Determination of Absorbed Dose in Reactors

268 Pages·2011·5.63 MB·English

THERMAL COLUMN HEATING
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DISTANCE FROM INNER FACE (сm .)
TECHNICAL REPORTS SERIES No. 1 27
Determination
of Absorbed Dose
in Reactors
Vfty S¡ INTERNATIONAL ATOMIC ENERGY AGENCY, VIENNA, 1971

DETERMINATION OF
ABSORBED DOSE IN REACTORS
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TECHNICAL REPORTS SERIES No.127
DETERMINATION OF
ABSORBED DOSE IN REACTORS
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 1971
DETERMINATION OF ABSORBED DOSE IN REACTORS
IAEA, VIENNA, 1971
STI/DOC /10/127
FOREWORD
There are many areas in the use and operation of research reactors
where the absorbed dose and the neutron fluence are required. These in-
clude work on the determination of the radiolytic stability of the coolant
and moderator and on the determination of radiation damage in structural
materials, and reactor experiments involving radiation chemistry and
radiation biology. The requirements range from rough estimates of the
total heating due to radiation to precise values specifying the contributions
of gamma rays, thermal neutrons and fast neutrons. To meet all these
requirements a variety of experimental measurements and calculations as
well as a knowledge of reactor radiations and their interactions is necessary.
Realizing the complexity and importance of this field, its develop-
ment at widely separated laboratories and the need to bring the experts in
this work together, the IAEA has convened three panel meetings. These
were: Чп-pile dosimetry1, held in July 1964 (published by the Agency as
Technical Reports Series No. 46); 'Neutron fluence measurements', in
October 1965; and 'In-pile dosimetry', in November 1966. The recom-
mendations of these three panels led the Agency to form a Working Group
on Reactor Radiation Measurements and to commission the writing of this
book and a book on Neutron Fluence Measurements. The latter was pub-
lished in May 1970 (Technical Reports Series No. 107).
The material on neutron fluence and absorbed dose measurements is
widely scattered in reports and reviews. It was considered that it was
time for all relevant information to be evaluated and put together in the
form of a practical guide that would be valuable both to experienced workers
and beginners in the field.
The present manual was edited by Dr. A.W. Boyd of the Chalk River
Nuclear Laboratories of Atomic Energy of Canada Limited, Chalk River,
Ontario, Canada.
The authors who contributed to this manual are:
Chapter I: Introduction A.W. Boyd
Chapter II: Reactor radiation A.W. Boyd
and reactor irradiation
experiments
Chapter III: Dosimetry problems A.W. Boyd
Chapter IV: Radiation spectra V. Kadlec (Nuclear Research
Institute, Rez near Prague,
Czechoslovak Socialist Republic)
Chapter V: Calculation of A.W. Boyd
absorbed dose
Chapter VI: Calorimetry B. Sautiez (CEN, Saclay, France)
Chapter VII: Ionization chambers K.K. Mehta (Whiteshell Nuclear
Research Establishment, Pinawa,
Man., Canada)
Chapter VIII: Chemical dosimetry I. Draganic (Boris Kidric Insti-
tute, Vinca, Yugoslavia)
Chapter IX: Solid state S. Hayakawa (Tokyo Institute of
dosimeters Technology, Tokyo, Japan).
The original contributions have been substantially changed during edit-
ing in order to make the manual consistent. The editor and contributing
authors wish to express their appreciation and thanks to Dr. A. Keddar
(IAEA), the present Scientific Secretary of the Working Group, for his help
in assembling the manual, and for his comments and help during the
planning of the book and the final editorial work.
Particular thanks are expressed to the editor. Dr. A.W. Boyd, who
accomplished much in a very short time.
CONTENTS
CHAPTER I. INTRODUCTION 1
CHAPTER II. REACTOR RADIATION AND
REACTOR IRRADIATION EXPERIMENTS 7
II. 1. Reactor Radiation 7
II. 1.1. Neutrons
II. 1.2. Gamma rays
II. 1.3. Beta particles (and electrons)
II. 2. Absorbed dose measurement required 18
II. 2. 1. Absorbed dose and absorbed dose rate measurements
II. 2. 2. Absolute measurements using dosimeters
II. 2.3. Relative measurements using monitors
II. 2. 4. Total dose and the dose from different components of
reactor radiation
CHAPTER III. DOSIMETRY PROBLEMS 27
III. 1. Problems of penetration and range of primary and
secondary radiation 27
III. 1. 1. Kerma and absorbed dose
III. 1.2. Self-shielding and secondary radiation loss
III. 1.3. Secondary irradiation from surrounding material
III. 1.4. Importance of gain and loss of secondary radiation
for the different states of matter
III. 2. Selection of the dosimetry method 29
III. 2.1. The degradation of the energy transferred and
the effects produced
III. 2. 2. Selection of the method
CHAPTER IV. RADIATION SPECTRA 37
IV. 1. The neutron spectra 37
IV. 1.1. Spectra in different regions of a reactor
IV. 1.2. Variations of spectra with time
IV. 2. Gamma spectra 54
IV. 2.1. Gamma sources and source spectra
IV. 2. 2. Scattered radiation and the low energy part of
the spectra
IV. 2. 3. Spectra in different regions of a reactor
IV. 2. 4. Variation of spectra with time
IV. 2. 5. Combined spectrum for an experiment of the
reactor and radiation equipment
IV.2.6. Gamma spectra measurements
CHAPTER V. CALCULATION OF ABSORBED DOSE 85
V.l. Introduction 85
V.2. Calculation of absorbed dose from reactor design 85
V.2.1. Average absorbed dose rate in the core of
large reactors
V.2.2. Average absorbed dose in the core of small reactors
V.2.3. Absorbed dose rate in annular fuel elements
V. 3. Calculation of absorbed dose from
surrounding materials 92
V. 3. 1. Dose from capture gamma rays from
surrounding materials
V.3.2. Dose from beta particles from
surrounding materials
V.3.3. Dose from alpha particles from
surrounding materials
V.4. Calculation of fast neutron scattering dose from fast neutron
flux density measurements 97
V.5. Attenuation and s elf-shielding 99
V. 5. 1. Attenuation and self-shielding of gamma rays
V. 5. 2. Attenuation and self-shielding of thermal neutrons
V. 5. 3. Fast neutron attenuation and self-shielding
V.6. Secondary radiation loss ' 105
V. 7. Analysis of total absorbed dose data to determine the dose
from components of the radiation 116
CHAPTER VI. СALORIMETRY 121
VI. 1. The value of calorimetry and the problems of its use
in reactors 121
VI. 2. Theory of calorimetry. Different types and
methods of measurement 123
VI. 2. 1. Different types of calorimeters
VI. 2. 2. Different types of utilization and of
calibration of calorimeters
VI. 2.3. Characteristics of the heat transfer or
insulating medium
VI. 2. 4. Sensitivity range of measurements
VI. 3. Technological problems 133
VI. 3.1. Problems concerning materials
VI.3.2. Temperature measurements
VI. 3.3. Calibration
VI. 3.4. Duration of the life of a calorimeter
VI. 4. Corrections and errors 143
VI. 5. Description of calorimeters 147
VI. 5.1. Calorimeters with a gaseous heat
transfer medium