for v obtained was 3.797±0.038

where the error, as before, includes only that for the

fission counting.

Other N.P.L. MnS0i4 bath dependent measurements

Two separate absolute determinations^''*^' of ^ for

^^^Cf have been made in which the neutrons were counted

188

A number of other measurements of v for ^^^Cf are generally included in the list of N.P.L. dependent

measurements. One of them was the calibration of the

Boron Pile with a Ra-y-Be source and the AWRE 2'*0py source, both of whose activities were calibrated in the

N.P.L. MnSOii bath. In the present report, the first has

been included as part of the absolute calibration of the

efficiency curve for the Boron Pile. Only the second

source measurement, revised previously to

3.697±0.034, is included here with the N.P.L. dependent

measurements

.

Another addition was a measurement from Moat et

al."*^' in which the neutron emission rate of a californium sample was determined using two cylindrical

wax detectors of different dimensions, each containing

several BF3 counters for the detection of thermalised

neutrons. The efficiency of the detector was determined

using the Harwell ^'^''Pu source which was calibrated

using the N.P.L. MnSOi^ bath, the Boron Pile and the

Aldermaston oil bath. The original value of 3.77±0.07

for v for 2^^Cf was corrected by Fieldhouse et al."*^' to 3.685±0.040 in a revision of the Harwell 240 Pu source and then to 3.727+0.056 by Hanna et al.^' to account for an improved estimate of the difference in

the 2'*°Pu and ^^^Cf fission neutron spectra. Because

of the very strong dependence of this measurement on

the "softened" _^'*''pu spontaneous fission neutron

spectrum, it has not been included in the final list of

values

.

Average value from N.P.L .

The average of the three N.P.L. dependent measurements is 3.731±0.015 for v for ^^^Cf. The error here

includes only that from the fission counting. The error for the neutron source calibration in the N.P.L.

bath was given by Axton^' as 0.33%. Thus the total error becomes ±0.020^
A small revision to the N.P.L. value of v for

2 52cf has been recomitiended by Smith et al.^^' following their accurate measurement of the Ojj/Oj^ ratio.

They used the variable manganese concentration technique

developed by Axton et al.'*'*', but for the neutrons

employed a Bragg reflected beam of approximately

0.02 eV. Thus corrections required for the ^^0(n,a)^^C

and ^2s(n,p)32p ^^.^ eliminated, together with any

dependence on resonance capture data for ^^Mn. Recent

experimental work''^^^ can be used to predict considerable

intermediate structure in the ^^Mn(n,Y) cross section

and specifically a strong valence component which could

lead to a radiative width for the 2.375 keV resonance

as large as 2 eV. The value they obtained,

0.02503±0.23%, compares with the revised value from

Axton et al.'*'*' of 0.0249510.35%. Use of the Smith

et al.'*^ value would lead to a positive adjustment

of 0.15% in the N.P.L. v value.

For the purposes of this review, the N.P.L. v value has provisionally been adjusted by +0.15% to 3.737±0.020.

De Volpi and Forges '*''*^ '

For this experiment the experimental procedure was

as follows. Three separate fission counters were prepared in which the californium sources, all yielding

approximately 10^ neutrons/s, were deposited in three

different ways on one electrode of a fast ionisation

chamber with a plate spacing of 1-2 mm and operated in

the current mode. The neutron emission rates were measured by inserting the fission chambers in the A.N.L.

manganese sulphate bath. Subsequently, the absolute

fission rates for each of the three fission counters

were determined in a coincidence system in which the

fission fragments were counted in coincidence with

prompt fission neutron detection in a Hornyak button^'''.

The coincidences were recorded as a function of the

angle (6) between the fission counter normal and the

neutron detection axis. The fission rate was given by

NF/C where N is the neutron count rate, F is the fission

rate and C the coincidence rate. The angular anisotropy

of the effective fission rate was removed by averaging

the four measurements at 6 = ±37.4 and ±79.2. Although

three different counters were employed, the final value

of V was essentially based upon one counter (parallel

plate) which had a fission fragment detection efficiency

>99%. In addition to the coincidence method, the

fission rate was also measured in a variety of other

.ways, including low geometry counting and simple extrapolation of the bias curve. Despite the considerable

care that has been taken in arriving at the absolute

fission rates, the estimated error for the third fission

counter, 0.13%, appears optimistic. From Table 6 of

ref. 4), the two sources of error for fission counter 3,

namely the error of 0.13 for counter 3 in particular and

that for the error common to the three detectors , appear

not to have been added. Furthermore, two of the entries

in the common errors, those for neutron detection

efficiency dependence on v and fission detection

efficiency dependence on v, could each be increased to 0.1%. In addition, there does not appear to be any

component added for the distribution in nodal values

listed in Table III of ref. 4). The error has provisionally been expanded to 0.23%. This compares to an effective fission counting error in the summed N.P.L.

measurements of 0.42%.

Axton^' in his 1972 review recommended the introduction of an additional 0.5% error into the neutron

source calibration in the MnSOi^ bath to account for

aliquotting errors. De Volpi, in a private communication to the 1972 Panel Meeting, answered criticisms on

this score, and the 0.5% error has therefore not been

included in the list of errors in this review. However,

from the text of ref. 8) and De Volpi ' s reply, there

appears to be a number of small unresolved differences

between A.N.L. and N.P.L. Finally, it should be noted

that the measured value of o^/a^^ from A.N.L. of

0.253110.00003 differs significantly from the measurement of 0.02503±0. 00006 by Smith et al.'*'''. The

explanation proposed by the authors'* to explain their

high value, impurities in the solution, could introduce

an unknown correction as would any densimetric error.

No error has been introduced here on this account. The

final value obtained using the A.N.L. MnSOi^ bath is

therefore 3.729±0.017.

Summary of Absolute v Measurements for ^^^Cf

The revised values of v for the spontaneous fission

of ^^^Cf from the eight measurements are listed in

Table 8. The average of these measurements, weighted

according to the experimental error, is 3.74510.008.

The errors are not all independent and the accuracy

should be expanded to ±0.010 to include various common

errors. This revised value represents an increase of

0.32% in the value recommended by the 1972 Panel

Meeting.

TABLE 8 RE-EVALUATED VALUES OF V FOR ^^^Cf

Total Neutron Total Neutron

Measurement Emission Emission

Original Revised

Liquid Scintillator

Measurements

:

7 ) Boldeman ' 3.. 747±0..015 3..75510..016

Asplund-Nilsson et al.^' 3..808±0..034 3..79210..040

Hopkins and Diven^' 3..78010,.031 3..77710..031

Boron Pile:

Colvin and Sowerby^^ 3..71310,.015 3..741±0..016

Fieldhouse et al.**-^' 3..72610..039

Manganese Sulphate Baths:

I 7 \ Bozorgmanesh ' 3.,744+0..023

Axton et al.^'"*^' 3..728+0..019 3..73710..020

De Volpi and Forges'*^ '

' 3..72910..015 3.,729±0..017

The most important question to decide is whether

the weighted error represents the true error of the

summed eight measurements, or whether there is any

evidence within the final numbers for a possible

experiment-dependent systematic error. Of the eight

measurements, only that from Asplund-Nilsson et al.^^

lies more than one standard deviation from the mean and

certainly the external error of ±0.005 is less than the

internal error. The weighted average of the three

liquid scintillator measurements is 3.763±0.014 which

is approximately one and a half standard deviations of

the comparison displaced from the average of the three

manganese sulphate bath measurements [3.73510.011]

.

Alternatively, the average of the four gated measurements, 3.75410.010, compares satisfactorily with the

four source measurements. Thus any suggestion in the

original data of an experiment dependent systematic error has now disappeared. Therefore it is considered

that the error of 10.010 represents a legitimate

estimate of the accuracy of the standard.

The recommended value for the total neutron emission

from the spontaneous fission of ^^^Cf is 3.74510.010

189
3 . The V Ratios

Most previous reviews of the ratios of neutron

emission for neutron fission of ^^^U, ^^^U, 239pu ^^^(j 2'*lpu to that for the standard, spontaneous fission of

^^^Cf, e.g. refs. 6 and 16) have concluded that the

different measurements were in satisfactory agreement.

However, it will always be necessary to effect those

small revisions that become apparent with the general

improvement in nuclear data.

The principal revisions made in this paper

include those for improvements in the measured fission

neutron spectra (Table 1) and in the delayed gamma ray

from fission data (Table 2) . The correction, considered for the standard, to account for variation in the

neutron capture detection efficiency with emitted

neutron source energy can be safely ignored as it will

be at least an order of magnitude smaller here (i.e.

<0.01%)

.

A new correction that must be considered for all

of the V ratio measurements arises from the resonance dependence of v. None of the measurements that are generally included in the evaluation of the v ratios were made with monoenergetic neutrons and thus the

measured value has some small sensitivity to the exact

shape of the incident neutron spectrum. Since the

existing experimental data on the resonance dependence

of V is neither complete fpr the four important

isotopes nor always consistent, a brief review is

included here of the data and the likely character of

the effects which are contributing to the variation.

Resonance Dependence of v

Work on the resonance dependence of v was stimulated by the conflicting evidence presented at the

second IAEA Symposixim on the Physics and Chemistry of

Fission. For the neutron fission of ^^^Pu, Weinstein

et al.^"^' found that v values for 20 resolved resonances below 100 eV fell into two groups strongly correlated

with the resonance spin. The average value of v for resonances with spin J = 0 was 3% higher than the

average for resonances with spin J = 1. A smaller resonance effect was noted for ^^^U. Ryabov et al.^^'

observed the opposite effect. For resonances in 239py

they found v for J = 1 resonances was 5% larger than

that for J = 0 resonances. A subsequent experiment by

Weston and Todd^**^ confirmed neither experiment. They

found that v for the two spin classes was similar

within V't%- However, they observed a fluctuation in

the different v values outside the statistical accuracy

of the measurements. Two later measurements^ ^

'

' in

substantial agreement with ref. 54) gave rise to an explanation^ of most of the fluctuations in v values

in terms of the (n,Yf) process. Essentially, they

found that the v values for the ^^^Pu resonances were inversely correlated with the fission widths. Fairly

clearly, V is smaller for the (n,Yf) reaction than for

direct fission. Furthermore, because the (n,Yf) process

is a multichannel process, its width is fairly constant.

Thus V for resonances with small fission widths should

also be small because of the increased relative

importance in these cases of the (n,Yf) process.

'

The relevance of the resonance energy dependence

of V to the evaluation of the thermal value will be

apparent from Fig. 2 from Leonard^ ^' where the v energy

dependence near thermal neutron energies has been

plotted from the data of refs. 52,59-61). The resonance

at 0 . 296 eV has a considerable impact on the energy

dependence. The value of v for this resonance is

approximately 0.029 neutrons less than that for thermal

fission. From the data in ref. 56) , 0.011 of this

depression can be expected because of the (n,Yf) process.

The remainder, 0.018, is similar in magnitude to the

difference, 0.014+0.007, found by Frehaut and

Shackleton^^' for the difference in v for J = o""" and

J = 1 resonances with the effects of the (n,Yf)

process removed.

Two effects can be expected to contribute to this

difference. Cowan et al.^^' have shown that the

symmetric fission yields for J = 1"*" resonances are about

1/3 those for J = 0"*". Since symmetric fission fragments

emit more neutrons^'*^ , v for J = O"*" resonances should be

larger than that for J = l"*". However, this effect is

likely to be extremely small. For ^^^U, Howe et al.^^^

have found no correlation between resonance v values and

the measured mass asymmetry^ '

) . A second contribution

arises because of the considerable difference in the

fission barrier for J = O"*" compound states relative to that for J = 1"*" states . It has been shown that the

fine structure in the Vp(En) dependence between 0-1 MeV

for neutron fission of 2 2^u^'*^can be explained if the

collective energy at the fission saddle point is weakly

coupled to the nuclear degrees of freedom at scission,

and appears predominantly in the fission fragment

kinetic energies. The fission barrier for J =

1"*^ com- pound states (Kit = l"*") is displaced by at least 1.2 MeV

with respect to the ground state fission band (Ktt = 0"'')

. Thus one might expect the average total fission fragment

kinetic energy for the fission of a J = l"*" compound

state to be as much as 1.2 MeV larger than that for the

fission of a J = 0'*' compound state. From energy con- servation, this is equivalent to a difference of

approximately 0.15 neutrons. In fact, the measured

effect is very much less than this. The reduction is

readily explained. The measurements of Bach et al.^^'

have shown that the inner peak of the fission barrier

is the higher for the ^'tOpjj compound nucleus. Thus the

difference in the average collective energy for the two

J states is a property of the first- barrier . It is well

established that mixing of K states occurs in crossing

the intermediate well and the outer peak of the fission

barrier^ ^' . Thus only a small component of the original

effect survives.

Possible resonance effects can now be evaluated for

neutron fission of the other fissile nuclei. For ^^^U,

the (n,Yf) process should have a width similar to that

in the neutron fission of 2 3 9py_ Here, however, there

are fully open fission channels for both J = 2'^ and

J =

3""" compound states and the (n,Yf) process for both

spin populations is only a minor effect. Similarly, it can be expected that any effect caused by the variation

in the mass distribution is extremely small. However,

using the data of ref. 64) it can be shown that

V ( J = 2"'")-v(J = 3"'') should be approximately 0.009. Thus

it is expected that there should be a difference of

<0.0045 between v at 0.0253 eV and that for the resonance at 0 . 19 eV where the spin has not been

identified.

For the neutron fission of ^^^U, the fission

barrier for both the 3 and 4~ spin states is considerably higher relative to the neutron binding energy than

that for neutron fission of either 2 3 3^ qj- 2 3 9p^_ Thus

on these grounds alone, the width for the (n,Yf) process

is extremely small. As indicated previously, Howe

et al.^^' have shown that v is also uncorrelated with

the resonance mass asymmetry. The data from ref. 64) can also be used to show that collective effects give

rise to a difference in v for the two spin states of

less than 0.001.

Similar arguments can be used to show that resonance effects in v for neutron fission of 2'*lpu are also very small, in agreement with the experimental

data from Simon and Frehaut^ . Thus the only case where care must be exercised is in the evaluation of v for thermal neutron fission of ^^^Pu.

190
V Ratios

In the consideration of the v values for thermal

neutron fission there are a n\imber of different

approaches that may be adopted. In all cases, the

values have been measured relative to ^^^Cf , but for a number of experiments, e.g. 1 and 7) in particular, it

is possible to eliminate the standard altogether and

refer the neutron counting rate for thermal fission

directly to the absolute calibration. This procedure

has been adopted in the past because of the apparent

disagreement between different absolute measurements

for ^^^Cf. However, since the revised v data for ^^^Cf

show a very acceptable level of consistency, it is more appropriate to evaluate the ratios and refer these

values to the average experimental value of v for ^^^Cf.

A brief discussion follows of the corrections that have

been made for specific experiments.

Boldeman and Dalton^^^

The adjustments applied to the measured v ratios

are listed in Table 9. For the evaluation of the

delayed gamma ray correction, the measured delay of

585 ns between the fission event and the opening of

the neutron counting gate was used. The listed delayed

gamma ray corrections are those relative to the correco r n tion for "^Cf. The correction for the resonance dependence of v corrects the measured result to the

equivalent 2200 ms~^ value. For the only case where

this correction was necessary, neutron fission of ^^^Pu,

the correction was based on the data of Fig. 2 and a calculated neutron leakage spectrum from a well

thermalised graphite system at 300 K. The estimated

correction is +0.10%.

TABLE 9 REVISION OF Vp RATIOS

2 3 3r 2 3 5, 239Pu 21*1 Pu

(%

Boldeman and Dalton^^'

Fission spectra (%

Delayed y-rays (%

Resonance dependence (% Mather et al.^^»''°'

Fission spectra

Delayed yrays

Colvin and Sowerby^'

Fission spectra

Resonance dependence {%

Condg^l'

Fission spectra

Delayed yrays

Hopkins and Diven

Fission spectra

Delayed Y~3:ays

,3)

+0.27 +0.21 +0.21 +0.21

-0.16 -0.16 -0.43 -0.43

+0.10

-0.04 -0.12 +0.01

-0.20 -0.20 -0.54

-0.06 -0.06

+0.06

-0.25

-0.08 -0.10

-0.10 -0.10

-0.03

+0.10

-0.03

-0.30

Mather et al .^^r'^°)

The liquid scintillator used in these experiments

was identical with that of ref . 7) . The adjustments

for the fission neutron spectra differences were based

on the data from this experiment. There is not

sufficient data in the original papers to make an accurate estimate of the contribution of the delayed

gamma rays from fission. However, the correction will

be slightly larger than that for ref. 7) because of the

earlier (undefined) opening of the neutron counting

gate. Corrections 1.25 times those applied for ref. 68)

have been used. The corrections are listed in Table 9.

2-880 •

2-860

2-840 -

0-01 0-05 0-1 ENERGY leVl

Fig. 2: Energy dependence of v for 2 3 9p^ fj-om Leonard^^^

Colvin and Sowerby '^

A very small correction for fission neutron spectra

differences is necessary if the Ullo efficiency curve is

used. The thermal v measurements were made using a beam

of thermal neutrons from GLEEP. For 2 39pu ^n identical

correction to that applied in ref. 68) has been used for

the resonance variation of v.

Cond£^'>

The fission neutron spectra difference correction

has been revised using the calculated energy dependence

referred to in the section dealing with the absolute

measurement of Asplund-Nilsson^^ . The delayed gamma

ray correction is also based on the data from this

section

.

Hopkins and Diven ^^

The adjustments applied in this experiment for

fission spectra differences were based on the assumption that the original corrections were derived using

the fission neutron spectra measurements of Bonner^^'

.

The thermal neutron measurements were made using a 400 keV neutron beam moderated by a polyethylene plug

in the neutron collimator. A correction was applied

for fission by neutrons above a cadmium filter cut-off.

The correction for the resonance dependence of v is

therefore extremely small.

Comparison of Ratios

Table 10 lists the revised ratios from the above

experiments. The listed ratios are those for prompt

neutron emission. The consistency noted in previous

evaluations still exists and there is no evidence within the sets to suggest any systematic error.

4 . Recommended Values

Table 11 lists the recommended values and their

errors from the revision above. The delayed neutron

components have been taken from Lemmel-"-^^ . The

conclusion to be drawn from this review of existing v data is that there is no evidence within the data of

any significant systematic error.

191
TABLE 10

V RATIOS

-P

Experiment 2 33, 2 35, 2 39 Pu 241 Pu

Boldeman and Dalton^^^ 0 659310 0015 0 6388+0 0012 0 .766210 0021 0.777110 0019

Mather et al.^^.'^°' 0 668 ±0 008 0 635710 0032 0 .771 +0 009

Colvin and Sowerby^^ U /xU uu U 0 .759810 nnc: t UUl) X n "7 "7 /I -4. n nm 1 UU / X Cond^^l' 0 638810 0053

Hopkins and Diven^' 0 6546±0 0058 0 641810 0053 0 .748510 0074

Average 0 6587±0 0013 0 638610 0010 0 .7647+0 0018 0.776910 0018

TABLE 11 14) R. L. Macklin

,

G. de Saussure , J. D

RECOMMENDED V VALUES

W. Kington and

S. Lyon, Nucl. Sci. and Eng. 8 (1960) 210.

Isotope VP

V

233u 2 46110 008 2 46810 008

235u 2 38610 007 2 402+0 007

2 39pu 2 85710 010 2 86310 010

2'tlpu 2 90210 010 2 91810 010

252cf 3 73610 010 3 74510 010

Acknowledgements

I wish to acknowledge the contributions to this

report of many authors too niimerous to mention. I am particularly indebted to J. J. Ullo, B. R. Leonard Jr.,

E. J. Axton and J. R. Smith for permission to use

unpxoblished data.

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193
MEASUREMENT OF THE ^"^Cf SPONTANEOUS FISSION NEUTRON SPECTRUM

M. V. Blinov, V. A. Vitenko, V. T. Touse

V. G. Khlopin Radium Institute

Leningrad, USSR

The neutron energy spectrum of ^^^Cf spontaneous fission has been measured by the timeof-flight method using two neutron detectors -- a ^LiHEu) crystal and a fast ionization

chamber with 235u layers. The influence of scattered neutrons and fission neutron emission

time on the spectrum shape was studied. In the range from 10 keV to 7 MeV, the neutron energy

spectrum is satisfactorily described by a Maxwellian distribution within the experimental

errors with T = 1.41 ± 0.03 MeV. pro

(Fission neutron spectrum; neutron detectors; neutron standard; time-of-f 1 ight; Cf)

Introduction

It is known that the neutron energy spectrum of

spontaneous fission of Cf has been recommended as a standard. Such important role of this neutron spectrum demands high precision of its shape measurement

over the whole energy region from several keV up to

10 - 15 MeV. The experimental data for the energy

interval above 1 MeV are in agreement, -^"^ however the

measurement accuracy is desired to be improved. As to

the neutron energies below 1 MeV, there is a significant difference between the data; the discrepancies

between the results attain 30-40% and more.^"-^^ That

makes much more precise measurements necessary.

The difficulties of measurements in this low

energy region are first of all due to the absence of

suitable detectors, which could be used in a wide

energy region, including the keV range. Plastic and

liquid scintillators permit neutron spectra measurements only in a limited part of this region. In the

past few years, lithium scintillation glasses have

come into use for neutron detection at energies below

1-2 MeV. But they present some serious difficulties

in the experiment. Hence we have developed a technique of °LiI{Eu) crystal use for time-of-fl ight

neutron spectrum measurements. This crystal was used

earlier as a neutron detector in experiments which did

not demand higbj-time resolution. The ionization

chamber with ^ U layers is a neutron detector with

smooth spectral efficiency and practically unlimited

energy sensitivity region. In order to use it in this

experiment, it was necessary to solve the problem of

ensuring the necessary efficiency and high time resolution simultaneously. Such a chamber has not yet

been used for fission neutron spectrum measurements.

252

In this report we present the results of Cf

fission neutron spectrum measurements, done by timeof-fl ight method using a ^LiI{Eu) crystal, and also

the preliminary data on spectrum shape, which was measured with a multiplate^ 235u ionization chamber.

Thus, we have used two detectors, which detect

neutrons by means of "Li{n,a) and 235u(n^f) reactions.

Both reactions are standards.

Experimental

Method and Instrumentation

The time-of-f1

i ght method was chosen, since at present it gives the highest accuracy in neutron

energy measurements.

252 The fragments of Cf spontaneous fission were

detected by a gas scintillation counter or by a fast

ionization chamber. Californium layers were deposited

on platinum backings 0.2 mm thick; 0.3 x 10^,

0.6 X 10 and 1.2 x 10^ fissions per second occurred in the layers. A ^Lil(Eu) crystal was used for neutron

detection below 2 MeV. This crystal was chosen

194

because of its higher y-equivalent in comparison with

that for lithium glass {3.5 MeV as compared to 0.8

MeV). This reduces considerably the danger of spectrum distortion due to y-quanta detection. Besides,

the lithium glasses contain large amounts of oxygen

and silicon, total neutron cross sections of which

exhibit large resonances. The main difficulty of the

work with lithium crystals is their large decay time

(about 1.2 microseconds), which even at their high

light yield, leads to the emission of a small number

of photons per nanosecond time interval. Thus, it was necessary to select crystals according to their

light output and photomul ti pi iers according to their

quantum yield of photocathode. In order to obtain a high time resolution, the electronics were constructed,

taking into account the specifics of light emission

features of ^LiUEu) crystal. A short description of

the procedure for the use of this crystal in a timeof-fl ight spectrometer was given in our publication.^*

The time resolution of 1.5 ns was obtained for lowenergy neutrons. It must be noted, though, that the

large decay time of the crystal in case of insufficient quality of electronics adjustment may lead to the erroneous spectrum -- intensity rise for low

neutron energies. The correct adjustment was con- trolled by the symmetry of a gamma peak.

The second detector was the fission chamber. Its low efficiency requires the use of multilayer con- struction. In our chamber ten layers of uranousuranic oxide (^^^U), 10 cm in diameter and deposited

on both sides of the electrodes were mounted (the

total 235u amount was 1.5 grams). The distance

between electrodes was 3 mm. The chamber was filled

with methane to atmospheric pressure. All the chamber

electrodes were connected to the same fast current

amplifier. The time resolution of the spectrometer

with this chamber was about 2 ns.

Investigation of the Effects Caused by Scattered

Neutrons

Scattered neutrons usually cause considerable

difficulties at low neutron energies, when continuous

neutron spectra are being measured. We paid special

attention to this problem and made several control

experimental setups. At first we tried fission fragment detectors and neutron detectors of the types

currently used in similar experiments. We found the

scattering effects were rather large, which could

give a neutron intensity increase at the low energy

part of the spectra about 30-50% or more. Then we measured neutron spectra, using detector setups of

different weight. The constructions, which caused no noticeable spectrum deformation within experimental

errors, were chosen in that way. Two spectra, which

show the detector scattering mass influence on the

spectrum shape, are presented in Fig. 1.

The gas counter after several modifications was made of a steel tube (diameter 40 mm, wall thickness
Figure 1. Time-of-fl ight spectra of Cf neutrons, measured with the Lil(Eu) crystal on 25 cm flight path in

two different setups: (o) - Setup with usual neutron scattering probability {the gas counter with

quartz window, the californium layer 6 cm from the window, model 30 PMT in the neutron detector).

(+) - The modified setup with significantly lower neutron scattering probability (the gas counter

without a quartz window, the californium layer 15 cm from the photocathode, model 71 miniature PMT in the neutron detector).

0.2 nm, length 150 mm) without the usual quartz window.

It was mounted directly on the photomultipl ier (PMT).

The californium layer was placed at the far end of the

tube.

The scattering effects of PMT mass and of crystal

protective capsule were observed for 6Lil(Eu) detector.

Multiple scattering in the crystal more than 6 mm thick

caused a noticeable change in spectrum shape. For mea- surements a crystal 4 rm thick and 15 mm in diameter

was used. It was glued directly to a miniature model

71 PMT. The crystal container was made of aluminum 0.5

imt thick. The last modification of multilayer uranium

chamber was made of a brass foil 0.2 mm thick. Uranium

was deposited on thin nickel foil.

The measurements taken to decrease (or remove) the

scattering masses of neutron and fission fragment

detectors permitted a decrease in scattering effects

(more than 10 times in comparison with currently used

experimental setups).

On Fission Neutron Emission Time

252

In Cf fission neutron spectrum measurements, it

is necessary to take into account the neutron emission

time in case it is more than 1 x

10"-^ second. Such

intervals between fission events and the emission of

some part of the neutrons are possible because of

large fission fragment angular momentum. A time

interval of 10~^^ to 10"!^ sec is large enough to allow the fragment to slow down noticeably in the

backing, on which the californium layer is deposited.

In such a case the energies of neutrons, emitted by a slowed-down and by a free-moving fragment, may differ

considerably. Due to this fact the backing material

and thickness may influence the spectrum shape. In Refs. 12 and 13 it has been shown that the upper limit

of neutron emission time in 235u fission induced by

thermal neutrons is about 4 x 10"^'^ to 10"13 seconds.

But in the case of the 252cf spectrum shape measurement, which must be used as a standard, a much more accurate determination of emission time must be made,

to permit a clearer interpretation of results.

195
We made a precise comparison of fission spectra of

neutrons, which were emitted bv fragments, moving in vacuum and in a dense medium. ^'^ The measurements were

made in wide neutron energy region from 20 keV to

7 MeV. It was found that spontaneous fission neutrons

are emitted earlier than 1 x lO"!^ sec after fission

event. The neutrons emitted about 10"13 sec and more

after fission event are less than 2 to 3% from the

total number of neutrons.

To determine the influence of iodine and of ^Li

,

which were present in our detector, and also the

influence of our crystal capsule, the same measurements were repeated with a ^Lil(Eu) crystal of the same size and light output. The background was found

to be 3% for 1 MeV, 2% for 0.5 MeV and 0.3% for 0.25

MeV. Measurement runs with ^LiHEu) were done roundthe-clock for one month and for 15 days with a uranium

chamber.

Neutron Spectra Measurements

252

Cf fission neutron spectra were measured at flight distances of 12.5, 25, 37.5 and 50 cm with a 6LiI(Eu) crystal and at 25 and 50 cm with a uranium

chamber.

The results obtained at four flight distances

with a lithium crystal were compared and found to be

identical within the experimental errors. A small

discrepancy was observed only in the region of the

240-keV peak, especially for 12.5 cm flight path, due

to the influence of energy resolution. These data

contradict the results of the work,^ where it was

found that the neutron spectra for various flight

distances differ significantly in the low energy part

of the spectra. The authors of Ref. 9 came to the

conclusion that below 1 MeV a considerable yield of

neutrons with emission times 10"^ to 10"^ sec exists,

and that at 100 keV the yield of degraded neutrons is

comparable to the yield of prompt ones. Our experimentslS did not lead to such a conclusion even at 10%

level of the value, published by the authors of Ref. 9. Fig. 2 shows the experimental time-of-fl ight distribution, obtained with 6Lil(Eu) crystal, measured in one

of the runs with a 25 cm flight path. It can be seen that for a 25 cm path a complete separation of the

gamma-peak from neutron distribution is observed, the

gamma-peak full width being 1.2 ns at half maximum and

5 ns at 0.1% of its height. This is evidence of the

absence of any "tailing" effects of the gamma-peak in

the region of neutron flight times. The intensity

ratio of the 240-keV peak to the valley between the

peak and the rest of the neutron distribution was 2.4

to 2.5, which also characterizes the time resolution

obtained. In Refs. 8 and 16 "fine structure" of the

252cf fission neutron spectra was reported for a wide

energy region. We did not find it below 1 MeV in this

experiment, and also in the measurements with stylbene

crystal for neutron energies between 1 and 5 MeV.

u) 05 (US oi: oio m

E.(MeV)

Discussion of Experimental Results

Fig. 3 presents the energy spectrum of neutrons

from 252(;f fission, which was measured by us with

5Li I (Eu) crystal

.

252

Figure 2. Time-of-fl ight Cf fission neutron

spectrum [6LiI(Eu) neutron detector, flight

path 25 cm].

u 16 18 2.0 E jheV)

252

Figure 3. Energy spectrum of Cf fission neutrons

(o) - Data obtained with ^Lil(Eu) detector

at 12.5, 25 and 37.5 cm flight paths. The

smooth curve is a Maxwell ian distribution

with T = 1.38 MeV. The errors plotted are statistical , only.

In calculating crystal efficiency we used the

estimated values of the 6Li(n,a) cross sections from

the ENDF/B-IV file. No correction was made for

multiple scattering in the crystal. In Fig. 3 a Maxwell distribution curve for T = 1.38 MeV is also

plotted. The experimental points lie within ±5% from

this curve. Taking into account the errors in the

6Li(n,a) reaction cross sections, one should note that

there is satisfactory agreement between the Maxwell

distribution and the experimental data.

Our preliminary data,'^'^ after correction for new

values of cross sections, agree with the present data

within the experimental errors. Further precision

spectrum measurements using the ^Li(n,a) reaction

depend greatly on more accurate cross section data for

this reaction.

In the calculation of the spectra, obtained with

the uranium chamber, we used the estimated fission

cross section for ^-^^U from Ref. 17. The energy

spectrum measured in this way is presented in Fig. 4. The results of these preliminary measurements in a wide energy interval can be approximated by a Maxwell

distribution with T = 1.41 ± 0.03 MeV, which is in

good agreement with the data obtained by the lithium

technique.

Thus, we may say that our experimental data agree

in the region above 1 MeV with the data.2-5 in the

low energy region (below 1 MeV) we did not find any

noticeable neutron excess over the Maxwell ian, which

does not confirm earlier Refs. 6, 8, 9 and 18. Considering the importance of knowledge of the precise

196
References

0 0.1 0,2 0.3 O.k 0.5 0.6 0.7 0.8

0 12 3 4 5 6 7 8 E„(MeV)

Figure 4. Energy spectrum of Cf fission neutrons.

Points - experimental data obtained with

uranium chamber at flight paths 25 and 50 cm. Continuous line - Maxwellian distribution with T = 1.41 MeV.

252

shape of the Cf spontaneous fission neutron spectrum as an international standard, we plan to continue

the measurements of this spectrum with higher accuracy

and in a wider energy region.

252

Along with the measurements of Cf fission

neutron spectrum at the Radium Institute, the determination of the average number of neutrons per fission,

V, is being performed in the spontaneous fission of

252cf. The work is being done in the laboratory,

headed by K. A. Petrzhak.

The method of separate determination of neutron

emission rate and fission events of cal iforni urn source

is used. The samples of about 0.5 microgramm ^^^Cf

have been prepared by vacuum evaporation on stainless

steel backings. The absolute fission rate is measured

by a surface barrier detector with a small level of

low energy "tail" of the fission-fragment spectrum in

low geometry, the solid angle being calculated. The

neutron yield is measured by manganese sulphate bath

techniques. The manganese solution is pumped con- tinuously from the spherical tank 1 m in diameter

through the counting system. To calibrate the counting

system, a 4Tre-Y-coincidence set-up was made.

At present the value of Z^l^^hf) = 3.738 is

obtained with estimated total error about 0.5%. The

work is still in progress.

Acknowl edgments

The authors wish to thank S. S. Kovalenko and

M. A. Bak for discussion of the results, and also

0. I. Batenkov, I. T. Krisyuk, N. M. Kazarinov, A. S. Veshchikov, V. I. Yourevich, B. M. Alexandrov, G. G. Verbitskaya for their help at different stages of this

work.

1. A. B. Smith, Prompt Fission Neutron Spectra

(Proc. Conf. Meeting Vienna 1972) IAEA, Vienna

(1972), 3.

2. G. V. Kotelnikova, B. D. Kuz'minov, G. N. Lovchikova, 0. A. Sal'nikov et al . , Neitronnaya

fizika (Proc. 3rd National Conference on Neutron

Physics, Kiev, USSR, 1975) Moskwa (1976), 5, 109.

3. L. Green, I. A. Mitschel, N. N. Steen, Nucl . Sci.

Eng. 50 (1973), 257.

4. H. H. Knitter, A. Paulsen, H. Liskien, M. M. Islam, Atomkernenergie 22_(2) (1973), 87.

5. H. Werle, H. Bluhm, I. Nucl. Energy 26^ (1972)

,

165.

6. J. W. Meadows, Phys. Rev. 157_ (1967), 1076.

7. L. Eki, D. Kluge, A. Laitaj, P. P. Dyachenko,

B. D. Kuz'minov, Atomnaya Energia 33 (1972), 784.

8. Yu. S. Zamyatnin, N. I. Krochkin, A. K. Mel'nikov,

V. N. Nefedov, Nucl. Data for Reactors (Proc. Int. Conf. Helsinki 1970) IAEA, Vienna 2 (1970), 183.

9. V. N. Nefedov, B. I. Starostov, Neitronnaya

fizika (Proc. 2nd Nat. Conf. on Neutron Physics,

Kiev, USSR, 1973) Obninsk 4 (1974), 163.

10. 0. I. Batenkov, M. V. Blinov, V. A. Vitenko,

I. T. Krisjuk, V. T. Touse, Neitronnaya fizika

(Proc. 3nd Nat. Conf. on Neutron Physics, Kiev,

USSR, 1975) Moskwa 5_ (1976), 114.

11. M. V. Blinov, V. A. Vitenko, Nietronnaya fizika

(Proc. 3rd Nat. Conf. on Neutron Physics, Kiev,

USSR, 1975) Moskwa 6 (1976), 252.

12. Yu. S. Zamyatnin, Phyzika delenia tyazhelykh

jader (Fission Physics of Heavy Nuclei) Moskwa

(1957), 75.

13. I. S. Fraser, Phys. Rev. 88 (1952), 536.

14. M. V. Blinov, V. A. Vitenko, I. T. Krisjuk,

Neitronnaya fizika (Proc.3nd Nat. Conf. on Neutron Physics, Kiev, USSR, 1975) Moskwa 5_ (1976), 131.

15. M. V. Blinov, V. A. Vitenko, I. T. Krisjuk, DAN 224 (1975), 802.

16. V. N. Nefedov, Preprint, NIAR (Atomic Reactor

Research Institute, Melekess) (1969), 52.

17. G. V. Antsipov, A. R. Benderskii, V. A. Kon'shin

et al . , Jadernye konstanty (Nuclear constants)

20 (1975), 3.

18. A. M. Andreichuk, V. A. Korostylev, B. G. Basova,

V. N. Nefedov et al . , Neitronnaya fizika (Proc.

3rd Nat. Conf. on Neutron Physics, Kiev, USSR,

1975) Moskwa 5^ (1976), 120.

197
PROMPT FISSION NEUTRON SPECTRA '

Leona Stewart

University of California

Los Alamos Scientific Laboratory

Los Alamos, NM 87545

and

Charles M. Eisenhauer

National Bureau of Standards

Washington, D.C. 20234

Recent measurements of the spectra of prompt neutrons emitted from neutron-induced fission

in ^-^^U and Pu are reviewed. Results are discussed in terms of departures of the data

from a simple Watt representation. An evaluation of the neutron spectrum from neutroninduced fission in ^-'-'U and spontaneous fission in ^^'^Cf, made by NBS in 1975, is also

reviewed. Recent measurements on the fission spectra of ^-^^U and ^39p^ seem to indicate

a harder energy spectrum than indicated by the earlier data available for the NBS evaluation. Possible reasons for this trend in the experimental data are discussed.

(Calif ornium-252; Maxwellian spectrum; plutonium-239; prompt fission spectrum; uranium-235;

Watt spectrum)

I . Introduction

The prompt neutrons emitted in the fission process

have been studied for more than three decades. It is

interesting to note, however, that the prompt neutron

spectra for many important fissile and fertile isotopes

have not been measured and the data available on other

isotopes show significant discrepancies. Recent experimental measurements of the spectra of emitted neutrons

have reduced some of these discrepancies. However,

since experiments have been limited to incident energies

of less than 2 MeV, the dependence of the spectra upon

incident neutron energy for neutro22^nduced2^^ssion is

still virtually unknown, even for U and Pu.

harder than for Pu. However, the difference in

average energy between ^-^^U and ^^^Cf is only ~ 5%, and

all of the spectrum determinations available today are "similar" in shape. Figure 1 shows a typical time-offlight spectrum by Johansson^ for 235u neutron-induced

fission. Also shown are best fits of his data to the

following forms:

Maxwellian Distribution:

(t)^(E') ~ v/e ' exp(-E'/T); E' = |t

Watt Distribution:

The differential fission spectrum (f) (E'), for an incident energy E and final energy E', of neutrons

arising from neutron- induced fission has been the

subject of many measurements. An alternate approach to obtaining information on the spectrum is that of integral measurements of the form R(E) = J'a(E ' )())g(E ' )dE ' , where o(E') is the cross section for a reaction such as

U(n,f), often used for reactor dosimetry studies.

Apparent discrepancies between differential

fission spectral measurements and integral cross sections and/or reaction-rate ratios have existed for

several years. These discrepancies were discussed at

a Consultants' Meeting^ in August 1971 and recommendations for measurements were made. In order to better define the differential spectrum (j)g(E')

for neutron- induced fission, a Specialists Meeting^ was called at Harwell in April 1975 to present and discuss

recent experimental data from Cadarache (France) , Geel

(Belgium) , Harwell (United Kingdom) and Studsvik

(Sweden) on ^-^^U and ^^^Vn. These results are described

in the next section.

252

The spectrum of Cf fission neutrons is very

similar to that for neutron- induced fission of ^^-'U and

^^^Pu. Furthermore, small near-point sources of ^^^Cf can be used as absolute flux standards. Therefore,

integral detectors with broad energy responses, such as the "^Pu(n,f) fission reaction, can be used to relate

the flux in a 235U fission environment to an absolute

flux determination using a ^^^Ci fission source.

Historically, both integral and differential

measurements predicted a "harder" spectrum, that is,

higher average neutron energy, for Pu than for

For ^^^Cf, the spontaneous fission spectrum is even

U.

,(E') ~ exp(-AE') sinh(i/BE');

E' +

2k 4A2

In general, good fits to either of these analytic

functions are obtained from all of the experimental

data available, except for small systematic deviations

which will be described in more detail in Sections II and III. For more than a decade, however, the Maxwellian with T = 1.29 to 1.30 MeV was most often the

recommended representation for low-energy neutroninduced fission on 235U.

A complete review of the measurement and theoretical contributions on fission spectra cannot be given

here — this subject, alone, could command an entire

symposium. Instead, only a few of the more recent

contributions to the field are discussed.

II. 235 239

Neutron-Induced Fission: U and Pu:

Since the Consultants' Meeting' on "Prompt Fission

Spectra" in 1971, much effort has been expended in

studies on the differential and integral spectra for

several important nuclei.

t,

Work performed partially under the auspices of the

United States Energy Research and Development Administration.

This paper was presented in place of a paper by

P. I. Johansson of Sweden, who was unable to attend

the Symposium.

198
10^

8

6

1

235

U

o JOHANSSON

MAXWELLIAN

(E = 2.00)

E,MeV

235 Figure 1. Comparison of Measured and Calculated Spectra of Emitted Neutrons from Neutron-Induced Fission in U

The experiments described at the Specialists

Meeting at Harwell are listed in Table I along with

other pertinent information. All of the experiments

were performed using time-of-flight, and Monte Carlo

corrections^ were applied for scattering and secondary

fission in the scattering samples. These corrections,

which were not applied in most of the earlier

experiments, tend to harden the spectra; that is give a slightly higher average energy. In addition, all but

one of the newer measurements record the neutron

spectrum out to energies greater than 13 MeV. As a result, the spectrum can now be predicted with more

confidence for neutrons emitted with energies up to 10 MeV.

199
23 5 219 TABLE I. Experimental Data for Neutron- Induced Fission of U and Pu

Incident

Target Neutron Energy No. (keV) Pts.

Range of

Emitted Neutrons

(MeV)

Authors

235,

10 - 58

400 + 33

- 39

< 520 + 20

100

180

530 + 32

2070

96

91

90

62

0.5 - 13.946

0.575 - 6.87

0.625 - 15.629

0.225 - 14.4

D. Abramson Cadarache

C. Lavelaine

M. M. Islam (Dacca), Geel

H. H. Knitter

J. M. Adams Harwell

P. I. Johansson (Studsvik)

P. I. Johansson Studsvik

B. Holmqvist

T. Wiedling

L. Jeki (Budapest)

r

239

10 - 58

Pu "S 215 + 32

100

180

530 + 32

2070

95

183

60

0.55 - 14.253

0.28 - 13.87

0.325 - 14.4

D. Abramson

C. Lavelaine

H. -H. Knitter

P. I. Johansson

B. Holmqvist

T. Wiedling

L. Jeki (Budapest)

Cadarache

Geel

Studsvik

Many attempts have been made to obtain analytic

representations of the fission spectra in order to satisfy the requests of the reactor designer. The two most often recommended are mentioned in the Introduction, namely the Maxwellian and the Watt distributions.

Furthermore, since it is impossible to compare the

experiments point-by-point, ratios of the experimental

data to analytic expressions permit one to compare

trends in the data.

235 The U data shown in Figure 2 are plotted as ratios of the experimental spectra to a Watt distribution with A = 1.0193; B = 2.3075; and E' = 2,027 MeV. 239 The -^^Pu experimental data are presented in the same manner in Figure 3 with A = 1.0364; B = 2.87 28; and

E' = 2.116 for the Watt distribution. The excess of

neutrons below one MeV in the Cadarache results are

said to be due to instrumental effects. Figures 2 and 3 were obtained by averaging over groups of emitted

neutron energies for each set of data listed in Table I. Data for all incident energies measured by Johansson

et al were combined in one set.

It should be noted that the experiments outlined

in Table I also included a study of the angular dependence of the fission spectra. For example, the Geel

spectra of ^35^ ^g^e observed at 45°. 90°, and 130°.

Cadarache made measurements on both ^^^U and 239py with

acceptance angles of + 30° and + 60°. All other spectra were observed at 90°. These data indicated that the

shape of the observed spectra were not dependent upon

the angle of the emitted neutrons.

Since the Maxwellian shape has been the recommended

distribution for fission-reactor analyses for many years,

the Johansson data on 235u were chosen to compare the

Watt and Maxwellian parametr izations . The ratios of the

2 + experimental points to these analytic representations'

are plotted in Figure 4. While the Cadarache and Geel

results do not show as clearly the preference for the

Watt distribution over the Maxwellian as do the Studsvik

(Figure 4) and Harwell data, all of the measurements

individually and collectively indicate slightly better

agreement with a Watt distribution. Therefore, the

Proceedings of the Specialist Meeting^ include recom- mended Watt distributions for 23S23Q -"-^U and -"^Pu. Consider2 S 2 ing these results and recent data on Cf , perhaps the

Watt distribution should be the recommended parametrization for all important isotopes on ENDF/B.

Ill . Comparison with NBS Evaluations

In 1975 Grundl and Eisenhauer"* described their

evaluation of differential neutron fission spectra for

^^^U and 252(;f^ based on all available experimental data

for these two isotopes. Tabulated data were obtained

and least-squares fits were made to a Maxwellian form by

consistent procedures for each data set. The goodnessof-fit was generally of the quality of the ^^^U spectrum

shown in Figure 1. However, the average energy deduced

for each fit varied by as much as 10% - considerably

greater than the quoted uncertainties of less than 5%.

The main purpose of the analytic representation was to obtain a consistent procedure for normalizing the data.

Reference Maxwellians for 235U and 252 Cf were determined

from a weighted average of individual data sets for each

isotope. Seven energy intervals were chosen and departures of the data from the reference Maxwellian shape were averaged. Uncertainties in the small but systematic

t It is interesting to note that the "best fit" to the

Johansson data using the Watt representation gives an average energy of 2.020 MeV while the Maxwellian fit

gives 2.001 MeV. These average energies were obtained

by fitting over limited energy regions while Johansson

obtained the same average energy (2.027 MeV) for both

shapes when fits were made over the entire energy range.

200
I.i

1.0

0.9

o

o

CADARACHE

oocr o o o ft o o oo o ° ° o

°

I.I

1.0

u:UJ

JDC.

UJ 0.9

Q.

X

UJ

LU I.I

1.0

GEEL

235

U

HARWELL

o ^ o o o

0.9

I.I

1.0

0.9

X Xy yX K

STUDSVIK

X Y X

Xy XX^ ^ X-XX X

X X X X

xx X X X X X X

4 , 6

E, MeV

8 10

Figure 2. Ratio of Experimental Data for Neutron- Induced Fission in to a Reference Watt Spectrum^ with

A = 1.0193, B = 2.3075 and E' = 2.027 MeV.

201
I.I

1.0

0.9

o

o

CADARACHE

o

o

o o

o o

u:

UJ

UJ I.I

:i.o

Q.X

111

UJ 0.9

GEEL

cxx X X

^x^VxXxXx 239

Pu

I.I

1.0

0.9

STUDSVIK

o o

4 6

E', MeV

6 10

239 2 Figure 3. Ratio of Experimental Data for Neutron- Induced Fission in Pu to a Reference Watt Spectrum with

A = 1.0364, B = 2.8728, and E' = 2.116 MeV.

departures, based on the spread of the data, were calculated and found to be statistically significant. The

NBS reference Maxwellian"* , together with departures from

this Maxwellian and assigned uncertainties, constituted

the evaluation.

Specification of the departures in seven energy

intervals had the disadvantage that the evaluated

spectra were discontinuous. Therefore, subsequent to the published evaluation, the same average departures

of the data from the Maxwellian were reanalyzed in

finer energy groups. This procedure resulted in a continuous analytic representation of the fission

spectrum. This NBS representation, which is called the

"segment-adjusted Maxwellian," was described in a recent

paper at the Vienna Consultants Meeting^.

The departures of the NBS evaluated spectra from a Maxwellian_with E = 1.97 MeV for 235u and from a Maxwellian with E = 2.13 MeV for ^^^Cf are shown in Figures 5 and 6. Some of the departures of the measurement of

Green^ for the ^^^Cf spectrum shown in Figure 6 give an indication of variations from the mean values. The

average energies of the evaluated spectra are 1.98 MeV

""Su and 2.12 MeV for ^^^Cf

.

for

Both figures show a deficiency of neutrons,

compared with the Maxwellian shape, above 6 MeV and

below 0.3 MeV. This behavior, which is consistent with

a Watt distribution, suggests that the reference

spectrum is closer to a Watt rather than to a Maxwellian.

However an integral measurement reported by McElroy

et al^ in a ^^Sy fission spectrum indicates that the

202
li.

UJ

K 0.

Q.X

UJ

1. 10

1.05

0.0

95 n(

UJ- 0.90

0.85

0.80

o o

XX O X X

o o

\ x=

X

X X

X

STUDSVIK

o MAX

X WATT

235

U

oo

X X

X X X ($

X

o o

1

4 6

E', MeV

8 10

235, Figure 4. Comparison of Ratios of Measured Spectrum of Neutrons Emitted from Neutron-Induced Fission of U to Analytic Fits of a Maxwellian and a Watt Shape. For the Maxwellian; E' = 2.001 MeV. For the Watt;

A = 1.0172, B = 2.2553 amd E' = 2.020 MeV.

10
MeV

o c n Figure 6. Percent departures of the NBS Evaluated Spectrum for Neutrons Emitted from Spontaneous Fission of Cf.

Solid horizontal lines are average departures of the data in 28 energy groups; dashed horizontal lines

are departure of Green's data; linear segments are piecewise continuous approximations to the average

departures.

Figure 7. Ratio of NBS Evaluated Spectrum for Neutron-Induced Fission in

A = 1.0193, B = 2.3075 and E' = 2.027 MeV.

235

U to a reference Watt spectrum with

spectrum at about 10 MeV may be closer to a Maxwellian

than a Watt. Finally, both figures indicate an excess

of neutrons around 0.5 MeV, which is inconsistent with

either of the representations discussed here.

235

For U, the apparent excess of neutrons has been

eliminated in more recent data by scattering corrections

in the ^35^ sample. For example, when Knitter's

correction for this effect was applied to Johansson's

204
data, the average energy of the experimental spectrum

increased from 1.99 MeV to 2.02 MeV. Since Johansson's

data represented only one of six experiments included in

the NBS evaluation, the impact of this change on the NBS

evaluation is small, incre||^ng the average energy of

the evaluated spectrum of U fission neutrons from

1.98 MeV to 1.99 MeV. It is suggested that the average

energy would increase still further if the older data

sets of Cranberg and Rosen^ , Barnard et al' and Werle

Bluhm^" were corrected for scattering effects in the

U sample. However, Johansson's sample was considerably thicker than many of those used in earlier work.

Therefore the effect of scattering in the sample would

probably be much less than the effect calculated for

Johansson's data.

In order to permit2^jdirect comparison, the NBSevaluated spectrum for U is plotted in Figure 7 as ratios to the same reference Watt spectrum assumed in

Figure 2. In general, the magnitude of the ratios shown

in Figure 7 is comparable to those shown in Figure 2. However, the ratios are systematically greater than

unity between 0.2 and 1 MeV and less than unity between

1 and 10 MeV. This follows from the fact that the

average energy of the evaluated spectrum (E = 1.98 MeV)

is lower than that of the reference spectrum (E = 2.03

MeV) obtained from the most recent measurements.^

252

For Cf , the excess of neutrons in the region

around 0.5 MeV is still not understood. Green^ has

formulated a model based on experimentally determined

parameters which predicts an excess of neutrons over the Maxwellian below about 0.75 Mev. On the other

hand, in the previous paper presented at this Symposium,

Blinov noted that an apparent component of low-energy

neutrons disappeared when the thickness of the Li-glass

detector was reduced. His statistics are not good

enough, however, to determine the excess over a Maxwellian to better than 5%.

IV. Summary and Conclusions

Recent experiments have served to establish with

greater confidence the fission neu|^^n spec^^^m for lowenergy neutron- induced fissj^^ of U and Pu and for

the spontaneous fission of Cf. Average energies of

emitted spectra inferred from experiments on neutron

induced fission have tended to increase with time.

Since the scattering corrections^ applied to the

recent measurements slightly harden all of the spectra,

better agreement with the older data might be obtained

if such corrections had been applied.

While these data, when considered separately,

show a slight increase in the average energy of the

emitted neutron, the highest incident neutron energy

investigated was only 2.07 MeV. Therefore, the trend

is indistinguishable from a fission spectrum independent

of incident neutron energy and the dependence of the

shape of the prompt fission spectrum upon the incident

neutron energy remains an enigma.

REFERENCES

"Prompt Fission Neutron Spectra", Proceedings of a Consultants' Meeting on Prompt Neutron Spectra;

organized by the International Atomic Energy Agency

and held in Vienna, 25-27 August 1971, IAEA, Vienna,

1972.

B. H. Armitage and M. G. Sowerby, ed, "Inelastic

Scattering and Fission Neutron Spectra," AERE-R 8636,

Proceedings of a Specialist Meeting; sponsored by

the Joint Euratom Nuclear Data and Reactor Physics

Committee and held at AERE, Harwell, April 14-16,

1975, Harwell, U.K., January 1977.

M. M. Islam and H. -H. Knitter, "The Energy Spectrum

of Prompt Neutrons from the Fission of Uranium-235

by 0.40 MeV Neutrons", Nuc . Sci. Eng. 50, 108

(1973).

J. A. Grundl and C. M. Eisenhauer, Proc. Conf

.

Neutron Cross Sections and Technology, Washington,

D.C. (1975). NBS Special Publication 425, p. 250.

Edited by R. A. Schrack and C. D. Bowman.

J. Grundl and C. Eisenhauer, "Benchmark Neutron

Fields for Reactor Dosimetry," IAEA Consultant's

Meeting on Integral Cross Section Measurements in

Standard Neutron Fields for Reactor Neutron

Dosimetry; International Atomic Energy Agency,

Vienna, Nov. 15-19, 1976 (to be published).

L. Green, J. A. Mitchell, and N. M. Steen, "The

Calif ornium-252 Fission Neutron Spectrum from 0.5

to 13 MeV," Nuc. Sci. Eng. 50, 257 (1973).

W. N. McElroy, R. Gold, E. P. Lippincott, A. Fabry,

and J. H. Roberts, "Spectral Characterization by

Combining Neutron Spectroscopy, Analytical Calculations and Integral Measurements", Consultants

Meeting on Integral Cross-Section Measurements,

IAEA, Vienna, Nov. 15-19, 1976.

L. Cranberg, G. Frye, N. Nere^g^, and L. Rosen,

"Fission Neutron Spectrum of U", Phys. Rev. 103 , 662 (1956).

E. Barnard, A. T. G. Ferguson, W. R. McMurray, and

I. J. van Heerden, "Time-of-Flight M^^^ure^^gts of

Neutron Spectra from the Fission of

Pu", Nuc. Phys. 71, 228 (1965).

U, U, and

10. H. ^I^le H. Bluhm,„"Fission-Neutron Biun^^2' Measurements

of U, Pu, and Cf", Jour. Nuc. Energy 26 , 165 (1972).

The fission spectral shape is considered to be well

determined over the range of emitted neutron energies

from about 1-8 MeV. Below one MeV, some experimentalists

see a large excess of neutrons over the Watt or the

Maxwellian shape; while above 8 MeV, the statistical

errors are often large due to the extremely low flux in

that region. In many reactor physics applications it

could be important to know about an excess of lowenergy neutrons.

205
ON QUANTITATIVE SAMPLE PREPARATION OF SOME HEAVY ELEMENTS*

A. H. Jaffey

Chemistry Division, Argonne National Laboratory,

Argonne, Illinois 60439

A discussion is given of some techniques that have been useful in quantitatively preparing

and analyzing samples used in the half-life determinations of some plutonium and uranium

isotopes. Application of these methods to the preparation of uranium and plutonium samples

used in neutron experiments is discussed.

(Absolute counting, aliquotting, a-activity, analysis (plutonium, uranium), isotope dilution,

sample preparation, specific activity)

Introduction

In an experiment which measures such properties

of heavy element isotopes as fission cross-sections

and the like, it is usually necessary to know the

sample's content of the isotope of interest with some accuracy. For some experiments, for example those

involving fission fragment counting, it is also desirable to have thin uniform deposits of the isotope.

Unfortunately, the two desiderata are often partially

incompatible.

I shall not review the various methods that may

be applied to this problem. A cursory examination of

the literature indicates that there have been many

papers on this and related subjects, and there have

even been a number of conferences. I would like to describe some techniques that my colleagues and I have

used extensively in a series of half-life measurements

of some uranium and plutonium isotopes. These methods

may be directly used for preparing certain kinds of

samples for fission and neutron measurements, and,

using the same principles, the methods may also be

modified for preparing a wider range of sample types.

Measurement of Heavy Element Half-Lives

by Specific Activity

After a period of relative dormancy, there has

been a flurry of activity in this field in the last

five or six years. Our group has measured and published results on 238u^ Sssy, 236u, and 233^,1-3 and

on 238p(j and 239p^J/^,5 techniques used have gone

through several stages of development, but they have

some common features throughout. In these experiments,

we used absolute alpha counting to measure the alpha

particle emission rate from known masses of the isotope

considered. We have examined many published experimental reports on the results of absolute counting as applied to half-life measurements. When two experimenters, each making absolute measurements to 0.2%

error or less, get results differing by over 1%, it is

clear that some "absolute" measurement is less than

absolute. In examining our own work, we checked all

the possible sources of error that we could think of;

there nevertheless remained the possibility that some important point was overlooked. However, in our last

measurement, 5 which involved ^^"^Pu, the half-life was evaluated by two methods: one, using our usual absolute counting technique; the other, using isotopic

dilution (mass spectrometric) to determine the mass of

235y grown into a known mass of ^^^Pu in a given time.

That the two results agreed within a statistical error

<0.1% gave us confidence that our absolute counting

method was rightly named.

This method stands on three legs: (1) measurement of the mass of the element used (e.g., uranium or plutonium), (2) accurate dilution and aliquotting for

sample preparation, and (3) counting the samples in an alpha counter of precisely known geometry factor.

Mass Measurement of Plutonium and Uranium Samples

Since it is desired to transform weights to moles

or inversely, a mass spectrometric analysis is necessary to provide the correct atomic weight of the element. In most cases, one isotope dominates, so only a modest accuracy is needed in the analysis.

For accurate mass measurement of good accuracy (in

the tenth percent range) two primary methods are useful : (a) preparation of a compound of known composition and purity or (b) oxidation-reduction titration.

A secondary method is that of mass spectrometric isotope dilution; secondary, in the sense that the spike

itself must ultimately be standardized by a primary

method.

The compound preparation technique has been widely

used in the half-life field. Among the compounds used

for uranium have been the pure metal and the oxide

U^Oq. a wider variety has been used for plutonium,

including the pure metal, the dioxide Pu02, the disulfate Pu(S04)2, cesium plutonium chloride Cs2PuCl6.

and the halides PUF3, PUCI3 and PuBrs- For accurate

determination of the element's weight from the weight

of the compound, it is necessary to know that the com- pound is really stoichiometric, i.e., that the material as formed actually contains the ratio of atoms described by the simple chemical formula (e.g., that, to

a high decree of accuracy, prepared PUCI3 indeed has

three times as many chlorine atoms as plutonium atoms)-

The metal is simplest in that no stoichiometry is involved, but it is harder to make without impurities,

which tend to be mostly oxides or materials coming

from the reduction process or from the crucible

material

.

For uranium it has been found that heating the

oxide carefully at 900°C in air for at least one hour

yields a composition very close to U3O8. Hence, the

NBS finds it feasible to provide natural uranium oxide

with a warranteed composition as a primary standard.

There has been less satisfaction with the stoichiometry of the various plutonium compounds, hence the

wide variety of compounds used.

With suitable experience and use of a large mass of metal, the impurities in metal formation can be adequately segregated. Plutonium is harder to make, one

of the reasons being the handling problem caused by

its high radioactive toxicity. For the same reason,

it is harder to analyze for small amounts of impurities, the methods used generally requiring volatilization (e.g., optical spectra). Nevertheless, the

problems have been sufficiently well solved so that

the Bureau also offers uranium and plutonium metal

samples of warranteed purity. No plutonium compounds

are so offered. The U content in the metal is known

better than the Pu content, the NBS setting a 1 a error of 0.009% for U and 0.025% for Pu.

206
Since both plutonium and uranium form well-defined

and stable (+6) and (+4) states in water solution, it

is feasible to measure the amount of oxidant or reductant required to change the element completely from one form to the other. For example, the method we used for

analyzing plutonium^ involved oxidizing it completely

to Pu(VI) and then reducing it quantitatively with

standardized ferrous ion solution according to the

reaction:

Pu(VI) + 2Fe(II) ^ Pu(IV) + 2Fe(III)

The concentration of Fe(II) is determined by oxidizing

it with pure standard potassium dichromate, whose com- position is warranteed by NBS. In the titration, the

point of exact equivalence is observed as a sudden

change in the current passing through the solution

(hence, an amperometric endpoint). Uranium can be

completely reduced to U(IV) and then quantitatively

oxidized with standard dichromate, with an endpoint

detected through a sudden potential change.'' The NBS

standard uranium or plutonium can be used as the primary standard instead of potassium dichromate; such a procedure serves to protect against possible uncertainties in the stoichiometry of the endpoint. Experience,

however, shows that this uncertainty is very small for

uranium (<0.02%) and quite small for plutonium

('^.0.05%).

The uranium and plutonium provided by NBS are excellent chemical standards, but the isotopic composition is usually not that desired by the experimenter. The titration method, can then serve as a means of transferring knowledge about the exact mass of an NBS sample to the experimenter's material.

The titration method for uranium can be made more precise than that for plutonium. For uranium titrations, the standard deviation per sample can be <0.02%,

whereas that for plutonium titrations is ^-0.05°^. For most purposes, such precision suffices. Such accu- racies are attainable only through the use of relatively large samples. The precision values described were

attained when 60 mg U and 20 mg Pu samples were used,

and smaller samples yield poorer precision. It may

also be noted that other precision titration methods

are available for both uranium and plutonium, but we

do not have experience with their use.

Aliquotting, Dilution and Sample Spreading

After the solution concentration has been evaluated, sample preparation requires that a known fraction

of the solution be appropriately transferred to a backing plate. Frequently the amount of isotope needed for

the experiment is so small that an intermediate dilution is required. This was, indeed necessary for the

half-life measurements of such highly radioactive

nuclides as 239pu or ^ssy.

That an intermediate dilution might be necessary

is due to the fact that accuracy requirements set a lower limit on the sample size which may be transferred. In our experiments, we use a Mettler semimicro single-pan balance which could weigh to 100 g and had a random weighing error of '^-0.02 mg. Since

all weight measurements were made by difference, the

expected error was 0.02 /2 s 0.03 mg. We required our minimum sample size to be '\-200 mg, which should yield

'V'0.015% net weighing error.

If, for example, there were 400 g of solution,

then 200 mg would represent 1/2000 of the total amount.

If the sample required was larger than this limit,

then the sample could be prepared by direct transfer

from the stock solution. However, for samples which

were a smaller fraction of the stock solution, an intermediate dilution would be needed. In the 239pu

measurement, for example, requirements for weighing

out titration samples resulted in a concentration of

'X'40 mg Pu/g solution. Since counting samples were to contain 3-4 yg Pu, intermediate dilution was necessary,

and by a factor of '\^2000, i.e., diluting the transferred volume (aliquot) containing ^-8 mg Pu with 400 g of acid solution. Where dilution is required, we have

found it feasible to dilute only once, by choosing the

dilution volume as large as necessary.

The technique used for accurately transferring an aliquot to a dilution flask is the same as that used

for sample transfer to a backing plate. It involves

the use of a polyethylene pycnometer of the type described briefly by Janet Merritt in Ref. 8 and more

fully in Ref. 9. The pycnometer is made from a small

polyethylene vial whose top is pulled out in the form

of a long, thin snout. Solution is sucked up into the

vial, the vial is weighed, some drops are delivered

into a dilution bottle or onto a plate (as in Fig. 1),

and the vial is reweighed. The weight delivered in

the aliquot is the difference in vial weights. The

technique is not quite this simple; some refinements

are necessary for good results, and these are described in Refs. 4 and 9. After some practice an operator can achieve quite good precision in the aliquotting process.

DISK

Fig. 1. Use of Polyethylene Pycnometer

The snout on a polyethylene vial is pulled out into a long capillary, the vial cleaned, dried and weighed.

The vial is carefully handled, the capillary inserted

into the stock solution, and liquid drawn up. The

capillary is wiped, the end cut off. Droplets are delivered with pressure, preferably with a mechanical

jig, and with no free-fall (with its potential of

splashing) when sample plates are prepared as in the

figure. Reweighing the vial gives the aliquot weight

with great accuracy.

The data of Table I and II show examples, the

first representing the results in an earlier phase of

developing the technique. Table I shows the counts of

samples taken from a dilution bottle and prepared as

in Fig. 1. The final result represents the activity

concentration of the original stock solution. Table

II corresponds to a different stock solution, s^ is

the usual estimate of the standard deviation per

sample and eav is the average counting error. We may

take 6^ = s^ - e|y, where 6 represents the error of

207
Table I. Aliquotting Precision - 1

Dis/min Ct

per g soln error

No. Ct/min Min xlO-^ Ei

1 26996.5 400 523.883 ±0.159

2 31768.9 400 523.441 0.147

•5O 1 nnn S?'^ 917

4 30425.8 900 523.441 0.100

5 28880.3 1000 523.547 0.097

6 39130.4 400 523.658 0.132

7 28887.3 1000 523.652 0.097

8 29806.8 1000 523.453 0.096

s = 0.192; £av = 0.115

Table II. Aliquotting Precision - 2

Di s/mi n Ct

per g soln error

No. Ct/min Min xlO--7 ci

1 40534.3 3800 419 978 ±0.034

2 48980.3 400 419 821 0.095

3 34436.1 900 419 913 0.075

4 41628.5 400 419 814 0.103

5 38192.6 1000 420 042 0.068

6 42040.3 400 419 830 0.105

s = 0.095; £av = 0-080

aliquotting, sample preparation and positioning it in

the counter. In Table I, 6 = 0.153 (0.029%); in Table II, 6 = 0.051 (0.012%).

For sample plates prepared as in Fig. 1, the

distribution of the activity over the place is made more uniform by evaporating with a spreading agent

such as tetraethylene glycol.'* Better uniformity, how- ever, is achievable by el ectrodeposi tion . We have

used the method of molecular plating, a high voltage

deposition from an isopropyl alcohol solution, using

the plating cell in Fig. 2. In this case, the aliquot

Fig. 2. Molecular Plating Cell

A: Aluminum sample plate and cathode. B: Teflon

wall. C: Base plate. D: Isopropyl alcohol solution. E: Platinum anode.

uranium up to 500 pg/cm^ in thickness. Plutonium

samples have not been made to such thickness.

The molecular deposition method is not quantitative, part of the element remaining in the isopropyl

alcohol. The amount, however, is generally a small

fraction of the total, and it can be accurately ac- counted for by washing the plating cell, evaporating

the isopropyl alcohol and depositing the residue for

counting with an a-counter.^ We have gotten higher deposition yields more consistently with uranium, the re- sidual being generally <1%; with plutonium, residuals

of <1% are achievable, but 2 or 3% are not uncommon.

Absolute Alpha Counting

Various methods can be used for absolute counting,

including coincidence techniques, liquid scintillation

counting and 4Tr proportional counting. We have used

the defined geometry technique, in which the a-particle

source is placed at a known distance from an aperture

of known dimensions. On the assumption of isotropic

emergence from the source, the fraction of particles

passing through the aperture is the same as the fractional solid angle subtended by the aperture at the source. The major factor limiting the accuracy of this

assumption lies in the scattering of a-particles either

from the sample mount or from the counter walls. With

suitable care in the counter's construction, a negligible fraction of the scattered a-particles reach the

aperture. The limit on the accuracy of the method is

then set by the accuracy of measuring the dimensions

which define the subtended solid angle.

In our half-life work we have used IGAC,^ an intermediate geometry alpha counter (Fig. 3). With a

SCALE

Fig. 3. The Intermediate Geometry a-Counter (IGAC)

Operated with flowing argon (10% CH^) gas filling en- tire chamber, 35 to 50 torr pressure (depending upon

a-energy). Geometry defined by accurately machined and

measured circular aperture K (with 0.025 mm thick edge)

and by measured distance between aperture plane and

sample surface H. Proportional counter chamber above

thin window L (^0.6 mg/cm^, plastic with evaporated

gold layer) with wires M spanning circular area.

is delivered into a temporary transfer tube or directly into the plating cell. The deposition gives quite

a uniform distribution, the range of density variation

being up to 20%, but generally <10% (for example, see

Ref. 1). Such uniformity makes practical the deposition of thicker samples than can be usefully prepared

by the method of Fig. 1. We have prepared samples of

relative solid angle G -vl/ll, it is intermediate in

solid angle between the 2tt counter (G ^^0.5) and the low

geometry counter, for which G^ is generally <0.01. By

reducing G^well below 0.5, we avoid the well-known

back-scattering from the sample mount and severe sample

self-absorption at oblique angles. Further, the large

208
aperture makes for accurate dimensional measurement

and negligible slit-edge scattering, and the relatively large value of allows counting of low specific

activity isotopes. At a given distance from the aperture plane (Fig. 3), the relative solid angle G subtended by an elementary area dS on the source is a function G(r) of the distance from the axis (defined

by the circular aperture) to dS.^^ The average relative solid angle for a particular sample is evaluated

by counting the sample in a 2tt counter with a series

of collimators. Each collimator contains annularlydistributed orifices at a given r-value, hence corresponds to the relative solid angle G(r).i'^i

By counting the same sample in both counters, we

have checked the IGAC geometry factor against that of

a low geometry counter (G '^1/100) with a 2 cm aperture

and a surface barrier detector; the two counters

agreed within counting statistics (0.05%).

One of the most important features of IGAC is

that it has a very flat plateau. This means that the

pulses at levels below the discriminator level constitute <0.005% of the total, even for quite thick samples. Figure 4 shows a pulse distribution for a relatively thick 238LI sample. Fig. 5 for a thinner sample.

In both cases, the number of pulses smaller than the

pulse selector level is negligible.

J \ I 1 I I L_ 20 30 40 CHANNEL NUMBER

50 60

Fig. 4. Pulse Height Distribution from IGAC

Proportional Counter for 238y Source

Discriminator setting is A. Inset shows the plateau

region BB^' on expanded scale. Subtracting background,

the net counts over the whole plateau is -4.6 + 10.6

counts/1000 min, compared to 2 x 10^ counts/1000 min

for the sample.

It is common practice to extrapolate a-counter

plateaus to zero pulse height and to assume that the

extrapolated counting rate, when corrected for the

geometry factor, corresponds to 100% counting efficiency. It has been our experience that when an extrapolation to zero pulse height is necessary due to

a sloping plateau, then the extrapolated result is

generally too low. In using IGAC, the pulse distribution (as in Figs. 4 or 5) is routinely used as a diagnostic. Net counts in the plateau region (i.e., a sloping plateau) usually correspond to low results, so such samples are automatically discarded. We usually

ascribe entry of small pulses into the distribution as due to poor sample spreading, with some clumping.

A continuum of small pulses has also been observed with liquid scintillation a-counting and with a faulty surface-barrier detector used with a low

20 30 40 CHANNEL NUMBER

50 60

Fig. 5. Pulse Distribution from a Thinner

and More Active Source

Discriminator setting at A. Within counting error, the

background accounts for all counts between B^ and £, so the plateau is flat in this region. Even without subtracting background, the counts of pulses smaller than

A (extrapolating to zero pulse height) is <2 x 10"^ of

the total number of counts.

geometry counter. For the liquid scintillator, the

pulse distribution showed a-particle pulses down to the

noise level, and extrapolation to zero pulse-height

yielded results 0.15-0.20% lower than that derived from

IGAC. Replacement of the surface barrier detector did

not sensibly change the peak half-width, but decreased

the number of small pulses and restored the correct

counting rate. This effect may have been due to a small defective region in the large area detector.

While IGAC (or other counters) may be used to determine the disintegration rate of an a-emitting

source, the experimenter generally wishes to determine

the rate for a specific isotope. Auxiliary information

is then required, usually supplied by a mass spectrometric analysis and an a-pulse analysis with a solid

state detector. In certain cases, however, the pulse

analysis does not help, because it cannot discriminate

between certain nuclide pairs which have overlapping

a-energies. The most important of these are: {^^°Pu,

239pu), (238pu, 21.1^^) (^ssy, 23tu)_ jhe members

of the pair must be distinguished by the mass spectrometric analysis or by other radiometric methods, such

as detection of the prolific ^'tiAm 60 keV y-ray with a Ge(Li) detector. With the mass spectrometric and pulse

analysis data, the fraction of the measured disintegration rate ascribable to the nuclide of interest may be

eval uated.

When some component in the sample gives rise to relatively short-lived decays, it is important that the

disintegration rate measurement be made at close to the same time that the auxiliary measurements are made.

The most common cases in which such problems arise are those in which samples of uranium contain 232y (73.5

yr) or samples of plutonium contain ^'tipu (14.5 yr).

The latter decays to ^^'^Am (432 yr) and the former decays to 228-|-h (RdTh, 1.91 yr) and then in turn to an entire chain of very short-lived decays giving four more a-particles. Depending upon the fraction of 232u

(or 2'tipij) present, neglecting the possibility of these

activity-growths can, and, indeed, has, caused serious errors

.

Consider the following examples:

1. A sample of 233u contains 0.0005% 232u;

hence, the 232u provides 1.072% of the total uranium

a-activity. Just after the uranium has been chemically

separated from thorium, the 233u a-activity is 98.928%

209
of the total activity, a fact that is readily determined by a-pulse analysis. Immediately after such

separation, the thorium daughter starts growing and

the subsequent decay chain is in almost immediate

equilibrium with the ^^^Th. Then, as a function of

time (T yr) after chemical separation, the a-activity

is expressed as

A(T) = A^[l + 5(0.01072)(l-e'°-^^^ ^)],

where Aq is the activity of the sample just after purification. For short times, approximately,

A(T) = A^[l + 0.01946 T]

The ^^^U a-activity is a decreasing fraction of the

total activity. Thus, in six months, th'e sample ac- tivity has increased by about 0.9%, so the ^^^U fraction has correspondingly decreased. Clearly, for

longer times or larger concentrations of ^^^U, the

growth of a-activity is more serious.

2. A Plutonium sample contains 97.1% ^^"^Pu, 2.5%

2'*0Pu and 0.4% ^'tipu. After separation from Am, the

sample activity is Aq and after T yr, it is

A(T) = A^[l + 0.010 T]

Analysis by Isotope Dilution

Into a sample containing isotope A of an element, we add a known amount of isotope of the same element.

After the two isotopes are chemically equilibrated,

the measurement of the ratio of A to yields a determination of the amount of A in the sample. This type

of measurement is called isotope dilution and the

added isotope is the spike. The procedure is most

commonly carried out mass spectrometrical ly, yielding

an A/B mole ratio, or by pulse analysis, yielding an A/B a-activity ratio.

Either method is valid only when care is taken to correct for inherent isotopic biases in the ratio

measurement. Such correction may be made, but high

accuracy (0.1% or better) requires a skilled operator.

The mass spectrometric method requires calibration for

inherent mass discrimination; pulse analysis requires

a valid correction for the fact that the high energy

peak has a tail which overlaps the low energy peak,

hence gives the lower peak an apparent higher intensity.

Recent examples from our work:

1. In a determination of the half-life,^

we measured the amount of ^^^L) grown in a given time.

Solutions of 233U and ^SBy <^q^q standardized as described above, and precise aliquotted spikes were added

to the 239pL| solution. Uranium was extracted and the

( 233u^ 235u^ 238|j) mixture was mass-spectrometrical ly

analyzed. The linear bias was evaluated by comparing

the observed 233u/238u ^q-^q ratio to the known spike

ratio.

2. In an evaluation of the ^ssp^ half-life, we measured the amount of 238py grown from the decay of a known a-activity of '^^'^Zm during a given time. A known

a-activity of ^'tOpu was added as a spike, the plutonium

was extracted, and the ^ftOpu/zsepu a-activity ratio

was determined by pulse analysis. The most serious

potential error source arose from the need to subtract

the low energy tail of the 238p,j peaks (5.491 , 5.448

MeV) from the apparent intensity of the ^'tOp^j peaks

(5.159, 5.114, 5.01 MeV). A "template" for subtraction

was derived from pulse analysis of a pure 238p^J sample.

In both cases, we tested the correction methods

by analyzing known artificial mixtures. Mass spectrometrically, samples made from mixed aliquots of titrated 2^^U, 2^^U and ^^^U were tested. For pulse analysis, we analyzed samples prepared from standardized

solutions of pure ^ssp^j and 2'*°Pu. The half-life

measurements were made when the calibration techniques were found to be satisfactory.

The first experiment was made somewhat more complex because a 238u spike was used. The ubiquitous

presence of natural uranium contamination in reagents,

dust, glassware, etc., required that super-clean techniques be used in all handling and chemical manipulations. It would be possible to avoid such meticulous

handling by using only one spike (usually 233u) and to calibrate the mass discrimination bias with a separate

source containing a known mass ratio (generally, an NBS 235u/238u source). However, one then loses an internal calibration based upon spike ratios taken under

the identical conditions used for the experimental

ratio. For the bias calibration with external source

to be valid, it must be shown that the bias calibration

of the mass spectrometer remains constant to the ac- curacy required over a number of succeeding runs.

Mass spectrometric isotope dilution analysis of

plutonium isotopes may be similarly used. Enriched

^'*^Pu is most commonly used for plutonium, because it

is generally in low concentration in the plutonium

samples of most interest.

The General Sample Problem

The techniques described above have the virtue

that they allow preparation of samples containing known

amounts of material, when analyzed solutions of the

desired isotopes are available. Samples prepared as in

Fig. 1 are suitable when small masses are needed or for

heavier samples, when sample self-absorption is not

important. Molecular plating may be used when heavier

samples are needed and uniformity is important.

When heavier samples of good uniformity are needed, other preparation techniques must be used. These,

unfortunately, are generally not quantitative. For

example, the well-known el ectrospraying technique can

be used to prepare thick, quite uniform samples, but

quantitative deposition is not possible. In many

cases, however, it is feasible to calibrate the

samples, either by counting or by destructive analysis.

If the samples are not too thick, they may be

counted either in a low or intermediate geometry

counter, depending upon specific activity. In either

case, the quality of the plateau must be carefully

examined, as mentioned above. If the isotopic halflives and the mole- or activity-ratios are known, the

isotopic masses can be calculated.

If the samples are not good enough for direct

counting, they may be dissolved after the experi-ment is

completed. Analysis may then be carried out either by

a-counting an aliquot as described above or by analyzing through one or the other form of isotope dilution.

References

*Work performed under the auspices of the Division of

Physical Research of the U. S. Energy Research and

Development Administration.

^A. H. Jaffey, K. F. Flynn, L. E. Glendenin,

W. C. Bentley and A. M. Essling, Phys. Rev. C, 4,

1889 (1971).

210
K. F. Flynn, A. H. Jaffey, W. C. Bentley and

A. M. Essling, J. Inorg. Nuol. Chem., 34, 1121

(1972).

^A. H. Jaffey, K. F. Flynn, W. C. Bentley and

J. 0. Karttunen, Phys. Rev. C, 9, 1991 (1974).

^H. Diamond, W. C. Bentley, A. H. Jaffey and

K. F. Flynn, Fhys. Rev. C, 15, March (1977). In press.

'a. H. Jaffey, H. Diamond, W. C. Bentley, K. F. Flynn,

D. J. Rokop, A. M. Essling and J. Williams (to be

publ ished)

.

'C. A. Seils, Jr., R. J. Meyer and R. P. Larsen,

Anal. Chem., 35, 1673 (1963).

A. R. Eberle, M. W. Lerner, G. C. Goldbeck and

C. J. Rodden, "Titrimetric Determinat'Lon of llrawLum

in Product Fuel and Sarap Materials after Ferrous

Ion Reduction in Phosphorous Acid; Manual and

Automatic Titration," Proc. Sympos. on Safeguards

and Techniques (Karlsruhe, July 6-10, 1970),

Vol. II, p. 27, IAEA, 1970. Also New Brunswick

Laboratory Report NBL-252 (1970).

^Janet S. Merritt, Vlucl. Instnm. Methods, 112 , 325 (1973).

'j. S. Merritt and J. 6. V. Taylor, "Gravimetric

Sampling in the Standardization of Solutions of

Radionuclides," Atomic Energy of Canada, Ltd.

Report AECL-2679, 1967 (unpublished).

'w. Parker, H. Bildstein and N. Getoff, Mcl. Instr.

Methods, 26, 55 (1964).

A. H. Jaffey, Rev. Sci. Instr., 25, 349 (1954).

211
BLACK AND GREY NEUTRON DETECTORS

F. Gabbard

University of Kentucky

Lexington, Kentucky 40506

Recent progress in the development and use of "black" and "grey'

detectors is reviewed. Such detectors are widely used for counting

neutrons in (p,n) and (a,n) experiments and in neutron cross section

measurements. Accuracy of each detector is stressed.

(Review; black & grey neutron detectors; neutron flux measurement.)

Introduction

Comprehensive early reviews of methods

of neutron flux measurement and associated

standard cross sections have been given by

Barschall, et al. Larsson^ and Perry-^. Advances in neutron flux measurement techniques

have been discussed in Conferences on Neutron

Cross Sections and Technology in 1968 and

1970.^'^ Reviews were given at these conferences by Batchelor, Gibbons^ and Landon.^ The

symposium on Neutron Standards and Flux Normalization, held at the Argonne National

Laboratory in October, 1970, treated, rather

extensively, the whole subject of neutron flux

measurement as practiced at that time. Further

work has been reported at the Conference on Nuclear Cross Sections and Technology'^ held at

the National Bureau of Standards in 1975 and

at the International Conference on the Interactions of Neutrons with Nuclei held at the

University of Lowell in July, 1976. Papers at

the Lowell Conference by H. Liskien and B. Zeitnitz deal with detectors and standards.

The basic techniques for neutron flux

measurement are well known and advances have

been largely in the refinement of these methods

toward easier use and higher accuracy. There

are some five (5) basic methods in current use

for the measurement of neutron flux. These

are

:

(1) Associated particle methods;

(2) Proton recoil methods;

(3) Neutron moderation methods;

(4) Associated radioactivity methods; and

(5) Methods of reaction cross section

comparison

.

s point to be emphasized^ is

"best" method for measuring

ch works at all energies and

s; the "best" technique detron energy and the geometrin of source and detector.

One obviou

that there is no neutron flux whi

in all geometrie

pends on the neu

cal configuratio

Since the advent of time-of-f light

methods of neutron spectroscopy, the time

response of fast neutron detectors has become

an important consideration in neutron-fluxstandards applications. Properties of an IDEAL

neutron detector include the following:

(1) Variable detection sensitivity;

(2) An efficiency independent of neutron

energy, i.e., a "flat" response;

(3) Insensitivity to background radiations; and

(4) Fast time response.

The advantage of variable detection

sensitivity is that different experiments

require the use of different intensity levels

in the neutron flux. Fast time response is

important in time-of-f light applications.

Such an ideal detector does not exist

and it appears unlikely that any single detector can have the optimum characteristic in all

of these properties. Nevertheless, these are important criteria for selection of a real

detector

,

The objective here is a review of recent

developments in the use of "black" and "grey"

neutron detectors for the measurement of neutron flux. The detectors which I shall describe

are of two types; proton-recoil detectors and

moderated neutron detectors. The recoil detectors use liquid or plastic scintillators

as the active medium and moderated detectors

utilize hydrocarbons, LiH, or carbon as the

moderating medium.

Selection of detectors

was based on personal knowle

velopment and/or new uses,

or "grey" neutron detectors

detectors which will not be

detail are the Large Liquid

the Activated Bath Detectors

ered by Dr. Axton in this se others at other sessions of

for description

dge of recent deAmong the "black"

or classes of

discussed in any

Scintillators and

These are covssion and by

the Symposium.

The time period covered in this review

is roughly from 1970 until now. During this

time there have been no startling breakthroughs

in the neutron counting game; however, there

has been steady progress on all fronts.

The Macklin Sphere

First, I want to recall, or call as the

case may be, your attention to the Graphite

Sphere Detector,^ or Macklin Sphere designed

and built by R.L. Macklin at the Oak Ridge

National Laboratory, Although this detector

is not new, there have been important developments in its use. This detector is a sphere

of reactor grade graphite with a radius of 78 cm. Neutrons originating at the center are moderated in the graphite and counted by eight

(8) BF^ counters near the surface of the sphere.

The detector is covered with cadmium to reduce

the effect of room scattered neutrons. Figure

1 shows a picture of the detector with Cleland

Johnson in the laboratory at Oak Ridge.

An attractive feature of this detector

is that its simple construction makes it feasible to calculate its relative efficiency as

a function of neutron energy through application of age-diffusion theory. Figure 2 shows

the calculated efficienty of the detector as a function of neutron energy. To be noted is the

212
Fig. The Macklin Graphite Sphere neutron

Detector with C.H. Johnson at the

Oak Ridge National Laboratory.

S8BJ in

|5

1.15

IK)

i05

100

$035

030

0B5

080

aool Q01 0.( 10 NEUTHON ENERGY (M«v)

(0

Fig. 2. The calculated efficiency of the

Macklin Sphere as a function of

neutron energy.

flat response curve at low energy in the keV

region. This feature makes the detector particularly well suited for use in measuring

total neutron yields from sources placed at

the center. The detector has been used extensively for neutron production cross-section

measurements and continues to be used, niost

recently by Johnson, Bair and co-workers

for highly accurate measurements of cross

sections for neutrons produced in (p,n) and

(a,n) reactions.

Bair, Johnson and co-workers have re- cently recalibrated the Macklin Sphere using

NBS II as the calibrating source. At the

time of calibration, the detector efficiency

for Ra-Be neutrons was measured to be

(0.003149 ± 0.000011). A local source was made by placing 492 . 2 mg of radium in a Be can. This source was compared to NBS II and is used

as a local calibration source in the continued

use of the Macklin Sphere. Johnson estimates

that recent measurements of {p,n) cross sec- tions are accurate to better than 1%.-'-'^

In addition to its continued importance

for measurement of neutron production cross

sections, the Macklin Sphere as a prototype

has good potential as a secondary standard.

Source comparison with this detector can be

made quick and precise since the BF^ detectors

make measurements of neutron emission rate

simple and in "real" time as compared to Activated Bath methods. The efficiency can be

raised to about 3% through use of 10 BF- counters in place of the natural BF-^ detectors.

The response curve for the detector is well

understood. The response can be quantitatatively checked with high precision in the keV

neutron energy range through the use of the

associated activity method with ^LI (p,n) "^Be , 51v(p,n)51cr and ^ 'Fe (p , n) ^ 'Co as neutron

sources

.

I also want to mention in passing a second detector designed by R.L. Macklin and

co-workers ' at Oak Ridge, This detector

can be made flat through appropriate tuning

of the position of BF., counters in the moderator assembly. This detector has an efficiency of '^'30% and will be useful in applications where very high sensitivity is needed.

The Poenitz Detectors

1 . The Grey Detector

An important group of black and grey

neutron detectors has been developed and used

by Poenitz. 13, 15 rp^e first of these detectors which Poenitz has called the "Grey Neutron Detector" consists of a homogeneous hydrogeneous material such as water, paraffin, or polyethylene with an entrance channel for a collimated neutron beam. The neutrons are thermalized in the moderator and subsequently

captured in the moderator material. ir-he capture Y~ rays are detected at the surface of

the moderator. Figure 3 shows a schematic

diagram of a typical experimental setup for

the use of this detector.

m NEUTRON }— BEAM

NTRANCE CHANNEL

MODERATOR

COLLtMATOR

Nal- DETECTOR

eV-RANGE keV-RANGE MeV-RANGE NEUTRON ENERGY

Fig, Schematic diagram of the Grey Neutron

Detector and a comparison of the efficiency response curve (b) with that

of a manganese bath (a).

213
Poenitz has calculated the relative efficiency for the "Grey Detector" which exhibits a flat energy response over a wide range

of energies. Comparisons of the relative efficency of this detector with that of a manganese bath are shown in the lower part of Figure

3.

The relative efficiency of the "Grey

Detector" was measured by two different methods, -'^ the associated activity method and a flux integration technique. Results showed

agreement with calculated efficiency to within

2% for neutron energies below 1 MeV and showed

somewhat larger deviations toward higher neutron energy.

14 The detector reported by Coates, et al. uses the same operating principle, i.e. the

'-"B-vaseline detector is a "grey" detector.

In this latter case, the 478 keV y-rays from

the l^B (n, ay) "^Li were detected. The time re- sponse of this detector is about 0.7 y sec which

is sufficiently fast for work with linac

sources

.

MULTIPLIEBS

ENERGY SPECTBA ENERGY SPECTR*

Fig. 4. Schematic comparison of a conventional scintillator detector (left) and

the Black Neutron Detector (right)

.

The energy range for which the Grey Neutron Detector appears , to be most useful is the

keV range. This detector has been used by

Poenitz in several experiments for measurement of relative cross sections . ' -'-^ The precision of these measurements ranges around (2-

3)% for keV neutrons.

2. The Black Neutron Detector

The "Black Neutron Detector" is a fast

time-response detector (designed by W.P.

Poenitz) for absolute neutron flux measurements. 5,17 rpj^g detector consists of a hydrogeneous scintillator material in the shape of

a cube, cylinder or sphere. The neutron beam

enters the detector system through a channel

which terminates near the center of the detector volume. The light produced in the scintillation material is detected by several

photo-multiplier tubes. A neutron entering

the detector undergoes several successive collisions and loses most of its energy to hydrogen or carbon nuclear recoil energy. The com- parison in the concept and energy response of

a conventional scintillation detector and the

Black Neutron Detector is illustrated in Figure 4. A very important difference is in the

light pulse spectra in the two detectors. The

energy spectrum for the Black Neutron Detector

allows accurate extrapolation to zero pulse

height because most of the energy of all neutrons is lost in the detector. Pulses counted

above a low-set bias will account for almost

all (90-95)% of the neutrons losing energy in

the detector.

1 7 A monte-carlo computer code. Carlo Black,

for the purpose of evaluating a Black Neutron

Detector of cylindrical shape was used by

Poenitz. The parameters in the evaluation

were the length of the cylinder, H, the radius

of the cylinder, R, and the length of the re- entrant hole, H(,, the radius of the re-entrant

hole, R^, and the detection threshold, E . The

prototype detector is filled with a scintillator having atomic densities of hydrogen and

carbon of N^ and N^, respectively. The calculated efficiency of this detector is shown

in Figure 5. The structure in the calculated

Fig.

NEUTRON ErCRGY (Mav)

The calculated efficiency of the

Black Neutron Detector versus energy.

The calculation was done with CarloBlack with H = 40 cm, R=13 cm, =

15 cm, Rc. = 1.26^cm, = 200 key^, N^ = 4.2

cm'

x 1 0^2 / cm- Njj-7 04

0 key.

X 10^^ /

efficiency curve is due to scattering resonances in carbon and gives fluctuations of 1% or so.

Figure 6 shows examples of calculated

energy spectra at neutron energies of 1 MeV,

2.5 MeV, and 4 MeV. The important features

are the low background counting rates at

pulse heights below 20. Figure 7 shows experimental spectra obtained with the Black Neutron

Detector. The two spectra at 1.5 MeV and 2.5

Mev shov7 the desirable features of low background and good peak definition. The energy

resolution spreading of the detector system

causes the shape of the spectra to be smoother

than the calculated curves of Figure 6.

The Black Nfeutron Detector has high sen- sitivity, flat response as a function of energy and a fast time response ('\^4-5 ns) . This

detector with a well known energy response can

be used in time-of-f1 ight experiments with

pulsed or associated particle-tagged sources.

214
Fig. 6. Calculated (Carlo-Black) pulse-height Fig. 8. Schematic of the NBS Black Detector

spectra evaluated for three different showing the scintillator and the elec — monoenergetic neutron beams. tron multiplier.

I I I ' I M I ' ' 1 » 100 IM

I

I I I I

I

I

1—I I

I

I I I 1 30 KM ao

Fig. 7. Experimental pulse-height spectra for

1.5 and 2.5 - MeV neutrons.

The detector, designed for use with collimated

beams, must be carefully shielded against stray

neutrons and yi^adiation . Absolute measurements-'-^ with the Black Neutron Detector have

been made to an accuracy of ^2%.

The NBS Black Detector

18 Lamaze, Meier and Wasson"" at the National Bureau of Standards have constructed a detector of a plastic scintillator after the

Black Neutron Detector design of W.P. Poenitz^.

This detector consists of a cylinder of NE 110

plastic scintillator 12,7 cm in diameter by

17.78 cm long with a re-entrant hole having a diameter of 2.54 cm and a depth of 5.08 cm. The scintillator is optically coupled to an RCA 8854 photomultiplier tube. A schematic

diagram of the detector is shown in Figure 8

.

17 The program Carlo-Black was used to

provide design criteria for the NBS Black Detector .

Fig. 9 shows the efficiency of this detector as a function of neutron energy for

three different bias settings. It is seen that the response curves are very nearly flat

on the energy interval between 200 keV and 1000

keV.

. 90

>-u

,z

UJu

tZ

u.bj so

TO

.r--«- 15 keV CUTOFF

50 keV CUTOFF

100 keV CUTOFF

200 •4(30 600 NEUTRON ENERGY, k«V

600 idoo

Fig. Calculated efficiencies of the NBS

Black Detector for different lower

level cutoffs. Calculations were

done with the computer program CarloBlack,

At low neutron energies (<250 keV) , only

a small number of photons are produced in the

scintillator and only a fraction of these produce photoelectrons . Therefore, the photoelectron statistics are important in the determination of the response below 300 keV or so. This effect is illustrated in Figure 10 which

shows comparison of a calculated and experimental spectrum at Ej^ = 250 keV and 560 keV. The

calculated curves were obtained using CarloBlack-'-"^ as modified by M. Meier to calculate

the spectral distributions at the two neutron

energies. The modification is the explicit

treatment of the Poisson statistics when N,

215
4,000

560 keV

CALCULATED RESULTS

ooooo EXPERIMEMTAL RESULTS

250 keV {

CALCULATED RESULTS

' EXPERIMENTAL RESULTS

I BIAS CHANNELS

Fig. 10.

20 40 60 80 100 120 140 160

CHAIMIMEL NUMBER

Comparison of calculated and experimental recoil spectra from the NBS

Black Detector at incident neutron

energies of 250 keV and 560 keV.

Calculations were done with CarloBlack as modified by M.M. Meier for use with small N in the Poisson statistics .

the number of photoelectrons , is small. The

result of Meier that the photoelectron production at low energies corresponds to about

15 keV/photoelectron is consistent with Cranberg's20 findings.

The NBS Black Detector has an efficiency

near 100% and has very fast time response

('^5 ns) . As is the case with the Black Neutron Detector built by Poenitz, this detector

is designed for use with a collimator or beam

defining device and must be well shielded. The

detector is being used in both the Linac and

Van de Graaff laboratories at NBS.

The Kentucky Polyethylene Sphere

The nuclear physics group at the University of Kentucky has developed a detector for use in the keV region. This detector has been

briefly described in the Conference on Neutron

Cross Sections and Technology held at NBS in

1975. ' This detector was constructed primarily for the accurate measurement of total

neutron production cross sections. The detector consists of a polyethylene sphere with a radius of 30 cm in which 8 IOBF3 counters are radially installed. A picture of this detector is shown in Figure 11. The detector is

designed so that it may be opened up for inFig. 11. The Kentucky polyethylene sphere

detector. The two hemispheres are drawn apart. In operation the

counter is closed.

sertion of targets into a chamber at the center

of the detector. The charged-particle beam

from a Van de Graaff accelerator can be passed

through the target and the resulting neutrons

counted. The counter is covered with 1 mm of

cadmium to reduce neutron background from the

room.

The efficiency of this detector has been

measured in the neutron energy range from 30

keV to 2.5 MeV. The resulting energy response

curve is shown in Figure 12. Absolute efficiency measurements were made near 300 keV and

1.35 MeV. The results gave (0.55 ± 0.017)%

and (0.56 ± 0.016)% respectively for the efficiency of the detector. The associated activity method was used in these efficiency measurements. A 35 cm-^ Ge(Li) detector was used

to compare the gamma-ray yield from the residual ^-'-Cr and ^'Co produced in the calibrating

0.8

0.6

0.4

0.2

Absolute measurements _ x.* Relative meosurements

_}f-H-f-M 1 J

i ^

Sphere Counter

Efficiency for neutrons from Pu-Be Source =0 . 47%"

0.5 1.0 1.5 2.0

Average Neutron Energy (MeV)

2.5

Fig. 12, The absolute efficiency of the polyethylene sphere counter as determined by the associated activity

method

.

216
targets of V and ^^Fe, with standard sources

of ^Icr and Co obtained from the National

Bureau of Standards. Accuracy of the standards

was ± 1.7%. The estimated accuracy of our mea- surements was ± 3%. Relative efficiency was measured using the 'Li{p,n) 'Be reaction. In

the energy range below an average of 1.5 MeV,

the relative neutron yield as a function of

neutron energy was determined using the 'Be

decay as observed from observati

keV Y-ray from ^Li.^^

on of the 47i

Figure 13 shows a comparison of the Li

(p,n)^Be excitation function as measured by

Gibbons and Macklin23 with the cross section

measured with the polyethylene sphere detector

.

Agreement is very good up to a proton energy

of 4.2 MeV. Toward higher energies, the yield

observed with the polyethylene detector falls

off relative to the yield observed with the

graphite sphere counter. This fall-off of efficiency of the polyethylene counter relative

to the graphite sphere is shown as a function

800

€00

400

200

T

'Li (p.n)' Be Reoclion ERROR t 4% — Dola by Macklin ond Gibbons

• This work — Ratio of the from lo the rear counts

0

I ^ I I I _L

2.0 3.0 4.0 5.0

INCIDENT PROTON ENERGY (MeV) Fig. 13. Comparison of the total (p,n) cross

section as determined with the polyethylene sphere and by Gibbons and

Macklin in the Macklin Sphere.

of neutron energy in Figure 12. The decrease

in counting efficiency for the polyethylene

sphere is quite rapid above 2 MeV.

The falloff of efficiency toward higher

energies is primarily a geometrical effect re10 suiting from the radial placement of the -^^BF^

counters; this means that the fraction of the

volume filled with detector decreases inversely as the square of the distance from the detector center. The average thermalization distance increases with neutron energy and the

spatial distribution of thermal neutrons spreads

radially. This latter effect partially compensates for the detector geometry. The key to

the energy response (Fig. 12) is the placement

of the •''^BF3 counters in their radial channels.

Figure 14 shows counting rate as a function of

counter position for three different average

neutron energies. The three lines come very

near to intersecting at the same point. If

the counters are placed at 1.2 cm, the apparent difference in efficiency is about 1% for

the three energies shown. Figure 12 shows that

the flat response extends nearly to 2 MeV.

In addition to reasonably high sensitivity and flat response, the polyethylene sphere

is quite insensitive to gamma-ray background.

It has moderate sensitivity to neutrons coming

tn +Jc

0ou

10

0

o 0.06 MeV

• 0.8 MeV

A 1.6 MeV

0 0.5 1.0 1.5 2.0 2.5

Position (cms)

Fig. 14. The counting rate in the polyethylene counter as a function of counter

position for three different average neutron energies. The detectors

were placed at 1.2 cm for operation.

from surroundings,

microseconds

.

The time response is a few

Now I want to turn briefly to the performance of the detector as a device for measuring total neutron production cross sections

from nuclear reactions, such as (p,n)and (a,n),

of interest in stellar nucleosynthesis and

nuclear reactor design problems.

The detector and the y^ay counter, Ge (Li ), were used to measure the neutron yield for y~

ray count^-'- from ^'-Cr for the ^'"V (p , n) 51cr re- action at incident proton energies of 3 and 5 MeV. This ratio was constant to a precision

of 0.5% indicating that a precise neutron count

is given by the polyethylene sphere even though

the ground state neutrons at a proton bombarding energy of 5 MeV is 3.4 MeV (Q = -1. 5 34); and,

the sphere efficiency is decreasing with energy

above a neutron energy of 2 MeV. This precise

result is obtained primarily because the neutron spectrum emitted in (p,n) and (a,n) re- actions roughly follows a lumpy Maxwellian distribution.^^ The lumps correspond to individual excited states and groups of states. Such

a spectrum is illustrated schematically by the

dashed curve in Fig. 15. This dashed curve was derived from the time-of-flight spectrum shown,

corrected for scintillator efficiency, and the

assumption that the peak of the Maxwellian falls

at about 0.5 MeV.24 The time-of-f light spectrum was taken with a 1.3 cm plastic scintillator. Fewer than 10% of the neutrons emitted

have energies higher than 2 MeV.

The energy range in which the efficiency

of the polyethylene sphere is flat can be increased by the insertion of trimmer counters

into the sphere. For the present applications

,

this has not been necessary.

217
En(MeV)

ZDoo

"V(p.n)"Cr

e'"" = 502 MeV

n.

I

300 400 CHANNEL

Fig. 15. Time-Qf-f light spectrum for neutrons

from ^'V{p,n) ^•'Cr. The dashed curve

is a schematic representation of the

neutron spectrum emitted at a proton bombarding energy of 5 MeV.

The LLL Li Detector

A neutron detector for time-of-f light

2 5 measurements in the 1 keV to 1 MeV energy

range has been designed by Czirr and Shosa'

at the Lawrence Livermore Laboratory. The

design concept is an outgrowth of the detector

designs of W.P. Poenitz . -'^ -' Neutrons incident

through a channel are moderated and captured

in a 4 m of compressed LiH and the capture

distribution is sampled with thin slabs of ^Li

glass scintillator. The mean capture time is

approximately 100 nsec.

Figure 16 shows a schematic diagram of

this detector. The outer dimensions of the

detector were determined primarily by the

available lengths of Li-glass slabs. Slabs

165 mm long can be reliably produced so the

detector length was fixed at 320 mm. The

design provides for use of two pieces of glass

to sample each longitudinal detection zone. Photomultiplier tubes view the glass slabs

end on. The radius of the sampled region used

in the design model was 200 mm. The glass

Fig. 16. Schematic diagram of the totally

absorbing detector showing end-on

and side views.

slabs were taken to be 2 mm in an effort to

compromise between requirements of strength

and good counting statistics and the requirement of small flux perturbations. The glass

slabs were separated by 10 mm in the model.

The radius of the entrance hole was chosen to

be 10 mm in a deuterated-polyethy lene , (CD„)

scatterer with a radius of 14 mm. The optimSm

depth of the re-entrant hole in the (CD^)

scatterer was found to be 115 mm.

The TARNP Monte Carlo transport code^^

which access the LLL cross-section library

was used for the design calculations.

Figure 17 shows the calculated efficiency

for the model detector. The efficiency is flat

over the energy range from 1 keV to 1 MeV. The

error flags are an estimate of the computational uncertainty at each point. The error in

the shape of the efficiency curve is estimated

as 0.7%. The computational accuracy is about

2% as shown in Fig. 17.

1.30

Fig. 17.

I0< I0»

INCIDENT NEUTRON ENERGY (eV)

Calculated efficiency of the LLL

totally absorbing Tirae-of-Flight

neutron detector. Calculation was done with computer program TARNP.

io«

0.8 -

0.6

= 0.4 -

0.2

10 100

TIME (NAKOSECONDS)

Fig. 18. Time response of LLL detector as a function of incident neutron energy.

218
The time response of the detector is shown

in Fig. 18. Essentially all of the neutrons

are captured in 1 usee and about half are captured during the first 100 nsec. This time

response is quite satisfactory for long flight

paths such as encountered at Linacs.

This detector is novel and exhibits good

characteristics. It should be v;ell suited for use at linac neutron sources. The materials

required for its fabrication tend to be expensive .

by Langsdorf, ° one very important region in a collimator is the throat which is the most nar- row part where neutrons are strongly scattered.

The heavy shielding material near the center

of the collimator (Fig. 19) serves two purposes.

It is useful to have strong attenuation in the

collimator throat; and, this material serves

as a detector shield for the 2.225 MeV y-rays

produced by the neutron capture in hydrogen.

The intensity of such y^ays is large near the

neutron source in hydrogen containing collimators .

A Word About Collimators

The black and grey neutron detectors which

have been discussed are capable of ultimate

accuracy of better than 1% for measurement of

neutrons originating in a target at the center.

Usage of these detectors with collimated neutron beams has been extensive. ^"-'-^ In the

first use mode, the neutrons detected are the

primary nuclear reaction products; in the second, neutrons produced in a neutron source are collimated and caused to react with a target.

The neutron flux is measured with the black or grey neutron detector. These detectors can accurately count all neutrons which reach the

central region of the counter near the end of

the re-entrant channel. One of the largest

potential sources of error is the perturbation

of the neutron "beam" by the material between

the neutron source and the detector. When collimation is used this perturbation takes two

forms; first, the resolution function or energy

definition is broadened and made less precise

through energy losses in scattering from air,

the target and the collimator material; and,

second, a general background may be produced

in the collimator materials which affects ac- curacy. An example of this is the production

of 2.225 MeV y-rays in hydrogen containing collimators which will affect detectors sensitive

to Y~rays. Most of these y-rays are emitted

after the neutrons are moderated.

27 Spencer and V-Joolf provide important

guidelines for the construction of neutron collimators. Langdorf's work remains as the

most complete treatment of neutron collimation.

Figure 19 shows a collimator for use at a Van

de Graaff accelerator. As has been emphasized

I I Lij Co, and Paraffin

Copper or Tungsten

Fig. 19. A neutron collimator designed for

keV neutrons.

Tests of a collimator 1 m long at the University of Kentucky by Cochran^^ have shown

that excellent geometrical definition of the

neutron beam can be achieved for neutrons with

energies in the range of 250 - 750 keV. This

work also shows that fewer than 5% of the neutrons in the collimated beam have energies

differing from the primary beam by more than

25 keV. With proper design, it should be possible to further reduce this fraction. Detectors such as the Black Neutron Detector and

the NBS Black Detector with fast time response

reject all but a fraction of a percent of the

wrong-energy neutrons by the time-of-f light

discrimination. Even so these "background"

neutrons remain a problem since their effect

will depend on the target and reaction studied.

Further work on the transmission and beam

quality of neutrons passing through such collimators is needed.

Summary

As counting devices for neutrons in the

energy range between 1 - 1000 keV, the black

and grey neutron detectors show high potential

for ultimate accuracy. High accuracy levels

can also be achieved in the MeV energy range.

Such detectors are useful in two operational

modes. First, they may be used to detect neu- trons originating in their center. Second,

they may be used to count the neutrons in a collimated beam.

With current techniques, the accuracy of

such detectors can be made better than 1% for

neutrons originating at the center of the detector. The accuracy limit with a collimated

beam of neutrons is less well established. It

is known that long, thin collimators provide

neutron beams of high quality; both with re- spect to their geometrical definition and with

respect to energy definition. Ultimate accuracy of measurement of the flux of collimated

neutrons in the keV energy range with current

techniques is estimated at 2%.

In the neutron energy range between 1 keV

and 250 keV, the moderated detectors even with

their corresponding slow time response are among the best detectors for measurement of

neutron flux.

Acknowledgments

I want to express gratitude for the opportunity to make this presentation. Thanks go

to J.B. Czirr, C.H. Johnson, R.L. Macklin, M.M.

Meier, and W.P. Poenitz without whose help the

review would not have been possible. Special

thanks go to W.P. Poenitz.

219
*Work supported in part by the National Science

Foundation and U.S.E.R.D.A.

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14. M.S. Coates, G.J. Hunt, C.A. Uttley, and

E. R. Rae, Reference 6, p. 401 (1970).

15. W.P. Poenitz, "Neutron Standard Reference

Data", International Atomic Energy Agency,

Vienna (1974); W.P. Poenitz, Nucl. Inst,

and Meth. 109 , 413 (1973)

.

16. W.P. Poenitz, Nucl. Sci. and Eng. 53 , 370 (1974).

17. W.P. Poenitz, ANL-7915, Argonne National

Laboratory (1972)

.

18. G.P. Lamaze, M.M. Meier, and A.O. Wasson,

Reference 7, p. 73 (1975).

19. M.M. Meier, Private communication (1977).

20. L. Wishart, R. Plattner and L. Cranberg,

Nucl. Instr. and Meth. 5^, 237 (1967).

21. K.K. Sekharan, H. Laumer, B.D. Kern, and

F. Gabbard, Nucl. Inst, and Meth. 133 , 253 (1976).

22. M. Mutterer, Reference 6, p. 452 (1970).

23. J.H. Gibbons and R.L. Macklin, Phys. Rev.

114 , 571 (1959)

.

24. J.M. Blatt and V.F. Weisskopf, Theoretical

Nuclear Physics , John Wiley, New York

(1952) , p. 365 ff

.

25. J.B. Czirr and D.W. Shosa, " A Totally- Absorbing Time-of-Flight Neutron Detec- tor . " Lawrence Livermore Laboratory,

Livermore, California 94550. Preprint

UCRL-79089.

26. Ernest F. Plechaty and John R. Kimlinger,

TARNP : A Coupled neutron-photo Monte

Carlo transport code, UCRL-50440, Vol.14

(1976)

.

27. L.V. Spencer and S. Woolf, Nucl. Instr.

and Meth. 97_, 567 (1971).

28. Alexander Langsdorf, Jr., in Fast Neutron

Physics , Part I , J.L. Fowler and J.B.

Marion, eds. (Interscience Publishers,

Inc., New York, (1960) p. 721.

29. J.L. Cochran, M.S. Thesis, University

of Kentucky (unpublished)

.

220
ASSOCIATED PARTICLE METHODS

Michael M. Meier

Developments in the associated particle method for the last ten years are summarized

with emphasis on the reactions 3 H(d,n)423He, H(d,n) He and 33H(p,n) He. Recent progress on

the associated particle calibration of a black detector in the energy range 250 to 1000 keV

is reported. Current accuracies for the time uncorrelated and time correlated approaches

are noted and prospects for future improvement are estimated.

(calibration; efficiency; neutrons; neutron beams; neutron flux; ^H, ^H, *He, ^He,

protons)

Introduction '^H(d,n)'^He Reaction

The associated particle technique (APT) has been . used as a monitor of neutron flux for about thirty

years and has been discussed in several reviews of

1" 3 neutron flux monitoring techniques. This survey

includes major developments in the method utilizing the

^H(d,n)*He, ^H(dn)^He and ^H(p,n)^He reaction from 1970

to the present. These reactions have been used for

flux monitoring at energies between about 200 keV and

25 MeV.

The APT has been used in two rather different

ways and there are distinct advantages associated with

each. In the first, the charged particle reaction

products are collimated and counted in a detector, and

the neutron flux is then calculated for the corresponding neutron angle. This method has the advantage

that the electronics can be rather simple, needing only

to identify the charged particles with which the neutrons are associated. Also, if the neutron flux does

not change rapidly with angle, or if its angular behavior is well known, then the neutron detector or sample

can subtend a large solid angle. This approach is

also useful when the neutron detection system is not

amenable to fast timing techniques or when high count

rates associated with large solid angle are necessary.

Essential to the success of the method is accuracy

in measuring the effective solid angles for both the

neutron detector and the charged particle collimator.

Angle measurement is also critical, since the differential ratio of solid angles enters into the calculation and may be a rapidly varying function of angle.

If the neutron detector subtends an appreciably larger

solid angle than the one associated with charged particle detection a correction dependent on the differential cross section may be necessary. A further

problem is the correction for background neutrons

which are not associated with the direct flux.

The second technique correlates the detected

neutron with the charged particle of interest by

coincidence or time of flight methods. One is restricted obviously to detectors which have fast response,

although in principle they need only be fast relative

to the data acquisition rate. The neutron detector

must also be larger than the associated neutron cone, which normally implies that the extent of this beam

must be experimentally determined.

This usually restricts the size of the charged

particle solid angle and with it the associated neutron

production rate. It is also clear that this is a calibration of only a part of the detector and uniformity

may be a problem. Advantages which accrue are the

elimination of geometry determination and suppression

of neutron backgrounds by coincidence or time of

flight. Once the experiment is properly set up there

is a minimum of calculation and correction to the data.

The efficiency of a detector is simply the ratio of

the correlated rate in the neutron detector to total

associated particle counts. In the following, these

two methods will be referred to as the " time-uncorrelated" and "time correlated" methods respectively.

Several properties of this reaction make it at- tractive to use for absolute flux monitoring. The + 17.6 MeV Q value provides a source of 14 MeV neutrons

and associated 4 MeV alpha particles which are therefore well separated in energy from scattered deuterons

at bombarding energies of only a few hundred keV. The

attractiveness of this technique to a laboratory with

a low energy accelerator is enhanced by the 110 keV resonance with a peak value of about 400 mb sr Also other (d, charged particle) reactions in target

components, e.g. titanium or zirconium and backing

materials, which would provide background for the alpha

particles are strongly suppressed by the Coulomb barrier. There are also serious difficulties encountered

in the use of low bombarding energies and thick

tritiated targets. First, the identification of alpha

particles is complicated by the presence of an intense

Coulomb scattered deuteron background which is inversely proportional to E^. If detected, these deuterons

can cause saturation of electronics, baseline shifts

due to pulse pileup and other problems associated with

high rates. One simple technique* to eliminate this

problem has been to interpose a metal foil between

target and detector. The foil thickness is selected to be slightly greater than the deuteron range, permitting

the alpha particles to pass through with slightly increased straggling.

Another problem arising from Coulomb scattering is

the broadening of the neutron cone beyond the limits

kinematically defined by the alpha detector. At low

bombarding energies this effect is largest and predominately due to deuteron scattering. For experiments in

which the neutron beam must be smaller than the detector, this effect must be taken into account.

Another set of problems arises because of the

time dependence of the tritium distribution in a solid

target. For these targets, where the total beam power

is dissipated in the target and substrate, tritium is depleted from the surface layer and the effective mean

deuteron energy is decreased. For a fixed charged

particle detection angle, the associated neutron angle

increases (see, for example Figure 2, Ref. 4) and

thereby causes misalignment of the neutron detector

relative to the beam center. This problem is not usu- ally important for a time uncorrelated experiment,

since the laboratory (d,n) cross section varies by less

than a percent for a typical five degree change in

neutron angle.

Finally, for time uncorrelated experiments, background neutrons may result from scattering in the tar- get and beam line structures and from production via

(d,n) reactions. In particular, drive in deuterons are always a source of background for experiments using a solid target. References 5, 6 and 7 are recent examples

of the APT with low energy deutrons for measurement of

fission cross sections, (n, charged particle) spectrometry and for evaluation of collimetor design.

If one wishes to extend the energy range of neutrons away from 14 MeV the use of higher bombarding
energies becomes necessary. Although the Coulomb scattering effects are thereby reduced, other difficulties

become important. Specifically, beam power is usually

increased, accelerating the tritium loss and the conse- quent neutron angle change. Also, other (d, charged

particle) and (d,n) reactions contaminate the alpha

particle and neutron spectrum respectively. A unique

scheme employed by Cookson et al.^ utilizes a tritium

gas target chamber to obtain neutron energies between

14 and 25 MeV. Many experimental difficulties are eliminated by utilizing such a system. First, problems

associated with the target backing disappear. The

Coulomb scattering, neutron beam broadening and neutron angle migration are all essentially eliminated.

Background in the charged particle spectrum is reduced

since the alpha particle collimation shields the sur- face barrier detector from reactions in the entrance

and exit foils and from the beam dump where (d charged

particle) reaction products would be produced. This

apparatus has been used to obtain an absolute calibration for the efficiency of a 2.5 cm thick by 10 cm diameter NE-213 scintillator. Neutron background originating in foils and beam defining slits is small and

easily corrected since the experiment uses the timeof-flight method for obtaining the detector efficiency.

^H(d,n)'^He Reaction

Many of the remarks concerning the ^H(d,n) reaction at low bombarding energies apply to the ^H(d,n)

reaction as well. For acceleration voltages lower than

500 keV monoenerge tic neutrons spanning the energy

range 2 to 3.5 MeV can be produced with appreciable

yield. In at least two important respects this is a less favorable reaction. First, the competing

^H(d,p)^H(Q =: 4.033 MeV) complicates the spectrum of

^He from the (d,n) reaction (Q = 3.269 MeV), the triton

and "^He differing in energy by only a couple hundred

keV for energies below 500 keV bombarding energy.

Second, the ^He energy is usually less than 1 MeV,

giving rise to Coulomb scattering problems for the re- coiling particles as well as the bombarding deutrons.

Apart from the degradation of energy resolution in the

^He peak, this effect also contributes to the cone broadening and migration. Finally, the cross section

for this reaction is only 20 mb sr"''' at 500 keV and

increases with increasing energy, becoming 90 mb sr""'' at 10 MeV bombarding energy.

Use of a deuterium gas target for this reaction

has been employed with the chamber discussed above.

The benefits of a gas target and higher bombarding

energies are the same as those discussed for the

H(d,n) reaction plus the fact that the yield increases

with increasing energy. A transmission surface barrier

detector is used in this experiment in order to detect

the He particles in the presence of tritons with

almost identical energy. The thickness of the detector

is selected to stop He particles and allow the tritons

to pass through, losing only a fraction of their energy.

Breakup and other reactions that produce neutrons are not a problem since the experiment time correlates

charged particles and neutrons.

Applications of the APT have been made at bombarding energies below 500 keV by employing electrostatic® and magnetic^ analysis of the reaction products to eliminate scattered deuterons and tritons.

Foils were used to stop the scattered deuterons for an experiment^ '• at 150 keV bombarding energy although

background in the charged particle spectrum is a difficult problem under these conditions.

Bartle and Quin have used a surface barrier detector AE - E pair.-'-^ The AE detector has a thickness

equal to the range of the ^He particles. Tritons and

elastically scattered deuterons are transmitted and

detected in the E detector which produces a veto for

data accumulation. To reduce the Coulomb scattering a deuterated polyethylene target 0.5 ^im thick is used in

a transmission position. Target rotation extends the

lifetime to about 100 hrs with a 0.5 nA beam collimated

to a 2 mm diameter. Background in the charged particle

spectrum becomes a problem when the ''^C(d, Uq) ^°B(g.s.)

alpha particles begin to interfere with the -^He peak.

This background can be reproduced by using a carbon

film of similar thickness in place of the polyethylene.

A simple stripping technique is then sufficient to separate the -^He peak. This system has been used to produce calibrated neutron beams from 2 to 10 MeV.

Other applications of the ^H(d,n) reaction include

a calibration^^ of a neutron detector in the energy

range 25 to 60 MeV. Also, a novel approach to the time uncor related method has been used by Ryves and

Sharma.14 Here the ^H(d,p) protons are monitored at an angle where (d,p) and (d,n) cross sections are comparable.

^H(p,n)^He

This reaction has been used extensively in the

past decade, its attractiveness stemming largely from

the possibility of monitoring flux at energies as low

as 100 keV. To obtain energies this low the associated

"^He particles must be detected at angles on the order

of 10°, where the ratio of Coulomb to (p,n) cross section is 10^ . It is necessary to cope with these conditions in order to obtain He particles with sufficient

energy to escape the tritiated target and be detected

unambiguously. Under these "favorable" conditions the

energy spread of 1 MeV He's is 15% for a 100 ng/cm^

titanium layer. Again, the neutron cone is broadened,

this time almost exclusively on account of the recoil

particle scattering. Apart from the large elastic

proton background, inelastic protons from the target

and collimating surfaces and tritons from elastic

scattering H(p,t) are present in the charged particle

s pec t rum.

Fort and co-workers''^ ^ have used this technique for

measurements in the range 100 to 500 keV. Transmission

targets 200 i^g/cm' thick are used and by employing se- quential electrostatic and magnetic fields the separation of ^He particles from background is very clearly

accomplished. The solid angle for this technique is

rather small, limiting the count rate, but providing a consequent advantage. The apparatus can be turned to accept a very limited band of He energies and in this

way actually select those He particles which have

undergone a minimum of Coulomb scattering in the target. By operating in this fashion the angular and

energy spread of the neutron cone can be minimized.

This system has been used to calibrate the efficiency of a Li - loaded glass scintillator. The mea- surement was complicated by the necessity for making a large multiple scattering correction in the glass to extract the ®Li cross section. This calculation included a Monte Carlo calculation of the Coulomb scattering of the "^He and its effects on the energy spread and

angular broadening of the associated neutron beam. The

largest uncertainty in this measurement was for the

conversion of the efficiency as measured at a distance

of 5 cm from a point source to that which would result

in a parallel beam.

Liskien, Paulsen and co-workers'^

® ' ''^^ have com- bined an electrostatic deflector and a pulsed beam

time of flight system to produce calibrated beams

between 250 keV and 1 MeV. The deflection is used

to suppress the elastic proton rate by sweeping the

222
^ He into an offset surface barrier detector. A gate

is generated by particles which have the source to detector flight time appropriate for a °He particle.

This logic signal is then used to generate gated pulse

height spectra for the surface barrier detector.

Experiments done with this apparatus have been

done using the time uncorrelated approach. As previously mentioned, the calculation of flux for this case involves determination of the effective solid angle

for charged particle collimation, solid angle of neutron detection and ( dOs ^j^/dO^) . That careful measurement of these quantities is essential is illustrated

by the fact that (dOa^^/dfi^) varies by 5°L for a one

degree change in the ^ He detection angle. The use of

deflection complicates the time uncorrelated measurement in two ways. First, assurance must be obtained

•a ++

that all the He which are collimated into the

deflection system are, in fact, detected by the solid

state detector. This can be ascertained by varying

the deflection voltage and operating on the plateau of

counting rate. Also, since neutral and sing ly- ionized

'He are associated with the flux at the neutron detector, a correction of order 107. must be made. The lat- ter can be partially checked by experiment. By varying the deflecting field and using a 2 mm collimator

Paulsen, Liskien and Cosack""^^ have measured the relative ^He to ^ He yields with their associated particle system. These measurements agree well with

calculation and give an added measure of confidence to the correction. Excellent agreement was obtained

using this method to calibrate a proton recoil proportional counter.

Leroy has used the associated particle apparatus

described earlier in the time uncorrelated way.^^ By

properly tuning the electrostatic - magnetic analyser,

it is possible to eliminate to first order the effects

of spatial spread of the "^He"'"'" beam which is due to its energy dispersion. Because of this, the effective

solid angle of the ^ He beam can be increased and a larger corresponding solid angle of neutron flux can

be measured. Count rates a factor of 50 to 100 higher

than with the time correlated technique are possible

with this method. The system has been used to check

the calibration of a long counter which had been previously calibrated with a MnSO^ bath. The calibrations

are in good agreement within the - 2°U error bars.

At NBS we have used a setup similiar to Liskien's,

employing pulsed beam and electrostatic deflection to identify the ^He particles. Two chambers for detection

of recoil particles at 10° and 25° are used to cover the

range of neutron energies 250 keV to 900 keV. The

Coulomb scattering is not severe at 25 , and no electrostatic deflection is used there. A two parameter

data analysis system is now used for obtaining the He data, providing an easy to use method for background

determination. Figure 1 shows such a spectrum from an experiment where electrostatic deflection is used.

Each of the 32 time channels has a width of about 7ns

and is comprised of a 128 channel pulse height spectrum. The background under the peak is about 57o of

the ^He rate and can be very accurately determined by

gating the spectrum with a logic signal generated by

neutron detection. This method provides a powerful

technique for determining the ^He lineshape. It is

especially useful in cases where no deflection is used

and the ^He peak can be incompletely resolved from

tritons from proton elastic scattering, T(p,T).

\ ACHiRGED PARTICLE SPECTRUM

\ n Ep > 2 2 MeV

Figure 1. Charged particle spectrum for 2.2 MeV

protons. The resolved group is -^He from

='H(p,n)3He

The system has been used to calibrate a black detector " which has a response calculable by Monte Carlo

technique. For the calibration the detector is placed

at the appropriate angle with the front face 11 cm from

the target. The anode signal from the photomultiplier

is supplied to two linear gates which are gated by

logic signals generated by ^He events. The first gate

operates normally, opening for 200 ns to permit the

neutron event to be analysed when an associated ^He

event occurs. The second gate opens one microsecond

earlier corresponding to the preceding proton pulse

and thereby measures accidental background that is not

physically correlated with true ''He events. •

After subtracting this background, the reduced

spectrum is normalized in area and channel width so as

to be directly comparable to the Monte Carlo calculation. The program used is "CARLO BLACK"^° by W. P. Poenitz modified to include the effects of Poisson

statistics for photoelectron production at the photocathode. Figure 2 shows data taken at 250, 500, 700

and 880 keV and the corresponding Monte Carlo calculations. The data were all obtained under the same con- ditions and in particular no gains or bias voltages

were changed in the neutron electronics. In reducing

the data, the same bin width conversion factor was used

for all energies so that the agreement between experiment and calculation could be compared as a function of

energy. The Monte Carlo generated response functions

are shown as smooth curves for each energy in the figure. Two curves are shown for each energy, corresponding to levels of photoelectron production which differ

by 50% for a given light output.

These comparisons are very encouraging in that

calculation and experiment qualitatively agree for a single set of parameters for all energies. The spreading of the peak seems to be well described by a single

Poisson parameter between the two shown here and the

peak amplitude shifts as calculated. This latter

indicates that the light tables used are valid in a relative sense over this energy region. The efficiencies for the detector when a bias corresponding to channel 20 is chosen are compared for calibration and

calculation in Figure 3. The agreement seems acceptable at the 2.57o level.

The black detector is now being used as a flux

monitor in absolute cross section measurements and for

dosimetry at the NBS Van de Graaff. Our confidence in

the system is enhanced by the understanding we have

obtained in bringing experiment and calculation into

the detailed agreement shown above.

223
2000

<

I

o

CO

>

UJ

1000

./Ll

^1

BLACK DETECTOR CALIBRATION

ASSOCIATED PARTICLE RESULTS

o 2 50 keV

A 500 keV

o 700 keV

• 880 keV

MONTE CARLO CALCULATIONS

22 5 keV / PHOTOELECTRON

34 keV/ PHOTOELECTRON

2° ^0 60 80 ^0 ^0 WO fso 180 200 220 240 260 280 300 320 340 360

CHANNEL NUMBER

Figure 2. Experimental and calculated response functions for the black detector

Cone lus ion

0.96

0.94 >o

z

UJy 0.92

Oo 0.90

0.88

0.86

100

BLACK DETECTOR * EFFICIENCY

X CARLY CALCULATION

° ASSOCIATED PARTICLE

CALIBRATION

300 500 700

En (keV)

900

Figure 3. Measured and calculated detector

efficiencies for the black detector

It is perhaps worthwhile to summarize by reviewing

the present APT accuracy and to speculate on future

improvements. The best accuracies obtainable with the

time uncorrelated method have been on the order of 37„. A major limitation of this method, as mentioned previously, is the difficulty in determining the solid an- gles for neutron and charged particle detectors. Uncertainties for the latter are quoted to be 0.57„. A

neutron detector in a divergent beam with finite source dimension may be even more difficult. Without some breakthrough in this area it is unlikely that accuracies much better than 17, overall will be obtained.

For the time correlated method the accuracy is

limited by the details of the particular reaction con- sidered. For ^H(d,n) Cance and Grenier estimate their

uncertainty in the APT to be 0.1%, only dependent on counting statistics of the alpha particles. This impressive accuracy is a good bit better than can be

expected for sample assay or efficiency calculation in

the near future. It is probable that experimenters

will therefore attack these latter problems rather than

improve this accuracy level.

The ^H(d,n) reaction with a gas target has the potential for a very high ultimate accuracy. ^ He identification is quite unambiguous with detectors that transmit the competing tritons. Also, the Coulomb scattering which disperses the neutron beam is due only to the tritium and not to any intervening foils. The

accuracies obtained in a scintillator calibration have

224
been about 27=, due mostly to uncertainties in extending

the calibration over the complete surface of the detector. Presumably the accuracy could be considerably

improved in the calibration of a detector with a re- entrant hole or in a cross section measurement where

sample assay and uniformity were well known.

The ^H(p,n) reaction has a current accuracy of 1 -

37. using the time correlated approach. Some of this

uncertainty is due to problems not inherent to APT, but

to sample and geometric difficulties. A good example

is the non- trivial problem of converting an efficiency

in a divergent beam back to the normal parallel beam

situation. APT measurements will, as a rule, have such

detector or sample problems and it would be inappropriate to attempt a listing of all of them here. There

are tv/o areas of concern specific to the APT which are common to all "^H(p,n) applications (and to some extent

the other two reactions) and which now limit our use of

the method. First is the identification of charged

particles. This has been done with an accuracy of 0.17. using very sophisticated deflection systems''" but at a large sacrifice in count rate. At NBS we now cope with

a 57o background known to 207, accuracy and can obtain

with some effort 17. - 0.57..^''^ The technique of using

the neutrons to gate the charged particle spectrum and

thereby determine the lineshape has been mentioned

above. We plan to combine this method with a computer

data collection system that utilizes "tagging" for very

accurate determination of background. It is likely

that we can obtain better than 0.27. accuracy in charged

particle background in this way.

A second problem is somewhat less tractable. Fig- ure 4 shows the angular profile of our neutron beam for

two different targets with a ^He energy of 1.2 MeV

10

>-o

z

UJo

UJ>

UJ

IT

10

10c3

10'

~1 '

\ ^ NEUTRON BEAM

PROFILE

En =250 keV

E, =1.17 MeV / 'He

lOOMg/cm

35 fiq/cm

120 130 140 NEUTRON ANGLE , DEGREES

150

Figure 4. Spatial profile of associated neutron beam

for two targets of different thickness

where the Coulomb scattering is rather severe. A problem that ultimately limits accuracy is the question of

how many associated neutrons are found outside the cone defined by a given detector or sample. Such neutrons

arise from scattering by the chamber walls and by cone broadening due to Coulomb scattering. The former are usually calculable, although for our 0.8 mm aluminum

wall the correction is on the order of one percent.

One way to experimentally check on the Coulomb

scattering problem is to perform the measurement with

differing target thicknesses. We have checked our black detector calibration and do not find differences

at the 27. level for the targets represented in the

figure. Tuning the electrostatic analyser to select

^He's which have undergone different amounts of Coulomb

scattering might also provide some measure of this

effect. It is clear that both the above problems could

be diminished by the development of new targets, for

example, tritium gas with very thin windows or tritiated polyethylene.

References

1. H. H. Barschall, L. Rosen, R. F. Tascek and J. H. Williams, Rev. Mod. Phys . 24, 1 (1952).

2. J. E. Perry, Fast Neutron Physics, Vol. 1, J. B. Marion and J. L. Fowler, Interscience Publisher,

New York (1960).

3. R. Batchelor, Neutron Cross Sections and Technology, NBS Special Pub. 299, Vol. 1, Washington,

D. C. (1968).

4. H. Marshak, A. C. B. Richardson and T. Tamura,

Phys. Rev., 150, No. 3, 996-1010 (1966).

5. M. Cance and G. Grenier, Trans. Am. Nucl. Soc. 22,

664-665, Nov. 1975.

6. C. Sellem, I. D. Perroud and J. F. Loude , Nuclear

Instruments and Methods 128 , 495 (1975).

7. K. M. Jones, E. P. Cytacki and C. S. Kelsey, Phys.

Med. Biol., 20 (1), 131-135 (1975).

8. J. A. Cookson, M. Hussian, C. A. Uttley, J. L. Fowler and R. B. Schwartz, Neutron Cross Sections

and Technology, Vol. 1, pp. 66-68 (1975).

9. R. B. Galloway and A. Waheed, Nucl. Instr. and

Methods 128, 505 (1975).

10. P. B. Johnson, J. E. Callaghan, C. M. Bartle

and N. G. Chapman, Nucl. Instr. and Methods,

100 , 141-148 (1972).

11. A. Greil and K. Trutzschler, Kernenergie, 12,

58-70 (1969).

12. C. M. Bartle and P. A. Quin, Nucl. Instr. and

Methods 121, 119-127 (1974).

13. R. A. J. Riddle, G. H. Harrison, P. G. Rose,

M. J. Saltmarsh, Nucl. Instr. and Methods 121 , 445 (1974).

14. T. B. Ryves and D. Sharma, Nucl. Instr. and

Methods 128, 455 (1975).

15. E. Fort, Nuclear Data for Reactors (Proc. Conf.

Helsinki: 1970) 1., IAEA, Vienna (1970); E. Fort,

J. L. Leroy and J. P. Marquette, Nucl. Instr. and

Methods 85, 115-123 (1970).

225
16. H. Liskien and A. Paulsen, Nucl. Instr. and

Methods 69, 70-76 (1969).

17. A. Paulsen, R. Widera, A. Berlin and A. Trapani,

Nucl. Instr. and Methods 91, 589-593 (1971).

18o A. Paulsen, H. Liskien and M. Cosack, Nucl. Instr.

and Methods 105, 103-107 (1972).

19. G. P. Lamaze, M. M. Meier and 0. A. Wasson,

Nuclear Cross Sections and Technology, Vol. 1, pp. 73-74 (1975).

20. W. P. Poenitz, ANL-7915, Argonne National

Laboratory (1972).

21. M. M. Meier, A. Do Carlson and G. P. Lamaze,

Nuclear Cross Sections and Technology, Vol. 1,

75-77 (1975).

22. J. L. Leroy, I. Szabo and J. Y. Tocquer, Neutron

Standard Reference Data, IAEA, pp. 63-73 (1972).

226
ASSOCIATED GAMMA-RAY TECHNIQUE FOR NEUTRON FLUENCE MEASUREMENTS

by

J. D. Brandenberger

University of California

Los Alamos Scientific Laboratory

Los Alamos, New Mexico 87545

The use and development of the ^ Li (p,n q^y)^ Be reaction as an example of the

associated gamma-ray technique for neutron fluence and detector efficiency

measurements are described. Present limits on energy range and accuracy are stated for this method, and current extensions of this work are oiscussed.

(Neutron fluence; detector efficiency; cross sections)

Introduction

Associated gamma-ray techniques have been used ex- tensively in nuclear physics. Two examples are the

gairnia-gaitma coincidence method that has been used for

decades in analyzing gamma-ray cascades of excited

nuclei in nuclear spectroscopic studies, and coincident measurement of annihilation radiation to study

the interaction of positrons with matter. Several

years ago, with the advent of high resolution Ge(Li)

detectors, it became convenient and enhanced experimental precision to monitor particle induced neutron

producing reactions by a gamma-ray associated with a neutron group. With respect to neutron measurements,

the 431-keV gamma-ray from the ^Li (p,n^Y) ^Be reaction

has been used at Lowell University to monitor the

relative fluence from the ^Li(p,nQ)^Be reaction.

While this gamma ray is not actually associated with

the n^ neutrons, the principle is similar because at

a given proton energy, the branching ratio of nj^/ng

neutrons is constant. The closest example to the

technique described here is the use by Presser and

Bass^in determining both the ^Li (p,n-|^Y) ^Be to ^Li (p,p^Y) ^Li branching ratio and the integrated

cross sections by observing the 431- and 478-keV

gamma rays, respectively. That work, however, re- quired proton fluence and target thickness measurements in order to obtain the absolute total reaction

cross sections. The present work uses associated

gamma-ray techniques in conjunction with a neutron

source reaction to measure angular distributions

which for simplicity and accuracy can be used to measure the neutron fluence and detector efficiencies while avoiding target thickness and proton

beam current measurements.

Use of the Method

For the moment, let us assume that a set of

^Li (p,n-j^Y)^Be angular distributions as a function of

proton energy are known with good precision either on

an absolute or relative scale. The measurement of

these angular distributions must be made before the

associated gamma-ray method can be used with this

reaction. This subject is left for later discussion

because of its critical relation to the accuracy of

the method.

Figure 1 shows the spectroscopic information involved in the example reaction. The n-j^ neutron

leaves ^Be in the 1/2", 431-keV level, which promptly

decays to the 3/2" ground state. Because of the spin

values involved, the decay of the level is by the

isotropic emission of the 431-keV gaima ray, and the

ratio of the number of these gamma-rays to n^ neutrons is unity from threshold until the 4.5-MeV level

is reached. Thus, from 0 £ E^^ 5 4 MeV one may monitor the n-|^ neutrons by observing the 431-keV gamma

ray. If the absolute efficiency of the gamma-ray detector has been determined for the 431-keV gamma ray

by comparison to gamma-ray standards that are now available from, for example, the National Bureau of

Standards (NBS) , the absolute n-^ neutron fluence may

be determined at each proton energy and detector

angle. A high-resolution gamma-ray detector must be

used in order to resolve the 431-keV gamma rays.

If the objective is to determine the absolute

fluence of n^^ neutrons, the above measurements are

all that is necessary. To calibrate a neutron

detector with this reaction, it is necessary to make

time-of-flight measurements of the n^^ neutrons

because of the presence of neutrons, gamma rays,

and room-scattered background. Subsequently, a second detector may be calibrated for monoenergic

neutrons by direct comparison with the first if only

monoenergetic neutrons are present. The method was developed to measure detector efficiencies as a function of energy with a combination of ease and

precision that has not previously been possible in

this energy region (E < 4.1 MeV) . The ultimate

motivation, of course, is the measurement of neutron

differential cross sections normalized to the n-p cross section at laboratory angles of about 40°.

Figure 2 shows a set of ^Li(p,n2)^Be angular distributions, 3 which are considered for the present to be precise. We can use these angular distributions,

or we can use the information ^ in Fig. 3 to generate

angular distributions from interpolated polynominal

coefficient ratios.

Each angular distribution can be expressed as \^ measurable max W{e,E ) P e, = 0

A„ {9,E )P„ (cose

A = A ,

where A„/A is known, and

IL o

W(9,E ) = A (E ) P op

The total number of n^^ neutrons N emitted by the source is obtained by dividing the number of

detected 431-keV gamma rays, Ny, by the gamma-ray

detector's absolute efficiency, e^. Thus,

N = Ny/Zy

or N = Kj'w(6,E^)dfi = N^/e^.

The fluence at a particular energy and angle is

Kl'J F(e,Ep) = (e,E

227
5/2" 7480

(5^ ) ^Li+n 6560

3/2"

4630

CM

10

ino

Q

o

+

1/2- 4774>r^„o-,4,

o.o;

(5/2' 7190

5/2" 6510 b606A-

"^Li*p 7/2' 4500

1/2" -a- 431

i3/2'

'Be

0.0

2.7x|0-'^s

53.28 d

'Li

9000

8000

7000

6000 — 5000

4000 — 3000

2000

• 1 1400

1200

1000

800

600

400

200

0

Fig. 1

The level and decay schemes of ^Li and ^Be.

228

0.0

-0.2

-0.4

0.4

0.2

o

<

"=-1 0.0

<

0.2

0)

CO 0.4

0.0

-J 0,2

0.4

1 \ I [—1 \

I

I r A^3 /A^0

1 •

I I I I I . I .

I I .

• i

• * • •

"^LK p,n, )^Be

Elbakr et al • Present work

> I I ! I i_J I L_j L

-0.6

J L

A ./A 0

• # i •

J L_i. I I I J \ I L I I I,

3.0 3.2 3.4 3.6 3.8 4.0 4.2 4.4 4.6 4.8 V.O

Ep (MeV)

Fig. 3

The Legendre Polynominal ratios that fit the angular distributions of Fig. 2. The dots are from Ref. 3 and the squares from Elbakr et al.

230
where rz^^and are the solid angle subtended by the

detector and its effective area, respectively. Thus, we simply have w(e,E ) F(0,E ) = N /e ^-t: P ^ ^ 4Tr -Ao

where the distance from the source is L. Note that

is a property of the gamma-ray detector that has

been measured, appears as the 1/R dependence of

the flux, and W/Aq is determined by the known ratios

of the polynominal coefficients. The only quantities

measured for each fluence determination are N ^ and 9,

which are extremely easy to determine.

count, assuring that each point on an angular distribution is monitored to a constant integral fluence of

n-^ neutrons. If an efficiency curve is found which

fits both the data points of each angular distribution and each of the integral cross sections, it is

assumed to be correct. This assumes that the efficiency curve is smooth or that enough data are taken

to establish any fine structure in the efficiency

curve. The precision of the efficiency curve is

taken to be not much greater, say 1.5 times greater,

than the standard deviation of the data points from

the assumed efficiency curve. There is no obvious

means to determine a valid statistical uncertainty in

the usual sense for this method.

To bring us to a final result, the absolute efficiency of a neutron detector whose yield is YCE^^)

is

e = Y(E^)/F(e ,Ep)xS

for a detector with effective (or nominal) area S,

and the relative efficiency is

e , = Y(E )/N xW(e,E )

.

rel n Y P

For determining Y and N^, it is assumed that corrections for attenuations by the target, target backing,

and target assembly have been made appropriately as the experimental arrangement requires. To obtain a detector efficiency in the range of applicability of

the method with the ^Li(p,n-|^Y) Be reaction, i.e. 0 to

4 MeV, typically two or three angular distributions

of the ^Li (p,n-]^Y ) ^Be reaction are necessary. This

can be seen in Fig. 4, where the angular distributions denoted by A, F, and L have overlapped nicely

in energy to give an efficiency curve. By knowing

the efficiency e, both elastic and inelastic neutron

cross sections can be normalized to the n-p cross

sections. If the absolute efficiency is determined

as outlined, (p,n) , (a,n), (d,n) and in general all

neutron source reactions below 4 MeV can be determined provided target thicknesses and appropriate

particle currents are known. It is important to realize (1) that the standard deviation of the points

from a smooth curve (and probably also the correct

efficiency curve) is <2%, which is probably indicative of the accuracy as well as precision of the

results, and (2) that the results are independent of

detailed knowledge of (a) the gamma-ray detector's

position and effective area, (b) the proton current

or target thickness, or (c) the neutron detector's

area.

Present Status

Figures 2 and 3 indicate the sole source of com- pleted measurements using the associated gamma-ray

method. These are from Ref. 3. A statistical analysis indicates that the standard deviation of the set of data points from the assumed efficiency curve derived by the fitting and normalizing procedure is

1.9% in the region above E^^ = 0.4 MeV. Many alternate efficiency curves have been considered and have

been rejected as being erroneous or at least inferior

to that shown. Thus, considering the average and

individual errors as 1.5 times the standard deviation

over most of the energy range gives an uncertainty of

about 3% from 0.6 MeV to 2.75 MeV. The higher energy

region, where only one angular distribution is represented, may have slightly larger errors, but this angular distribution is required to be compatible with

a large number of points from several angular distributions at lower energies and is consistent with the

behavior of efficiency curves well past the maximum.

Further Developments in Energy Range and Accuracy

Further work is underway by Ron Harper, E. C. Hagen, B. D. Kern and the author to extend the energy

range and accuracy of the method. Data have been

taken, and continue to be taken by Harper and Kern,

to extend the range from 70 keV to 4.1 MeV with an accuracy of 1-2%. The difficulties are (1) maintaining a very low detector threshold at a constant

value, (2) maintaining appropriately thin targets for

energies near the neutron threshold that are uniform,

and (3) integrating the n-|^ neutron time-of-flight

peaks in a consistent and precise manner. These difficulties may become real obstacles as the work is

extended to lower energies.

Conclusions

Method of Establishment of the Primary Standard

To establish th^ standard^for neutron flux determination using the Li(p,njY) Be reaction, it has

been shown in the previous section that a set of an- gular distributions is required for use with the as- sociated gamma-ray method. These angular distributions may be relative with respect to angle, but each

angular distribution must have the same normalization

factor to the absolute value. Such a set of angular

distributions must be established concurrently with

the neutron detector's efficiency by measuring a large set of such angular distributions and demanding

that the gross redundancy in measurements at the same

or very nearby neutron energies be consistent with a single detector efficiency curve. The method is outlined in detail by Brandenberger et al. ^ In principle the establishment of these angular distributions is simple, in practice it is very time consuming. Each point on all the angular distributions

is normalized to a constant associated gamma-ray

It appears that the present work is accurate to = 3% for energies above 600 keV. Such accuracy as indicated by the data cannot be accepted without

equally rigorous substantiation by independent ob- servers. In any case, the method, per se, opens the

possibility of measuring differential cross sections

in the 0.1 to 4.1 MeV energy range to accuracies

heretofore not attained, say 1.5 to 3%. Several

other uses quickly come to mind. It should be possible to:

1. further develop secondary standards such as the

C(n,n)C reactions,

2. make more precise measurements of neutron-producing reactions, such as the T(p,n)^He reaction,

using a calibrated detector.

231
I

I

/

•• -I'

I

I

>

/ i: I

/ J» 8 • iti z

E

K

Ihi

-."

. ^

- Jl

v.

Fig. 4

The efficiency curve of the neutron detector used in Ref. 3. Each set of letters

represents a 12 or 13 point angular distribution at a given E . P

232
3. calibrate other detectors including counters of

the McKibben long counter type, the moderating sphere

type, "^He and "^He counters, etc., as well as scintillatiOTi counters operated in the time-of-f light mode.

References

1. A. Mittler, Lowell University, private communication, April 1973.

3. J. D. Brandenberger , F. D. Snyder, J. D. Dawson,

T. W. Burrows, and F. D. McDaniel, Nucl. Instr

.

Methods 138 (1976) 321.

4. S. A. Elbakr, I. J. Van Heerden, W. J. McDonald,

and G. C. Neilson, Nucl. Instr. and Methods 105

(1972) 519.

2. G. Presser and R. Bass, Nucl. Phys. A182 (1971)

321.

233
ASSOCIATED ACTIVITY METHOD

K. K. Sekharan

Cyclotron Institute

Texas A&M University

College Station, Texas 77843

A brief description of the associated activity method is given and its application for the

calibration of flat response neutron detectors in three laboratories is described. The

uncertainties involved in the associated activity method have been discussed and suggestions

for improving the accuracy of the efficiency for neutron detection have been made.

(Description of Method; Possible Improvements; Uncertainties in Efficiency)

Introduction

A neutron has to be detected always by some nuclear process such as '^B(n,a)''Li reaction or (n,p)

scattering. As the nuclear processes are energy dependent the efficiency of the neutron detector for

detection of neutrons also tend to be energy dependent. Hence the efficiency of a neutron detector has

to be determined experimentally and/or by theoretical

calculations as a function of the neutron energy.

The "Associated Activity Method" is a precise experimental method for determining the efficiency of a neutron detector. A survey of the literature shows

that this method has been employed for detector calibration in at least three laboratories. Recently,

K. K. Sekharan, H. Laumer, F. Gabbard and B. D. Kern

have determined the efficiency of a spherically

shaped 4it neutron detector''' using associated activity

method. They have used ^^V(p,n)^^Cr and ^''Fe(p,n)^'Co

reactions for the determination of the absolute efficiency of the detector. W. P. Poenitz has calculated

2 the efficiency of the grey neutron detector by calculating the intensity of the 2.2 MeV capture gamma

rays. However, he has used the associated activity

3 technique to determine the relative efficiency of

this detector as a function of neutron energy.

Another group to use this technique to calibrate

their long counter was J. M. Adams, A. T. Ferguson,

4 51 and C. D. McKenzie. They have also used Cr and

57 Co for the activity measurement. Taschek and

5 Hemmendinger employed the associated activity method to measure the total cross section of ''Li(p,n)'Be

reaction.

Description of the Method

There are many (p,n) reactions in which the re- sidual nucleus decays back, at least partly, to an excited state of the target nucleus by positron emission or electron capture which results in the emission of a gamma ray when the nucleus is de-excited to the ground state. Some of these residual nuclei decay with long half lives whereas others are short

lived. As one residual nucleus is produced for every

neutron produced during the (p,n) reaction the gamma

rays associated with the decay of the residual nu- cleus have a definite relationship to the total num- ber of neutrons produced in the nuclear reaction.

Hence it is possible to determine the efficiency of

a neutron detector by counting the neutrons (with the

detector whose efficiency is to be determined) for

the entire length of time for which the target is bombarded with the proton beam and then determining the

gamma ray activity of the bombarded target. From the

ratio of the neutron yield to the gamma ray yield the

efficiency of the neutron detector can be determined.

It turns out that the (p,n) reactions which can

be conveniently used for the associated activity method are ''li (p,n) '^Be, ^^V(p,n)^^Cr and ^^Fe(p,n)^^Co

although possibilities of employing some of the other

(p,n) reactions cannot be ruled out. Available information^''' about the residual nuclei of the above

7 51 57 reactions ( Be, Cr and Co) are tabulated in Table

1. It can be observed from the table that the half

lives of these residual nuclei are very long compared

to the time required to activate the target by proton

bombardment or the time required to measure the gamma ray activity. Appropriate corrections can always

be applied for the finite time required for the proton

bombardment of the target, the counting of the gamma

rays from the residual nucleus and also for the time

elapsed between proton bombardment and gamma ray

counting.

Table 1

Half lives of residual nuclei and the associated gamma

ray energies.

Residual

Nucleus

Half Life

(days)

Acti vation

Time

(hours)

Associated

Gamma Ray

Energy (keV)

^Be

51cr

'ho

53

27.704+.002

270.9 ±.6

2

3 to 4

6

477.4

319.8

121.94

136.31

However, one should remember that the charged particle

accelerators are not a source of constant current.

Therefore, the rate of build up of the radioactive re- sidual nucleus will not be constant and it is advisable to make the bombarding time as small as possible

compared to the half life of the residual nucleus. A brief description of the techniques employed in the

three laboratories in calibrating the detectors is

given below.

^Li(p,n)^Be reaction was used to determine the

relative efficiency of the 4ir neutron detector'^ at the

University of Kentucky in the neutron energy range

30 keV to about 1600 keV. Neutron yield was measured

from LiF targets at several energies using a fresh

target for proton bombardment lasting one to two hours

at each energy. 478 keV gamma ray yield was measured

234
from each target immediately after the proton bombardment. For the absolute efficiency calibration a va- nadium metal target was bombarded with protons cor- responding to an average neutron energy of 300 keV.

57 Similarly, an isotopically enriched Fe target

57

( ^^2'^3 evaporated on to a carbon backing) was bombarded with protons to produce neutrons of average

energy 1350 keV. The gamma ray detector was cali51 57 brated using Cr and Co sources (standard sources

supplied by the National Bureau of Standards) which

enabled the determination of the absolute efficiency

of the neutron detector at 300 and 1350 keV. The

relative efficiency in the energy range 30 keV to 1600 keV was then normalized to the absolute efficiencies at 300 and 1350 keV to obtain the efficiency

for the entire energy range. The efficiency of the

detector as a function of the neutron energy is shown

in Fig. 1. The efficiency of the detector for neu- trons form a Po-Be source was found to be 0.47%.

^Li(p,n)^Be reaction cross section was measured with

this detector and compared to the cross section published by R. L. Macklin and J. H. Gibbons.^ In Fig. 2, the dots are the present measurements and the

dashed lines are cross section data published by

Macklin and Gibbons. The detector is being used ex- tensively for measurement of (p, n) and (a, n) cross sections, some of which are important for astrophysical calculations.

3 The grey neutron detector is applicable in the

neutron energy range thermal to several MeV. Fast

neutrons incident at the center of the detector, are slowed down in a moderating medium such as water or paraffin and a fraction of these neutrons are captured in the hydrogen medium emitting 2.2 MeV gamma

rays which are detected by a Nal (Tl) detector

mounted at the surface of the moderating medium. As- suming that the detector is large enough to prevent

leakage of neutrons from the moderating medium the

efficiency of the detector as a function of neutron

energy has been calculated taking into account the

gamma ray absorption in the medium. The relative

efficiency of the grey neutron detector was determined by the associated activity method and compared

to the calculated efficiency in the energy range 30

INCIDENT PROTON ENERGY (MeV)

to 1000 keV. The ^Li(p,n)^Be and the ^^V(p,n)^^Cr

Fig. 2. Li(p, n) Be reaction cross section.

Dashed lines are cross sections published by Macklin

and Gibbons.

reactions were the neutron sources for the calibra7 51 tion. The activities of Be and Cr in the

irradiated targets were counted by measuring the yield

in the photopeak of the 478 and 320 keV gamma rays

7 51 from Be and Cr respectively. The results are shown in Fig. 3. The performance of the detector was

also checked by measuring the ^Li(p,n)''Be cross seco tion. The grey neutron detector has been used for

measurement of cross sections of neutron producing

9 10 reactions important for fast reactor calculations. ' The conventionally determined efficiency of the

long counter was compared to the efficiency determined

by the associated activity method by J. M. Adams,

4

A. T. G. Ferguson and C. D. McKenzie in the neutron

energy range 50 to 1300 keV. They measured the angu51 51

lar distribution of the neutrons from V(p,n) Cr

57 57 and Fe(p,n) Co reactions and compared the neutron

yields from these reactions to the activities of the

bombarded targets. A Nal (Tl) detector was used to measure the yield in the photopeak of the 320 keV

57

gamma ray from Fe. They observed deviations in the

efficiency of the long counter from that determined

by the conventional method.

AVERAGE NEUTRON ENERGY (litVl

30 60 1U 245 320 430 €20 SiO 995 ilSO III II 1

SPHERE COUNTER

1 1 1 1

1

• Abtolula Ucotuismcnl*

• Reloiive Mwjsufement*

_ ' 1

•

} • f i 1

Efficiency lor Pu- B< Ncuttoni • -47% —

1 1 III 1

INCtOENT PROTON ENERGY (Mevt

Fig. 1. The absolute efficiency of the 4Tr neutron

detector. The error flags are an estimate of the

standard error at each point.

is

3

'

UJ2 1.06

1 1

1 \ \ ASSOCIATED ACTIVITY TECHNIQUE 0 V^'(p,n) Cr^' A Li^(p,n)Be^

FLUX INTEGRATION TECHNIQUE

H (n,y) H /

[^NORMALIZATION POINT

J \ 1 600 800 1000 1200 NEUTRON ENERGY. keV

Fig. 3. The ratio of the experimental efficiency

to the theoretically calculated efficiency of the

grey neutron detector.

235
Uncertainties in the Determination of the

Efficiency of the Detector

The important factors in the determination of

the efficiency of the detector using the associated activity method are (1) the knowledge of the

half life of the radioactive target whose activity is to be measured after bombarding the target with

charged particles and (2) the knowledge of the

strength of the sources used to calibrate the gamma

ray detector. The beauty of the associated activity

method is that the accuracy of the efficiency determination is independent of the target uniformity and errors in the target thickness determination which

are, in general, two difficult factors in efficiency

calibration. Even the absolute efficiency of the

gamma ray detector need not be known. One needs to know only the relative efficiency of the particular

geometry in which gamma ray counting from the bombarded targets and the standard sources are done.

However, one should exercise great caution to avoid

uncertainties arising from geometrical factors in the

placement of bombarded targets and calibrated gamma

ray sources.

The easiest method to determine the absolute

number of gamma rays from the irradiated targets is to compare the gamma ray yield from the target to that of the standard sources of the same radioactive

element. As several days may be elapsed between the

time of the determination of the strength of the

standard sources and the determination of the relative efficiency of the gamma ray detector used for

gamma ray activity measurement, the half lives of the

standard sources must be known as accurately as possible. The half lives of ^Be, ^^Cr and ^''co have

been reported extensively in the 1 iterature.^'^

Various methods have been used to determine the half

lives of these nuclei. Any improvement in the ac- curacy of the values of the half lives of these or similar radioactive nuclei will be useful for

improving the accuracy of the efficiency of the neu- tron detector. Hence it is necessary to obtain con- sistent values for the, half lives of those radioactive nuclei from several different methods.

Improvements in the accuracy of the knowledge of the

strength of the gamma ray sources employed for the

calibration work will also reduce the overall un- certainty in the efficiency of the neutron detector.

The effect of strongly anisotropic angular distribution of neutrons on the efficiency of the flat

1 3 response 4n neutron detector has been studied ' and

found to be less than 1% in the case of the strongly

forward peaked resonance'''^ at 2.28 MeV in the

^Li(p, n)^Be reaction. However, for the development

of a neutron detector whose efficiency is known to an accuracy of about 1%, the effect of the angular distribution on the detector should not be neglected.

Possible Improvements in the

Efficiency Determination

The efficiency of the above mentioned detectors

as a function of the neutron energy is known to about

1300 keV. It will be useful to determine the efficiency of the detectors for somewhat higher energies

though complications may arise due to flux leakage

from the moderating medium. A Monte Carlo type calculation can be used for this purpose. In fact,

W. P. Poenitz has calculated the efficiency of a 12 black neutron detector using Monte Carlo method.

A practical method for extending the energy

range over which the efficiency of the 4^ neutron

detector^ is known, will be to compare the calculated

efficiency to the experimentally determined efficiency in the energy range 30 to 1300 keV and then

extend the calculation to higher energies. Good

agreement between the calculated and experimental

efficiencies in the energy region 30 to 1300 keV will

be adequate justification for using calculated efficiencies at higher energies. The Monte Carlo method

can also be used to investigate the effect of

strongly anisotropic angular distributions of a nuclear reaction on the efficiency of the ^-n detector.

Thus better knowledge of the half lives, improvements

in the uncertainty of the source strength of standard

sources used in associated activity method and theoretical calculation of the neutron efficiency are ma- jor factors which will improve the accuracy of the

neutron detector efficiency.

References

'•K. K. Sekharan, H. Laumer and F. Gabbard, Proc. Int. Conf. Nuc. Cross Sec. Tech., NBS SP 425, 108(1975);

K. K. Sekharan, Ph.D. Dissertation, Unpublished; K. K. Sekharan, H. Laumer, B. 0. Kern and F. Gabbard, Nucl

.

Instr. and Meth., 133, 253(1976).

^W. P. Poenitz, Nucl. Instr. and Meth., 58, 39(1968).

•^W. P. Poenitz, Nucl. Instr. and Meth., 72, 120(1969).

4

J. M. Adams, A. T. G. Ferguson and C. D. Mckenzie,

Activation Technique for the Absolute Calibration of

a Long Counter, A. E. R. E.- R. 6429(1970)

.

5 E. F. Taschek and A. Hemmendinger, Phys. Rev. 74 , 373(1948).

Nuclear Decay Data for Selected Radionucl aides,

ed. M. J. Martin, ORNL - 5114(1976).

^E. De Roost and F. Lagoutine, Atomic Energy Review,

11, 642(1973).

^R. L. Macklin and J. H. Gibbons, Phys. Rev., UA, 571

(1959).

g H. 0. Menlove and W. P. Poenitz. Nucl. Sci. Eng., 33 , 24(1964).

P. Poenitz, J. Nucl. Energy, 22, 505(1968).

'^J. H. Gibbons and H. W. Newson, Fast Neutron Physics

Part I, ed. J. B. Marion and J. L. Fowler, Interscience Publishers Inc., New York, 164(1960).

^^W. P. Poenitz, Nucl. Instr. and Meth., 109, 413

(1973).

236
ACCURACIES AMD CORRECTIONS IN NEUTRON BATH TECHNIQUES

E.J. Axton

National Physical Laboratory

Teddington, Middlesex, United Kingdom

The corrections and currently attainable accuracy in the neutron source emission rate measurements with the manganese sulphate bath technique are discussed in detail, followed by a review of other bath techniques and their adaptation for neutron flux measurement.

1 . Introduction

Over the last 25 years much has been written

about the use of bath techniques for neutron measurements and it would be impossible to review all this

work in the space of 15 minutes. Probably of greatest

interest is the use of manganese sulphate baths to measure neutron source strengths, particularly with

respect to fission yield measurements. Therefore, in

spite of the title of this session, it is proposed to begin by assessing the ultimate accuracy attainable

with an ideal system and if there is time left to discuss the limitations of practical systems, and some other applications of the bath technique,

2. Design

Figure 1 is a schematic diagram of a typical

system. This shows some of the features which are essential to the attainment of high accuracy. Some of

the design features are necessarily compromises. For

example the bath has to be large to reduce neutron

escape but the specific activity goes down with the

cube of the radius. The presence of the central

cavity increases the neutron escape but reduces the

capture of neutrons returning to the source. Spherical

geometry maximises the specific activity whilst minimising the escape. It also eases the problems of

calculation and measurement of the corrections. The

bath should not be shielded imnecessarily, but if it is

shielded provision should be made for the insertion of

detectors to assess the neutron escape. Provision

should also be made to measure the thermal neutron flux

at the cavity boimdary. The solution should have a high concentration to minimize the effect of uncertainties in the cross section ratios, and should be

circulated through the activation detectors so that

they can be operated whilst the source is in the bath

in order to improve the counting statistics. Two

independent detection systems should be employed so that a change or failure in one of them is instantly

detected. A further advantage is that twice as much

data is acquired. The bath detectors are calibrated

by the insertion of calibrated samples of ^^Mn which

have been measured by the /l%^y coincidence method, A very stable high pressure ionisation chamber system is

essential to check the consistency of the coincidence

counting. Standard sources in fixed geometry should

be used regularly to check the stability of the bath

detectors. The concentration of the solution should

be measured regularly by at least two methods. Two bath sizes should be used to check the validity of the

evaluated corrections. Finally a standard neutron

source should be measured periodically to check the

whole system,

3. Source strength determination

The source strength Q is derived from the following equation;

A Q = • f E (l - 0)(1 - S)(l - L)

The fraction f of neutrons captured by manganese is

^r.i^ + G r s) NMn a (^ + G r s)

Where A is the satiiration manganese counting rate

measured by the bath detector system with efficiency

E, 0 is the fraction of neutrons which are captured

in the n,a and n,p reactions in sulphur and oxygen,

S is the fraction of neutrons recaptured by the source, and L is the fractional neutron escape from the

boundaries of the bath, CTjj, cfg '^Mn are the

thermal neutron capture cross sections of hydrogen,

sulphur, and manganese respectively, and NH/NMn is

the hydrogen to manganese atom ratio in the solution,

s is the normalised above l/v resonance activation

integral of manganese, r is the epithermal flux parameter averaged over the bath, and G is the resonance

self shielding factor. In the following sections the

components of this equation will be examined in detail,

3. 1 Measurement of bath efficiency E

The difficulties associated with this apparently

simple operation are almost invariably underestimated

as there are so many things which can go wrong.

Basically a quantity of radioactive of high purity

is divided into portions by weight, some of which are used to activate the baths and others are measured in

the 471:0 Y coincidence equipment to determine the

specific activity. Differences in the specific

activity of these solutions can be caused by sticking

of activity to glassware and by evaporation of water

during the weighing of small liquid samples either in

the bath loading or counting operations. In absolute

measurement of solutions, an international comparison revealed differences in apparent activity depending

on whether a liquid scintillator or gas flow proportional counter is used for the 3 channel. In the

proportional counter the wrong result can be obtained

if the plastic source mount becomes non-conducting. If a liquid scintillator is used it is necessary to check

experimentally that no significant after pulsing occurs. Both methods require a correction for the fact that the

p counter is not WOfc efficient (the decay scheme

correction). In the case of the liquid scintillator

this correction is larger and more difficult to derive.

Unsuspected radioactive contamination can occur in

which case the measurement will be time dependent.

Inevitably the time will come when the bath efficiency

will appear to have changed. This could be due to a fault in the bath counting or in the coincidence counting, or in the dispensation of the sample or its purity, or to a simple mistake. A third party is then required

to arbitrate. This is the role of the high pressure

ionisation chamber referred to in section 2, The calibration figure for the ion chamber based on the coincidence equipment should remain constant, a typical

standard deviation for a large number of calibrations

being about ± 0,15^. Only if this reproducibility is

237
Small spherical bath

Neutron escape rrxDnitor

Large spherical bath

Pump

Spherical cavity

Stirrer

Cavity flux monitor

Dual detector system

Fig. 1. Schematic arrangement for a typical bath system.

achieved, and the resultant calibration figure is in

agreement with that obtained using a standard from an appropriate national laboratory caji there be confidence

in the measurements, and their relationship to those of

other laboratories throughout the world. For the

actual bath efficiency .measurements, if precautions are taken to prevent loss of activity on the way e.g. along

the walls of entrance pipes, or by splashing, and no electronic drifts or faults have occurred a typical

standard deviation of ± 0.2>fo can be obtained and

sufficient repetitions of the calibration may be carried out to reduce the standard error of the mean to an acceptable level. Ultimately one is left with an additional systematic uncertainty ~ ±0.2^ for the

bath calibration. This is the current level of agreement obtained by national laboratories in international

comparisons, de Volpiv''J and is attributable in part to the difficulties of weighing small quantities of

liquids.

3.2 The H/Mn atom ratio

This may be determined by three methods.

(1) The volumetric method by titration with EDTA.

(2) The gravimetric method by drying and weighing

weighed samples of solution both as the monohydrate after drying at 100 °C and as the

dehydrate after drying at 300 °C.

(3) Determination of solution density.

The third method is not absolute but has to be

calibrated by one of the first two, both of which are

capable of reproducibility ~ 0.05% standard deviation.

In the absence of impurities the two methods should

give identical results. In the presence of impurities

the volumetric method should give the correct specific

manganese concentration provided there are no other

titratable ions present, whilst the difference between

the results of the two methods gives a guide to the

impurity level, Needless to say, impurities should be

avoided wherever possible. The H/Mn ratio can be

determined sufficiently accurately to make a negligible

contribution to the overall uncertainty.

3.3 The manganese to hydrogen thermal neutron cross section ratio

This is the dominant factor in the computation of

the fraction of neutrons captured by manganese. The

ratio of the individual cross sections taken from

Mughabghab and Garber(2) is O.O2496 ± 1.62^. However

the ratio can be determined more acurately by variation

of the concentration of the solution. Equation (1) is

rearranged as follows to form a straight line equation.

A m

= (1 + G r s) . E (1 - 0)(l - S)(l - L) Mn

1 + —

Mn

1 + Mn (1 + G r s)

Measurements are made for a variety of values of

NH/UMn from say 30 to 300 and the source strength is

derived from the intercept of the line and the cross section ratio from the ratio of the slope to the intercept. All the quantities on the left hand side are concentration dependent and therefore have to be

238
evaluated for each concentration. By suita"ble choice

of source energy and the bath size 0 can he made zero and L very small. Nevertheless L can vary hy a factor of 2 over the range of concentration used so provision must be made to measure and calculate L at each point if it is significant. A typical uncertainty in the cross section ratio which can be obtained

by this method would be ± O.jfo to 0.4^. The NPL

value is 0.02495 ± 0.34^ standard error(3). A

measurement in progress^4j indicates agreement between

this value and the BMX 325 3rd edition value. A

further value of 0.02531 ± 0.12^ has been published,

but this is qualified as an effective ratio for the

particular bath and solution used to measure itwj.

The sign of the change of efficiency with concentration

can be positive or negative depending on the type and

mode of operation of the bath counters. In principle

a different bath and coixnting system could be used for

each point provided the appropriate corrections were

made. Thus if the same source were used, all available data could be analysed in one global fit.

3,4 The effect of the impurities

The effects of the impvirities are rather complex.

Impurities which absorb neutrons and produce measurable

activity would introduce a time dependent bias to the

result which would be very difficult to interpret.

Assuming that no measurable activity is produced, the

linearity of the fit should be unaffected if the

impurity concentration is proportional either to that

of the Mn or to that of the H. Neglect of an impurity in the supply of the solid MnSO^ would produce a low value for the H/Mn cross section ratio,

but if the atom ratio was determined by a gravimetric

method it too would be too low and this would tend to cancel the error in the cross section ratio. However,

an impurity in the water would produce the correct

source strength but a high value of the cross section

ratio. The latter situation is unlikely in practice

because the initial concentrated solution is probably

made up from Mn SO^ 4H2O, in other words 2jfo of the

water is already present, and an impurity in the added

water would produce a non-linear situation,

4. Corrections

4. 1 Fast neutron capture in o^tygen and sulphur

This correction is both concentration and spectrum

dependant. Its evaluation is almost entirely dependant on calculations. The oxygen loss in water for

type of source obtained (1,69 ± 0,25)^ in water and

(3.05 ± 0.3)^ for the oxygen and sulphur loss in concentrated manganese sulphate solution. Agreement is

obtained between the NPL manganese bath and the AEKE

boron pile(Si9) if a correction of (3.O ± 0.5)^ is

applied to the bath measurements. This information

could be interpreted as another measurement of the

correction. Unfortunately the Ra Be source is not a very good source for this experiment from the point of

view of comparison with calculations as one is dependent

on assiimptions regarding the intensity of the low energy

component in the spectrum, which is fairly large and

known to vary from source to source. Finally one should mention the experimental method developed by

de Volpiw) whereby the correction is derived from the

change in slope of the line obtained in a dilution

experiment. Unfortunately the method is not very

sensitive. Nearly 80^ of the oxygen effect in a con- centrated solution is due to the oxygen in water. Thus

the change in the slope registers only the sulphur

effect plus 20^ of the oxygen effect, which varies from

about 1.4^ for AmBe source down to 0.359^ f°r ^ californium spectrum. Thus a change in the slope of

Ra Be source was measured by de Trover and Tavernier^ as (2.25 ± 0.3)^. Ryves and HardenW) using the same

only 1 to 4 times the uncertainty in the slope can be

expected. Calculations have been published for a number of source spectra by Ryves and HardenvT) and by

LouwrierV ^ based on standard slowing down theory and

by Murphey( "I ) using Monte Carlo techniques. In the case of Ryves and Harden, and Murphey the oxygen cross sections used are now thought to be too high particularly in the energy range above 7 MEIV. Likewise the

oxygen cross sections used by Louwrier are thought to be too low. Table 1 shows a comparison of calculations based on both methods for a number of source spectra using cross sections obtained from the data

centre at Saclay at the end of 1972 and are believed to be up to date. The calculations are for concentrated

manganese sulphate solution with NH/NMn = 30.

Table 1

Oxygen and Sulphur correction

Source Mean E

Monte

Carlo

Slowing down

theory

Am Be 4.46 3.09 3.17

Ra Be 3.94 2.20 2.45

Cf-252 2.09 (ET=1.39)^ 0.53 0.73

Cf-252 2.15 (EIt=i.43) 0.62 0.79

Am B 2.76 0.49 0.64

3- Maxwellian temperature in MeV,

The agreement seems to become progressively worse

as the neutron energy decreases. The only other calculations known to have been performed with up to date cross sections are those of Ullo and GoldsmithV 1 2, 1 3

;

in their re—appraisal of the t) measurements.

Table 2 compares their results with recent calculations

at MPL.

Table 2

Axton

0 S 0+S

Ullo

0 S 0+S

Macklin

0+S

Steen

0+S

Smith

et al.

NH/NMn=

77.7

0.30 0.09 0.39 0.19 0.03 0.22

Macklin

et al.

ra/NMn=

192.6

0,24 0.04 0,28 0.25

So it appears that even, calculations by the same

technique, using the same cross section data do not

agree. Whatever cross section data is used the oxygen

effect should follow the O/H atom ratio, and the

sulphur effect should follow the S/H atom ratio. The

figures on the right of the table do not appear to do

this. There is no way of assessing the absolute validity of these results other than by agreement between

different calculations. These programs are necess- arily complex and it seems almost impossible to be

certain that no hidden faults exist in them. It would

be a good idea to hold an international comparison of

Monte Carlo calculations to resolve the point.

4.2 Neutron Escape

As with the oxygen and sulphur loss correction the

escape correction is dependent on the concentration of

the solution and on the source spectrum. In addition

it depends on the size and geometry of both the bath

and the central cavity. It is also varied by the

239
presence of shielding materials. There are so many

variables that it is difficult to produce a comprehensive correction set which would suit all occasions, and

it is equally difficult to compare corrections which

have been evaluated for different systems. Many, > authors have used expressions of the type A e"-^-"^^ ' to evaluate their escape corrections using constants

derived for other systems by other authors. Whilst

such expressions give a good estimate of the order of

magnitude of the correction, they are not transferable

when the geometry of the bath or the central cavity are

changed, or if the concentration of the solution or the

shielding is changed. On the other hand the correction may be checked by measurement. In principle the

escape correction can be derived from measurements in

different bath sizes and extrapolated to infinite

radiusOU This is not very practicable, but the use of

two bath sizes provides a valuable bench mark for

testing measurement and calculation procedures. For

day to day monitoring at NPL a flat response detector

(long counter) is used to monitor the neutron flux at the surface of the bath, which is then integrated over

the bath surface. The counter efficiency is chosen

empirically for each source spectrum to give agreement

between the two bath sizes. Although there are many reasons why the long counter would appear to be unsuitable for this purpose it seems to give satisfactory

results for the two baths. Table 3 compares these

measurements with Monte Carlo calculations at NFL which

are only in a preliminary state. The fast neutron

program records fast neutron escape and records the

position of neutrons which become thermalised. Shell sources of thermal neutrons so produced can be used to determine thermal escape either by diffusion calculations (M/C + Diff) or by a Monte Carlo program

(M/C). The programs are still being developed so that

the results must be regarded as provisional. Agreement becomes progressively worse as the source energy

decreases, and the thermal fraction increases. Part

of the problem is due to the uncertainty in the source

spectrum. Some Am B and Am Li sources are believed to be contaminated with Be.

At the present state of the art, a systematic

uncertainty of ± ^0'fc of the correction for an Am Be source and 2Qffo for a Californium source seems reasonable.

Table 3

Ueutron escape from an unshielded sphere with a

8.9 cm. dia. cavity and MH/Mn = 30

Total leakage (^) Thermal fraction

Meas. m/c M/C +

Diff.

Meas. m/c M/C +

Diff.

98 cm. tank

Am Be

Ra Be

Am B

Cf (ET=1.39)

Cf (ET=1.43)

Am P

1.46

1.15

0.33

0.30

0.05

1.49

1.14

0.33

0.32

0.42

0.05

1.41

1.07

0.34

0.30

0.38

0.05

17

93523

18

1528

22

20

19

13

10

19

1712

25 cm. tank

Am Be

Ra Be

Am i

Cf (ETt=i.39)

23.2

19.7

14.6

10.0

23.4

18.8

12.

1

21.2

15.8

10.0

2122

29

32

1222

34

20

13

20

Table 4

Comparison of neutron escape calculations with

Ullo and Goldsmith evaluations

Total escape

m/c M/C + Diff. Ullo

Smith

et al bath

0. 16 0.13 0.23 ± 0.11

Macklin 0.28 0.21 0.26 ± 0.03

4.3 Source capture of moderated neutrons

This correction is both concentration and source spectrum dependent. In all cases it is necessary to know accurately the amount and geometry of each

material in the source assembly and the appropriate

reaction cross sections. There are two approaches to this problem. One can derive the cavity boundary

thermal and epithermal neutron flux in terms of the

Wescott parameters either by measurement with foils or by diffusion calculations^ ''4) or by Monte Carlo calculations. It is then necessary to evaluate an effective cross section for the source assembly with

appropriate flux depression and self shielding

corrections for each region of the source assembly.

Spiegel and Murphey^''^) describe a method of treating

the problem. Alternately sufficient detail can be

programmed into the Monte Carlo calculation to reproduce the source interactions directly. In either case

a convenient test of the calculation .is the ability to determine the cavity boumdary flux, which can be

measured to about ± 2^ with foils. Table 5 compares

the result of Monte Carlo and diffusion calculations

with sub-cadmium thermal neutron flux measurements by

Bardell^''^) using gold foils. Again the agreement

gradually worsens as the energy decreases. The low

value for Am B is due to flux depression in the

absence of which these results would lie on a smooth curve as a function of neutron energy.

Table 5

Thermal neutron flux at cavity boundary

for unit source (/J)

Measured

with gold

foils

Monte Carlo/

Diffusion

calculations

Am Be 0.139 0.134

Ra Be 0. 180 0.178

Am B 0.140 0.128

Cf (ET=1.39) 0.223 0.218

Cf (ET=1.43) 0.223 0.214

Am P 0.249 0.208

Smith et al

HH/mn=77 ET= 1.323

0.175

Macklin et al

1IH/Mn=192 ET= 1.323

0.204
4.4 The manganese resonance correction Table 7

This correction can be obtained by deriving an effective cross section for manganese in terms of the

Wescott'^"^^ parameters G, r, ajid s. Calculations

by this method have been published by Axton and

Ryves^ Alternately one could enter the detailed

manganese resonance data directly into a Monte Carlo

calculation

.

Both methods depend on knowledge of the manganese activation integral. If the resonance parameters

are put into a Monte Carlo calculation it is necessary

to choose the y width so as to give the preferred

resonance integral. The value given in BNL 325

edition 3 is 8 barns. Recent measurements at NPL in

terms of the gold resonance int egral ' 9 ) indicate a value of about 7.6 barns. There are some much higher

values elsewhere in the literature. The derived

source strength is not very sensitive to the precise

value of the integral. With concentrated solutions

(H/Mn =30) a change of 10^^ in the integral would

alter the source strength by about 0.0'jfo, Table 6 gives examples of the magnitude of the correction.

The difference between the Axton, and Ullo and

Goldsmith values is almost entirely accounted for by

the difference in the value of the resonance integral

used. Ullo and Goldsmith used 9*15 barn. The

details of their calculation are not given in their

paper.

Table 6

Manganese resonance corrections

NH/HMn NPL Ullo and Goldsmith

TTPL 30 1.0071

Smith et al 77.7 1.0083 1.0100

Macklin et al 192.6 1.0093 1.0111

4.5 Photoneutron production

In some cases a small correction is necessary to allow for photoneutron production from deuterium in the

solution. This correction is usually negligible with

light water baths except in the case of high energy

high intensity y emitting sources such as Na Be.

5. Uncertainties

Provided all the necessary precautions are taken

and all the checks* are successful, typical uncertainties would be estimated as shown in Table 7. Uncertainties are evaluated for two concentrations and

two source spectra. The figures shown are for a 98 cm. diameter bath with an 8.9 cm, diameter cavity. The

random uncertainties are quoted at the 68^ confidence

level. Five measurements of the bath efficiency, and

five irradiations of the bath by the source are assumed.

The existence of the second counting channel, in

addition to providing internal consistency checks, also

doubles the amount of data for analysis. It is

assumed in all cases that the source is sufficiently

strong to provide adequate counting statistical

accuracy and that the background uncertainty is negligible even for the dilute solution.

*see Appendix

Typical estimates of uncertainties

Systematic

Uncertainties {%)

llH/mn=30

Am Be Of

m/mn=i92

Am Be Cf

Bath efficiency 0.2 0.2 0.2 0.2

Cross section ratios

cr /cr F Mn 0.145 0.145 0.287 0.287

/ V^'Mn 0.122 0.122 0,039 0.039

Manganese resonance U»U ( n n7( Oxygen and sulphur 0.3 0.1 0,18 0.060

T ,fia Vfl £rp 0.15 0.03 0.20 0.04

Source capture 0.015 0.020 0,02 0.03

Random uncertainties

Bath efficiency 0. 13 0.13

Source activation of 0. 10 0.10

bath

6. Other baths used for source strength measurements

6. 1 Noyce et al (20) introduced heavy water into their

manganese bath in order to reduce the cross section

dependence. It was necessary to introduce a proportion of light water into the mixture to prevent

excessive neutron escape. In addition to the usual

corrections, a correction for the photoneutron production in the bath was necessary. To minimise this

correction an intermediary Sb-Be source was calibrated instead of the NBS standard Ra Be photoneutron

source. In a subsequent comparison measurement ± ^'fo accuracy was achieved for the standard source.

6.2 de Troyer and Tavernier(6) Jaritsina et al(2l).

Van der Eijk^^^' and Michikawa(23) used a water bath in

which the neutron density as a function of distance

from the source was integrated numerically. The

neutron density at each point was measured by gold foil

activation. The necessary corrections are similar to those described for the manganese bath. The method is

time consiMiing and less amenable to automation but provides a valuable check on MnSO^ bath calibrations.

In principle it might be possible to develop the method

to the same accuracy. Present uncertainty in the

gold/hydrogen cross section ratio is ± 0.7^.

6.3 Pieldhouse et al (24) used a cylindrical bath in

which the neutron density as a function of distance was determined from BP^ counter measurements. The system

was calibrated by associated particle counting of a particles from the T(dn) reaction. The experiment

did not achieve its aimed accuracy of ± ^fo,

6.4 Smith et al(25) and Macklin et al(26) used

cylindrical manganese sulphate baths for their r] measurements. In these experiments only ratio's of source strength are involved so none of the uncertainties associated with the absolute calibration of

the bath detectors are relevant. On the other hand

the problem of calculating the correction for neutron

241
absorption and multiplication in the complex source assembly are correspondingly greater and it seems difficult to believe that these effects can be calculated to the accuracy quoted. For example in the Goldsmith and

Ullov''3j re-evaluation of the Oak Ridge measurement,

the correction quoted for cadmium absorption is

(1.85 ± 0.04)/o.

7. The use of baths with accelerators

7.1 Scott(27) used a spherical manganese sulphate bath

to determine the thick target yield from the '^Li(p,n)

reaction, the source in this case being the accelerator

target situated at the centre of the bath. In addition to the usual corrections it was necessary to allow, in the continuous flow detector system, for time

dependent fluctuations in the neutron emission rate.

Overall accuracy of 2—3^ was achieved,

7.2 Poenitz(2^) used a spherical manganese sulphate

bath to measure the neutron flujc at zero degrees at energies close to the threshold of the endothermic

reactions '^Li(p,n) and T(p,n). By careful control

of the accelerator energy the neutron production was

limited to a forward cone of half—angle 9 degrees,

thereby removing the need to collimate the bath with

bulk shielding. The largest contribution to the

uncertainty in the neutron flux was ± 1.39° due to neutron escape.

7.3 The use of 3»8 minute. 52^ speeds up the measurement cycle considerably compared with 155 minute Mn. However, the short half—life introduces problems in the

absolute measurement of the activity and its comparison

with other laboratories. Furthermore, the uncertainty

in the V/H cross section ratio of ± 1^ enters into

the source evaluation with greater weight as the V

cross section is only 5 barns. For these reasons, if an absolute system is required it is preferable to calibrate the system with standard neutron sources.

7.4 Robertson et al(29) used a collimated vanadyl

sulphate bath to measure the relative zero degree

neutron yield from the '^Li(p,n) reaction. It was necessary to shield the bath from room scattered

neutrons and to admit the direct neutrons into the bath

through a collimator. An important feature of the

experiment was the demonstration by Monte Carlo calculation supported by inverse square measurements that

the collimator could be taken as having, for each

energy, a fixed entrance aperture at a fixed position

over the energy range considered. The system was used

to measure relative neutron flux with a standard

deviation of ± 2^,

7.5 Leroy et al^-^O) yged a small shielded and collimated manganese sulphate bath to measure the neutron

flux from an accelerator in the energy range from 10 keV

to 1 MeV,

The bath detector system was calibrated by means of a Ra Be neutron source which had been calibrated by

BIFM. Neutron capture in the neutron source assembly

and neutron escape into the reflector were obtained by

Monte Carlo calculations as was the effective solid

angle subtended by the bath aperture. An additional

correction was necessary for air attenuation of the

neutrons and the overall precision was reported to be + 1.8^. Above 1 MeV the neutron escape fraction became

unacceptably large.

8. Conclusion

With sufficient care, the manganese sulphate bath

can be used to achieve an uncertainty estimated at + 0,35?^ for a Californium neutron source at the 68?^ confidence level or + O.^fo at the 99^ confidence level,

242

Adapted for the measurement of neutron flux density,

typical uncertainties of + 1,8-2,0^ at the 6&fo confidence level have been achieved with bath

detectors,

9. References

1. de Volpi, A., 1973, Metrologia, 6, 65.

2. Mubhabghab, S.L. and Garber, D.I., 1973, BNL 325,

Edition 3.

3. Axton, E.J., Cross, P., and Robertson, J.C., 1965,

J. Nucl. Energy, 22, 409.

4. Smith, J.R. , 1977. Private communication.

5. de Volpi, A. and Porges, K.G. , 1969, Metrologia,

1, 128.

6. de Troyer, A. and Tavernier, G.C., 1954, Acad.

Roy. Belg. Bull. Class. Sci., 40, 15O.

7. Ryves, T.B. and Harden, D., 1965i J. Nucl. Energy

a/b, 22, 607.

8. Colvin, D.W. and Sowerby, M.G. , 1965, Physics and

Chemistry of Fission, (Proc. Symp, Salzburg, I965),

2, 25. IAEA Vienna,

9. Colvin, D.W. , Sowerby, M.G. and Macdonald, R.I.,

1967, Nuclear Data for Reactors, (Proc. Symp.

Paris, 1966), 2» 307. IAEA Vienna,

10, Louwrier, P,F, , I964, PhD Thesis, Institute for

Nuclear Physics Research (IKO) Rototype,

Amsterdam,

11. Murphey, W.M. , I965, Nucl, Inst, Meth, , 37, 13.

12, Ullo, J.J. and Goldsmith, M. , 1976, Nucl. Sci.

Eng., 60, 239.

13. Goldsmith, M. and Ullo, J.J., 1976, Nucl. Sci.

Eng., 60, 251,

14. Ryves, T,B, , I964, J, Nucl, Energy A/B, _i8, 49.

15, Spiegel, V, Jr. and Murphey, W.M. , 1971,

Metrologia, 34.

16. Bardell, A.G. , 1977, to be published.

17. Wescott, C.H, , Walker, W.H. and Alexander, T,K,

,

1958, Proc, 2nd. Conf, The Peaceful Uses of Atomic

Energy, Geneva, 15/p/202 United Nations N,Y,

18, Axton, E.J, and Ryves, T,B, , 1967, J, Nucl,

Energy, 21_, 543.

19, Ryves, T,B, and Axton, E,J,, to be published,

20, Noyce, R,H, , Mosburg, E,R, Jr, , Garfinkel, S.B.

and Caswell, R.S. , 1963, J. Nucl. Energy A/B,

11. 313.

21, Jaratsina, I, A,, Constantinos, A, A, and

Porminikh, V,I,, I964, Energie Atomique, 26, 253.

Jaratsina, I. A., Andreev, O.L, , Katchine, A.E. and Stukov, S.M. , 1964, Energie Atomique, ^6, 255.

22. Van der Eijk, see Naggiar, V., I967, Metrologia,

3, 51.

23. Michikawa, T. , Teranishi, E., Tomimasu, T. and

Inoue, Y, 1959» Bull. Electrotechnical Laboratory,

Japan; 23, 223.
24. Pieldhouse, P., Culliford, E.R. and Mather, D.S.,

1967, J. Nucl. Energy, 21_, 131.

25. Smith, J.R. , Reeder, S.D. and Pluharty, R.G.,

1966, IDO 17083, Phillips Petroleum.

26. Maoklin, R.L. , de Saussure, G. , Kington, J.D. and

Lyon, W.S., 196O, Nucl. Sci. Eng., 8, 210.

27. Scott, M.C. and Ashraf Atta, M. , 1973, Nucl. Inst.

Meth. , 106 . 465.

28. Poenitz, W. , 1966, J. Nucl. Energy, 20, 825.

29. Robertson, J.C., Ryves, T.B. and Hunt, J.B.

,

1972, J. Nucl. Energy, 26, 507.

30. Leroy, J.L. , Huet, J.L. and Gentil, J., Nucl.

Inst. Meth., 88, 1.

31. Davy, D.R., 1966, J, Nucl. Energy, 20, 277.

Appendix

The following internal checks give valuable

insight into the hehaviour of the system and can save time being wasted on abortive irradiations. If any

check fails the cause should be ascertained and

rectified.

1. The product of bath efficiency (E) and total

volume of bath and detector system should be

constant for all bath sizes.

2. Fixed geometry y source checks on the bath

detectors should yield a constant count rate for

each channel.

3. Results should be processed for both growth

(source in bath) measurements and decay (source

removed) measurements, growth/decay ratios should

be constant and equal to unity for both channels.

Channel l/channel 2 ratios should be the same for

growth, decay, and efficiency measurements.

Deviations in the early stages of growth or near

the changeover point immediately reveal faults

due to change in pumping speed, timing errors or inadequate mixing.

243
INTERNATIONAL COMPARISON OF FLUX DENSITY MEASUREMENTS FOR MONOENERGETIC FAST NEUTRONS

V.D. Huynh

Bureau International des Poids et Mesures

Pavilion de Breteuil, F-92310 SEVRES, France

An international flux density intercomparison of fast neutrons has been organized by the Bureau

International des Poids et Mesures during the past three years. Nine laboratories have carried

out measurements. Three neutron energies were selected (250 keV, 2.5 MeV and 14.8 MeV),

to v^hich two optional energies were added (565 keV and 2.2 MeV). A polyethylene sphere

with a small BFg counter at the center was used as a transfer instrument at all energies except

for 14,8 MeV. A "^He proportional counter was used at the two lower energies as the second

transfer instrument. At 14.8 MeV a fission chamber (^"^^U) and the iron foil activation method

using 56Fe(n,p)56 Mn reaction were used. The results concerning the sensitivity of each transfer

instrument measured by all the participating laboratories are summarized.

(International comparison; fast neutron flux density)

Introduction

At the request of Section III (Mesures neutroniques)

of the Comite Consultatif pour les Etalons de Mesure

des Royonnements lonisants, the Bureau International des

Poids et Mesures decided in 1972 to organize an international comparison of flux density measurements for monoe ne rgetic fast neutrons. Nine laboratories have

carried out measurements. Table 1 lists the neutron

energies chosen, the transfer instruments used and the

participating laboratories with their responsible physicists. V.D. Huynh of the BIPM took part in the measurements performed in all the participating laboratories

in order to control the good functioning of the transfer

instruments and to ensure the homogeneity of the compa rison

.

Three neutron energies were selected, namely

250 keV, 2.50 MeV and 14.8 MeV, to which two optional energies (565 keV and 2.20 MeV) were added.

Three transfer instruments were used: a Bonner

sphere (polyethylene sphere with a small BF2 counter

at the centre), a "^He counter and a fission chamber 000

(^"^"U), The polyethylene sphere was used at all energies (except for 14.8 MeV), the "^He counter for the

two lower "en« cgies and the f,iss1on c ham be r at 1 4 . 8 Me V

.

Another method was used at 14.8 MeV also, namely the

activation of iron foils by the 56Fe(n,p)56M n reaction .

Transfer instruments

1 . Bonner sphere

The BIPM studied and supplied the polyethylene

sphere + BF3 counter (Fig. 1). This counter is placed at

the center of the sphere, the diameter of which is 20.3 cm. The active part of the counter has a diameter

of 1 cm and is 3 cm long. The threshold and the stability of the counter are checked by means of a 100 mC I Am-Be source which can be introduced into the sphere

and placed near the counter.

Toble 1

Conditions of t"he international comparison of flux density measurements

for monoenergetic fast neutrons (neutron energies, transfer instruments and participating laboratories)

N e u t ro n energies

250 keV

565 keV

(optiona I)

2.20 MeV

(optiona 1)

2.50 MeV

14.8 MeV

Transfer instruments

polyethylene sphere + B

3 He counter

polyethylene sphere + BF_

3^He counter

polyethylene sphere + BF^

polyethylene sphere + BF^

fission chamber ( ^ U)

re(n,p) Mn

Participating laborotories*

NRC, CEN, BCMN, NPL,

ETL, PTB, NBS

NRC, CEN, NPL, ETL, PTB, NBS

NRC, CEN, NPL, ETL,

PTB, NBS

CEN, BCMN, NPL, PTB

BIPM, NRC, CEN, BCMN,

NPL, ETL, PTB

CEN, BCMN, NPL, ETL, BIPM

NPL, BCMN, IMM, ETL, BIPM

given in chronological order of participation; the acronyms stand for:

Bureau Central de Mesures Nucleoires, Euratom, Geel,

Belgium (H . Liskien and A. Paulsen) Bureau International des Poids et Mesures, Sevres, France (V.D. Huynh)

Centre d'Etudes Nucleoires, Codoroche, France (I. Szabo)

E lec trotec hni CO I Laboratory, Tokyo, Japan

(E . Teronishi and T. Michikawa)

; Institut de Metrologie D.I. Mendeleev, Leningrad, USSR

(I. A. Jaritzina and V.T. Stchebolev)

: Notional Bureau of Standards, Washington, D.C., USA

(M . M . Me ie r and O.A. Wasson)

: National Physical Laboratory, Teddington, Great Britain (E.J. Axton, J.B. Hunt and A.G. Bardell)

; National Research Council of Canada, Ottawa, Canada

(K.W. GeigerondL. Van der Zwan)

: Physika I isch-Te c hnisc he Bu nde so nsta 1 1 , Braunschweig,

Federal Republic of Germany (R . Jahr and M. Cosack).

BCMN

BIPM

CEN

ETL

IMM

NBS

NPL

NRC

PTB

244
Fission chamber (^"^^U)

The fission chamber was constructed and supplied

by NBS. it consists in fact of two chambers called

chamber T ("top") and chamber B ("bottom") (Fig. 3),

The supports are two platinum foils set back to back

and having each a ^SSjj deposit of 1 mg/cm^ over an

area of 1 .3 cm^. The gas circulating through the

chamber is pure methane.

EC

1 y ^'^m

^txx^-^

Figure 1 - BIPM polyethylene sphere with BF^ counter

(all dimensions in mm)

P - polyethylene sphere, 0 203

F - polyethylene sheath containing the BF„ counter,

type 0,2NE3/1 F (LMT)

B - polyethylene plug which can be removed to place

the source C - active part of the counter, 0 10, I 30

T - polyethylene plug containing the source

S - 100 mCi Am-Be(a,n) source, 2.2 x 105 n/s

2 o H e proportional counter

3

The He counter was supplied by NRC . It has an

active length of 10.2 cm and a diameter of 2.44 cm.

It is filled with helium-3 at a pressure of 10^ Pa.

Figure 2 gives an example of the response of the counter

to 565 keV neutrons. In the comparison the bias was set at a pulse height which corresponds to 764 keV + (0.6 x E ) , n where E is the neutron energy considered (in keV).

n

thermal 764 keV

peak .

(764+En)keV

bias , '. (764 + 0.6 En) keV

... \ . V

A.

Figure

channel number

2 - He counter spectrum for - 565 keV

0 25

-E

aluminium teflon

Figure 3 - Fission chamber constructed by NBS

(all dimensions in mm)

a - fission chamber:

T - top chamber

B - bottom chamber

EC - collecting electrodes 2

2 D _ 238(j deposits, 1 mg/cm , surface 1,3 cm , on platinum foils (ground potential)

b - envelope of the fission chamber:

P - position of the deposits

E - entrance of methane

S - exit

Figure 4 - Response of the fission chamber

245
Figure 4 shows the extrapolation method used for the

low energy counts of the spectrum: a straight line is drawn between the point corresponding to the minimum

of the "valley" and the one corresponding to zero amplitude ("true zero") of the multichannel analyzer.

4. Iron foil activation method

Iron foils are irradiated in the 14,8 MeV neutron

beam and the induced "saturated" P activity according

to the 56Fe(n,p)56M n reaction is measured. This comparison was organised by E.J. Axton (NPL). Before the

real comparison started, foils were circulated

among the participating laboratories to enable them

to compare the results of P counting measurements. In the light of the agreement obtained the method may be

conside red va I id .

Determination of the sensitivity

of the transfer instruments

Let us recall that the comparison consisted in measuring the sensitivity S. of each instrument defined

as the quotient of the count rate of the detector to the

neutron flux density. In practice, all laboratories

measured the neutron flux density by means of a quantity SPn which is the number of neutrons emitted by

unit of solid angle in the direction of the detector;

therefore, it is easier to compare directly the ratio

^'S~i ~^c^^T\ ° chosen source-detector distance d,

where is the number of counts of the transfer instrument. This gives the relation

a difference turned up in all laboratories (Table 3).

Moreover, this difference is more important in the cases where the contribution of scattered neutrons is larger.

One can see in Fig. 5 that for the 250 keV energy

there are two groups of results, one for the laboratories

having a higher correction for scattered neutrons and

the other for those having a smaller correction

(Table 2a). In addition, it happens that the group with

the larger correction obtains also a higher value for the sensitivity of the Bonner sphere. Consequently,

if for each laboratory we correct the contribution of

scattered neutrons by means of the value determined

by the shadow bar, the discrepancy between the

two groups of laboratories should be reduced.

2 As a matter of fact, the 1/d law for the determination of the contribution of scattered neutrons is valid only if the variation of the scatter with distance

is known or constant. Since this is not true, we should

use the shadow bar. Table 4 summarizes the results

obtained in the laboratories which used a 50 cm long

shadow bar to estimate the contribution of scattered

neutrons for the Bonner sphere.

In conclusion, we can say that, on the one hand,

a great effort is still necessary to improve the absolute

measurement of flux density for reaching an accuracy

of 1% and, on the other hand, a new suitable transfer

instrument has to be found which has a reasonable

efficiency and a low sensitivity to scattered neutrons

and y rays

.

For the sake of convenience, £^ will be called the

sensitivity of the transfer instrument in what follows.

Results

The results obtained in the various participating

laboratories are summarized in Tables 2a, b, c, d, e and f. The Bonner sphere, the "^He counter and the

fission chamber were placed at 1.50 m, 50 cm and

10 cm from the target, respectively. The l/d^ law was

used to estimate the contribution of scattered neutrons

for the Bonner sphere. The statistical and systematic

uncertainties given in the tables were obtained by a quadratic sum of the various contributions as given by

each laboratory. These results are also indicated in Figures 5 to 12.

It appears that there exists in general a reasonable

agreement among the results (within 5% or better),

except for 250 keV (Bonner sphere) where the discrepancy ranges from 10 to 20%. At 250 keV the corrections for the contributions of scattered neutrons are

inconsistent: the l/d^ law was used by all laboratories

to estimate the contribution of scattered neutrons, but some laboratories (NPL, ETL and PTB) used also a shadow bar which gave practically the same correction,

except for the cases where the contribution of scattered

neutrons was high, in particular for 250 keV where

Table 2a

Results of the comparison for the 250 keV neutron energy

Labora tory and

tra nsfe r instrument

D U

9. o

"Absolute"

detector*

NRC PC CEN DC BCMN PC NPL LC

ETL PC

PTB PC NBS BD

NRC PC CEN DC

NPL LC

ETL PC

PTB PC NBS BD

Correction

for scattered neutrons (%)

11.3

6.9 20.7

9.1

10.2

6.4

1.0

Sensitivity [|counts/(n/sr)^

5.71 X 10

4.81 X 10"

5.31 X 10'

4.71 X 10'

5.24 X 10"

4.71 X 10"

4.76 X 10"

3.35 X 10

3.05 X 10"

2.90 : 3.05 ;

1010"

3.09 X 10"

3.20 X 10 -6

PC - recoil proton proportional counter DC - directional counter LC - long counter BD - black detector

Uncertointy

syst. Stat. (\<r) (%) (%)

3.5

2.3

2.8

3.2

2.5

3.1

3.7

3.2

2.2

3.1

2.5

3.1

3.6

246
Table 2b

Results of the comparison for the 565 keV neutron energy

Table 2e

Results of the comparison for the 14,8 MeV neutron energy

La bora tory and

transfer instrument

"Absolute "

detector*

Correction

for scattered neutrons (%)

Sensitivity [^counts/(n/sr)]

Uncertainty

syst. stat.(10") (%) (%) Labo rotory N e u tro n

e ne rgy (MeV)

(Fission cham

" A bso 1 u te

"

detector*

ber at 10 cm)

Sensitivity [[counts/ (n/sr)^

Uncertainty

syst. stat.(l<r) PC 8.9 5.96 (%) {%) X 10-' 1.9 0.4 DC 3.5 6.21 X 10-' 2.2 0.1 o

LC GEN 14.8 AP AO . O O XV IfJ- 1 w 3.3 0.5

6.0 5.88 X 10-' 3.1 0.4 -8

PC 5.84 X BCMN 14.8 T 6.50 X 10

° 3.1 1.0 9.0 10-' 2.3 0.6 -R

PC 4.6 5.84 X 10-' 3.1 0.2 NPL 14.67 PP 6.68 X 10

Q 2.9 0.6

BD 1.0 5.97 X 10-' 3.6 0.7 ETL 14.75 AP 6.55 X 10-° 2.8 0.3 BIPM 14.68 AP 6.36 X IQ-^ 2.0 0.3

PC 2.19 X 10-' 2.6 1 .4 * AP - associoted particle DC 2.20 X 10-' 2.2 0.1 T - telescope LC 2.17 X 10-' 3.1 0.5 PR - single stage proton recoil device

PC 2.08 X 10-' 2.2 0.6 PC 2.08 X 10-' 3.1 0.3 BD 2.21 X 10-' 3.5 1 .0

NRC

GEN

NPL

ETL

PTB NB5

NRC GEN

NPL

ETL

PTB NBS

* PC - recoil proton proportional counter DC - directional counter LC - long counte r BD - black detector

Table 2f

Results of the comparison for the 14.8 MeV neutron energy

('°Fe(n,p)-'°M n reaction)

Table 2c

Results of the comparison of the 2.20 MeV neutron energy

(polyethylene sphere + BF counter at 1 . 50 m)

Labo ra tory

Laboratory

CEN

BCMN

NPL

PTB

"Absolute

detecto r*

DC

T

LC

T

Correction

for scattered neutrons (%)

3.9

6.1 6.0

4.0

Sensitivity

Qcounts/(n/sr)

6.88

7.61

10

10-

7.24 X IQ7.24 X 10 -6

IJncertainty

syst. Stat. (la-)

(%) (%)

3.1 2.7

3.2

2.2

0.1 0.7

0.4

0.8

DC - directionol counter

T - telescope LC - long counter

Table 2d

Results of the comparison for the 2.50 MeV neutron energy

(polyethylene sphere + BF^ counter at 1 .50 m)

Labo ra tory "Absolute "

detector*

Correction

for scattered neutrons (%)Sensitivity

[^counts/(n/ sr)]

Uncertainty

syst. stat.(lcr)

(%) (%)

BIPM AP 15.8 7. 16 X 10''^ 1.8 0.1 NRC SS 8.7 7.23 X 10''^ 4.2 0.8 CEN DC 4.0 6.58 X 10"^ 3.2 0.1

PC 8.5 6.90 X

10''^ 2.6 0.4

BCMN

1

T 8.5 7.15 X 10'' 2.7 1.0 NPL LC 6.5 6.97 X lO"*^ 3.2 0.5

ETL SD 48.5 8.07 X 10"'^ 8.6 1 .0

PTB T 3.6 7.13 X 10''^ 2.2 0.7

NPL

I BCMN

H i ETL BIPM

NPL IBGMN

ETL

\ BIPM

NPL - ^BCMN

ETL

, BIPM

5

I NPL

9 °c / I M M

5 5 l| NPL

CO J o c / IMM

5 \ NPL

0 °c / IMM

* AP

T CM

NP

PR

Neutron

energy (MeV)

14.67

14.8

14.75

14.68

14.67

14.8

14.75

14.68

14.67

14.8

14.75

14. o8

14.67

14.8

14.67

14.8

14.67

14.8

" Abso lute " detector*

PR

TAPAP

PR

TAPAP

PR

TAPAP

PR

AP, CM, NP

PR AP, CM, NP

PR AP, CM, NP

Sensitivity

I coun ts/(n/cm2)]

23 X 10-^ 2 6 0.4

22 X 10-^ 2 5 0.8

21 X 10-^ 2 8 0.3

17 X 10-^ 2 5 0.1

3.21 X IQ3.22 X IQ3.14 X IQ3.11 X 10-

3 22 X 10-^ 2 6 0.3

3 18 10-^ 2 5 0.9

3 17 X 10-^ 2 8 0.3

3 13 X 10-^ 2 5 0.2

3.25

3 .24

3.22 3.17

3.19

3.19

1010"

IQIQ1010-

Uncertainty

syst. stat.(la~) (%) (%)

2.6

2.5

2.8

2.5

0.3

0.9

0.3

0.2

2.6

0.8

2.6

3.1

2.6

1 .0

0.5

0.3

0.3

0.4

0.4

0.4

associated particle

telescope n-a coincidence method

method based on n-p scattering cross section

single stage proton recoil device

AP - associated particle SS - stilbene spectrometer

DC - directional counter PC - recoil proton proportional counter

T - telescope LC - long counter SD - semiconductor detector + hydrogen radiator

247
The resuUs of the comparison are given in graphical form in Figures 5 to 12

I 1 systematic uncertainty statistical uncertainty

NRC

CEN

NPL

ETL

PTB C

NB5 C

NBS

4.5 5.0 5.5 6.0 . 10"°

counts/(n/sr)

Figure 5 - = 250 keV, sphere + BF^ counter at 1 .50

2.0 2.1 2.2 2.3

ounts/(n/sr)

Figure 8 - E - 565 keV, He counter at 50 cm

n

NRC

CEN

CEN

NPL

ETL

PTB

I I H 1%

BCMN

NPL

PTB

2.5 3.0 3.5 X 10"°

counts/(n/sr)

Figure 6 - E = 250 keV, He counter at 50 cm

n

7.0 —I—

8.0 X 10"*

counts/(n/sr)

Figure 9 - E = 2 .20 Me V, sphere + BF- counter at 1 50

NRC

CEN

NPL

ETL BCMN (t«latcop«]

PTB I

1

NBS C

5.5 6.0 6.5 X 10

"

countj/(n/lr)

a. 5. 10'°

count»/(n/»f)

Figure 7 - E no= 565 keV, sphere + BF counter at 1 .50 m Figure 10- E no= 2 .50 Me V, sphere + BF„ counter at 1 50 m

248
ETL

6.0 H

7.0x10

counts/(n/sr)

Figure 11 - E = 14.8 MeV, fission chamber at 10 cm

n

Table 3

Contribution of scattered neutrons (%)

(polyethylene sphere + BF^ counter at I . 50 m)

250 keV 565 ke V 2 20 MeV 2 50 MeV

Laboratory

(a) (b) (a) (b) (a (b) (a (b)

BIPM 15 8 17.8 NPL 9.1 11.2 6.0 6.3 6 0 6.2 6 5 5.8

ETL 10.2 17.1 9.0 11.2 48 5 58.

1

PTE 6.4 9.3 4.6 4.8 4 0 5.3 3 6 5.2 NBS 1 .0 1 .0 1 .0 1 .0

a) according to the 1/d law

b) measured with a shadow bar

Table 4

Results of the comparison. Polyethylene sphere + BF^ counter at 1 .50

(Correction for scattered neutrons measured with a shadow bar)

3.2 3.

H 1%

Fe foil 47

3.4x10 3.0 3.2 3.4x10

J coufits/(n/cm ) H 1%

Fe foil 49

I I

H 1%

Fe foil 57

3.4x10" 3.0 3.4x10

2 ounls/(n/'cm )

Figure 12 - = 14.8 MeV, ^*^Fe(n, p)^^Mn reaction

Correction U nee rta i nty Neutron Laboratory for scattered Sensitivity

e ne rgy neutrons Fcounts/Cn/sr)j syst

.

stat.dO")

(MeV) (%) (%) (%)

NPL 1 1 .2 4 60 X 10-^ 3.2 0.6

ETL 17.1 4.84 X 10-^ 2.5 0.6 0.250 . PTB 9.3 4 56 X 10-^ 3.1 0.4 NBS I .0 4 76 X 10-6 2.8 0.8

' NPL 6.3 5 86 X 10-6 3.1 0.4

ETL 1 1 .2 5 70 X 10-6 2.3 0.6 0.565

PTB 4.8 5 83 X 10-6 3.1 0.2 NBS 1 .0 5 97 X 10-6 2.8 0.7

2.20

' NPL 6.2 7 22 10-6 3.2 0.4

PTB 5.3 7 14 X 10-6 2.2 0.8

BIPM 17.8 6 99 X 10-6 2.5 0.1 NPL 5.8 7 02 10-6 3.2 0.5 2.50

ETL 58.

1

6 57 10-6 4.0 1 .0

PTB 5.2 7 01 10-6 2.2 0.7

249
CALIBRATION AND USE OF FILTERED BEAMS

R. B. Schwartz

National Bureau of Standards

Washington, D.C. 202 3A

Using a combination of resonant scatterers and filters, very pure beams of 2 keV, 24 keV,

and 144 keV neutrons are produced at the NBS reactor. The calibration of these beams and

their application to dosimeter calibration will be discussed.

(Dosimeter calibration; monoenergetic neutrons; neutrons; neutron beams; resonant scatterer.)

Introduction

The existence of "windows" in neutron cross sections has made possible the development of filters

which transmit neutrons in relatively narrow energy

bands, starting from a continuous reactor spectrum.

These "windows" are, of course, just minima in the

total cross section arising from destructive interference between s-wave resonance and potential scattering, and, as such, have been known and understood

for many years. In fact, the 24 keV minimum in iron

is rather infamous, since it has long been known that

a fast neutron shield made of iron becomes a copious

source of 24 keV neutrons. It remained for the MTR

group,' however, to turn this "infamous" behavior into

a virtue, when Simpson and his co-workers produced an intense 24 keV neutron beam by passing a reactor beam

through an iron filter. They also produced beams of

2 and 144 keV, by using scandium and silicon filters,

respectively. Unfortunately, the MTR was shut down

before this work could be completed, and the filters

themselves were transferred to NBS. Several other

laboratories subsequently also developed filtered beams

and Table I, taken from the compilation of Tsang and

Brugger,^ lists the current filtered beam facilities.

Table I. Filtered Beam Facilities,

neutrons/ cm^-sec.

The numbers are neutron flux in

Filter Material Fe-Al Si °2

Energy 186 eV 2 keV 24 keV 144 keV 2.35 MeV

FWHM 2 eV 700 eV 2 keV 25 keV 500 keV

BNL 3.5 X 10* 1.3 X 10*

BOMBAY 1.5 X lO*

KYOTO T.O.F.

MTR 5 X 10* 6 X 10^ 10^

MURK v. 1 X lo' 1.2 X 10* 2.5 X 10* 1.7 X 10*

NBS 2.6 X 10* 2 X 10^ 4.5 X 10^

RPI T.O.F.

USSR 5 X 10* 2 X 10* 8 X 10*

(The data for the original MTR beams are also listed,

for the sake of nostalgia.) The numbers are the re- ported fluxes for the appropriate beams, in n/cm^-sec.

The entries for the RPI and Kyoto 24 keV beams indicate

that these are used in conjunction with a pulsed linac

neutron time-of-flight facility, and hence the fluxes

are not comparable with the other beams, which are all

from reactors.

+Work supported in part by U.S. ERDA.

The "Filter Material" refers to the principal com- ponent of the filter. Small amounts of additional

materials may be added to suppress secondary peaks; some examples of this will be discussed later. Additionally,

lead or boral may be added to reduce gamma ray or thermal

neutron contamination. "Energy" is the nominal energy

of the beam so produced and "FWHM" is the full width at half-maximum of the beam. The values for FWHM for the

186 eV, 2 keV, and 24 keV beams are calculated from the

appropriate cross sections; the 144 keV and 2.35 MeV

results are measured values. The values given are typical, but may vary from facility to facility depending

upon the actual filter construction.

As we see, there are now several sets of filtered

beams, each with its own reason for existence. In this

paper we shall very briefly discuss some of the other

beams, but shall primarily be concerned with the development and use of the NBS filtered beams. This is because:

a) I know more about the NBS beams than any of the others,

and b) the principal applications of the NBS beams are, possibly, closer to the general theme of this Symposium

than the others.

Applications of Filtered Beams

For studying the properties of neutron capture

gamma rays (as opposed to using capture gamma ray

spectroscopy as a tool for studying the capturing states)

,

a neutron source with a moderately broad (but well defined)

energy spread may be preferable to the high resolution

approach afforded by time-of-f light techniques. Such a broad source, encompassing many capturing states, will

average over the Porter-Thomas fluctuations and allow

the average properties of the gaimna rays to be seen without the complications of resonance-to-resonance

fluctuations

.

Filtered beams are ideal for these experiments and,

in fact, one of the first complete sets of high-quality

keV neutron capture y-ray data was obtained with the

2 keV beam at the MTR. This type of work is now being

carried on using 2 keV and 24 keV beams by Bob Chrien

and his group ^ at BNL; the study of capture y-ray spectra

also provides part of the motivation for the Bombay

effort."* The Kyoto^ and RPI^ groups use iron filters in

time-of-flight systems to essentially eliminate background

and gamma-flash. This allows very precise total and

capture cross section measurements at the iron windows.

(The similarity of these two programs is not entirely

accidental, since Bob Block seems to have carried the

Wisdom of the Occident with him on his sabbatical journey

to the East.) The MURR group^ seems to have been most

innovative, using the well-known minimum in (liquid)

oxygen at 2.35 MeV, and the less well-known minimum at 186 eV in ^^^U to produce beams at these high and low

energies, as well as the more usual 24 and 144 keV

beams

.

250
At NBS, the principal motivation for the development of the filtered beams has been for use in the

calibration and development of neutron dosimeters, both

"active" types such as remmeters and Bonner spectrometers,

and "passive" devices such as albedo dosimeters. The

lower energy beams are of particular importance, since

typically 30% to 40% of the dose in the working environment around a reactor is due to neutrons below 30 keV,^

but there are few (if any) suitable neutron calibration

sources anywhere in this range. Preliminary data have

been given earlier:^ These will now be brought up to date.

NBS Filtered Beams

Table II. Comparison of 2 keV filtered Beams

2 keV (Scandium) Beam

The problem with scandium as a filter material is

that in addition to the window at 2 keV, there are several other windows at higher energies , which can give

a relatively high transmitted flux in the energy region

between 8 and 800 keV. Since a remmeter should give a higher response to high energy neutrons, the presence of

a significant high energy background would be disastrous

for the calibration of such instruments. Specifically,

for the BNL Sc-Ti filtered beam^ the higher energy neutron flux is listed as 25% of the 2 keV flux; the rem

dose due to these higher energy neutrons is almost twice

the dose due to the 2 keV neutrons. This type of background which is very difficult to avoid in any facility

in which a scandium filter looks at a reactor core, is

essentially eliminated in the NBS installation by the use of a through-tube in conjunction with a resonant

scatterer. The through-tube passes 10 cm outside of the

edge of the reactor core, and the collimating system

containing the filter only sees a scatterer at the center

of the tube (see Fig. 1). We use manganese to scatter

REACTOR

SHIELD

VMLL

NBS REACTOR

FUEL

SCANDIUM

FILTER

MANGANESE

SCATTERER

Fig. 1. Schematic representation of the

NBS Reactor showing the filter,

collimater, and scatterer in the

through tube (not to scale)

.

the neutrons, making use of the scattering resonance

at 2.375 keV. Although the resonant energy does not

exactly match the energy of the 2 keV window, the reson- ance is sufficiently broad (r= 400 eV) that at 2 keV

the scattering cross section is still >100 barns.-''' Physically, the scatterer is a 3 mm thick Mn-Al alloy

containing 57 atomic percent manganese. This design

eliminates unwanted core neutrons and gammas. The

advantage of this design can be inferred from Table II,

Lab.

2 keV

Flux

°L of Other

Energy Neutrons

y-Intensity

(mR/hr)

BNL 3.5 X 10® 25% 340

MTR 5 X 10® 40% ~ 1

NBS 2.6 X lO'^ 37<. 8

which compares 2 keV beams. We note that the NBS beam

has a considerably lower intensity than the other two,

but also a much lower fraction of higher energy neutrons

and a much lower gamma ray background than the BNL beam.

For our purposes, this is a very advantageous trade-off.

(Greenwood and Chrien,^ BNL, describe a scandium-cobalttitanium filter which they estimate would have 9% higher

energy neutrons, and a 2 keV flux of 2 x 10^ neutrons/

(cm^-sec) . This filter does not appear to have been

actually used for any experiments, however.) The spectrum of the NBS 2 keV beam is shown in Fig. 2. These

data were taken with a 1 atm. hydrogen proton-recoil

counter. The solid line shows the spectrum obtained with

a 110 cm long Sc filter; the dashed curve shows the effect

of adding one cm of Ti to the Sc. Even without the Ti,

SCANDIUM FILTERED

NEUTRON BEAM

no cm SCANDIUM

PLUS I cm TITANIUM

200

NEUTRON ENERGY. keV

Fig. 2. Neutron spectrum through scandium filter.

The solid curve represents the spectrum

with a 110 cm scandium filter alone, and

the dashed curve the spectrum with the

addition of 1 cm of titanium.

the main secondary peak at 29 keV has an area of only 3%

of the 2 keV peak. The addition of the Ti reduces this

peak (as well as the ones at 7, 15, and 40 keV) by a factor of about 2-1/2, at the cost of only 17% of the

2 keV peak.

251
In terms of total flux, the higher energy contaminants amount to approximately 6% of the 2 keV flux without the titanium; the addition of the titanium reduces

their contribution to 3% of the 2 keV flux.

Before leaving the matter of 2 keV beams. It must

be pointed out that the value of the cross section

minimum in scandium of 50 mb originally reported' '^^ is

almost certainly wrong. A very careful combined BNL-RPI

measurement^^ gives a total cross at the minimum of

700 mb, rather than 50. This higher value is consistent

with a very rough transmission measureoient made with

the NBS filter, and suggests that the optimum filter

thickness for a scandium filter may be rather thinner

than heretofore believed.

Table III. Comparison of 144 keV Filtered Beams

144 keV y-Intensi ty

Lab. Flux (mR/hr)

rl i K ~ JUU MURR 2.5 X 10® ~ 450

NBS 4.5 X 10^ ~ 40

Our silicon filtered 144 keV spectrum is shown

Fig. 4.

24 keV Beam

The NBS 24 keV beam is produced by an iron-aluminum

filter in another through tube, using a graphite scatterer. Both iron and aluminum have broad cross section

minima near 24 keV, but their other minima do not, in

general, overlap. Hence, it is not too difficult to produce reasonably clean 24 keV beams by judicious

choice of the thickness of iron and aluminum, and since

iron is a good gamma-ray absorber, these beams generally

do not have serious gamma ray background problems. In this instance, therefore, the use of the through tube

does not seem to be any great advantage. Our ironaluminum 24 keV spectrum is shown in Fig. 3. The magnitude of the higher energy components is typical of the

other 24 keV beams listed in Table I.

2O

cr.

100

1 \ 1

1 IRON-ALUMINUM FILTERED BEAM

25 cm IRON

36 cm. ALUMINUM

RELATIVE AREAS

ENERGY % OF FLUX

24. J keV 97.0%

74 023

135 1.6 280 1.0 364 0.16

I ~50 100 150 . 200 250 3&0- NEUTRON FNERGY , ko^/ VArVj/ . ,

40C

Fig. 3. Neutron spectrum through the iron-aluminum

filter.

144 keV Beam

Other than the 144 keV window, the only window in

silicon is at 54 keV, and the effect of this window can easily be minimized by an additional thin titanium

filter. The main problem with silicon-filtered beams

is the Y-ray background. In our case, the silicon filter

is on the other side of the through tube from the ironaluminum filter and, hence views the same graphite

scatterer. Table III compares the 144 keV beams; we see

that the use of the through tube-plus-scatterer greatly

reduces the y-background, but at the expense of a loss

in neutron intensity.

X 100

o

cc

hUJ

-z.

LU>

<

_)UJ

q:

50

10

"1—I—

r

SILICON FILTERED NEUTRON 8EAM

136 cm SILICON

2 cm TITANIUM

144 keV - 98.5%

54 keV - 1.5%

I i I I I ' I 50 !00 150

NEUTRON ENERGY , keV

200

Fig. 4. Neutron spectrum through the silicon filter.

Beam Intensity Measurements

The 2 keV and 24 keV beam intensities are determined

by counting with a one-atmosphere ''^BF3 counter. We use standard commercial counters, 5 cm diameter, varying in

sensitive length from 6 to 33 cm, and employing a thin

ceramic end window. The beam is brought in through the

center of the ceramic window, parallel to the center

wire. Since the beams are 2-1/2 cm diameter or less,

there is minimum wall effect, and it is only necessary

to know the ^''b density and the sensitive length to determine the counter efficiency. The ^''b density was determined by measuring the transmission of the counter

along a diameter, using beams of 3.86 and 54.4 millivolt

neutrons produced by NBS crystal spectrometers. The

transmission was measured relative to an identical, but

empty, counter to take into account the effect of the

0.9 mm thick stainless steel walls. At these low neutron

energies, the "^B absorption dominates the transmission,

and thus the '"b content can be determined quite accurately. The sensitive length of the counter was determined by scanning along its length with a finely collimated thermal beam. The small correction for losses in

the 2 mm ceramic window was determined by an explicit

measurement using a dummy window. We thus end up with

counters whose absolute efficiencies are known in terms

252
of the ^"^B cross section and the explicitly measured

physical properties of the counter. Measurements of

the 2 keV beam intensity, using 3 counters of lengths

6. A, 31, and 33 cm, showed a spread in values of less

than ± 1%, which suggests at least internal consistency

in our use of these counters. In the absence, however,

of any cross checks on these measurements (e.g., foil

activation) we feel that we can only quote an accuracy

of ± 5% for the 2 keV beam intensity.

This technique is better suited to the 2 keV than

the 25 keV beam, since the "'B(n,a) cross section is

'V 3-1/2 times higher at the lower energy. However, a proton recoil counter can also be used for the 25 keV

beam, as well as for the 1A4 keV beam. In this case, we use a counter which is physically similar to the

31 cm long BF3 counter, but filled with hydrogen to an accurately known pressure. Since the hydrogen cross section is very well known, the area under the proton

recoil spectrum is a direct measure of the beam

Intensity.

The 25 keV beam intensity measurements performed

with the two different types of detectors differed by

7%. We consider this to be satisfactory agreement at this stage, although we are working to understand, and

reduce, this difference. The flux quoted in Table I is

the mean of these two values and ± 7% is probably a reasonable measure of the uncertainty in this flux, as well as for the uncertainty in the 144 keV flux.

Use of the Beams for Calibration

importance of this can be judged from the albedo

dosimeter calibration curve; without the NBS 2 keV datum,

there would simply be no way of knowing the low energy

response of the dosimeter.

Future Development

We feel that "Phase I" of the NBS filtered beam

facility is complete: we have three clean beams whose

intensities are well enough known for the initial uses

to which they are being put. "Phase II" will consist

first, of optimizing the filter configurations, and,

second improving the calibrations. We feel that the

2 keV beam intensity can probably be increased with no appreciable loss in beam purity, and that the 24 keV

beam purity can be further improved. More careful work

on the corrections associated with the counter calibrations (e.g., end effects, backgrounds, etc.) together

with alternate calibration methods (foil activation,

counting with standard fission chambers) will reduce the

uncertainty in our intensity measurements. We feel that

accuracies in the 2%-3% range are a reasonable goal for

the near future.

Conclusion

At the conclusion of his talk at the 1970 Neutron

Standards Symposium, '^ Orval Simpson predicted that

"... one fine day a reactor will offer standard neutron

sources of 2, 24.5, and 144 keV produced from filtered

neutron beams." We should like to suggest that perhaps

that fine day has arrived.

The first extensive use of the beams for dosimeter

calibrations was by Dale Hankins of Livermore. Some of

his results are shown in Fig. 5.^^ The circles are the

z£6

UJ>

-1:^-1 1 I I I I I I

1 1—I

I I I I I I

I \I I I

1 1

I ^ALBEDO DOSIMETER ^ REMMETER ond DOSIMETER

CALIBRATION

0-E HANKINS, UCRL-78307 119771

I I M I ll I I I I I I I ll \ 1 l^^+ljj

Fig.

fo 100 NEUTRON ENERGY, keV

5. Energy response of 9-inch sphere and of

albedo dosimeter.

data obtained with the L.L.L. cyclograph; the squares

are the data taken with the NBS beams. While the ordinate is in arbitrary units, there has been no normalization between the two sets of data. The line is an eyeguide only. We note that the NBS 144 keV point is in

excellent agreement with the data taken with the cyclograph. The 30 keV cyclograph data represent the low

energy limit of that device; the beam quality is rather

poor at this energy and, for some of Hankins ' runs , there

was a suspicion that the data were systematically some- what low.-"-"* This is borne out by comparison of the 30 keV

cyclograph and 25 keV NBS datum for the 9 inch sphere

calibration. The 2 keV NBS data then allow the calibrations to be extended a decade lower in energy. The

Acknowledgements

The importance of using the resonant scatterer

technique to clean up the 2 keV beam was first pointed

out by Ivan SchrfJder, and we are grateful for his con- tinued advice, assistance, and close collaboration.

It is a pleasure to thank E. D. McGarry and C. Heimbach for making the proton recoil spectra measurements shown in Figs. 2, 3, and 4.

References

1. 0. D. Simpson, J. R. Smith, and J. W. Rogers, Proc.

Symp. Neutron Standards and Flux Normalization,

Argonne, Illinois (1970). CONF-701002, p. 362.

2. F. Y. Tsang and R. M. Brugger, private communication.

3. R. C. Greenwood and R. E. Chrien, Nucl. Instr. and

Meth. 138, 125 (1976).

4. E. Kondaiah, R. P. Anand, and D. Bhattacharya,

Proc. Int. Conf. Photonuclear Reactions and Applications, Asilomar, California (1973), CONF-730301,

paper 2D11-1.

5. R. C. Block, Y. Fujita, K. Kobayashi, and T. Oovaki,

J. Nucl. Sci. and Technology, 12_, (1975).

6. R. C. Block, N. N. Kaushal, and R. W. Hockenbury,

Proc. Nat. Topical Meeting on New Develop. Reactor

Phys. and Shielding, CONF-72091, 1107 (1972).

7. F. Y. Tsang and R. M. Brugger, Nucl. Instr. and

Meth. 134, 441 (1976).

8. J. R. Harvey and S. Beynon, Proc. First Symposium

on Neutron Dosimetry in Biology and Medicine,

Neuherberg/Munchen (Germany), (1972) p. 955.

253
9. I. G. Schroder, R. B. Schwartz, and E. D. McGarry,

Proc. Conf. Neutron Cross Sections and Technology,

Washington, D.C. (1975). NBS Special Publication

425, p. 89; E. D. McGarry and I. G. Schroder, loc.

cit. , p. 116.

10. M. D. Goldberg, S. F. Mughabghab, B. A. Magurno,

and V. M. May, "Neutron Cross Sections, Vol. IIA,

Z=21-40," BNL 325, Second Edition, Supplement

No. 2 (Physics TlD-4500) Feb. 1966.

11. W. L. Wilson, "Neutron Cross Section Measurements

and Gamma Ray Studies of '^Sc;" M. S. Thesis,

University of Idaho Graduate School, Dec. 1966

(unpublished .

)

12. H. 1. Liou, R. E. Chrien, K. Kobayashi, and R. C. Block, Bull. Am. Phys . See 22^, 53 (1977).

13. D. E. Hankins, UCRL-78307 (1977).

14. D. E. Hankins, private communication.

254
MUCH ADO ABOUT NOTHING: DEEP MINIMA IN ^^SC AND ^^FE TOTAL NEUTRON CROSS SECTIONS*

R. E. Chrien and H. I, Liou

Brookhaven National Laboratory, Upton, New York 11973

R. C. Block and U. N. Singh

Gaerttner Linac Laboratory, Rensselaer Polytechnic Institute, Troy, New York 12181

and

K. Kobayashi

Kyoto University Research Reactor Institute, Kumator i-cho , Osaka-fu, Japan

The deep minima in ^^Sc and 56pg neutron total cross sections have been measured at the

Gaerttner Linac Laboratory by using thick, ultra-pure samples in transmission experiments.

The samples are used to produce quasi-monoenergetic beams at the BNL High Flux Beam Reactor.

For the 45Sc minimum near 2.05 keV we obtain CTtotal" 0-71 + 0.03 barns, in sharp contrast to

a previously reported value of ~ 0.05 barns. The 56Fe measurement was carried out with a

6 kg, 58 . 58-cm- long sample of 99.87% isotopically pure sample of 56pe; a minimum cross sec- tion of 0.0085 + 0.004 barns at 24.39 keV is inferred. This may be compared to a value of

0.420 barns for natural iron.

(Neutron total cross sections, ^^Sc measured from 0.4 to 22 keV; deduced neutron resonance parameters; ^^Ye mea- sured from 0.4 to 1000 keV)

I. Introduction

Many elements exhibit deep neutron total s-wave

cross section minima due to the interference between

resonance and potential scattering amplitudes. These

minima are of considerable Interest in shielding applications and in the design of transmission filters for

the production of quasi-monoenergetic neutron beams.

For the latter application both steady state re- actors ^ ' 3 and pulsed sources^'^ have been used. The

use of such filters for dosimetry measurements has

been discussed by Schwartz^ at this meeting. Scandium

and iron-aluminum filters have been also rather widely

used for capture 'i-vay and cross section studies. The

empirical optimization of scandium and iron filters

has been discussed by Greenwood and Chrien.^ A sum- mary of known filter facilities has recently been

prepared by Tsang and Brugger.^

In spite of the wide applicability, the total

neutron cross sections of ^^Sc and ^^Fe, in particular,

are rather ly poorly known. In the installation of the

HFBR scandium filter, it became obvious that the flux

measured at this filter could not be reconciled with

the accepted cross sections. Furthermore no accurate

measurement of 56pe total cross sections has been

published. For these reasons we carried out experiments to establish the cross sections at the 2.05 and

24.4 keV minima in ^^Sc and ^^Fe . Thick metallic

samples of Sc (99.9% pure) and separated Fe (enriched

to 99.87% in SSpe) were obtained from the HFBR Tailored

Beam Facility. Neutron transmission measurements were carried out at RPI's Gaerttner Linac Laboratory, and

the data subsequently were analyzed at BNL.

II. Method

A full description of the RPI Linac Facility has

been reported, and only a brief description is given

here

.

The standard water-cooled Ta and CH2-moderated

neutron TOF target and the lOs-Nal neutron detector

at the 28.32-m flight path were used for these measurements. The linac was operated at a repetition

rate of 500 sec"-'-, an electron energy of ~ 70 MeV, a peak electron current of 1 A, and an electron pulse

width of either 19, 35, or 66 ns . The counting data were recorded vs. TOF with the 31.25-ns TOF clock

interfaced to the PDP-7 on-line computer. The

transmission samples were cycled automatically into

and out of the neutron beam by the programmed com- puter, and a cycle was repeated every 10 to 20 minutes

to average out neutron source intensity fluctuations.

The composition of the '^^sc and 56pg samples are listed in Table 1. The ^^Sc samples were prepared by

Table 1 Sample Properties

(A) Scandium Samples

Sample No. Dimensions (cm) 1/N (barn/atom)

1 2.54 diameter x 10.2 2,.738

2 2.54 diameter x 15.2 1,.844

3 2,54 diameter x 30.5 0,.922

4 2.54 diameter X ^.2 152,,5 5 2.54 diameter X 0.55 50,.8 6 2.54 diameter X 2.4 11..7

Impurity Atom Per Cent

H

0

Ta

(B) ^^Fe Sample

Dimensions (cm)

3.22 X 3.85 X 68.6

Isotope

5^Fe

56Fe

57Fe 58Fe

0.49

0.18

0.037

1/N (barn/atom)

0.175

Wt. Per Cent

0.05

99.87

0.07

0.006

Impurity Atom Per Cent Impurity Atom Per Cent

0

Cu

H

CN

.617

.120

.059

.020

.012

Cr

W

Ni

Si

P

.011

.010

.005

.002

.002

vacuum sublimation onto a cooled Ta plate, and the

sublimed metallic material was subsequently pressed

into steel containers 2.54 cm in diameter. The H

and 0 impurities were determined by the vacuum fusion

method, and the heavier elements by mass spectrometry.

255
for both samples, at the Ames Laboratory Analytical

Center. In addition the Ta content of the Sc was checked by emission spectrography , by resonance x-ray

fluorence, by neutron activation, and by neutron

transmission. This wide variety of methods for Ta

analysis was necessitated by the discovery of a sharply inhomogeneous distribution of Ta impurity

throughout the bulk of the Sc filter. Both activation and transmission methods were able to sum over

large sections of the filter and thus produce a more reasonable average impurity content than was the case

for the other techniques, which were applied to small

samples of the filter material.

The ^^Fe sample was prepared at Oak Ridge

National Laboratory from elec tromagnetically enriched

iron In the form of iron oxide. The metallic sample was obtained by reducing the oxide in a hydrogen

atmosphere. The isotopes and impurities listed in

Table 1 were determined from measurements of the iron

oxide and metal respectively.

In the final analysis, these cross section mea- surements are only as reliable as the impurity content

determination. At the Sc minimum, the impurity cor- rection is about 112 mb out of a measured 822 mb

;

while in the ^^Fe, the correction is about 51.5 mb

out of a measured ~ 60 mb . In each case ENDF/B IV evaluated cross sections were used in the correction.

RELATIVE NEUTRON INTENSITY FOR 2 , 8

50 40 30 20

45

Two sets of Sc transmission measurements were carried out. In one measurement the 10.2-cm-long

scandium sample No. 1 was placed in the neutron beam

to produce a TOF-filtered spectrum of neutrons which

is peaked near the interference minimum in scandium.

This removes most of the neutrons from the beam and

results in a very low background of neutrons with

energies far from the minimum. Then samples No. 2 and

3 were cycled into the filtered beam to obtain an accurate measurement of the cross section near the

interference minimum. The other measurement was a conventional transmission measurement. The 10.2-cmlong scandium filter was removed from the beam and

samples No. 4, 5 and 6 were cycled into the beam.

This latter measurement enabled the cross section to

be determined near the peaks of the resonance as well

as near the minima.

For the 56pg measurements a 20.3-cm-long filter

of pure Armco iron was placed in the neutron beam to produce a TOF-filtered spectrum peaked near the iron

minimum. The 68.6-cm-long ^^Fe sample was then cycled

into and out of the filtered neutron beam. The filtering effect in iron Is illustrated in Fig. 1 near

the 24 keV minimum. Here the ^'-'B-Nal detector relative counting rate is shown for Armco iron filters

varying in thickness from 5.1 cm(2") to 50.8 cm(20")

.

For this experiment the 20.3 cm(8") thickness was used,

and from Fig. 1 we see that the peak transmission

through the filter is 487o and that most of the neutrons

at energies several keV away from the peak are removed

from the beam.

. 14". AND 20" IRON FILTERS

(KeV)

10

z

<

X

o

oc

a.

Ill <

01

o

o

UJ>

q:

I I I I Ii I i r

r2' (84% TRANSMISSION)

8 (48% TRANSMISSION)

14" (28% TRANSMISSION)

20" (16% TRANSMISSION)

500 600 700

CHANNEL NUMBER

Fig. 1: 10„ The ""B-Nal detector counting rate vs. TOF with Armco iron filters 5.1-cm(2"), 20 . 3-cm(8'') , 35 . 6-cm(14"")

and 50.8-cm(20") thick. The peak transmission through each filter is shown in parentheses.

256
The TOF counting data were corrected for deadtime losses In the electronics and for background,

and the neutron transmission was then determined. The

total cross section Oj. was obtained from the neutron

transmission with the following equation

= -(1/N) In T (N . CT . )/N air air (1)

where N is the thickness of the scandium or 56pg

sample, T is the neutron transmission, Ngir is the

thickness of air displaced by the sample, and o g^±-^ is the neutron total cross section of the air. The cross section of air was obtained from the oxygen

and nitrogen cross sections plotted in BNL-325. 10

III. Results

A. 45 So

The ^^Sc total cross section obtained from

equation (1) is plotted in Fig. 2 over the neutron

energy range from ~ 0.4 to 22 keV. This plot is a "blend" of all the Sc data, where the data near the

peaks in the cross section are from the thinnest

sample and the data near the deep minima are from the

thickest sample. The data have been corrected for the

presence of the contaminants listed in Table 1.

in the higher energy ^^sc minima, and thus the length

of the ^^Sc filter should be limited to enhance the

transmitted neutron flux near 2 keV relative to that

transmitted at higher energies.

The ^^Sc total cross section has been fit by

an approximate R-matrix formalism, ^3 a^d the solid curve through the experimental points in Fig. 2 is a "best fit" to the data. The curve has been reso- lution broadened; the resolution FWHM is approximately

3 channels near the 2 keV minimum. In Table 2 are listed the resonance parameters derived from this fit.

Table 2

45

Sc Resonance Parameters

s-wave level parameters = 0.4 eV

Eo (eV) Tn (eV) J Eo (eV) J -500 3 11575 290 4 -220 0.67 (rn°) 4 14525 20 3 3295 75 3 14740 26 4 4330 340 4 15560 28 4 6684 130 3 15850 5 3 8023 145 4 18580 32 3 9092 300 3 18870 62 4 10625 10 3 20500 80 4 10735 6 4 20780 710 3

i

s z

a-*

<D

EiKEvi

Fig. 2: The neutron total cross section of ^-^Sc . The

experimental data are shown as solid points

with error bars (standard deviations determined from the counting statistics'* . The

solid curve is a resolution-broadened "R- matrix" fit to the data using the resonance parameters listed in Table 2.

The measured minimum cross section of ^^Sc near

2 keV is (0.71 + 0.03)b, where the error is derived

from a combination of the counting statistics and the

uncertainties in the H and 0 corrections. The mini- mum occurs at an energy of (2.05 + 0.02) keV. This

minimum cross section of (0.71 + 0.03)b is in serious

disagreement with earlier measurements reported by

Wllsonll of ~ 0.05 b, and it is also in disagreement

with the evaluated minimum cross section^^ of ~ 0.085b

which was based on these older measurements. Our

result of 0.71 b has serious implications for the

effective use of expensive scandium for filtered

beams. The 0.71 b cross section at 2.05 keV is

comparable to, or larger than, the cross sections

.This fit was determined by the following procedure:

(a) The spins of the positive energy resonances

were obtained by fitting the peak cross sections and

the shape of the interference between resonances.

(b) Negative energy levels were introduced to fit the cross sections at thermal and in the low energ

region

.

(c) The neutron widths were obtained for all the resonances such that (i> the calculated R-matrix cross section curve produced an acceptable overall fit to the data, and (ii) the R-matrix minimum total cross section near the 2 keV minimum equaled the observed

value of 0.71 b.

(d) A single radiation width was then determined

for all the resonances such that the thermal capture

cross section resulting from the sum of contributions

from all the resonances equaled the evaluated value^'-^ of (26.5 + 1.0) b.

The best fits to the data are obtained when the

J=3 channel spin contributes significantly to thermal

capture. The "best fit" parameters listed in Table 2 produce a thermal capture cross section which has

approximately equal contributions from J=3 and J=4

channel spins.

Thermal neutron capture Y-ray spectral measure- ments by Bolotin"'' favor a significant J=3 channel spi

contribution. He observed a primary gamma-ray transition to the 1" state in ^^Sc at an excitation energy

p-wave level parameters

257
of 142 keV. Thermal capture in scandium consists of

a mixture of capture into 3" or 4" states and

Bolotin's observed transition strength of 1.3 gammas

per 1000 captures indicates that this gamma ray is an E2 transition from a 3" to a 1" state. The partial radiation width for this E2 transition can be

calculated from the observed transition strength and

the resonance parameters deduced from the R-matrix

fit to the total cross section. The E2 width calculated from the "best fit" parameters in Table 2 is

about 6 times larger than the typical E2 width observed

in this mass and energy range, and this is reasonable

considering the fluctuations of the observed E2 widths.

However, when the E2 width is determined from the R- matrix parameters which produce predominantly J=4

channel spin thermal capture, the E2 width is about

500 times larger than the typical E2 width.

Such a 500 times larger E2 width is very un- likely, and thus Bolotin's measurements favor thermal

capture which has a significant J=3 channel spin con- tribution. We have independently confirmed Bolotin's

observation in a separate experiment to be reported

elsewhere

.

Channel NunoSR

A polarized neutron experiment by Roubeau et al.-'-^ claims to have measured the difference in scattering

lengths at thermal between the (I + 1/2) J=4 and

(I - 1/2) J=3 spin states. They report a value of

(a+ - a_) = + 1.2 x lO"-*-^. cm. The implication of

their result is that a J=4 state dominates thermal

scattering. This result is inconsistent with our present result, since it would mandate a deep minimum

near 2 keV, as a result of interference between the

dominant J=4 resonance at 4.330 keV and a strong J=4

bound state. An attempt to fit our data subject to this constraint results not only in a poor fit, but

requires absurd R-matrix parameters, (e.g.,

Rj. ^/Rj-fi, 8) . We conclude that the experiment

of Roubeau et al. is in error.

Fig. 3: The B-Nal detector counts vs. TOP channel

number for a 20.3-cm(8") Fe filter placed

in the neutron beam. The TOF channel width

is 31.25 ns.

B. 56Fe

The ''^B-Nal detector counts per TOF channel are plotted in Fig. 3 for the 20.3-cm (8") Armco iron filter in the neutron beam, and in Fig. 4 for the 20.3- cm Armco iron filter plus the 68.6-cm (27") ^^Ye. sample. The 68.6-cm 56pg sample produces a quasimonoenergetic peak of neutrons which peaks near an energy of 24 keV

.

The total cross section of 56pg near the 24 keV

minimum is plotted in Fig. 5. For iron both impurity

cross section corrections and resolution corrections

are very important, which was not the case for

scandium. The total correction for impurities

amounted to 51 mb in the region of the minimum; and we infer a minimum observed cross section of ~ 15 mb

at an energy of 24.39 + 0.04. This cross section is,

however, seriously affected by our resolution function,

which has a FWHM of about 200eV at the minimum (~ 2.1

channels). Using the parameters of Pandey and Garg''-^ we fitted a series of resolution broadened cross sec- tion curves to the minimum, using the radiation width

of the major s-wave resonance near 27 keV as a parameter. We find the best fit for a radiation width

of about 2.2 eV , which provides an independent mea- surement of that important parameter. We also include

a p-wave potential scattering contribution in this

calculation. From the best fit, we infer a minimum

cross section of 8.5 + 4 mb , of which 1.3 mb is due to l=\ scattering, and the balance is due to the (n.y)

reaction.

11- 'Te -g"Fe

CHANNEL NUMBER

Fig. 4: The B-Nal detector counts vs. TOF channel

number for a 20.3-cm(8") Fe filter plus a 68.8-cm(27") 56Fe sample placed in the neutron

beam. The TOF channel width is 31.25 ns

.

258
25,0

Fig. 5: The neutron total cross section of Fe near the Ik .Yl keV tnlnlmum. The

solid curves are resolution-broadened "R-matrix" calculations using the resonance parameters shown in the figure.

259
The minimum cross section measured for

this 5^Fe sample is almost two orders of magnitude

smaller than the ~ 420 mb cross section measured for

1 / 1ft elemental Fe . ' This results in a much more intense beam of quasl-monoenergetic ~ 24 keV neutrons

from this filter than from a filter of Fe of the same

length. For example, for the 68.6-cm long filter with

1/N = 0.175 b/a, the transmission through the ^^Fe

filter at 24.4 keV is 72% (allowing for impurities),

whereas the transmission through the same thickness

of Fe is only 97=. Thus a quasi-monoenergetic beam of ~ 24 keV neutrons can be obtained with a SBpe filter

which has excellent transmission through the 24.4 keV

minimum. The filter can be used in very thick con- figurations to reduce unwanted fast neutrons and gamma

rays

.

Summary

The neutron total cross section has been measured

for ^^Sc and ^^Ye with particular emphasis upon mea- suring the cross section in the minima. The "^Sgc cross section has a prominent minimum at (2.05 + 0.02) keV

which is (0.71 + 0.03) b. This cross section is an order of magnitude larger than estimated from earlier

measurements, and this has serious implications in the

design of a '^5sc filter for reactors. Although the

design of a ''''^Sc filter system depends upon the application of the system (e.g., for neutron capture spectra, dosimetry, etc.), this higher cross section of

0.71 b should lead to the selection of a thinner Sc filter than one based on the former ~ 0.05 b value.

S The Fe cross section has a prominent minimum

at 24.4 keV which is (8.5 + 4.0) mb . This is con- siderably smaller than the ~ 420 mb minimum in elemental iron, and thus thick filters of 56Fe can provide intense quasi-monoenergetic beams of 24 keV

neutrons with a very small contamination of gamma

rays and fast neutrons.

"Z . M. Bartolome, R. W. Hockenbury, W. R. Moyer

,

J. R. Tatrczuk and R. C. Block, Nucl . Sci. and Eng.

37, 137 (1969) . 1%. I. Garber and R. R. Kinsey, BNL-325, 3rd Ed.,

Vol. 2, (1976).

L. Wilson, M.S. Thesis (Univ. of Idaho, 1966),

unpublished

.

'-^B . A. Magurno and S. F. Mughabghab , Proc . Conf. on Nuclear Cross Sections and Technology, NBS Spec.

Pub. 425, Vol. 1, 357 (1975) . 13r. g. Thomas, Phys . Rev. 97, 224 (1955); F.W.K. Firk,

J. E. Lynn, M. C. Moxon, Proc. Phys. Soc . 82, 477

(1963) . I'^H. Bolotin, Phys. Rev. 168, 1317 (1968).

15p. Roubeau, A. Abragam, G. L. Bacchella, H. Glaetti,

A. Malinovski, P. Meriel, J. Piesvaux and M. Pinot,

Phys. Rev. Lett. 33, 102 (1974).

S. Pandey, J, B. Garg, J. A. Harvey and

W. M. Good, Proc. of Conf. on Nuclear Cross Sections

and Tech., NBS Special Publ. 425, Vol. 2, 748 (1975).

l^J. A. Harvey, Proc. of New Devel. in Reactor Phys.

and Shielding, CONF-720901, Book 2, 1075 (1972).

l^K. A. Alfieri, R. C. Block and P. J. Turinsky,

Nucl. Sci. and Eng. 51, 25 (1973).

Acknowledgement s

The authors wish to acknowledge the valuable

contribution of B. Beaudry in arranging for the vacuum fusion and mass spec trographic analyses of

the samples at Ames Laboratory. Valuable discussions

were held with Larry Passell, Harvey Marshak,

Wally Koehler, Marty Blume and Said Mughabghab. The

able assistance of Frederick Paffrath in the sample

preparation is much appreciated.

References

*Work supported by the Energy Research and Development

Administration

.

'"0. D. Simpson and L. G. Miller, Nucl. Instr. and

Meth. 61, 245 (1968); and U.S. Atomic Energy

Commission Report I N-1218 (1968) . ^R . B. Schwartz, contribution to this conference.

3r. C. Greenwood and R. E. Chrien, Nucl. Instr. and

Meth. 138, 125 (1976) . ^R. C. Block, N. N. Kaushal and R. W. Hockenbury,

Proc. of New Devel. in Reactor Phys. and Shielding,

CONF-720901, Book 2, 1107 (1972).

5r. C. Block, Y. Fujita, K. Kobayashi and T. Oosaki,

J. of Nucl. Sci. and Tech. 12, 1 (1975); Y. Fujita,

K. Kobayashi, T. Oosaki and R. C. Block, Nucl. Phys.

A258 , 1 (1976).

^R . B. Schwartz, contribution to this conference.

^F . Y. Tsang and R. M. Brugger, private communication.

University of Missouri.

8r. W. Hockenbury, Z. M. Bartolome, J. R. Tatarczuk,

W. R. Moyer and R. C. Block, Phys. Rev. 178 , 1746

(1969).

260
THE U-235 NEUTRON FISSION CROSS SECTION

FROM 0.1 TO 20,0 MEV

W. P. Poenitz

Argonne National Laboratory, Argonne, Illinois 60439

The status of the U-235 fast neutron fission cross section is discussed based

primarily on material contributed and considered at the NEANDC/NEACRP Specialists

Meeting on Fast Neutron Fission Cross Sections held at Argonne National Laboratory in

June 1976. However, some newer measurements and evaluations are discussed as well.

Specifically, recent measurements at ANL over the energy range 0.2-8.2 MeV, using

several BND's, are reported. Data from the last 10 years are found to be in good

agreement with an evaluated average of these data. Suggested problem areas are investigated in terms of their actual significance. It is found that the presently

suggested version of ENDF/B-V for the U-235 (n,f) cross section does not represent

the data base well and a reconsideration is recommended.

(U-235 (n,f), 0.1-20.0 MeV, Status, Data, Evaluations)

Introduction

An NEANDC/NEACRP Specialists Meeting on Fast

Neutron Fission Cross Sections of U-233, U-235, U-238

and Pu-239 was held at Argonne National Laboratory in

June, 1976^. One of the major sessions dealt with

the absolute U-235 cross sections. A workshop on absolute cross sections considered essentially U-235

and presented conclusions and recommendations. The

material presented at this meeting, and considered

by the workshop, forms the base for the present investigations. However, additional data were included

and further extensive analysis was carried out. This

has led to the present status report and recommendations which go beyond those of the above meeting.

Recent Experiments

Most of the experiments considered here have

been described in detail in the literature, and some

of the detectors were discussed in other sessions of

this symposium. Thus, the present account will be

extremely short.

WoMon {'76 AML. Rei^.Z)

This was a very important and extensive contribution

on U-235 (n,f) at the '76 ANL Meeting. Though the

actual normalization of the data was in the 7.8-11.0

eV interval, the experiment follows a suggestion by

C. Bowman^^ that all cross sections be measured as shape data and then normalized at thermal energy.

This would utilize the well known thermal cross section.

The measurements were carried out at the NBSLinac using the time-of-flight technique and a "white

neutron-source" spectrum. First, the ratio U-235

(n,f ) /Li-6(n,ct) was measured from 7.8 eV to 30 keV

using a multiple layer ionization chamber and a Liglass detector. Secondly, the ratio of U-235(n,f)/

H(n,n) was measured with the same ionization chamber

and a recoil detector. The U-235 (n,f) cross section

was obtained from the U5/Li-measurements using an R- Matrix fit^^ for the Li-6 cross section and normalizing in the 7.8-11.0 eV interval to an absolute

value reported by Deruytter and Wagemans^. The U-235

(n,f) cross section obtained from these measurements

in the 10-20 keV interval was then used to normalize

the cross section obtained from the U5/H-measurements

.

The two sets of data differed in the remaining over- lap-range below 10 keV by 5% but agreed rather well

in the 20-30 keV range (better than 1%). The quoted

uncertainty of the results is typically 1.6%, not

including normalization uncertainties.

Leaqzu clI. {'76 ANL, Re.f,.4)

The shape of the ratio U-235 (n,f)/H(n,n) was measured between 1.2 MeV and 20 MeV. A multiple gasscintillation chamber was used and both fission fragments were detected in coincidence. The neutron flux was measured with a special proton recoil telescope.

Both detectors were positioned at a 59 m flight-path

in the white neutron-source spectrum obtained from the

KFK-Cyclotron. The systematic uncertainties of these

measurements are between 4 and 8%.

Szabo and MoAqueJXz {'76 AML, Rei^.S)

An extension of previous measurements to the 2.3-5.5

MeV energy range was reported. The efficiency of the

directional neutron monitor was determined relative to the D(d,n) reaction cross section and normalized below

2.2 MeV where it was known from two previous independent calibrations. The paper also contains a listing

of all previously reported data in its finalized form.

BoAton U al. {'76 ML, Rz{,.6]

A summary of the previously reported data was presented

and some time was devoted to considering the possibility of an energy shift around 6 MeV between the data

from LASL and LLL. It was shown that the suggestion

of an energy shift results from insufficient consideration of statistical uncertainties. The experimenters

have in the meantime extensively investigated their

energy scale calibration using threshold reactions, resonances in carbon, and the AjI(p,y) reaction. The

largest difference which could be found was 20 keV, but

under more realistic operation conditions the estimated

uncertainty would be much less.

These measurements will be reported later in this

session. Four values were obtained using calibrated

photo neutron sources and one value was obtained with

a Cf-252 source. The uncertainties are 2% for the

former and 1.6% for the latter.

Cawce and GKunldfi {'76 ANL, Re.{,.8]

An absolute value was obtained at lA MeV using the

associated particle technique and the T(d ,n) a-reaction.

The experiment was carried out in time-of-flight and

in coincidence with the associated a-particle. The

result has an uncertainty of 1.9%.

H&aXon at. {'76 ANL,

This experiment will be reported later in this session.

A value was obtained for a calibrated Cf-252 source with 2.2% uncertainty.

ZhuAavUv at al. {'76 LomM, Raf^.lO]

Several values were measured relative to the B-10(n,a)

cross section using filtered reactor neutron beams.

One of the values falls in the energy range considered

here and was obtained at 144 keV using a Si-filter.

261
Previous measurements with a Black Neutron Detector

were extended to lower and higher energies using

various sized detectors. The basic experimental setup was similar to that previously reported. An

ionization chamber with two U-235 samples back-toback was placed 'x- 8 cm from the neutron source (Li-7

(p,n) below 4.5 MeV and D(d,n) above). The neutron

flux was measured with a END positioned behind a collimator at a distance of 5 m. At low energies,

a small END with only one photo-multiplier was used

and the modification of the code "Carlo Black" by

G. Lamaze from NES for the Poisson statistics of

photo electrons was utilized. At higher energies, a 16 multiplier-detector was used. An additional shape

measurement with improved resolution was made by detecting the higher energy (> 6 MeV) yi^ays associated

with the fission process in a metallic sample with a large-liquid scintillator. The energy resolution of

this experiment was 6 keV in the 200-300 keV range.

Data were grouped in 10-keV intervals and normalized

using the absolute values. The latter have a typical uncertainty of 2%, but are larger at higher

energies.

Data Consideration

It was noted at the '76 ANL Meeting that all more

recent data can be found in a ± 3% band. Though

statistics or systematic errors cause some points to scatter outside this band, the general systematic

trends are with few exceptions within this band.

Some. Sp2.cJ.^ic F^oblmi

As a result of the '76 ANL Workshop on absolute

cross sections, it was stated that the ±3% band

would have to be replaced with a ±5% band in a local

region from 200 to 400 keV. Another notable difference appears to be between the data by Szabo and

other sets above 3 MeV. Fig. 1 shows the relevant

data sets. The difference at 280 keV between the

data by Wasson^ and those by Szabo^ and by Poenitz^^

is about 13%. The systematic difference between the

data by Szabo^ and by Barton et al.^ and Czirr and

Sidhu^^ is about 6% around 4 MeV.

100 1. 00 EN/MEV 10. 0

Fig. 1. Comparison of some data in the 0.1-10 MeV

energy range. The difference between the data by

Wasson and those by Poenitz and by Szabo is about 13%

at 280 keV. The difference between the data by Szabo

and those by Barton et al. and by Czirr and Sidhu is

about 6% at 4 MeV.

1 I I I I I I I I I I M I J, I

U-235 O

o

WASSON SZA60

POENITZ 77 6N0 1

100 1. 00 EN/MEV 10. 0

Fig. 2. Comparison of the data by Szabo with new re- sults by Poenitz.

Fig. 2 shows the results from the new ANL measurements which support the data by Szabo in both

areas under question. Structure around 270 keV may

be construed from the new data set but the dip would

be only 4% and 10 keV wide based on the ANL

measurements compared with 9% and 'v 30 keV for the

Wasson data. A better comparison can be made for the

ratio of the 250-300 keV interval vs. the 200 - 250

keV interval:

Poenitz (13) GND '72 .958

Szabo (5) '70-'73 .967

Gayther (14) '75 .940

Wasson (2) '76 .907

Poenitz (11) END '77 .955

. 050 . 100 EN/MEV

Fig. 3. Comparison of fluctuation in the U-235 (n,f) cross section obtained in the 50-400 keV energy range.

262
The applicable uncertainties for these values are unknown but presumably are 2%. The problem around

280 keV suggests a comparison of structure in other

energy ranges, in order to consider the possible influence of such structure on cross section averages.

Fig. 3 shows the three available data sets^» ' ^ (a

smooth cross section, 0.7585-0.26356 ' In (E) , was

subtracted). Some structure is common to all three

sets, but some dips or peaks occur only in one or two

of the three data sets. Correlations with the Adscattering cross section could not be found.

A BcUAJs i^Q/i StcLtuA , PiobZm AnaZyi-u and

Vcuta Compa/biiion

Many problems, including some discussed during the

'76 ANL Meeting, appear to be based on a poor understanding of the result of an experiment on the part

of "data interpreters". This situation is most

vividly demonstrated in the still continuing practice

of displaying or discussing differences in data without displaying or taking into account the uncertainties of these data sets or the uncertainty in the

knowledge of the considered quantity. It is imperative to judge the significance of existing differences by relating them to their uncertainties.

Another important criterion for determining the significance of a suggested problem is the magnitude of its

practical implication.

For further investigation of the status of U-235

(n,f), an evaluated cross section is desirable. The

comparison of individual data sets and conclusions

about possible problem areas should be done using

such an evaluated "best value set". For practical

purpose we restrict the data base here to values re- ported since the experiment by White'- ^ (.'^ 12 years

ago), thus all discussion is restricted, by definition, to this data base. Following standard procedures and previously described techniques'^ an evaluated "best value set" was determined. It represents

practically a weighted average of the data base

(which was, with one exception, not altered).

An extensive display of existing data sets was made in the supplements of the proceedings of the ' 76 ANL Meeting-'. Here we consider instead the quantity

a_. - a_ D/U

/(Aa.)^ + (Aa P~ 1 ev where a ± Aa is an individual measured cross section ana a ± Aa the evaluated "best data set"

cross section at fKe same energy. In other words,

D/U represents the deviation of a measured value from

the best value in terms of the uncertainties of both,

the measurement and the knowledge of the measured

quantity. The above definition of D/U permits us to define criteria for the existence or non-existence of

significant problems:

1. Fluctuations of D/U values should be expected to follow a normal distribution. This applies to both,

individual point-scatter and larger range, systematic

variations. Fluctuations must be expected for

statistical and for many systematic uncertainties.

2. Average D/U-values should be within ± 1. The cause is most likely an error of the normalization

which should not exceed the estimated total uncertainty.

Vata CompaAAJiOn

Figs. A and 5 show the D/U values for the data which

contributed to the U-235 (n,f) evaluation. The shape

data by Czirr and Sidhu'^, and by Leugers et al.*^, were normalized to the absolute 14 MeV value by Cance

and Grenier^. The shape data by Gayther et al.'"*

were normalized to the absolute values from the U. of

Michigan^. We can conclude that, for the high energy

range (E>1 MeV) , no significant problem exists for

U-235 as a whole nor for individual data sets. However, it may be noted that all data but those by

Czirr and Sidhu'^ tend to suggest lower values between

2 and 5 MeV. It should also be noted that the suggestion of an energy shift of the data by Barton et al.^ cannot be supported by the present analysis as significant.

At lower energies (< 1 MeV) the data set by

Kaeppeler'^ fails to follow anything close to a normal distribution. The 100 keV-averaged data by

Wasson^ still reflect the influence of the structure

previously discussed but, in general, follow the shape

measured by Gayther et al.''*, Poenitz' '

' ' ^ and Szabo^.

The data by Wasson appear to be inconsistent with the

evaluated result as they are systematically lower by

a D/U of 2 or more. However, a proper no-rmalization

and accounting of normalization uncertainties (which

will be discussed later) shift these data into a satisfactory D/U range.

In Table 1, the average deviations of all absolute values which contributed to the normalization

of the evaluated cross section are listed. The

average uncertainty and the D/U values are also

listed. Again, there is no indication of any major

problem after adjustments of the data by Wasson were

made. The uncertainty of the normalization of U-235

(n,f), based on the values given in the Table, is

0.6%. This uncertainty covers the unweighted average

normalization factor. A check for a dependence of

the results on the time of the measurements, which in

a similar analysis was shown at the '70 ANL Symposium^

may still indicate some trend, however, exclusion of

pre-1975 data would change the average normalization

by no more than its uncertainty.

TABLE I. Average Difference of Absolute Cross Section

Values from the Evaluated Data Set of U-235(n,f). All

Data Since and Including the Experiment by White.

Dependent Data were Lumped Together.

First Author Ref.

Average

Deviation

%

Average

Uncertainty

%

* D/U

Smith 22 -7.5 6.3 -1.3

White 16 +2.6 2.6 +0.9

Kaeppeler 18 +3.0 3.0 +1.0

Kuks 23 +5.5 3.6 +1.4

Poenitz 13 -1.0 3.6 -0.3

(W/0 BND)

Szabo 5 -0.2 3.0 -0.1

Barton 6 +0.4 1.5 +0.2

U. Michigan 7 -0.2 2.0 -0.1

Cance 8 -1.5 1.9 -0.8

Adamov 24 +3.6 1.6 +2.0

Heaton 25 -1.3 2.2 -0.6

(Wasson 2 -4.0 1.7 -2.3)

not used

Wasson See text -2.8 3.8 -0.9

(revised)

Poenitz (BND) See text -1.3 2.4 -0.5

Data rel.BlO 10,15 -3.2 3.1 -1.0

Data rel.Li6 26 -4.8 5.2 -1.0

Values of D/U < ± 0.7 are expected with 50%

Probability.

263
3^

1 1 1 1 1 MM 1 1 1 1 1 1 1 1

1

r\j

O U-235

<!> m

SMITH

KUKS

oo

t

1

1

1

1 III III MMMM 1 1 1 1

—

1 1 1 II II

1 1 1 II II

:r

1 1 1 1 1 Mil 1 1 1 MINI

O

<!> BARTON

fM

1

>-

1

1 1 1 1 INI 1 1 1 1 1 1 II

1

CM

O

CM

I

<!> LEUGERS

CANCE

<!><!>

CM

CM

1

I

<r>

<!> 5ZAB0

I I I I 1 1 1

1

100 1. 00 EN/MEV 10. 0 20.

Fig. 4. Comparison of experimental data with an evaluated cross section.

The difference relative to the uncertainty is shown.

264
3^

1 1 1 1 i MM 1 1 1 1 1 1 1 1 1

CM _ <^ <>

<i> o —

O <!> o —

CM

t — O WHITE

1

1 1 1 1 1 1 1 1 1 Mill 1 1

CM

O

(M

(

<!>

<!>

<t> <ls!^ 0

O POENITZ '74 GNQ

6ND

A. ACT.

-<u

z

ID

o

^ 7'

<I><J><I>

<3>

U. MICHIGAN

GAYTHER

CM

O

CM

t

1 1 1 P Mill 1 1 II II

- • , •••^

• 1 • 1 1 1 1 I 1 1 1

• WASSON

K WASSON

<!> KAEPPELER

1 1 1 1 1

CM

O

CM

POENITZ 77 BND

<!>

<!>

<!xi>

100 1. 00 EN/MEV 10. 0 20. 0

Fig. 5. Comparison of experimental data with an evaluated cross section.

The difference relative to the uncertainty Is shown.

265
The only data which were changed for the present

evaluation were those by Wasson. The original

normalization, which was based on one specific experimental value, was replaced by the normalization to an average value obtained from a large number of experiments for the 7.8 eV-11.0 eV interval (obtained by the

'76 ANL Workshop2^), The uncertainty for the normalization consists of several components which are listed in Table 2.

the Wasson Data Above 1001 keV. Uncertainty

Source Ref

.

%

1. Average Stated Uncertainty 2 (Statistical + Systematic) 1.6

2. Added Uncertainties Related to the Normalization

Uncertainty of Norm. Integral 27 2.4

Statistical, 7.8-11.0 eV 2 0.8

Statistical, 10-20 keV, U5/Li 2 0.8

Statistical, 10-20 keV, U5/H 2 0.6

Systematic, U5/Li, eV-keV 2 1.0

Systematic, U5/H, 10-20 keV 2 1.6

Li-6 Cross Section, eV-keV 34 0.5

H Cross Section, Tens to 34 0.7

Hundreds keV

Systematic Discrepancy in Overlap

Range None Added

Total Uncertainty (1. and 2.) 3.8 %

CompcLfois on M-iih OtheA. Evalucutioyn

In Fig. 6 two other evaluations are compared

with the present evaluation result. The recent evaluation by Konshin^'' represents the data base very

well. The proposed ENDF/B-V^l is generally lower below 1 MeV and higher above 1 MeV than the existing

data base would require. The deviation is substantial below 1 MeV and caused in part by placing un- justified emphasis on a preliminary data set. It was specifically suggested by the '76 ANL Workshop that

the use of preliminary data should be restricted or avoided. The upper part of the figure shows a com- parison with the ±3% band considered at the '76 ANL

Meeting.

practical importance beyond their implications in high

resolution differential measurements^^. The knowledge

of U-235 (n,f) is backed by a sufficient number of

independent techniques and results.

0() TechnA.qaeA

A sensitivity study based on the present level of un- certainty and the number of independent techniques

and experiments involved, lead to the requirement that

any new measurement should have an uncertainty of less

than 2%. Any difference which is found in respect to the average of presently available data should be

backed up by an absolute D/U of at least 2 in order to be significant. Table 3 summarizes the status of

present techniques for obtaining absolute values above

100 keV. The table suggests that the preferable

techniques will be those involving the determination

of absolute masses and efficiencies rather than the

reliance on normalization at low energies. The present status and expectations for future improvements

of techniques indicate that improvements in the cross section can only be expected from a large number of

independent measurements.

TABLE III. Present Status of Experimental Uncertainties for Absolute Values Above 100 keV.

Technique, Reference

Quoted

Uncertainty

Improvemnt

%

Calibrated Bath, Absolute

Mass Determination

Poenitz(13) 1972,74 3.6

Szabo(5) 1970,73 2.5

Heaton(25) 1976 2.3

Davis(7) 1976 1.7

Associated Activity

Poenitz(13) 1972,74 3.6

Associated Particle,

Absolute Mass Determination

Szabo(5) 1970,73 2.5

Kuks(23) 1973 3.8

Cance(8) 1976 1.9

Hydrogen Recoil, Absolute Mass

Determination

White(16) 1965 2.3

Kaeppeler(18) 1972 2.1

Barton(6) 1972,76 1.5

Conclusions

ChanqeJ, oj the U-2'5S[n,i]Cn.o{>f> S&cJxon

At the panel discussion of the 1970 Helsinki Conference, R. F. Taschek pointed out^^ that one can draw

a fine line between the data by White^^ and other

English^^ and LASL measurements^'' which would be consistent with all but one data set. The presently

evaluated average cross section is 7-8% lower in the

hundreds of keV range and a further lowering by 1%

might be expected.

StatLLi ofs U-235 [n,{',] Above WO keV

The fact that no data exist which conflict with the

weighted evaluated data set for U-235 (n,f) suggests

that the evaluated uncertainty represents a reasonable

estimate of the real uncertainty. The evaluated un- certainty is less than 2% below 2 MeV, less than 3%

below 8 MeV, and increases to 5% at 20 MeV. Fluctuations superimposed on the average cross section are probably less well established but should have little

Black Neutron Detector, Mass

by a-Counting

Poenitz(ll) 1977 2.0

Low Energy Normalization

Wasson 1976 3.8

1.5

2.8

The QaeiitLon o^ the Need {^on. Futu/Le U-235

Me,asuA&ment!>

Nuclear data request lists usually state a 1% uncer- tainty requirement for U-235 (n,f) up to 14 MeV. In the light of these requests and the fact that the

present uncertainty is 2-3%, the need for continued

measurements is obvious. However, economics and the

considerations mentioned above suggest that we look

also at the limitations involved in the application of

the U-235 (n,f) cross section at its present uncertainty level. The major application is its use to

266
o

CM

1 ANL76. -^/- 3 PC

I—

UJ o

6 7

IJJ o

—<I> <^<I><;i>

EVaL. KONSHIN

II.

1 III Mill 1 1 Mill 1 1

ENDF/B-V

o

^<^^

1

1 III Mill 1 1 1 1 1 1 1 1

1

100 1. 00 EN/MEV 10. 0 20. 0

Fig. 6. Comparison of other evaluated cross sections with the present re- sult. The difference relative to the uncertainty Is shown.

convert measured ratios to cross sections. Such

ratios, their present uncertainty and Importance are

Ratio Importance

to U-235(n,f) Uncertainty for k

ef f Pu-239(n,f) V 3% High

U-238(n,f) < 2% Medium

U-238(n,Y) 5% Medium

U-233(n,f) - 3% High

Th-232(n,f) Medium

Th-232(n,Y) Medium

Pu-2A0(n,f) Low

Pu-241(n,f) Low

Others Low

This indicates that for U-238(n,f) and Pu-240, Pu-241

(n,f) the knowledge of U-235 (n, f ) is sufficiently

precise; and for other values, the uncertainty of the

ratios is probably an equal or dominant factor. The

workshop at the '76 ANL meeting considered the question

whether absolute measurements should be made for

Pu-239 (n,f) directly, rather than measuring U-235(n,f)

and the ratio Pu9/U5. No conclusion was reached at

that meeting, but considerations mentioned above suggest that absolute measurements of Pu-239 (n,f) are more profitable for the near future than further

absolute U-235 (n,f) measurements.

Re.-e.valaatLon Oj^ U-235{n,£]

A re-evaluation of U-235 (n,f) for ENDF/B-V is recommended. This is based not so much on any new measurement but on the fact that ENDF/B-V does not represent

the existing data base. The true cross section is un- known and the presently suggested version of ENDF/B-V

may actually be closer to the true values than the

weighted average of the data is. However, such assumption is purely speculation.

ACKNOWLEDGEMENT

This work was supported by the U.S. search and Development Administration.

Energy Re267
References

1. Proc. of NEANDC/NEACRP Specialists Meeting on

Fast Neutron Fission Cross Sections, Argonne

National Laboratory, ANL-76-90 and Supplement,

(1976).

2. 0. A. Wasson, p. 183 of Ref.l, (1976).

3. A. J. Deruytter and C. Wagemans, J. Nucl. Energy

25, 263 (1971).

A. B. Leugers et al., p. 246 of Ref.l, (1976).

5. I. Szabo and J. P. Marquette, p. 208 of Ref.l,

(1976) ; also see other references cited therein.

6. D. M. Barton et al., p. 173 of Ref .1 , (1976) , Nucl.

Sci. Eng. 60, 369 (1976).

7. M. C. Davis et al. , p. 225 of Ref.l, (1976).

8. M. Cance and G. Grenier, p. 237 of Ref.l, (1976).

9. H. T. Heaton II et al. , p. 333 of Ref.l, (1976).

10. K. D. Zhuravlev et al. , Proc. of Int. Conf . on Interactions of Neutrons with Nuclei, U.Lowell,

ERDA, CONF-760715 (1976).

11. W. P. Poenitz, to be submitted to Nucl . Sci. Eng.

,

(1977) . 12. J. B. Czirr and G. S. Sidhu, Nucl. Sci. Eng. 57, 18

(1975), Nucl. Sci. Eng.58, 371 (1975).

13. W. P. Poenitz, Nucl. Sci. Eng. 52, 370, (1974).

14. D. B. Gayther, Proc. of Conf. on Nuclear Cross

Sections and Technology, Washington, NBS Spec.

Pub. 425, (1975).

15. R. Gwin et al. , Nucl. Sci. Eng. 59, 79, (1976).

16. P. H. White, Nucl. Energy A/B 19 , 325, (1965).

17. W. P. Poenitz and P. Guenther, p. 154 of Ref.l,

(1976).

18. F. Kaeppeler, Proc. of Panel on Neutron Standard

Reference Data, Vienna, IAEA, p. 213 (1974).

App&ndix J: Eva£aate.d VcUa

The evaluation of a best value cross section should

include all absolutely measured cross sections and all

ratios in an "consistent data set" evaluation procedure (Ref. 19). It also should extend to average cross sections at lower energies. Such evaluation is in

progress and will be published elsewhere. The evaluation for the present purpose was restricted to recent

data, to energies above 100 keV and to U-235(n,f),

thus the validity of the values given below is restricted as well.

E/MeV o/b E/MeV a/b E/MeV a/b

0.100 1.587 0.900 1. 178 5.800 1.050

0.125 1.510 0.940 1. 215 5.900 1.073

0.150 1.454 0.960 1. 230 6.000 1.096

0.175 1.412 1.000 1. 226 6.200 1.180

0.200 1.346 1.050 1. 209 6.400 1.255

0.225 1.326 1.100 1. 210 6.800 1.484

0.250 1.297 1.200 1. 213 7.000 1.550

0.275 1.274 1.400 1. 221 7.400 1.671

0.300 1.256 1.500 1. 233 7.800 1.742

0.350 1.225 1.700 1. 254 8.000 1.758

0.400 1.200 1.900 1. 271 8.500 1.758

0.450 1.178 2.000 1. 274 9.000 1.744

0.500 1.156 2.250 1. 263 10.000 1.734

0.550 1.149 2.500 1. 242 12.000 1.741

0.600 1.143 2.750 1. 220 13.000 1.892

0.650 1.135 3.000 1. 206 14.000 2.054

0.700 1.130 3.500 1. 160 15.000 2.111

0.750 1.128 4.000 1. 147 16.000 2.117

0.800 1.129 4.500 1. 110 17.000 2.051

0.820 1.134 5.000 1. 079 18.000 2.004

0.850 1.147 5.500 1. 039 20.000 1.967

kppuicLix 11 •• CnJXznA-a. jon. VIU-lmltajtiovii>

Establishing criteria for the existence or nonexistence

of problems of any cross section can be obtained by

simulation techniques (as for example used in testing

19. W. P. Poenitz, Proc. of Symposium on Neutron

Standards and Flux Normalization, Argonne National

Laboratory, AEC Symp. Series 23, p. 331, (1970).

20. V. A. Konshin, Review and Evaluation of the U-235

Fission Cross Section, Intern. Nucl. Data Committee,

INDC(CCP)-94/U, (1976).

21. M. R. Bhat, p. 307 of Ref.l, (1976).

22. R. K. Smith et al.. Bull. Am. Phys. Soc. 2, 196

(K4), 5704, revised by G.E. Hansen et al. (1969).

23. I.M. Kuks et al., Conf. on Neutron Physics, Kiev,

Vol. 4, 18, (1973).

24. W. M. Adamov et al., Conf. on Neutron Physics,

Kiev, Vol. 4, 21, (1973).

25. H. T. Heaton II, p. 333 of Ref.l, (1976).

26. J. B. Czirr and G. S. Sidhu, Nucl. Sci. Eng. 60,

383 (1976).

27. B. R. Leonard and 0. A. Wasson^ p. 452 of Ref.l,

(1976) . 28. R. F. Taschek, Sec. Int. Conf. on Nuclear Data for

Reactors, Helsinki, IAEA Vol.11, Panel Discussions,

p. 929, (1970).

29. W. D. Allen and A.T.G. Ferguson, Proc. Phys. Soc.

A70, 573 (1957).

30. R. L. Henkel, LA-2122 (1957), B.C. Diven, Phys.

Rev. 105, 1350, (1957).

31. C. D. Bowman et al. , p. 270 of Ref.l, (1976).

32. C. D. Bowman, Proc. of Symp. on Neutron Standards

and Flux Normalization, Argonne National Laboratory, p. 246, AFC Symp. Series 23, (1970).

33. G. R. Hale, priv. communication to G.Lamaze (1976).

34. G. R. Hale and L. Stewart, private communication

(1977) .

and comparing economic or technical models). This approach is presently under investigation and the results

will be published elsewhere. The two criteria stated

in the text are crude approximations based on the more

commonly expected behavior of experimental variables.

kppmdix III- Rdcmt ANL BNV ReAulti

The recent experimental values shown in Fig. 2 are listed below. Until their final publication these values must be considered preliminary, though no change is

anticipated at the present time.

E/MeV a/b Aa/mb E/MeV a/b Aa/mb E/MeV o/b La/mb

0.215 1.342 50 0.929 1.158 35 3.756 1.096 25

0.225 1.311 38 0.969 1.214 25 3.955 1.095 27

0.235 1.309 34 0.988 1.203 25 4.153 1.069 29

0.245 1.297 29 1.088 1.201 38 4.351 1.068 30

0.255 1.294 27 1.138 1.209 25 4.449 1.065 31

0.265 1.283 27 1.185 1.209 29 4.396 1.083 23

0.275 1.225 26 1.236 1.199 29 4.618 1.075. 22 -

0.285 1.246 27 1.288 1.221 26 5.179 1.028" 22

'

0.295 1.257 27 1.336 1.217 26 5.536 1.026 24

0.305 1.255 26 1.388 1.190 25 5.618 1.033 28

0.193 1.354 37 1.433 1.225 26 5.714 1.049 27

0.213 1.342 30 1.486 1.208 26 5.809 1.032 27

0.233 1.304 26 1.578 1.247 26 5.898 1.073 30

0.253 1.297 26 1.778 1.245 29 6.006 1.087 33

0.271 1.261 23 1.978 1.262 28 6.107 1.154 32

0.293 1.254 28 2.178 1.262 28 6.185 1.161 31

0.684 1.138 23 2.222 1.286 28 6.384 1.260 36

0.735 1.144 23 2.472 1.224 28 6.579 1.387 30

0.793 1.119 22 2.718 1.201 26 6.803 1.515 35

0.814 1.126 22 2.968 1.150 26 7.025 1.541 37

0.835 1.151 28 3.162 1.134 26 7.478 1.713 54

0.850 1.135 23 3.262 1.137 25 7.875 1.795 59

0.871 1.151 24 3.360 1.134 26 8.275 1.793 62

0.893 1.155 24 3.463 1.133 26

0.909 1.154 24 3.658 1.093 27

268
PROPAGATION OF UNCERTAINTIES IN FISSION CROSS SECTION STANDARDS IN THE INTERPRETATION

AND UTILIZATION OF CRITICAL BENCHMARK MEASUREMENTS^

C. R. Weisbin and R. W. Peelle

Oak Ridge National Laboratory

Oak Ridge, Tennessee 37830

USA

This work explores the constraints imposed on the 235U(n,f) standard (proposed ENDF/B version V)

by information deduced from clean integral measurements and demonstrates how uncertainties in

fission cross section standards propagate in an uncertainty analysis and interpretation of those

experiments. The question of what a significant improvement in the accuracy of the ^^^U(n,f)

standard would accomplish is addressed in the limited context of analyses of GODIVA and JEZEBEL

measurements

.

The CSEWG integral benchmark results and uncertainties were updated in accordance with more recent information. Sensitivity coefficients were developed and used to estimate calculated re- sults which should be obtained using the subsequent release of ^^^[]{n,f), ^^^Uin,f), and 239pu(n,f)

at version V status. Covariance files were evaluated and processed for all important cross sec- tions with the sole exception being inelastic scattering for all levels and the continuum.

Uncertainties due to the '^^^[J{r\,f) standard were estimated to comprise more than half of the

calculated uncertainty for criticality and ^28g^/25o\ spectral index in JEZEBEL as well as GODIVA; though the JEZEBEL assembly contained no ^^^u. We are not able at this time, to predict criticality or {^^^ec/^^af^ q to anywhere near the accuracy obtained by direct measurements,

and therefore the integral results are significant to our analysis capability. Inclusion of

integral information from GODIVA and JEZEBEL in an adjustment procedure was effective in

reconciling all parameters other than^28gf/25o^^^ p^gasurement in JEZEBEL for which current calculation and measurement are in disagreement. The adjustment procedure made changes of less

than one standard deviation in the cross sections for ^^^U[n,f), ^^^U{n,y), 238u(n,f), ^^^[}iT\,y),

and 239pu(n,f) including an increase of '^^1.5% for the ^^^[i{r\,f) cross section above 1.3 MeV. This specific adjustment result could change with inclusion of inelastic covariance files and

must be viewed cautiously at this time.

(criticals, cross section, ENDF/B, fission, standards, uncertainties)

I . Introduction

It is well known that if uncertainties in fast

reactor performance parameters were to be developed

solely from consideration of uncertainties in differential nuclear data, and even assuming all methods

approximations to be negligible, the resulting un- certainties in predicted fast reactor performance

would still be significantly larger than those re- quired by designers.^ Thus, in addition to the

customary calculation/experiment comparison, data

adjustment schemes have been developed^ which can incorporate information from both evaluated differential data and relatively clean geometry, fast

integral data. The latter generally lacks energy

resolution but provides bounds within which the

reactor performance predictions should lie. In general, the cross section adjustment schemes minimize a quadratic function of the weighted difference

between the measured and computed values of the

performance parameters and between the reference

and adjusted cross section values. The fitting

procedure provides an estimate of the accuracy of the

adjusted cross sections and of the reactor properties

calculated using them.^"** Although the standard

deviations of the adjusted cross sections may not be

significantly smaller than those assumed for the un- adjusted cross sections, performance parameter predictions can be significantly improved because of the

large effect the adjustment has in altering the offdiagonal elements of the cross section covariance

matrix. 2"*+ it is important to note, however, that

the adjusted covariance matrices depend upon the

initial estimated accuracy for both the differential

and integral measurements.

Adjustment of the nuclear data base, and associated uncertainty analysis, for improvement of performance parameter prediction has sometimes been

received with considerable skepticism. In part,

this was because adjustments have been made without

detailed consideration of the differential data co- variance i nformation. 3 This was not at all a simple

error of omission. The evaluation of differential

cross section covariance matrices is a formidable

task^ involving correlations of evaluated cross sec- tions with cross sections in other energy ranges and

reaction types. Evaluated data which is derived (e.g.,

by deducing the elastic cross section from the total

and non-elastic data) or evaluated through measurements relative to standard cross sections introduce

additional complexity. Although the level of effort

required is considerable, the rational assessment of

uncertainty information for ENDF/B files was considered

of high urgency^ to improve upon existing analyses

by permitting systematic sensitivity investigations

to propagate uncertainties in a credible fashion.

In a recent paper, we presented the first results

of our uncertainty analysis for fast reactor benchmarks.^ The FORSS system^ was employed to compute

sensitivities^ in 126 energy groups using transport

theory. Multi group covariance files were developed

for 23§u(n_5_f), 238u(n,Y), 2 39pu(n,f), 2 39pu(n,Y),

and 2 39p,j(^) using evaluations generated at ORNL^ and

formats and procedures established for the ENDF/B

Research performed at Oak Ridge National Laboratory for Union Carbide Corp. under contract with U. S. Energy

.Research and Development Administration.

Notice: By acceptance of this article, the publisher or recipient acknowledges the U. S. Government's right to retain a non-exclusive, royalty-free license in and to any copyright covering the article.

269
system^" for the definition and processing of evaluated and correlated energy dependent uncertainty

information. Using the measurements in ZPR-6/7 for

k and central 238u capture/^ 39pu fission with assigned

uncorrelated one standard deviation (lo) uncertainties

of 1 and 2 percent resulted in (la) predictions for the same parameters of 0.8% and 1.8% when the integral data

was included in a cross section adjustment procedure

which made changes of less than one standard deviation

to the basic multigroup file. This technique was subsequently applied^^ to a realistic two-dimensional

model of an LMFBR, including several reaction rate

ratios measured in ZPR-6/7 and ZPR-3/56B. The (lo)

uncertainties in predicted multiplication factor

and breeding ratio of a 1200 MWe LMFBR were reduced

from 3% and 7.2% (differential nuclear data only) to 0.9% and 3.2% after inclusion of the information from

CSEWG benchmark assemblies. In all cases, the cross section covariance files referred to the average cross sections in an infinitely dilute situation. The only

cross section reaction correlation considered was for

239pu(n,f) and ^sgp^^n^^) developed from measurements

of their ratio. Later work^^ -jp ^|^g analysis of the

TRX-2 thermal lattice required covariance file evaluations for the four low energy resonances of ^^^U (i.e.,

covariance of r^, and Fy), and the thermal region. However, even with this expanding data base, none of the

studies above gave detailed consideration to the effects

of correlated uncertainties resulting from measurements

relative to standards. (It should be noted that a more comprehensive covariance file is expected with the 1978

release of ENDF/B-V.)

A file of cross section uncertainty information

for use in reactor performance uncertainty analysis

should take into account the propagated effects of

uncertainties in the standard cross sections used.^^

In particular, this applies to the ^^^U(n,f) standard above a few hundred keV; the 239p^J(n,f) and

238u(n,f) cross section measurements, among others,

are often made relative to 2^^U(n,f). The purpose of

this paper is to explore the constraints imposed on the 235u(n^f) standard by information deduced from

clean integral measurements and to demonstrate how

uncertainties in fission cross section standards propagate in an uncertainty analysis and interpretation

of those experiments. The question of what a significant

improvement in the accuracy of the ^^^U{<t],f) standard

would accomplish is addressed in the limited context of

analyses of GODIVA and JEZEBEL measurements.

II . Data Testing Assemblies Description

The CSEWG Benchmark Specifications^'* contain a recommended calculational model as well as pertinent

experimental results for the GODIVA and JEZEBEL

assemblies. ^5 The GODIVA model is a bare sphere of

enriched uranium metal, having a core radius of 8.74 cm

and atomic densities of .045/. 002498/. 000392 atoms/barncm for 235u/238u/23'ty^ respectively. Similarly, the

calculational model for JEZEBEL is a bare sphere of

Plutonium metal, having a core radius of 6.385 cm and

atomic densities of .03705/ .001751/ .00011 7 atoms/barn-cm

for 239p,j/2itOpu/2'tipL,^ respectively.

The experimental results for the performance parameters studied are given in Table I; the bracketed quantities refer to reaction rate ratios measured at the

center of the sphere and are effective microscopic

quantities, densities divided out.

It should be noted that the experimental values

for the ^28(j^/2 5(j^^^ ratios have been modified from the

CSEWG recommendations (0.205 for JEZEBEL and 0.156 for

GODIVA) in accordance with the recent recommendations

Table I. Accurate Measurejncnts and Associated Uncertainties

Have Been Reported for GODIVA and JEZEBEL

GODIVA JEZEBEL

K 1 .UU+U . UU

J

k 1 .00+0.003

0.16+0.005^ 0.21+0.008^

1.42+0.071'' 1.49+0.075''

0.47+0.02

^The experimental values for the Q^^a^/^^a^ ^ ratios have been modified from the CSEWG recommendations (0.205 for JEZEBEL and 0.156 for GODIVA) in accordance with recent recommendations of Hirons."

''The uncertainties in the (^^a^/^^a^^ ratios have been increased'^

from -v. 2% to 1-5% (lo) due to revised uncertainty estimates -associated with the deduction of the number of fissions through the 6- counting of ^^Mo.

""Macroscopic central reaction rates, denoted by < have been divided by respective densities; i.e., all ratios are "microscopic" reaction rate ratios, is the capture (n,Y) cross section; o^^ is

the fission (n,f) cross section.

have been increased from '\'2% to '^-5% (lo) due to revised

uncertainty estimates^^ associated with the deduction

of the number of fissions through B-counting of ^^Mo.

III. Analyses of GODIVA and JEZEBEL: Generation

of Sensitivity Coefficients

The cross section library employed in these calculations was a 126 group processed MINX/SPHINX library^^

generated from ENDF/B-IV data at ORNL. Group constants were developed simultaneously in CCCC^o AMPX,2i and

MATXS22 formats to allow for self shielding, preparation of user libraries for transport codes and creation

of a data base for the FORSS sensitivity profile

modules. The ANISN code^^ was used to generate regular

and generalized fluxes and adjoints in the S^g, P3,

40 spatial mesh interval description provided in the

CSEWG specifications^** and to compute the desired

responses. The volume integrated flux spectrum for

GODIVA is illustrated in Fig. 1. Previously determined

correction factors^"* to account for higher order transport approximations were applied and the results are

03-10-77

g 10

id s

10

10'

VOLUME INTEGRflTEO FOFIWBRD FLUXES FOR GODIVR n rrn m—r-n—rrn—rrn—nn—

r

FLUX 6.7036E 00 126 GROUPS

/

/

I 'III III III III III I

ENERGY IHEV)To^'^ 5 10° 2 s 10' 2 =10=

Fig. 1. The Flux Spectrum in GODIVA Is Applicable for Testing ^^^U{n,f) Cross Sections in the

Fission Source Energy Range.

of Hirons..^'^ The uncertainties in the ^9of/ 2 5,Of

270
listed in Table II along with results of other laboratories participating in the CSEW6 data testing effort^**

and the experimental uncertainties from Table I. It is important to note that the revised recommendations for

the /28af/25a Aexperimental results^^ change previous

GODIVA overpreaiction into reasonable agreement but

makes the JEZEBEL discrepancy even worse. The plutonium

fission ratio is underpredi cted in the JEZEBEL assembly

consistent with underprediction of critical ity there.

This somewhat complicated series of results is not the

final picture. With the coming issue of ENDF/B-V,

several of the principal cross sections are due for

revision. We assess this effect in a subsequent section

by making use of the sensitivity profile§~-d^eloped

from ENDF/B-IV as discussed below.

Table II. Calculation/Experiment Data Testing with ENDF/B-IV

Indicates Discrepancies Particularly for JEZEBEL

Table IV. Relative Sensitivity of JEZEBEL Performance Parameters

to Various Cross Section Reaction Types

BNL LASL ORNL

Experimental Uncertainties

{%: lo)

JEZEBEL

^eff 0.992 0.996 0.992 0.3

0.908 0.922 0.905 3.8

0.934 0.936 0.933 5.0

GODIVA

'eff 1.005 1 .007 1.003 0.3

1.031 1 .035 1.031 3.1

0.971 0.970 0.972 5.0

0.937 0.954 4.2

C/E is the ratio of calculation to experiment for the quantity in brackets. The {C/E)'s for (^°a^/^^o^)^ are relative to the experimental values in Table I which updates the CSEWG recommendations changing

previous GODIVA overprediction into reasonable agreement but making the JEZEBEL discrepancy even worse.

Energy dependent sensitivity profiles were calculated with the JULIET module'' of the FORSS system. Spaceenergy-and angle integrated sensitivities for each

of the performance parameters in GODIVA and JEZEBEL are presented in Tables III and IV for many of the cross sections of interest. Comprehensive libraries^ of energy

dependent coefficients in a computer retrievable format^^'S have been documented and released for distribution by RSIC and NNCSC. As examples. Fig. 2 illustrates

Table III. Relative Sensitivity of GODIVA Performance Parameters

k C»-f/""^c

Reaction Relative Sensitivity Reaction Relative Sensitivity

. Relative

Reaction Sensitivity

0.967 '\ 1.0 "•^c^ 1.011

""f 0.729 -1.0 25a^ -1.0

'\,n 0.082 ^\,n'.d -0.192 •'"^ „| H -0-''22

n ,n ,a

0.031

n ,n ,c -0.151

n ,n ,c

n,n ,d 0.027 "°f 0.101 „ -0.011n ,n

0.023 -0.078

n,n ,c 0.012 0.054

"°c -0.008

n,n' ,d

n,n ,c

0.012

-0.010

-0.007

This is percent change in response per percent change in cross section uniformly over all energy.

is fission [(n,f) MT=18]

is capture [(n.y) MT=102]

"n n' X total discrete inelastic when x=d (MT=4); discrete inelastic to

level n when x=n (MT=51-90); total continuum cross section when x=c (MT=91).

The fission spectrum temperature is not included in this list; otherwise ranking of reactions is in order of importance.

the sensitivity profile of k for GODIVA with respect to the ^^^U fission cross section. Figure 3 presents the

sensitivity profile for the ^ssy capture/fission ratio

with respect to the ^ssy fission cross section. The

dashed line is positive, i.e., an increase in 2^^LI(n,f) at high energies effectively competes with

the ^'^^U{r\,f) cross section, thus increasing the

(2^a^/28a.^ ratio [the solid line is negative; it

pBaks near ^^^U(n,f) threshold and at all energies

below, competes only with the 238^ capture rate].

Figure 4 presents the sensitivity of {^^of/^^aA

in GODIVA with respect to the ^ssu fission cross section. Increases in 238Li(n^f) clearly have a positive

direct effect on this performance parameter since the cross section appears explicitly in the performance

parameter definition. The net contribution to the flux

O2~16-77C00IVR K 0-235 FIS flREfl= 6.5985E-01

to Various Cross Section Reaction Types >-

k LD

CC

s

Reaction'' RelativeSensiti vi ty Reaction'' Relative

Sensi ti vi ty

^ Relative

Reaction Sensitivity Reaction''

Relative

Sensi ti vi ty

a:

i:1—

2s;;' . 0.983

0.660 "»f

0.998

-0.819

"of 1.00 "of -0.979

28(7 0.998

-0.,997

UJ

1—

io->

"°„,„ 0.113

n n' .d -0.269 '*°rn\d

2 5o n.n .d 0.385 s

n,n' ,d 0.062 2 5-n ",c

-0.164 ""n.n'.c -"-"'^ 2S/T"f -0,266

""c -0.037 25(5 n -0.087 ^ -0.011 '\.n >c 0.230 CC

UJ

n.n' ,c 0.014 2S~ 0.068 "oj, to. 0059 "°n,n 0,126 Q. 10-2

2B~ 0.0097 25o

C 0.034 -0.096 >—1— 5

2S_

n.n' ,2 0.0083 2 8(3 n n' .d -0,026 25„

c -0.054 >

0.0083 ""n n',Z -0.009 "°n,„ .d 0,037 1—

0.0073

n n" .c -0.009 "°n,n .2 0,016 CO 10-3

0.0068

0.0067

28(1 n -0.006

n.n

26-

n ,n

n,n

.c

,3

0,013

0,009

0,007

REL.

SEN 5

This is percent change In response per percent change 1n cross section unifonnly over all energy.

Is fission [(n,f) m=183

is capture [(n.r) MT=102]

°n n* X tota} discrete Inelastic when x=d (HT=4); discrete Inelastic to level n when x^n (HT=51-90);

total continuum cross section when x=c {MT=91).

T>.c fission spectrun tenperature is not included in this list; otherwise ranking of reactions Is in order

of importance.

10 5 10^ 5 10-' ENERGY (I-IEV)

10 '

Fig. 2. Relative Sensitivity Per Unit Lethargy of

kgff in GODIVA with Respect to the 235u(n,f) Cross

Section.

271
02-22-77 COOlvn U-23a / U-J3a FIS U-235 FI5 I fiBEfl- -2.6S70E-01

10 ° y-O 5 CCCC

I—UJ

10-' I—— s

CC

LlI a. 10-2

cn 10-3

CO ^

UJ

10-'

1 I I ml _] I I I I nil. 10^ s 10-' ENERGY (MEV)

I I ml

5 10 ° 10

Fig. 3.

Cross Section

Relative Sensitivity per Unit Lethargy of

in GODIVA with Respect to the ^35y(n^f')

the present study, we've taken McKnight's five group

estimated cross section changes^^ for 235u(n^f) ^^^^ 233pu(n,f) and have projected the impact on all of

the GODIVA and JEZEBEL performance parameters con- sidered in this work. The cross section changes

proposed were +1.44, -0.24, -2.57, -3.32, -2.13

percent change in ^^of relative to ENDF/B-IV for

3.679-10.000, 1.353-3.679, 0.498-1.353, 0.183-0.498,

0.067-0.183 MeV energy regions. Similarly, the per

cent changes for same energy ranges were

-1.07, +0.83, +1.02, -0.30, and -2.13, respectively.

In addition to McKnight's tabulations^ we have included the proposed changes associated with ^asy^p^f)

for ENDF/B-V. For the three highest energy groups, we estimate -2.5, -3.5, and +12% respectively (highest energy group first).

As a result of lowering the 2^^U(n,f), the

238U(n,f), and raising (slightly) the 239pu(n,f)

cross section, the results on the whole are brought

into much better agreement. Critical ity, for both

systems, is underpredicted as is the {^^of/^^o^^

ratio in JEZEBEL. Other performance parameters are close to or within one standard deviation of the

experimental value. Table V summarizes the projected calculated results (ENDF/B-V ^asy, zsey, and

239pu fission) and measurements for GODIVA and JEZEBEL.

Table V. ENDF/B-IV, Projected ENDF/B-V and Measured

GODIVA and JEZEBEL Performance

02-2J-T! caoivR U-?36 riS / U-?35 FIS U-?38 FIS I'O 9.9783E-01

10 "

[HflRGI

5

'i_

t

1

UJ

_J 10-'

1

1

UNIT 5

J*

r

.1

PER 10-2

r

IVITY

S

SENSIT 10-3

s

1

_jUJ j

1 Mini. 1 1 1 1 ( Mil r t t 1 1 Mil 1 1 Jl 1 1 Mil t 1 1 1 m

10 S 10-3 S 10-2 5 iQ-' 5 10 °

5 10 ENERGY (MEV)

Fig. 4. Relative Sensitivity Per Unit Lethargy of

the C®°f/^^'^f)c in GODIVA with Respect to the ^ssy

Fission Cross Section.

"ENDF/B-V" Experimental

(Updating 2 38u, 235u, Uncertainty

ENDF/B-IV ^3^Pu Fission) {%: lo)

OEZEBEL

"eff

<^^°f/""f>C/E

<"<'f/^^°f)c/E

"eff

<''<'f/"<'f>C/E

<^^°f/"<'f>C/E

<"V"°f>C/E

0.9920

0.905

0.933

1.0033

1.031

0.972

0.954

GODIVA

0.9944

0.892

0.949

0.9937

1.010

0.990

0.991

0.3

3.8

5.0

0.3

3.1

5.0

4.2

effect is essentially negligible with additional fission neutrons increasing both relative ^^^y and ^^^U

fission rates in roughly the same proportion.

IV. Changes to GODIVA and JEZEBEL Calculated Performance

Associated with ENDF/B Re-Evaluation to Version V

At the May, 1976, ENDF/B-V Task Force meeting, evaluators made "first cut" estimations for what changes

might be made in the cross sections for the principal

isotopes [235y(n^f)^ 238u(n^Y)^ 239p^(n^f)j for Version

V relative to ENDF/B Version IV. The Data Testing Subcommittee then made projections based upon available

sensitivity coefficients of the possible impact on fast

reactor criticals analysis (e.g., ref. 7, p. 71). Subsequently, McKnight and Poenitz^^ re-examined this

question with alternative evaluations discussed at the

1976 ANL Fast Neutron Fission Cross Section Meeting. For

272

V. Evaluated "Pointwise" Covariance Files

Covariance files have been evaluated for ^ssy^^^f)^

238u(n,Y), 235y(^), 238y(n^f)/235u(n^f) ^atio, 238u(n,

nMst), 2 38y(n,Y), 238u(v), 2 39pu (p , f

)/2 35u( p , f ) ratio

239Pu(n,Y)/239pu(n,f) ratio, 239pu(^), 2'*opu(n,Y),

2'tOPu(v), 2tipu(n,f), and 2'tipu(n,Y). The evaluations were based primarily on "external" methods of analysis

which examine the scatter among existing data sets.

These sets are assumed to represent fairly the statistical ensemble of hypothetical sets of measurements

which could have been obtained in the experiments which

form our present data base. The "pointwise" covariance

files were represented on convenient energy grids

believed to be adequately fine to reproduce the broad

range behavior important for estimation of uncertainties

in integral quantities.

The source of data for new uncertainty files evaluated for 28a^/25o^ and '^af/sSof was the recent com- pilation prepared by W. Poenitz.26 in the method used,

weights were assigned to each experiment which were

estimated to reflect the reciprocal variance of a typical point from the data set. Since uncertainties
in most sets are a function of energy, an overall judgement was used. The evaluated ratio for ^^o^/^^o^ was

taken to be the proposed evaluated fission ratio for

ENDF/B-V;27 -j.e., the ratio was used which, when multiplied by the ^ssy fission cross section for ENDF/B-V

gives the proposed ^^^U fission cross section for

ENDF/B-V. For the '^'^a-^/'^^o-f fission ratio, the

ENDF/B-V evaluated ratio was not in hand early enough

so the evaluated ratio had to be taken from ENDF/B-IV.

Only in these two recent fission ratio evaluations was

a distinction made between the ensemble of hypothetical

measurements and the ensemble of evaluations based

upon these measurements. Accounting for this apparently

straightforward difference introduced cal culational com- plexities because of the varied pattern of measurements

made by individual authors and because of unequal

weights assigned. In practice, these files for the

ratios were group averaged and subsequently combined

with multigroup covariance files based upon existing

235u clata^2 (properly weighted).

In general, the remaining files were obtained as indicated in the SUR report, ^ essentially applying

the definition of the covariance matrix directly to com- piled sets of experimental cross sections. Additional

information relating to the pointwise covariance files

can be found in ref. 7. In cases for which only a few

data sets exist or could be compiled, the ensemble variances were statistically poorly determined by the small

sample; however, variance fluctuations over small

energy regions may be unimportant after averaging over

the assembly spectrum. It is also important to recognize that such "external" covariance evaluation

methods do not include any systematic bias; for

example, the possibility that the ^aspu half life may

be uncertain to 2% could systematically affect all

"absolute" ratio measurements requiring exact foil

weights. Finally, in all cases, the uncertainty files

refer to the infinitely dilute average cross section.

VI . Multigroup Covariance Files

The differential covariance files developed in

the preceding sections were put into ENDF/B- IV format^° and processed with the PUFF covariance file

processing code.^s The complete covariance matrix

used is illustrated in Fig. 5.

The ^38y and 239pu fission ratios relative to ^^sy

combined with the 239pu capture/fission measurements

introduce several off-diagonal components that are directly related to the 235u(n,f) standard. For

example, if the 239p,j fission cross section is measured

relative to 235u (tSg^ = r ^^af, r being the 239pu/235u

fission ratio) and the 239py capture cross section is

measured relative to the 239pL, fission data (i.e.,

^ac = a '^f; a is the Pu capture/fission ratio).

Then it follows that:

(1) The relative covariance matrix for 239p,j fission

is the sum of the relative covariance matrices

of r and '^^of.

(2) The relative covariance matrix of 239pu fission

and capture is the sum of the relative covariance

matrices of r and ^s^^.

(3) The relative covariance matrix of 239pu fission

and 235u fission is the relative covariance

matrix for ^^Of.

(4) The relative covariance matrix of ^^gp^ capture

is the sum of the relative covariance matrices

for a, r, and ^^OfOf Si V

>t ^ «

,

.

(J

as V

11_V '\ V \

o o o o

V

25 _ V V V V

o o o o

V

V

It V V V V V

o o o o

o o o o

11 V V V V V

V

'JO V V V V V

*| V

v

C V

Fig. 5. The 235U(n,f) Covariance Matrix Impacts Several Sub-Matrices of the Complete Multigroup Uncertainty Matrix.

1. Each box contains a 6x6 energy group covariance matrix. The blank boxes con- tained null elements.

2. ^indicates evaluation used in this study.

3. ^indicates representation includes a processed 23Su component.

(5) The relative covariance matrix of 239pu capture

and 235y fission is the relative covariance

matrix for ^^Of.

Similarly arguments can be made to deduce relative

covariance matrices between ^ssy fission and ^sby

fission and between 238u fission and 239pu capture

cross sections. Additional off-diagonal components

are introduced due to the uncertainty in ^sa^f ^ affecting a V evaluation. Perey has shown^^that when

the covariance matrix for a cross section is to be

determined from the covariance matrices of ratios

and other cross sections, it is necessary in the

reduction to group form to weight the error files

with the cross section whose covariance is being determined. This was implemented properly in this work

for the matrices which required only one reaction

type for weighting. However, off-diagonal group

matrices [e .g . , ( 239pu(n,Y) , 239pu(n,f)) ] for which

the 235U covariance file should have been weighted

bilinearly [e.g., by the 239pu(n,f) cross section at one energy and by the 239pu(n,Y) cross section at

the other] were approximated by weighting with the

235u cross section itself.

It is important to note that each block of Fig. 5

is, in fact, a 6x6 energy group covariance matrix.

Tables VI-XI present several of the important co- variance files in the form of relative standard

Table VI. Relative Standard Deviation and Correlation Matrices for ^ssy Fission

Energy Ranqe (eV)

% Ret. Std. Dev. Group 1 2 3 4 5 6

2.000E+07-3.679E+06 3.1 1 1000 3.679E+06-1.353E+06 2.3 2 545 1000 1.353E+06-4.979E+05 2.7 3 184 516 1000 4.979E+05-1.832E+05 2.8 4 49 243 749 1000 1.832E+05-4.087E+04 2.7 5 9 72 311 515 1000 4.087E+04-1.000E-05 3.3 6 0 0 0 282 609 1000

273
Table VII. Relative Standard Deviation and Correlation Matrices for ^"U Capture

I Rel

.

Std. Dev. Grou \ 3 5 6

2.000E+07-3.677E+O6 61.8 1 1000

J.D/yt+UO- 1 .J3JL+UD 60 .

0

2 729 1 000

1.353E<-06-4.979E+05 39.7 3 592 762 1000 4.979EtO5-1.832E+05 24.1 4 358 417 727 1000 1.832Et05-4.087E+04 10.9 5 148 193 291 501 1000 4.087Et04-l .OOOE-05 8.4 6 96 99 150 247 663 1000

laDie Vlli. Keiative O Lu J] Ua I u Dev1 at Ion and Correlation Matrices for ^'^[) Fission

% Rel. Energy Ranqe (eV) Std Dev 1 2 3 4 5 6

3.00OE+07-3.679E+06 3.1 1 1000

3.679Et06-1.353Et06 2.5 2 542 1000

1.353E+06-4.979E+05 2.6 3 272 620 1000

4.979E+05-1.832E+05 3.1 4 5: 241 352 1000

1.832Et05-4.087E+04 7.8 5 -2 32 40 474 1000

4.087E+04-1.000E-05 107.0 6 0 0 0 6 7 1000

Table IX. Relative Standard Deviation and Correlation Matrices for Capture

t Rel. Energy Ranoe (eV) Std. Dev. Group 1 2 3 4 5 6

2.000E+07-3.679E+06 51.8 1 1000

3.679E+06-1 .353E+06 17.8 2 921 1000

1.353Et06-4.979E+05 19.8 3 448 611 1000

4.979E+05-1 .832E+05 12.7 4 402 547 856 1000

1 .832E+05-4.087E+04 7.6 5 326 427 618 769 1000

4.0B7E+04-1 .OOOE-05 9.6 6 67 149 402 404 500 1000

Table X. Relative Standard Deviation and Correlation Matrices for 2"Pu(n,f)

Energy Range (eV)

% Rel. Std. Dev. Group 1 2 3 4 5 6

2. OOOE+07-3. 679+06 3.1 1 1000

3.679Et06-1.353E+06 2.4 2 543 1000

1 .353E+06-4.979E+05 2.7 3 184 510 1000

4.979E+05-1 .832E+05 2.8 4 50 242 748 1000

1.832E+05-4.087E+04 2.8 5 9 74 310 508 1000

4.087Et04-1.000E-05 3.4 6 0 0 0 271 569 1000

Table XI. Relative Standard Deviation and Correlation Matrices for "5Pu(7)

Energy Range (eV)

% Rel. Std. Dev. Group 1 2 3 4 5 6

2.000E+07-3.679E+06 1.3 1 1000 3.679E+06-1.353E+06 0.7 2 891 1000 1.353E+06-4.979E+05 0.5 3 -277 57 1000 4.979E+05-1.832E+05 0.8 4 -664 -458 793 1000 1.832E+05-4.087E+04 0.8 5 -664 -458 793 1000 1000 4.087E+04-1.000E-05 0.8 6 -664 -458 793 1000 1000 1000

deviations and correlation matrices (matrix elements

multiplied by 1000 for ease of reading) in the six group

structure selected for analysis. The GODIVA spectrum

(Fig. 1) was used as the weighting function. Note

that the energy dependent correlation matrix for a specific reaction is symmetric

VII . Uncertainty Analysis

The GODIVA and JEZEBEL sensitivity profiles calculated in Section III were collapsed to the six-group

structure used in multigroup covariance file generation

(Section VI) and folded' with the nuclear data co- variance files for each of the performance parameters

analyzed. The computation was then repeated assuming

that the ^^^U{r\,f) cross section was known exactly.

The results are indicated below in Table 12. It is

vitally important to note that the evaluated uncer- tainty files which form the basis of Table 12 included only those represented in Fig. 5. It is clear

from Tables III and IV that the uncertainty

Table 12. Increased Precision In ^^^\J{n,f) Could Significantly

Impace Our Interpretation of k and (^^^af/^^a^ Measurements

1n GODIVA and JEZEBEL

ENDF/B-V

" C3 Iculdtton" Uncertainty (%: lo)

All Nuclear Data

Uncertainty {»: lo) P^^Uin.f) Assumed

Known Exdctly]

JEZEBEL

eff

0.9944 1.5 . 0.5

\ "f °f/c 0. 188 1.3 0.3

\ f' f/c

1 .414 0.4 0.3

GODIVA

"eff

0.9937 1.7 1.0

<"°f/"''f>c 0.163 1.5 0.7

<^'<'f/""f)c

1.401 0.4 0.3

<^'"c/''"f>c

0.466 14.3 14.0

in discrete and continuum inelastic scattering could

add significantly to the projected uncertainties in

Table 12. This is particularly true for the threshold reaction rate ratios, (^^Of/^^<^f)c ^""^ /^^0(~/28gA and is also important for the uncertainties in kgff. In addition to the lack of covariance

file data for inelastic interactions in ^^^U and ^^^Pu,

covariance files were not constructed for the fission

spectrum shape. This might also have made a significant contribution. Clearly, the uncertainties listed

in Table 12 are probably lower estimates. However,

the numbers presented are themselves quite interesting.

The most startling result obtained is probably the

0.4% uncertainty for the {'^^of/^Sof^^ ratios, particularly in view of the '\^3% uncertainties for each of

the reactions (Table VI and Table X). However, there

is good reason for this low number. Recall that the

239pu(n,f) was evaluated through a ratio measurement

relative to 23^LI(n,f). Thus, the direct component of

the 235u(n^f) uncertainty vanishes to first order. The

large number of rather independent measurements tend to reduce the estimated uncertainty on the evaluated ratio

to 0.5%. In addition, combining uncorrelated data over the spectrum of the assembly reduces the uncer- tainty further.

For the(^^af/^^afy^ ratio, the situation is not

as clean. As shown in Table III, the indirect effect

of 235u(n,f) on the GODIVA {^^or/^^a^^ spectral index

is considerable. [The ^^^u{r],f) uncertainty would cancel if the sensitivity for ^'^^U{n,f) and 235u(n,f) were equal and of opposite sign.] In addition to the

direct effect, which would correspond to a sensitivity

of -1, there is an approximate +0.2 indirect effect

so that the sensitivity to 238u(n,f) [0.998] is not

equal in magnitude to the sensitivity to ^^^U{n,f)

[-0.819]. Thus, a component of the ^^^Uin,f) covariance remains. In the case of JEZEBEL where the

sensitivities to ^^^U{r\,f) and ^^^\J{n,f) are +1 and -1 respectively (and thus, the ^^^U contribution should

cancel ) there is a reasonably large sensitivity to 239pu(n,f) which in turn depends strongly on ^^^[}{n,f).

The resulting uncertainty in ^28^.^/250^^ -jp jezEBEL

is then comprised largely of the uncertainty in

238U/235U fission ratio (small) and the uncertainties

arising from 239pu from ^^^U{n,f). The nuclear data

uncertainties in kgff, approximately 1-2 percent,

depend on. many factors but ^^^U{r\,f) is important.

Finally (^^o^/^^a^ cannot be computed with high con- fidence due primarily to the large uncertainties

assigned to ^ssu capture at fission spectrum energies.

274
The third column of Table 12 indicates that eliminating all uncertainty in the ^^^UCnjf) cross section

could significantly improve our confidence in calculated

criticality and (^^of/^^o-^c • "^^^^ remark must be

qualified until such time as uncertainties in ^^^U and

239pu inelastic and in imprecision in the fission spectrum temperature have been included in the analysis.

Hence, the significance of increased precision in the

^^^J(n,f) cross section on our analysis capability for

GODIVA and JEZEBEL is limited by the lack of covariance

file components for other cross sections in the system

and, of course, on the reliability of our estimates

of uncertainties in cross sections evaluated for this

analysis

.

Inclusion of the measurements (and their uncer- tainties) in GODIVA and JEZEBEL, listed in Table I, in a data adjustment scheme^ which included simultaneously the seven integral parameters led to the

new set of calculated results (and uncertainties)

listed in Table 13. The integral measurements have

a pronounced effect on the resulting uncertainties

for keff and { °c/^^''f)c since in these cases the

assigned integral expenment uncertainties are con- siderably smaller than those projected to be due to nuclear data. It is interesting to note that the

optimal adjustment could do little to resolve the

Table 13. The Adjusted Data Set Contines Information from

Both the Differential and Integral Data

ENDF/B-V

"Calculation" Reported Measurement Adjusted Data Set Reported Measurement

Uncertainty {%: lo) on Calculated Result Using Adjusted Set

JEZEBEL

"eff

<"°f/"°f>C/E ("V"°f)c/E

"eff

<*'V"°f>C/E

0.9944

0.892

0.949

0.9937

1.010

0.990

0.991

0.9999

0.901

0.950

0.9999

1.019

0.991

0.998

0.3

1.1

0.3

0.3

1.2

0.3

4.0

{28a^/25o A JEZEBEL; increases in this ratio tended

to disturb the current agreement of the same ratio for

GODIVA. (This might not have been true had the ^^^Pu

and/or ^ssu inelastic files or the covariance of

the fission spectrum shape' been included.) Since

the experimental ratio of ratios [(^^of/^^ofJgODIVA/ {^^af/^^af)j£l^Q^l^ is more surely well known the

discrepancy in one of these ratios would appear to be

real and of necessity be related to the calculated

flux spectrum. The adjustment was characterized by

a x^/degree of freedom of 1.55; there is a '\^Z0%

probability that x^/degree of freedom is at least that

large if the assigned integral and differential errors

are correct. This minor overall inconsistency cannot

be given much weight until uncertainties are included

for all important quantities.

ments cannot be divorced from associated adjustments

to the other four cross section type reactions; the

adjustments must yet be confirmed by review and exten- sion of the current covariance files before use in

associated analysis.

Conclusions

Uncertainties due to the ^^^U{n,f) standard were estimated to comprise more than half of the calculated

uncertainty for criticality and v^af/^^af)^ spectral

index in JEZEBEL as well as GODIVA, though the JEZEBEL

assembly actually contains no ^ssy. are not able,

at this time, to predict criticality or {^^Of-/^^a^

to anywhere near the accuracy obtained by direct mea- surements, and therefore the integral results are significant to our analysis capability and, particularly in the case of criticality, could eventually

lead to improvement in our knowledge of the ^^^U{n,f)

cross section. Inclusion of integral information from

GODIVA and JEZEBEL in an adjustment procedure was effective in reconciling all parameters other than the

(_^^of/^^af)c measurement in JEZEBEL. The adjustment

procedure made changes of less than one standard

deviation for the 239p,j gp^j 235li fission and capture

cross sections including an increase of '^^1.5% for

the "ENDF/B-V" 235u(n,f) cross section above 1.3 MeV. This specific adjustment, as well as the strength of

the JEI£QEl{'^^af/^^of)c discrepancy, could change

with inclusion of inelastic covariance files and must

be viewed cautiously at this time.

Acknowledgements

The authors take pleasure in acknowledging the

significant contributions of R. Q. Wright and J. L. Lucius in the generation of cross sections, transport

results, and sensitivity profiles. E. M. Oblow and

J. H. Marable offered valuable insight on the validity

of the sensitivity profiles. The guidance of F. G. Perey in the generation of multigroup covariance files was essential as was the efforts of J. D. Drischler in

processing of such data and in carrying forth the data

adjustment calculated. The authors appreciate the

advice of G. Hansen with respect to the measured un- certainties in GODIVA and JEZEBEL. Many of the original

covariance files generated by F. Defillipo were developed from information generated by R. B. Perez,

G. deSaussure, R. Gwin, and L. Weston. The authors

gratefully appreciate the careful review of this paper

performed by J. H. Marable, G. deSaussure, and E. M. Oblow. Timely information important to this study pertaining to the proposed ENDF/B-V evaluations was graciously provided by L. Stewart and W. Poenitz prior

to its formal general release. Finally, this paper ^

would not be nearly as readable without the careful

typing and organization provided by Virginia Glidewell

and would probably not exist at all without the

financial and moral support of P. B. Hemmig and F. C. Maienschein.

The actual adjustments made in this study are

less significant since the adjustment is very sensitive to additional degrees of freedom which would be

provided with the inclusion of the fissile material

inelastic covariances (which may be difficult to estimate) and the uncertainty on the fission spectrum

shape. For the record, however, only the cross sections for 235u(n,f), 235u(n,Y), 238u(n,f), 238u(n,y),

and ^39pu(n^f) ^^g,,.g modified in any significant way,

and at most, these adjustments were 0.4-0.6 of one standard deviation. This particular adjustment would

suggest that the ^^^U{n,f) cross section above '^^^.3 MeV be increased by 'x.l.SX; below 500 keV an additional

0.4% reduction was suggested. These suggested adjust275
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277
AND 23^U AS POSSIBLE STANDARDS FOR THE MeV REGION

S. Cierjacks

Institut flir Angewandte Kernphysik

Kernforschungszentrum Karlsruhe, F.R. Germany

The aspects of using the fission cross sections of ^"Np and ^'*U as possible standards in the MeV region are considered. In comparison to other neutron standards their

application is particularly advantageous for experiments involving white-source techniques.

Major distortions in fast neutron measurements due to frame-overlap problems and contributions from slow neutrons events can be avoided by spectrum cut-off at threshold energies.

The present data basis for both nuclei is discussed and critically examined. Some suggestions

are made of how to achieve an ultimate accuracy of 2 % with measurements employing ^"Np or ^^"U as secondary standards.

(^^'Np, ^'^U(n,f) as neutron standards, summary of experimental results, E = 0.1 - 20 MeV)

Introduction

Recent developments in fast reactor design, fast

reactor safety, waste disposal projects and in fusion

reactor technology have turned the interest in accurate

neutron data to a substantial extent towards neutron

cross sections in the MeV region. As experienced largely in the eVand the keV-range absolute flux measurements necessary for absolute cross section determinations are difficult to perform, so that the use of a cross section standard turnes out to be most preferable

also in thi-s range. Among the presently internationally

recommended six standard cross section of H(n,n),

«Li (n,a),i°B (n,a),''C (n,n), '"Au (n.y), and

(n,f) there are only two, H (n,n) and ^"U (n,f), which

are suitable for general application in the MeV re- gion. For the other four reactions either a resonance structure of increasing complexity occurs or cross sec- tions drop-off very rapidly with increasing energy ; both facts excluding them as useful standards. The use of the two remaining neutron reactions, though acceptable in principle, is also not fully satisfactory in

this range. The difficulty for the application of the

H{n,n) cross section is mainly connected with the fact,

that the use of pure hydrogen is difficult and that the

use of solid radiator foils creates either threshold

problems and/or background problems if no telescopelike flux counters are used, or suffer from counting

statistics due to the low efficiency of counter telescopes. Involving the fission cross section of ^^^U

instead, causes complications due to the appearance of

structure in the cross section up to the several hundred keV range or due to background problems occurring

from small impurities of slow neutrons. The latter fact

is mainly of importance in experiments with white neutron sources and high pulse repetition rates. In the

present paper, therefore, two possible new standards for

the MeV region are proposed. In connection with such a proposal the recent assessments in determining accurate

cross sections for these nuclei are discussed.

Possible Standards in the MeV-Region

In general an ideal standard reaction should meet

the following requirements:

i. The cross sections of the reactions should be large

and accurately known over the energy interval of

interest.

ii. The cross section should smoothly vary with neutron

energy.

iti.Good samples should easily be preparable, target

material with the required chemical and isotopic

purity ought to be readily available.

iv. A suitable neutron detector based on the standard

reaction should have good overall properties, such

as high efficiency, high stability, low sensity

against other than neutron irradiations. For use with time-of-f1

i ght devices also a good timing

characteristic is important.

In addition the above general requirements another property becomes highly desirably if a standard is to be

used in the MeV range and/or with white neutron sources

and broad continuous energy spectra.

V. The reference cross section process should be a threshold reaction, since MeV cross sections are typically small compared with eV- but also keVcross sections. Small admixtures of slow-neutrons

thus may introduce large distortions from small impurities of slow neutrons. Another favourable aspect

in this context is that such a standard could directly be used in connection with the threshold

method. The latter has favourably been employed in

a number of recent fission cross section ratio mea- surements ^

'

.

In relation to our criterion (i) fission cross sec- tions are reasonably high compared with other partial cross sections in the MeV-range. From the large number

of fissile isotopes only a small number of nuclei is of

interest for our purpose, since highly radioactive material and isotopes occuring with low abundance in the

isotopic mixture are to beexcluded from the consideration. On this basis there remain except from ^^^U

mainly three nuclei and their fission reactions: ^'^Th,

^^®U and ^^'Np. From these ^^'Th turned out to be less

suitable in view of the pronounced structure effects observed recently at Saclay in most of the energy range

between 1-2 MeV as a consequence of isomeric fission

in this nucleus. In how far structure effects to a les- ser extent play also a role in the fission cross section

of the remaining two isotopes will be discussed in the

next paragraph. Considering the criterion iii, ^'^Np

and ^'^U favourably fulfill this requirement. The monoisotopic element 2 3 7fjp ^5 nowadays widely used in medicine as an energy source for heart pace-makers. Thus

^"Np material with high chemical purity is easily

available. Likewisely highly enriched ^ can easily

be obtained from nuclear industry. Suitable fission

foils and fission detectors are standard production of

various laboratories in different countries.

From the viewpoint of detector availability, fission reactions have several advantages in the sense mentioned under item iv: Gas scintillation chambers as well as ionization chambers have high efficiencies,

good stability and good timing characteristics to be

used for measurements in the 1 nsec range. Their sen278
sitivity against y-rays and charged particles can be

largely reduced. Since suitable detectors can be run

in approximately 2 u or 4 it -geometry, their sensitivity

against anisotropy effects is small as long as sufficiently thin fission foils are employed.

Structure and Fluctuations in ^^^Np, ^^"U

As mentioned in the previous paragraph a pronounced structure in the fission cross section can occur in nuclei exhibiting subthreshold neutron fission. To a lesser extent fluctuations can also be observed as a consequence of statistical fluctuations

of average level spacings and level widths in an energy range of strongly overlapping resonances. Structure of the first type can be caused by vibrational

levels in the second wall of the double humped fission barrier, where the total width might be smaller

than the spacing of class-!I states. Since the damping

width increases rapidly with excitation energy, gross

structure .effects should be minimum for nuclei having

about the same barrier height for the first and the

second well with a deep minimum within between. An inspection of the systematics of fission barrier parameters - e.g. that given by Michaudon^) - shows that

the above condition is best fulfilled for the uranium

isotopes and neptunium. The thorium isotopes have about

the same height for the inner and outer well, but only

a rather shallow minimum within between. For nuclei

starting from plutonium, the height of the outer barrier

is systematically lower than that of the inner one; this effect increases with increasing target mass. Structure effects connected with isomeric fission

should therefore be minimum in the range of ^'^U and

^''Np. Fluctuation due to statistical effects in the

matrix elements determining the neutron reaction can also be observed provided the fission cross section is

measured in a sufficiently high resolution experiment.

As demonstrated by Bowman') such random fluctuations

can be predicted with resonance parameters obtained

from measurement in the resolved resonance region.

Bowman has also shown that the well documented fluctuations in ^'^U in the several hundred keV region can

be well understood in terms of Adler-Adler"* ) or R-matrix resonance parameters.

For ^^'Np and ^'^U very high resolution measurements have recently been carried out at the Saclay

laboratory. The result for ^^^Np obtained by Plattard

et al.') with a resolution of 0.3 ns/m in the range

from 25 keV - 2 MeV is shown in Fig. 1. The data ex- hibit over the whole energy range a structureless fission cross section. All oscillations are purely due

to counting statistics of the measurement. The data

show also that no intermediate resonance occurs near

0.3 MeV as assumed by Brown et al.^). In Fig. 2a new unpublished measurement for ^^^U of Blons and coworkers') is shown. This measurement which was also car- ried out at the Saclay linac with a 0.3 ns/m resolution reveals substantial fluctuation in the whole energy region from threshold to 'v 4 MeV. Unexpectedly

large fluctuation amplitudes exceeding partially 10 %

of the average fission cross sections are being observed. These fluctuations might have a significant influence on many of the past and future measurements in this range because the resolution and point spaces be- come important factors in the measurements. If this

observed structure turns out to be real it becomes

doubtful to some extent, whether ^'^U would be a good

standard, since one of the major requirements, that

a standard should possess a smooth excitation function,

is no longer fulfilled.

Status of Present Data

^^°U Fission Cross Sections

Ratios Relative to "^U.

The status of the cross section ratio relative to ^^^U has only recently been reviewed at the 1976 ANL Specialists Meeting on Fast Neutron Fission Cross

Section of U and ^"Pu^). Since only little new informations was obtained in addition during the meantime,

mainly a brief summary of the conclusions from the

Argonne Meeting will be given here. Present data availableforthis nucleus have been summarized in the Proceedings of the Meeting"). In Fig. 3 the more recent

data sets obtaind since - say about 1970 - are shown.

Some new data not available at the ANL Meeting have been

included. Fig. 4 shows a selection of most of the older

data available from CCDN Saclay. In these graphs several relative measurements have been renormalized at 2.5

MeV to the corresponding Ponitz'') value. With respect

to the ratio data the Working Group on Ratios concluded

at the time of the Argonne Meeting, that a large number

of experiments can be brought now into what one would

consider an excellent agreement in the energy region

from threshold to 10 MeV. This is achieved after some energy changes and after some adjustments for possible

mass changes. It was expected by the Group that a new evaluation of the ratio data would yield ratio numbers

of as good as 2 %. Above 10 MeV data becomes more sparce.

There appear to be discrepancies in the white source

data. Here are presently two cyclotron and two linac

measurements extending to 20 MeV and bej'ond. Between

cyclotron and linac measurements there occurs a divergence above '^lO MeV, showing increasingly higher ratios for the cyclotron measurements with increasing

energy. The difference approaches about 10 % at 20 MeV. This discrepancy is not yet well understood, in particular since the agreement below 10 MeV is so well in the same measurements. One possible ^r^nrce of the discrepancy might be due to the use of both ion chambers

and gas scintillation chambers in cyclotron and linac

measur^nents

.

Absolute Fission Cross Section of ^'°U

Figs. 5 and 6 compare the fission cross sections

presently available from the NEA Nuclear Cross Section

Centre at Saclay. In Fig. 5 the more recent data sets are shown, while Fig. 6 contains mainly a collection of

all old cross section measurements. The data in both

figures are shown together with the newest evaluated

curve of KEDAK 3'°). No adjustments of energy scales

and absolute values were made, although several mea- surements were arbitrarily normalized at particular neutron energies. A more detailed inspection of the data

shows that a large portion of the data is in rather good

agreement over most of the energy range between 2 and

13 MeV, an exception being the threshold region below

2 MeV.

Also after adjustment of energies due to the obvious energy shifts in several data sets there seem to exist further significant differences in the overall

shape in the threshold region. Excluding the threshold

region it can be expected that a simultaneous evaluation of ratio and shape measurements would produce fission cross sections with an accuracy of better than

3 %. in the range below about 10 MeV. Above this energy fission cross section measurements are more scarce

than ratio determinations relative to^^'^U. Excluding

discrepant very old measurements of Katase and keeping

in mind that the measurements of Wilson were not inten279
ded to be precise determinations of the ^^^U fission

cross section, there exist only two cyclotron measurements in addition which agree rather favourably with

each other within 5 % (compare Figs. 5 and 6). Unfortunately, however, this agreement is misleading since

no linac measurements are available in this range. As long as we do not know the sources for the discrepancy

between the linac and the cyclotron measurements in the

ratio determinations we cannot exclude that the absolute

cross sections have systematic uncertainties of as much

as lO % between 10-15 MeV as well as the ratios'*".

^^^Np Fission Cross Sections

Ratios Relative to "^U.

There have only very few ratio measurements of

^^'Np relative to ^'^U been made in the past. Only two

data sets are presently available from CCDN-Saclay.

These values together with a new measurement of Behrens

et al.") from the Lawrence Livermore Laboratory are shown in Fig. 7. The latter data which have been obtained with the threshold method agree favourably with

the 1967 White data by better than 2 % below 10 MeV. The additional data point given for 14 MeV deviates

from the LLL results by as much as 10 %. The third data

set of Stein turns out to be on average 4 % higher than

Behrens' ratios. Despite the general good quality of the

Livermore results for fission cross section ratios, only restricted conclusions might be drawn from these

^"Np ratio measurements. To judge the present data basis, one has to rely also on the numerous absolute fission cross section determinations.

Absolute Fission Cross Sections of ^^'^Np

In contrast to the few ratio measurements a large

number of absolute or shape measurements exists for

^^^Np. The latter fact reflects its importance as a threshold detector. Existing data contained in the CCDN

library are shown in Figs. 8 and 9. As in the case of

^"U the data are plotted separately with respect to new and old results. Here the END-FBI\_curve is included in both graphes as a reference for interrelations

of measurements in the figures. It should be noted

that the fission data are given on a double logarithmic

scale. No adjustments of nnssible energy and mass changes were made .This spread is mainly due to the fact, that some rather discrepant 14 and 2-3 MeV values exist for

this nucleus. The absolute fission cross section mea- sured in the 14 MeV range are listed in Table I. On a first view these data appear to have very large scattering. But, if we exclude the 1960 Pankratcv results

but take his new 1963 values instead, and remove all

c'.o uata with unassigned uncertainties (Rago, Henkel),

there remains only a severe discrepancy with Iyer's

value .All remaining determination are then consistent

with the newest and most accurate Of-value of Adamov

(2.43 + 0.043 b). A similar, only slightly better situation exists in the 2-3 MeV region (Table II). If we exclude for the same reasons Henkel 's measurements

and the very old Klema data, all remaining results are

"""Note added in proof: at the time of the Conference a preliminary evaluation of Pbnitz'^) became available,

which is consistent with above accuracy estimates. A comparison of "absolute" and (U8/U5)*U5(ENDF-B/IV) re- sults showed that a consistent-data-fit would have

lowered the present U5 (ENDF-B/V) data between 2-13 MeV. This would support the data of Hansen et al . and the

recent data of Szabo (both above 3 MeV). The difference

between the "two" sets could also be resolved by lower

II8/U5 values in the above energy range. This assumption

would support the ratio measurements of Stein and

recent data by Cance.

Table I. Absolute fission cross sections of ^"Np in

the 14-15 MeV range

Author, Lab. Year En (MeV) 0(b) Flux Stand.

Henkel ,LAS 1957 14 .0 2.,6 no information

railK-ra LUV ,

1 Qfifl rD 2

C 1 b % Of I ' 1 KUR 15. 0 2 .62+ 5 %

pa ril/ v7k "frw/ r a 1 1 K. r a LUV ^

1 /I U 2 . 4 +

cb 7h Of ^

' ) KUR 14. 7 2 .54+ 5 %

11 Qfifi ~u u 14. 1 2 .33+ 0. 1 rr 1967

Rann NRD 1J. 968 _/U Q 14. 0 2 .31+ Of ^o;

14.,2 2 .27+ ?

Protopopov, 1958 14.,6 2 .4 + 0,,2 USSR

Iyer, TRM 1969 14.,1 2 .98+ 0..3 0.(8)

Adamov, LEN 1977 14 .8 2 .43+ 0..047 ass. particle

cons i stent wi th the most accurate val ues of Jiacoletti of

1.59 + 2 %. Even Plattard's high value is not discrepant, considering the large error in the absolute cross section determination. Due to the scatter in the norma- lization cross section data in Fig. 9 deviate in a broad band of ± 10%. But also if one restricts consideration to more recent data sets as given in Fig. 8 data scattering exceeds bv far the 3 % limit necessary

to justify its use as a fast neutron standard cross section.

Table II. Absolute fission cross sections of ^"Np in the 2-3 MeV range

Author, Lab. Year E^(MeV) a(b) Flux. Stand.

Klema, LAS 1948 2.,5 1..45+3 % o.p(5),a.(8).

3.,0 1..48+3 % a;(9)

Henkel , LAS 1957 2.,0-2.,5 1.,48 no inform.

2,,5-3..0 1..45

Schmitt, ORL 1959 2.,82 1 .65+10 % Of(8)

White, ALD 1966 2.,25 1..67+0.1 Of (5)

1967

Grundl ,LAS 1967 2.,0-2..5 1..652+0.05 0^(8)

2.,5-3..0 1..575+0.05

Jiacoletti

,

1967 2.,75 1..59 +2 % Of(5)

LAS

Brown, LAS 1970 2..0-2,.5 1..62 +10 % 0^(5) Davey

Kobayashi

,

1973 3.,5 1..65 +0.17 InllS

KTO

Plattard, SAC 1976 2,.0-2,.1 1..79+12 % a ^(5)

Proposal for Additional Measurements

It has been demonstrated that the availability of

Np and^'lJ as additional neutron standards for the

MeV range would be desirable. Furthermore it was shown

that the fission cross sections for these data are not

yet sufficiently known to carry out precise flux mea- surements in the accuracy region of < 2 %. In order to reach this long-term goal, further experimental and

280
evaluational effort is necessary. In this context the

following measurements are suggested:

i. Ratio measurements for Np relative to ^^^U over

the whole energy region from threshold to 20 MeV

employing the threshold method. Resolutions might

be moderate of the order of 5 %. A high accuracy

for the absolute neutron energy of 1 % should be

obtained. Useful measurements must aim a statistical accuracy of better than 2 %. Measurements

can rather easily be carried out with a cyclotron or some of the existing linacs.

ii. Additional absolute measurements of the fission

cross section of ^'^Np at^l4 MeV and at an energy

point between 2-3 MeV. The measurements at both

energies would serve as a cross check of good

normalizations at either one or the other energy

point.

iii. Accurate shape measurements for ^"Np and ^^^U

between about 10 and 20 MeV with moderate resolution but good statistics to establish a reliable

shape curve in this range.

iv. Extended high resolution cross section measurements

for 2^^U above 4 MeV and for ^^^Np above 2 MeV

with resolutions better than 0.3 ns/m. Measurements

should be made relative to a smooth cross section

such as H(n,n) and should extend to an energy for

U at which the percent standard deviation of the cross section fluctuation decreases below 2 %. For Np the measurements should cover at least the

plateau region between about 2-5 MeV. For the

absolute values a 10 % accuracy should be adequate.

Such investigations are necessary for U to investigate the existence and the extent of the cross section fluctuations in the few MeV range. For

Np such measurements serve to verify that the

smooth shape of the fission cross section curve continues above 2 MeV as would be most likely.

Such measurements can in principle by carried out

with most of the white-source neutron facilities.

Summary

To summarize, the ^"Np (n,f) and the^'ll(n,f) re- actions have several favourable characteristics which

justify their use as fission standards for fast neutron

cross section measurements. An inspection of the ^^'U

(n,f) data indicate that the numbers might already be as accurate as 3 % in the range from threshold to 10 MeV. But care must be taken with respect to fluctuations

at least below 4 MeV.

From this point of view ^"Np is much more favourable, since it experiences a smooth behaviour as demonstrated in a recent very high resolution measurement.

Unfortunately the fission data for this nucleus are not

yet accurate enough, to permit precise cross section

determinations. Above 10 MeV the data situation is un- satisfactory for both, Np and U. In this range, which

gains increasing importance in connection with fusion

data needs further experimental effort is necessary

to achieve the ultimately desired accuracy of s: 2 %.

Acknowledgements

The author is indebted to Dr. L. Edvardson from

the NEA Neutron Data Compilation Centre at Saclay for

supplying renormalized cross section data for ^ U and for providing graphs for almost all cross sections

shown in this paper. The cooperation of Drs. J. Behrens,

J. Blons, M. Cance, G. Grenier and S. Plattard of

readily providing their data for this survey partially

even before publication is gratefully acknowledged.

The author wishes also to thank Dr. Derrien from

CCDN Saclay for his assistance in the collection of

lackingdata sets.

References

1. J.W. Behrens, G.W. Carlson, NSE, to be published.

Preprint UCRL-78704.

2. A. Michaudon, Neutrons and Fission, Invited Paper

Int. Conf. on the Interactions of Neutrons with

Nuclei, Lowell, Mass., July 1976.

3. CD. Bowman, Nuclear Data for Reactors, Conf. Proc. Helsinki, IAEA, 1970, Paper CN-26/41.

4. D.B. Adler, F.F. Adler, USAEC Report, ANL-6792,

p. 695, 1963.

5. S. Plattard, J. Blons, D. Paya, Nucl . Sci.Eng.61^

(1976) 477.

6. W.K. Brown, D.R.Dixon, D.M.Drake, Nucl .Phys. A156 ( 1970)609.

7. J. Blons, private communication.

8. Proc. NEADNC/NEACRP Specialists Meeting on Fast Neutron Fission Cross Sections, ANL 28-30 June, 1976,

ed. by W.P. Pbnitz and A.B. Smith.

9. W.P. Ponitz, J. Nucl. Energy 26, (1972) 683.

10. B. Goel, F. Weller, KEDAK-EvaTuation File, KFK-Report

2386, Part II. 11. J.W. Behrens, J.W. Magana, C.J. Browne, Report,

UCID-17370 (1977).

12. W.P. Ponitz, private communication, 1977.

13. V.M. Adamov et al.. Contribution to this Symposium.

Data Basis Used in the Graphs

Figs. 3,4 U-238 / U-235

Name, Symbol Reference Source

COATES Har 75 75 Wash GB 7 CCDN, energy

scale revised

CIERJACKS KFK 76 76 ANL 2 KFK Listing

BEHRENS ILL 75 75 Wash 2 591 CCDN

DIFILIPPO ORL 76 76 ANL ORL CCDN

FURSOV FEI 73 73 Kiev 3 73 FEI CCDN

CONDE ULL 76 76 ANL UPP CCDN

POENITZ ANL 72 JNE 26 483 CCDN

POENITZ ANL 72 JNE 26 483 CCDN

POENITZ ANL 72 JNE 26 483 CCDN

POENITZ ANL 72 JNE 26 483 CCDN

POENITZ ANL 72 JNE 26 483 CCDN

MEADOWS ANL 75 Meadows 75 D CCDN

MEADOWS ANL 72 NSE 49 310 CCDN

MEADOWS ANL 72 NSE 49 310 CCDN

MEADOWS ANL 75 Meadows 75 D CCDN

CANCE BRC 76 76 ANL 1 BRC Paper

CANCL BRC 75 75 Kiev 1 BRC CCDN

CANCE BRC 77 Grenier P.C. 77 Li sting

WHITE ALD 67 JNE 21 671 Paper

Name, Symbol

PANKRATOV

PANKRATOV

VOROTNIKOV

KALININ

ADAMS

NETTER

GRUNDL

LAMPHERE

KATASE

BATCHELOR

KUR 60

KUR 63 KUR 71 KUR 58

ALD 61

SAC 56

LAS 67

ORL 56

KYO 61 ALD 65

Fig. 7 Np-237/U-235

BEHRENS

WHITE

STEIN

LRL 76 ALD 67

LAS 68

Fig. 8,9 Np-237

ENDFB/IV

KOBAYASHI

KOBAYASHI

KTO 73

KTO 73

Reference

AE 9 399 AE 14 177

71 MOSCOW

58 GEN 16 136

JNAB 1485

Wetter PC NSE 30 39 PR 104 1654

KATASE PC 61 NP 65, 236

Source

EANDC(J)-26

EANDC(J)-26

CCDN

CCDN

CCDN

CCDN

CCDN

CCDN

CCDN

CCDN

CCDN

CCDN

77 UCID-17370 Report

JNE 21, 671 CCDN

68 Wash 1 627 CCDN

CCDN

CCDN

CCDN

281
JIACOLETTI LAS 70 NSE 48 412 CCDN

Figs. 5,6 U--238 GRUNDL LAS 67 NSE 30 39 CCDN

KEDAK 3 PLATTARD SAC 75 NSE 61, 477 Journal

RI 74

Ged. KFK-2386( 77) Report KALININ KUR 58 58 GEN 16 136 CCDN

ALKHAZOV SCHMITT ORL 59 PR 116 1575 CCDN

VOROTNIKOV KUR 71 YFI 20 9 CCDN BROWN LAS 70 NP A 156 609 CCDN

VOROTNIKOV KUR 75 75 Kiev FEI CCDN RAGO NRD 68 HP 14 595 CCDN

CIERJACKS KFK 76 76 ANL 2 KFK Listing PROTOPOPOV USSR 58 AE 4 190 CCDN

BLONS SAC 77 BLONS PC 77 Computer- HOCHBERG KUR 59 LA-1495, 52 CCDN

cards PANKRATOV KUR 60 AE 9 399 CCDN

WILSON LAS 67 WASH 1074 CCDN WHITE ALD 66 EANDC(UK)-77 Report

GRUNDL NBS 72 ANS 15 945 CCDN HENKEL LAS 57 LA-1495, 52 CCDN

FLEROV USSR 58 AE 5 657 CCDN IYER TRM 69 BARC-474, 1 CCDN

KLEMA LAS 47 PR 72 88 CCDN

237

Np(n,{)

Fig. I Fission cross section of ^"Np

between 0.1 and 2 MeV Ref. 5)

Fig. 2 Fission cross section of U between 0.4 and 3,5 MeV (Ref. 7)

EpfMeV)

12 20 28 36

282
/ ID UD />>0 m CMCNCMCMOJ

(nge2)*P/(n9e2)*P

283
sI/)

c

01

sco

raa.o

C_3

i.o

•I—

4->

<_) <U

CO

O

u

o•(

to

(nge3)*P/(n«,,)*P

284
285

T n I u r n [ti n I' r'ii| n iT'T nT'|TT Ti nT n |Ti n'ii'T n| M i iMW T|TTTriTirTtTrTrirTnT T^rTnTn rT M TFnnr MP '

i i ii|Tri r n m i

|

I

' i ii m i ir p i n i m i [i m i m i i] ii iii i m '"i" TJTTTTTTTTTp-

_J O (/) _ (T

to H _]

UJ UJ z

ex I I — i.u

llJ a >-

CD 3

>0;

c

UJ

I "t im 1 li 1 1[ 1 1 till M 1 il I II m II [1 1 1 1 11 . 1 im [ LI L 1 1 ! I II 1 1] il I II riL LI L li I L 1 1, II 1 1 , 1 11 1 M il 1 1 1 1 II ri I li I II m I li [I I I Ml L I [ I IT) Lno

(nge2)*P/(dN,e3)^P

287
288
(q)*D

289
STANDARD INTEGRAL MEASUREMENT FACILITIES

A. Fabry

C.E.N./S.C-K. , Mol, B-2400

Belgium

The usefulness of integral measurements in standard and reference neutron fields is

examined in terms of the validation of microscopic differential-energy neutron fission cross section standards needed for fission reactor technology. This synthesis

encompasses a summary description of the identified neutron fields and of the status

of their spectral characterization, discussion of the corrections and uncertainties

involved in such experiments, and an appraisal of the accuracy of integral fission

cross sections, in particular at the light of interlaboratory comparisons. The significance of such integral measurements to the testing and improvement of evaluated

nuclear data is illustrated by a limited confrontation with the ENDF/B IV cross section file.

(Cross sections, ENDF/B, fission, integral measurements, standard neutron fields)

1 . Introduction

The aim of this paper is to examine how far

integral measurements in standard neutron fields

may contribute to obtain accurate microscopic differential-energy neutron fission cross section

standards. Other important applications of standard neutron fields to fission reactor technology

are discussed elsewhere''"'^. A standard radiation

field is tentatively defined-^ as a permanent , stable

and reproducible radiation field (neutron or gamma

or mixed) that is characterized to state-of-the-art

accuracy in terms of flux intensity and energy

spectra, and spatial and angular flux distribution.

Important field quantities must be verified by

interlaboratory measurement.

The concepts of microscopic and macroscopic

integral measurements have been delineated satisfactorily in reference^. In particular, a microscopic integral neutron cross section o is simply

the convolution of a microscopic differential-energy neutron cross section o(E) and of an energy

distributed neutron field or spectrum, 4'(E). All

quantities of relevance to the design, operation

and safety of nuclear reactors are macroscopic

integral ones : critical masses or enrichments,

reactivity worths, fuel pin power and burn up,

damage in structural-components, all directly

relate to integral reaction rates, e.g. microscopic

integral cross sections of various types, in different neutron spectra. Because differential-energy neutron cross sections are not known well enough

to afford sufficiently accurate predictions of

reactor macroscopic properties, it has been current

practice to adjust them within their uncertainties'' »^ > 9, 10 so as to match the results of a variety

of microscopic and macroscopic integral measurements

- critical masses, fission rate ratios, central

reactivity worths, material bucklings, ... - performed in zero-power "clean" '''' critical or exponential assemblies and even mock-up'''' of actual

reactors. Differential neutron spectrum measurements''^, mostly by time-of-flight , 6Li( n,a) and

proton recoil techniques, are also done in such

"controlled environments" (terminology as defined

in5); they are sometimes taken into account''^ in

the automated cross section adjustment procedures.

Neutron spectra are predominantly sensitive to

elastic and inelastic scattering cross sections;

such nuclear data are often tested in separate

neutron spectrometry experiments on simple source- driven one-material macroscopic arrangments'''*' ''^''^

.

All in all, this semi-empirical approach leads

to adjusted multigroup cross section sets associated to appropriate reactor physics computational

codes. It is not generally claimed that such adjusted cross sections are more accurate than unadjusted ones but they have proven successful and

necessary in establishing in a timely manner design

parameters for specific demonstration power plants

such as LMFBRs in the 300 MWe range''7.

In the long term however, the nuclear industry will have to keep a flexible perspective relatively to various competitive reactor and fuel

cycle options. Adjusted cross section sets developed in the frame of such specific projects as the

LMFBR need thus to be validated for wider applicability; this is because the type of approach briefly evocated above may introduce various fictitious

correlations between the many differential nuclear

data and the integral observables simultaneously

analyzed : though the accuracy of predicted reactor

integral quantities is improved, this is likely to

result of compensating errors which are propagated

in a consistent way. As far as fission cross section standards are concerned, accuracy requirements are and will remain severe, whatever the

nuclear strategy. The potential shortcomings" of cross section adjustment procedures can in this

case be largely alleviated if accurate integral

data sensitive only to fission cross sections are considered and given adequate weight. Integral

microscopic fission cross section measurements in

standard neutron fields serve this purpose. Such

measurements provide straightforward, relevant and

general integral constraints that must be satisfied

by differential data.

In this paper, standard neutron fields are identified, the status of their spectral characterization is briefly discussed as well as the quality of integral fission cross section measurements.

290
Some ratios of integral fission cross sections are largely insensitive to neutron spectral shapes and

can be considered as standard integral observables,

even if not obtained in standard neutron fields : this review consequently encompasses also some

reference-5 neutron fields less well characterized

than the standard ones. Finally a quick outlook

at selected integral results in the various considered facilities reveals a few interesting trends

relatively to the ENDF/B IV cross section file.

2. Identification of Standard Neutron Fields

The definition of standard neutron fields given in the introduction does not specify what is

meant by a state-of-the-art accuracy characterization in terms of the space, angle and energy dependent absolute flux spectrum ^'(E, r, To a large degree, this vague qualification stems from

the fact that the accuracy level at which a neutron

field may be considered as a standard depends on

the application and energy range considered.

For instance, if the field is used to calibrate or validate an instrument aimed at measuring

scalar reactor neutron spectra with an energy resolution of 5% to a requested accuracy of +_ 5% in the

energy range 10 keV - 1 MeV, there is no need for

the standard to be characterized much better than

according to these specifications; in this example

nevertheless, the accuracy of the flux spectrum

above 1 MeV and below 10 keV should be sufficient

to allow adequate corrections for the eventual sen- sitivity of the instrument to higher and lower

energy neutrons; furthermore, the gamma dose to

neutron fl\xx ratio should not be excessive (typically, 10"^ rad hr~''/cm~2sec~'' constitutes an upper

bound). Thus, a number of more or less subtle con- siderations are involved in deciding whether a particular permanent, stable and reproducible neutron

field is indeed a standard in terms of a given

application.

For the integral testing and potential improvement of differential-energy cross sections, it is

the shapes of the cross sections and of the flux

spectrum as a function of energy that determine the

sensitivity of the integral cross sections to flux

spectral uncertainties. The less rapid is the

variation of a cross section, or flux spectrum, or both, with energy over the reaction response range

(e.g. over the energy range contributing to 90% or so of the reaction rate) and the less stringent is

the accuracy with which the spectral shape must be

known. Although a +_ 1 -2% accuracy goal for differential-energy fission cross sections means as well a +_ 1 - 2% target for the integral fission

rate measurements and for the determination of the

total absolute flux in the standard neutron fields,

the corresponding needed accuracy for the flux

spectral characterization is generally much less

severe, of the order of + 5% over the energy response range; this is still a difficult target to

meet, but it is manageable provided the standard

flux spectrum displays little or no energy structure.

A neutron field can thus be standard relatively to

a given type of nuclear reaction and not relatively

to another : for example, the thermal -neutron induced uranium-235 fission neutron spectrum is presently a standard for the validation of fission

cross sections, but it is not for the validation

of (n, 2n) cross sections of relevance to fusion

reactor technology.

In the context of this paper - fission cross section standards needed for fission reactor technology - five types of neutron fields only have

been identified as standard.

They are, in order of decreasing spectral

average energy :

1. the californium-2 52 spontaneous fission neutron

spectrum

2. the theimal -neutron induced uranium-235 fission

neutron spectrum, ^^^^ 3- the i_ntermediate energy _standard rieutron _fields,

ISNF, a family of boron-lO tailored, carbon mo- derated neutron spectra driven by fission neutrons ; the spectral hardness is variable within

some limited energy range and the gross spectral

shapes are comparable to LMFBRs

k, 1/E or near ''/E shape neutron spectra

5. Maxwellian thermal neutron spectra

3. Appraisal of Standard Neutron Fields

Californium-232 Fission Spectrum Facilities

252

Intense spontaneous fission sources of Of provide a close approximation to an isolated point neutron source. Typically, this involves simply a small aluminium-steel capsule containing a few mgr

of Cf202S0^ and associated, lightweight detecting

instruments suspended in a low scatter environment

(a large room or outdoor). Such facilities have

been developed in recent years by a few laboratories,

most noticeably NBS (USA)^^^ piB (Germany)''^ and

lEP (Hungary )20, Integral fission cross section

measurements in the ^^^Cf fission neutron spectrum

have been performed at NBS21 ,22 an accuracy

level suitable for standard integral data.

The absolute source strength is determined to

+ 1 %(ti) by total absorption techniques (the manganese sulfate bath at NBS) ; current international

comparisons of ^^^Cf source strengths support so far

this accuracy assessment. The total absolute free

field neutron flux intensity at carefully established distances from such source is deemed known to

within + 1 - %.

The error component due to distance assessment is

typically as low as + 0- 6 % ; this is achieved''8 by

simultaneous exposure of two nearly identical detectors on opposite sides of, and equidistant from,

the source : the first-order distance error is then

associated with the separation of the detectors.

The maximum error component related to intrinsic

neutron field perturbation by scattering in the source capsule and supports is + 0.7 %•

In all neutron fields discussed in this paper, $(E, r, H) is separable and space-angle variations are

well understood.

^^^All uncertainties in this paper are quoted for a 68.3 % confidence interval (I0).

291
The fraction of the neutrons that undergo one inelastic scatter in the source capsule is indeed as low as (0.7 + 0.3) %.

The only other component of error in such measurements relates to neutron return from the environment : albedo from boundaries, irradiation support

structures, source capsule and air scatter. These

corrections and their uncertainties are very small : for a source-to-detector distance of 5 cm,

(l.'f + 0.6) % for ^^^U fission with cadmium enclosure and (O.'t + 0.7) % for

^

U fission.

The uncertainties in the determination of absolute

fission rates themselves are discussed in a subsequent section. _ _ Integral fission cross sections in the Of fission neutron spectrum are reported with an overall

combined uncertainty of + 2.2 - 2.8 %, depending on the isotope.

The spectral characterization of the californium-252 spontaneous fission neutron spectrum is

excellent over the bulk of the energy range. A

recent evaluation^-5 of eight documented spectrometry measurements indicates an accuracy of + 1 to

2 % between 250 keV and 8 MeV. This is by~far the

best known distributed energy neutron field to-day.

If spectral uncertainties are weighted by differential-energy cross sections, they can be expressed

as uncertainty components of the corresponding integral cross sections : for ^^^u and ^-^°U fission,

these components are 0.24 % and 0.95 % respectively^-^. This is a satisfactory situation from the

viewpoint of the differential-energy fission cross section validation by integral measurements.

Uranium-235 Fission Spectrum Facilities

Integral cross section measurements in the thermal

neutron induced uranium-235 fission neutron spectrum represent an effort covering almost two decades °»24,25,26^ jj^^ ^g^j fission cross section

data have been reported and most of them are ratios. This is because the generation of pure

uranium-235 fission neutron spectra is not straightforward and the determination of the absolute flux

is difficult.

Four facilities are in operation to-day : at KYOTO

University (Japan)27, LJUBLANA (Yugoslavia)^^

,

NBS (USA)^9 and CEN-SCK (Belgium)^". Background

responses in the two first ones prevent their application to non-threshold fission cross section

work. The CEN-SCK facility is the most versatile

and extensively investigated one ; it will thus be

briefly discussed.

This facility involves a one-meter diameter spherical cavity within the vertical graphite thermal

column of the BRl reactor at MOL. Two types of

arrangments have been developed to produce fission

neutron spectra :

1. central coaxial source disc aEsemblies31 for

passive measurements (solid state track re- corders, activation foils) : two bare discs of

^35u, typically diameter I9 x 0.1 mm, surround

coaxially a cadmium box containing the samples

to be exposed (Fig. 1) ; this type of arrangment is also used at NBS (30 cm diameter graphite cavity)

2. central cylindrical source shell30 concentric

to an extruded cadmium sleeve extending from

TABLE I. CORRECTION FACTORS FOR BACKGROUND RESPONSES AND NEUTRON FIELD PERTURBATIONS

IN THE CAVITY URANIUM-235 FISSION SPECTRUM NEUTRON FIELD MEASUREMENTS : STANDARD CYLINDRICAL SOURCE SHELL

TYPE OF

CORRECTION

NUCLIDE

TYPICAL ASSOCIATED

2^^u Pu UNCERTAINTY

241

Pu ^^^Np

Wall return background 0.8636 0.7810 0.8923 0.7995 0.9975 0.9978 + 0.1 to + 0.3 %

Photof ission , epithermal and thermal neutron penetration

(a) Impurity isotopes

0. 9907 0.9939 0. 9840 0.9939 0.9967 0.9935 + 0.1 %

0.999 0.997 0.999 1.000 1.000 0.998 + 0.1 %

Cadmium sleeve

perturbation

0.991 0.991 0.991 0.991 1.001 1.010 + 1.0 %

Instrumental

perturbation

1.000 1.000 1.000 1 .000 1.003 1.006 + 0.3 %

Net correction 0.8472 0.7668 0.8693 0.7875 0.9982 1.0056 + 1.05 % to + 1.10 %

For typical fissionable deposits

292
cavity bottom to reactor top : this allows fission rate traverses with a variety of active

(fission chambers) and passive detectors.

Arrangement 1 is currently applied for absolute cross section measurements. The source strength is derived from '''**^a-^ *^La absolute radiometric counting of the uranium-235 discs, using literature

values for the fission yield and the number of

neutrons per fission ; the accuracy is of + 2 %.

The flux follows by straighf orward geometry calculations^^. This arrangement has been chosen for

international radiometric integral fission cross section measurements for 235u and ^^Su. Selected

high quality sources and samples will be assembled

and exposed at CEN-SCK by end 1977, will be transferred to lAIA for dismantling, inspection and geometry control, and will be subsequently shipped for

counting by various laboratories in Europe, the U.S.

and Japan.

Table I presents, for the standard cylindrical

source shell (diameter 33 x 77 x 0.1 mm ) in

arrangement 2, an inventory of all corrections an un- certainties involved in cross section measurements

for six important fissionable isotopes33 (excluding

uncertainties in the determination of absolute fission rates, to be discussed separately). The dominating correction is for non-threshold reactions

and relates to the background responses to neutrons

returning from the cavity walls ; these fractional

responses, of 15 to 20 % in the table, can be

varied easily by modifying the diameter and height

of the shell source, in a range as large as from

2 % (using miniature fission chambers-^^) to 100 %.

These responses can be measured directly by moving

the instruments from cavity center up to half its

radius : this is because the fission flux decays

rapidly with distance to the source while the ca- vity wall return neutron flux in the energy range

of response for non-threshold fissionable isotopes

(< 1 keV) displays the remarkable property of a relatively great space-angle uniformity^^. Indeed,

the return neutron spatial and angular distribution

within the cavity would be perfectly flat if the

angular flux at the vacuum-moderator boundary was perfectly isotropic. The actual patterns can be

parametrically vizualized by considering the reciprocal configuration : a central point detector and

a source shell of increasing radius R in a cavity

of aurbitrary radius R . Figure 2 displays the relative central wall return neutron flux ratio

0 (f, P )/0 (E,0) as a function of the reduced

W 6 W source radius p = R /R for three cavity sizes and

various_neutron energy groups of approximate mid

energy E. The short history of neutron cross section technology presents here an amusing paradox : years ago, measurers of differential-energy cross sections,

alert to struggle with room return neutrons in

their experiments, tended to feel that integral measurers applying fission neutron sources in mo- derating cavities might well mishandle their cor- rections for wall return backgroxmds ; to-day,

integral measurers do not hesitate to use such

background neutrons as one of the primary spectral

components of intermediate-energy standard neutron

fields, as outlined in the next section. The

dominating uncertainties of the corrections needed

in -^^U fission spectrum integral experiments are actually the ones due to intrinsic neutron field

perturbation by the materials close to the source

and detectors, as shown by Table I.

The spectral characterization of the '-^XS thermal

fission neutron spectrum is not comparable in quality to the californium case. The previously con- sidered evaluation^-^ of spectrometry data lists

uncertainties of + 3 % to + 5 in the energy range

250 keV - 8 MeV, and these propagate into computed

integral cross section errors of 0.2*+ % and 1.7 %

for ^-^^U and U fission respectively, e.g. about

twice the ^^^Cf error for 238u fission. This eva- luation however does not include finite sample size

corrections which are important in some spectrometry experiment s35. Alth ough the long-standing inconsistencies^ between differential and integral

data in the thermal neutron induced 23 5u and 239pu

fission neutron spectra tend to be generally resolved^^, there are still some discrepancies affecting

nuclear reactions of importance to reactor dosimetry. These suggest^^'^ a slightly harder ^^Sy

fission spectrum than presently evaluat ed23 and

close to the fits of pulsed Van de Graaffs spectrometry measuremente35. Nevertheless, the impact of

these differences is small insofar as the validation of fission cross sections is concerned and

v2-5 "^sy be considered a good standard to this particular respect ; the fact that it is a less accurate standard than >,82 balance^ by its greater

practical importance as basic source of neutrons in

chain reacting systems, an argument whose weight is

obvious at the light of the introductory section.

Intermediate-energy Standard Neutr on Fields

The concept of primary intermediate-energy standard

neutron field, contemplated since about a decade,

has become reality in 1976 only when the ISNF-1

assembly was put into operation at NBS (USA), as

the result of a cooperation between NBS, CEN-SCK

(Belgium) and LOS ALAMOS SCIENTIFIC LABORATORY

(USA). The first ISNF fission cross section mea- surements are reported at this conference-^" ; the

NBS-ISNF facility is also described and analyzed to some extent39. this section, a brief and somewhat general discussion of the concept and its

physics is done.

ISNF does not designate a specific standard neutron

field, but a family of standards derived from the same fundamental principle : controlled moderation

and absorption of fission spectrum neutrons in a one-dimensional macrosccpic arrangment of materials

with differential-energy cross sections known to

the accuracy of primary standards - at least over

the energy range where they significantly participate to the mechanisms of spectral generation. The

last condition is crucial to the definition of

primary intermediate-energy standard neutron fields,

as opposed to secondary ones^ (exemplified by ZZ).

This condition bears upon a theoretical predictability argument : it is required that the ISNF space,

angle and energy neutron flux density be fully and

accurately characterized by solution of the stationary Boltzm^nn linear integro-dif f erential transport equation ; neutron spectrometry techniques are consequently not mandatory for spectral characterization purposes, but can be validated usefully by

measurements in the ISNF and contribute to the

certification of the fields. Other conditions of

ISNF definition such as adequate simulation of

LMFBR neutron spectra, reproducibility, ... are outlined elsewhere^. A parametric investigation

has resulted into some specialization of the concept : selection of boron-10 and carbon as major

293
(« R^)i centered

within a gramaterial constituants of the systems, and restriction to the class of spherical geometry arrangements

specified by Fig. 3- Thus the ISNF fields involve

a fission spectrum source shell of radius Rg concentric to a boron-10 shell (the absorber) of outer

radius (< Rg) and thickness >. in a spherical cavity of radius K(

phite moderator (the reflector).

In practice, the infinite reflector thickness is

50 cm, but 35 cm is satisfactory for most applications, and the source shell can be replaced by a few adequately distributed point sources, either

thin uranium-255 discs driven by thermal neutrons

or cali f ornium-252 capsules.

As illustrated by Fig. 4, ISNF neutron fluxes (full

line) are formed as the superposition of fission

neutrons (dotted line) and cavity wall return

neutrons (dashed line minus dotted line, not shown)

tailored by the boron-10 absorber. This superposition can be formulated semi-analyti cally in

terms of reduced radii and thicknesses = R^/Rci

(-g = Rs/Kq and 0;^ = Xa/Rc* Consider first a con- figuration in which the boron shell is replaced by

a thin cadmium absorber, the so-called ISNF/CV standard field-^. The central scalar neutron flux

spectrum "!-(£, pg,Rc) for a fission shell source of

unit total strength may be expressed as :

kuji^ '^(E,Pg,R^) X(E)

where ,\(E) is the normalized fission spectrum,

Vw(E,Rc) is the wall return neutron flux spectrum

for a central point fission source in a cavity of

radius Rq normalized so that ,io',5eV ''^W^^'^C^'^^ ~

'' and the function g (E,Pg) is displayed on figure 2.

The constant Yq ^^'^ ^^e relaxation length ^ are characteristic of the moderator (11.37 and 8.75 cm respectively for carbon). This equation quantifies

the superposition of an uncollided source spectrum

and the collided component associated to the reflector, these two basic spectral shapes being

weighted by the intensity scaling factors 1/p^ and

Yq/(1 + VR^)2. This relates the ISNF to the properties of carbon wall return neutrons, which are accurately predictable by discrete-ordinates transport theory approximations-^^ and MONTE CARLO methods.

The parametric survey calculations for the ISNF have

shown that detailed design criteria are optimized

if Pg and pa/i- S 0.5. Under these conditions,

insertion of the boron-10 absorber modifies the

above equation as follows :

HE,P3,P^,5^,R^) = ^4S^Jt7(5^) . [;v„(E,R^)g(E,P3

pling CAg(iA/Ps) = 0.99 to 1.0 for Pa/Ps = 0.5 to zero : this is a simple shadowing effect related to

the decrease of the first collision source density

in the carbon wall caused by the intensity rescaling of the uncollided flux ; b) the absorber-reflector coupling Cap(^A)^^E» ^a) ) > that varies

smoothly with energy between ^ 0.7 and unity, depending on (for Pa 0.15, the deviation from

unity is less than 5 % over the whole energy range);

this expresses the fact that the probability of

neutron absorption by the boron-10 shell is en- hanced by multiple cavity crossings of the collided

component neutrons ; this may also be seen as a shadowing effect, but the important point is that

it is governed by the intrinsic shell transmission

itself.

Summarizing, the ISNF physics is dominated by

simple and well understood mechanisms acting in a largely decoupled way and amenable to almost textbook formulation : streaming of virgin fission

neutrons in vacuum, reflective scattering of neutrons from a point source at the center of a spherical moderating cavity and transmission of neutrons

by a spherical absorbing shell.

1/E and near 1/E Standard Neutron Fields

The slowing-down of fission neutrons from

point sources homogeneously distributed in a non- absorbing infinite moderator generates a collided

neutron spectral flux that varies as the inverse

of the neutron energy above the thermalization

range. Such idealized way of creating 1/E neutron fields has unfortunately no direct implementation able to compete with californium facilities

when they meet the challenge of materializing a free field point source fission flux i Measurers

of resonance integrals, e.g. of ^'Sq o(E) not

only have to be careful in assessing what their

cadmium cut-off energy Eq actually is'^'' , how accurate their neutron self-shielding corrections are

- to list only two major uncertainty components _ but most importantly, they have to ascertain that

their integral measurements are indeed performed

in a pure 1/E standard field, e.g. place realistic

bounds on errors due to spectral shape deviations

from the ideal law. Such deviations may be small,

but it does not matter if they are not, provided

they are accurately known. Carbon cavity wall

return neutron flux spectra do not obey the 1/E law.

Yo

(1+VRc)

2J ^a^^'^a^^as^T-) ^ar^^a'^a^^'^a^)

s

The function Ta(E,6a) is the intrinsic neutron

transmission of the boron-10 shell in infinite vacuum ; deep physical insight has been brought into

this type of macroscopic neutron interaction by the

famous analytical work of Bethe, Beyster and

Carterho It is this tailoring exponential -like

term that drastically shapes the neutron spectrum

at intermediate and low energies while the other

terms essentially result in relatively small res- calings of the intensity factors for the two superposed spectral components as identified by the previous equation. For the uncollided flux, the rescaling is a transmission factor T^(6a) ^ 0. 97i almost energy-independent from a practical standpoint.

For the collided flux, the rescaling involves two

coupling functions : a) the source-absorber coubut generally speaking, and most certainly in terms

of integral fission cross sections, the deviations

are excessively well known, e.g. ISNF/CV type fields

as defined in the previous section are relevant.

They do not seem necessary from the standpoint of

this review because fission resonance integrals*^ are usually considered as well known for standard

isotopes and they also agree generally with differential data. This section could thus have been

limited to such statement. But even keeping out of

mind the various technological applications of

ISNF/CV standard fields, it seems valuable to re- port here the integral-versus-dif f erential cross section comparison displayed in Table II ; o and

OW denote cross sections averaged over the spectra

294
TABLE II. a /o : INTEGRAL CROSS SECTION RATIO

FOR i^ISSION BY GRAPHITE WALL RETURN^*)

AND URANIUM-235 FISSION SPECTRUM

NEUTRONS

NUCLIDE MEASURED COMPUTED DIFFERENCE

9.07(+ ^%) 9.37 - 3.3 %

6.93(+ ^%) 6.67 + 3.9 %

16.1 (+ 5%) 15.9 + 1.3 %

^h.k (+ 5%) 1^.8 - 2.8 %

(*)

1 meter spherical cavity, central fission

source

X(E) and v-w(E,Rq) entering the ISNF equations,

previous section. The measured integral data are

the by-product of a careful fission spectrum average cross section study-^-^, and state-of-the art

accuracy was consequently not searched for, so that error estimates are largely conservative.

The confrontation is most gratifying.

Thermal Maxwellian Standard Neutron Fields

Maxwell

rators have b

croscopic int

The data have

can be expres

2200 m/sec fi

Westcott g|.(T

M(E,T)dE wher

Maxwellian di

20° C).

ian thermal neutron spectra in modeeen used extensively to measure miegral fission cross sections.

been compiled and evaluated'*-5. They

sed as the product gx.(T)o of the

ssion cross section o anS the ___

) factor, equal to oj(E)(VE/\fE^)

e Eq = 0.0233 eV and M(E,T) is a

°

stribution of temperature T (usually

Comparisons of integral versus differential

gj-(T)o^^ values reveal consistency within mutual

uncertainties for ^^^U 0.8 %) , ^^%u 0.6 %)

and ^'^''Pu (~ 0.9 %) , but an unexplained bias of

1.5 for ^'^U, outside of experimental errors of

O.k to 0.7 %.

interlaboratory comparisons are and will remain of

utmost importance to realistically assess the un- certainties. It must be regretted that such com- parisons have generally not brought together the

differential and integral communities, who seem to

i±i± more or less ignore each other - but not always

fortunately - to this very practical respect.

Interlaboratory comparisons of fission rate measurements have been organized from time to time by

integral experimenters, usually for specific programmatic reasons. The most ancient and extensive

effort of this type, reported in I966, involved

the Idaho Division of Argonne National Laboratory,

the Zebra reactor group at Winfrith, the Vera team

at Alderraaston, the Los Alamos fast criticals group

and, to a modest extent, the Cadarache fast reactor

group ; most of the measurements have been done in

ZEBRA and FLATTOP-25 using parallel plate fission

chambers ; the isotopes were 233U, 235u, 238u and

Another exercice of this type took place

in the frame of the SCHERZO 556 unit k-infinity

lattice benchmark^realized at Winfrith, Karlsruhe,

Fontenay and Cadarache, but this was limited to the

238u/235U fission rate ratio. More recently, the

Karlsruhe, Petten, Mol and NBS groups have published^ the results of intercomparisons performed in

the MOL-ES secondary standard neutron field^ and

encompassing the absolute fission rate of 235u,

^39pu, 230y ^j^jj 237j^p . further data have been obtained since, including for higher plutonium isotopes, and the 239pu/ -'-'u fission rate results

have been updated as shown in Table III. In general, it seems that integral measurers may be con- servative in their error estimates. The combined

interlaboratory uncertainties are quoted as

+ 1.3 %, + 1.6 % and + 1.8 % for the absolute fission rates of 235u, 258u ^nd ^^^Pu in ES, which is

consistent with the estimates'*^ of the US Interlaboratory LMFBR Reaction Rate (ILRR) program , but

further, currently planned intercomparison work

and on-going mass assay efforts may well lead to the conviction that the actual uncertainties are rather of the order of + 1 9b.

2^q 255

TABLE III. INTERCOMPARISON OF ^^Pu to "^^U FISSION RATE RATIO MEASUREMENTS IN THE

MOL-Ei; FACILITY

Maxwellian thermal neutron spectra are often

used for normalization of integral fission cross section ratio measurements in other neutron fields.

Not surprisingly, a critical source of uncertainty

in such procedure lies in the thermal field flux

spectral characterization : this can be achieved

by chopper time-of-flight measurements, spectral

index comparisons to monochromatic beams (using

lutetium, europium and dysprosium activation foils)

and by transport theory computations.

^. Accuracy of Integral Fission Rate Measurements

Differential and integral measurers of microscopic fission cross sections face the same stringent accuracy requirements and very similar

difficulties regarding a major component of their

experiments : the determination of absolute fission

rates and of their uncertainties. Careful experimenters apply independent approaches at the two

major steps, mass assay of fissionable deposits

and detection of the fission events. Nevertheless,

Laboratory

(a) Departure from

recommendation^^ ^

Petten 1.162(.7/1. 8) - O.^t %

Karlsruhe 1.161 (.5/1. 5) - 0.5 %

Karlsruhe, mj iss 1.169(.5/1. 5) + 0.2 %

renormalized 7

^6

NBS-MOL 1.178(. 5/2. 0) + 0.9 %

Mol^^ 1.173(.6/1. 5) + 0.5 %

Number in parentheses are respectively the

(b)

random and systematic errors in percent

1.167(+ 2 %)

295
TABLE IV. INTEGRAL TESTING OF ENDF/BIV : FUNDAMENTAL FISSION CROSS SECTION RATIOS IN STANDARD

AND REFERENCE NEUTRON FIELDS

^Pu(n, f

)

^^'^Np(n,f) U(n, f) ^^^Pu(n,f)

235,w „^ U( n, f

)

^^«U(n,f) 238„rU( n, I 1 ^^^U(n f)

NEUTRON

MEASURED^ MEASURED MEASURED^ MEASURED MEASURED^ MEASURED MEASURED

COMPUTED COMPUTED COMPUTED COMPUTED

(a)

1 . 500 (+1 .6%) 1 .040 4.16 (+2.5%) 0.972 3.76 (+1.7%) 0.955 0.993

X^^Ca) 1.505 (+2.2%) 1.049 4.30 (+3.0%) 0. 964 3.9^ (+2.0%) 0.939 0.985

JEZEBEL^ ^

^ 1.^+9 (+5.o%r'^ 1.068 '+.59 (+3.0%)^^^ 0. 950 4.74 (+3.2%)^^^ 0. 917 0.980

GODIVA^*^^ 1.42 ( f

)

(+5.0%)^'^ 1.030 5.19 (+3.0^)^^^ 0.985 6.21 (+3.2%)^®^ 1.033 1.064

I.I67 (+2.0%) 1.025 6.92 (+2.5%) (0.926)^^ 17.8 (+1.5%) 0.9^8 0.972

CFRMF^'^^ 1.1^5 ( + 1.5fo) 1.025 7.29 (+2.3%) 0.967 20.6 (+1.4%) 0.993 1.018

BIG-10^^^ 1.199 (+1.5%) 1.023 8.52 (+2.3%) 0.965 26.8 (+1.7%) 1.014 1.037

^^^*>(E) for computation : NBS evaluation^'^

(b)

<J'(E) for computation : transport theory, ENDF/B-IV cross sections

(c) 2

4>(E) for computation : reference

C d ) 37 All experimental data taken from reference , unless otherwise indicated

(e) 49 Reference . The data relate to monoenergetic calibrations at '^2.5 MeV where absolute ratio measurements were performed for the normalization ; they are backed by interlaboratory comparisons in Flattop^^

(f) ^ ^ 50 Reference

(g) In the energy range of major response, the neutron spectrum is to be significantly revised

5. Integral data testing of selected ENDF/BIV

fission cross sections

239 235 The Pu/ U fission cross section ratio

belongs to the category of data qualified in

section 1 of standard observables for reference

neutron fields. This is maybe true also for the

237np/238u ratio, but not for the 235u/238u ratio.

Nevertheless, all three ratios are gathered in

Table IV for an array of standard and reference

neutron fields ; the Los Alamos bare metal critical spheres GODIVA and JEZEBEL are well known and

the other reference fields are the ones investigated by the already mentionned ILRR program. The

ENDF/BIV cross section file was used to obtain the

computed integral values. This table represents

a state-of-the art confrontation between differential and integral fundamental fission cross section data and illustrates clearly the major and

most interesting trends. This deserves an extensive discussion which lies outside the scope of

the present paper.

References

1. A. FABRY, P. VANDEPLAS - Fast Reactor Physics I,

389, IAEA (I968)

2. A. FABRY, G. and S. DE LEEUW - Nucl. Techn. 25,

3^9 (1975)

3. J. A. GRUNDL, C. EISENHAUER - "Benchmark Neutron

Fields for Reactor Dosimetry" in IAEA Consultants' Meeting on Integral Cross Section Measurements in Standard Neutron Fields for Reactor

Neutron Dosimetry, November 15-19 (1976)

4. W.N. Mcelroy et al. - Special Issue, Nucl.

Techn. 25 n° 2 (Feb. 1975)

5. J. A. GRUNDL, A. FABRY - "Report of Workshop

Session on Reactor Dosimetry Benchmarks" in

First ASTM-EURATOM Symp. on Reactor Dosimetry,

Petten, Netherlands (1975)

6. J. A. GRUNDL - Proc. Symp. on Neutron Standards

and Flux Normalization, p. 417, Conf-701002

(I97I)

7. The Physics of Fast Reactor Operation and Design, Proc. Int. Conf. BNES, June 24-26 (I969)

8. Proc. IAEA Helsinki Conf. "Nuclear Data for

Reactors" (1970)

9. Third Conf. Neutron Cross Sections and Technology, CONF -710301 (1971

)

10. Proc. Int. Symp. on Physics of Fast Reactors,

Tokyo, October 16-19 (1973)

11. W.G. DAVEY, W.C. REDMAN - Techniques in Fast

Reactor Critical Experiments. Gordon and

Breach Science Publ.

12. Y.A. KAZANSKIJ, A. A. VAN'KOV, E.I. INYUTIN - Atom. En. Rev. 13, 807 (1975)

296
13. E.D. PENDLEBURY - CONF-7I03OI, p. 1 (I97I)

14. A. PARVEZ, M. BECKER - Nucl. Sci. Eng. 62,

571 .(1977)

15. G. and S. DE LEEUW - in Report BLG ^95, p. 1-44

(1974)

16. H. BLUHM, G. FIEG, H. WERLE - Nucl. Sci. Eng.

5it, 300 (197^^)

17. J.L. ROWLANDS et al. - ibid. J^, 1133

18. J. A. GRUNDL et al. - Nucl. Techn. 32, 315 (1977)

19. W.G. ALBERTS et al. - NBS Spec. Publ. 425, I,

273 (1975)

20. M. BUCZKO et al. - Int. Syrap. on Californium252 Utilization, Paris, April 26-28 (I976)

21. H.T. HEATON II et al. - NBS Spec. Publ. 425, I,

266 (I975)

22. D.M. GILLIAM et al. - NBS Spec. Publ. 425, I,

270 (I975)

23. J. A. GRUNDL, C. EISENHAUER - NBS Spec. Publ.

425, I, 250 (1975)

24. ki FABRY - Report BLG 465 (1972)

25. A. FABRY - Report NEACRP-L-140, also in AERER8636 (I977)

26. A. FABRY et al. - "Reactor Dosimetry Integral

Reaction Rate Data in LMFBR Benchmark and Standard Neutron Fields : Status, Accuracy and Implications" ibid. 5

27. I. KIMURA et al. - J. Nucl. Sci. Tech. 10. 574

(1973)

28. M. NAJZER et al. - Proc. Conf. Nuclear Data for

Reactors, II, 57I (I97O)

29. J. A. GRUNDL - Proc. of "Prompt Fission Neutron

Spectra" p. 107, IAEA (I972)

30. A. FABRY, J. A. GRUNDL, C. EISENHAUER - NBS Spec.

Publ. 425, I, 254 (1975)

31. A. FABRY, K.H.CZOCK - Report IAEA/RL/27 (1974)

32. J. A. GRUNDL - Nucl. Sci. Eng. 31, I9I (1968)

33. A. FABRY, I. GIRLEA - "Quality control and calibration of miniature fission chambers by exposure to standard neutron fields. Application to

the measurement of fundamental integral cross section ratios" ibid. 3. 34. A. FABRY, J.D. JENKINS - Trans. ANS _15, 940

(1972)

35. J.M. ADAMS - "Comparison of ^"^^U and ^"^^Pu Fast

Neutron Fission Spectra" Appendix A in AERER8636 (1977)

36. M.F. VLASOV, A. FABRY, W.N. McELROY - Proc. Int.

Conf. on Interactions of Neutrons with Nuclei,

Lowell, Mass. (I976)

37. A. FABRY et al. - "Review of Microscopic Integral Cross Section Data in Fundamental Reactor

Dosimetry Benchmark Neutron Fields" ibid. 3. 38. D.M. GILLIAM - Integral Measurement Results in

Standard Fields" this conf.

39. C.M. EISENHAUER et al. - "Transport Theory Applications in Neutron Standards" this conf.

40. H.A. BETHE, J.R. BEYSTER, R.E. CARTER - J. Nucl.

En. 3, 207 ; 3, 273 ; 4, 3 ; 4, 147 (1956)

41. S. PEARLSTEIN, E.V. WEINSTOCK - Nucl. Sci. Eng.

29, 28 (I967)

42. S.F. MUGHABGHAB, D.I. GARBER - Report BNL 325

( 1 973

)

43. H.D. LEMMEL - NBS Spec. Publ. 425, I, 29O (1975)

44. P.I. AMUNDSON et al. - Report ANL 7320, 679

(I966)

45. M. DARROUZET et al. "Studies of Unit k Lattices

in Metallic Uranium Assemblies ZEBRA 8H, SNEAK

8, ERMINE and HARMONIE UK" - ibid. 10

46. M. PINTER et al. - NBS Spec. Publ. 425, I, 258

(1975)

47. D. GILLIAM, National Bureau of Standards - Private communication (1976)

48. J. A. GRUNDL et al. - Nucl. Techn. 25, 237 (1975)

49. J. A. GRUNDL, G.E. HANSEN - Proc. Nuclear Data

for Reactors, vol.1, 321, IAEA (I967)

50. H. ALTER et al. - Report ENDF-202 (1974)

297
Lock spring

Aluminium

Fig.l. TYPICAL SOURCE -DETECTOR ASSEMBLY

FOR ACTIVATION MEASUREMENTS IN THERMAL-NEUTRON INDUCED FISSION NEUTRON

SPECTRA

o

iiij

1.3

1.2

luT I.I

X3

UJ>

UJ

cc

0.9

0.6

Pj 40 GROUP DISCRETE -OROINATES COMPUTATIONS —•— CAVITY DIAM. ZRj 30cm A CAVITY DIAM. 2Rj •SOciT.

• CAVITY DIAM. 2R^ "lOOcm

R, RADIUS OF SHELL SOURCE

e" APPROXIMATE GROUP MIU-ENERGY

0.5 LO REDUCED SOURCE RADIUS p = Rj/Rj

Fig 2. SPATIAL DISTRIBUTION OF CARBON WALL

RETURN NEUTRON FLUXES

Fig. 3. GEOMETRICAL PARAMETERS OF THE ISNF

298
INTEGRAL MEASUREMENT RESULTS IN STANDARD FIELDS

D. M. Gilliam

National Bureau of Standards

Washington, D.C. 20234

Measured spectrum-averaged fission cross sections are reported for several benchmark

fast neutron fields. For the Cf-252 spectrum, absolute cross section measurements at NBS

and the University of Michigan are compared. Cross section ratios (relative to Pu-239) are reported for U-235, U-238, and Np-237 for the following fields: U-235 fission spectrum,

BIG-10, CFRMF, SIGMA SIGMA, and the Intermediate-Energy Standard Neutron Field (ISNF) at NBS.

Fission product yields measured by the Inter laboratory LMFBR Reaction Rate Program (ILRR) are reported in two categories: (1) Consensus Yields from thorough interlaboratory studies

and (2) Subsidiary Yields for isotopes studied less intensively (usually by a single laboratory). All measured yields were determined by Ge(Li) counting of fission activation foils,

with specific fission rates determined by counting fissions from a separate light deposit in an ionization chamber. The fission product yields for the CFRMF and BIG-10 fields are reported

first separately and then combined, to provide a single set of yields for a Fast Reactor Spectrum.

(Nuclear data; neutrons; standard fields; cross sections; fission yields)

Introduction

The preceding paper by A. Fabry has discussed

facilities employed in standardization of measurement

of neutron doses for reactor fuels and materials re- search. The present paper summarizes results from

tests in many of these facilities in which there has

been a direct NBS involvement in the past five years.

Most of the NBS involvement in this area has been

through the USERDA Interlaboratory LMFBR Reaction

Rates (ILRR) program, centered at the Hanford Engineering Development Laboratory.

Spec trum-Averaged

Fission Cross Sections

Absolute Cross Sections

The availability of small Cf-252 neutron sources

and manganese bath facilities has made it possible to determine the neutron flux in a fission-spectrum neutron field so that absolute spectrum-averaged cross section measurements can be made. Table I shows the

TABLE I

ABSOLUTE FISSION CROSS SECTIONS IN A

Cf-252 NEUTRON FIELD

Cmb)

U-235 U-238 Pu-239 Np-237

NBS

(Refs. 1,2) 1205 ± 27 319 ± 8 1808 ± 41 1332 ± 37

University of

Michigan

CRef. 3) 1215 ± 17 1790 t 34

Average of Measurements

1210 ± 2.04 319 ±2.5% 1800 ±2.2% 1332 ± 2.8%

Calculated Spectrum-Averaged

Cross Section*

1241 315 1789 1351

Measured/ Calculated

0.975 1.013 1.006 0.986

'Calculated cross sections were derived from ENDF/B-IV data and the neutron c-nergy spectrum given in the NBS evaluation iRe£. A) .

results of absolute fission cross section measurements

made in Cf-252 neutron fields at NBS and at the University of Michigan. The results for U-235 and Pu-239

agree within 1.0%, well within the total uncertainty

estimates of either laboratory. At both laboratories,

the neutron irradiations were performed with two counters positioned symmetrically about the source in

order to avoid sensitivity to the positioning of the

highly radioactive source. The calibration of the

manganese baths at both laboratories is based on the same standard neutron sources, NBS-I and NBS-II. The mass assay of fissionable deposits at the two laboratories have been intercompared by both fission and

alpha counting methods; and in the case of Pu-239, the mass of the University of Michigan deposit was based

almost entirely on comparison with the NBS reference

deposit by alpha counting. The good agreement of the

results is not so surprising in view of the several

areas of collaboration between the two laboratories,

and yet there are also significant areas of independence in the separate measurements. The fission counting methods were entirely different. At NBS miniature

fission ionization chambers were employed; while at the University of Michigan, track-etch detectors were

used. The scattering perturbations from the source capsule, the experiment support structures, and the

laboratory walls were all very different at the two sites. The averages of the experimental results are compared with calculated values in Table I, also. The

calculated values for U-235 and Pu-239 are not very

sensitive to the assumed neutron energy spectrum.

Thus the discrepancy of 2.5% for U-235 is primarily a disagreement of the integral measurements with the

ENDF/B-IV data. Since the total estimated uncertainty

in the integral result for U-235 is 2.0%, the disagreement cannot be considered a strong one. The measured

cross sections for U-238, Pu-239, and Np-237 agree

with calculated values within 1.4%,, indicating a generally satisfactory situation in terms of both the

ENDF/B data and the evaluated neutron energy spectrum

for Cf-252.

Cross Section Ratios

The Intermediate-Energy Standard Neutron Field

(ISNF)

Table II presents the results of initial measurements in the Intermediate-Energy Standard Neutron

Field (ISNF) which has recently become operational at

NBS. The fission cross section ratios for U-235,

U-238, and Pu-239 were measured by back-to-back fission counting in an NBS double fission ionization

chamber at the center of the ISNF facility. _The most

accurate measurement of the ratio a£(U-238)/5j^(U-235)

299
was achieved by use of a natural enrichment uranium

sample which allowed the deposit mass ratio and counter efficiencies to be determined by observing the

fission rate ratio in a thermal neutron beam prior to

TABLE II

ISNF MEASUREMENTS:

FISSION CROSS SECTION RATIOS

ajCU-238)/aj(U-23S)

Comments on Separate

Measurements and Averages

0.0914 ± 0.7%

0.0921 ± 2.0%

0.0925 ± 3.0%

Natural uranium vs. 99.75% enriched U-235

1600/1 depleted uranium vs. 99.75% enriched U-23S

Thick U-238 deposit,

250,000/1 depletion; same U-235

0.0918 ± 1.1%*

O ClQ") 1 + 1 T * U . U7^0 — 1 . 1^

Average in situ without

correction for scattering

in the detector

Corrected average for unperturbed ISNF field

aj(Pu-239)/Oj(U-235)

Comments on Separate Measurements and Averages

1.155 ± 1.8%

1.152 ± 1.9%

99.98% enriched Pu-239, same U-235

99.12% enriched Pu-239, same U-235

1.155 ± 1.8%* Average (the scattering

correction is nil in this case)

*The error estimate given here does not include any allowance for epi- thermai neutron perturbations. These perturbations are expected to be small, based on Fabry's experience with SIGMA-SIGMA (Ref. 12). However, these ISNF results cannot be considered as final until the epithermal neutron perturbations have been determined experimentally.

the ISNF run. (The isotopic ratio of the U-235 and

U-238 in the natural uranium sample is known to better

than 0.5%. ) In all, U-238 deposits with three different depletion levels were used, giving_results that . had a spread of only 1.27,. The ratio Cff(Pu-239)/

crf(U-235) was determined using two Pu-239 deposits

from two separate batches. The spread of these results

was only 0.37,. In Table III, the thermal neutron

TABLE III

THERMAL-NEUTRON-INDUCED FISSION RATES

IN THE ISNF

Fissionable

Isotope

Fission Rate

in ISNF Spectrum

Cs mg )

Fission Rate

Due to Thermal Neutrons pliis Background

Cs mg )

Background and Thermal- Induced Fission Rate/ Fast-Induced

Fission Rate

U-235 3170 45 ± 11 1.4%

U-238 290 8 ± 0.5 2.8%

Pu-239 3660 47 ± 10 1.3%

"background" and other background contributions are shown to range from 1.37, to 2.87, of the fission rate

due to the ISNF spectrum. A description of the ISNF

system may be found in the compendium by Grundl. The

intermediate-energy field is formed in a spherical

cavity within the thermal column of the NBS reactor by

exposing disks of U-235 to the high thermal flux and

then using a spherical shell of B-10 to cut off the

low-energy end of the resulting degraded fission neutron spectrum. The background component shown in

Table III is residual count rate within the B-10 shell

when the driving disks of U-235 are removed. This

background arises from leakage of thermal and epithermal neutrons through penetrations in the shell and

from other effects, probably mostly fission induced by

capture gammas.

Other Dosimetry Benchmark Fields

Table IV summarizes the spectrum-averaged fission

cross section ratio results from measurements in dosimetry benchmark fields. The results are given in terms

of cross section ratios, because it is not possible to determine the absolute neutron flux precisely for most

of these fields. The choice of 5f( Pu-239) as the

normalizing cross section was made because this cross section varies the least (± 6.57,) over the set of

TABLE IV

MEASURED FISSION CROSS SECTION RATIOS

IN DOSUETRY BENCHMARK FIELDS

aj(x)/Oj(Pu-239)

Field and

Re ferences x:U-235 X U-238 X Np-237

Cf-252 fission 0.672 0 177 0 740 neutrons ± 1.6% + 1.7% + 2.2%

(1,2,3,7)

U-235 fission 0.664 0 169 0 734 neutrons ± 2.2% + 2.2% + 3.0%

(10,7)

BIG-10 0.835 0 0311 0 265

(5,7) ± 1.5% + 1.9% + 2.2%

CFRMF 0.873 0 0428 0 309

(6,7) t 1.6% + 1.9% + 2.3%

SIGMA-SIGMA 0.857 0 0483 0 333

(7.8,9) ± 2.1% + 2.5% + 2.8%

ISNF 0.866 0 0799

(7; Table II ±1.8% + 1.8%

above)

fields considered. Table V compares the ratios given

in Table IV to calculated ratios derived from ENDF/B-IV

cross section data and the spectra compiled by Grundl

and Eisenhauer.'' The ratio of measured to calculated

results for U-235 are clustered within - 1.77. about a mean value of 0.97, underscoring the result seen previously in Table I

- that the integral measurements give

lower fission rates for U-235 than those calculated

from ENDF/B IV data. The measured/calculated values

in Table V for the threshold reactions U-238(n,f) and

Np-237(n,f) show much larger fluctuations: +4.8% and

± 37., respectively. These discrepancies are more suggestive of errors in the tabulated neutron energy spectra than of consistent cross section biases. Of

course, one expects to find the calculated threshold

detector responses to be very sensitive to the neutron

enery distribution, since these isotopes see only a

small part of the total spectrum.

300
TABLE V

FISSION CROSS SECTION RATIOS IN DOSIMETRY BENCHMARK FIELDS—MEASUREMENT VS. CALCULATION

{a^(x)/a^(Pu-239) }/{o^(x) /a^(Pu-239] } Measured Calculated*

x:U-235 x:U-238 x:Np-237

C£-2S2 fission neutrons 0.970 1.007 0.981

U-235 fission neutrons 0.953 1.016 0.991

BIG-10 0.979 0.960 0.930

CFRMF 0.978 0.993 0.951

SICm-SIGMA 0.986 1.034 0.957

ISNF 0.964 l.OSS

•Calculated cross section ratios are derived from folding

together the ENDF/B-IV cross section data and the neutron

energy spectra as tabulated in the compendium by Grundl

CRef. 7).

Fission Product Yields

Consensus Fission Yields

Table VI summarizes the ILRR Consensus Fission

Yields results. The term Consensus Fission Yie Ids is

used to denote the very satisfactory agreement which

was achieved after an intensive interlaboratory effort

in which the absolute fission product gamma activities

were determined by three independent Ge(Li) counting

laboratories. In eighteen cases of yield measurements

for fission of U-235, U-238, and Pu-239, the spread of

the three individual results about their mean exceeded

- 1.97. in only two instances. The uncertainties

quoted for these radiometric fission yield data are about twice as small as some evaluators were willing

to concede to be possible until very recently The

gamma counting laboratories participating in this effort were from EG&G Idaho at the Idaho National Engineering Laboratory, Argonne National Laboratory, and

the Hanford Engineering Development Laboratory. The

specific fission counts (fissions/unit mass of a given

isotope) for the fission activation foils were determined by NBS by means of a double fission ionization

chamber. In Table VI the separate interlaboratory

averages for BIG-10 and CFRMF are first given separately and then combined to provide a single ILRR

recommendation to ENDF/B under the Fast Reactor Spectrum Category. Table VII shows the differences in

yields measured in the two separate fields. These differences are generally comparable to or smaller than

the estimated uncertainties of the measurements,

despite the somewhat harder spectrum of CFRMF.

TABLE VII

COMPARISON OF FAST FISSION YIELDS: CFRMF/BIG-10

CFRMF Yield Principal Isotope of Fission Foil

BIG-10 Yield

U-235 U-238 Pu-239 Np-237

Zr-95 -0.3 % -0.8 % 0.2 Z -4.5 % Fission

Product Ru-103 -2.1 % -0.6 % -0.6 % -0.5 7. Nuclide

Ba-140 0.0 % -0.7 7, 0.6 % 1.4 7.

However, the Ru-103 yield from U-235 has shown significant energy dependence in the ILRR measurements (See

the footnote of Table VI).

TABLE VI

ILRR CONSENSUS FISSION YIELDS

Fission

Product/ Principal

Isotope

of Foil

ILRR Mean and Combined Errors (exclusive of decayscheme errors) ILRR Average for a Fast Reactor Spectrum and Total Error

Fast Neutron Field

BIG-10 CFRMF

Zr-95/U-235 0 0646 + 1 .9% 0.0644 + 2 07. 0 0645 + 2.2% RU-103/U-235 0 0337 + 1.9% 0.0330 + 2 07. 0 0333 + 2

.

37*

Ba-140/U-235 0 0607 + 1 9% 0.0607 + 2 07. 0 0607 + 1.9%

Zr-95/U-238 0 0520 + 2 .57. 0.0516 + 2 4% 0 0519 + 2.6% RU-103/U-238 0 0635 + 2 27. 0.0631 + 2 4% 0 0634 + 2.5% Ba-140/U-238 0 0601 + 2 .27. 0.0597 + 2 4% 0 0600 + 2.2%

Zr-95/Pu-239 0 0479 + 2 07. 0.0480 + 2 1% 0 0480 + 2.3% Ru-103/Pu-239 0 0709 + 2 07. 0.0705 + 2 17. 0 0708 + 2.3% Ba-140/Pu-239 0 0329 + 2 07. 0.0532 + 2 17. 0 0530 + 2.0%

Zr-95/Np-237 0 0597 + 3 ItZ 0.0570 + 3 77. 0 0593 + 4.0% Ru-103/Np-237 0 0591 + 4 17. 0.0588 + 3 77. 0 0589 + 4.3% Ba-140/Np-237 0 0566 + 4 07. 0.0574 + 3 4% 0 0571 + 3.6%

'significant neutron energy dependence has been observed in the yield of Ru-103 from fission of U-235. The measured yields in the BIG-10 and CFRMF spectra differ by (2.1 + 1.0)%. These

fast reactor yields are about 10% higher than the ILRR average

for well-thermallzed neutrons. The error seated does not account for deviations due to energy dependence.

Table VIII gives the decay scheme data used in deriving the yields given in Table VI. To get the most

TABLE VIII

DECAY SCHEME DATA AND ERRORS

Fission

Product Nuclide

Half-Ufe*

(days)

Ganma-Ray

Energy*

(keV)

Gamma-Ray

Intensity*

(7.)

Resul tant Yield Uncertainty

(7.)

Zr-95 64.1 - 0.3 724.179

756.710

44.1 - 0.5

54.6 - 0.5 |l.l

Ru-103 39.43 - 0.10 497.08 89.0 - 1.0 1.2

Ba-liiO

La-140

12.789 - 0.006

1.6775 - 0.0008

537.35

1596.18

24.4 - 0.3

95.40 i 0.08

0.2

*These data (In columns 2, 3, and 4) were given in a private communication from R. G. Helmer, Nuclear Physics Branch, ECJAG Idaho, Inc., Idaho National

Engineering Laboratory, Feb. 23, 1977. An earlier review by Helmer and

R. C. Greenwood was published in Nuclear TechnoloRy , 25 , No. 2, February

1975, p. 258.

''In equilibrium, the ratio of the La-140 to Ba-140 activities is

T)j(Ba)/ rTt^(Ba) - T^(La)j» 1.15097 + 0.00012

accurate results in using the yields of Table VI to get fission counts from fission activation foils, one should use the same decay scheme data assumed in deriving the yields. If one does use the data of Tables

VI and VIII together in subsequent fission counting,

the errors in gamma ray intensities will cancel out

exactly and there can be considerable cancellation of errors in the half- lives.

301
Table IX compares the ILRR radiometric results

with mass spectrometric results from yield measurements

at EBR-II.''^ No comparison is made for Ru-103 data

because there is no high precision mass spectrometric

data available for this chain; a plating-out problem

is known to exist in the destructive analysis procedure, preventing accurate results for this case. The

Zr-95 and Ba-140 yields from the EBR-II and ILRR tests

agree within 27= in five of six cases. The 4.8% discrepancy for Zr-95 yield from U-238 is beyond the la error limits. There is no EBR-II data on Np-237 yields

for comparison.

TABLE IX

COMPARISON OF ILRR RADIOMETRIC YIELD DATA WITH EBR-II MASS SPECTROMETRIC DATA*

Fission Product Nuclide/

Principal

Isotope of

Fuel Specimen

ILRR Radiometric

Result

EBR-II

Mass Spectrometric

Result

Maeck . ILRR

X 100%

Zr-95/U-235 0.0645 ± 2.2% 0.0642 ± 0.9% -0.5% Ba-140/U-23S 0.0607 ± 1.9% 0.0618 ± 0.8%t 1.8%

Zr-95/U-238 0.0S19 ± 2.6% 0.0494 ± 1.6% -4.8% Ba-140/U-238 0.0600 ± 2.2% 0.0592 ± 2.0% -1.3%

Zr-95/Pu-239 0.0480 ± 2.3% 0.0471 ± 1.3% -1.9% Ba-140/Pu-239 0.0S30 ± 2.0% 0.0538 ± l.l%t 1.5%

*W. J. Maeck, Ref. 13.

tlhe mass spectrometric data from Ref. 13 has been adjusted slightly

for differences in the yields of the radioactive fission products observed in the ILRR measurements and the stable fission products observed in the mass spectrometric measurements.

the decay scheme data for the subsidiary yields. Some

comments were made previously about using the yield

data and decay scheme data together for the most accurate results. These comments are all the more important in the case of the subsidiary yields, because the errors in the gamma ray intensities are so much larger

in these cases.

TABLE XI

DECAY SCHEME DATA AND ERRORS FOR SUBSIDIARY YIELDS

Fission Product

Nuclide Half-life*

Gamma- Ray

Energy*

(leV)

Gamma-Ray

Intensity*

(%)

Zr-97 16.88(6)\ 743.3(1)^ 92.9(3)^

1-131 8.05 d 364.5 82.(3)*

Te-132 77.9(5) h 228.16(6) 89.(5)

Cs-137 30.03(15)yr 661.647(12) 85.3(4)

Ce-143 33.0(2) h 293.26(2) 47.(4)

Ce-144 284.4(4) d 133.53(3) 11.0(2)

*Except for the 1-131 data, these data were given in a private

communication from R. G. Helmer, Nuclear Physics Branch, EG6G

Idaho, Inc., Idaho National Engineering Laboratory, Feb. 23, 1977.

The number in parentheses is the uncertainty in the least significant digit.

Arbitrary error assignment.

Table X gives some subsidiary ILRR fission yield

results. There are two distinctions between the "subsidiary" yields and the "consensus" yields. The gamma

counting for the subsidiary yields was usually done by

TABLE X

ILRR SUBSIDIARY FISSION YIELDS

Fission Product/

Principal

Isotope

of Foil

Cumulative

Fractional Fission Yield

Comparison of Results from

Two Fast Neutron Fields ILRR Average for a Fast Reactor Spectrum and Total Error

Fast Neutron Field CFRMF

BIG-10

X 100%

1 BIG-10 CFRMF

Zr-97/U-235 0 0600 0 0606 1.0 0.0603 ± 2.4% I-131/U-235 0 0334 0 0328 -1.8 0.0331 ± 4.3% Te-132/U-235 0 0481 0 0486 1.0 0.0483 ± 6.1% CS-137/U-235 0 0628 0 0623 - .8 0.0626 ± 5.8% Ce-143/U-235 0 0514 0 0521 1.4 0.0517 ± 8.7% Ce-144/U-235 0 0596 0 0569 -4.S 0.0583 ± 4.5%

Zr-97/U-238 0 0568 0.0568 ± 3.0% I-131/U-238 0 0328 0 0324 -1.2 0.0326 ± 4.3% CS-137/U-238 0 0600 0.0600 ± 4.5% Ce-143/U-238 0 0424 0 0418 -1.4 0.0421 ±8.7%

Ce-144/U-238 0 0495 0.0495 ± 4.3%

Zr-97/Pu-239 0 0541 0 0547 1.1 0.0544 ± 2.5% Te-I32/Pu-239 0 0537 0 0551 2.6 0.0544 ± 6.1% Cs-I37/Pu-239 0 0676 0.0676 ± 2.6% Ce-143/Pu-239 0 0390 0.0390 ± 8.7% Ce-144/Pu-239 0 0379 0.0379 ± 5.5%

Zr-97/Np-237 0 0644 0 0632 1.9 0.0638 ± 3.8% Cs-137/Np-237 0 0650 0.0650 ± 5.2%

only one laboratory rather than three, and the decay

scheme uncertainties for the subsidiary yields isotopes are generally larger than the decay scheme un- certainties for the consensus yields. Table XI lists

References

1. H. T. Heaton II, D. M. Gilliam, V. Spiegel, C. Eisenhauer, and J. A. Grundl. Proceedings of the

NEANDC/NEACRP Specialists Meeting on Fast Neutron

Fission Cross Sections of U-233, U-235, U-238, and

Pu-239, June, 1976, Argonne, Illinois. ANL-76-90,

ERDA-NDC-5/L, NEANDC(US) - 199/L . p. 333.

2. D. M. Gilliam, C. Eisenhauer, H. To Heaton II,

and J. A. Grundl. Proceedings of a Conference on Nuclear Cross Sections and Technology, March, 1975,

Washington, D. C. NBS Special Publication 425,

Vol. I, p. 270.

3. M. C. Davis, Absolute Measurements of Fast Fission

Cross Sections for U-235 and Pu-239, Ph.D. Thesis,

University of Michigan, Ann Arbor, Michigan, 1976.

4. J. A. Grundl and C. M. Eisenhauer. Proceedings of

a Conference on Nuclear Cross Sections and Technology, March 1975, Washington, D. C. NBS Special Publication 425, Vol. I, p. 250.

5. D. M. Gilliam, J. A. Grundl, G. E. Hansen, and H. H. Helmik, LMFBR Reaction Rate and Dosimetry 10th

Progress Report, (Hanford Engineering Development

Laboratory Report) HEDL-TME 75-130, UC 79 b,d.

1976. p. NBS-15.

6. J. A. Grundl, D. M. Gilliam, N. D. Dudey, and R.

J. Popek. Nuclear Technology, 25 (1975), p. 237.

7. J. Grundl and C. Eisenhaur, "Benchmark Neutron

Fields for Reactor Dosimetry," Proceedings of the

IAEA Consultant's Meeting on Integral Cross Section Measurements in Standard Neutron Fields for

Reactor Neutron Dosimetry, Vienna, November,

1976.

302
8. M. Pinter, W. Scholtyssek, et al., Proceedings of

a Conference on Nuclear Cross Sections and Technology, March, 1975, Washington, D. C, NBS Special Publication 425, Vol. I, p. 258.

9. A. Fabry, et al., "Review of Microscopic Integral

Cross Section Data in Fundamental Reactor Dosimetry Benchmark Neutron Fields," Proceedings of the

IAEA Consultant's Meeting on Integral Cross Section Measurements in Standard Neutron Fields for

Reactor Neutron Dosimetry, Vienna, November, 1976.

10. A. Fabry, J. A. Grundl, and C. Eisenhauer, Pro- ceedings of a Conference on Nuclear Cross Sections and Technology, March, 1975, Washington,

D. C, NBS Special Publication 425, Vol. I, p. 254.

11. C. M. Eisenhauer "Neutron Transport Calculations

for the Intermediate-Energy Standard Neutron

Field (ISNF) at the National Bureau of Standards,"

Proceedings of this Conference.

12. A. Fabry, G. de Leeuw, and S. de Leeuw, Nuclear

Technology, Vol. 25 (1975), p. 349.

13. W. J. Maeck, editor, "Fact Reactor Fission Yields

for U-233, U-235, U-238, and Pu-239 and Recommendations for the Determination of Bumup on FBR

Mixed-Oxide Fuels: An Interim Project Report,"

Allied Chemical Corporation Report, Idaho National

Engineering Laboratory, ICP-1050-1, UC-79c, (1975).

14. D. M. Gilliam, " Inter laboratory Consensus Fission

Yields and Subsidiary Fission Yields," LMFBR

Reaction Rate and Dosimetry 11th Progress Report,

Hanford Engineering Development Laboratory Report,

HEDL-TME to be published, 1977.

303
ABSOLUTE FISSION CROSS SECTION MEASUREMENTS USING FIXED ENERGY NEUTRON SOURCES

G. F. Knoll

The University of Michigan

Ann Arbor, Michigan 48109

Quasi-monoenergetic neutron sources of fixed energy can permit generation of absolute

cross section data at well-known and isolated neutron energies. These data can serve

as independent normalization points for cross section shape measurements made using

other techniques. Several examples of their application to fission cross section measurements are given.

(absolute cross sections, neutron sources)

Introduction

Most of the cross section data discussed at this

symposium have been generated using techniques in which

the neutron energy either covers a continuum of values

or is arbitrarily chosen by the experimenter. In particular, time-of-flight experiments carried out

using white neutron sources are capable of closely

reproducing the shape of cross sections over extended

energy ranges. Other experiments make use of nearly

monoenergetic neutrons generated in thin targets by

charged particle beams from accelerators. The neutron

energy can be varied over wide ranges and, with the use of a flux monitor with known efficiency variation,

the cross section shape can also be determined through

repeated measurements at different neutron energies.

In either case, however, the determination of the

neutron flux in an absolute sense is a separate problem generally involving more difficult procedures

than are required for a simple shape determination.

This paper deals with a different category of

neutron sources in which the energy is fixed by fundamental nuclear parameters. They are monoenergetic to first approximation, but all show some energy spread

about the average neutron energy. In some cases, the

neutron spectrum may also contain one or more contaminant groups of neutrons whose energy differs widely

from the primary yield. With the application of cor- responding corrections, however, these fixed energy

neutron sources allow the measurement of cross sec- tions at well-defined energies and can serve as "spot

checks" on cross section shapes deduced from other

methods. The principal merit of this approach is

that it is often simpler to determine the absolute

neutron yield from fixed energy sources, and the re- sulting absolute cross section values can be used for

normalization purposes in conjunction with other

shape data.

Correction for the finite energy spectrum width

requires information on both the shape of the neutron

energy spectrum emitted by the source and the shape

of the measured cross section over the same energy

range. These neutron sources are therefore best

suited for the measurement of cross sections which do

not change greatly across the source energy spectrum.

However, any fluctuation in the cross section on an energy scale which is fine compared with the inherent

source width will be averaged out. It is conventional

to make these spectrum corrections relative to the

median or average energy of the source, and thus to reduce the data to a single cross section value at this energy.

Commonly used neutron sources which fall into the

fixed energy classification are:

A. Nuclear reactions with a large Q-value com- pared with the energy of the incident particle.

B. Filtered beams in which sources with a broad

energy spectrum are transmitted through thick samples

of absorbers with cross section minima.

C. Nuclear reactions with negative Q-value in

which the incident particle energy is just above threshold.

D. Neutrons produced by photodisintegration of

target nuclei.

In the first category, the primary examples are the D-D and D-T reactions produced in low energy accelerators. The corresponding neutrons at about 3 and

14 MeV have been widely used in fast neutron cross section measurements. Since data of this type are involved in many of the discussions presented at this

symposium, further elaboration will not be given here.

Filtered beams are also discussed in a preceding paper

by R. B. Schwartz in this symposium. Associated particle methods are commonly used to determine the absolute neutron yield in the first case, but it is

somewhat more difficult to gain absolute information

about the neutron flux provided by filtered beams.

Topics C and D listed above are the major subjects

of this paper. In both cases, circumstances can be

created in which the absolute neutron yield can be

determined to a high precision. These methods are therefore most applicable to the measurement of absolute neutron cross sections which are not dependent on other cross section data or assumptions about the efficiency of flux monitor detectors.

Endothermic Reactions Near Threshold

If a neutron producing reaction with a negative

Q-value is observed using an incident particle beam

below the threshold energy, no neutrons are produced.

If the energy of the particle beam is raised, neutrons

first appear when the particle beam just reaches the

threshold energy. Because of the motion of the centerof-mass of the incident particle and target, neutrons

at threshold are emitted only in the extreme forward

direction. Their energy can easily be calculated from

the center-of-mass motion, and typically amounts to several tens of keV. As the incident particle energy

is raised further, the neutrons observed in the laboratory frame of reference are emitted within a forwarddirected cone with ever widening apex angle. For particle energies sufficiently above threshold, the cone broadens into a full sphere and neutrons are observed

in all directions.

If the energy of the incoming particle beam can be

sufficiently well regulated, conditions can be maintained in which the entire neutron yield is confined

within a cone of small apex angle. For example, using

the ^Li(p,n)^Be reaction with a proton energy 2 keV

above threshold, the neutron yield is confined to

304
within a 12° cone. Under these conditions, the

neutron energy at the cone edge is approximately 30

keV and is between 43.2 and 18.7 keV along the cone

axis. The neutron energy spectra have been calculated

for this reaction by Poenitz and Brudermueller-'- for cone angles of 5 and 10 degrees for which the energy

spread is somewhat reduced.

As illustrated in Fig. 1, it is possible to conduct a cross section measurement under these con- ditions in which all neutrons, since they are con- fined to the narrow cone, pass through the sample

target. This type of geometry can then greatly simplify calculation of the incident neutron flux. In the example given above, the total neutron yield can be determined from a fresh target through the use of

the associated activity technique (see preceding

paper in this symposium by K. K. Sekharan) . Since

each of these neutrons passes through the sample target, the integrated neutron flux can be easily determined from the sample thickness.

In addition to the ^Li(p,n) reaction with a threshold neutron energy of 30 keV, other endothermic

neutron-producing reactions can be employed in the

same way. A lighter target nucleus increases the

center-of-mass motion energy and consequently raises

the neutron energy at threshold. For example, the

2H(p,n)3He reaction provides threshold neutrons at

64 keV.2

Photoneutron Sources

Neutrons produced by photodisintegration of

deuterium or beryllium can also be used in cross sec- tion measurements. If the incident gamma rays are monoenergetic above the photodisintegration threshold

energy, the neutron spectrum will also be nearly monoenergetic, subject only to slight broadening due to kinematic and scattering effects. The examples of

their application to cross section measurements given

here will be drawn from recent experience at The

University of Michigan. •^»'*'^ However, extensive

application of photoneutron sources in other cross section work has previously been made by Perkin et al.^, Robertson et al.^'^, and several other groups.

The advantages of photoneutron sources in cross section measurements are summarized below:

1. The sources are "quasi-monoenergetic" in

that the neutron spectrum in most cases consists of

a single kinematically-broadened peak with a low

energy tail due to neutron scattering within the source (see Fig. 2) . The absolute energy scale can

be accurately predicted from the high-precision nuclear data available on neutron binding energies and

gamma ray photon energies. Because of the simple

geometry of the sources, the energy spectrum can be

calculated with a reasonable degree of reliability.

Some data are also available from direct spectrum

measurements. ^''•^ The degree of low-energy tailing

can be reduced at the expense of a lower absolute

neutron yield by decreasing source dimensions.

2. Because of the small physical size of these

sources, they can be readily calibrated by total

absorption bath measurements. If the sources are constructed in simple spherical geometry, a straightforward neutron transport calculation gives the absolute neutron flux at any distance from the source

surface. Since these geometric calculations can be

carried out to high precision, an absolute neutron

flux can be accurately generated for cross section

measurements

.

3. Using readily available gamma ray sources produced by reactor activation, neutrons ranging in

energy from 24 keV to 964 keV can be conveniently

produced. This is an energy region in which competing

methods of monoenergetic neutron source generation

often have some difficulties.

The use of such sources also has some significant

drawbacks

:

1. In order to produce neutron yields in the 10^

to 10^ per second range, gamma ray activities of the

order of tens of curies are typically required (see

Table I). Therefore, all manipulation and handling of

the sources must be done under remote conditions within

heavily shielded working areas. Furthermore, targets

and detectors used for cross section measurements are subjected to high gamma ray exposure rates (as large

as 10^ R/hr) . These high gamma fluxes preclude the

use of many conventional detectors for reaction product counting. Therefore, most photoneutron cross section measurements have been carried out using foil

activation or particle track-etch methods which are inherently insensitive to gamma rays.

124

2. With the exception of Sb, useful gamma

ray sources have half-lives of no more than several

days. Therefore, sources must be reactivated immediately prior to each use requiring close proximity

to a reactor if elaborate handling and transportation

schemes are to be avoided.

Fission Cross Section Measurements

at The University of Michigan

Source Description

The five photoneutron sources currently in routine

use at The University of Michigan are listed in Table

I. In all cases, the source consists of a spherical

gamma ray emitting core surrounded by a 3 mm thick

spherical shell of either beryllium or deuterated

polyethelene. For most of the sources, the shells are detachable and can be fitted to a number of different

cores. This modularity allows separate reactor irradiation of the core to prevent radiation damage to the

deuterated polyethelene shells, and also permits maximal use of each set of shells. Physical dimensions

of the sources are standardized (36 mm OD) to simplify

their handling during the irradiation and calibration

phases.

Table I

Properties of Photoneutron Sources

Median Initial

Neutron Neutron

Source- Energy Source Yield

Target (keV) Half-Life (n/sec)

^"^Na-Be 964 15.00 hr. 5 X lO'' 140, ^ La-Be 770 40.23 hr. 2 X 10^

^Sa-D 265 15.00 hr. 2 X lO'' ^^Ga-D 140 13.95 hr. 6 X 10^

124sb-Be 23 60 days 5 X lo''

Target Irradiation

A typical cross section irradiation geometry is

shown in Fig. 3. Virtually all our measurements have

employed the dual target or "compensated beam" geometric arrangement in which nearly-identical targets

are positioned on either side of the spherical

305
photoneutron source. The summed reaction rate from

both targets is then a much less sensitive function

of the exact source location than if a single target

geometry were used. It is then necessary to measure

only the distance between target foils to high precision, a measurement which can be carried out in the

absence of the highly radioactive source. Dimensional

uncertainties typically result in uncertainties in

the flux of several tenths of a percent.

To minimize the effect of thermalized neutrons on the measurements, the irradiations are carried out

within a cadmium enclosure which is located at the

center of a laboratory in which all walls, ceiling,

and floor have been covered with three inches of

anhydrous borax. Despite these low albedo conditions,

a measurable reaction rate is induced in samples with

high thermal fission cross sections by neutrons which

have been moderated in the room structures. We therefore routinely carry out multiple measurements in

which the spacing between targets and source is varied.

A plot of the apparent cross section measured under

these conditions is shown in Fig. 4. We eliminate the

effect of room return neutrons by extrapolating these

plots to zero spacing.

All our fission cross section measurements are carried out with detectors located behind limited

solid angle apertures. The efficiency of this detector geometry must then be calculated from geometric

data and assumed anisotropy of fission fragments with

respect to the incident neutron direction. When using

track-etch detectors, the limited solid angle geometry

prevents difficulties arising from fragments incident

at low angles relative to the film surface which may

not give rise to etchable tracks. We also avoid com- plications arising in 2tt geometry due to fission fragments emitted near the plane of the target foil. We

are, however, dependent on angular distribution data

cind the associated uncertainties are one of the dominant sources of error in our fission cross section

measurements

.

Source Calibration

We have relied exclusively on the manganese bath

technique to provide calibration of our photoneutron

sources. Methods for the absolute calibration of

such baths are the subject of a preceding paper in

this symposium by E. J. Axton. At The University of

Michigan, we have carried out all source calibrations

relative to the secondary national neutron standard,

NBS-II. We therefore avoid many of the uncertainties

associated with absolute bath calibration, but are dependent on prior calibrations carried out on NBS-II.

Based on an evaluation by Gilliam"^, we currently

assume this uncertainty to be 0.5%.

The activity of the manganese bath is continuously sampled and counted by a sodium iodide scintillation detector. Multiscaler data are recorded

over the entire cycle of activity build-up, saturation,

and decay after source removal. After appropriate

calculations are made to account for source decay,

bath dynamics, and mixing delays, we normally obtain

data over the entire cycle which is self-consistent

to within statistical uncertainties. A single best

value for the neutron induced activity is then obtained by statistical weighting methods.

Small corrections (< 1%) must then be made in the

manganese bath data to account for differences in source self-absorption and activation of the bath by

photoneutrons produced from the natural deuterium contact of the bath. The self-absorption was directly

measured by noting the decrease in saturated bath

activity as additional absorber was placed at the source position. The photoactivation of the bath was simply determined for each source by removing the

target shells from the core. Since the uncertainties

in these corrections are small, our overall calibration uncertainty is dominated by the uncertainty in

the NBS-II emission rate.

Fission Fragment Recording

All fission fragments are recorded on polyester

track-etch films and are manually counted using a projection microscope. In order to reduce counting

errors, all important films are counted twice by independent observers and any discrepancies resolved by a third count. With proper etch conditions, visual discrimination against alpha-induced pits and other background is relatively simple.

Fission Cross Section Results

Results obtained to date using four of the photoneutron sources with targets of ^Sby ^j^jj 239p^ ^re

given in Table II. Some of these values are slight

revisions of previously-reported results, reflecting

improved counting data and updated values for some corrections. Additional data are currently being

generated from the same target foils using the Sb-Be

source. After these measurements are completed, we plan to apply the set of photoneutron sources to measurements of the ^^'u and 237i^p fission cross sec- tions. Since many of the same techniques are common

to all these measurements, additional data should

help establish the validity of the entire set of cross sections.

Table II

Fission Cross Section Results Obtained

with Photoneutron Sources

En (keV)

140

265

770

964

^"^^U (bams)

1.471 ± .024

1.274 ± .020

1.162 ± .022

1.195 ± .025

References

^'^^Pu (bams)

1.469 ± .041

1.515 ± .035

1.670 ± .037

1.643 ± .033

1. W. P. Poenitz and G. Brudermuel ler , Report EANDC33"U" (1963).

2. L. W. Weston and W. S. Lyon, Phys. Rev. 123 , 948

(1961)

.

3. D. M. Gilliam and G. F. Knoll, Annals of Nucl.

Energy 2_, 637 (1975).

4. J. C. Robertson, M. C. Davis, J. C. Engdahl , and

G. F. Knoll, Trans. Am. Nucl. Soc. 2l_, 503 (1975).

5. M. C. Davis, J. C. Robertson, J. C. Engdahl, and

G. F. Knoll, Trans. Am. Nucl. Soc. 22, 663 (1975),

6. J. L. Perkin, P. H. White, P. Fieldhouse, E. J. Axton, P. Cross, and J. C. Robertson, J. Nucl.

Energy 19_, 423 (1965)

.

7. J. C. Robertson, Nucl. Phys. 71, 417 (1965).

306
8. J. C. Robertson, T. B. Ryves, E, J. Axton, I. Goodier, and A, Williams, J. Nucl. Energy 23,

205 (1969).

9. K. Mueck and F. Bensch, J. Nucl. Energy 27, 857

(1973).

10. T. B. Ryves and J. C. Robertson, J. Nucl. Energy

25, 557 (1971),

V2A End Window

/ U3O3 Deposit

Sealing Flange

UF Target

Proton

Beam

Aperture

A I Cylinder

Gold Foils

a 5 mm Platinum

Foil Backing

,Quartz Window

56 UVP Photomultiplier

To Trigger

Time-to-AmpL Conv.,

Multichannel Analyser

Reflector

Fig. 1 Kinematic collimation provided by operating the Li(p,n) reaction 2 keV

above threshold (from G. F. Knoll and W. P. Poenitz, J. Nucl. Energy 21 , 643 (1967)).

307
10^

10"

c/.(E)

10"

10"

Ga-CD,

J

COMPUTED PHOTONEUTRON EMISSION SPECTRA

(Normalized /(/>(£) dE = 1

)

all energy

1

2 —

J

J

Na-CD,

i

J I I I I I II I I I I I I I I I I L

Na-Be

La -Be

'III

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2

NEUTRON ENERGY (MeV)

Fig. 2 Neutron spectra produced by several photoneutron sources as calculated by Monte Carlo.

Fig. 3 Dual foil "compensated beam" geometry used for fission cross section measurements.

308
235

1.6

THE APPARENT "''UCn.f) INTEGRAL CROSS

r SECTIONS OVER PHOTON EUTRON SPECTRA

1.5

1.4

CO

<

m

lb*'

1.2

1.173

1.1

a = 1.508

a = 1.298

o- = 1.173

1

140 keV

265 keV

770 keV

r

20 40 60 80

[SOURCE-DETECTOR DISTANCE]^ [CM ^]

Fig, 4 Plots of the apparent fission cross section vs. the square

of the source-target spacing showing extrapolation to eliminate the contribution of room-return neutrons.

309
IMPACT OF ENDF STANDARDS ON FAST REACTORS

Ugo Fari nel 1

i

Comitato Nazionale Energia Nucleare

C.S.N. Casaccia, Roma, Italy

ENDF/B is widely used throughout the v/orld by fast reactor designers, either directly or as a basis for adjustment procedures. Neutron standards have at least potentially a value that

goes beyond their intrinsic function in the production of evaluated nuclear data files; they

could be used as standards also for integral measurements, especially in connection with

benchmark experiments, and could provide a reference set for adjustment procedures. The

conditions under which this new use of the standards would be possible are briefly reviewed

in the paper.

(Adjustment; benchmark; cross-sections; fast reactors; integral experiments; standards)

Introduction

When I v/as asked to prepare a paper discussing vyhat was the impact of ENDF standards on fast reactors, my

first reaction was to answer by a four letter word - please don't misunderstand me, the four letter word was just f^lONE. Cooler thinking convinced me that this was perhaps not the case, that the question deserved closer

scrutiny and that even a negative answer would have to

be amply qualified. In any case, asking a European to report on this subject is either meant to be provocative or to invite provocation -and I decided to interpret it in this second way.

The question does of course not concern ENDF/B as

a whole: anybody connected with fast reactor design, in

any part of the world, will share grateful appreciation

and due respect for the virtues of ENDF/B-IV in dealing

with fast reactor problems. If ENDF standards help measure and evaluate nuclear data and prepare END Files, God bless them. But this is an indirect and

somewhat trivial consideration; the real problem is,

does the concept of neutron standards - in the sense meant in this Conference - have a more direct bearing

on the real world, I apologize for the professional

bias, I mean on the macroscopic world to which nuclear

reactors belong?

I believe that it does - or at least that it

might. But before discussing these possible impacts, at

least from a European point of view, I have to venture

on very slippery ground and refer to a subject which is

very touchy on this side of the Atlantic: cross section

adjustment, a procedure that may meet here with an ad- verse environment.

The Utopia of the IDAlists

In Europe we do not believe in the Happy Utopia of IDA - the Ideal Differential Approach. According to the

tenets of this faith, there shall come a day (the Day

of the Final Evaluation) in which all the relevant

cross sections will be known to any desirable detail

and degree of accuracy. By using the highly sophisticated calculational methods then available on the

Utoputer (Utopian Computer) it will be possible to obtain results as accurate as needed, and an estimate

of the errors to go with. Although tremendous progress

has been made in nuclear data measurement and evaluation, notwithstanding the development of powerful

codes, despite the fact that new breeds of bigger and

faster computers appear every day on the shelves, I do

not think that IDA is any closer now than it was some years ago /!/.

There are mainly two reasons behind this. The first

is that reactors are not waiting for IDA to be built.

Light Water Reactors are now designed and operated

with methods and data that make IDAlists frown their

brows. It is an upside-down world that of commercial

reactor design, where diffusion performs better than

transport and old data better than new ones. Still,

reactors are built and do work, although basic data and

methods are still unable to account for the observed

temperature coefficient, to name but one.

Adjusting to Adjustment

A similar situation applies to fast reactors. Here

the different time-scale, and emphasis, in Europe and

in the United States may account at least partially for

the difference in philosophical approach. With two demo

reactors in operation and advanced work on the

commercial plants, hardware is obviously attracting

most of the attention and of the effort, and little

space is left for fundamental research. Our reactor

physicists are hardly pressed for numbers that have to go into the design, and they would become highly

unpopular should answer something like "Wait a minute,

I need a better cross section for hafnium, let me file

it into WRENDA". We have to make use of anything that

is available, which often means integral experiments.

In this way one makes mistakes (who doesn't anyway?)

but one also makes reactors.

Incidentally, something like this might conceivably repeat itself for fusion reactors in a not too far

future, with an exchange of roles, due to the more

relaxed time-scale and fundamental approach prevailing

in Europe, as compared with the faster and more technologically-oriented CTR program in the'Ufiited

States.

The second reservation on IDA is more basic. As time goes, the amount of detail that is available on nuclear data grows tremendously, and to the reactor

designer who has to use them the situation may well seem to run out of hand. We had difficulties in

adjusting to the idea of 5 000 energy points for the

elastic cross section of iron, when Francis Perey comes

up and tells us that this is oversimplified, the cross section is much more complex and structured than that,

and the amount of detail available (and needed) is one order of magnitude greater, or more. At this point of course we are unable to deal with such a hyperfine

structure and the first thing we do is straightforward

summation to bring the number of points down to an amenable few thousands at most. Why should then one go

to such a resolution in the first place if all this

information is going to be thrown away? Somebody might

compare this situation with that of the designer of a capacitor for routine circuitry who is given a detailed

account of the electronic levels, including Lamb shifts,

of his dielectric to come out with his capacity.
A Hundred Mays to Cook a Reactor

How does all this connect with the problem of

standards? Just a moment of patience. I will first disclose to you a Party Secret: the Party of Adjusters is

not monolitic in Europe as it would appear from the

outside. There are hidden contrasts and diverging

points of view. Incidentally, our intelligence reports

speak of similar divergences in the Non-Adjusters'

Party of North America, but this is a different

question.

European adjusters exhibit a very wide range of

behavioural schemes, the extremes of which can be

exemplified by the Italian Consistent Approach on one side and the Formulaire a la Frangaise on the other

(the two extremes co-existing more or less peacefully

within the same joint fast reactor program). The

Consistent Approach 111 tries to take into account all

the information available from the differential

measurements and the theoretical models by using the

results of integral experiments only to make corrections

that are consistent with the original information, and

therefore preserve all the physical meaning. This

method implies the use of calculations for neutron

diffusion or transport that are sophisticated enough

not to introduce any bias. In this case, I believe one

can really talk of cross-section improvement rather

than adjustment (or "fudging").

The other extreme - the French formulaire or Book

of Recipees /3/ - follows a route close to that used

for thermal reactors. It boldly faces the fact that errors are likely to be introduced by the calculational

methods commonly used in design as much as by the

nuclear data, and uses integral experiments - as well

as a limited set of sophisticated calculations with

basic data - to generate multigroup cross-sections that

are applicable to a certain range of compositions and

problems to yield, by low-cost, simple calculational

methods, a final result that has an accuracy estimated

to be adequate for the design needs. In this context,

there have been two different cross section sets for

sodium according to the amount of steel or other

materials present with the sodium Obviously in

this case you cannot talk about improvement of crosssections - the only thing you can say in favour of this

approach is that it appears that you can design, build

and operate fast reactors with it.

Standards as Fixed Points. ..but How Fixed?

This distinction of a variety of positions in the

adjusters' front is important in order to discuss the

role of neutron standards It is clear that brute-force

approaches correcting for limitations in calculational

methods at the same time as for uncertainties in the

nuclear data have little, if any, interest in neutron

standards. On the other hand, standards may play a preeminent role in the more fundamental methods. The

role of standards with respect to these methods could

be twofold. On one side, they could be standards for

integral cross sections or reaction rates much in the same way as they are for differential measurements. On

the other side, they could provide a basis for the

evaluation of errors in the nuclear data, a process

which is essential for sensitivity studies and

uncertainty determination, and for any adjustment

procedure.

of this group the cross sections of interest for

reactor dosimetry should be subdivided into two

categories. Category 1 reactions are a limited set with

better known cross sections, covering all the energy

range of interest, and for which differential and

integral measurements in the standard neutron fields

are consistent within acceptable margins. Category 2 includes all the other reactions important for reactor

dosimetry. The basic idea is that Category 1 cross sections should be extracted from differential

measurements and evaluation, and that integral

experiments in benchmark fields should only assess the

validity of these evaluations - much along the

classical lines of CSWG and ENDF/B in the United

States; for Category 2 reactions, on the other hand, a more "European"approach is suggested. The original

draft conclusions of the IAEA Consultants predicted

that for most Category 2 reactions "the improvement of cross sections is expected to derive essentially from

integral measurements in benchmark fields". Strong

opposition from US differential ists prompted a milder

formulation of improvements being "expected to derive

from a combination of integral and differential

measurements which should yield internally consistent

data". What if they don't?

Whichever the formulation. Category 1 reactions hold a preeminent position in the procedure, and it is

suggested that integral experiments in benchmark fields

are aimed at measuring the ratio between Category 2 and

Category 1 average cross sections in a given neutron

field rather than the absolute values of the cross

sections themselves. In this sense, Category 1 reactions have the role of standards.

I believe that this approach is fruitful and

represents a reasonable compromise between differentialists'

and integral ists ' points of view and that it could be

usefully extended to other areas of reactor physics

(and not only reactor physics: think for instance of

damage cross sections) where integral experiments play

an important role. The situation would be still more promising if the same standards could be used in differential and in integral experiments. This would make it

much easier to reach consistency. Unfortunately, this seems impossible at least at the present time and for

various reasons.

The first reason is that most, if not all, of ENDF standards are very difficult to use in an integral

experiment. In most such experiments, only activation

measurements are made. This rules out the use of Li, B,

C and H. Even fission standards are used only indirectly with activation detectors, since energydependent yields must be known or assumed. The

situation could be improved by the development of

sufficiently precise and standardized techniques such

as alpha-sensitive solid state track detectors, or helium production assay, but at the moment these are

not simple, reproducible and precise enough to be used

in connection with standards. Even if it were possible

to measure these standards in integral experiments, it

would certainly not be possible to restrict their use

to the energy range for which they are defined as standards in differential measurements.

Finally, in my personal appreciation, the word

"standard" carries with it some implication of

stability; one would very much like to measure a length

with a ruler that does not stretch and shrink during

the operation. The desire to improve continuously the

quality of the standards, however, implies frequent

revisions of the standard cross sections, in contrast

with the stability. This contrast is particularly

evident for the U-235 fission cross section: the

extreme direct usefulness of this cross section in

Let us consider these two aspects separately. The

idea of establishing a set of international standards

for integral measurements has been proposed for

instance by the IAEA Consultants on Reactor Dosimetry

Cross Sections /5, 6/. According to the recommendations

311
reactor calculations (as distinct from its use as a standard) on one hand results in more frequent and more accurate measurements of this cross section, on the

other hand it pushes the use of new values as soon as they become available; or in other cases it encourages

the adjustment procedures in order to force agreement

with integral results. I know of many calculations

carried out now that use recent values for this cross

section, that will no doubt be reflected in ENDF/B-V

but that could not be considered the standard values at

the present time.

The Role of Standards in Error Evaluation

Standards also have a crucial role in the

assessment of uncertainties in the integral results of

calculations and in all adjustment procedures. I think

this aspect has been covered in detail in the preceding

paper II I . I just want to stress the importance of an accurate assessment of the errors and of the covariance

of errors in any consistent adjustment procedure. The

splitting of this error in the part pertinent to the

measurement relative to the standard, and in the part

inherent to the standard itself, greatly helps in

arriving at a consistent formulation of the errors and

the correlations.

References

/!/ U.FARINELLI, "The Role of Integral Experiments in

the Production of Nuclear Data for Reactor Core

and Shield Design and for Irradiation Experiments",

in Nuclear Data in Science and Technology, Vol.11,

p. 103-119, I.A.E.A., Vienna (1973)

HI A.GANDINI, M.PETILLI and M.SALVATORES, "Nuclear

Data and Integral Measurements Correlation for

Fast Reactors. Statistical Formulation and

Analysis of Method: The Consistent Approach",

International Symposium on Physics of Fast

Reactors, Vol.11, 612, Tokyo (1973)

/3/ J.C.MOUGNIOT et al., "Review of Neutronic Considerations arising from Studies for a Proposed 1200 MWe

Fast Reactor", ibid. Vol.1, paper A 44

/4/ P.BOUTEAU et al., "Formulaire de Propagation de

Neutrons dans les Milieux Acier-Sodium pour les

Protections de la Filiere Rapide", Proceedings of

the Specialists' Meeting on Sensitivity Studies

and Shielding Benchmarks, p. 148, Nuclear Energy

Agency, Paris (1975)

/5/ Proceedings of a Consultants' Meeting on Nuclear

Data for Reactor Neutron Dosimetry held at Vienna,

10-12 September 1973; I.A.E.A., Nuclear Data

Section INDC (NDS)-56/6, Vienna (1973)

i /§/ Proceedings of a Consultants' Meeting on Integral

Cross-Section Measurements in Standard Neutron

Fields for Reactor Dosimetry held at Vienna, 15-19

November 1976, I.A.E.A., Nuclear Data Section (to

be published)

77/ C.R.WEISBIN, "Propagation of Uncertainties in

Fission Cross Section Standards in the Interpretation and Utilization of Critical Benchmark

Measurements", this Symposium.

312
ABSOLUTE 2^^U, ^^®U, ^^''Np FAST NEUTRON^FISSION CROSSSECTION MEASUREMENTS*

''

Vo Mo Adamov, Bo M, Alexandrov, I« Do Alkhazov,

L» Vo Drapchinsky, So So Kovalenko, Oo lo Kostochkin,

Go Yuo Kudriavzev,' Lo Z, Malkin, Ko Ao Petrzhak,

Lo Ao Pleskachevsky, A« Vo Fomichev, Vo lo Shapakov

Vo Go Khlopin Radium Institute,

Leningrad, USSR

The fission cross sections of U, U and Np for Cf fission spectrum neutrons

and I4o8-MeV neutrons have been absolutely measuredo Coincidence method fission fragment - associated particle was used. Measurement accuracy is better than 27oo Error sources are discussedo

(Fission cross sections; absolute measurements;

335,, S38„ 237 v neutrons; U; U; Np)

Absolute measurements of induced fission cross sections of ^ ^U, ^ and ^ Np for both ^^^Cf fission

spectrum neutrons and 14o8-MeV neutrons have been madeo

The method of coincidence between fission events in the

target of the nuclide and the particles associated with

neutrons, inducing fissions, was usedo For the fission

spectrum neutron measurements, the neutron^ source was

^ ^Cf, and the associated particles were ^ ^Cf spontaneous fission fragmentSo Each^^^Cf fission fragment

corresponds to v neutrons - a value known to a precision''^

of tenths of 17.. For the measurements on 14o8-MeV

neutrons, the T(d,n)*He reaction from a neutron generator was used and the associated particles were a-parti- . cleso

The main advantages of the coincidence method are

as follows; there is no need for either direct neutron

flux determination or 1007o efficiency of associated

particle detection; the influence of slowed-down and

scattered neutrons can be minimizedo Additionally, in

the case of 14o8-MeV neutron measurements, effects due

to neutrons from other reactions are eliminated, and

there is no need to determine the solid angles and

other geometrical factors.

In the fission spectrum neutron measurements, the

experimental set-up was used, where a flat target of a fissionable nuclide was irradiated by neutrons from a ^^^Cf source of the smallest possible dimension.

Source and target were fixed 4 mm apart with an ac- curacy - 10 microns. The mass of structural materials

was negligible (Fig, 1 and 2).

The double ionization current pulse chamber was used as a fission fragment detectoro

The following expression was derived for calculation of fission cross sections for the ^^^Cf fission

neutrons

:

2S2

Cf; fission spectrum neutrons; 14o8-MeV

o N

271 A R^ f N P" G K N

O T Cf

(1)

where:

A

N

fission cross section of the measured

nuclide,

number of fissions of the measured

nuclide,

atomic weight of the measured nuclide,

Avogadro constant.

NUCLIDE MEASURED

FRAGMENT

FRAGMENT

/ NEUTRON

as2 Cf

Figo lo Experimental geometry in the cross-section

measurements for ^^^Cf fission-spectrum

neutrons.

Work supported by the International Atomic Energy

Agency (Research Contract Noo 1718/RB)

This paper was not presented orally at the Conference.

313

Cf source is deposited

on the lower backing (not

shown)

o

Fig. 2. Assembly for holding the ^^^Cf source and

fissionable nuclide target.
Cf

G

R

- mass of fissionable nuclide,

- ^^^Cf prompt neutrons average number,

- number of ^^^Cf fission events,

- geometrical factor,

- fissionable nuclide layer radius.

- the correction for overlapping in the

f fission fragments channel.

The neutron flux is introduced in Eq (1) through

the value of " (^^^Cf) and the geometrical factor which

takes into account the mutual position of the ^ ^Cf

neutron source and the target.

An exact calculation of the geometrical factor

requires

:

(a) the fissile layer radius, the distance between ^^^Cf layer and that of the fissile

nuclide, and the thicknesses of the backing

material have to be accurately determined,

the layers of ^^^Cf and of the fissile nu- clide have to be strictly parallel;

(b) the layer of fissile nuclide be uniform to better than 1%;

(c) the backings be as thin as possible and their

surfaces - mirror-polished;

(d) the mass of structural materials be minimized in the vicinity of the ^^^Cf source

and target.

For present experimental geometry

(2)

where: - the main term, taking into account

fissions by neutrons, not scattered in the

backings, in case of a "point" source

;

'Gf

- the correction term, taking into account

the ^ ^Cf source real size;

- the correction term, taking into account

fissions, induced by neutrons, scattered

in the backings

»

Expressions for 1^^, I^j are multidimensional

integrals and were calculated by numerical integration. The factors in the geometrical error estimation were made by varying of formulae parameters in

their error limits

»

The experimental geometry in case of neutron

generator is shown in Figo 3. Associated a-particles

were detected by a thin plastic scintillatoro The

detection angle was 16^ o The fission-fragment

detector was an ionization current pulse chamber. As mentioned above, there was no direct neutron flux

determination. The associated alphas were detected

in the solid angle defined by the input diaphragm of

the detector.

Provided the two main geometrical constraints are

met, 1) the solid angle of the cone of neutrons,

associated with the detected alphas, is sufficiently

smaller than that subtended by the fissionable deposit,

and 2) the fission target nonuniformity is less than

P M T

PLASTIC

SCINTILLATOR

TARGET NUCLIDE

IONIZATION

CHAMBER

COINCIDENCES

Fig. 3. Experimental geometry in 14.8-MeV neutron

fission cross-section measurements.

314
17.,; the induced fission cross section, for 14.8 MeV

neutrons, is determined from the expression:

the results of other authors.

To compare the data obtained with differential

f " N N (1 - 3)(1 - 6) n a

(3)

where: - number of coincidences (without random

ones)

,

- number of associated alphas,

- number of fissile nuclei per 1 cm^

,

- correction for undetected fission

fragments in 27i-geometry

,

- neutron flux attenuation correction.

The fission-fragment detection efficiency decrease

in 27t-geometry is due to counting plateau slope of

ionization chamber and fission fragment absorption in

the target layer. The plateau slope was determined

using fragment pulse-height spectrum analysis. The

fission fragment losses were calculated, taking into

account fission anisotropy.

The neutron flux attenuation due to scattering and

absorption on structural materials was determined both

experimentally (the materials thicknesses being doubled)

and by calculation in the same way as in the preceding

case

.

To ensure the main geometrical constraints, the

ratio N /N was measured at the different distances

c a between tritium and fission targets. These ratios were found to be constant within the errors limits for

the distances used.

Two special channels for time and pulse-height

spectra analysis were included in electronic equipment

to distinguish the true coincidences and to improve

the pules-height extrapolation to zero.

The fission targets were prepared by high frequency sputtering. The deposit nonuniformi ty had been

checked by means of a-counting scanning and was found

to be less than 17o. The masses of fissionable deposits

were determined by a-counting in a 2it-ionization

chamber, and with a low geometry surface-barrier detector. All the dimensions for low geometry were determined by optical methods; errors were found to be a few

microns (the solid angle calculation accuracy being

0.27o). Moreover, the solid angle calculation was

checked experimentally with an ^*''^Am standard source,

the latter had been calibrated by the a-y coincidence

method to an accuracy 0.17..

Isotopic analysis was carred out for all the fissionable deposits, using a semiconductor spectrometer

with energy resolution 27 keV. For ^^^U, which had

impurities undistinguishable with a- spectrometry, massspec trometric analysis was carried out.

The following a-decay half-lives were used to calculate deposits masses:

measurement results, numerical calculations of

2 38 2 5 2 and U cross sections had been made for Cf fission- sjject rum neutrons, compiled differential ^^^U

and ^ U cross sections data being taken from ref. 4. The fission neutron spectrum had been approximated by

a Maxwellian distribution with parameter, T = 1406 keV.

Calculated values obtained are in a good agreement with

our data. One should keep in mind, however, in this

comparison, that the accuracy of compiled data is much

less than that of our results.

References

1. Second IAEA Panel Meeting on Neutron Standard

Reference Data, 20-24 November, 1972, IAEA,

Vienna, 1972, IIIC 3.

2. A. H. Jaffey, K. F. Flynn, L. E. Glendenin,

W. H. Bentley, A. M. Essling, Phys . Rev. C4,

1889 (1971).

3. F. P. Brauer, R. V. Stromatt, J. D. Ludwick,

F. P. Roberts and W. L. Lion, J. Inorg. Nucl.

Chem. 12, 234 (1960).

4. I. Lalgner, J. J. Schmidt, D. Woll, Tables of

evaluated cross sections for fast reactor

materials, Karlsruhe, 1968.

5. H. T. Heaton II, J. A. Grundl, V. Spiegel,

Jr., D. M. Gilliam, and C. Eisenhauer,

Nuclear Cross Sections and Technology, Proc.

of a Conference, Washington, D. C, March 3-7,

1975. Vol. I, p. 266.

6. J. A. Grundl, V. Spiegel, Jr., C. Eisenhauer,

Trans. Amer. Nucl. Soc. 15, 945 (1972).

7. J. B. Czirr, G. B. Sidhu, Nucl. Sci. Eng. _57, 18 (1975).

8. M. G. Sowerby, B. H. Patrick, Annals of Nucl.

Sci. and Eng. 1, 409 (1974).

9. M. Cance, G. Grenier, Proc. of the NEANDC/

NEACRP Specialists Meeting on Fast Neutron

Fission Cross Sections of U-233, U-235, U-238

and Pu-239, June 28-30, 1976, at Argonne

National Laboratory, p. 237.

10. W. Hart, UKAEA, ANS(S)R, 169 (1969).

11. V. M. Adamov, L. V. Drapchinsky, G. Yu. Kudriavzev, S. S. Kovalenko, K. A. Petrzhak,

L. A. Pleskachevsky , A. M. Sokolov, Preprint

Radievogo Instituta, RI - 52, Leningrade,

1976.

1/2 (^^^U)

1/2 (^''^U)

/3 3 7

(7.0381 - 0.0048) x 10® y. (4.4683 - 0.0029) x 10^ y. T^ 1/2 (^^^Np) = (2.11 - 0.01) X 10®y.

The sources and values of errors, associated with

fission cross sections determinations, are listed in

Tables 1 and 2. The fission cross sections, determined

in the present work, are listed in Table 3 along with

315
Table 1. The errors of data, associated with determination of fission

cross sections for ^^^Cf fission spectrum neutrons (X)

Nuc lides

Error Sources

835,,

U Np

V (2^^Cf) 0.35 0.35 0.35

Geometrical factor 0.71 0.71 0.71

Mass deter- Solid angle 0.30 0.30

mina t ion O La L is LI cs 0.35 0.45

at low Part of measured

geometry- isotope a-activity 0.41 0 . 10

Ha If- life 0.20 0.47

Statis tics

Mass deter- Extrapolation to zero mination pulse height 0.51

in 271; - Absorption and

geometry scattering correction 0.30

Part of measured

Vc isotope a-activity 0.30

Statistics A DA U . 0(J 1.11 0.95

Number of Extrapolation to zero

fissions pulse height 0 .40 0.37 0.41

of isotope Correction for

measured absorption in layer 0.30 0.25 0.28

Fissions in other

isotopes 0.30 0.14

Total fission cross section error 1.44 1.68 1.59

U half-life error is taken into account in determination of a part of

the measured isotope a-activity.

316
Table 2. The errors of data, associated with determination of

14.8 MeV neutron induced fission cross sections (7=)

Nuclides

335y 338 y Np

Mass deter- Solid angle 0.35 0.35

Tninat ion S t a 1 1 s tics 0.35 0.45

at low Part of measured

geome try isotope cx~ activity 0.41 0.10

Half-life 0.20 0.47

Statls tics 0.55

Mass deter- Extrapolation to zero mination pulse height 0.51

in 2tc ~ Absorption and

geome try scattering correction 0.30

Part of measured

isotope cx~activity 0.30

Statistics 1.0 1 r\ i • U

1 /i

Number of random

coincidences 0.20 n on

Neutron flux

Number of attenuation 1.0 1.0 1.0

fissions Extrapolation to zero

pulse height 0.40 0.20 0.30

Correction for

absorption in layer 0.30 0.30 0.30

Fissions in other

isotopes 0.20 0.12

Number of associated particles 0.05 0.05 0.05

Total fission cross section error 1.60 1.70 1.94

317
Table 3. Results of absolute fission cross section measurements (and comparison

with other authors)

Nuc lides

Fission cross sections (mb)

^^^Cf fission spectrum neutrons 14.8 MeV neutrons

This work

Other authors This work Other authors

Experiment Calculation

1266 - 19 1281 1204 - 29^ 2188 - 37 2075 - 40 (14.6 MeV)^

2191 - 40 (14.6 MeV)^

2063 - 39 (14.6 MeV)^

338^ 347 ± 6 352 324 - 14^ 1207 ± 20 1193 - 20 (14.6 MeV)^

1149 - 25 (14.6 MeV)^

1442 - 23 2430 - 47 2360 - 90 (14.1 MeV)^°

2500 - 70 (14.0 MeV)^°

318
NEUTRON KN3RGY STANDARDS

G. D. James

UKAEA, Atomic Energy Research Establishment,

Harwell, England.

In recent years there have been several examples of discrepancy between the neutron

energy scales from different spectrometers. Some of the work undertaken to review and

improve neutron energy determination is noted and some suggestions on how errors can

be reduced are listed. Tne view advocated by Youden that the only worthwhile estimates

of systematic error are those made experimentally is presented. Comparison of energy

determinations for a few resonances show that at best resonance energies can be quoted

to an accuracy of about one in 10 000. A list of '^i narrow resonances, over the energyrange 0.6eV to 12.1MeV, which should prove suitable as energy standards is given. At present, not all the energies listed are known to the highest accuracy attainable.

(Accurate neutron energy determination, energy standards)

^ . Introduction

Neutron energy standards are required to help

ensure that all neutron spectrometers produce data

on energy scales that agree to within the estimated

errors of measurements. Discrepancies in neutron

energy scales, of iwhich there have been several

examples in recent years, present additional problems

for evaluators and compilers, for the users of data,

for example those who wish to use transmission data

for the calculation of neutron shielding, and for

analysts who wish to undertake a combined study of

partial and total cross section data. The work

required to revise and correct energy scales could

be saved if a set of accurately measured resonance

energies became available. These could then be

checked on each spectrometer, preferably during

each experiment. At least one such list has been

published^''' but improvements in neutron spectrometer

resolution demand that this should be revised or supplemented by much narrower resonances.

As energy measurements become progressively more

accurate, several examples of energy discrepancies

have arisen some of which have been successfully

removed but some of which are still outstanding. A discrepancy of 20keV at 6MeV between measurements

made using the 2H(d,n) reaction compared with those

made using the 3H(p,n) reaction was explained by

Davis and Noda^ in terms of a field dependent

calibration factor. More recently, a discrepancy of

25keV at 1.5MeV in the energy scale of 230u/235tj

fission cross section measurements as measured on white scarce spectrometers^3-6) removed when

inadequacies in the method of calibrating one of the

spectrometers by the resonance technique were

revealed and the results were revised using a direct

calibration in terms of measured flight path length^^'.

Measurements of the same ratio^ ' using a mono- energetic Van de Graaff neutron source give results

which are about 20keV above the mean of the broad

spectrum measurements. A discrepancy of 7keV, again

indicating that data from mono-energetic sources tend

to lie higher than white sources, is presented by

data on the energy of the peak of the resonance in

the total cross section of 6Li near 250keV. The

data available on this cross section are presented in

sect. 5.1. This resonance is very broad and neither

it nor the 238u threshold are suitable for accurate

energy comparison. Nevertheless they easily reveal

discrepajicies which amount to about ^%. Within the

group of white source spectrometers the discrepancies

are not so large. Bockhoff et al. (9) have shown,

however, that the energies of ^Soy resonances below

2.7keV as measured at Geel and Columbia^ 10J differ

by 0.1%. This difference is almost independent of

energy and is equivalent to a 10cm discrepancy in the

lengths of the 200m flight paths used in each experiment. Tnis is far greater than the errors of

measurement which are about 1cm or less.

As a framework for considering how to improve

the measurements and comparison of neutron energies,

the methods in use on three spectrometers and the

random and systematic errors in each of the methods

are briefly reviewed in sect. 2. Some considerations

which can help to reduce inaccuracies are described

in sect. 5 and the important concept of making experimental measurements of systematic errors, so ably

advocated by Youden'''''''', is presented in sect. h. Examples of the accuracies which are achieved in

energy measurements at present are discussed in

sect. 5 where the results available from transmission

measurements for 6Li, 23Na, 238u and 12c are

discussed. In sect. 6 a list of forty-one

resonances over the energy range 0.6eV to 12.1MeV is

presented. Tliis list was drawn from examples of narrow resonances kindly suggested to me by members

of an INDC sub-group on neutron energies*. It does

not yet carry the imprimatur of the sub-group but it

is likely that the list finally adopted will contain

most if not all the resonances given in sect. 6. The

establishment of a recognised list of resonances is

important in that it will encourage the measurement

and intercomparison of the energies of these resonances by scientists who recognise the need to establish accurate energy scales for their spectrometers. The conclusions that can be drawn from the

studies presented in this paper are discussed in

sect. 7.

2. Neutron resonance energy determination

Two methods of neutron energy determination are

in use; those based on neutron time-of-flight and

those based on the uses of mono-energetic charged

particles to produce mono-energetic neutrons. The

*J. Boldeman, F. Corvi, J. A. Harvey, G.D. James,

J. Lachkar, A.B. Smith and F. Voss

319
former require pulsed sources and spectrometers

based on electron linear accelerators, cyclotrons,

synchrocyclotrons and pulsed Van de Graaff machines

are in operation. The latter method is based

exclusively on Van de Graaffmachines which alone

can produce charged particle beams with energies

known to sufficient accuracy. In this section the

methods used to determine energy on three spectrometers which are typical of the range of instruments

in use are presented so that the relative magnitudes

of random and systematic errors can be more readily

appreciated.

2. 1 Neutron energy determination at ORELA

In the ORELA neutron time-of- flight spectrometer,

short bursts of l'+OMeV electrons of minimum width

5ns strike a water cooled tantalum target to produce

bursts of fast neutrons which are moderated in water

to produce a pulsed source of neutrons covering the

energy range from several MeV down to the Maxwellian

spectrum at thermal neutron energy. Neutrons are detected by a detector set at a known distance from

the source. Part of the interval between the time

when the neutrons are produced and the time when

neutrons are detected is measured and recorded by a time digitizer in units of timing channel width

which have a minimum value of 2ns. Another part of

the interval is determined by the delay between the

electron pulse and the opening of the first timing

channel - the so called initial delay. As an example (12) to illustrate the contribution of

systematic and random errors, in an experiment to measure the transmissions at a path length of

155^'f1-10mm , the following equations are used to derive neutron energy Ec and the relativistically

corrected energy Er

72. 2977 (155^4^41 -10mm)

t-(2695-2ns)

Ep = E^(1 + 1.5965 10

...1(a)

...1(b)

Here 2695±2ns is the initial delay and t is the part

of the flight time recorded on the time digitizer.

Examples of the results obtained for a few selected

resonances in Pb,Al,S, Na and '^^"V are given in

table 1 which gives the contribution to the quoted

uncertainty made by dEcjj, the error in locating the

peak of the resonance, dEt the error in energy due

to the 2ns uncertainty in the initial delay and dEj^

the error in energy due to the 10mm uncertainty in

the neutron flight path. Below the Al resonance at 5903. 5eV the error is dominated by the uncertainty

in flight path length. Above this energy the error

is dominated by the error in locating the peak of

the resonance by the fitting programme SIOB.

2.2 Neutron energy determination on the Harwell

Synchrocyclotron

In the Harwell synchrocyclotron neutron timeof-flight spectrometer, pulses of protons are accelerated to am energy of I'+OMeV and deflected

downwards to strike a 2cm thick tungsten target to produce fast neutrons by spallation. The salient

features of the target and detector system are illustrated in fig. 1. Fast neutrons from the

proton target reach a 2cm thick beryllium faced

water moderator tank from which a pulse of

moderated neutrons travels towards the detector.

Table 1

ORELA 150m Neutron Energies

Isotope

Resonance

Energy

(eV)

dEch dEt

(eV) (eV)

dEL

(eV)

BNL-325*^''5)

(eV)

206pb 3357.'+ ±0.5 0.1 0.07 OA3 3360 ±10

27ai 5903.5 ±0.8 0.3 0.16 0.76 5903 ± 8 32s 30378 ±6 k 2 30350 ±90

23Na 53191 ±27 26 it 7 53150 ±30

206pb 71191 ±18 l^t 6 9 71000 ±100

32s 97512 ±28 22 10 13 96600 ±500

32s 112186 ±33 27 12 V4 1 1

1

koo ±500

27ai 257.3 ±0.3* 276 ifl 33 257.5 ±1.8*

238u 1^+19.76*0,18 0.0*+ 0.02 0.18 iiH9.5 ±0.3

238u 2489.1^?±0.32 0.03 0.0'+ 0.32 2k8S.k ±0.7

'keV

Since late 1973 transmission measurements have

generally been carried out using either a 2.5cm thick

Li-glass detector at 50m or using a1.2omthick NE110

detector at 100m. The 50m measurements give results over the energy range lOOeV to lOOkeV and the 100m

measurements give results over the energy range lOkeV

to lOMeV. During 197^*- the distances between reference

marks at 1m, km, 13m, and 98m from the neutron

source were measured to better thain 0.5mm accuracy by

a tellurometer* . The equations used to determine

neutron energy are set out in Appendix 1 where the

symbols used have the following meaning. L is the

distance from the face of the moderator to the point

in the detector where a neutron interacts to give a detectable pulse. It is made up of the moderator

face to detector face distance P and the distance D traversed inside the detector. The neutron time-offlight is the difference between the time ^no when a neutron leaves the face of the moderator and the time t when it is detected. Using t'^, the time whenlf-rays

and fast neutrons leave the proton target, and ty when

the Y-flash is detected, t can be re-stated in terms

ofti, the recorded time between the detected Y-flash

in channel Xo and a neutron in channel X, t2 the

T-ray time-of- flight over a distance Ly, t-j the

transit time of fast neutrons from the proton target

to the water moderator and t/^ the neutron slowing

down time delay. The channel width is denoted by w,

D-y. is distance travelled by Y-rays in the detector

and Sy is the distance travelled by y-rays to reach

the face of the moderator from which they emerge.

In calculating the transit time'. of fast neutrons it

is assumed that detected neutrons of energy E are produced by neutrons of energy 2E. The slowing down

time delay is calculated from the formulae of

Groenewold and GroendijkC '. In an experiment to determine the energy of the peak of the resonance

near 250keV in the total cross section of Li

carried out in 197'*-, the Li-glass detector was used,

exceptionally, at 100m. In the same experiment the

energies of three peak cross sections in 23Na and

27ai were also determined. A careful assessment of

the errors arising in the measurement have been

made(15) and are presented in table 2.

*Tellurometer(U.K. ) Ltd., Roebuck Rd. , Chessington,

Surrey, England KI9 IRQ

Tellurometer-U.S.A. ,89 Marcus Blvd, Hauppauge ,NY1 1787

320
Table 2

Harwell Synchrocyclotron 100m Neutron Energies

Isotope

Al-27

Li-6

Al-27

Na-23

Resonance

Energy

(keV)

119. 753-0. 0it2 2^+2.71 -0.3'+ 257.16 -0.13

299.19 -0.12

dEch dEt dEL

(eV) (eV) (eV)

3 18 28

330 68 58

90 7h 62

15 97 72

2.3 Energy measurements at the Argonne Fast

Neutron Generator .

Recently, Meadows^''^'' has undertaken a close

scrutiny of the techniques involved in energy

determination at the Argonne FNG when used as a source

of mono-energetic neutrons. Yield curves and neutron

energy spectra were calculated for some (p,n)

reactions commonly used as neutron sources for energy

calibration purposes. These calculations take into

account the energy spread of the incident proton beam

and the statistical nature of the proton energy loss.

Meadows shows that when thresholds are observed by

detecting neutrons at O deg. to the proton beam

direction, the best results are obtained by plotting

the square of the yield against proton energy and

extrapolating to zero yield. A linear plot of

neutron yield against proton energy can be in error by 1 - 2keV if the energy spread of the proton beam

is large. A calibration was established by locating

the 7Li(p,n)''Be threshold (188O.6O* 0.07keV) and the

'l''B(p,nr ''C threshold (30l6.'t ± 1.6keV). Using this

calibration a measurement of a carbon resonance gave

th' value for a mean over five target thicknesses of

2078.2 - 2.8keV. This error allows for systematic

uncertainties. The error in the mean derived from

the spread of the five readings is 0.67keV. The

calibration was also confirmed by a neutron time-offlight measurement which at En= 2.7739 - O.OCtSMeV

gives Ep = ^.'t62't - 0.0046MeV to be compared with the

value = h.kGG t O.OO^MeV derived from the analysing magnet calibration.

3. Reduction of uncertainties

Several techniques are available which can be

used to reduce uncertainties in some of the quantities

involved in the determination of neutron energy. Five

of these are discussed in this section.

3. 1 Effect of resolution on s-wave resonance energies

In total cross section and transmission measurements, s-wave resonances show a marked asymmetry

caused by resonance-potential interference which

causes a minimum on the low energy side of the resonances. With worsening energy resolution, the

energy at the observed peak of the cross section, Em,

shifts to higher energy and the energy difference

between the peak and the interference minimum,

observed at Em, increases. This effect is illustrated

for the 299keV resonance in Na. in fig. 2 which is

taken from a paper by Derrien^^"^'. Derrien shows

that provided both ^\ and Em are known, the effects

of spectrometer resolution can be allowed for to derive E^, the energy at the resonance peak observed

with perfect resolution. Values of Er derived from

six measurements of different resolution are illustrated in fig. 3. The mean of the six values of

Er is 298.65 - 0.32keV. Even after carrying out the

correction, however, only the three measurements with

the best energy resolution agree within the error in

the mean. Taking these three values only, the mean

value of Ej. is 298.550 - 0.082keV which corresponds

to an error of one in 365O. This effect of energy

resolution on asymmetric s-wave resonances makes them

less suitable for accurate energy comparison than

symmetric resonances.

3.2 Cumulative probability plot

Fitting programmes such as SIOB in use at Oak

RidgeCiS) are not always available for the determination of the position of the peak of a resonance in

energy or time. When the observed shape of a resonance is Gaussian, or at least close to symmetric,

the error in deducing the position of the resonance peak can be considerably reduced by the use of a cumulative probability plot (ogive curve) in which,

after the subtraction of a suitable background

representing the neighbouring value of the data, the

data (cross section, transmission or observed counts

per timing channel) are treated as probability values

and plotted on probability graph paper against timing

channel number. The linear portion of the graph is

fitted by the least squares method to derive the

timing channel corresponding to the centre of the resonance at 50% probability. Half a channel must be

added to the result obtained because of the binning

of data into channels. This technique readily allows

peak positions to be estimated to within an error of

0.1 channels or less and an example is shown in fig. k. The extent to which the resonance shape is not normal

is immediately apparent. Only the linear portion of

the curve is fitted and the error derived reflects

any departure from normality.

3.3 The &L/At method

In deriving neutron energies by the direct method

set out in sect. 2.2 and in Appendix 1, it will be seen that several of the quantities involved can only

be estimated with relatively large uncertainties*

These quantities are the distance D traversed by a

neutron within the detector before detection, the

energy, transit distance and transit time, tj, of the

fast neutrons which generate the neutrons under

consideration, the slowing down time delay th. and any

differences, caused by differences in pulse 'shape,

between the detection of neutrons and coincident

gamma rays. It is shown in Appendix 1 that all these

quantities can be removed from the equations used to determine energy if the neutron flight times t-j and

are measured with the same detector at two path

lengths Li and li^' However, a disadvantage of the

method is that the effective path length is reduced to AL = L^- Lg. In principle the method allows (t^ + t/^. ) to be measured but only at the expense of re-introducins

the poorly known quantities D, L-jf1 , L'j(2 andS. However,

by careful design, such as the use of thin detectors

to reduce D, improved measurements of (tj + ti,.) could

be made.

3.'+ y2 - H Method

This method, described by Meadows^''^\ for

reducing the error in determining the energy at the

threshold of a (p,n) reaction has been described in

sect. 2.3.

3.5 Effect of counting statistics on the energy

uncertainty

In their measurement of the Li total cross section peak energy, the statistical quality of the

data was not good but James et al. ^''^'were able to

321
derive accurately the error in determining the peak

energy caused by the errors in the counts per timing

channel. This was done using random number (Monte

Carlo) techniques to vary each measured transmission

datum by an amount controlled by the normal law of errors and by the standard deviation of each datum.

The data set obtained in this wa/ was fitted and another

estimate of peak cross section energy obtained. This

process was repeated ten times to get a measure of

the spread of the values derived. In this numerical

experiment, uniform pseudo- random variates were con- verted to normal variates using the simple but exact

Box-Miiller transformation^'' 9 •'. It was found that even

for the broad resonance near 250keV in Li the peak

could be located to an error of 0.33keV.

k. Experimental determination of systematic errors

When a constant has been measured in several

indepeadent studies, it becomes necessary to undertake some data evaluation with the aim of determining

a best value and of setting some limit to the error in

this best value. In a series of papers which are models of clarity, W.J. Youden*'''''^ has shown that the

aim should be to establish 'enduring values' which are best values with an error range within which future

best values will lie. By examining fifteen values of

the astronomical unit measured between l895 and I961

and 21 measured values of the velocity of light,

Youden shows that, as so often happens, the data differ

from one another by more than would be expected from

the ascribed estimates of uncertainty. He attributes

this to a poor assessment of systematic errors and

argues that the only worthwhile estimates of systematic errors are those which are made experimentally.

To achieve this it should be noted that when an experiment is done at another laboratory everything

gets changed whereas an investigator at one laboratory makes only minor changes to his equipment.

Youden argues elegantly for making a direct experimental assessment of systematic errors by planning

experiments in which everything which could make a

difference to the experiment, and even a few things

which are regarded as not affecting the measurement,

is changed. Often the things or quantities changed

will have only two descriptions or values but the

benefit in bringing systematic errors to light will

be immense. The time devoted to the experiments need

not increase inordinately. Clearly, more will be

learnt by making six changes and accumulating data

for a sixth of the time after each change than if no changes are made. Youden also clearly shows that if

the changes are properly planned a large reduction in

the error in the mean can be achieved by changing more

than one thing at a time. Many papers on the way such

experiments should be planned are now available, some

in the collection of Ku(20),

5. Comparison of data for certain resonances

In this section the data available on single resonances in Li, Na and C and five resonances in

23ou are presented. Average values derived by taking

the best value from each laboratory are given and

used to produce a comparison factor G which equals

the mean value divided by the error in the mean. The resonances selected differ markedly from one another.

The resonance near 250keV in Li is broad and will not be

used as an energy standard but it is interesting to note the accuracy achieved for the quoted peak cross section energy. This resonance also shows the

importance of measuring a well defined quantity such

as the energy at the maximum of the cross section

rather than the value of the 'resonance energy' which

appears in the theoretical expression for the cross

section. The resonances in Na and ^^°U are s-wave

resonances, which in the case of Na have been

corrected by Derrien for the effects of instrumental

resolution, whereas the resonance in carbon has 1^2

and is symmetric.

(: 5. 1 The resonance in "Li total cross section near 250keV

c Values of the energy at the maximum of the °Li

total cross section near 250keV are available from

four white source time-of- flight measurements and

from four mono-energetic measurements on Van de

Graaff accelerators. The data are presented in

table 3 and illustrated in fig. 5 in chronological

order. The data of Harvey and Bockhoff have been

reanalysed using the formulae used initially by

Uttley and then by Jajnes so that there is no difference in the method of analysis for the four time-offlight values quoted. The average of these values

is 2^t'+.0 ± 0.5keV corresponding to G = k&S. This

contrasts with the four mono-kinetic beam Van de

Graaff measurements all of which lie at or above

244keV. In fairness to the early measurers it must

be stated that the energy at the peak was not a prime

consideration in these papers and no errors are quoted

on the values given. The errors of ilkeV given in the

table are judgements made by experienced workers.

The value of 0.33keV given by James et al.^''^'' is

derived from a careful assessment of all the errors involved in their measurement including, as described

in sect. 3-5 above, the effect of limited counts per

timing channel on the energy derived. It is likely

that an experimental determination of systematic

errors as advocated in sect. 4 above would lead to an upward revision of this error.

Recently a value, given preliminarily as 2^2 - 2keV, has been measured by time-of- flight on a Van de Graaff in an experiment which allows the

succeeding gamma flash to fall at the peak of the

6Li resonance. If confirmed, this would reduce the

average of all measurements made by the time-of- flight

method of 2^3.60 t 0.58keV.

Table 3

Neutron energy at the maximum of the ^Li total

cross section

Author

Hibdon & Mooring(2l)

Farell & Pineo(22)

Meadows & Whalen^23)

Uttley(2'+)

James et al. (15)

Harvey et al.(25)

Harvey (James)

Bockhoff et al.^9)

BCckhoff (James)

Knitter(26)

Average of values marked

Year

E(max)

(keV)

1968

1972

1975

1975

1975

1975

1975

1975

1976

.6

.5

2k6

250.

252.

2^+3.

*2't2.71 246

•2H.8

245.0

*245.0

247.0

2kk.O

10.33

1

1

1

13

0.5

5.2 'The resonance in ^^Na at 298.330 - 0.082keV

As discussed in sect. 3»1, the data available on

the observed peak energy Em for the s-wave resonance near 299keV in 23^^ have been corrected by Derrien^'''''' for the effects of instrumental resolution. The

values of Em and the values of which would be

observed with perfect resolution are illustrated in

322
fig. 3 and listed in table 'f. The mean of the three

values with the best energy resolution corresponds to

Table k

Observed and corrected peak energy values for

the 298.33keV resonance in Na

Origin %(keV) E^(keV)

Saclay II

+ 0.37

Karlsruhe 299.26 + 0.20

Columbia 298.5 + 1.0

Harwell S.C. 299.2 + 0.2

Cadarache 302 + it Saclay I 303 + 3

Average

Average of values marked

298.39

*298.66

297.35

*298.6

299.

299.5

298.65

298.550

0.37

0.20

1.0

0.3^+ h.O

3.0

0.32

0.082

5.3 Five resonances in 238u

It is suggested in sect. 6 that resonances in

238u could be selected for use as standard over the

energy range 6eV to 3keV. The data available on five of these resonances are presented in table 6

.

All the measurements made at the Harwell synchrocyclotron used a Li-glass detector. A 50m measurement was made in August 1976. For other reasons,

the flight path was then raised by 3cm and the source geometry was altered so that the water moderator

stood in front of the proton target instead of underneath it. Energy measurements of 23og resonances

were then made, in November 1976, both at 50m and

at 100m. Results were also derived using the AL/At

method discussed in sect. 3.3. The /iL./4t method is

based on the same data as the two other results

obtained in November and the correct method of

deriving the best value aind an error in this best

value remains to be formulated. For the present,

the four results obtained for each resonance are regarded as independent and the average value is

given. However, the error quoted in the average

value is that for an individual reading derived from

the spread of the measurements. Results from Oak

Ridge for three of the resonances were not available

to me when the table was prepared. An unweighted

average from all laboratories together with an error

in the mean derived from the spread is also given.

The agreement between Harwell and Oak Ridge is well

within the quoted errors and all the Harwell values

are within the quoted range of the average from all

laboratories. The discrepancy between Geel and

Columbia has already been aoted^^' •'. The results

for the resonance at 2489.^7 * 0.5eV are illustrated

in fig. 6.

The resonance in carbon at 2078.O5 - 0.32keV

The data available on the energy of the resonance

in the total cross section of carbon near 2078keV

are presented in table 5 • Two measurements were made on the Harwell synchrocyclotron in August 1976 one using a Li-glass detector at 50m and one using

an N3110 detector at 100m. The value obtained from

the measurements by the A.L/At method are also given

although, with two dissimilar detectors, the

'distances D travelled by neutrons within the

detectors are not eliminated from the equations.

Source distance uncertainties are, of course, still

eliminated. The measurements of Heaton et al. and

of Meadows were made on Van de Graaff accelerators

by the mono-energetic beam technique. All the other

measurements are made by neutron time-of- flight. The

excellent agreement between the two methods is clear.

The three central values are quoted more precisely

than the others. Taken alone these values have an average of 2078.11 1 0.l6keV corresponding to G

greater than 10 000.

Table 5

Measured peak energy for the resonance in carbon

at 2078.05 - 0.32keV

Reference

Harwell ^Om

" 100m

" AI/At

" Average*

Davis & Noda^?)g.

Heaton et al.

James

Meadows^''^^

Perey et al.^29)

Bockhoff et al. ^9)

Cierjacks et al.^^OJ

Average* *

Date

Aug 1976

Aug 1976

Aug 1976

1969

1975

1977

1977

1972

1976

15

Energy - keV

2079.2 ±1.1

2078.31 - O.itif 2077.^5 - 0.8k

2078.33 ± 0.89

2079

2079

2078.33

2078.2

2077.8

2077

2077

2078.05 ± 0.32

*Error in individual reading derived from spread of

the data

''*Error in the mean derived from spread of the data

6. Resonances for use as energy standards - a selected list

To compare the energy scales of different

spectrometers and thereby help to establish accurate

energy standards it is necessary that those concerned

with this task should all make measurements on the same set of resonances. To promote this development

the INDC set up a sub-group with the task of producing

a list of suitable resonances. Seventy-six narrow resonances were suggested in the energy range 0.6eV

to 12.1MeV. This list is probably too long and I have reduced it by selecting about six resonances in

each decade which have the smallest value of

(r +A)/E, Here V is the resonance width, A is the

Doppler width and E is the resonance energy. The

reduced list of forty-three resonances is presented

in table 7. The resonances are all in the total cross section of the eleven elements listed. Apart from

sodium and iridium, all the elements are commonly

occurring, readily available and easily fabricated

into suitable transmission samples. The energies

given are listed as nominal energies to emphasise the

fact that they do not represent evaluated data.

Nevertheless the best values of energy available are given wherever possible. No errors are given for the

U-238 resonances of ref.(l2) but it is known that the errors are likely to be of the order of 1 in 5000-

The distribution of the resonances listed is shown in

fig. 7. It will be seen that the largest gaps on a logarithmic energy scale are between 0.6eV and 5eV

and between 6keV and 30keV. It may be necessary to augment the list within these ranges.

As more data become available more stringent

evaluation of the energies of the resonances listed

in table 7 will be possible. Evalnators may then

reasonably demand that the only data to be considered

are those for which full details of all steps in the

323
Table 6

Measured peak energies for five resonances in ^^^U

Origin

Harwell 50m Aug 1976

" 50m Nov 1976

" 100m Nov 1976

" Al/At Nov 1976

Harwell average*

Oak Ridge('^2J

Geel^27;

Columbia^''°)

Average**

G

*Error quoted here is that in an individual reading derived from the spread of four values

**Error in the mean derived from th^ spread is quoted

Energy -- eV

1 if 5. 63^ + 0.037 463.23 + 0.12 708.44 + 0.19 1420.12 + 0.045 2490. 16 + 0.40

145.578 + 0.051 462.93 + 0.15 708.62 + 0.27 1419.56 + 0.35 2489.96 + 1.1

1^5.593 + 0.033 463.09 + 0.12 708.09 + 0.13 1419.80 + 0.34 2489.26 + 0.79

1 if 5. 606 + 0.071 463.24 + 0.25 707.59 + 0.29 1420.02 + 0.46 2488.61 + 1.54

145.603 + 0.024 463.12 +

0. 14 708.18 + 0.45 1419.88 + 0.25 2489.50 + 0.71

145.68 463.62

1419.76 ± 0.19 2489.18 ± 0.43

+ 0.10 + 0.20 708.59 + 0.25 1420.7 + 0.3 2490.8 + 0.4

l'+5.57 + 0.15 462.8 + 0.4 707.9 + 0.4 1419.2 + 0.3 2488.4 + 0.7

145.617 + 0.033 463.13 + 0.24 708.22 + 0.20 1419.88 + 0.32 2489.47 + 0.5

44l2 1929 3541 4437 4978

energy determination are published. At present, such

a demand would mean that almost no resonance energy

data could be evaluated.

7. Conclusion

Ihis report has reviewed the methods of neutron

energy determination in such a way as to indicate the sources of error, and methods whereby some of the errors

can be reduced have been described. The errors quoted on published energy values are often derived

from reasoned judgements by experienced scientists.

Better estimates of the systematic errors would

clearly be derived by adopting the suggestion made

by Youden that as many experimental components as possible should be deliberately varied preferably

more than one at a time in a planned way. From a comparison of the data available on certain resonances it appears that at best energy measurements from different laboratories agree to a fraction

approaching one in 10 000. Furthermore, the work of

Meadows has shown that with the development of careful methods there is no discrepancy between mono- energetic and white source measurements. A list of

forty-three narrow resonances suitable for energy

comparison and calibration has been drawn up which

may encourage experimentalists to measure the energies

of the resonances listed both as a means of improving

energy standards and also as a means of keeping a continual check on the energy scales of their

spectrometers. The confirmation of at least one energy from a list such as that in sect. 7 should

be encouraged whenever changes are made which could

result in a changed energy scale. Very few papers

have been published giving the full details of

energy determination which could reasonably be

demanded by an evaluator. It is hoped that this

situation will change and that soon energy values

from fully documented published sources will be

available for all the resonances used in energy

int ercompari sons

.

Acknowledgements

I am grateful to the members of the INDC subgroup on Neutron Energy Calibration, J. Boldeman,

F. Voss, F. Corvi, J. A. Harvey, J. Lachkar and

A. B. Smith, for their invaluable advice, for their

suggestions on suitable narrow resoaances and for

many of the energy values quoted in this paper.

However, the sub-group has had no opportunity to comment on the reduced list presented in sect. 7 and

it does not carry their imprimatur. Both

J. A. Harvey stnd K. H. BBckhoff readily sent me their

data on- the total cross section of -"Li so that all

white source m'easurements could be analysed by the same method. The help given by D. K. Olsen in

selecting suitable resonances in 23o'j greatly

appreciated. His list was, however, augmented

slightly and the responsibility for this rests with

me.

It is a pleasure to acknowledge the interest

taken in this work by Basil Rose who in the course

of lively discussions suggested the use of the ogive

curve method and stimulated the development of the

4L/At method. Experiments on the synchrocyclotron

eajoy the unstinting support of Colin Whitehead and

the active participation of P. H. Bowen, A. D. Gadd,

D. B. Syme and I. L. Watkins. Many of the energy

calculations were carried out by A. D. Gadd.

References

1. S. Wynchank, F. Rahn, H. S. Carmada, G. Hacken,

M. Slagowitz, H. I. Liou, J. Rainwater and

W. W. Havens, Jr. Nucl. Sci. & Engng.51 (1973)

119

2. J. C. Davis and F. T. Noda. Nucl. Phys. A134

(1969) 361

3. M. S. Coates, D. B. Gayther and N. J. Pattenden.

Proc. 4th Conf. on Neutron Cross Sections and

Technology, Schrack and Bowman (Eds.) NBS Spec.

Pab. 425 Vol. 2 (1976) 568

4 J. W. Behrens and G. W. Carlson. Proc. NEANDC/

NEACRP Specialist Meeting on Fast Neutron Fission

Cross Sections, ANL-76-90 (ANL, Argonne, 1976)

47

5- S. Cierjacks, B. Leugers, K. ICari , B. Brotz,

D. Erbe, D. Grbschel, G. Schmalz and F. Voss. Ibid. 94

324
Table 7

Narrow resonances suitable for use as energy

standards

Energy Range Isotope Nominal Energy Order* Reference

(eV)

0.1-1 Ir-191 0.6551 O.OOI^t 1 13

1 - 10 U-238 6.672 1 12 10 - 100 U-238 10.236 5 12

U-238 20.86't k 12

U-238 36.671 3 12 J-238 66.015 2 12

iJ-238 80.729 1 12

100 - 1000 U-238 1^5.617 + 0.033 7 Table 5

U-238 8 12

U-238 311.18 - 0.25 5 13

U-238 397.58 12 U-238 it63.i8 + 0.2k 3 Table 5 U-238 619.95 2 12 U-238 708.22 + 0.20 1 Table 5 U-238 905.03 6 12

1000 - U-238 1 '419. 88 + 0.32 7 Table 5 10 000 U-238 5 12

U-238 2^+89. ^+7 + 0.5 3 Table 5 U-238 2672.2 h- 12

Pb-206 3360 + -10 8 13

U-238 3^+58.

1

6 12

U-238 it512.0 2 12 U-238 5650.6 1 12

Al-27 5903

+ 8 9 13

(keV) (keV)

10 - 100 S-32 30.378 + 0.006 3 12

Na-23 53.191 + 0.027 6 12

Si-28 67.73 + 0.02 5 31 Pb-206 71.191

+ 0.018 k 12

Fe-56 90.13^

+ 0.016 2 32

S-32 97.512 + 0.028 1 12

100 - S-32 112. 186 + 0.033 3 12 1000

fe-56 266.3^+7 + 0.053 2 32

S-32 1 31

S-32 818.7 6 31 1000 - 0-16 1651

+

2 6 13

10 000

C-1Z 2078 5 33

G-12 2818 +k 3 13

0-16 3211.1 ± 1.5 1 13

C-12 6293 + 5 2 13

10 000 - C-12 12100 + 100 1 100 000

Order of ( H +L)/E within an Energy Range

6. P. A. R. Evans, G. B. Huxtable and G. D. James,

Ibid. Iif9

7. M. S. Coates - private communication 1975

8. J. W. Meadows, Ibid. 73

9. K. H. Bockhoff, F. Corvi and A. Dufrasne,

NEANDC(E) 172 "U" Vol. Ill- EURATOM (1976) 17

13. S. F. Mughabghab and D. I. Garber, BNL325,

Third Ed. Vol. 1 (1973)

ih. H. J. Groenewold and H. Groendijk, Physica _1_3

15. G. D. James, D. B. Syrae, P. H. Bowen,

P. E. Dolley, I. L. Watkins and M. King, Report

AERE-R7919 (1975)

16. J. W. Meadows, Report ANL/NDM-25 (1977)

17. H. Derrien, Report NEANDC(E)-l64-L

18. G. de Saussure, D. K. Olsen and R. B. Perez,

(unpublished)

19. G. E. P. Box and M. E. Miiller, Ann. Math.

Statistics 29 (1958) 610

20. H. H. Ku, NBS Special Pub. 300 (1969)

21. C. T. Hibdon and F. P. Mooring, Neutron Cross

Sections and Technology, NBS Special

Publication 299 Vol. 1 (1968) 159

22. J. A. Farell and W. F. E. Pineo, Ibid. 153

23. J. W. Meadows and J. F. Whalen, Nuc. Sci. & Engng. k2_ (1970) 351, i+S (1972) 221

2.k. G. A. Uttley - private communication

25. J. A. Harvey and N. W. Hill, Nuclear Gross

Sections and Technology, NBS Special Publication

^+25 (U.S. Dept. of Gommerce, NBS, Washington,

1975) 2kh

26. H. H. Knitter . C. Budtz-J^rgensen, M, Mailly

and R. Vogt Report EUR-5726e (1977)

27. F. Corvi - private communication and ref. (9)

28. H. T. Heaton II, J. L. Menke, R. A. Schrack and

R. B. Schwartz, Nucl. Sci. & Engng. 56 (1975) 27

29. F. G. Perey, T. A. Love and W. E. Kinney, Report

0HNL-'f823 (ENDF-178). The value quoted is

derived from the data described in this report.

30. S. W. Cierjacks et al. Report KFK-1000 (1968).

The value communicated by F. Voss is based on the

data in this report.

31. J. A. Harvey - private communication and Report

ORNL/TM-5618

32. Unpublished results from the Harwell synchrocyclotron to be regarded as preliminary.

33. H. Weigmann, R. L. Macklin and J. A. Harvey,

Phys. Rev. C ^h. (1976) 1328, with errors provided by J. A. Harvey - private communication.

10. F. J. Rahn, H. Gamarda, G. Hacken, W. W. Havens,

Jr., H. I. Liou, J. Rainwater, M. Slagowitz and

S. Wynchank, Phys. Rev. G6 (1972) 185't

11. W, J. Youden, Technometrics _lA- (1972) 1

12. D. A. Olsen, Unpublished data communicated by

J. A. Harvey

325
Appendix 1

DETERMINATION OF NEUTRON ENERGY

DIRECT METHOD

L = P + D

t = t - t

n n

Fast

Neutron

Source ^9o

V =-n D

(tn - t ) + (t - t ) - (t - t

(x - X ) w + L /O. 29979 - (fast transit time

o y + slowing down time delay)

L = P + D + S

Y Y Y

g = L/ (t X 0.29979)

E = 469776.3 (g^ + 0.75B^)

AL/At METHOD

Perform experiment at and

= P^ t D

t^ = (X^ - X^^ +6)w + L^j^/0. 29979 - (t^ + t^)

= P^ + D

to =

(>^o - + 6)w + L /0. 29979 - (t + t ) 2 2 02 y2 3 4 L=P, + D+ S • L =P +D+S

Y1 1 Y Y ' y2 2 Y Y

AL = - P^

At = CX^ - X^^) w - (X^ - Xq2' " + - P2)/0. 29979

B = AL/ (At X 0.29979)

E = 469776.3 (B^ + 0.756'*)

Note: 6 D D S and t^ + t, are removed

Y Y 3 4

Moderated

Source

Detector

Fig. 1 Arrangement of target and detector in the

Harwell synchrocyclotron neutron time-offlight spectrometer illustrating some of

the symbols used in Appendix 1.

,8

z

o

^ fi

(/) z

<

a:

h- .4

.2

• ,S ACLA Y II

SACLAY I

i i'

." i

1-

9

»

290 300 310

ENERGY keV

Fig. 2 The transmission of sodium near the resonance at 299keV measured with good

resolution, Saclay II, and poor resolution,

Saclay I, as presented by Derrien^'^''^^. . It illustrates the effect of resolution on the

observed energy of the transmission maximum

and minimum.

326
I Sodium Resonance at 298-65 ±0-32 keV \i Resonance at 244 0 ±0-5 keV

. --0-

-O- X -O- X o—

X

-X

Saclay I Cadarache

Karlsruhe

Harwell

Saclay II Columbia

Average

Hibdon & Mooring

Farell & Pineo

Meadows &Whalen

Uttley

James

Harvey

Harvey - James

Bockhoff Bockhoff -James

Knitter

295 300

E - keV

Fig. 3 Ihe energy of the sodium resonance at 298.65 - 0.32keV as measured in six laboratories (X) and the values obtained by

Derrien(l7) after correcting for the effect

of spectrometer resolution (O).

Cumulative Probability

U-238 Resonance at 2489eV

240 250

E - keV

Average of TOF values

Fig. 5 Energies at the peak of the ^Li total cross section near 250keV.

(James) denotes results obtained after

fitting the data by the formulae used by

Uttley and by James. The average of results

obtained by time-of- flight experiments is

Zkk.O ± 0.5keV

N = 13754-99 ±0-12

U-238 Resonance at 2489-47 ±0-50eV

13760^

N

13750.

(a)

(b)

50m Aug'76 50m

100m Nov '76 dL/dt

Average

Bockhoff James

Olsen Rahn

10 50

PROBABILITY

90 99

Fig. k Cumulative probability distribution for the

transmission of a resonance in U-238 at 2^89eV. A least squares fit to the data

plotted on probability graph paper enables

the resonance timing channel N to be located

to 0. 12 units.

2487 2490

E -eV

Average

Fig. 6 An illustration of the data for the resonance in the total cross section of

23°U at 2h&9.k7 ± 0.5eV presented in table 6

327
Energy Distribution of Selected

Resonances

_1 1 I I—Ul I I I I I I I I II I I II L I II I I ll II E-eV

U U U UUPbUAl SI Na SI Pb F* S \\\\\

ll I I I II I II I II I I I l> I I I 1 I I i ll I H I I

1 W 1 E- keV

S F*S SOCOC C

U I L j I I 1 1 1

1

I I II I I I 1 1 1

1

E-MeV

Fig. 7 Energy distribution of narrow resonances selected as suitable for use as standards.

The largest relative gaps are between 0.6eV

and 6eV and also between 6keV and 30keV.

328
NEUTRON TRANSPORT CALCULATIONS FOR THE INTERMEDIATE-ENERGY STANDARD NEUTRON FIELD (ISNF)

AT THE NATIONAL BUREAU OF STANDARDS

C. M. Eisenhauer and J. A. Grundl

National Bureau of Standards

Washington, D.C. 20234

and

A. Fabry

C.E.N. - S.C.K.

2400 Mol, Belgium

The intermediate-energy standard neutron field (ISNF) established in the thermal column of

the NBS reactor is designed to produce a benchmark neutron spectrum concentrated between 10 keV

and 3 MeV. Design principles and physical description of the spherically-symmetric system are summarized. Neutron transport calculations of the ISNF spectrum at the center of the facility

by the discrete-ordinates method are discussed and results are given for 40-group and 240-group

computations. The sensitivity of the calculated spectrum to variations in important physical

parameters such as material densities and carbon and boron-10 cross sections is explored.

(Benchmark spectrum; discrete ordinates; intermediate energy; measurement assurance; neutron

standard; reaction rates)

Motivation and Design Principles

The Intermediate-energy Standard Neutron Field

(ISNF) is an irradiation facility at NBS designed to produce a strong component of neutrons in the energy

range of interest for fast neutron reaction rate

technology. It is intended for use in the following

measurement activities.

1. Measurement of integral fission and capture cross sections.

2. Long-term measurement assurance for neutron

reaction rates required for the design and operation

of power reactors.

3. Calibration of passive neutron detectors used

in materials neutron dosimetry to characterize neutron

fields associated with radiation effects testing.

4. Validation of neutron spectrometers and other

energy-sensitive neutron detectors.

The ISNF arrangement is simple: a spherical

cavity in^^he themal column of the NBS reactor, a thin

shell of B lightly suppor^^^ at the cavity center,

and fission-source disks of U placed symmetrically

around the periphery of the cavity. The ISNF field at the center of the cavity consists of a fission-neutron

component and a component of neutrons returned from

J^e graphite, both of which are attenuated by the

B shell. The general properties of the neutron

field at the center of the ISNF are as follows

:

Total neutron flux

Typical maximum fluence

Neutron field gradient

Spectrum:

Average energy

Median energy

7 X 10 n cm

-.14 0.5 X 10^^ n cm

2% over 4 cm

1.0 MeV

0.56 MeV

sec

2

Energy range 90% between 8 keV and 3.5 MeV

Neutron transport in the ISNF is determined almost

entirely by the ^35^ fission neutron spectrum and by

neutron scattering in carbon and neutron absorption in

boron-10. In particular, the neutron spectrum at the

center of the facility is governed by the kinematics of

elastic scattering in carbon and by two accuratelyknown cross sections: O^.^^. for carbon and o^j^^ for

boron-10. Moreover, only the energy dependence of the

boron-10 absorption cross section is critical because

the neutron transmission of the shells is checked

experimentally with a 2-keV monoenergetic neutron beam.

The one-dimensional geometry of ISNF permits calculation

of the spectrum with an accuracy limited only by the

uncertainties of input cross sections and densities of

materials, and allows investigation of (1) sensitivity

of the spectrum to variations in system parameters and

(2) perturbation of the spectrum due to extraneous

structural materials.

Physical Description of Facility

The initial ISNF facility has been set up in the

graphite thermal column of the NBS reactor. A square

opening, 30 cm x 30 cm provides access to the center of

the graphite column. Split graphite blocks enclosing

the ISNF cavity and containing cylindrical access penetrations may be inserted through the biological

shield of the reactor and into the thermal column

opening . A detailed schematic cross section of the ISNF

arrangement within the cavity is shown in Figure 1. Three levels of access are available: (1) instruments

and/or irradiation samples may be inserted or removed

through a 5 cm dia. cylindrical channel; (2) the

boron-10 shell and the graphite pieces to which it is

attached slide in a 20 cm dia. cylindrical access; and

(3) the entire 30 cm x 30 cm cavity block may be withdrawn or inserted with the reactor at full power. Four

O Q C of the eight ^'-'U fission source disks are indicated in

symmetric array near the surface of the cavity in

Fig. 1. Each disk is made of enriched uranium metal

16 mm dia. x 0.15 mm thick (93% enrichment, 0.476 or 0.511 gm each) and positioned at 1 cm from the surface

of the 30 cm dia. cavity. Eight disks are used in

normal operations producing a total fission neutron

source strength of about 6 x 10-'--'- sec"-'".

Figure 1. Intermediate-Energy Standard Neutron Field

arrangement at the center of the NBS Reactor

thermal column.

329
The crucial component of the ISNF is the boron-10

shell and serious difficulties were encountered in its

fabrication. Two shells of different thickness, formed

as stepped hemispheres at the Los Alamos Scientific

Laboratory [1] by techniques of powder metallurgy, are now available. The shells are 14.26 cm outside diameter.

The thick, and thin shells have thicknesses of 1.293 cm and 0.638 cm. The mean boron-10 thickness of each

shell is 1.26 and 0.61 g/cm . Fabrication of the shells

required that a significant amount of aluminum be added

to the 95%-enriched boron-lQ metal powder.

TABLE I Chemical and Isotopic Analysis of Boron

and Aluminum Powders [1]

Constituent Concentration (weight%)

Boron-10 Aluminum

aluminum 99.65

boron-10 93.58

boron-11 4.87

carbon 0.6

silicon 0.4 0.15

iron 0.3 0.2

lead 0.1

uranium 0.15

(99 .y/, 0.6% ^-^^u)

Table I presents an analysis of the boron and aluminum

metal powders employed. Materials other than "'"^B in

and around the shell, including a protective shell of

0.7 mm thick spun aluminum metal, add up to 29 atom

percent of aluminum and 4.8 atom percent of boron-11.

Rings and support rods for mounting the shell in the

cavity introduce another 1.4 atom percent of aluminum

into the cavity. The uniformity and absolute absorber

thickness of the shells is determined by means of

neutron transmission measurements in the 2 keV filtered

beam at the NBS reactor. Transmission measurements in

the 2 keV beam for several orientations of the shell

indicate that the thickness is uniform to better than 1%.

The flanged cadmium shield indicated in Figure 1,

which fits over the stems of instruments placed inside

the shell, prevents streaming of thermal neutrons

through the hole in the shell access plug. Such streaming, particularly by epicadmium wall-return neutrons, is

a characteristic problem for any driven, one-dimensional

neutron field. Careful experimental investigation is

required to demonstrate that the problem is under

control. Shell access plugs without a hole are available to investigate this effect with passive detectors

which do not involve instrument stems.

An inventory of ISNF components and their physical

properties is given in Table II. Included are some characteristic nuclear parameters appropriate for the

fission neutron transport problem.

TABLE II Inventory and Physical Properties of Principal

ISNF Components

Graphite Thermal Column —dimensions

:

—graphite density: —cavity diameter:

Boron-10 Shells [1] —dimensions: Thick

inside diameter: 11.68 cm outside diameter: 14.26 cm thickness 1.293 cm

1.4 X 1.3 X 0.94 m" 1.71 g/cm3

29.8 cm

—densities

:

boron-10

boron

aluminum

0.975 g/cm;

1.025 g/cm;

0.992 g/cm"

Thin

12.99 cm

1A.26 cm 0.638 cm

0.954 glcm

1.002 glcm

0.971 g/cm-^ —thickness in mean free paths Z t for neutrons

(thick shell only; for thin sheSl multiply by

0.49)

1 MeV 0.1 MeV 25 keV 2.7 keV

absorption

scattering

.025

.30

.145

.33

.29

.20

235,

.88

.23

3 . Fission Source Disks (93.5% Enriched ""^^U Metal) —dimensions: 16 mm dia. x 0.15 mm thick

total fission neutron source strength (8 disks,

reactor at 10 MW)

<; mil -1 ~ 6 X 10 sec

Calculation of Central Flux Spectrum (ISNF-1)

The main method for calculating the flux in the

ISNF facility has been the discrete ordinates method.

We have used the ANISNW version of this code which

includes revisions by the Westinghouse Astro-nuclear

Laboratory [2]

.

In making one dimensional discrete ordinates

calculations of the ISNF system, certain schematizations

must be made. Figure 2 gives the parameters which were

used in the calculations for the thick shell. The

linear diagram at the top shows the radial coordinates

of ISNF-1. The l^B-Afc shell has inner and outer radii

Vacuum

B-A£ A£

Vacuum

Carbon

r(cm) 0 5^838 7.^^131' 2005

10„

.0582 at/(b-cm); N^^ = .0602 at/(b-cm)

N^j^ = .0216; N^ = .0038

13.^914.0

Source N^ = 0.0860 at/(b-cm)

Adjustments made in Shell Composition to preserve Total Mass for Calculation in Spherical Geometry

lSNF-1 (thick """^B shell)

A£ density = .0021 x .9916 = .9838 g/cm^ (1.27 g/cm^)

1°B density = .9747 x .9916 = .9665 g/cm:: (1.25 g/cm^

Other = .0762 x .9916 = .0756 g/cm (0.10 g/cm ) Total density 2.0259 g/cm^

Volume = 4/3 it [(7.131) - (5.838) ] = 685.5 cm Total Mass = 1388.8 g

Protective AJ!. Shell

Thickness t = 0.0695 cm Inner Radius = 7.131 cm

^ Density = 2.70 g/cm. , Mass = 4/3 TT [(7.2005) - (7.131) ] 2.70 = 121 g

Figure 2. Parameters used in the one-dimensional discrete ordinates calculation of the ISNF-1 configuration.

330
of 5.838 cm and 7.131 cm. This is surrounded by an AH

protective shell with an outer radius of 7.2005 cm. The source was assumed to be distributed in a 1 mm

annulus with inner radius of 13.9 cm. This assumption

is justified by the fact that the field at the center

of a spherically symmetric system is the same for a spherical shell source as for a number of point or thin disk sources placed at the same radius and having

the same total source strength. The radius of the

cavity is 14.92 cm and the carbon was assumed to extend out to a radius of 65 cm. Calculations show

that neutrons reflected from graphite beyond 65 cm would add less than 0.2% to the flux above 8 eV. The density of the various components in the B-A£

shell were adjusted from those provided by Los Alamos

Scientific Laboratory in order to preserve the total mass (1389 g) of the two 3-Al hemispheres plus the

heaviest of 3 access plugs. The outer radius of the

protective M shell was adjusted to maintain the

measured mass of 121 g.

Calculations with 40-group ENDF/B-III Cross Sections

Cross Sections

The cross sections used in calculations at NBS are based on ENDF/B-III data averaged in a 40-group energy

structure. This library was derived from a 100-group

set (the GAM-II structure) weighted by fine-group

fluxes determined from one dimensional discrete

ordinates calculations and averaged over spatial zones appropriate to the ISNF geometry.

Results

The three calculated spectra of interest at the

center of the ISNF are shown in Figure 3. The quantity

plotted is E(()(E) vs the log of the energy E. Thus

the relative area under each histogram for any energy

interval is proportional to the relative number of

neutrons in that interval. The dotted histogram shows

the ^-^^U fission neutron flux at the center of the

ISNF. It dominates the ISNF spectrum above a few MeV.

lO"^ lO"^ lO"" lO"' lo'^ lO"' I 10 NEUTRON ENERGY E.MeV

Fig. 3 Calculated central flux lethargy spectra for the

intermediate-energy standard neutron field

(ISNF-1) , the fission neutron source spectrum, and

the cavity spectrum with no '^B shell (ISNF/CV) .

The dashed curve shows the central spectrum (designated

ISNF/CV) resulting from the addition of neutrons

returned from the graphite. The spectrum below 1 MeV

is greatly enhanced by the neutrons slowed down in the

graphite, and below ~ 1 keV the spectrum extends smoothly

down to thermal energies with a near-l/E behavior.

This spectrum is realized for experiments by replacing

the ISNF -'-'^B shell with a cadmium shell. The solid line

shows the spectrum at the center of the ISNF facility,

with the thick shell in place (designated ISNF-1)

.

Here, neutrons below ~ 1 keV have been attenuated

severely by absorption in -'^B. The corresponding group

fluxes ([)(£) AE for the three spectra are given in

Table III. Detector responses in these three neutron spectra

have been calculated with the DETAN [3] computer code.

Calculated cross sections averaged over both ISNF spectra

are given in Table IV for some fission detectors.

Corresponding cross sections averaged over the fission

spectrum are also shown. Experimental measurements are beginning to be made in the ISNF-1 configuration.

Early results show an observed 235u/ " U cross section

ratio that is about 10% different from predictions of

ENDF/B-IV tabulations. These measurements are discussed

in another paper at this conference [4] . Similar

discrepancies have been found for other fast reactor

spectra.

Sensitivity Studies

Since the spectrum at the center of the ISNF is

established by means of calculation, it is essential to assess the sensitivity of both the spectrum and integral

detector responses to variations in physical and geometric parameters of the system.

B Absorption

The most important parameter for determining the

shape of the ISNF spectrum below about 10 keV is the

absorber thickness of in the B-Al shell. A given

uncertainty in either the '^B cross section <J(b/at) or the mass thickness X(at/b) produces a larger

uncertainty in the flux at low energies than at high

energies because of the exponential nature of transmission of neutrons through the shell. In the

calculation of ISNF-1 we have used a mass thickness of

1.25 g/cm2 of -'•'^B and ENDF/B-III absorption cross

sections. An uncertainty of 4% in the product of cross section and thickness produces an uncertainty of about

3% in the flux between 1 and 10 keV. With the 2 keV

beam-transmission measurement as verification of shell

thickness in mean free paths, the uncertainty in this

quantity is judged to be less than + 2%.

Graphite Density

Present measured values of the graphite density

exist only for removable thermal-column stringers,

because the bulk of the thermal column cannot be

^ disassembled. Results vary from 1.67 to 1.71 g/cm , while the nominal value used in ISNF calculations is

1.71 g/cm^. When the graphite density is increased by

3%, the flux in the cavity increases gradually from

0.1% at energies above 6 MeV to 2.7% at epithermal

energies. This can be interpreted as a nearly uniform

2.7% increment in the cavity-return component of the

flux. The corresponding increment in integral response

of a 235u fission foil is 1.4%, while that of a 238u

foil is 0.5%. Since a density change is equivalent to a change in total cross section, the error in the total

cross section of carbon, presently set at + 0.5%, would

account for proportionately smaller uncertainties in

flux and detector response.

331
TABLE III. Group Fluxes (})(E)AE at Center of ISNF

Upper Energy Fission ISNF/CV ISNF-1

Group Bound (eV) Spectrum / 1 ( )-n 1 1 1 \ (no -^"B shell) (thick -^"B she

1

1 A O O C A C 1 . 03250-5 1.08436-5 9.81571-6

Z D. 0653+06 4 . 47391-5 5. 39036-5 4.89605-5

J J . 6/ 00+06 / £ O A O T C 4 . 68987-5 7 . 01945-5 5.30993-5

/. H 2. 7/53+06 3 . 64821-5 7 . 34228-5 6.65930-5

5 2. 2313+06 9. 21736-5 2.00399-4 1.85559-4

0 1. 35J4+06 i . 22965-5 8.00682-5 7 .50595-5

7 1 . 1080+06 4 . 04492-5 1 . 17128-4 1.10525-4

oo o . 20o5+05 2 . 18288-5 1

.

32631-5 6. 86174-5

9 D. 7206+05 2 . 61317-5 1.03554-4 9.41721-5

lU 1 . 34049-5 6. 35855-5 5 . 70434-5

11 1. 51932-5 8 . 79786-5 8.00351-5

1 O LL •3 m n"7_i_n

c

1 . 04694-5 7 . 86573-5 7.08074-5

13 O T1"7T 1_AC 2 . 2371+05 4 . 84693-5 4.75933-5 4 . 16005-5

14 1. 8316+05 5 . 01832-6 6. 50755-5 5. 39525-5

13 1 . J569+05 O O *7 A O Zl ^ 2 . 27986-6 1 A / A 1 1 C 3 . 94921-5 3.39017-5

lb 1. 1109+05 2. 03273-6 4 . 57084-5 3.39677-5

17 8 . 6517+04 ^ 1 A / / A £ 1 . 39449-6 4 . 20754-5 3 . 54945-5

lo D. /joO-rU4 9. 57525-7 A A O A C C 3 . 90805-5 3 . 02092-5

ly C O / "7 C 1 A / 6 . 54398-7 3 . 65123-5 2 . 62202-5

20 4 . 0868+04 4 . 51031-7 3.43059-5 2.02771-5

21 3 . 1828+04 T A A C A "1 "7 3. 08527-7 3 . 23951-5 2 . 32677-5

22 1 / "T O O 1 A / 2.4788+04 2. 12484-7 3.07273-5 1.98570-5

23 1 A A C 1 A / 1. 9305+04 1.45919-7 2 . 92581-5 1.76938-5

24 T r A / 1 A / 1. 5034+04 9. 99573-8 2.79549-5 1.57295-5

25 1 T ~7 A A 1 A / 1. 1709+04 6. 87815-8 2 . 67891-5 1 . 40237-5

26 A 1 T O O 1 A T 9. 1188+03 7 . 96820-8 C A O A C O C 5.02952-5 2 . 28531-5

27 5. 5308+03 3 . 75963-8 4 . 69476-5 1. 78851-5

2o 3 . 3546+03 T "7 T T A A O 1. 77700-8 4 . 40797-5 1 . 31942-5

29 O A O / T 1 A O 2 . 0347+03 8 . 38826-9 4 . 15821-5 A 1 O A 1 C z' 9. 28015-6

30 "1 A O / -1 1 A 1 1. 2341+03 3 . 96618-9 3 . 93778-5 6 . 11757-6

31 7 . 4852+02 1. 87239-9 3 .74039-5 3 . 68933-5

32 4. 5400+02 8 . 84187-10 3 . 56270-5 1.97961-6

33 2 . 7536+02 4. 17368-10 3.40149-5 9. 11945-7

34 1 . 6702+02 t AT1AA 1A 1. 97399-10 3 . 25422-5 3 .42626-7

35 1. 0130+02 9. 31675-11 3.11880-5 9. 82393-8

36 6.1442+01 4.40192-11 2.99415-5 1.95149-8

37 3.7266+01 3.06191-11 5.70116-5 1.43028-9

38 1.371CH-01 4.64120-12 2.67281-5 6.66496-12

39 8.3153+00 4.10660-12 1.51313-4 2.32821-11

40 4.1400-01

TABLE IV Spectral Indexes for ISNF

and Fission Spectra

a(x)/o("\)

Reaction

239^

235

238.

237

Pu(n,f)

U (n,f)

Np(n,f)

U Fission

1.435

0.238

1.064

ISNF-1 ISNF/CV

1.113 1.205

0.0843 0.0111

0.480 0.0535

Source Radius

The flux at the center varies somewhat with source

radius. The main sensitivity occurs at high energies

where uncollided neutrons predominate. Flux changes at these energies are largely associated with the

inverse-r^ dependence. The variation below about

800 keV is generally less than 0.6% when the source radius of 13.95 cm is varied by + 2 mm. The normal

position of the sources is 1 cm from the cavity surface

with an estimated uncertainty of less than 2 mm.

Scattering Anisotropy in Graphite

The effect of the angular distribution of scattering in graphite was examined by repeating the ISNF

calculation with truncated V^, scattering cross sections. The ratios of fluxes calculated by truncating after rather than P^, as in the usual

ISNF calculations, indicate that the V-^ and P2

components are important in determining the intensity

of the scalar flux at the center. The ratio of fluxes

with truncation after P^ vs. after P^ remains

constant at about 1.06 at low energies and does not

reach 1.10 until a neutron energy of a few MeV is

reached

.

The ENDF/B-III Legendre harmonic coefficients for

elastic scatter in carbon, in particular the P-j^ component between 0.35 MeV and 1.35 MeV, which are responsible for lowering the flux of neutrons returned

to the center of the ISNF, compared with that obtained

for isotropic scatter, have not changed significantly in

the ENDF/B-IV tabulation. For example, the elastic

scatter in the backward direction o(-l) changed by

only 1% at an energy of 2 MeV. Therefore difference in

the ISNF central flux due to differences in scattering

cross sections for carbon in ENDF/B-III and ENDF/B-IV

are not expected to exceed 1%.

332
Aluminum in Boron Shells

The fabrication of the ''"'^B shells required that a significant amount of aluminum be included with the l^g

In order to assess its effect a calculation of the

spectrum at the center of the ISNF was performed with

all aluminum removed. Fluxes with and without aluminum

generally differ by less than 4% with the exception of

energy groups where aluminum resonances scatter neutrons

preferentially into an adjacent group. Variations of

the spectrum due to uncertainties in the amount of

aluminum actually present in the ISNF would be much

smaller. The effect will be discussed in more detail

in the next section.

Table V presents a summary of the effects of

varying the parameters discussed above.

TABLE V. Summary of Sensitivity Studies

Percent Effect

Group Fluxes Total Flux Cross Section a(n,f]

Parameter Variation 1-10 keV 10-100 keV 0.1-1.0 MeV 0.4 eV-18 MeV

235u 238y

Thickness + 4% - 3.3 - 1.4 - 0.5 - 0.8 - 0.9 + 0.8

Graphite

Density + 3% + 2.4 + 1.9 + 1.2 + 1.1 + 0.3 - 0.4

Source Radius + 2 mm - 0.3 + 0.2 + 0.0 - 0.2 - 0.1 - 0.6

Angular Distribution

of Scatter in Carbon P^ P

3 o + 6.2 + 6.7 + 9.5 + 8.3 - 0.6 - 0.3

Aluminum in

Shell

29 ^ 0 atom% + 1.9 + 1.8 + 0.1 + 1.0 + 0.2 + 1.2

Cross Sections and

Group Structure

ENDF/B-III IVI 40 -> 240 J+ 1.1 + 0.4 + 1.5 + 0.7 - 0.3 - 1.2

Results for 240-group Calculations

Two important limitations of the 40-group calculations are the accuracy of the ENDF/B-III cross sections and the coarse structure of the energy groups

in which the cross sections and the resultant spectra

are tabulated. In order to investigate these limitations calculations were performed by LaBauve and Muir

[5], using ENDF/B-IV cross sections in a 240-group

energy structure. They used exactly the same input

parameters for dimensions and material concentrations

as used in the 40-group ENDF/B-III based calculation.

Table VI shows a comparison of the two calculated spectra

in terms of the integral responses predicted for the

total flux and several fission detectors. The largest

discrepancy is 1.5% for a 238u fission detector. This

is a rather remarkable invariance since the 40-group

calculation contains only 4 groups in the energy range

238

in which 90% of the U response occurs. One reason

for the good agreement may be that the integral responses

are calculated with the DETAN code, which transforms the

histogram flux spectrum into a series of linear segments

with interpolated slope in f lux-vs . -lethargy

.

This comparison of coarse and fine group structure

is mainly a check on the group-averaging techniques

used in discrete ordinates calculations, since the

difference in the total cross sections of boron and

carbon cross sections for ENDF/B-III and ENDF/B-IV is

negligibly small. Significant differences in the

angular distribution of scattering in carbon exist in

the two data files, but they occur above 5 MeV where

the impact on the ISNF spectrum is negligible.

Figure 4 shows the 240-group central ISNF-1 flux

spectrum calculated by LaBauve and Muir. It is similar

to the 40-group spectrum shown in Figure 3, except that

the resonance structure due to the aluminum in the

10B shell is now resolved. The effect of this structure

TABLE VI Comparison of Calculated Integral Cross

Sections for 40-Group and 240-Group ISNF Spectra

Flux

235U(n,f)

238

U(n,f)

239

Pu(n,f)

237

Np(n,f)

a(b)

ENDF/B-III

NBS 40-Group

.0013749

1.6362

0.13798

1.8176

0.7834

ENDF/B-IV

LASL 240-Group

.0013850

1.6313

0.13632

1.8219

0.7855

Percent

Difference

+ .7%

- .5%

- 1.5%

+ 0.2%

+ 0.3%

ISNF CaiTRAL SPECTRIN

2140-GROUP CALCULATIOH

Fig. 4 Calculated Central flux lethargy spectrum for

ISNF using 240-group cross sections.

333
on reaction rates whose energy dependence is smoothly

varying in this region or for which only a small part

of the response occurs in this region will be small.

For example, the effect of the aluminum in the shell on

the total flux is 1%; the effect on integral cross sections for wide-energy-range detectors such as 235u(n,f) and ^^^Pu(n,f) is less than 0.2%. The effect

on threshold integral detectors such as U(n,f)

depends generally on the threshold energy. For

U(n,f). for which 95% of the response lies above

1.5 MeV, the effect of the shift in spectrum due to downscatter is 0.8%.

The effect of the aluminum on the flux spectrum

can be interpreted almost entirely as a redistribution

of elastically scattered neutrons in a localized energy

range. In order to demonstrate this hypothesis, the

detailed behavior of the ISNF spectrum near an aluminum

resonance was calculated in a simple single scatter

analysis. This was done by assuming that the effect of

the aluminum could be approximated by concentrating it

in an aluminum shell with an equivalent number of

atoms per square cm placed just inside the boron shell.

The scattering problem then is treated as resonanceenergy removal of neutrons directed radially inward and

inscatter of the others with a spectrum degraded in

energy by the average logarithmic decrement for

elastic scatter = 0.072). The expression for the

ratio of the spectra with, and without, aluminum

estimated by this procedure is

ISNF U)/A£(ISNF-1) ^ _ ^ ~^t'. "^T*^_

ISNF a)/oA£(ISNF/NA) U - e

) + e l^^t

Figure 5 shows the results of applying this single scattering analysis to the prominent resonance at 35 keV.

[The region from 20 keV to 50 keV accounts for 6,5% of

the ISNF flux.] In this figure we see the histogram

representation of the smoothly varying ISNF/NA spectrum,

the negative removal contribution and the positive

inscatter contribution. The figure shows that the

spectrum "ISNF(CALC)" obtained by single scatter

analysis gives a very good representation of the ISNF-1

spectrum calculated by discrete ordinates method. The

differences indicated by cross hatching in figure 5 account for about 1% of . the ISNF flux between 20 keV and

50 keV.

0 20 40 70 100 E , KeV

Conclusions

The spectrum of the ISNF at NBS has been thoroughly

studied by means of discrete ordinates transport calculations. Sensitivity of the spectrum and integral

detector responses to variations of all significant

physical parameters of the system has been investigated.

For parameter variations considered realistic, spectrum

changes in a coarse energy-group description are mostly

less than 1% and never more than a few percent.

A quantitative assessment of the accuracy of the

neutron spectrum calculated for the ISNF will be based

on the sensitivity studies described in this paper,

combined with a final estimate of uncertainties

associated with the corresponding physical and nuclear

parameters. The initial goal is to achieve a coarse

(about 40 groups) specification of the spectrum that is

accurate to better than + 5% in the energy range 0.4

keV to 8 MeV. The lower bound corresponds to an energy

below which the response of a 1/v-detector is less than

5% and the upper bound to the limit of good fission

spectrum spectrometry data [6]

.

Acknowledgements

We wish to thank H. Scheinberg and J. Kostacopoulos

for their efforts in the extremely difficult task of

fabricating the boron-10 shell, and R. LaBauve and

D. Muir for their 240-group discrete ordinates calculations of the ISNF spectrum.

References

[1] H. Scheinberg, Los Alamos Scientific Laboratory

(Private Communication)

.

[2] R. G. Saltesz, "Revised WANL ANISN Program User's

Manual", Westinghouse Astronuclear Laboratory

Report WANL-TMI-1967, April, 1969.

[3] C. Eisenhauer and A. Fabry, "DETAN 74: Computer

Code for Calculating Detector Responses in Reactor

Neutron Spectra", available upon request from the

National Bureau of Standards.

[4] D. Gilliam, et al., this conference.

[5] R. LaBauve and D. Muir, Los Alamos Scientific

Laboratory (Private Communication)

.

[6] J. A. Grundl and C. M. Eisenhauer "Fission Spectrum

Neutrons for Cross Section Validation and Neutron

Flux Transfer", NBS Special Publication 425,

October 1975.

Fig. 5. Comparison of discrete-ordinates spectrum

(ISNF-1) with spectrum (ISNF(CALC)) calculated

by considering effect due to removal and inscatter in aluminum on slowly varying spectrum

(ISNF/NA) for no aluminum. Hatched area

indicated difference of two spectra.

334
STANDARDIZATION OF FAST PULSE REACTOR DOSIMETRY

A. H. Kazi

Army Pulse Radiation Facility, Aberdeen Proving Ground, Maryland 2)005

E. D. McGarry

Harry Diamond Laboratories, Adelphi, Maryland 20783

D. M. Gi 1 1 iam

National Bureau of Standards, Washington, D.C. 2023't

A dosimetry method, developed at the National Bureau of Standards and known as the Flux

Transfer Technique, is proposed for accurately determi ning the total fast flux in the

vicinity of a fast-pulse reactor or bare-critical assembly. The method is to determine the

fast flux from the comparison of free-field 239pu fission-rate measurements made at the

reactor facility to calibration measurements made in a standard ^^^Zf neutron flux. Use of

the technique at the Army Pulse Radiation Facility shows that total fluxes can be measured

in and near the reactor to a determinable accuracy of ^5%. For comparison, but with less

accuracy of iSZ , the same total fluxes were verified using ^^^Hp, 23'ty^ ^3(>\j and 23o(j_

The technique has several important advantages. It uses a recognized standard neutron

source for calibration. Accurate fission rates are measured and compared with dual-isotope

fission chambers that are easily calibrated. The method is independent of errors in foil masses. The method does not require the use of an unfolding code and the effects of cross section errors are lessened because of the need to evaluate only cross section ratios. The

method is simple and is readily amenable to absolute error analysis and i nter 1 abora tory

cal ibrations.

(Pulse reactor calibration, neutron flux standard, dosimetry, Ca 1 i forn i um-252

,

radiation effects)

I ntroduct ion

Fast Pulse Reactors

Fast pulse reactors are bare, all-metal fuel

critical assemblies that are operated either at steady

state power or at super-prompt critical ity to produce

near-fission-spectrum neutrons in a microsecond time

frame. '"^ At present there are six such reactors

operating in the US. They serve as sources of

neutrons for a wide variety of DOD, ERDA and NASA

programs, and are used extensively for TREE tests. There are two non-US reactors which fall into this

class: The French reactor CALIBANS and a fast-pulse

reactor in the USSR." These reactors have the least

complex physical structures of all the operating

nuclear reactors, and their small sizes and lack of

moderators enable them to be positioned far from

neutron reflecting or moderating surfaces. Also,

many of these reactors have simple geometry in-core

irradiation facilities or "glory holes". All these

conditions facilitate calculating spectra for direct

comparison with measurement. Since most current fast

pulse reactors are made of the same uranium-molybdenum

fuel alloy, the in-core and free-field leakage spectra

of these assemblies are expected to be nearly identical.

Neutron Dosimetry Requirements

There are three key neutron dosimetry requirements for fast-pulse reactors: (l) accurate measurement of the total neutron fluence that is delivered

to a particular location; (2) high-resolution measurements of the neutron spectrum; (3) known energy

dependence of the response functions characteristic

of the particular phenomena being investigated.

the spectrum-averaged cross sections of the neutron

detectors. However, the third requirement encompasses

more than neutron dosimetry, as discussed so far, and

includes the response functions of the experiments of

interest. Typical examples are neutron dose in tissue

for biological experiments and atom-displacement damage

effects in semiconductor electronics for experiments

involving Transient Radiation Effects in Electronics

(TREE experiments). For such experiments a spectrum

measurement or calculation is required to determine a given response per unit neutron exposure. This is obtained by integrating the response function over the

spectrum. For fast-pulse reactors a few reference

spectrum determinations are adequate since for many

response functions of practical interest the neutron

f luence-to-response conversion factors are fairly weak

functions of variations in fast-pulse reactor neutron

spectra. This was confirmed in a sensitivity analysis

performed at the Army Pulse Radiation Facility (APRF)

,

which examined the sensitivity of various neutron

response functions to realistic variations in neutron

spectra encountered at fast pulse reactors.^'

Therefore, the key output characteristic of these

reactors is the total neutron fluence which must be

accurately determined for each exposure at each

location. Present dosimetry is such that absolute

errors in the fluence determinations are so dependent

upon complex procedures and calibrations that accuracies

are unknown. Furthermore, different dosimetry techniques give results that vary by SOt for the same exposure . 9 • Obviously there is need for a standard

calibration procedure. Such a standardization technique

is described herein. The method was developed by the

National Bureau of Standards (NBS),'' and has been used

at APRF^^ to measure total fluxes to an accuracy of -St.

The first two requirements are fundamental and

the first is integrally related to the second through

335
Flux Transfer Technique Fission Chambers

Basic Method

239 Basically, the method uses a Pu fission chamber

as a flux-measurement standard after it has been calibrated against a 252(;f source. The neutron-energy

spectrum of a properly encapsulated 252(;f deposit is the most accurately known neutron reference source

ava i 1 ab 1 e .

' 3 ,

1 The fission cross section of 239py

is energy independent to within t(>% between 10 keV

and 5 MeV.'5 |n the absence of extraneous moderators,

the fraction of neutrons with energies below 10 keV

in and near a fast-pulse reactor is insignificant.

Furthermore, greater than 37% of the neutrons in a fast-pulse reactor spectrum, like those in a Cf

spectrum, have energies between 10 keV and 5 MeV.

All these factors, taken together, are the basis for

the standardization technique.

Detai Is of the Method

Let " r'' refer to a reactor spectrum and "s"

refer to the standard 252(;f spectrum. For a given

isotope the ratio of the fission rates, Fj-ZFg is:

F a 0

r _ _r_ _r

F ~ a

' 0

s s s

where {Of/a^) is the ratio of the spectrum-averaged

fission cross sections for the isotope of interest

in the reactor and 252(;f spectra, and 0 is the flux.

Note that the mass of the isotope has cancelled out

of the ratio.

0 Now let K = -rA

s F

s

The total flux, 0^, at the reactor is therefore

derived from the known Cf flux, 05, by

a 0 = K • F • — r s r a

r

Note that the equation for 0^ only involves the

ratio of the spectrum-averaged cross sections. This

ratio is more accurately known than the absolute

value of either cross section, because errors in cross-section magnitudes tend to cancel from the

ratio. Since the 239pu cross section is nearly flat

(energy independent) over the energy range of interest,

the ratio is very close to unity, typically 1.04 for

a fast pulse reactor. This says that the 239pu Flux

Transfer Technique is insensitive to the shape (energy

dependence) of the spectrum between 10 keV and 5 MeV.

However, the cross section of ^39pu is certainly

not energy independent much below 10 keV. The method

must therefore account for the presence of low-energy

neutrons caused by structural or other extraneous

material near the reactor. Herein, the technique is expanded to provide for several capabilities. First,

by virtue of their dua 1 -detector construction, the

fission chambers can provide accurate ratios of

various isotopic fission rates. Such ratios give

information about the spectrum and are most useful in detecting spectral changes. Furthermore, the chambers

may be covered with cadmium or boron-10 to demonstrate

that the bare-chamber response has not been compromised by low-energy neutrons.

The fission rates are measured with dual fission

chambers and a triple-sealer pulse-processing system

developed by NBS. The chambers have been described

in detail'' and have been extensively evaluated at a variety of radiation sources .' ^' ' 7 Figure 1 shows

a fission chamber with and without a boron-10 cover. Figure 2 shows an enlarged cross-sectional view of

the chamber construction. To reduce neutron absorption

and scattering effects, the electrodes and structural

elements are made of aluminum and insulators are made

of a hydrogen-free polymer (Teflon).

Five isotopes were used for the present measurements: 239pu; 237np; 23'tu; 236y. and 238u. All

deposits were 12.7mm-dia oxides vacuum evaporated on polished nickel or platinum disks of 19.1mm-dia and

0.33mm thickness. Isotope data and foil masses are summarized in Table I.

As seen in Fig. 2 the NBS double-fission chamber

contains two independent fas t- ion i zat ion chambers.

The backings of the thin-fission deposits are back to back so that the deposits are only one millimeter apart.

The chambers are operated at about 135 volts with con- tinuous methane-gas flow. The electrical signals

from the chambers have the voltage pulse height shown

in Fig. 3- Note the very broad fission-fragment peak

that is well separated by a low valley from the noise

and alpha background, near zero pulse height. In the

absence of undue electronic noise, the magnitude of

the valley region is proportional to the thickness of

the fission foils. The absolute efficiency of each

fission chamber is near 100% for foils that are thin

compared to the range of a fission fragment. Small

corrections are described in Ref. 11.

Cal ibrat ion

Use was made of the NBS ^^^Cf Fission-Spectrum

Irradiation Fac i I i ty . ' ^ >

' This 252(;f neutron source

is a 0.1-cm3 deposit and singly encapsulated in a 0.3'»-cm3 steel container. The location of the ^^^Cf

deposit in the capsule is known to within tO.Smm

relative to the surface. The source strength in February 1976 was 3.3xlo9 n/s t].3Z. The 252cf neutron

source strength was de termi ned ' ^ > ' 9 with a manganous

sulfate bath relative to the internationally compared

radium-beryllium photoneutron source, NBS-1, presently

known to t] .]%.

For calibration, the fission chambers were mounted,

two at a time, l06-mm apart on a light aluminum frame

on opposite sides of the 252cf source. This is shown

schematically in Fig. k. By placing the source between

two similar detectors in carefully controlled geometry,

positioning errors were held to within \ .5% of the

source-to-detector distance and, therefore, represent

only 3% uncertainty in the 252(;f calibration flux at the fission chamber. Repositioning of the source

relative to the chambers can be accomplished to io.Smm.

Prior to irradiation, all important distances can be

measured without the source in place. Chamber-tochamber distances are determined optically to tO.Smm

on a machinist's bench.

Fast Pulse Reactor Dosimetry Evaluation

The Army Pulse Radiation Facility (APRF)

The APRF reactor is representative of current

generation fast-pulse, or as they are often called,

fast-burst reactors. It consists of an unmoderated,

336
unreflected cylinder 22.6cm in diameter by 20cm high,

made of 90% uranium (93-21 235u) and 101 molybdenum.

The reactor is suspended from a transporter so that

it can be located at floor level or as high as 1 ^m

above a borated concrete floor anywhere within, or 30m outside, a 30m dia by 26m high, low-scattering

weather shield. Fig. 5 shows the reactor at 6m above

the floor inside this reactor building. The APRF can

be operated with two in-core irradiation facilities,

called glory holes. The smaller glory hole is 38mm

in useable diameter and 20cm long. The larger glory

hole is 106mm in useable diameter and also 20cm long.

When using the larger glory hole, the reactor is operated with reflector control.

Total Flux Measurements

The fluxes determined from ^-^^Pu fission-rate

measurements are summarized in Table II. The errors

quoted in the table result from a \.3% error in the 252(;f source strength, a 2.5% error in APRF power

determination, a 3-3 to 3-3% combined error in the

252cf calibration factors and APRF fission-rate

measurements, and a 0.7% error in effective crosssection ratio. This last error was determined from

a sensitivity analysis in which the 252cf spectrum

was shifted by t^O keV and the APRF spectra by to.l

MeV and the corresponding effects on the spectrumweighted cross-section ratios were calculated. The

resultant error of 0.7% for 239pu is small, because

the -'"Pu fission cross section is flat over the

energy range of interest. As mentioned, this was

the principal reason for selecting 239pu as the

flux-standardization material.

Flux Measurements With Other Fissionable Isotopes

Flux measurements were also made using ^-'^Hp,

23'»u, 236u and 230u. jhg average fluxes obtained at the three locations were 1 .22xl07 + 7.3%, 1 .00x10^ ± 7.3% and 7-91xlo7 + 7.3I neutrons/cm2-sec-watt

respectively. The estimated errors in these measurements are based on data similar to the 239pu data,

except that the cross-section ratio error is estimated

at about tSZ. This larger error is because approximately 40% of the neutrons in the APRF spectrum are in the energy range 6OO keV to 1.5 MeV where the cross sections of these other fissionable isotopes are not energy independent like 239py. Indeed, these other

isotopes are nearly threshold functions with the ^^^Hp

and 234u thresholds near 6OO keV, the 236u threshold

near 0.95 MeV and the 238y threshold near 1.5 MeV. However, the total fluxes measured with all isotopes,

including 239py^' agree to well within the estimated

errors. This lends credence to the validity of the

basic flux transfer data given in Table II. The

implication is that the spectrum-averaged crosssection ratios and hence the calculated APRF spectra

are satisfactory for use with detectors with thresholdtype response functions.

Effects of Cadmium and '^B/jC Shields

Fission-rate measurements were made with a 0.8mm

thick cadmium-covered fission chamber containing 239pLj

and at all three locations. The cadmium cover

should eliminate any thermal neutron response of the

23"pu. The measured fission rates differed by less

than 0.2% from bare-chamber results, indicating the

absence of any detectable therma 1 -neutron contribution.

Measurements were also made with a boron-covered

239Pu/237np chamber (2.2 g/cm2 thickness of '^B). This

boron shield should effectively remove all neutrons

below 1 keV. The measured 239pu/237np f I ss I on-rate

ratios were 2% below the bare chamber results. These

data confirm that the APRF spectra at the selected

locations have a very small low-energy (< 1 0 keV)

component. Additional verification of this fact is given in a later section.

Fission-rate measurements with boron-covered ^^'^Hp

in the 106mm dia glory hole are essentially a measurement of the attenuation of the flux due to the boron.

This was determined to be ]h.]%.

Spectral Indices

Table III compares measured and calculated spectral

indjcgs for 239p,j relative to 237|gp gpjj 237f,jp relative

to U at three APRF exposure locations. Because of

the close, fixed position of the two different isotopes

in each fission chamber and simultaneous collection of

fission rates on a dual-scaler counting system, one obtains accurate fission-rate ratios at the same time

as measured fluxes. The APRF spectral indices were determined from the fission-rate ratios measured at APRF and two independent determinations of the spectral

indices for 239p^j/237Np and 237Np/238u at the 252cf

source. The indices for 252(;f neutrons are 1.35 - 2.5%

and A. 16 t 2.7%, respectively. 20

The data of Table III show that transport-calculated

spectra and ENDF/B-IV cross sections give cross section

ratios reasonably consistent with experimental data.

The sensitivity of the ratios is, to first order, proportional to the difference in the low-energy thresholds

of the two isotopes in the ratio. Thus agreement is expected to be better for the 237NP/238U ratio, wi th 600 keV and 1.5 MeV thresholds, than for the 239pu/237Np,

which is primarily sensitive to spectrum behavior below

0.6 MeV. The 237Np/238U measured and calculated

results are within the uncertainties of the measured

results. The errors on the calculated ratios are not known as they involve absolute uncertainties in the cross sections as well as spectra.

The disagreement between measurement and calculation

for the 239pu/237np ratio is not due to thermal neutrons,

as indicated by measurements with cadmium and boron

covers. For leakage-spectrum measurements, one could

argue that the difference is due to the presence of

neutrons scattered from external core-support structures.

However, because the in-core ratios are also higher

than calculated and not subject to spectral perturbations

from external structures, the discrepancy is more likely

to result from the ca 1 cu 1 at i ona 1 model.

Further Application

The evaluations, thus far, indicate that the procedures, when calibrated against a recognized-standard

252(;f source, can provide a practical and acceptable

technique for accurately determining total fluence and

spectral indices at fast-pulse reactors. The fission

chambers are not presently designed to operate in pulsed irradiations; however, there is no evidence of

spectral differences between pulsed and steady-state

reactor operations. In practice, fast-neutron pulsed

fluence levels are determined from the 32p beta radioactivity induced in sulfur. Because the sulfur

reaction has a high-energy threshold, sulfur-monitored

result is usually reported2' as fluence with energy >3 MeV. The sulfur fluence is then multiplied by a previously determined (>10 keV/>3 MeV) flux ratio to yield the total fluence. In the absence of extraneous

moderator, the number of neutrons with energies below

10 keV is insignificant. The required flux ratio is determined using a number of techniques including

transport calculations, spectrum measurements, fissionfoil analysis and spectral unfolding. The resulting

flux ratios vary considerably, from 6.7 to 10.

337
This spread is not due to real differences in reactor

spectra, but is the result of differences in measurement techniques. It is evident that there is also a need for a standard calibration procedure for sulfurfluence monitoring.

Sulfur Fluence Measurements

For steady-state irradiations, can be used

to provide a flux-transfer calibration for sulfur by

the fol lowing:

where 0tr is the fluence in spectrum "r" and C refers

to the respective sulfur responses (e.g., counts/time

on the beta counter used to measure the sulfur

activity). The cross section ratio Og/o^ refers to the ratio of the sulfur cross section integrated over

the entire energy range of the spectrum.

A separate 252cf calibration of the sulfur „ resulted in fluxes of S.OxlO^ n/cm^-s "w , 1 . Ox 1 0 n/cm

•S'w and 1.2x10^ n/cm^'s-w, respectively, for the

large- and small-glory hole and reactor-surface

locations. There is very good agreement between these

results and the total fluxes reported in Table II. The estimated error on the sulfur-monitored fluxes is t^% and is composed of tJi.kZ from the sulfur dosimetry,

from power level normalization, and -dX on the

sulfur cross section ratio. The good agreement

between the sulfur-flux values and the flux-transfer

data suggest that the >3 MeV portions of the calculated

APRF spectra are quite good.

Spectrum Measurement

Further confirmation of the quality of the calculated leakage spectrum is given by results of protonrecoil measurements, recently completed22 and shown

in Fig. 6. The uncertainties in the measured results

in the 200 keV to 2 MeU range are approximately t8%.

Uncertainties below 200 keV increase from about -\^%

at 100 keV to tZ5% at the low-energy end. Assuming

the calculated results (solid curve in Fig. 6) above

2 MeV, the measured-to-calculated total flux ratio

above 30 keV is only 1.035. There is 211 more measured flux in the 30 keV to 200 keV interval but

only 1% of the total flux is below 200 keV. This

measured difference only accounts for half of the

total difference in the previously discussed 239pu/

237np spectral index.

>3 MeV Sulfur Dosimetry

Although it is not necessary for the sul f ur-f 1 uence measurement technique, it is of interest to extend the

procedure to sulfur measurements of >3 MeV fluence to better evaluate the past history"^ of the 10 keV/3 MeV

controversy

.

Based on present knowledge of the 252cf spectrum,

the fraction of '^^'^Zf flux >3 MeV is 0.237.^ The

present total -fl uence sulfur-calibration results are corrected to reflect cross sections >3 MeV in the

252cf and APRF spectra. Such adjustments yield

(>10 keV/>3 MeV) flux ratios for both glory holes

and the leakage spectra of 8.8 io.6 and 1 .h t 0.5,

respectively.

Concl us ions

The evaluations at APRF indicate that the procedures, when calibrated against a recognized

standard Cf source, can provide a practical,

acceptable technique for accurately determining

total fluence and spectral indices at fast pulse

reactors. By calibrating with the 252^;^ source accurately positioned between two high-quality, thindeposit fission chambers, the uncertainty in the 252Qf

calibration flux at the deposits can be held to -2%.

Subsequent measurements at the APRF, with 239pu fission

deposits, show that free-field fluence at a fast-pulse

reactor can be calibrated to an absolute accuracy of

ts%- Additional experiments with cadmium and ^^g^c covers confirm the applicability of the Flux Transfer

Technique for the situation where the neutron flux Is free of low-energy neutrons. Also these results show

that when the spectrum under investigation is similar

to a fission spectrum, a direct neutron-flux transfer

can be undertaken. For other classes of spectra,

computed fission-spectrum-averaged cross sections are required to obtain a flux In the spectrum under Investi

gation. In both cases, the neutron Flux-Transfer

Technique relaxes the requirements to establish

absolute activation detector efficiencies and

uncertainties associated with absolute cross section

sea 1 es

.

Comparisons of the Flux Transfer Technique were

made with ^^^Hp, several uranium isotopes and the

32s(n,p)32p reaction all calibrated against 252cf.

Total fluxes from all measurements agree to within -1%.

The technique provides several other capabilities.

First, by virtue of their dual-deposit construction,

the fission chambers give accurate ratios of various

isotope fission rates at specified locations near reactors. Such ratios provide information about the

neutron-energy spectrum and are most useful in detecting spectral changes that result from perturbations by

structural materials or experiments. Second, the

method may be used to establish secondary standarization of activation foils, which are suitable for

neutron dosimetry during pulsed reactor operations.

Such secondary standardization can be used to calibrate

data obtained with unfolding codes.

Reactor neutron dosimetry includes questions of

both the magnitude (total fluence) and the shape of

the neutron spectrum. Sensitivity analyses, performed

at APRF, indicate that for many applications the

question of total fluence is paramount. This problem

is satisfactorily solved with good accuracy by the

present flux transfer technique. Since essentially all

fast pulse reactors are similar in composition and con- struction, the method is proposed as a means of Intercallbrating all fast-pulse reactor dosimetry. It has

the tremendous advantage of being an easy task to repeat, thereby permitting periodic restandard i zat ion

and interlaboratory comparison. The method is con- ceptually simple, accurate and readily amenable to absolute error analysis.

Acknowl edgments

Credit is due to Dr. V. Spiegel for doing a large

portion of the calibration work at the NBS 252cf source

Mr. Thomas Wright for assistance with data collection

and computer analyses, and Lt George Davis for performing the sulfur calibrations. We also thank

Mr. Charles Elsenhauer for obtaining ENDF/B cross section data and calculating the effects of energy

shifts. We appreciate Mr. Robert Dallatore's careful

fabrication of fission chambers and irradiation hard- ware and equally appreciate the many helpful suggestions given by Dr. James Grundl and the enthusiastic

support of the APRF staff in operating the reactor.

338
References

1. ProceeJings of the National Topical Meeting on

Fast Burst Reactors, Albuquerque, N.M.,

January 1969, CONF-690102 (December 1969).

2. J. T. Mihalczo, "Superprompt-C r i t i ca 1 Behavior of

an Unmoderated , Unreflected Uranium-Molybdenum

Alloy Reactor," Nuclear Science and Engineering

16:291-298 (July 1963)

•

13. J. A. Grundl et^. a_j_. , "A Ca 1 i forn i um-252 Fission

Spectrum Irradiation Facility for Neutron

Reaction Rate Measurements," Nuclear Technology

32 (3) (March 1977).

14. H. T. Heaton II, et. aj_. , "Absolute ^^^U Fission

Cross Section fo r^52cf Spontaneous Fission

Neutrons," Proc. of Conference, Nuclear Cross

Sections and Technology, Vol I, NBS Special

Publication 425:266 (October 1975).

3. A. H. Kazi, "Fast-Pulse Reactor Operation with

Reflector Control and a 1 06mm Diameter Glory

Hole," Nuclear Science and Engineering, 60:62-73

(May 1976).

4. R. W. Klingensmith et^. aj_. , "TREE Simulation

Facilities," DNA-2432H, Defense Nuclear Agency

(1973).

5. J. Cortella et^. a_l_. , "Operational Characteristics

of the CALIBAN Fast Pulse Reactor," pp 157-170,

Proceedings of US/Japan Seminar on Fast Pulse

Reactors, University of Tokyo, Takai, Ibaraki,

Japan (January 1976).

6. G. G. Doroshenko e_t. a_l_. , "Spectra of Fast Neutrons

from a Pulsed Reactor," Atomnaya Energiya, hO : 6, pp k(>0-h(>k (June 1976)

.

7. T. A. Dunn, A. H. Kazi, J. Saccenti, "Fluence-toDose Conversion Factors for APRF Fast Pulse

Reactor Neutron Spectra," USA Ballistic Research

Laboratory Report 1832 (September 1975).

8. A. H. Kazi, T. A. Dunn and J. C. Saccenti,

"Sensitivity of Fl uence-to-Dose Conversion to Changes in Fast Pulse Reactor Spectra," Proc. First ASTM-EURATOM Symposium on Reactor Dosimetry,

Petten, Netherlands (September 1975).

9. F. N. Coppage, "The Influence of Dosimetry on Damage Equivalence Ratios," IEEE Trans. Nuc. Sci.

NS-22, 6:2336 (December 1975).

15. J. A. Grundl and C. M. Eisenhauer, "Fission

Spectrum Neutrons for Cross Section Validation

and Neutron Flux Transfer," Proc. of Conference,

Nuclear Cross Sections and Technology, Vol I, NBS Special Publ ication 425:250 (October 1975).

16. J. A. Grundl and J. W. Rogers, "Absolute Fission

Chamber Measurements in CFRMF," LMFBR Reaction

Rate and Dosimetry 7th Progress Report, HEDL-TME

73-31, Hanford Engineering Development Laboratory

(1973).

17. D. M. Gilliam, "Integral Measurement Results in Standard Fields," paper #4. This conference.

18. R. H. Noyce, E. R. Mosburg, Jr., J. B. Garfinkel

and R. S. Caswell, "Absolute Calibration of the

National Bureau of Standards Photoneutron Source

III Absorption in a Heavy Water Solution of

Manganous Sulfate," Reactor Sci. Technol. (j. Nucl.

Energy, Part A/B) 17:313 (1963).

19. V. Spiegel, Jr. and W. M. Murphy, "Calibration of

Thermal Neutron Absorption in Cylindrical and

Spherical Neutron Sources," Metrologia, 7.= 34

(1971)

.

20. D. M. Gilliam et. a_I_. , "Fission Cross Section

Ratios in the ^^Cf Neutron Spectrum," Proc. of

Conference, Nuclear Cross Sections and

Technology, Vol I, NBS Special Publication

425:270 (October 1975).

21. T. A. Dunn, "APRF Reactor 3 MeV (Sulfur) Threshold

Neutron Fluence Measurements," USA Ballistic

Research Laboratories Report 1786 (May 1975)-

10. J. L. Meason et^. aj_. , "The Neutron Spectral Distribution from a Godiva-Type Critical Assembly,"

IEEE Trans. Nuc. Sci. NS-22, 6:2330 (December

1 o^c•^ 22. E. D. McGarry, C. R. Heimbach, A. H. Kazi and

G. W. Morrison, "Fast Pulse Reactor Neutron

11. J. A. Grundl, D. M. Gilliam, N. D. Dudey and Spectrum Measurement and Calculation," IEEE

R. J. Popek, "Measurement of Absolute Fission Trans. Nuc. Sci. NS-24 (1977). To be published.

Rates," Nuclear Technology 25: 237-257

(February 1975).

12. E. D. McGarry, A. H. Kazi, G. S. Davis and

D. M. Gilliam, "Absolute Neutron-Flux Measurements at Fast Pulse Reactors with Calibration

against Cal ifornium-252," IEEE Trans. Nucl.

Sci., NS-23, 6:2002 (December 1976).

339
TABLE I

ISOTOPIC COMPOSITION AND NOMINAL MASSES

OF FISSIONABLE DEPOSITS

MASS

(MICROGRAMS)

TYPE OF DEPOSIT

Pu-239 Np-237 Np-237 U-234 U-236 U-238

427 265 170 30 37 175

PERCENTAGE

OF ISOTOPES

Pu-239 99.978

Pu-240 0.002

Pu-241 0.005

Pu-242 0.005

Np-235 - 0.004

Np-236 - O.OOS

Np-237 99.98 99.991

(Pu-239) 0.02 -

U-234 99.887 0.002 -

U-235 0.064 0.005

U-236 0.035 99.990 0.001

U-238 0.014 0.003 99.999

FIG. 2 CROSS-SECTIONAL VIEW OF NBS DUAL

FISSION CHAMBER CONSTRUCTION

FIG. I. NBS FISSION CHAMBER WITH AND WITHOUT 80R0N-I0 COVER

340
FIG. 5. APRF REACTOR POSITIONED 6 M ABOVE BORATED

CONCRETE FLOOR FOR CORE SURFACE LEAKAGE

FLUX MEASUREMENT

TABLE II TOTAL APRF FLUXES MEASURED WITH PU-239 USING THE FLUX TRANSFER TECHNIQUE

EXPOSURE LOCATION NEUTRONS /CM^-SEC-WATT

67 MM FROM CORE SURFACE 1.22 XIO^ » 4.4%

CENTER OF CORE INSIDE 38 MM GLORY HOLE 1.04 X 10^ * 4.4%

CENTER OF CORE INSIDE

106 MM GLORY HOLE 8,30 XIO^* 4.9%

TABLE ill - COMPARISON OF MEASURED AND CALCULATED

SPECTRAL INDICES FOR THE APRF REACTOR

SPECTRUM-AVERAGED CROSS-SECTION RATIOS FOR:

ISOTOPE

RATIO

67.MM FROM

CORE SURFACE 36-MMGLORY HOLE 106.MM

GLORY HOLE

MEASURED CALCULATED MEASURED

MEASURED

CALCULATED MEASURED

MEASURED CALCULATED MEASURED

"^P 5.39 + 3.4%

1.017

5.72 + 3.4%

1.000

5.77 + 3.4%

0.991

"Spu

1.69 0.9291.78

0.91 S

1.81 0.901 ± 3.2% ±3.2% ±3.2%

"T 1 r "T 1 r "1 1 r

J I L 0.02 0.04 0.06 0.08 0.1 0.2 0.4 0.6 0.8 1 NEUTRON ENERGY (MeV)

6 8 10

FIG. 6. MEASURED AND CALCULATED APRF NEUTRON LEAKAGE SPECTRUM

341
DOSIMETRY STANDARDS FOR NEUTRONS ABOVE 10 MeV

H.H. Barschall

University of Wisconsin, Madison, Wisconsin 53706

Dosimetry of neutrons in the energy range 10-50 MeV is needed for applications in radiation

damage studies and in biomedical work. Dose determinations use either a fluence measure- ment or the Bragg-Gray principle. Better knowledge of activation cross sections, kerma

factors, and energy per ion pair is needed to reduce uncertainties in dosimetry.

(Activation; Bragg-Gray; dosimetry; energy per ion pair; fluence; kerma factor; neutrons)

Introduction

This report discusses dosimetry standards for neu- trons in the energy range 10-50 MeV with special em- phasis on 14-MeV neutrons. Neutron dosimetry in this

energy range is of importance for applications to studies of radiation damage in various materials and to biomedical work.

The effect of 14 -MeV neutrons on materials is un- der intense study for the design of fusion reactors,

in particular for the choice of a material for the

first wall of the reactor. ^ Radiation damage problems

are expected to be much more severe in fusion reactors

than in fission reactors because of the much larger re- action cross sections for the neutrons of higher energy.

Production of hydrogen and helium by nuclear reactions

is likely to produce swelling, blistering, and embrittlement, and generally new elements produced. in transmutations may affect the mechanical properties of the

materials adversely.

In the biomedical area the interest in dosimetry

used to be restricted to applications in radiation protection. More recently the interest has shifted to radiotherapy^ where much higher accuracy is needed. If

the dose administered to a tumor is 51 too low, the cure rate is greatly reduced, while a 51 too large dose

may result in severe side effects.

• Definitions

Both material scientists and radiobiologists wish

to correlate observed effects and radiation dose. Dose

is defined^ as the energy imparted by ionizing radiation to the matter in a volume element, divided by the mass in that volume element. Dose differs in general

from a similar quantity, kerma(K), which is defined^ as the sum of the initial kinetic energies of all the

charged particles liberated by indirectly ionizing particles in a volume element divided by the mass in that

volume element. For neutrons kerma and dose differ

primarily near the surface of the irradiated material

within a distance of the order of the range of the

secondary charged particles in the material. Both kerma

and dose are measured in grays (1 Gy = 1 J/kg) , although the older unit, the rad, is still more widely

used (100 rad = 1 Gy] . Another quantity related to the

dose is the neutron fluence $. The neutron fluence is

defined as the track length of all the neutrons in a volume element divided by the volume, and it determines

the number of nuclear reactions in that volume. Kerma

and fluence are related by the kerma factor, which is

defined as the kerma per unit fluence. The kerma factor depends only on the nuclear constants of the material. For a single element of atomic weight A

K/. = N

^ a. E,

1 where N is Avogadro's number and is the partial

cross section for producing charged particles of average kinetic energy E.

.

Neither dose, nor kerma, nor fluence are, however.

good measures of either the radiation damage or of the

biological effects induced by neutrons. As mentioned

before, the radiation damage is strongly influenced by

the gas production and transmutation cross sections

which vary rapidly with neutron energy. When a material scientist says that a sample has been exposed to a certain neutron dose, he usually means fluence, not

dose. If one looks at the Dosimetry File'' published by

the National Neutron Cross Section Center, one finds

only neutron cross sections which might help to determine fluences, but there is no information in the

Dosimetry File on how to get dose; in fact, the word

dose appears only in the title. On the other hand,

engineers who design fusion reactors need to know the

dose in calculations of both nuclear heating and

shielding.

For biomedical studies the dose is the quantity

which is always quoted. In order to take into account

the biological effect of the dose the radiobiologist

multiplies the dose by the Relative Biological Effectiveness (RBE)

.

Measurement of Fluence

Neutron fluence can be measured most easily for a parallel beam of monoenergetic neutrons, for example,

at a distance of, say, 50 cm from a small source of DT

neutrons. In radiation damage studies the needed flu- ence is so high that small samples have to be placed

as close to the neutron source as possible. In this

geometry the neutrons are neither monoenergetic nor

monodirectional , since the neutron energy varies with

direction of emission, and the sample has a diameter

equal to, or smaller than, the size of the source. In such experiments thin foils of an activation detector

serve to determine neutron fluence and are placed

around the sample that is under study.

The principal requirements for the activation detector are that it must have a half life longer than

the time of bombardment, and that the resulting activity can be counted absolutely. In addition, the reac- tion that produces the activity should have a threshold not too far below the neutron energy being studied,

and it should have a cross section that varies slowly

with neutron energy in the energy range of interest.

Three reactions which are frequently used are listed^

in Table I. The (n,2n) reactions have thresholds

around 10 MeV, the (n,p) reaction is exoergic, but has an effective threshold near 2 MeV. The Nb(n,2n) re- action is preferred for 14-MeV neutrons.

The absolute counting of the resulting activities

does not introduce as much uncertainty as the lack of

knowledge of the reaction cross sections.

The only neutron cross section that is accurately

known for fast neutrons is the n-p scattering cross

section, and most absolute reaction cross section

measurements are comparisons with the n-p scattering

cross section. The best known reaction cross section

is probably the fission cross section of ^'^U. Fig. 1,

taken from a report on a specialists' meeting,^ shows

342
TABLE I

Activation Reactions used in Dosimetry

Reaction QCMeV) T^/2Cdays) YCMeV)

^^Ni(n,p)^^Co

^^Co(n,2n)^^Co

^^Nb(n,2n)^^Nb'"

0.39 70.8 0.81

10.5 70,8 0.81

-8.95 10.2 0.93

Fig. 1. Fission cross of ^^^U between 8 MeV and 22 MeV.

This figure is taken from ref. 6.

that measurements of this cross section at different

laboratories differ by more than 10% although the ex- perimenters quote much smaller errors. Activation

cross sections are not as well known as the ^^^U fission cross section; measurements at different laboratories often differ by an order of magnitude. The ac- tivation cross section that is reported to be best

known for neutrons around 14 MeV is that for the re- action ^

''AI (n,a) ^'*Na. Fig. 2 shows some values of

this cross section as well as the ENDF/B-IV evaluation

given in the dosimetry file."* The measurement quoted

with the smallest uncertainty (less than II) is the

absolute determination carried out in 1970 by Vonach

et al . ^ by the associated particle method. This

measurement is, however, ignored in the 1975 dosimetry

file, and the ENDF/B-IV evaluation differs by 51 from

this presumably most accurate measurement.

In most practical situations in which a foil serves

to measure fluence in radiation damage studies, the

variation of the activation cross section with neutron

energy may introduce more uncertainty in the fluence

determination than the error in the cross section at one energy. Fig. 3 shows the activation cross section

as a function of neutron energy for the three reactions

mentioned earlier. Experimental data on the Nb(n,2n)

cross section are shown in Fig. 4. The two most recent

extensive measurements^'^ around 14 MeV differ by 101.

There are no measurements available above 20 MeV.

Even for D-T neutrons the energy spread is typically 1 MeV, and the energy distribution within this

spread is usually poorly known and varies over the

sample. Hence there is substantial uncertainty in the

average neutron energy. If deuterons on Li or Be produce the neutrons, the energy spread is of the order of

the bombarding energy. Hence activation of foils gives

only a very rough measure of fluence. By using several

reactions with different thresholds the contribution of

130

100 —

13.5 14.0 14.5 15.0

En (MeV)

Fig. 2. Cross section of the reaction ^ 'Al (n,a) ^'*Mg

for neutrons between 13.5 and 15.0 MeV. Only

experimental results which are quoted with

small uncertainties are shown. The solid

curve is the ENDF/B-IV evaluation given in

reference 4. Additional measurements may be

found in this reference.

750

- 500

-O

b

250

'Co

\

\\\ \ \

^ \— ^^Nb(/J,

\

2/7) 52r"Nb

/ /

10 20

En (MeV)

30 40

Fig. 3. Activation cross sections of three frequently

used detectors as a function of neutron en- ergy from ref. 5. The solid curves follow

evaluated data sets, the dashed portion is an extrapolation based on theoretical con- siderations.

neutrons in different energy ranges can be estimated.

Dose and Kerma Determinations

If the fluence in a sample has been determined,

the kerma can be calculated if the kerma factor is

known. Again the only nuclide for which the kerma

factor is accurately known is ^H, since the only

343
500

E 400

300

93... , » >92.,,m Nb(n,2n) Nb

O LIVERMORE (1972)

X GEEL (1970)

• ZAGREB (1962)

13 14 15

En(MeV)

16

Fig. 4. Measured cross sections of the reaction

"Nb(n,2n)'^Nb"^ between 13 and 16 MeV. The

most extensive recent measurements are given

in re£. 8 and 9.

charged particles that are produced are recoil protons

and their angular distribution is close to isotropic

in the CM system. The kerma factors for and '*He are also well known, but they are of smaller practical

importance. For other nuclides the nuclear data

needed for calculating kerma factors are not well

known for neutrons above 10 MeV energy. Above 20 MeV

there are hardly any measurements of the needed nu- clear data, and kerma factors are calculated from nu- clear models. Abdou has developed a conputer code^°

for calculating kerma factors from the ENDF/B data

file.

Because of their importance to the biomedical applications the kerma factors of the principal con- stituents of tissue, H, C, N, and 0, have been calculated most carefully. The most widely used values

are the kerma factors published by Bach and Caswell''

in 1968. These have been revised recently by Caswell,

Coyne and Randolph'^ on the basis of more recent nu- clear data. Table II gives a conparison of the kerma

TABLE II

Kerma Factors for 15.5-MeV Neutrons

-9 2 (in 10 rad cm )

H C N 0

Bach, Caswell

(1968)

47 5 2 36 1.39 2.13

Caswell, Coyne,

Randolph

(1976)

47 1 3 02 2.68 2.07

Howerton

(1976)

47 1 2 17 2.90 1.48

factors of H, C, N, and 0 at 15.5 MeV according to the

two evaluations by the MBS group, and also according

to a recent evaluation by Howerton'^ at Lawrence Liv- ermore Laboratory. The differences between these

three evaluations indicate the uncertainties in the

kerma factors for the most carefully studied nuclides.

Bragg- Gray Approach to Dosimetry

In biological work dose in tissue is usually

measured with a tissue-equivalent (TE) ionization

chamber. According to the Bragg-Gray theory of cavity

ionization, the ionization produced in a small gasfilled cavity in a solid material is related to the

dose D at the position of the cavity by

Q W S D = g

^^^11 ^ where Q is the collected charge, W the

gas

energy required to produce an ion pair in the gas, M

the mass of the gas in the chamber, and S the mass stopping power. TE chambers were originally developed

for measuring y dose. The walls of the chamber con- sist of TE plastic, usioally Shonka A-ISO plastic,''*

in which most of the oxygen in tissue is replaced by

carbon. Such plastic is, however, not exactly tissue

equivalent for neutrons, since C and 0 have very different kerma factors for neutrons and the ratio of the

two kerma factors varies with neutron energy by an or- der of magnitude. For 15-MeV neutrons the difference

between tissue and TE plastic introduces a 51 correction.

TE chambers are usually calibrated with y-rays.

Such a calibration is not generally applicable for

neutrons because W and S are likely to be different

for the electrons produced by y-rays, and for the protons, a particles, and '^C and '^0 recoils, produced

by neutrons. In the past data on the dependence of W

on the mass and energy of the ionizing particle have

been contradictory. During the last year two

groups' ^''^ have obtained new results which clearly

indicate that W increases with increasing mass. There is also good evidence that for heavy ions W

increases at low energies. Some ^f the recent re- sults obtained for H"*", He"*", and C in TE gas are shown in Fig. 5. Except for low energy C ions these

w

(aV)

TE GAS

O A D FONTENAY -AUX - ROSES

• * BROOKHAVEN

He"

600 eoo ENERGY (keV)

Fig. 5. Results of two recent determinations of the

average energy loss per ion pair as a function of particle energy in methane -based^

tissue-equivalent gas for H""", He"*", and C

ions. The data are from references 15 and

16.

344
data are consistent to within 3%. A calculation of

an average value of W for the different type particles

produced by energetic neutrons will, however, have an appreciably larger error. Although there is also substantial uncertainty in

the knowledge of the ratio of the mass stopping powers

in solids and gases, the uncertainty has probably relatively little effect on the neutron calibration, be- cause some of the errors will tend to cancel in taking

the ratio of two ratios.

Comparison of Dose Measurement Based on Fluence

and on Bragg- Gray Approach

Neutron dose in tissue can be measured by two independent methods , i.e., by measuring fluence or by

measuring charge in a TE chamber. It is interesting

to conpare these two procedures for monoenergetic neutrons for which fluence can be measured fairly accurately. In an experiments^ performed at the Rotating

Target Neutron Source at Lawrence Livermore Laboratory

a 1-cm^ -volume commercially obtained TE chamber was exposed to 15-MeV neutrons. Neutron fluence was measured both by foil activation and by proton recoil

counters. Fluences were converted to dose using the

kerma factors of ref. 12. Charge collected in the

diamber was converted to dose on the basis of the y-ray

calibration with corrections for the difference in W

for Y rays and neutrons. In separate experiments both

air and TE gas served as chamber fillings. For air

there is a small correction for the difference in

stopping powers for electrons and protons. Table III

TABLE III

T-E Ion Chamber Conversion Factor (rad/nC)

(20°C, 760 Torr)

Chamber Gas

Radiation Method Material Air T-E Gas

60

Co y's

15-MeV

Neutrons

Bragg- Gray

Approach

Neutron

fluence

based

calibration

3,.19 2.78

A-150 3,,45 2.90

Muscle 3.,22 2.71

A-150 3.,31 2.81

Muscle 3.,09 2.62

dosimetry for such sources is very much more difficult

than for D-T sources, and there is an almost complete

lack of the required nuclear data. There are hardly

any of the relevant cross sections known for neutron

energies above 15 MeV so that neither the fluence nor

the kerma can be determined if the bombarding energy

is above 15 MeV.

In the biomedical application a tissue equivalent

dosimeter yields fairly reliable information, but for

the materials application there are large uncertainties. Since the sample is placed near the target,

there is a wide variation of both neutron spectrum and

intensity even over a small sample. The Li target has

to be so thick conpared to the distance to the sample

that there is a significant difference in distance between source and sample for the more energetic neutrons, which are produced farther from the sample,

from that for the lower energy neutrons, which are produced close to the sample. Hence the neutron spectrum at the sample is not the same as that observed

at a distance from the sample by, say, time-of- flight

measurements. In addition, the neutron spectrum

varies rapidly with angle of emission with respect to the incident deuterons; the average neutron energy de- creases rapidly with increasing angle. Before measurements with a D-Li neutron source can be interpreted

meaningfully, more measurements of the neutron intensity and spectra are needed as a function of deuteron

energy and emission angle. Furthermore, the chargedparticle production cross sections in the materials

to be studied and the activation cross section of the

dosimeter materials need to be measured as a function

of neutron energy up to the highest neutron energy

that the source produces.

The uncertainty in the dose detemination for

radiotherapy with continuous-spectrum neutron sources

is not as serious a problem because one does not know

the RBE accurately either. Radiobiological experiments measure the product of dose and RBE. Uncertainties in dose produce corresponding uncertainties

in RBE. While a better knowledge of the two factors

separately would be of considerable interest, it is

not essential for the therapy applications. Experiments are underway, however, which should improve the

knowledge of kerma factors for tissue for neutron en- ergies above 14 MeV. The results of such experiments

will help in the understanding of the dependence of

RBE as a function of average neutron energy.

References

1. Proceedings of the International Conference on Radiation Test Facilities for the CTR Surface and

Materials Program. Report ANL/CTR- 75 -4 (1975).

lists the results of the comparison both for dose in

muscle and A-150 plastic. The two procedures yield

calibrations which differ by only 4%, which is much

closer than would be expected on the basis of the un- certainties in the atomic and nuclear constants involved in the calculations.

Continuous Spectrum Neutron Sources

Since it is easier to obtain high neutron intensities by producing broad spectra than from the DT re- action, currently most neutron radiotherapy sources

use Be targets bombarded by either protons or deuterons. For the materials applications there is a plan to build a large facility in which the reactions

of deuterons with a thick lithium target serve as a high intensity neutron source. These sources produce

a continuous neutron spectrum extending from very low

neutron energies to the bombarding energy. Neutron

2. H.H. Barschall, Am. Scientist 64^, 668 (1976).

3. International Commission on Radiation Units and

Measurements, Report 19 (1971).

4. ENDF/B-IV Dosimetry File, Report BNL-NCS- 50446

(1975).

5. M.J. Saltmarsh, et al . , Report ORNL/TM-5696 (1976).

6. Proceedings of the NEANDC/NEACRP Specialists

Meeting. Report ANL- 76-90 (1976).

7. H. Vonach, et al . , Z. Physik 23_7, 155 (1970).

8. A. Paulsen and R. Widera, Z. Phys. 258 , 23 (1970).

9. D.R. Nethaway, Nucl. Phys. A190 , 635 (1972).

345
10. M.A. Abdou and C.W. Maynard, Nucl. Sci. Eng.

56, 360 (1975).

11. R.L. Bach and R.S. Caswell, Rad. Res. 35, 1 (1968).

12. R.S. Caswell, J.J. Coyne, and M.L. Randolph,

Proc. Workshop on Physical Data for Neutron

Dosimetry, Rijswijk, The Netherlands, EUR 5629e,

Luxembourg, 1976, p. 69.

13. R.J. Howerton, private communication. E.F. Plechaty, D.E. Cullen, R.J. Howerton, and

J.R. Kimlinger, Lawrence Livermore Laboratory

Report UCRL-50400, Vol. 16 (1976).

14. F.R. Shonka, J.E. Rose, and G. Failla, Proc.

Second U.N. Conf. on Peaceful Uses of Atomic

Energy, 21, 184 (1958).

15. N. Rohrig and R.D. Colvett, Rad. Res. 67, 613

(1976).

16. M. Chemtob, B. Lavigne, J. Chary, V.D. Nguyen,

N. Parmentier, J. P. Noel, and C. Fiche, Phys.

Med. Biol, (to be published)

17. H.H. Barschall and E. Goldberg, Medical Physics

4^ (to be published)

.

346
SUMMARY SESSION: W. W. Havens, Jr., Chairman

PANEL MEMBERS: R. Caswell, V. Farinelli, H. Llskien, H. Motz, R. Peelle, and L. Stewart

Introductory Remarks : Dr. Bowman

We now come to the final session - the summary of the Symposium. Bill Havens will

be chairing this session and he has been charged with maintaining a tight schedule.

Since we plan to incorporate the record of this session into the proceedings, I am

required by federal government regulations to remind you that the discussion of this

conference is being recorded. Now, I'll turn this session over to Bill Havens.

Prof. Havens : In the interest of time, we are not going

to follow the format listed on the program. We are

going to incorporate the summaries of the discussion

groups into the appropriate portion of the summary of

the program. And the format will be that the leader of

the discussion group for the particular workshop will

be allowed five minutes at maximum to present the sum- mary of the workshop. Then Alan Smith will summarize

that portion of the conference. Then we will go to the

panel for discussion of that particular portion of the

conference. Alan Smith and I put together a list of

times which would be devoted to the particular subject

matter of the conference, and we'll try to adhere

rather rigidly to the allotted times. I don't know

what I will do if somebody doesn't stop when I get up

to cut them off but, hopefully, forcible means will not

be necessary. So in the interest of time we'll start

immediately

.

The first session was on light nuclei and the

nearest workshop to that was the one Bryan Patrick

chaired, the Workshop on Li and B. So, I'll call on

Bryan Patrick to give a five-minute summary of the

workshop

.

Dr. Patrick : I wonder if I'm allowed to take, as a foreigner, the Fifth Amendment to get around this re- cording business. No? We had a fairly lively discussion on ''Li and '•''b in which a large number of people

participated, and I hope this isn't too disjointed a representation of what went on.

The use of a large body of nuclear physics information In the form of angular-distribution, polarization

and inverse-reaction data in addition to the direct

measurements (a , a , and a ) in the R-matrix

n,t n,a n,n

calculation of Hale on the ^Li system, although not a new idea, is a welcome development which should be

encouraged. Most of the recent experimental work on

^Li seems to have concentrated in the region of the

250-keV resonance. It's encouraging to note that the

latest (n,a) measurements agree to the order of 3 percent with each other and that they also agree with

the R-matrix calculation of Hale. However, we should

not put too much emphasis on this. There have been

many measurements over that resonance, and there has

been much disagreement. The reasons for these problems

are still not fully understood, and we should hesitate

to conclude that the cross section is known to even 3 percent in that region. Before proceeding with further

measurements using ^Li glasses, we really need to know

why the disagreements exist. Perhaps some of them are

due to the problems of ^Li concentrations and distribution in scintillation glasses that were discussed

during the symposium. This is a situation that for a standard like ^Li, which is used so much, is very

disturbing, and I think really unacceptable if we're

going to use this cross section as a primary standard.

On the credit side, the agreement between the R-matrix

calculation and the inverse reaction data of Brown et

al. at 0=0° and 180°, which were completed after the

calculation was done and which were therefore not

Included as input to the calculation, is to be noted.

We spent some time considering the question, "Is ^Li a suitable primary standard?" There are people

who put forward the view that, with all the problems

we have had measuring this cross section, it is not a good primary standard. It was concluded that for

essentially, practical reasons, although for other reasons too, it is a good primary standard up to "^150 keV. Above that energy it is questionable.

Now the problems with ^Li and ^ °B are bound up

with detector technology. There have been few real

advances in recent years , and this is an area which

clearly needs improvement if higher accuracies are to be achieved in the cross sections. It was felt that

we're probably fairly close to the limit of accuracy

using present ^Li glass techniques for measuring the cross section and maybe it is time that ^"B was

investigated. Measurements have concentrated on ^Li

recently and ^"b has not had much attention, and perhaps

it's time to look at "'b again, particularly to cover

the region above '^150 keV where ^Li becomes a poor

standard, to 'vSOO keV where hydrogen takes over. This

region between is probably the most important problem

at present. There was also some suggestion that -'He detectors should be considered again, and maybe a fresh

look would prove to be fruitful.

If you want some measurements, which are thought

to b6 the most desirable, we've made a shopping list,

mainly from Gerry Hale. The ^Li(n,a) and total cross sections at thermal energy are still important. We

need to determine these more accurately if we are to know the 1/v part of the cross section better. In this

connection, the levels in ^Li responsible for the lowenergy ^Li(n,a) cross section have not yet been clearly

identified. This is not just an academic problem as their position determines the sign of the deviation

from 1/v of that component. The minimum at about 80 or

90 keV in the region below the resonance is another

Important region where accurate (n,a) and ^ measurements would help to define the cross section'over a wider energy interval and again the cross sections at the peak of the resonance are still relatively poorly

known and must be more accurately defined. Elastic

scattering over the resonance and angular distributions

at low energies are also important . ^Li above 500 keV

to 1 MeV is used as a standard by certain groups and

that's a region where the R-matrix calculation may be

somewhat in error. And finally, it's felt that the

^Li(n,a) to '•'^B(n,a) cross section ratio measurements

above 50 keV, extending up to at least several hundred

keV would also be extremely useful. Thank you.

Prof. Havens: We'll have Alan's summary.

Dr. Alan Smith : I think before I start on the light

nuclei which will include B, H, C and Li, I should outline my position here. I am somewhat awed by standing

before you. I try to avoid making a precise measurement if at all possible, and there are those who say

I never do. I think that maybe the best you can expect

from me is an unbiased opinion, and one good sound reason is that the ultimate in unbiased judgment is

complete ignorance in some cases. But I also think

347
I bring to it a look which maybe isn't as close. And I bring also some age, which is not anything but my biological credit; but I have gone to more of these standard

conferences than have been mentioned in this proceedings.

Somebody forgot the Oxford conference of 1963, he didn't

count back that far. Well, that starts me off because

I was reminded this morning by Bob Block that, at the

'63 conference. Bob Batchelor had a little audit done

by the Queen's exchequer, and his conclusion at that

time was that one million pounds had been spent on the

^Li(n,a) cross section to that date. We're still at it,

and the pound has been devaluated, but even so I think

it must be up to about 8 or 10 million dollars at least,

and the question which I have now that really bothers

me is, "Where are we, what do we want, and how much more

effort is warranted?"

I looked at the new results, I counted crudely,

and I see only a few new total cross section measurements that are really relevant to standards; maybe two.

I see two new (n,a) measurements, and I believe they were both relative. I see two scattering measurements,

several angular distributions of the a particle, and one study of inverse reactions. That's apparently

in the last 10 years. I asked myself what the impact

has been, and I think that the major effect is that we have perhaps a better knowledge of the (n,a) cross section below 150 keV. Beyond that I don't know. I concur with the Workshop's report that this is where

we should focus our attention.

But looking at it from the pragmatic point of view,

I asked myself why we're making all these measurements

at higher energies. And I asked myself another question.

Looking back in history, have we any evidence that

perhaps we know it a good deal better than we think we

do? I look back, for example, at an early measurement

made 10 years ago by two English gentlemen who did a very good job using a simple-minded theory, and they came up with an (n,a) cross section. Bob Peelle plotted

it on a slide here the other day. As near as I can figure out, it's within 1-1/4 percent to 1-1/2 percent

of the current ENDF/BtV. So I wonder what we have

gotten from these measurements. Have we got more confidence and to what degree? I have an uncertain

feeling about how we improve that situation further.

Now, I am a little concerned at some of the statements that we see for what our needs are. You can look

at some of the compilations of requests and you can see

that you want the (n,a) cross section of ^''b to 1/2 percent from 0 to 3 MeV, and X think that's not a very

rational request; I think that's what the workshop said.

I think you may need it to that accuracy to 150 keV,

but I think we should focus on what the range of

interest is.

Some of you know that some years ago I wrote a memorandum extolling the virtues of physical interpretations and theoretical extrapolations for nuclear

data purposes and some resistance to pragmatic adjustments. You also may know I took my knocks for that memo and I'm still black and blue in some places. So

therefore I really welcome the outstanding work at Los Alamos to use established theory in a long-term

way of knitting together this process. But I am a

"•ittle concerned that apparently, if I understood

correctly, after 10 years we are still looking for the

positive parity state in the ^Li system. That, I thinic, was Ernie Rae's search, too, in 1970. I also wonder

about how far we extrapolate our fit. It is, I think,

a parameterization and a theoretically sound one, if

that's all it is.

I was interested particularly in Jack Harvey's

statement, and his very nice experiments, and those

here of Ivan SchrBder at the Bureau, where we now have

a different reaction mechanism proposed to account for

the (n,a) angular distributions. I don't know, I'm

not a theorist, but it bothers me that maybe we shouldn't

get too hung up on one type of theory until we're sure we really understand it. And I think then if you take

that approach maybe we shouldn't worry too much about

the exact energy of that resonance. And I am certainly

a guilty party for probing around in that thing, too.

I wonder if I should continue. I would be happy

to do so, in fact I'd love to tickle some toes, but I would like some guidance. (Granted by Havens) I too share the workshop's opinion that •'^B is probably a more promising material. One of the reasons for this

is that in my happy life I have not dealt either with

B or Li, so after hearing Larry Weston's excellent

talk on the difficulties with the Li detection system,

I decided I never wanted to get involved. Maybe ^"b

is easier. Somebody else mentioned something about the

total cross section of ^°B. I think Bob Block needs

that for correction procedures, and I think he's

correct certainly. But that again reminds me of a statement made by Bob Batchelor, and I think I do have

a positive suggestion there. Bob Batchelor in 1963

recommended the old shell transmission experiment as

a way of getting the (n,a) cross section. And in fact some of the results were shown here. They didn't

come out very good, but I think that was a mistake of

the times in two ways: the correction procedures were poorly done and we didn't know the elasticscattering angular distribution. Maybe it's worthwhile

doing that one again, now using the calculated distributions which we have from our R-matrix theory. Maybe

Batchelor 's recommendation of 15 years ago is still a good one. The question I guess is "Do we have those

Batchelor spheres in the AWRE vaults?"

There are two other light element cross sections,

the hydrogen and carbon, that I would like to say a word about. Lovely experiments on (n-p) scattering

from Harwell and their visitors; I was much impressed.

I'm not a specialist in that area certainly, and it's

been I guess five years since I even looked at that

type of reaction at even lower energy. But I looked

at the curves and so forth and I guess my question is,

"Excellent and beautiful though the results are, do

they really change our assumption of the Breit-Hopkins

expression of the standard?" I didn't see a quantitative statement. Is there real doubt in that basic and

most elemental standard? I'd like to see an answer to that question. I also think that we better give a little harder look at this n-p cross section at lower

energies, and this is partly what the workshop said,

but I'd go lower. And this impacts upon the ^Li. I am impressed by Mr. Czirr's detector which is said to be flat to, I believe, 1 keV to a percent or so, and

that I think takes much of the crunch out of the need

for standards such as (n,a) of Li and B from say 150

to 500 or 600 keV. That was a very impressive detector.

My final light element comment has to do with

carbon. I don't have any real serious objections, in

fact none whatsoever on the paper presented. I particularly though wish to quote again the remark of one man from the audience who, yesterday I believe it was, said that, "If you are measuring angular distributions you better damn well measure carbon; if you can't

get that you're in trouble". I strongly support that

comment. I might also suggest that in my view the

total and differential scattering cross section of

carbon, probably from 1.8 MeV on down, is known to about 2 percent and that is very nearly as good as the hydrogen. That's also an energy standard which, we heard today, is well established. I will extend

my remarks also from scattering from carbon to total

348
cross sections. And some of you may know that total

cross sections are a mess in many areas. I would also

suggest that those who measure total cross sections, even though self-normalizing, use carbon as a reference

check point.

Prof. Havens : I would like to call on some of the

panel members to comment on the summary and the workshop report. Lee Stewart.

Lee Stewart : Let me make one remark on hydrogen.

Remember that Ugo Farinelli suggested the one thing he

didn't like about ENDF standards was that we change

them too often. Hydrogen has not changed since ENDF/BIII and yet I come to this symposium and I find very

few people who use it; the answer is that hydrogen is too hard to implement. Well my comment, having been

in this business for many, many years (26 to be exact),

is we've done the easy experiments, now let's start

and do the hard ones, because we have been measuring

cross sections to standards other than hydrogen for

many years. Hydrogen is the only standard in the MeV

range that is both smooth and precisely known compared

to all other standards; there is no structure. So let's

do use it.

The only other comment I'll make refers to Alan's

comment that we do not need further measurements in

^Li. One problem has hurt us for a long time; there

are two measurements on ^Li elastic scattering at low

energies. They differ by 30 to 40 percent; one comes from Harwell and one comes from Argonne, and we have

never been able to resolve that 30 to 40 percent discrepancy. Therefore we cannot subtract the elastic

from the total to get the (n,a). Thank you.

Prof. Havens : Unfortunately that's all the time we

have allowed for this particular subject. The others

will have to go more rapidly. Yes, it's a railroad iobi

The next session was Capture Standards, Fission Parameters, and Thermal Standards. And since there was no particular workshop which was associated with this, I'll

call on Alan Smith to give the summary at this time.

Dr. Smith : Well I have ten minutes to cover this

collection, and I will move very fast and try to catch

up a bit. There was really only one capture standard

discussed in detail and that was gold. It has its

strong shortcomings in structure at low energies, and

the results are poor at higher energies, but is really

fairly well known, 4 percent or so from 200 keV to 3 MeV,

and it was said there was no discrepancy between the

direct detection and activation measurements. That,

I think, is a change from the past. I guess my real

problem with this general capture area is, "What happened

to the other capture standards. Ho and Ta?" They have

been proposed. Another question is, "Do we know some

of the applied capture processes even better than our standard, and if so, should we neglect the standard?

For example, "Is '^^^U capture, in fact, better known

than that of gold?" If I look in the same energy region

between the ENDF/B-IV and ENDF/B-V, the changes are

less than 4 percent.

I think then I should pass on to some of the other

things that go under the general heading of fission

physics, the fission neutron spectra. Here we had some new interesting results. One has been a question that's

been in my mind for a long time. What is the energy

distribution at very low energies? The Russian results seem to show it's a close fit to a Maxwellian. This

gives me a bit of trouble in the context of some of the

integral experiments which seem to show quite the

opposite, an abundance of neutrons in the low-energy

tail. I think that one must stress very importantly

the acceptance of the ^^^Cf fission spectrum as the

standard fission spectrum as it already is accepted as

a standard for v. I think it is regrettable that in some past meetings the emphasis was put on '^^^U and

2^5pu for pragmatic purposes. The basic standard should

have been emphasized instead. Once you have that, the

ratio obviously suffices for other isotopes.

Thermal standards I will not discuss to any great

extent, the main authority on the subject unfortunately

had to leave. I would only point out that there's one associated problem that bears on this and a number of

other aspects of fission, and that's the foil assay in

fabrication. I think that I have seen an order of

magnitude difference in accuracy uncertainties in our standard foils. The Plutonium Subcommittee is working

in the 0.1 percent range, while the rest of us are still at 1 percent. I think there's a transfer of

technology problem here that_will impact upon our precise thermal constants, our v, and a number of other

areas. I'm embarrassed to say that the speaker from

the Plutonium Subcommittee comes from my own laboratory,

and my foils aren't anywhere near as good as his.

Very carefully discussed was v for Cf. I think

the thing that disturbs me a bit here is, "Are we at the end of the line with present techniques?" There

was one new measurement as I understood it in the last

10 years. The rest are reassessments and recorrections

.

I understand another one which will combine n and v is

in progress or planned. That might be welcomed, but I question whether we should continue those type of

measurements with those existing techniques if we can't

get a real breakthrough. And one thing that comes to mind, in which somebody with much more knowledge of

the field should look at. You have an awfully intense

and sensitive thermal column in the Lucas Heights

reactor, it's an Argonaut reactor. What can you do with

a few precise fissile foil assays a la the Plutonium

Subcommittee and a good pile oscillator in that sensitive

thermal column? Is that a new method? I know, for

example, that very high accuracies are claimed for

weighting of flux distributions in fast criticals.

Perhaps this is some new technique that can be explored,

maybe it's been looked at.

Prof. Havens: I understand that Dr. Boldeman had a rump session on v, and I'd like to call on him. Could

you take the microphone since this session is being

recorded and give a very brief statement of the results

of the discussions on v?

Dr. Boldeman : Richard Smith with the eta bath is

measuring v and, having seen the things that he's

planning to do and the information that he's got

already, I'd be very surprised in fact if he doesn't

turn up with a particularly nice result, whatever that

happens to be.

Bo Leonard suggested to me at one stage that when we

combine the manganese sulphate bath measurements and

the liquid scintillator measurements we should recognize

that there are common errors in each of these. Measurements of a given type therefore should be first averaged

together and then the systematic error applied to the

average. The results from the different sets of

measurements should then be averaged with weights

according to their errors to get the final result.

Well, I thought I would try that, so I did a bit of a quick calculation this morning. Using the conventional

approach the result that I got was 3.745, + 0.010. I expanded the error to this value because I thought

there are lots of things like the uncertainties in the

fission-neutron spectrum which really occur in every

measurement. You might remember Ted Axton showed a slide for a manganese sulphate bath where, for average

energies of 1.39 or 1.43 MeV; there's a 0.1 percent

349
change. So even in the manganese sulphate baths the

spectrum matters. We expanded this error then to that.

So then I thought we would assume that manganese sulphate

baths have a common error and liquid scintillators have

a common error. I therefore averaged all liquid scintillator measurements together and separately averaged all

manganese sulphate baths together. These two values were then averaged with weights given by their

uncertainties to the boron pile and the Fieldhouse

measurements. The result now became 3.748 +0.01. So

it doesn't strike me that anything you do with the

measurements or the way which you write them will make

very much difference. If I were to guess, as Lee

Stewart sometimes thinks it's reasonable to do, I reckon the final answer is about 3.75.

Despite the very acceptable level of agreement

between the different absolute measurement of v for

^^^Cf, the question of a possible systematic difference

between the MnSoij. bath determinations and those obtained

using liquid scintillators, and therefore the real

accuracy of the mean of the measurements, continues to concern some evaluators. It should be reemphasized

that of the eight absolute measurements only one lies more than one standard deviation away from the mean. Furthermore, the averages of the liquid scintillator

and MnSoi^ bath values differ by only 1-1/2 standard

deviations of the comparison. Consequently, their

fears are unfounded and there is no evidence to postulate an experimental technique-dependent error.

I'd like to take up that thought of oscillating

things in and out of reactors; maybe somebody can do

something about that.

Prof. Havens : I'd like to then call on Dr. Liskien to comment on the summary and Dr. Boldeman's results.

Dr. Liskien : I think I have nothing to comment on this,

but I would like to remind the Symposium that we had

in this session and this group of subjects two potential

newcomers as standards, namely ^SSy and 237]^Jp_ ^ person asks me why we do need these cross sections, the answer is of course yes we want to conveniently measure

fluxes. Then he will certainly answer yes, OK, so I understand you need one cross section for the whole

energy range, but in fact I have to explain that we

need perhaps more. Now comes the point that is suggested

that we add again at least one of the two. Of course we all know the reason is that we don't want to have a thermal response in the standard. But my real point is

that when I remember correctly that the interest for

such a standard is coming from the reactor In-pile

dosimetry, and I wonder if this is also true now for

2^^, if what is really wanted is not the total fission

cross section for these isotopes, but, in fact, the

partial production cross section for the production of

a convenient fission product because many of these

measurements are done not on-line by counting fission

fragments but by off-line activity measurements. I would like to see in the future this point clarified;

really, what is needed there?

Prof. Havens : Any other member of the panel like to comment on what is needed? Dr. Peelle.

Dr. Peelle: I'd like to comment in terms of another

question which has concerned me. In my reading of many

papers on the subject of the neutron spectrum from

fission, I've never seen a really good analysis of what

properties of the spectrum are actually required. Is the average energy enough? What parameters, what

moments have sensitivity in the applications? And if we knew that, we might know better how to deal with,

how to parameterize, and how to seek better accuracy

in fission-spectrum measurements.

Prof. Havens : Would you like to comment on that?

Dr. Farinelli : I might make a comment on this last

question. Well it depends very much on the problem.

For instance, the high-energy tail of the fission

spectrum is very important when you use threshold

detectors with a high threshold. And why do you use

these detectors? For instance, because you are interested in high-energy neutrons, if you have to simulate CTR conditions in a fast reactor. So this

may explain why the status of the fission spectrum

which was acceptable at high energies up to a certain

time ago, perhaps a couple of years ago, is no longer

satisfactory, and it's certainly one application for

which you would need better resolution in the highenergy tail.

Prof. Havens : We have about one more minute. Lee.

Lee Stewart : Another remark is that Jud Hardy did some calculations in a thermal reactor and using the same average energy but just changing the shape, from a Maxwellian to a Watt, he got differences of 0.3 percent in k. So if you harden the spectrum, of course,

you get a change in k. Just measuring the average

energy in some configurations is certainly not enough; we need to know the shape and whether the shapes are different among the different isotopes.

Prof. Havens : I'm sure the problem is different for

shielding than it is for k and therefore the answer to the question is you want to know as much detail as possible in the final analysis.

The next topic for which there is a workshop, which

is closely related, that we divided the conference up

into was Flux Intercomparisons . I will call on Dr. Axton to present the results of the workshop he conducted

on the future of BIPM flux intercomparisons.

Mr . Axton : First let me state what the object of the

inter comparison is and the criterion for judging whether

or not it was a success. The object of the Intercomparison is to identify systematic errors in absolute

measurements of neutron flux density. There are two

types of measurements involved in this comparison;

firstly there are the absolute measurements of flux

density made by different methods at different laboratories, and secondly there -are the comparison measurements between those absolute measurements which are made directly as part of the comparison. The criterion

for success is that the errors of comparison must be

small compared with the absolute errors, otherwise you

know nothing about the systematic errors in the absolute measurements. The intercomparison falls Into ten

subgroups because there were five intercomparison

energies and there were two transfer methods for each

energy. The conclusions of the discussion group were

as follows: (1) Some transfer methods were good and

other transfer methods were not so good; (2) When the

transfer method is good the results are encouraging.

Flux measurements by different methods agree and flux

measurements at different laboratories agree. For

example, associated-particle techniques and proton

telescope techniques agree at 14 MeV. At 2.5 MeV,

associated particle counting, the stilbene crystal

spectrometer, proton telescope, the manganese bath

and the proton proportional counter all give very good

results. When the transfer method is bad, then the

information we obtain is "that the transfer method

is bad." Conclusion (3) is that we should carry on

and improve the comparisons, in other words, do some

more. (4) Future comparisons should satisfy both

dosimetry interests and cross section measurement

350
interests. Conclusion (5): The energy range should

be extended down to 144 keV to include reactors, and

this energy is still feasible with Van de Graaffs.

The range should also be extended up to 50 MeV for

cross-section interests. Conclusion (6): There is a need to improve transfer methods. Conclusion (7):

Cross section measurements should be included as transfer methods. This would allow participation from

the white spectrum linacs and also those from the Van

de Graaffs. There is need for a study period to iron

out the problems involved in this particular aspect of

the comparison. Thank you.

Prof. Havens : Alan, will you summarize this part of

the comment of conference?

Dr. Smith : I only have a couple of remarks on this

one. First of all, I'm happy to see that the concept

of a transfer basic physical standard of cross section

has entered the program. One of my problems with this

type of measurement over many years is that the

detectors tend to be environmentally associated or instrumentally associated, and therefore they're not

suitable for many places. The second thing is that I am impressed by the real good agreement. I think it's

quite good, and I think it was Caswell who pointed out

that the similar dosimetry comparison was almost

identical. I think they did quite a good job.

Prof. Havens : Dr. Caswell, will you comment?

Dr. Caswell : I would like to make a couple of comments,

one of which might involve some interaction with the

audience. The one question that the people who

organized this have wondered about is that this has

been essentially an intercomparison between Van de

Graaffs and Cockcroft-Waltons . The white source machines have not been involved although there were

earlier some expressed desires to have them involved.

I think that people who are involved should be very

interested to know to what extent the white-source

people would want to participate and what is the

mechanism envisioned. A very precise measurement of

a specific cross section, preferably one that is not

well known, could play the role of an intercomparison

in this way.

Secondly, we have used the word "BIPM" , and I'm not sure everyone in this audience knows what organization

this is; but I thought I'd take a minute to explain.

When you take an elementary physics course, in the

old textbooks they would have a picture of the

standard meter bar that sat in some institution near

Paris, and also of the world standard kilogram that

also sat in some institution near Paris. This

institution is, in English, The International Bureau

of Weights and Measures, in French the Bureau Internationale des Poids et Mesures, and it was set up originally by the Treaty of the Metre in 1875, at that time as the keeper of the artifact standards

like the meter bar, the kilogram, and so on. There's

only one artifact standard left. For example, the

meter bar is passg; it's now a wavelength of light,

and the one that's left is the kilogram. Incidentally,

there have been some interesting troubles recently

discovered in the transfer of information on the unit

of mass where the effect of the density of the air^ on the balance was wrong. So even that field hasn t settled down.

Prof. Havens : Would any other panel member like to comment on the flux intercomparisons? No one. I'll

proceed to the next sub-category. Personnel Dosimetry,

Biomedical Needs. Prof. Barschall, would you give

the results of the workshop on that subject?

Prof. Barschall : For biomedical applications, the

quantity which one wants to know is the absorbed dose.

Let me repeat from this morning what we mean by absorbed

dose: the energy absorbed per unit mass. There are

two types of instruments which can be used to measure

directly absorbed dose. One is a tissue-equivalent

ionization chamber and one is a calorimeter. There was agreement among those working in the area of personnel

protection and those working in the area of therapy,

that it would be extremely important to be able to have an ion chamber or calorimeter calibrated at the

National Bureau of Standards and certified by the

National Bureau of Standards as to the dose calibration

for neutrons of these instruments. There was agreement

that calibration should be available for two neutron

energies; one in the energy range of the fission neutrons,

and this might be done with a Cf source, but there's

also a need for a calibration at higher energy. For

this purpose, the most suitable source appears to be a

14 MeV d-T neutron source. I should like to emphasize

once more that the interest is in the calibration in

terms of dose per Coulomb rather than in terms of

fluence per Coulomb, which is why perhaps the National

Bureau of Standards might prefer to make the calibration.

Now since actually most centers, especially those active

in therapy, use continuous neutron spectra, if not fission spectra, they'll have to make corrections for the

energy dependence of the kerma factor in order to apply

the calibration which they might get from the Bureau

to their particular spectrum. For this reason, there's

an important need for a better knowledge of the energy

dependence of the kerma factors, particularly of carbon,

nitrogen and oxygen. It is not sufficient to know just

the cross sections for various kinds of processes; what

is more important to know is the charged-particle

spectra which are the spectra of the charged particles

produced by various kinds of processes, and so this is

the quantity which should be measured. And this information is not only needed to evaluate the biological

effects of the neutron; it is also needed to calculate

the effective value of W, the energy per ion pair, for

the mixture of charged particles that is produced by a given neutron spectrum. High priorities should be

given to determining the charged-particle spectra for

the 14-MeV neutrons, not only because the instrument

should be calibrated at this energy, but also because

it would be most desirable to tie down the kerma factors

at a particular energy in the neutron energy range of

interest in therapy. A knowledge of the kerma factor

is also important to calculate the dose delivered in

therapy to various organs which have different chemical

composition.

Prof. Havens : Alan, would you comment on this?

Dr. Smith : I only have a few brief remarks because

we had agreed that Randy is far more cognizant of this

field than I ever hope to be. I, however, have had

occasion to observe the Bureau's performance in this area for many years and I must say I am impressed by

the magnitude of this problem and the impact on society

and I don't think that's what comes through to people

out here. It came through to me, sitting here watching

the Bureau trying to handle it and its a very difficult

problem and I think the problems in personnel protection

for radiation dosimetry are fundamentally ones of

public implementation, technology transfer and assurance and these are administrative management social

problems but they are enormous in scope and I think

this should be recognized. I don't think that there

is really a basic standard need for that one. On the

bio-med dosimetry problem I had some troubles, really.

I was awestruck by the accuracies that are quoted. I am just glad that I'm not in a tumor hospital with

someone probing into my innards with a 5% accuracy.

What bothers me is, "Has anybody measured a kerma to

351
5% in an ideal laboratory situation of a dummy torso?"

Can it be done with the current technology much less

in a clinical application? I have trouble again with

this debate that apparently is going on, what can we

do with standards and how are we going to clinically

utilize them? I think there is a great deal of

ambiguity there and I have one final other remark.

I think that from the point of view of my interests we

should be rather specific in what we want, I saw these differences in calculated kermas which if I understood correctly were attributed to differences

in carbon cross sections used at Livermore and the

Bureau of Standards. What differences in the carbon

cross section? What specifically is causing the

trouble? I would like to see quite definitive lists

if I could, of say the ten most wanted things to characterize the 14-MeV biological dosimetry problem.

I agree the cross sections were bad. They've got to be determined and stated in the sense the physicists

understand, not values like biological effective

something or other but I would like to see a list.

What is the specific set of ten cross sections? Then

maybe I could do something about it.

Prof. Havens : Right now. Randy, would you care to comment on these?

Dr. Caswell: Well, first I'll send Alan a list in a few days and then I will go back to some observations

I wanted to make earlier. First, I have had the good

fortune I think to be at four out of five neutron

standards meetings mentioned. I think this is the

first time that any physicist or other representatives

of the biomedical community have really been in

attendance so I think we should recognize we are just

beginning a dialogue between the neutron standards

people and the biomedical people who are concerned

with neutrons. We need to learn to speak each other's

language and I would like to congratulate Prof.

Barschall on the choice of elements that he presented

which are really the essential guts you need to know

to go from one to the other. So, his paper and Dr. Broerse's paper were both very nice from the standpoint

of general introductions to the field. On the cross section question, I think that although there are problems in the values of the cross sections from time

to time, the much bigger problem is that ENDF/B is an incomplete set. In general, it does not give you the

charged-particle energies which are what you need for an energy deposition calculation. It is much more

likely to give you neutrons than gamma rays so one way of playing the game is to take the energy available,

subtract off the gamma-ray energy, subtract off an escaping or scattered neutron energy from a reaction

and then say the rest has gone to charged particles.

Well, how well you can do this depends on whether you

are given both the neutron and gamma ray energy so

for the kind of application we are talking about the

biggest problem is really definitive secondary chargedparticle energies. The elements in which the cross sections are needed are in carbon, nitrogen, and oxygen.

An example that has come up several times in neutron

dosimetry meetings is: There is a ^"^CCnjn'Sa) reaction

which is known and measured. You would think therefore

there might be an -"^ ^0 (n,n ' 4a) measurement and in fact

it's got something like a 14- or 15-MeV threshold. To

my knowledge, no one has ever measured it even once

at one energy so that we are really in an area where

there are lots of cross sections that are not known.

I guess I'll make two more comments and then shut up. One is that in the personnel monitoring area I think

accuracy is required which is easy for the neutron

standards people. There is still an energy gap

between thermal and 2 keV where there are lots of

neutrons that people are interested in and where

Auxier showed that successive intercomparisons improved

performance. I think in his case that's true; it may

not always be true in successive intercomparisons. It may be just that the field is getting better and intercomparison had nothing to do with the agreement but I think that the case shown by Auxier is clearly one where

the existence of the intercomparisons improved the

measurement performance in the field.

Prof. Havens : Any other member of the panel like to comment on it?

Lee Stewart : At Los Alamos, we have had some interest

and expertise in calculating kerma factors and have

found, in addition to the problems in the ENDF/B cross sections, problems in the calculational methods. Phil

Young, who is in the audience, can perhaps tell you how

not to calculate kerma factors. As far as the data

files are concerned, we have found that some ENDF

materials make very good refrigerators in that the

calculated kermas are often quite negative for many of

the individual reactions. I have asked other people

about their calculations and the Livermore Group, for

example, told me they just set a negative kerma to zero, wherever negativity occurs in the reaction calculation. Therefore, some of our problems are in the

ENDF/B representation of the data and others may be

traced to the calculational methods employed.

Prof. Havens : Shall I call on Phil Young to tell us how

not to calculate kermas?

Dr. Young : Well, we looked at several elements in ENDF

and tried to estimate what the kermas were by doing

what Randy suggested. Taking differences between total

energy, subtracting out neutron energy and gamma ray

energy, and frequently we would get things like negative

numbers. Energy is not conserved accurately enough in

the files to get kermas by subtraction techniques for

many of the elements, mainly for medium and heavy

nuclei. That was the conclusion.

Dr. Peelle : I have formed an impression based on looking a little bit at the NASA work on this question

some years ago as well as what's happened more recently.

I think the personnel dosimetry problem, as one goes to higher neutron energies rising above successive reaction

thresholds, is a very nice application of nuclear

physics and related phenomena—related fields to some practical matter and a very difficult one especially

as one gets above 15 or 20 MeV where I believe to do

it right requires physics we haven't even figured out

yet. So it's a challenging problem; its not just a matter of throwing off and transferring the information we really have to solve a practical problem. It's

going to require some deep thought and is a very

interesting as well as an important problem.

Prof. Havens : Any other comments?

Dr. Caswell : There were remarks made in some of the

talks that the standards laboratories are not offering

calibration services for biomedical application and

therapy. I would like to say they are beginning now. NBS has been doing it for a little while. PTB and N3S are sort of in the status of beginning to offer known

monoenergetic fluences for calibration. I think this

is a step forward. I think doing the business for the

therapy dosimeters at the high accuracy and following

multiple methods to get the same answer has not been

done and I might comment that we have two proposals

which we had going for two years now neither of which

has gotten anywhere to do that.

Prof. Havens : Well, we'll proceed now to the next subject which is "Benchmarks, Core Dosimetry"; Chuck

Weisbin, would you report on the results of your workshop?

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Dr. Welsh in : Let me apologize in advance to my

colleagues if I misinterpret or leave out any of the

comments. The discussion was quite lively on the interaction between differential and integral data. There

seemed to be in our workshop essentially unanimous

agreement that integral data must be considered in the

process of cross-section standards definition. In order

to achieve this worthwhile goal, important procedural

issues must be addressed including which integral data,

that is, what constitutes a standard integral benchmark

and how is it to be included in the evaluation process.

The methodology available for addressing the subject is

fairly demanding in that it requires uncertainty information including correlations for both the integral and

differential data and there I don't just mean standards,

I mean all differential data. It expects quantitative

estimates of possible methods bias as well as computed

sensitivities of integral data to the various ENDF/B

cross sections and their associated uncertainty files.

Will they both represent only the differential data or will they both have to represent all available information? The characterization and definition of what is an uncertainty file may be difficult. It is clear that

the methodology is becoming available but significant

advances must be made in order to provide more credible

input. The methods developed must be sophisticated

enough to provide specific recommendations to an evaluator, for example, in distinguishing between

modifications to the cross-section normalization and

shape. In developing such methodology and data the

very practical question is raised as to what kind of

time frame would be required for an endeavor of this

scope and indeed is it worth it. The process of assuring

compatibility is clearly iterative and desperately

needs documentation to improve best the communication

between people, the real barrier, rather than fundamentally conflicting data. A strong driving force,

perhaps much too strong in the direction of reconciliation of differential and integral data, would be to insist that a standard can only be so designated when

it is judged to be consistent with all relevant information—both differential and integral.

Prof. Havens : Thank you. That may rule out all

standards. Alan, would you summarize this?

Dr . Smith

:

I would like to back up a bit to the benchmark fields particularly those under the auspices, I believe, of the Inter-Laboratory LMFBR Reaction Rates

(ILRR) program. I think this has been one of the best

correlated national data efforts where people have really

gotten together and through cooperation have come up

with some very fine results. I heartily concur with

what has just been said that these benchmarks in the

real world of microscopic data are an essential part

of the whole process. I don't think it's a goal in

itself and I am worried that we have really not a measure

of the sensitivity of the benchmark measurement in such

things as the ISNF facility. It came out in the last

paper this morning. There were some crude sensitivities,

but it seems to me that's the first place to apply the

sensitivity analysis. That's a first step toward the

integral test and I also quarrel with an opinion which

I've heard, and it wasn't expressed at this meeting,

but I think I should express the view. A benchmark is

a benchmark that tests some underlying principle; it is

defined so in the dictionary. I think that you should never conceive of a benchmark as a method to measure

microscopic data and I think that is something that is

done periodically and I've had some words on that in the

past and couldn't leave the opportunity go.

I think too that I could not, in this vein, quarrel

with the Farinelli syndrome which is I guess to, "build

it and try and fly it". If a cavity and testing it

all out in the benchmarks is going to be the way to an engineering goal, then I don't think I am in a position

to challenge but when you say that the benchmark is going to give you microscopic basic information, that it

won't do.

I was impressed by the interest in core dosimetry

both at and in the core and beyond the core but I was

left a little bit with the feeling that the accuracies

that are required are really not all that great. That

the people that are involved in that work, particularly

the commercial people—their needs are mostly met. There are some "fine structure" things remaining but

there are no gross holes.

Quite a different area was the fission product,

the burnup control, the analysis of core performance,

and beautiful magnetic-spectrometry results from Idaho,

and the excellent correlation of the chemical yield

work here. Clearly the mass fields are changing

violently with energy and I don't think it was fully

appreciated what the impact of these, which I think

really are proper fission-product standards, might

be. I seem to recall that the way of monitoring the

basic number of fissions in some of the best delayedneutron experiments in the world was a counting of

merely a hundred products. If that is true and they

varied with energy, 1 question how good those delayedneutron yields are now in view of these recent spectrometry results. The impact of those 1 think was very

great. 1 think the Cf fission spectrum, as I mentioned

earlier, should be accepted as a standard but, I got the

general feeling that the averaged ratio rates given in

various standard fission spectra, are running about two

to three percent for Cf and that's probably not good

enough. We are looking for another factor of two.

Now, again with the previous speaker, I concur with his error-file concept. It's a wonderful world,

but there are several things that begin to bother me a little bit. I wonder if the sensitive things are not

the standards. It seems to me that in some of the

examples shown yesterday, the fission cross sections,

which were the standards in the Godiva and Jezebel

assemblies, they were not sensitive really. The other

holes were those holes of uncertainty where we're

guessing. What is the inelastic cross section of 235^?

Well I'd like to know and I guess some other people

would like to know and that is a very sensitive item

and how do I place an error on that aside from saying

I don't know. I think that you've got to be a bit

careful on how you specify these error files.

I also am a bit surprised that somebody hasn't

given more emphasis to where you assign these uncertainties. Certainly the measurers and evaluators have

simply got to give a great deal more attention to their error definition. Discrepancies are rampant

throughout; people are just not representing things

completely. It is a little like Ben Diven says, I guess, you correct it and put an error on everything

you can think of but what do you do about the things

you can't think of. Well you can't think of everything so people get ultra conservative but I think too that you might give consideration, particularly in

the benchmark area, rather than assigning your

discrepancies or your uncertainties and correlations

to the microscopic data, follow the English course,

as I understand it, and work with the correlated error

files on the multigroup cross-section sets.

Prof. Havens : Dr. Farinelli would you comment on the

subject?

353
Dr. Farinelli : Well, I think Randy Caswell mentioned

before that this was the first Standards Meeting in

which there was an attendance and the presence of biomedical people. To a certain extent I think it was

the first meeting in which there was a fairly consistent

representation of integral people as well. My opinion

is that dif f erentialists and integralists are much

closer today than they were some years ago and that we

are now starting to speak a common language. We seem

to be able to understand and the fact that we have

lively discussions is probably a representation of the

fact that we have at least a common understanding - a common language. Otherwise, we couldn't even discuss.

I think it is very important that we develop a common

basis for understanding. It's quite clear that we must

have some common way of looking, for instance, at errors,

at error correlations if we want to assess what is the

actual reliability of the data when they are applied

to practical calculations.

I don't think it is the best procedure that

diff erentialists and evaluators just well take their

own responsibility to a certain point and then they

leave the whole question to the others and say, "Well

it's your business from now on." I think there should

be some sort of hybridization of integral and differentialists to evaluate the final errors and so on and

there are some other areas which we have not had time

to mention but which are certainly interface areas and one of them I think is the area of how to make group

cross sections, how to process the nuclear data. The

way in which you process nuclear data involves both

the way in which you have represented the nuclear data

to start with and the way in which you are going to use

the group cross sections and this is an in-between area which I think may be the cause of some of the discrepancy

that we still observe.

Prof. Havens ; I guess we have come full circle; when

the reactor business first started it was the measurers

who were the cross-section evaluators as well as the

reactor theorists. In the Fermi group back in Columbia

in 1940-1941, and then as time went on, we developed

various specialities and were so interested in our specialities that we didn't talk to our nearest neighbor.

We then find out that it is impractical and then come back again, but in the original days of the Manhattan

project there was no need for communication because

the people that did the measurements also used the

measurements. They made them because they had to have

them immediately to proceed to the next step, so I'm

surprised sometimes at what the graduate students in

both physics and nuclear engineering know and I'm also

surprised sometimes at what they don't know. Would

any other member of the panel care to comment on this

subj ect?

Lee Stewart : I just might answer for some of the people

here who have done the work. In CSEWG we have made

many, many code comparisons to insure that we are processing the data in the same way or at least getting

almost the same answers and I mean very close, depending

on the methods that we use. There has been a great

deal of work in this area in this country in CSEWG.

Prof . Havens : Any further comments from the panel?

Dr. Peelle : I have heard Chuck Weisbin give a talk on

the subject of methods uncertainties and I'm sure

Dr. Farinelli has made talks on that subject. Maybe

we could hear just one comment which relates to this

thing very clearly. How big are the uncertainties

liable to be in calculated quantities, using the

transport theory?

Dr. Weisbin : I don't know. Bob, if you're asking for

a number or just a qualitative statement. Number one, they are not negligible. That's the first thing. I think I concur with Lee's comment that they are getting

better, and there really are two types. There's the

type of methods bias that we can know and pin down,

for example P^ versus P^, or 50 groups versus 100

groups versus 200 groups. There are harder methods

biases like taking a three-dimensional real problem

and modeling it in 2-D, and saying what is the error

due to our modeling when that is the best we can do.

So now I go back to Al Smith's comment what do you do

about the errors that you don't know about and of course those are harder. I would say that in general

if you wanted a number though; well, let's take the

number, k. I would say that methods errors today in

the calculation of criticaUty are not less than three

tenths of a percent.

Prof. Havens : Do you agree?

Dr. Farinelli : I concur.

Prof. Havens : I remember one time when there were

about five measurements of the ^3 5^ cross section at thermal, all of which were good to two percent and all

of which varied by about 15%. They were poor measurements because the errors had been incorrectly determined .

Prof. Havens : We'll proceed on to the next subject

which is rather a difficult subject to bring together,

"Techniques and Methods." However, there was a workshop which is close to that although not exactly, that's

the first on the list of the workshops, "Establishing

Neutron Energy Standards," and would Dorien James please

give a summary of that panel's discussion?

Dr. James : Perhaps I could say again that the work of

the INDC subgroup in selecting a list of narrow

resonances simply for use as energy standards could be

regarded as complete and the list is given in Table 7 in my paper. The next stage is to evaluate the data

available on all these resonances and establish the best

values, and it is hoped that workers on each spectrometer

will now and in the future publish their best values for

these resonances. Mughabghab and Bhat have suggested

to me that the list of best values should be published

in the next addition of BNL-325. I've agreed to bring

these results together. Now Joe Fowler suggested that someone should write a review article giving all the

details of all the measurements involved in deriving

the best values. He twisted my arm into doing this.

He threatened to tell Basil Rose, if I didn't do it.

You must wipe that out, Charlie. I guess you've got an expletive delete button on these federal tapes that you

are producing. Now this would, I hope counter Alan

Smith's strong objection to using any data as a standard

which is not fully documented. I agree with that

sentiment but I think that if we demand of all the measurers that they fully document their results, then

maybe we would have to wait a long time. It looks from

the evidence that white sources are now in fairly good

agreement except perhaps the Geel-Columbia ^^^U

discrepancy which I mentioned this morning. The careful work of Meadows has shown that white sources and

monoklnetic sources can be brought into agreement also.

Now many white sources have certain substances either

always in the beam, for instance, ORELA bind the '^''b filter with sulfur so they always have sulfur in the

beam and on the synchrocyclotron we use background

filters so that our measurements can have silicon oxide

in each run. These enable certain resonances on the

list to be continually checked and it is suggested that

the values obtained for these resonances should be

included in the publication of any data. Anytime you

354
publish data you should also report your value for the resonances on the list, just to confirm that you are

still checking. Finally, there might be a slight

problem with poorer spectrometers because they may not

see all the narrow resonances on the list, but in time

of course they will be able to use the energies of

broader resonances given by the laboratories which do

quote their values for the narrow resonances.

Prof. Havens : Thank you. Alan, would you put together

the summary?

Dr. Smith : I regret the action Dorien has taken because

if he is going to write a comprehensive review with all

the errors, I'm not going to see a good friend for 18

months, I guess. You're out of business, but I think

it is certainly a proper thing to do and I encourage

it. I think at the higher energies which I'm familiar

with, the whole problem has essentially gone away. I think that the remarks this morning that there are discrepancies but with an exception of perhaps one fission cross section, for most practical reasons, they are negligible. What really bothers me Is that for

this example, and I think also one dealing with the fission neutron spectrum, considerable effort has been

expended because of one or two measurements which

turned out to be a little in error. I think one should

be careful before you immediately take some action that

you carefully review any standard measurement and try

and find out if there is a problem or something. It

is kind of a double-edged weapon. There was a great

deal of effort put on fission cross sections and I Include fission spectra and I think that was very

warranted; we didn't know much. We know a lot more

now. But in fact the disaster that we thought was

there from some preliminary estimates was not there.

I think there is a word of caution not to get carried

away when somebody has a preliminary result on some of

these standards that goes somewhere different. Take

a good look at them, before everyone takes off and

makes some measurements.

I would like to give justice particularly to all

the people who talked about instruments. If I have any

competence, it is didling with instruments, and I realize how difficult and how complex these things are.

It was a very diverse discussion here with many contributions, and I can't begin to do justice to them. There were many of them that were very elegant and many of

them were carried to very high degrees of accuracy; but

if I recall, only two of the instruments were not

described at the last standards conference. Really what

is talked about is refinements - very hard to get and

very well done but not a really new instrument with two exceptions which are really modifications of other

schemes. One that I mentioned before was the flat

counter from Livermore which has a tremendous potential

for getting over this gap between the (n,a) resonance

and one MeV. I think one of the great advances in the

instrument area in the last 10 years which I never would have believed 10 or 15 years ago, is the ability

to use the hydrogen recoil counter reliably at low

energies. Just a technological achievement, but if

somebody would have told me in 1970 that you could

monitor a beam in a linac with precision at 50 keV, I would have laughed. It's done and I think that it is

done well.

I think certainly that the sharp improvement in

reference fields is a real key to the dosimetry problem

particularly the very high purity fields which have

the peculiar problems for dosimetry that they have

here at the Bureau. I am not sure their impact was made clear; you may not find those so valuable for some research purposes, but it is clear that they are essential to the proper calibration of dose. I think

we still have a big hole - the one thing which I have

always been looking for - a flat detector with a high

sensitivity. It's the crux of a lot of problems and

I don't see it coming. For example, how do you solve

the delayed-neutron issue which has been a very serious

bottleneck for a number of years.

Prof. Havens : Thank you. That reminds me of one time

a navy Admiral was trying to describe everything that was desirable in an anti-submarine detector, and I think

it was Panofsky who got up and said, "Admiral, we can

only use nature we can't coerce it." Seems to me that

Alan is trying to coerce nature into getting the ideal

detector. I don't know how we can do that.

Dr. Smith : I think that we should point out that one

of our medical people wanted a spectrometer which

would go from one kilovolt to 20 MeV with an absolutely

determined response so they have their problems too.

Prof. Havens : We all have problems , but we can only

use nature, we still can't coerce it. Dr. Motz would

you take the lead on this?

Dr. Motz : Several other comments have been made about

detectors and the fact that perhaps boron detectors

deserve more attention and Lee has mentioned that

hydrogen recoil isn't used as extensively as one would

expect. Another detector that I think has hardly been

mentioned is the ^He counter. It has some disadvantages but it certainly has a good predictable response.

It's a question of sensitivity and packaging but

certainly one that deserves some attention as a standard reference.

I am very pleased to hear that Dorien is going

to document the energy standards and I wish him luck.

I hope he can come up with the enduring values that

he was trying to explain to us during his talk. I would be especially interested in seeing that, Dorien,

because I have some degree of scepticism about the

high degree of accuracy that is claimed in the final

results. I think that it does take a careful analysis

to convince one that they are really known as well as you were indicating.

I wanted to make one comment about high-energy

neutron sources. With respect to the presentation on associated-particle methods, it was mentioned that a lot of the defects which occur in neutron sources and

associated-particle work are really associated with

the solid targets that are used and many of these

problems go away when one uses a gas target. There

are in fact two recent reports out about purity of

8-14 MeV neutron sources considering most of the

parameters that one could change such as beam stops

and collimators defining beams and things of this sort. This work has been done at Los Alamos both for the

t-P and the p-T sources and compared with d-D source

and the second report is from Bruyeres-le-Chatel in

France, similarly for p-T sources.

Prof. Havens : Would any other member of the panel like

to comment on this subject?

Dr. Liskien : In the field of standards one has to use

any occasion to check consistencies and I see a broad

field where this is not fully exploited and I mean

linac measurements which typically use shape monitors

and then go from the eV to keV region to normalize,

using hydrogen telescopes or similar devices as shape

monitors. I could imagine that it must not be too

difficult to go one step further from a shape monitor

to absolute recoil counting and thereby provide

additional information. I think nobody should be

afraid that other inconsistencies could come out when

you have two methods of normalization. Another point

355
is that I think in fact we have not really seen many

new techniques in this symposium. I think one, not a new method, but at least a way of applying a technique

has been demonstrated and I think it should be mentioned.

I mean associated-particle counting, not the particle

itself, via its activity or by counting but really by

its gamma-deexcitation as Dr. Brandenberger has demonstrated. I think we should see the people that are doing absolute flux determination, and inquire what

they really can do with this method with other sources.

Prof. Havens : Any further comments from members of the

panel?

Dr. Peelle : I would like to just accent, another time

the use of diverse techniques. We have heard in many

contexts, that this is the key to long term success

in the standards' field. One example is Dr. Knoll's

paper in which he described the work he and his students

have done at a few energies with y-n sources. That's

not a new idea but it appears that it is being pushed

in a logical way to give highly credible results which

seem very independent of most of the other work that

is done, and therefore very valuable.

Prof. Havens : Lee Stewart do you have any comments?

Lee Stewart : If I may just mention a situation at Los

Alamos many years ago; a young scientist started working

with the n-p cross section and he accordingly calibrated

his detector. Then he decided to measure the hydrogen

cross section with this calibrated detector and he was

quite discouraged when, at some angles he could not

reproduce the hydrogen cross section better than

approximately 15%. Naturally, he went back to the

drawing board. Therefore, I believe the business of

intercomparison is extremely important.

Prof. Havens : Any comments from the audience? Then

we will proceed to a subject which has been controversial ever since this field started; that's the ^SBjj

fission both at low and high energy. Now Dr. Cierjacks was the leader of a discussion group on the '^^^V cross section. Would you give the summary. Dr. Cier jacks?

Dr. Cierjacks : The fission cross section of 235^ ig

in quite good shape presently, nevertheless there are

a few problems left. As a result of the discussions we had in the working group, we came up with a few

observations and recommendations which I should like to read to you.

Thermal Value: The group felt uncomfortable about the

fact that evaluations gave about 0.3% accuracy for the

actual thermal value while deviations are somewhat

larger than the quoted values. Coming now to the

situation of the 9-eV resonance region. Here we felt

that more evaluations and more measurements are needed.

The data in hand when published and evaluated could

well result in a +1% accuracy of the fission integral

in that resonance region. Let's go on to the remaining

eV range; there we have two statements. First, another

normalization point in the resonance region should be

chosen to allow for a cross check of the data. A

suitable resonance range would perhaps be between 33

and 41 or between 41 and 60 eV. Second, an additional

reference band would be desirable in the range of perhaps 0.3 to 1 keV. With respect to the keV range, the

following recommendation was made. Absolute whitesource measurements with adequate accuracy ought to be

encouraged in that region. Finally we had a recom- mendation for the MeV region which might not necessarily be limited to ^^^U. Consideration should be

given to fission reference cross sections in the range

from about 10 to 40 MeV.

Prof. Havens : Alan, could you give your comments on

this subject?

Dr. Smith : As many of you know, eight or nine months

ago there was a special workshop conference at Argonne

on fast fission cross sections of U-235, U-238, and

Pu-239. All the available data were reviewed. There

is a book over there in the coffee room. One of the

outcomes was that certain areas were identified as being critical, other areas were in good shape, and we

hoped that as a result some work could be done on those

selected problem areas. In fact some was, and some is

still in progress. I am pleased to say that I think

that was an effective result of that conference, and some of it is reported here. I think as a result of

this work, depending on how you want to view your

various interpretations, the fission cross section of

2^^U from a few keV to less than 2 MeV is known to about 2 percent. Up to 8 MeV I would estimate 3 percent, at 20 MeV I would estimate 5 percent. That puts

it within the requirement, for example, of the biomedical people. One of the issues at the conference

was structure in the fission cross section, it's there.

There were examples shown at this meeting where careful

search shows that it isn't particularly correlated in some cases, in some cases it is. I think the issue is

not really important at the present time. I think that now the thing to do is to continue the work, which the

previous speaker emphasized, to try and get a normalization of the white source measurements using an absolute calibration. I am very leery of this lowenergy extrapolation. I think that the consistency

with this and the higher-energy normalization has got

to be achieved. A gap exists, and until you can get across it, I think you've got a real problem. The

work is particularly going on here at the Bureau in

that area, and I guess other places, and I suggest

that these people get together and they keep in better

contact with one another and that they attempt to put

together a preliminary evaluation of these results in

a better correlated manner. I agree with the opinion

that was quoted at the conference at Argonne that I don't think any formal evaluation should use unfinalized or unpublished results. I think preliminary

results are out. I'll extend that statement a bit

farther, and I think I already made it today, that no evaluation should be distributed for final use or considered final until it is fully documented too. And I think a lot of the problems that are involved now are the interpretations of evaluations, how you

want to interpret the data, and it's all rolling

around in a ball of wax.

Now I think with this approach there are some other serious concerns. I think that you might, as

I think it was expressed at this conference, wonder

whether you need this cross section any better until

you can master the ratio problem. I looked at those

ratios again last night, and well, depending upon who

you want to believe, it's one to three percent uncertainty in the ratio of ^SSy j-jjg other fission cross

sections. It certainly is not an order of magnitude

better and maybe not even a factor of two, or maybe

even worse. I think before you push too hard on the

problem of the absolute ^^^U cross section, if it is

correct to two percent over most of this range, you'd

better master the ratio problem.

I also think that one should give far more attention to the weighting of the absolute source measurements; I believe it was Bob Peelle who just

mentioned them here a minute ago. I think that there

is an awful tendency in fission cross section measurements, both ratios and absolute ones, to look at the mass of points and to draw a line through them, and to ignore these highly precise and quite independent

check points. I think that's a mistake that has been

made in a number of contexts, and it's acute in this

case. Only four or five points which appear to be

356
excellently done, and they get lost in the wilderness;

but they're very essential normalization points. The

same thing tends to be true of white-source versus monoenergetic-source machines.

I think that I would then like to go on to the

problem of the high-energy fission standards. I certainly subscribe to the statements made by Siegfried

Cier jacks a minute ago. Yes, we need a high energy

fission standard from 14 MeV to 20 MeV, with perhaps 2 or 3 percent accuracy. What bothers me I think is the

problem of structure in the ^^^U cross section. The

23^U cross section has structure, I'm sure, and some very excellent measurements show it. Apparently it is

not so prominent at all in ^^^Np. But I think before

we change standards one should be very careful to confirm the exact magnitude and nature of this structure.

Furthermore, I had a little inspection of the data

I could find last night, and I concluded that with any

reasonable working experimental average you in fact

could not be perturbed by that structure, and I am a little afraid that this is going to be a reflection of

a syndrome, and I'm sure people will take exception to this, but a syndrome for resolutionmanship , as Henry

Newson used to call it. In many of these things, let's

not be possessed with the structure. I think Paulsen

was right; gold's got structure, but when used with

judgment, it's a good capture standard, and the same may be true for ^38^ fission.

I think, too, that we should, of course, in this

high energy range, before we put a big crunch on a new

standard which does not have a direct application such

as ^^^U, always remember that this is a region, if no- where else, where hydrogen should work. Do we really

have a need for another reaction standard in this

area? Well, that's where I'd like to stop on fission

cross sections.

I'd like to make one more remark which is applicable

to fission cross sections, but more generally applicable

than that. It is aimed at our evaluators who, I must

confess as my operating budget gets smaller and smaller

and as I tend to spend more time in my office punching

square holes in long cards, I seem to be joining.

There seems to be an awful rigidity to evaluations.

I don't think there's a sin in being smarter today than

yesterday and that the result of new information or new thoughts ought to be incorporated into an evaluation - even a standard. I think we should be more adaptable in adjusting to changing conditions and

changing knowledge. I guess that's in a direct opposition to the way the world is going, but I seem to have this feeling that we're awfully rigid in what we're

doing. Even if somebody comes up today with a new

standard result, it's going to be four years, if we're

lucky, before anybody gets it out of their private

files and used. And I think that's a very serious

problem. There's a need for a great deal more responsiveness

.

Prof . Havens : Thank you.

Dr. Peelle : There were many remarks in the last few

days about the problem of normalizing cross sections

near thermal energy, in addition to my own. One point

I didn't make during my talk but which Dr. Bhat made

at Argonne was, that in spite of the great discomfort

with some of the intermediate results that one looks

at in making such a normalization effort in ^^^U, by

the time the CSEWG Task Force with its imperfect

understanding of everything got to about 100 keV , the

seeming discrepancy on the average with the monoenergetic sources was of the order of one or two percent which is well within the uncertainty of the

measurements. So perhaps the central limit theorem

or law of averages has helped out. Perhaps it has not

been so bad as it could possibly have been with the

uncertainties along the way.

We were challenged to come up with an uncertainty

on the request for a one percent -^^^U cross section.

Based on the evaluation of the statistics of the various

request lists I've seen over the years, the answer is

1+0.6 percent accuracy without any statement whatsoever as to what that means. I think it must mean, since most of the uses are somewhat integral in nature,

that this kind of accuracy is needed in the integral

over a substantial energy region including at least a few lethargy units. Now how we verify that as a desired

accuracy, I don't really know. Most of those numbers were derived by people a decade or more ago that I didn't talk with. Since that time, in the case of

^^•'U, the cross section I think on the average has

changed by several times that amount. But until we

get stability somewhat near that value, I don't think we need to worry too much about the precise number.

But how do we know how well we must do? That's

a question we'll have to face within a few years as the standards field becomes more mature. It seems to me that the work that Chuck Weisbin reported on, although of preliminary nature in a sense because all

the necessary inputs that he said he needed were not

present, does start to show in a fairly vigorous way

what the impact of a change in a standard or an uncertainty in a standard really is. In the analysis

of integral benchmarks, which are dear to the heart

of the applications people, changes at the 1% level seem to have a significant impact, even in an Integral

experiment that contained no uranium. So the first

tentative systematic study that I've seen suggests

that there is a definite impact of knowing a standard

to high accuracy perhaps better than the present knowledge .

I think that since 1970 there's been a dramatic

improvement in the ^^^U fission cross section. Dr. Poenitz suggested that indeed over some broad band it's

something like 7 percent in seven years. Now I don't

think it will do that in the next seven years, but

who's to be sure, especially if we achieve the flexibility in using our knowledge that Dr. Smith asked us to show.

Prof. Havens : Would any other member of the panel like

to comment on the ^^^U cross section? Lee Stewart.

Lee Stewart : Well I suppose that you can plot data

and, if you know what you want to show, you plot it

that way to show it. But I believe I would find it

o q c very difficult to accept a 2 percent error on ^^^U

above a few keV, remembering the work that we've done

for sometime now in CSEWG. In addition, if I remember

correctly, and if people have not changed their data,

the structure that we put into version IV in '^^^XS was based on a comparison of four separate experiments;

one from Harwell, one by Charlie Bowman from Livermore,

one from Los Alamos by Lemley, and Reg Gwin's from

Oak Ridge. The structure did line up. We did find

that they did not identically line up, but saw the same peaks and valleys in all four experiments. The main

reason, however, we chose to put in the structure is

because there are wide minima, there are 15 percent

deviations, and therefore a person doing a monoenergetic experiment will measure a cross section depending

upon his resolution, if you assume that those data are correct. Therefore I believe that we do need the

structure in the cross section.

Prof. Havens: Further comments?

357
Dr. Poenitz: When structure shows up, I always take

the practical point of view; I ask, what is the effect

in a calculation? For example, what is the effect on

a calculation using the ^^^U capture cross section

when the structure differs by an order of magnitude?

While the structure may be interesting, I don't believe

it has any practical implication, and the calculations

agree with this.

The major comment I wanted to make is in regard to the question of the hydrogen cross section as a standard.

First of all, there is no standard cross section anyway

because standards can only exist for basic units, like

length and time. I noticed you have the derived quantities, and so each new measurement necessarily changes

that so-called standard; we should probably call it a reference. From that it follows immediately that there

can be no difference among the standards. There is no primary standard. The argument that the hydrogen cross section, for example, is the best known, is a true

argument or true statement, but it does not mean that

the measurement using the hydrogen cross section is preferable to any other one because the cross section is

not the only quantity involved. The efficiency of the

detectors is the second one. So the absolute measurement of the fission cross section may bring us a much

better reference cross section than the use of the

hydrogen would permit us to do. As a matter of fact the

question comes up now in the high energy range where we

don't know the angular distribution of the hydrogen so well. It may be much better to use say a fission cross section where we have a very high precision in the

efficiency of the counter.

Prof. Havens : Anyone care to comment on that?

Dr. Peelle : A small piece of it I'd like to comment on. It's true that we did one study for the unresolved

resonance region in ^^^V, looking at fast reactors,

and found the fluctuations which have been seen there

with indeed much less resolution than is available

could not be shown to have an impact of any importance

relative to the error in the average cross section.

However, I think that one has to have a little reservation about that study' and recognize that in any such

question about the importance of fluctuations you have

to ask about the particular situation. In general, I think we've gotten in trouble with our reactor engineer

friends several times by having given them the impression that cross sections are smooth. They believed

the graphs they saw in the book. Even though we know

they weren't smooth, they didn't know they weren't

smooth. It may be best to represent them to the extent

that we know them.

Prof. Havens : Thank you.

Dr. Smith: What does Ugo Farinelli think about that?

Dr. Farinelli : This is I think a good example of the

influence of the averaging procedure. It depends on whether you make the average from very detailed structur

in the differential measurements, or whether it's the

measurement itself that makes a certain average. I have always thought that it would be conceivable to make measurements that measured directly the average

in the sense that it was to be used, especially when

the detail is so great that even if theoretically you

have the methods to weigh the structure differently

according to the composition, which should be done,

practically you don't do it because of the complexity.

If you have say 100,000 points, you are not going to weight this over any spectrum; you're simply going to make a very straightforward average, independent of

composition. So I think a measurement can yield the same result. Well, it may not be perfectly correct.

but this is going to be the same value that is used

by the integral man who is doing the calculations.

So I agree that in principle the man who makes the

measurements would like to give something which is

perfect and then it's the responsibility of the man

who uses his data if he fudges with them. But from

a practical point of view I don't see the difference.

Dr. Peelle : I did make a quick effort to check that

point on the question of thick-target transmission and

self-indication measurements for ^^^U, and to compare

that to what one sees in a processed cross section

taking into account what are thought to be the fluctuations. Even though the two studies are based on the same phenomena of the self-protection in the resonances,

I could not demonstrate a direct correlation for either

of those two types of measurements and the processed

cross sections which are now found. Not only was

there none, but 1 could not show there should be. I think I may have erred, but sometimes it will be

necessary for the calculational expert to take the

details in some form or another and deal with them.

Prof. Havens : Francis Perey.

Dr. Perey : I'm very much interested about the statement regarding structure details. As you know I've

struggled with the problem, but you tell me that I must average the cross section in the measurement process or the evaluation process and give you a smaller

number of points. We've tried to do that for years

and never succeeded. I don't know how you're going

to homogenize your calculations, whether you're going

to use a couple of mm or one mm of fuel cladding or one meter of iron in your shield. Which average

should I use to put into the file when I have structure,

when I know the structure exists?

Prof. Havens : Do you want to answer that Ugo?

Dr. Farinelli : Yes. You don't know what to do, but

you should not think that people who make the calculations know what to do. It's not really conceivable

that somebody making the calculations for the reactor

uses a different cross section if he moves from one point of the reactor to the other. Do you really think

that somebody is going to take a different cross section

for iron or nickel in the cladding or in the follower

of the control rod, in the core or in just a little way

off? I don't think so. It's common to have just one

cross section for iron, at least in the core. You may

have two if you go to the blanket or to the shield, but

certainly in the core you have just one cross section.

How are you going to calculate it? With an average

composition over all the core or over a system of cores which are not too different from each other. From one point to the other there will be differences, but the

differences will not be reflected in any set of cross sections used. So the problem is there.

Prof. Havens : Lee Stewart.

Lee Stewart : Let's get back to ^^^U. It is a standard,

and I dislike penalizing a monoenergetic experimentalist

who does an experiment with very good resolution and

therefore measures a cross section that is 10 percent

lower than anybody else who does one with very poor

resolution. And for that reason alone, if none other,

I think we should be willing to put in the structure

in 235u that has been measured by at least four people.

That was my primary reason for thinking that the structure should go in. It did bring in line, by the way,

a lot of older experiments. It did show us where some

of the problems are.

Prof. Havens: Chuck Weisbin.

358
Dr. Welsbin : I think this one study that we're talking

about here in intermediate structure unfortunately can

go back to something we did. This was one study, one

case, not a large sampling. I think, however, in the

future years we are going to see advanced techniques

which can make use of the structure, Ugo, in the sense

of the work that Red Cullen is doing on probability

tables, where he tabulates the probability of a given

cross section versus that cross section in a given

energy range. This is a very efficient representation

that can make use of the structure so that if evaluators

did the smoothing ahead of time, and just gave us a single average, we could take advantage of that methodology. However, if we do have the access to the

structure, we can put it in an efficient fashion, and

I think we will find out we get different answers,

at least in some cases, not yet shown, admittedly.

Prof. Havens : This problem of how much detail you

should put in the cross section and then what you can

use and what you assume and how you average has existed,

I'm sure, ever since nuclear energy has existed in the

form of reactors, and if you had seen some of the

averages that Hans Bethe made in 1943 and 1944 you

would be really flabergasted at some of the assumptions

that went into the formulae. I remember very well when

I first showed Neils Bohr a resonance in ^^^U he said,

"This can't happen". So, there are some problems with

the detail you put in the cross section and the theoretical models, not only of the nucleus but of the

reactor as well, and you're limited to what you can do

in a practical sense.

Dr . Bowman : I think that's an interesting story

because what you're saying is that as years have gone

by and calculational techniques via computers have

improved, that we've been able to handle more and more structure. It seems like a lot of the discussion

here this afternoon assumes that we are at the end of

that calculational development and calculational

capability. But, if I read the latest issue of Science

where they talk about what's coming in the speed of

computers and so forth and assume that that's going

to be implemented in the nuclear power industry, it

would seem to me that there's quite a likelihood that we will develop these techniques. Maybe the technique

that Chuck was referring to a minute ago is just the

next of a series of steps into the future.

Prof. Havens : I think with the development of reactor

design and reactor calculations over the years, the

capability of computers has very often been the

limitation of what you can do; and as that's improved,

the calculational accuracy of what could be predicted

has also improved and will continue to improve. However, I was asked to try to bring this session to a close shortly. Before we finish the session, I'd

like to ask Alan Smith if he has anything further

to say.

Dr. Smith : I do. I am brought back to Farinelli's

point of view. I really think the greatest value of

standards is the all-important ability to calculate

the performance of the system within a measurable

range. And if you can't do that, you could never

expect to make the extrapolation of the calculation to an unmeasurable range which you'll have to do to prove

acceptance in safety. So I think there's been a change

in the values in my mind of standards from primarily

economic now to safety and public acceptance. Now it

may change back after you've passed a period to economic grounds, but I think that's the key. Your

calculational ability is the thing that's important

now. It's your ability to predict reliably and show

that you can do it. Without that, I don't think your

technology is going to fly.

Prof. Havens : Well the conference will be brought to

a close, but before we do that I'd like to thank the

panel members and Alan Smith for summarizing the conference. I'd also like to take this opportunity to thank Charlie Bowman for organizing the program, which,

when he first said we're going to spend four days on standards, I said, "Oh, my God, not that." But cer- tainly the interest of this conference has held up a

lot more than many of the meetings that I have attended.

The fraction of the audience here on Thursday afternoon

to that here on Monday morning is very much higher than

it is in most technical meetings that occur. The standards, someone said, only are needed after you've made

several measurements and after a field has been pretty

well developed. I think that's true if you only have

one measurement. Then you don't need a standard, the one measurement is the standard. I don't want to get

into a debate with Poenitz about standards because we have talked about standard reference data and there are various definitions of standards. I will suggest that

you read the definitions of standards in some of the

literature put out by the National Bureau of Standards,

and you'll be surprised at what some of the standards

are. Some of them are arbitrary, some of them are natural, in that they are physical measurements, but

they vary all over the lot depending on the field that

you're working in. But I think that Charlie and the

Bureau have done an outstanding job. It's the first

conference I have attended here at the Bureau, and I think that it has been a very well run conference, and

the Bureau has excellent facilities which we've all

enjoyed. 1 think we should give a vote of thanks to the Bureau, and we certainly appreciate the activity

in this field that has been demonstrated here at this

conference. Thank you, Charlie.

I now declare this meeting adjourned.

359
LIST OF PARTICIPANTS

Alberts, W. G. Physikalisch-Technische Bundesans tal

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August, Leon S. Naval Research Laboratory

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Bartlett, William T. Battelle, Pacific Northwest Labs.

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Bergqvist, I Dept. of Physics,

University of Lund

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Bhat, Mulki R,

Brookhaven National Laboratory

T197C

Upton, New York 11973

Blinov, Mickaii

t Chlopin Radium Institut

Leningrad

USSR

Block, Robert

Rensselaer Poly. Institute

Troy, New York 12181

Boldeman, John William

AAEC

Private Mailbag Southerland

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Bowman, Charles D. Center for Radiation Research

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Brandenberger , Jerry D. Los Alamos Scientific Laboratory

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Cance , Michel

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Carlson, Allan D. Center for Radiation Research

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Cierjacks, Siegfried W. D-75 Karlsruhe, Weberstrasse

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Cosack, Martin

Physik Tech. Bundesans tal

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Dierckx, Ronald

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Eisenhauer, Charles M. Center for Radiation Research

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Engdahl, John C. Department of Physics

University of Michigan

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Fabry, Albert

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Farinelli, Ugo

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Fuchs , Jean Eric

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360
Gabbard, Fletcher

Department of Physics & Astronomy

University of Kentucky

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Geiger, Klaus W. National Research Council of Canada

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Gilliam, David M,

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National Bureau of Standards

Washington, D. C. 20234

Lane, Raymond 0. Physics Dept., Accelerator Lab.

Ohio University

Athens, Ohio 45701

Leiss, James E. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Leonard, Bowen R. Jr. Battelle - Northwest

P. 0. Box 999

Richland, Washington 99352

Lippincott, E. P. Wes tinghouse Hanford Company

P. 0. Box 1970

Richland, Washington 99352

Liskien, H. CEC, CBNM Euratom

Steenweg naar Re tie

B-2440 Geel

BELGIUM

Lisowski, Paul W. Triangle Universities Nuclear Lab.

Dept. of Physics, Nuclear Bldg.

Duke University

Duke Station

Durham, North Carolina 27705

Lone, M. A. Chalk River Nuclear Labs.

Chalk River, Ontario KOJ IJO

CANADA

Lurie, Norman

IRT Corporation

7650 Convoy Court

San Diego, California 92111

McDonald, Joseph C. Sloan-Kettering Institute

127 5 York Avenue

New York, New York 10021

McElroy, William N. Westinghouse Hanford Company

P. 0. Box 1970

Richland, Washington 99352

McGarry, Emmert Dale

Harry Diamond Lab

2800 Powder Mill Road

Adelphi, Maryland 20738

361
Maeck, William J. Allied Chemical Corp., INEL

550 2nd Street

Idaho Falls, Idaho 83401

Magumo, Benjamin A. Brookhaven National Lab.

T-197

Upton, New York 11973

Marshak, Harvey

Heat Division

National Bureau of Standards

Washington, D. C. 20234

Meier, Michael M. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Moazed, Cyrus

Harry Diamond Laboratories

2800 Powder Mill Road

Adelphi, Maryland 20783

Moore, Michael S. Los Alamos Scientific Lab.

P.O. Box 1663

Los Alamos, New Mexico 87545

Morgan, W. C. Batte lie-Northwest

Battelle Boulevard

Richland, Washington 99352

Motz, Henry T. Los Alamos Scientific Lab.

Box 1663

Los Alamos, New Mexico 87 544

Mughabghab, Said F. Brookhaven National Lab.

Upton, Long Island

New York 11973

Navalkar, M. P. Bhabha Atomic Research Centre

Neutron Physics Section

Trombay

Bombay, Maharasmtra

INDIA 400 085

Nelson, Charles E. Triangle Universities Nuclear Lab.

Duke University

Durham, North Carolina 27706

Nelson, Ronald 0. Los Alamos Scientific Lab.

P.O. Box 1663

Los Alamos, New Mexico 87544

Nilsson, Leif

Tandem Accelerator Lab.

August Sodermans vag 70

Uppsala, SWEDEN

Otroshenko, Gennady

Kurchatov Institute on the

Atomic Energy

Moscow, USSR

Czer, Odelli

Electric Power Research Institute

P. 0. Box 10412

Palo Alto, California 94303

Patrick, Bryan H. U. K, Atomic Energy Authority

Harwell, Didcot

Oxon. 0X11 ORA

ENGLAND

Paulsen, A. CEC, CBNM Euratom

Steenweg naar Re tie

B-2440 Geel

BELGIUM

Peelle, Robert W. Oak Ridge National Laboratory

Building 6010

Oak Ridge, Tennessee 37830

Perey, F. G. Oak Ridge National Laboratory

P. 0. Box X Oak Ridge, Tennessee 37830

Peters, Gerald J. Naval Surface Weapons Center

White Oak Laboratory

Code WR41, Building 2-013

Silver Spring, Maryland 20910

Phillips, Leigh F. Brookhaven National Laboratory

Building 535

Upton, New York 11973

Poenitz, W. P. Argonne National Laboratory

9700 S. Cass Avenue

Argonne, Illinois 60439

Pullen, David

Dept. of Physics

University of Lowell

1 University Avenue

Lowell, Massachusetts 01854

Qazi, M. Nuruddin

Pinstech, P. 0. Nilore

Rawalpindi, PAKISTAN

Quam, W. M.

E.G. &G

.

P. 0. Box 98

Goleta, California 93017

Rahn, Frank J. Electric Power Research Inst. Box 10412

Palo Alto, California 94303

Renner, Cleide

Oak Ridge National Laboratory

P. 0. Box X Oak Ridge, Tennessee 37830

Riel, Gordon K. Naval Surface Weapons Center

Nuclear Branch

White Oak

Silver Spring, Maryland 20910

Robertson, James Craig

Dept. of Nuclear Engineering

University of Michigan

Ann Arbor, Michigan 48105

Rogers, J. W. E.G.&G. Idaho

P. 0. Box 1625

Idaho Falls, Idaho 83401

Rogosa, George L. Division of Physical Research, J-309

U. S. Energy Research and Development

Adminis tration

Washington, D. C. 20545

Rohrig, Norman

Brookhaven National Laboratory

Building 902B

Upton, New York 11973

Rosen, Mervine

Naval Research Lab.

Washington, D. C. 20370

Saltmarsh, M. J. Oak Ridge National Laboratory

P. 0. Box X Oak Ridge, Tennessee 37830

Schell, Michael C. Mid-Atlantic Neutron Ther.

6104 Breezewood Ct. # 304

Greenbelt, Maryland 20770

Schrack, Roald A. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Schwartz, Robert B. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Sekharan, K. K. Cyclotron Institute

Texas A&M University

College Station, Texas 77840

Shapiro, Philip

Naval Research Laboratory

Code 6643

Washington, D. C. 20375

Shneiderov, Anatol

The Polycultural Inst, of America

1673 Columbia Road, N. W. , Unit 309

Washington, D. C. 20009

Slater, William E. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Smith, Alan B. Argonne National Laboratory

9700 South Cass Avenue

Argonne, Illinois 60439

Smith, Donald L. Argonne National Laboratory

9700 South Cass Avenue

Argonne, Illinois 60439

362
Smith, J. Richard

E.G.&G. Idaho, INEL

Box 1625

Idaho Falls, Idaho 83401

Spiegel, Valentin

Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Stewart, Leona

Los Alamos Scientific Laboratory

P. 0. Box 1663

Los Alamos, New Mexico 87 545

Taschek, Richard F. Los Alamos Scientific Lab.

P. 0. Box 1663

Los Alamos, New Mexico 87545

Theus, Richard B. Naval Research Lab.

Washington, D. C. 20375

Twining, Bruce

Division of Magnetic Fusion Energy

U. S. Energy Research & Dev. Admin.

Washington, D. C. 20545

Uttley, C. A. Atomic Energy Res. Est. Harwell

Didcot, Oxon 0X11 ORA

United Kingdom

Vlasov, Mercury F. Nuclear Data Section

International Atomic Energy

Agency

P. 0. Box 590

Vienna, AUSTRIA

Wang, Joseph

Naval Surface Weapons Center

White Oak Laboratory

Code WR-Yl

White Oak

Silver Spring, Maryland 20910

Wasson, Oren A. Center for Radiation Research

National Bureau of Standards

Washington, D. C. 20234

Wattecamps, E. CEC, CBNM Euratom

Steenweg naar Re tie

B-2440 Geel

BELGIUM

Weisbin, Charles R. Oak Ridge National Laboratory

P. 0. Box X

Oak Ridge, Tennessee 37830

Weston, Lawrence W. Oak Ridge National Laboratory

Bldg. 6010, P. 0. Box X

Oak Ridge, Tennessee 37830

Whetstone, Stanley L.

U. S. Energy Research & Dev. Admin.

MS J- 309

Washington, D. C. 20545

White, Roger M. Accelerator Laboratory

Ohio University

Athens, Ohio 45701

Yockey, Hubert P. Army Pulse Rad. Facility

Aberdeen Proving Ground

Aberdeen, Maryland 21005

Yolken, H. Thomas

Nuclear Safeguards Meas., IBS

National Bureau of Standards

Washington, D. C. 20234

Young, Phillip G. Los Alamos Scientific Laboratory

Los Alamos, New Mexico 87544

Zeitnitz, B. Universitat Bochum

Postfach 102148

D-4630 BOCHUM 1 WEST GERMANY

Zijp, W. L. Netherlands Energy Dev. Agency

3 Wes terduinweg

NL-PETTEN

THE NETHERLANDS

363
AUTHOR INDEX

Adamov, V. M. p. 313

Alexandrov, B. M. -- p. 313

Alkhazov, I. D. -- p. 313

Ambler, E. -- p. 1 Auxier, J. A. -- p. 101 Axton, E. J. -- p. 237

Barschall , H. H. — p. 342

Blinov, M. V. — p. 194

Block, R. C. -- p. 255

Boldeman, J. W. -- p. 182

Brandenberger , J. D. -- p. 234

Broerse, J. J. -- p. 106

Carlson, A. D. -- p. 85

Caswell, R. S. -- p. 121 Chrien, R. E. -- p. 255

Cierjacks, S. -- p. 278

Coachman, J. S. -- p. 61 Czirr, J. B. -- p. 54

Derrien, H. -- p. 14

De Saussure, G. --p. 174

Drapchinsky, L. V. -- p. 313

Edvardson, L. -- p. 14

Eisenhauer, C. M. -- pp. 156, 198, 329

Fabry, A. -- pp. 290, 329

Farinelli , U. -- p. 310

Fomichev, A. V. -- p. 313

Gabbard, F. -- p. 212

Gilliam, D. M. -- pp. 299, 335

Gold, R. -- p. 137

Grundl , J. A. — pp. 156, 329

Hale, G. M. — p. 30

Harvey, J. A. -- p. 10

Huynh, V. D. -- p. 244

Jaffey, A. H. -- p. 206

James, G. D. -- p. 319

Joneja , 0. P. -- p. 61 Kazi , A. H. -- p. 342

Knitter, H.-H. -- p. 3 Knoll , G. F. -- p. 304

Kobayashi, K. -- p. 255

Kostochkin, 0. I. -- p. 313

Kovalenko, S. S. -- p.' 313

Kudriavzev, 6. Yu. -- p. 313

Lachkar, J. C. -- p. 93

Lamaze, G. P. -- p. 37

Lemmel , H. D. -- p. 170

Liou, H. I. — p. 255

Liskien, H. -- p. 2 Maeck, W. J. — p. 146

Malkin, L. Z. — p. 313

Marston, T. V. -- p. 137

McGarry, E. D. -- p. 342

Meier, M. -- p. 221 Navalkar, M. P. — p. 61 Paulsen, A. — p. 165

Peele, R. W. — pp. 174, 269

Petrzhak, K. A. -- p. 313

Phiske, M. R. — p. 61 Pleskachevsky , L. A. -- p. 313

Poenitz, W. P. -- p. 261 Rahn, F. J. -- p. 137

Roberts, J. H. -- p. 137

Schroder, I. G. -- p. 10

Schwartz, R. B. — p. 250

Sekharan, K. K. — p. 234

Shapakov, V. I. -- p. 313

Singh, U. N. — p. 255

Srikantaiah, R. V. -- p. 61 Stahlkopf, K. E. -- p. 137

Stewart, L. -- p. 198

Touse, V. T. -- p. 194

Uttley, C. A. -- p. 47 Wasson, 0. A. -- p. 115

Wattecamps, E. -- p. 67

Weisbin, C. R. -- p. 269

Weston, L. W. -- p. 43

Vitenko, V. A. — p. 194

Zijp, W. L. -- p. 128

364
ELEMENT QUANTITY TYPE ENERGY DOCUMENTATION LAB COMMENTS

s A MIN MAX REF VOL PAGE DATE

H 001 DIFF ELASTIC REVW -CONF 20 + 7 30 + 7 77NBS 47 477 HAH UTTLEY . GfiPHS , TBLS. EXPT CFD Th CALCS.

LI 006 EVALUATION REVW -CONF 10-5 20 + 7 77NBS 30 477 LAS HALE+R-MATfilX ENDF B5 . G R P HS . CF D EXPTS

LI 006 TOTAL XSECT REVW -CONF 1 0 + 0 1 0 + 7 77NBS 3 477 GEL KNITTER. 3 MEAS CFD. TOT CFD SEL GRPH.

LI 006 TOTAL XSECT REVW -CONF 70+1 1 0 + 7 77NBS 1 0 477 SAC DERRIEN+EXPTS C F D . TB L , G R P HS . R EF TBL.

LI 006 ELASTIC SCAT REVW -CONF 1 0 + 5 70 + 5 77NBS 3 477 GEL KNITTER. INTEG DIFF CFD TOT CS GRPH

LI 006 ELASTIC SCAT REVW -CONF 10 + 3 75 + 6 77NBS 1 0 477 SAC DERRIEN+EXPTS CFD GRPH. REF TBL.

LI 006 DIFF ELASTIC REVW -CONF 20 + 5 14 + 7 77NBS 3 477 GEL KNITTER. 4 MEAS CFD.ANG DIST.LEG COFS

LI 006 DIFF ELASTIC REVW -CONF 10 + 3 50 + 6 77NBS 1 0 477 SAC DERRIEN+ANG DISTR . NDG . TbL OF REFS.

LI 006 POLARIZATION REVW -CONF 20 + 5 14 + 7 77NBS 3 477 GEL KNITTER . ANAL , POL POWER MEAS. CFD. NDG

LI 006 POLARIZATION REVW -CONF 20 + 6 50 + 6 77NBS 1 0 477 SAC DERRIEN+ANAL PWR.NDG.TBL OF REFS.

LI 006 RES INT ABS REVW -CONF 50-1 77NBS 128 477 RCN ZIJP.TBL.l/E SPEC. FOR TOT HE PRODUCT

LI 006 N, TRITON REVW -CONF 30 + 3 1 4 + 7 77NBS 3 477 GEL KNITTER. INV N DG . I NTEG , D IFF CS CFD.

LI 006 N, TRITON REVW -CONF 1 0 + 0 1 0 + 5 77NBS 1 0 477 ORL HARVEY+ANG A N ISOT R OP Y . D IF F GRPHS.

LI 006 N, TRITON REVW -CONF 25-2 1 8 + 7 77NBS 1 0 477 SAC DERRIEN+INV NDG. DIRECT GRPHS. REF TBL

LI 006 N, TRITON REVW -CONF 25-2 77NBS 128 477 RCN ZIJP. TBL. COMPARISON OF 2200M/S CS.

LI 006 N, TRITON REVW -CONF 1 0 + 2 1 0 + 5 77NBS 174 477 ORL PEELLE+THR NORM TECH ST U D Y . TBLS , G R P

H

LI 006 RESON PARAMS EVAL -CONF 2^ + 5 25 + 5 77NBS 319 477 HAR JAMES. ALL FOR 250KEV RES CFD. TBL.

B 0 1 0 TOTAL XSECT COMP -CONF 10 + 2 20 + 7 77NBS 67 477 GEL WATTECAMPS.GRPHS, TBLS. TBL REFS GVN

B 010 ELASTIC SCAT COMP -CONF 40+2 20 + 7 77NBS 67 477 GEL WATTECAMPS.GftPHS, TBLS. TBL REFS GVN

B 0 1 0 DIFF ELASTIC COMP -CONF 40 + 2 20 + 7 77NBS 67 477 GEL WATTECAMPS.GRPHS, TBLS. TBL REFS GVN

B 010 RES INT ABS REVW -CONF 50-1 77NBS 128 477 RCN ZIJP.TBL.l/E SPEC. FOR TOT HE PRODUCT

B 010 N, ALPHA REAC COMP -CONF 10 + 4 20 + 7 77NBS 67 477 GEL WATTECAMPS.GND+IST EXC CS GRPhS.TBLS

B 0 10 N, ALPHA REAC REVW -CONF 25-2 77NBS 1 28 477 RCN ZIJP. TbL. COMPARISON OF 2200M/S CS.

B 010 N, ALPHA REAC REVW -CONF 1 0 + 2 1 0 + 5 77NBS 174 477 ORL PEELLE+THR NORM TECH ST U D Y . TBLS , G fi P

H

C 012 TOTAL XSECT EVAL -CONF 25-2 15 + 7 77NBS 93 477 BRC LACHKAR. GRPHS, TBLS PHASE SHIFT ANAL.

C 012 DIFF ELASTIC EVAL -CONF 50 + 4 1 5 + 7 77NBS 93 477 BRC LACHKAR. GRPHS , TBLS PHASE SHIFT ANAL.

C 012 POLARIZATION REVW -CONF 1 0 + 5 26 + 6 77NBS 93 477 BRC LACHKAR.TBLS.ee, PHASE SHIFT ANAL.

C 012 N, GAMMA REVW -CONF 25-2 20 + 7 77NBS 93 477 BRC LACHKAR. THR VAL GVN. INV STUDY ALSO.

C 012 RESON PARAMS EVAL -CONF 1 0 + 7 1 0 + 8 77NBS 319 477 HAR JAMES. NARROW RES FOR E STANDARD GVN.

C 0 1 3 TOTAL XSECT REVW -CONF 10 + 5 23 + 7 77NBS 93 477 BRC LACHKAR. CS CONTRIBUTION TO NATURAL C

C 013 N, GAMMA REVW -CONF 25-2 20 + 7 77NBS 93 477 BRC LACHKAR. THR VAL GVN. INV STUDY ALSO.

0 0 1 6 RESON PARAMS EVAL -CONF 10 + 6 1 0 + 7 77NBS 319 477 HAR JAMES. NARROW RES FOR E STANDARD GVN.

NA 023 RES INT ABS REVW -CONF 50-1 77NBS 1 28 477 RCN ZIJP. TBL. AVG OVER 1/E SPECTRUM.

NA 023 N, GAMMA REVW -CONF 25-2 77NBS 128 477 RCN ZIJP. TBL. COMPARISON OF 2200M/S CS.

NA 023 RESON PARAMS EVAL -CONF 10 + 4 1 0 + 5 77NBS 319 477 HAR JAMES. NARROW RES FOR E STANDARD GVN.

365
ELEMENT QUANTITY TYPE ENERGY DOCUMENTATION LAB COMMENTS

s A MIN MAX REF VOL PAGE DATE

MG il| d d n t a mN , r n (J 1 U N

D LT 11 I.I n t V W P (~1 M C - L U N r r X oo "7 KI D C

( ( Vi DO 1 o

i\ n 1 D P KI n L N ^iJr.Lo Avu UVbn U235 r ISS SPEC.CFD

A I A LA O T M D D n T n M N , r r( U 1 U ND I? \/ U - p n M c L U N r F T Q Q f X O O7 7 M n C

I ( N DO 1 o ft \[ n 1 D P M n L In 71 TP PC TDI IIOOC ^XJr^.Lo L dL , U d CTQC CDCP uir\rc c r\ r ioo orbO nULo Leu.

A I n o 7 Ki p D n T n M N , r n U 1 U N R R 1/ U n L V H — p n w c u u w r r X L}o 7 7 KI n Q

( (Moo 19ft 1 d 0

ll 7 7 R P W It L. IN 7T1P ZiXjr.UoPC AVLi AUP AUFR uvc>n y^cdoc rloo orJ:*L LJf U

n L 0 9 7 R R V W — pV nu ivw rR R T <; *^ r X o o 7 7 KI R ^ ( ( IM O O 19ft 1 c O hi 1 Rn PU KI IN 7TTP PC TRI 119'iR rioo ort:#u riLJLo \^vv»

A I W flIPPA RRiir R R U lu — p^ nu KIii rR r X o o 7 7 KI R ( I Vi DO 1 9 ft ll 7 7 nR P KJ L- IN 7T1P ZiXOr.UOP<s nVLl AUP nUFR L/VCiH ^0 ^ rj.oo orCtU urLJ

fl I 0 9 7 RR^nW PARflM'^ R U fl I p n w R 1

1 nu J.+ nu 1 n J. ii

I U + 4 7 7 K] R <^

( ( n D o h 7 7 RAP ri n n UAric«o*iNAnrvun TAMRC MARRAU nciO RFC run JL OiJiNLJAnL/ uVn.

O X n 9 A U c 0 nc*oUN riinHrlo 17 U A I C< V A L — p n w F u u iM r I u + 4

1 n j_ p i \J + Z> 7 7 KI n c

/ 1 N D o

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Dr U ^ 1 M p R n Tn w R r \/ Ui n c, V n — PLi O M c u w r F T Q Q r X o o

7 7 M n Q

f ( n D o 19ft 1 d 0

Jl 7 7 D P M n IN 7TTP PC A\/P A17FD iioQC CTCc CDCP pen

Q O J 9 KI p R n T n w R c u lj n Ci V W — p n w c* L u r F T ^ ^ r X o o 7 7 M D Q 19ft 1 C O

)l 7 7 D P w n L n 71ZiXclTaLiO TP PC AVu fi \I P AUFD UVCn IIOOC \i c. rITfCQ loo orJLL.LrL) QDUP pen

QoA Q O ntoUW rAniiMC) C. V A L - P A M C L U W r1 A • Jl 1 U + 41 A . A 1 U + D

7M"7 Ikl D C N oo O 1 A

5 1 9

Jl 7 7 U A D n A n JAMES. NARROW RES FOR E STANDARD GVN.

C P o LA Jl C U 4 !5 THTAI iUlAL VQCPT AobLi c- A r 1 — p n Ki c u U N rJl A 1 O 9 9 ll il + H7 7 W D C

( ( vi DO Jl 7 7 R P T n r X CHRIEN+LINAC . TRNSi.GRPHS. CFD.

c n A )l C U *4 ^ DITQ TMT ADQ nfco XrJl Ado D C \/ U K b V Wp A M r - U UN f C A 1 t> L) - 1

"7 M C Q

/ / N DO 19ft Jl 7 7 nPPMIn ZIJP.TBL. AVG OVER; 1/E SPECTRUM.

o^ rL. n ii c; U 4 t) MN P A M M A

, u A rl It A Rn REl UV UW — p nu KI R N r 9 c; 9 7 7 KJ R ( / IM D O 19ft 1 c. 0 il 7 7 nR Pu INW ZIJP.TBL. COMPARISON OF 2200M/S CS.

RR^^nw n Ci o u w rPARAMO /I n ii ri o R Y P T p n Ni R ~0 + J9 1 J. il 7 7 M R *N

( f li D O 5 c; c il 7 7 Ri\ rP T± CHRIEN+S WAVE WN ,J,WG.P WAVE G«WN.

T T R R U Uj fl C< V n — p^ nu i"( Ki rR Rr XT o^ o 7 7 KI R ^ ( ( i * D O 1 C. O ll 7 7 nun R P W ZI JP . Tl (N , X ) 46SC

.

U235 FISS SPEC AVG.

T T W PROTON nR C(R uV yW — p^ nw I*KI rR R T ^ <^ r X o o 7 7 W R *s 19ft 1 c. O Jl 7 7 nR PV INW ZIJP.TI(N,X)46SC. U235 FISS SPEC AVG.

T T nR CiR uV yn — pu nu NiiM rR R T r X o o 7 7 KI D c

I / N D O 19ft 1 d O nR Pu KiIN Z1JP.TI(N,X)46SC. U235 FISS SPEC AVG.

T T 1 X nU iiH 0 M p R n T n M R P 1/ U n c. V w — p n w F L u In r F T Q Q f X OO 7 7 M U C

( f N DO 19ft 1 d 0

Jl 7 7 D P Kj n L N ZIJP.CS AVG OVER CF252 FISS SPEC CFD

TI XT n ii 7 Ki p R n T n M Rn CiR UV Uw — pu nu KIw rR r X o o 7 7 KI R <?

1 f n D O 19ft 1 d 0 ll 7 7 nR Pu KiIN ZIJP.CS AVG OVER U235 FISS SPEC.CFD

Ti 1.T n u 7 w p R n T n w Rn CiR UV Wn — p^ nu iiw rR r X o o 7 7 Ki R ^ ( ( IN DO 19ft 1 d o ll 7 7 nR PU KiIN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

T T n\J iiH RO w P R n T n M nR RCi \7V Un — pv-i nu ww rc F T Q <i r X o o 7 7 KI R Q f / rj D O 19ft 1 d 0 ll 7 7 Rn P Ki Ll IN ZIJP.CS AVG OVER U235 FISS SPEC.CFD

T T OUR U H O M p R n T n M Rn CiR \/V Uw — p n Ki [? u u N r F T ^ r X o o 7 7 KI R Q

f ( IN DO 19ft 1 d 0 11 7 7 R P KI Ia Lf IN ZIJP.CS AVG OVER CF252 FISS SPEC.CFD

U D D M9M Vt ^ n nDnVyiXUH RPflrTTOKj n

R cR VV Uw — pLi nu KIN Fr r X o o 7 7 KI R Q

f ( Ci DO 19ft t d 0

ll 7 7 nun R P M ZIJP.CS TBL.U235 FISS SPEC MDLS CFD.

M M nU R R D D

D r \f Ui n Cj V W - P A M C L U N r F T Q C

r X o o

7 7 M D C

( 1 n DO 19ft \ do Jl 7 7 nPPMu N ZIJP.CS AVG OVER CF252 FISS .SPEC.CFD

r c< U D H M P R n T n M nR R U LI Ci V n — p n M F 0 u N r F T Q r X o o 7 7 M p Q

( ( IN ISO 19ft 1 d 0

Jl 7 7 Rn P Ki Ll IN ZIJP.CS TBL.U235 FISS SPEC MDLS CFD.

F E nU ^ ii D H W P R Pi T O M nR RCi VV UW — P A K! F u u In r F T Q c; r X o o 7 7 KI U Q

1 f In DO 19ft i d 0 1 I I nR Pu KlIN ZIJP.CS AVG OVER U235 FISS SPEC.CFD

rF CiR n c, U ii

_J H R W \l Li n Ci V M — P A M F L u n r F T Q Q r X oo 7 7 W R Q

( ( ti DO 19ft 1 d O

)l 7 7 R P Ki n U IN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

F E nW J u CiR YA rP T1 — PLi Au KIw rR li n J. 9 1 n J. ii

1 U + *4 7 7 KI R f / IN D O 9 R c; do J

ll 7 7 Rn rP XT CHRIEN+LINAC .TRNS .GRPHS. CFD.

F E nU R ft _J u M P R fl T n M nR CiR VV Un — P A W F U u w r F T Q Q r X o o 7 7 M D C

( / M D O 19ft 1 d 0

il 7 7 R P KJ A L/ IN ZIJP.CS TBL.U235 FISS SPEC MDLS CFD.

R R nU R AD

M rnUlUN P D n T A U D I? 1/ U n Ci V w P A M C - U N f t X o o7 "7 W tJ Q

{ f N DO 19ft Jl 7 7 'ill D P M n L N ZIJP.CS AVG OVER CF252 FISS SPEC CFD

R R nU R A D D Ci V A L - C 0 N F 10 + 41 A ^, 1 0 + D

7 T H D O

3 1 9 4 7 7 H A R JAMES. NARROW RES FOR E STANDARD GVN.

FE 058 RES INT ABS REVW -CONF 50-1 77NBS 128 477 RCN ZIJP.TBL. AVG OVER 1/E SPECTRUM.

FE 058 N, GAMMA REVW -GONE 25-2 77NBS 128 477 RCN ZIJP . TBL . COMPARISON OF 2200M/S CS.

CO 059 RES INT ABS REVW -CONF 50-1 77NBS 128 477 RON ZIJP.TBL. AVG OVER 1/E SPECTRUM.

CO 059 N, GAMMA REVW -GONE 25-2 77NBS 128 477 RCN ZIJP. TBL. COMPARISON OF 2200M/S CS.

366
ELEMENT QUANTITY

S A TYPE ENERGY DOCUMENTATION LAB COMMENTS

MIN MAX REF VOL PAGE DATE

CO 059 N2N REACTION R E VW -CONF FISS 7 7NBS 11 ? Ru U 7 7 R C N 7Li A.T .1O LP • cs A V G OVER CFPR? \^ c ^ J L, rxoo KTSS oiCf^.K^rL' SPFT TRD

CO 059 N, ALPHA REAC REVW -C ONF FISS 7 7N BS 12 8 4 7 7 R C N Z IJ P . CS TBL. U235 FISS SPEC MDLS CFD.

NI 058 N2N REACTION REVW -CONF FISS 77NBS 12 8 4 7 7 R C N Z I JP . cs TBL . U235 FISS SPEC MDLS CFD.

NI 058 N, PROTON REVW -CONF FISS 77NBS 1 28 4 7 7 R C N Z IJ P . CS TBL. U235 FISS SPEC MDLS CFD.

Nl 058 N, PROTON REVW -CONF FISS 77NBS 1 28 477 RCN ZIJP . cs AVG OVER CF252 FISS SPEC CFD

CU 063 RES INT ABS REVW -CONF 50-1 77NBS 1 28 4 7 7 RCN Z UP . TBL . AVG1 OVEFi 1/E SPECTRUM.

CU 053 N, GAMMA REVW -CONF 25-2 77NBS 1 2 8 4 7 7 RCN ZIJP. TBL..COMPARISON OF 2200M/S CS.

CU 063 N , GAMMA REVW -CONF FISS 7 7N BS 1 28 4 7 7 RCN ZIJP. CS AVG OVER U235 FISS SPEC. CFD

CU 05 3 N2N REACTION REVW -CONF FISS 7 7N BS 12 8 4 7 7 RCN ZIJP. C S AVG OVER U235 FISS SPEC. CFD.

CU 06 3 N2N REACTION REVW -CONF F ISS 7 7 H R 5 1 ? ft U 7 7 R r N 7£j J.T .1o rP . rV o AVG OVER CF252 FISS SPEC. CFD

C u 05 3 N ALPHA REAC REVW -CONF FISS 7 7N BS 1 ? ft Ji 7 7 r\ V 7 ±T .1o rP » CS AVG OVER U235 FISS SPEC. CFD

Z N 064 N PROTON REVW -CONF FISS 7 7N BS 128 4 7 7 RCN ZIJP. CS AVG OVER U235 FISS SPEC. CFD

ZR Ogo N2N REACTION REVW -CONF FISS 77NBS 1 28 4 7 7 RCN ZIJP. CS AVG OVER U235 FISS SPEC. CFD.

NB V 7 J N2N REACTION REVW -CONF FISS 77NBS 1 28 4 7 7 RCN ZIJP. CS AVG OVER U235 FISS SPEC. CFD

RH 103 TOT INELASTI REVW -CONF FISS 7 7N BS 1 28 4 7 7 RCN ZI JP . CS AVG OVER U235 FISS SPEC. CFD

RH 103 TOT INELASTI REVW -CONF FISS 77NBS 1 28 477 RCN ZI JP . CS AVG OVER CF252 FISS SPEC CFD

IN 1 1 5 TOT INELASTI REVW -CONF FISS 77NBS 128 477 RCN ZI JP . CS AVG OVER U235 FISS SPEC. CFD

IN 1 1 5 TOT INELASTI REVW -CONF FISS 77NBS 1 28 477 RCN ZI JP . CS AVG OVER CF252 FISS SPEC CFD

IN 1 1 5 RES INT ABS REVW -CONF 50-1 77NBS 128 477 RCN ZI JP . TBL . AVG OVER 1/E SPECTRUM.

IN 1 1

5

N , GAMMA REVW -CONF 25-2 77NBS 1 28 477 RCN ZIJP. TBL .COMPARISON OF 2200 M/S CS.

IN 1 1 5 N, GAMMA REVW -CONF FISS 7 7N BS 1 28 4 7 7 RCN ZI JP . CS AVG OVER U235 FISS SPEC. CFD

IN 1 1 5 N , GAMMA REVW -CONF FISS 77NBS 1 28 477 RCN ZIJP. CS AVG OVER CF252 FISS SPEC CFD

I 1 27 N2N REACTION REVW -CONF FISS 77NBS 128 477 RCN ZI JP . CS TBL. U235 FISS SPEC MDLS CFD.

IR 1 9 1 RESwN PARAMS E VA L -CONF 10-1 10 + 0 77NBS 3 1 9 477 BAR JAMES . NARROW RES FOR E STANDARD GVN.

AU 197 RES INT ABS REVW -CONF 50-1 77NBS 1 28 477 RCN ZI JP . TBL . AVG OVER 1/E SPECTRUM.

AU 197 N , GAMMA REVW -CONF 25-2 77NBS 128 477 RCN ZIJP. TBL .COMPARISON OF 2200 M/S CS.

AU 197 N, GAMMA REVW -CONF FISS 77NBS 1 28 477 RCN ZI JP . CS AVG OVER U235 FISS SPEC. CFD

AU 1 97 N , GAMMA REVW -CONF FISS 77NBS 128 477 RCN ZIJP. CS AVG OVER CF252 FISS SPEC CFD

AU N GAMMA REVW -CONF 10 + 5 14 + 7 77NBS 1 65 477 GEL PAULSEN

.

NORM .GRPHS. STANDARD STATUS.

PB 205 RESON PARAMS EVAL -CONF 1 0+0 1 0 + 5 77NBS 319 477 HAR JAMES . NARROW RES FOR E STANDARD GVN.

TH 232 N, GAMMA REVW -CONF 25-2 77NBS 128 477 RCN ZI JP . TBL .COMPARISON OF 2200 M/S CS.

TH 232 RES INT CAPT REVW.-CONF 50-1 77NBS 1 28 477 RCN ZI JP . TBL . AVG OVER 1/E SPECTRUM.

TH 232 FISSION REVW -CONF FISS 77NBS 128 477 RCN ZI JP . CS AVG OVER U235 FISS SPEC. CFD

U 233 ABSORPTION EVAL -CONF 25-2 77NBS 170 477 lAE LEMMEL.2200M/S VAL VS MAXW VAL.TBL.

U 233 N, GAMMA EVAL -CONF 25-2 77NBS 170 477 lAE LEMMEL.2200M/S VAL VS MAXW VAL.TBL.
ELEMENT QUANTITY

S A TYPE ENERGY DOCUMENTATION LAB

MIN MAX REF VOL PAGE DATE

COMMENTS

U 233 FISSION EVAL-CONF 25-2 77NBS 170 477

U 233 FISSION EVAL-CONF 25-2 77NBS 170 477

U 233 FISSION EVAL-CONF 25-2 77NBS 182 477

U 233 FISSION EXPT-CONF NDG 77NBS 304 477

U 233 ALPHA EVAL-CONF 25-2 77NBS 170 477

U 233 ETA EVAL-CONF 25-2 77NBS 170 477

U 233 NUBAR,(NU) EVAL-CONF 25-2 77NBS 170 477

U 233 FISS YIELD REVW-CONF FAST 77NBS 146 477

U 233 FISS YIELD REVW-CONF FISS 77NBS 299 477

U 235 ABSORPTION EVAL-CONF 25-2 77NBS 170 477

U 235 N, GAMMA EVAL-CONF 25-2 77NBS 170 477

U 235 FISSION REVW-CONF 25-2 77NBS 128 477

U 235 FISSION REVW-CONF FISS 77NBS 126 477

U 235 FISSION REVW-CONF FISS 77NBS 128 477

U 235 FISSION REVW-CONF FISS 77NBS 155 477

U 235 FISSION REVW-CONF 25-2 45+0 77NBS 174 477

U 235 FISSION EVAL-CONF 25-2 77NBS 182 477

U 235 FISSION EVAL-CONF 10+5 20+7 77NBS 261 477

U 235 FISSION REVW-CONF FISS 77NBS 299 477

U 235 FISSION REVW-CONF FISS 77NBS 299 477

U 235 FISSION EXPT-CONF 14+5 96+5 77NBS 304 477

U 235 FISSION EXPT-CONF FISS 77NBS 313 477

U 235 FISSION EXPT-CONF 15+7 77NBS 313 477

U 235 RES INT FISS REVW-CONF 50-1 77NBS 128 477

U 235 RES INT FISS REVW-CONF 70+0 10+3 77NBS 174 477

U 235 ALPHA EVAL-CONF 25-2 77NBS 170 477

U 235 ETA EVAL-CONF 25-2 77NBS 170 477

U 235 NUBAR,(NU) EVAL-CONF 25-2 77NBS 170 477

U 235 SPECT FISS N REVW-CONF 10-3 10+7 77NBS 156 477

U 235 SPECT FISS N REVW-CONF 10+4 21+6 77NBS 198 477

U 235 FISS YIELD REVW-CONF 25-2 77NBS 146 477

U 235 FISS YIELD REVW-CONF FAST 77NBS 146 477

U 236 N, GAMMA REVW-CONF 25-2 77NBS 128 477

U 238 RES INT CAPT REVW-CONF 50-1 77NBS 128 477

U 238 FISSION REVW-CONF FISS 77NBS 128 477

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

AUA BOLDEMAN . RE-EVAL. TBL. RECOMMENDED VAL

MHO KNOLL. PHOTO N SOURCE TBD.NDG

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

INL MAECK.MASS SPEC YLDS MEAS TBD

NBS GILLIAM. ILRR R ES U LTS . TBLS

.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

RON ZIJP. TBL. COMPARISON OF 2200 M/S CS.

RON ZIJP.CS AVG OVER U235 FISS SPEC.CFD

RCN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

NBS GRUNDL+U235 , CF252 FISS SPEC. TBLS.

ORL PEELLE+THR NORM TECH S T U D Y . TBLS , G RPH

AUA BOLDEMAN . RE-EVAL. TBL. RECOMMENDED VAL

ANL POENITZ .TBL, GRPHS. MUST UPDATE ENDF5B

NBS GILLIAM. CF252 SPEC AVG CS. TBL. CFD.

NBS GILLIAM. REL PU239.DATA FOR 5 FIELDS

MHG KNOLL. PHOTO N SO U R C E . 4 ES . TB

L

RI ADAM0V + CF252 S PEC . TB L . ABSL CS.CFD

RI ADAMOV+ABSL CS. TBL. CFD OTH.

RCN ZIJP. TBL. AVG OVER 1/E SPECTRUM.

ORL PEELLE + 2E R A NGES . 7 - 1 OE V , . 1 - 1 KE V . TBLS

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

NBS GRUNDL+DET SPEC RESPONSE . GRPHS

.

LAS STEWART+GRPH, TBLS. CFD WATT, MAXW SPEC

INL MAECK.REL YLDS MEAS IN PROGRESS. TBC

INL MAECK. GRPHS. MASS SPEC YLDS MEAS TBC

RCN ZIJP . TBL . COMPARISON OF 2200 M/S CS.

RCN ZIJP. TBL. AVG OVER 1/E SPECTRUM.

RCN ZIJP.CS AVG OVER U235 FISS SPEC.CFD

368
ELEMENT QUANTITY

S A TYPE ENERGY DOCUMENTATION LAB

MIN MAX REF VOL PAGE DATE

COMMENTS

U 238 FISSION REVW-CONF FISS 77NBS

U 238 FISSION REVW-CONF FISS 77NBS

U 238 FISSION EXPT-CONF 10+5 20+7 77NBS

U 238 FISSION REVW-CONF FISS 77NBS

U 238 FISSION REVW-CONF FISS 77NBS

U 238 FISSION EXPT-CONF FISS 77NBS

U 238 FISSION EXPT-CONF 15+7 77NBS

U 238 SPECT FISS N REVW-CONF +6 60+6 77NBS

U 238 FISS YIELD REVW-CONF FAST 77NBS

U 238 FISS YIELD REVW-CONF FISS 77NBS

U 238 RESON PARAMS EVAL-CONF 10+0 10+4 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION REVW-CONF 25-2 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION EXPT-CONF 10+5 20+7 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION REVW-CONF FISS 77NBS

NP 237 FISSION EXPT-CONF NDG 77NBS

NP 237 FISSION EXPT-CONF FISS 77NBS

NP 237 FISSION EXPT-CONF 15+7 77NBS

NP 237 SPECT FISS N REVW-CONF +5 ^0+6 77NBS

NP 237 FISS YIELD REVW-CONF FAST 77NBS

PU 239 ABSORPTION EVAL-CONF 25-2 77NBS

PU 239 N, GAMMA EVAL-CONF 25-2 77NBS

PU 239 FISSION REVW-CONF 25-2 77NBS

PU 239 FISSION REVW-CONF FISS 77NBS

PU 239 FISSION REVW-CONF FISS 77NBS

PU 239 FISSION REVW-CONF FISS 77NBS

PU 239 FISSION EVAL-CONF 25-2 77NBS

PU 239 FISSION EVAL-CONF 25-2 77NBS

PU 239 FISSION REVW-CONF FISS 77NBS

PU 239 FISSION EXPT-CONF 14+5 96+5 77NBS

PU 239 RES INT FISS REVW-CONF 50-1 77NBS

128 477 RCN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

156 477 NBS GRUNDL+U 235 , CF252 FISS SPEC.TBLS.

278 477 KFK CIERJACKS . POSSIBLE STANDARD . DATA GVN

299 477 NBS GILLIAM. CF252 SPEC AVG CS.TBL.CFD.

299 477 NBS GILLIAM. REL PU239.DATA FOR 5 FIELDS

313 477 RI ADAM0V+CF252 SPEC . TBL . ABSL CS.CFD

313 477 RI ADAMOV+ABSL CS.TBL.CFD OTH.

156 477 NBS GRUNDL+DET SPEC RESPONSE . GRPHS

.

146 477 INL MAECK.MASS SPEC YLDS MEAS TBD

299 477 NBS GILLIAM. ILRR RESULTS . TBLS

.

319 477 HAR JAMES. NARROW RES FOR E STANDARD GVN.

128 477 RCN ZIJP.CS TBL.U235 FISS SPEC MDLS CFD.

128 477 RCN ZIJP. TBL. COMPARISON OF 2200 M/S CS.

128 477 RCN ZIJP.CS AVG OVER U235 FISS SPEC. CFD

128 477 RCN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

156 477 NBS GRUNDL+U 235 , CF252 FISS SPEC.TBLS.

278 477 KFK CIERJACKS . POSSIBLE STANDARD . DATA GVN

299 477 NBS GILLIAM. CF252 SPEC AVG CS.TBL.CFD.

299 477 NBS GILLIAM. REL PU239.DATA FOR 5 FIELDS

304 477 MHG KNOLL. PHOTO N SOURCE TBD. NDG

313 477 RI ADAM0V+CF252 SPEC . TBL . ABSL CS.CFD

313 477 RI ADAMOV+ABSL CS.TBL.CFD OTH.

1 56 477 NBS GRUNDL + DET SPEC RESPONSE . GRPHS

.

146 477 INL MAECK.MASS SPEC YLDS MEAS TBD

170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

1 28 477 RCN Z I J P . TBL . COMP A R ISON OF 2200 M/S CS.

128 477 RCN ZIJP.CS AVG OVER U235 FISS SPEC. CFD

128 477 RCN ZIJP.CS AVG OVER CF252 FISS SPEC CFD

156 477 NBS GRUNDL+U235 , CF252 FISS SPEC.TBLS.

170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

182 477 AUA BOLDEM AN . RE-E VAL . TBL . RECOMMENDED VAL

299 477 NBS GILLIAM. CF252 SPEC AVG CS.TBL.CFD.

304 477 MHG KNOLL. PHOTO N SOURCE . 4ES . TBL

128 477 RCN ZIJP.TBL.AVG OVER 1/E SPECTRUM.

369
ELEMENT QUANTITY

S A TYPE ENERGY DOCUMENTATION LABMIN MAX REF VOL PAGE DATE

COMMENTS

PU 239 ALPHA EVAL-CONF 25-2 77NBS 170 477 lAE L EMMEL . 2 2 0 OM / S VAL VS MAXW VAL.TBL.

PU 239 ETA EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

PU 239 NUBAR,(NU) EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

PU 239 SPECT FISS N REVW-CONF 1 0-3 30 + 6 77NBS 156 477 NBS GRUNDL + DET SPEC R ES PO NS E . G fi P HS

.

PU 239 SPECT FISS N REVW-CONF 10+4 21+6 77NBS 198 477 LAS STEW ART+GRPH , TBLS . CFD WATT, MAXW SPEC

PU 239 FISS YIELD REVW-CONF 25-2 77NBS 146 477 INL MAECK.REL YLDS MEAS IN PROGRESS. TBC

PU 239 FISS YIELD REVW-CONF FAST 77NBS 146 477 INL MAECK . GRPHS . MASS SPEC YLDS MEAS TBC.

PU 239 FISS YIELD REVW-CONF FISS 77NBS 299 477 NBS GILLIAM. ILRR R ES U LTS . TBLS

.

PU 240 FISS YIELD REVW-CONF FAST 77NBS 146 477 INL MAECK. MASS SPEC YLDS MEAS TBD.

PU 241 ABSORPTION EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL . 2200M/S VAL VS MAXW VAL.TBL.

PU 241 N, GAMMA E V A L-C ON F 2 5 -2 77NBS 170 477 lAE LEMMEL. 2200M/S VAL VS MAXW VAL.TBL.

PU 241 FISSION EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL . 2200M/S VAL VS MAXW VAL.TBL.

PU 241 FISSION EVAL-CONF 25-2 77NBS 182 477 AUA BOLDEM AN . RE-E V A L . TBL . RECOMMENDED VAL

PU 241 ALPHA EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL . 2200M/S VAL VS MAXW VAL.TBL,

PU 241 ETA EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL . 2200M/S VAL VS MAXW VAL.TBL.

PU 241 NUBAR,(NU) EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL . 2 200M/ S VAL VS MAXW VAL.TBL.

PU 241 FISS YIELD REVW-CONF FAST 77NBS 146 477 INL MAECK. MASS SPEC YLDS MEAS TBD.

PU 242 FISS YIELD REVW-CONF FAST 77NBS 146 477 INL MAECK. MASS SPEC YLDS MEAS TBD.

AM 241 FISS YIELD REVW-CONF FAST 77NBS 146 477 INL MAECK. MASS SPEC YLDS MEAS TBD.

CF 252 NUBAR,(NU) EVAL-CONF 25-2 77NBS 170 477 lAE LEMMEL. VAL FOR 2200M/S DATA FIT.

CF 252 NUBAR,(NU) EVAL-CONF SPON 77NBS 182 477 AUA BO LDEM AN . RE -E V A L . TBL . RECOMMENDE D VAL

CF 252 NUBAR,(NU) EXPT-CONF SPON 77NBS 194 477 RI BLINOV+PRELIM VAL GVN.MEAS TBC,

CF 252 SPECT FISS N EXPT-CONF SPON 77NBS 194 477 RI BLIN0V+T0F.2 DETS . GRPHS . MAXW . TBC

.

CF 252 SPECT FISS N REVW-CONF SPON 77NBS 198 477 LAS STEW ART+GRPH . CFD WATT, MAXW SPEC.

370
NBS-114A (REV. 7-73)

U.S. DEPT. OF COMM. 1 . PU B L IC ATION OR R E PORT NO . 2, Gov't Accession

BIBLIOGRAPHIC DATA No.

SHEET NBS SP-493

3. Recipient's Accession No.

4. TITLE AND SUBTITLE NEUTRON STANDARDS AND APPLICATIONS

Proceedings of the International Specialists Symposium on

Neutron Standards and Applications, Held at the National

Bureau of Standards, Gaithersburs. MD. March 28-31. 1977

5. Publication Date

October 1977

6. Performing Organization Code

7. AXOfXTO^O Editors:

C. D. Bowman, A. D. Carlson, H. 0. Liskien and L. Stewart

8. Performing Organ. Report No.

9. PERFORMING ORGANIZATION NAME AND ADDRESS

NATIONAL BUREAU OF STANDARDS

DEPARTMENT OF COMMERCE

WASHINGTON, D.C. 20234

10. Project/Task/Work Unit No.

11 z^" 1 r- XT 11. Contract/Grant No.

12. Sponsoring Organization Name and Complete Address rSfree^ City State, ZIP)

National Bureau of Standards; Central Bureau of Nuclear

Measurements; U.S. Energy Research & Development Adm. ; Electric

Power Research Institute (U.S.A.); American Nuclear Society;

American Phvsical Societvi Nuclear Tilriererv Afpnpv ^^TihiTTcnp^ • International Union of Pure & Applied Physics; with the

cooperation of the International Atomic Energy Agency

13. Type of Report & Period

Covered

14. Sponsoring Agency Code

15. SUPPLEMENTARY NOTES

Library of Congress Catalog Card Number; 77-14317

16. ABSTRACT (A 200-word or less factual summary of most significant information. If document includes a significant

bibliography or literature survey, mention it here.)

These proceedings contain forty-seven papers, which were presented at the

International Specialists Symposium on Neutron Standards and Applications held

at the National Bureau of Standards on March 28-31, 1977. The topics

addressed at the Symposium include light-element cross section standards,

capture and fission cross section standards, integral neutron standards, flux

measuring techniques, and medical and personnel dosimetry.

17. KEY WORDS (six to twelve entries; alphabetical order; capitalize only the first letter of the first key word unless a proper name; separated by semicolons)

Cross section standards; dosimetry; fission; flux; measuring techniques; neutrons;

standards.

18. AVAILABILITY [x] Unlimited 19. SECURITY CLASS

(THIS REPORT)

21. NO. OF PAGES

1 1 For Official Distribution. Do Not Release to NTIS

UNCL ASSIFIED 379

IxJ Order From Sup. of Doc, U.S. Government Printing Office

Washington. D.C. 20402. SD Cat. No. Cl^lOil^Q'^

20. SECURITY CLASS

(THIS PAGE)

22. Price

1 1 Order From National Technical Information Service (NTIS)

Springfield, Virginia 22151 UNCLASSIFIED

$8.50

USCOMM-DC 29042-P74

iV U.S. GOVERNMENT PRINTING OFFICE : 1977 O—246-327

i