understanding neutron radiography reading vii nrhb part 1 of 2

Understanding Neutron RadiographyReading VII-NRHB Part 1 of 2Principles And Practice Of Neutron RadiographyMy ASNT Level III, Pre-Exam Preparatory Self Study Notes 15 July 2015

Charlie Chong/ Fion Zhang

Nuclear Power Reactorsapplications



Submarine Nuclear Pile

The Magical Book of Neutron Radiography



ASNT Certification GuideNDT Level III / PdM Level IIINR - Neutron Radiographic TestingLength: 4 hours Questions: 135

1. Principles/Theory• Nature of penetrating radiation• Interaction between penetrating radiation and matter• Neutron radiography imaging• Radiometry

2. Equipment/Materials• Sources of neutrons• Radiation detectors• Non-imaging devices


• Electron emission radiography• Micro-radiography• Laminography (tomography)• Control of diffraction effects• Panoramic exposures• Gaging• Real time imaging• Image analysis techniques

3. Techniques/Calibrations• Blocking and filtering• Multifilm technique• Enlargement and projection• Stereoradiography• Triangulation methods• Autoradiography• Flash Radiography• In-motion radiography• Fluoroscopy


4. Interpretation/Evaluation• Image-object relationships• Material considerations• Codes, standards, and specifications

5. Procedures• Imaging considerations• Film processing• Viewing of radiographs• Judging radiographic quality

6. Safety and Health• Exposure hazards• Methods of controlling radiation exposure• Operation and emergency procedures

Reference Catalog NumberNDT Handbook, Third Edition: Volume 4,Radiographic Testing 144ASM Handbook Vol. 17, NDE and QC 105

Fion Zhang at Shanghai15th July 2015

http://meilishouxihu.blog.163.com/


Greek Alphabet


sound sound

A a [a] nu

B [b] [v] - [ks] [ks] --

[o]

/)fj [d] lT

EE [e] p [r] [r]

z~ [zd]A

11 ] [i] TT [t] [t]

08 ] [I]

I I [I..] [i] ¢qp [f]

K ~ [c] X ~ [~]

[I] 4Jtp [ps] [ps]

Q

Greek Alphabet


Charlie Chong/ Fion Zhang http://greekhouseoffonts.com/

A Alpha r Gamma NNu T Tau (al-fah) (gam-ah) (new) (taw)

B Beta (bay-tah)

HEta (ay-tah)

Q Omicron ( om-e-cron)

y Upsilon (up-si-lon)

X Chi I Iota II Pi Q Omega (kie) (eye-a-tah) (pie) ( oh-may-gah)

p q

~Delta K Kappa 8 Theta ~ Xi (del-ta) (cap-pah) (thay-tah) h d (zie)

E Epsilon A Lambda p Rho 'I' Psi (ep-si-lon) (lamb-da h) (roe) (sigh)

<l>Phi MMu L Sigma Z Zeta (fie) (mew) (sig-ma) (zay-tah)


GREEK ALPHABET

A a ALPHA [a] &.A cpa

Hfl ETA [c:] ijra

Nv NU [n] vii

TT TAU [t] raii

BP BETA [b] f3ryra

ee THETA [th] er,ra

XI [ks] ~El

Yu UPSILON [H] i5 1/flAOV

ry GAMMA [g] y&.f-tf-ta

IL IOTA [i] iwra

Oo OMICRON [o] o f-tlKp6v

~() DELTA [d] 0£;\ra

KK KAPPA [k] Kanrra

Iln PI [p] rrc:l

By Ue11 Crowder • b£'11trowdeuu·t • Last modUicd .2 \fay .201.2

EE EPSILON [e] £ 1/flAOV

AA LAMBDA [I] ACxf-tf38a

Pp RHO [r] pw

PSI [ps]

1/fEl

z~ ZETA [dz] (ryra

M~ MU [m]

f-tii

SIGMA [s] (Jtwa

Ow OMEGA [::>:]

Wf-tiya


INTRODUCTIONRadiography with neutrons can yield important information not obtainable by more traditional methods. In contrast to X-rays as the major tool of visual non-estructive testing, neutrons can be attenuated by light materials like water,hydrocarbons, boron, penetrate through heavy materials like steel, lead,uranium, distinguish between different isotopes of certain elements, supplyhigh quality radiographs of highly radioactive components. These advantageshave led to multiple applications of neutron radiography since 1955, both fornon-nuclear and nuclear problems of quality assurance. The required neutronbeams originate from radioisotopic sources, accelerator targets, or researchreactors. Energy "tailoring" which strongly influences the interaction withcertain materials adds to the versatility of the method.


Since about 1970 norms and standards have been introduced and reviewed both in Europe (Birmingham, September 1973) and the United States(Gaithersburg, February 1975). The first world conference on neutronradiography will take place in December 1981, in San Diego, U.S.A. . InEurope the interested laboratories inside the European Community haveentered into systematic collaboration through the Neutron RadiographyWorking Group (NRWG), since May 1979. This Handbook has been compiledas one of the common tasks undertaken by the Group. Its principal authorsare J.C. Domanus (Rise National Laboratory), and R.S. Matfield (Joint Research Centre, Ispra).

This Handbook documents the availability, not only of a large number of research reactorbased facilities in the Community, but also of advancedequipment and solid expertise for the interpretation of neutron radiographs,serving present and future needs of Europe's industry.


1. PRINCIPLES AND PRACTICE OF NEUTRONRADIOGRAPHYThis part of the Handbook is about neutrons, radiography, and the techniquethat has been developed to bring them together. It is written in three chapters,a description of the subject for the assistance of the clients of neutronradiography services; a discussion on the problems facing the designer ofneutron radiography equipment and a description of some of the applications.The special terms used are explained in Appendix 1.1.


1.1. INTRODUCTION TO NEUTRON RADIOGRAPHY1.1.1 HistoricalHistorically, radiography came first in 1895 with the discovery by Röntgen of aradiation which he called X-rays. He rapidly realised the technical implicationsand in the same year took an X-ray 'photograph' of a weld in a zinc plate. Thesignificance of X-rays for the detection of unseen flaws was immediately seen by other workers, and experimental X-radiographs were soon produced inlaboratories in Europe and the U.S.A. It was later found that the attenuation ofX-rays increased smoothly with atomic number, indicating that the X-raysinteracted with the orbital electrons around the atomic nucleus. The discovery of the neutron is credited to Chadwick who, in 1932, related and hypothesised on the work of Bothe, Becker, Curie and others and assumed that the penetrating radiation produced by bombarding beryllium with alpha particleswas neither positively nor negatively charged; so he called it the neutron (from Latin neuter meaning neither).


He had indentified a particle which, together with the proton, was one of the basic building bricks of matter. The radiographic applications for neutrons were not acted upon quite so rapidly as had occured with X-rays and several years intervened before the first neutron radiography experiments were started in Berlin by Kaliman and Kuhn [Ref. 1]. They started work in 1935 with a small accelerator source, said to be equivalent to a 2-3 gramme Ra-Be source, and they defined the basic principles of neutron radiography and recorded them on a large number of patents filed over the next ten years or so.


The publication of their work was delayed by the second World War and it was not until 1947 that they revealed the thoroughnes of their investigationsby describing most of the basic techniques in use today. They suffered thedisappointment of being preceded by Peters [Ref. 2] who published theresults of similar experiments in 1946. The next development had to await theadvent of nuclear reactors, and the first reactor neutron radiographs wereproduced in 1 956 by Thewlis and Derbyshire [Ref. 3] at Harwell. They carriedout their work with the BEPO reactor (BEPO stood for British Experimental Pile with the “O” ), and its intense neutron beam allowed them to produce radiographs of much better quality than those of Kaliman and Peters.

More reading on BEPOhttp://www.research-sites.com/UserFiles/File/publications/project-info/harwell-BEPO.pdf


BEPO stood for British Experimental Pile with the “O”


BEPO stood for British Experimental Pile with the “O”

http://petapixel.com/2013/02/18/photos-from-the-worlds-first-underwater-nuclear-explosion/

■ https://www.youtube.com/embed/_dwX8FIuiIo


They also demonstrated the applications of neutron radiography to specificproblems by showing the flaws in a uranium cylinder, a defect in a piece ofboral (boron-aluminium sandwich) and the fine structure of plant tissue. Thetechnique developed slowly for several years until problems associated withthe radiography of radioactive materials encouraged its more active revival.

Several researchers reported their work in the early 1960's. But it wasprincipally the work of Berger [Ref. 4] of Argonne Laboratories in the U.S.A.,followed by Barton [Ref. 5] at Birmingham University that led to its revival.Interest expanded rapidly and Krolick [Ref. 6] et al reported in 1968 that therewere 33 centres throughout six different countries all active in neutronradiography. At that time there were 46 reactor facilities in use, threeaccelerators and above five isotopie sources in use of being built. Thesituation is much the same today in that the reactor sources predominate, andthere are still very few accelerator or isotopie sources. The number of activecentres however, is now probably over 50.


1.1.2. Basic ConceptsAll material objects are formed from a substance which we call matter. This isan arrangement of atoms which can take many forms varying from the regularpattern of a crystal lattice to the free moving single atoms within a gas plasma.

No one has ever seen an atom although the electron microscope allows us toget very close to seeing it and modern theory represents it as a tiny nucleussurrounded by a diffuse cloud of electrons, the outer boundary of which is notclearly defined and may not even be spherical. The nucleus is itself a group of closely bound neutrons and protons, the overall diameter of which is some10,000 times smaller than the size of the atom. For our purposes we willimagine the atom as consisting of an extremely small, extremely dense,nucleus surrounded by an enormous empty space (on the nuclear scale) inwhich a retinue 随从 of electrons maintain their regular orbital motions.


The radiographic process requires free neutrons and so they must be dislodged from the nucleus. This is achieved by bombarding the nucleus andcausing it to change into smaller nuclei and a number of free neutrons. These liberated neutrons are electrically neutral (i.e. no charge) and so are able to pass through the electron cloud surrounding an atom without disturbinginteractions.


Unlike the X-ray which interacts with the electron cloud, the neutron interaction is not characterised by a rational dependence on the atomicnumber of the object, the relationship between the two being quite random.There are practically no generalisations that can be made which relateneutron characteristics to atomic mass or atomic number, and eachinteraction of a neutron with an atom of a particular nuclide is unique, thenature of that reaction being only related to the energy of the neutron. Toproduce a neutron radiograph we must have a continuous supply of freeneutrons, and these must be directed onto the object to be radiographed. Thisobject will modify the neutron beam by (1) scattering or (2) absorbing the radiation, and the beam reaching the detector will have an intensity patternrepresentative of the structure of the object.


1.1.3. Neutron SourcesNeutrons are produced in three ways: from an accelator, a radioisotope, or anuclear reactor. In each case they are removed from an atom by a nucleartransmutation process and they emerge over the enormous energy range of 1013 electron-volts, that is from 10-4 to 109 eV. The energy of most interest forneutron radiography is about 0.03 eV, (thermal neutron: 0.03~0.1 eV?) for it is at this energy that the detectors used for neutron radiography are usually most efficient, except where the resonance characteristics (epithermal neutron / resonance energy neutron?) of the detector foil can be utilized (see1.1.9).


1.1.3.1 AcceleratorsThis is a general name given to machines that accelerate a beam of chargedparticles (protons, deutrons, alphas etc.) and directs them onto a target (seeFig. 1.1). An interaction takes place between the bombarding particles andthe target atoms, and this results in the expulsion of other particles. Withparticular combinations of incident particle and target material the ejectedparticles are neutrons. To remove a neutron from a target atom the energy ofthe bombarding particle must exceed the nuclear potential barrier surroundingthe nucleus.

Protons 11P, deutrons 1

2H, 24He alphas etc.

expulsion of other particles


This energy varies with both the target material and the charge on thebombarding particle, and so the target used in a particular type of acceleratoris matched to the energy of incident particle that the machine can produce.Typical of this system is the machine which uses a Penning ion source toionise the atoms of deuterium gas and uses a Cockcroft-Walton generator(100-400 kV) to accelerate them onto a tritium target ( as tritium gas absorbed in the porous Ti or Zr) .


Fig. 1.1 The Principles of a Particle Accelerator.

ACCELERATING ELECTRODES •

TARGET


The reaction takes place, that is a deutron (21D) strikes tritium (3

1T) which releases a 14.6 MeV neutron (1

0n), and is converted to helium (42He),with a

contribution of 3.6 MeV.

When higher potentials are available, such as those from a Van der Graaffgenerator, then a beryllium or lithium target is used, and the reactions are

3Li, 4Be, 5B

4

14.6 + 3.6 MeV?


An alternative system is to accelerate electrons onto a tungsten target andthereby produce X-rays. If these are directed onto a second target with a high(X,n) reaction cross-section, such as beryllium or uranium, again neutrons willbe produced. This last system has the potential of being used as a dualpurpose generator of both X-rays and neutrons. Two such machines havebeen reported, the first is a 5,5 MeV Linac which was built as an X-raymachine [Ref. 7] and then modified to produce neutrons, and the second is alarge 20 MeV Linac [Ref. 8] which was designed as a dual purpose X-ay/neutron generator. The first machine used a tungsten target, and the X-rayand gamma ray emission from this produces a 9

4Be + γ → 84Be +1

0n reaction in the beryllium. The second interchanges a tungsten and uranium target, thefirst producing X-rays and the second generating neutrons by the reaction:

23892U + γ → 237

92U + 10n


The life and output of the target used in accelerators varies with the system, and the energy of the bombarding particle. Fig.1.2 shows this variation ofneutron yield with energy for the deuterium-tritium and the deuterium -beryllium reactions ¡usually referred to as 'DT' and 'DB').

The beryllium target is used in the form of pure metal, and, providing it is adequately cooled, will not deteriorate significantly with use.

Tritium targets are produced by absorbing tritium gas in a titanium or zirconium layer on a copper plate. The neutron output is high but the lifetime (usually defined as the time required for the neutron output to fall to half its initial value) is relatively low. The early machines of this type used a continuously pumped vacuum system in which the tritium is fed to the target through a controlled leak. An alternative system used a large rotating target which increased the lifetime by simply providing a larger target area.


Fig. 1.2 Neutron Yield from Deuteron Reactions (After Hawkesworth [ Ref. 11 ]).

t 1010

0 ~

I u Ql

.!!! ,os -c: --

0 ....J !6! >-

T(dnl He-4 2·5mg /cm2 T.n TARGET

Be9(dn)810 TI-!ICK TARGET

10 6 ~--L-~~~~U----L~~~wu~ 0-1 lO 10

BOMBARDING ENERGY [Mev 1 -•


There is considerable variation in the reported life of these systems but times between 10-100 hours are usually quoted. Later designs use a sealed accelerator tube in which the problem of the depletion of the tritium in the target was overcome by feeding a mixture of deuterium and tritium into the ion source. Tritons as well as deuterone are accelerated into the target so that the net amount of tritium in the target remains about the same, and hence the neutron yield is reasonably constant. More detailed descriptions of the various types of particle accelerator are given in the reviews by Olive et al [Ref. 9], Krolick et al [Ref. 6] and Holland and Hawkesworth [Ref. 10], and details of source accelerator systems are given in Table 1.1

mixture of deuterium and tritium

titanium or zirconium layer

Cu

absorbing tritium gas


Table 1.1 Accelerator Neutron Sources

Manufacturer Type Particle Targets Operating Beam Fast Neutron materials Voltage. current neutron energy

kV mA output. -1 n s

MeV

Elliot Automation P Tube Deuteron Tritium 120 2 1011 14 20th Century Electronics NGH 150 Deuteron Tritium 150 3.5 4 X 10

11 4 Sames T Deuteron Tritium 400 3 1 011 2 High Voltage Eng. Co. Van der Graaff Deuteron Beryllium 3,000 0.6 1) 1.6 Mulfard Linac Electron Beryllium 5,500 0.2 2 X 1011 1.4

Peak thermal flux in water 10 10 n cm-2s-1


1 .1 .3.2 RadioisotopesRadioisotopes are produced by bombarding nuclei with charged particles in an accelerator or a nuclear reactor. A nucleus becomes radioactive when itchanges from a stable, unexcited, state to an unstable, excited, condition.Now, for a nucleus to be stable it must contain a particular neutron-protonratio. This ratio varies from 1 to about 1 .6 (excluding hydrogen) as the atomicnumber increases. The stable condition is referred to as the ground state, andif extra energy can be imparted to the nucleus it is said to have been raised toan excited state, from which it eventually decays back to the ground state,usually with the emission of gamma rays. In the excited condition there is nochange in the neutron-proton ratio unless the energy imparted to the nucleussufficiently exceeds the energy that binds it together for it to eject one of itsneutrons or protons. The nucleus then become unstable because it has thewrong neutronproton ratio for its particular atomic number.


So, by bombarding atoms with charged particles of sufficient energy it ispossible to raise the nucleus to a state of instability from which it will decayback to its stable state at a characteristic rate measured by the half-line (thetime take for the radioactivity to halve).

Unfortunately there are few radiosotopes which emit neutrons, and neutron production is achieved in the same manner as with accelerators, that is by allowing the gamma rays or alpha particles emitted from the radioactive isotope to bombard a neutron emitting target. The disadvantage of the radioisotope is that the activity is continuously reducing. When the radioisotope has a long half-life this is not inconvenient, but a radioisotope such as antimony loses half of its activity every 60 days and must be regularly reactivated. This, of course, adds to the cost of the neutron generator.Beryllium has the lowest neutron binding energy (1.6 keV) of all the nuclides,and is used as a target with both alpha- and gamma-emitting radioisotopes (see Fig.1.3).


Fig. 1 .3 Isotopie Neutron Sources.

http://fas.org/sgp/othergov/doe/lanl/lib-www/la-pubs/00377082.pdf

BERYLLIUM BLOCK -._ ......

AADIOACTlVE SOURCE

POLONJUM tPo:no)

ANTlMONY

GAMMA RAYS

Q. BERYUlUM

__ 1_· ----.t..,...,. BERYLLIUM

+ n

n 11 M•V

n • 21i K.eV


The neutrons produced from these reactions vary in energy up to amaximum of about 11 MeV; the lowest energy ,and ás we shall see later themost useful, coming from a combination of antimony and beryllium. This is an( γ,n ) source and like all gamma sources has the disadvantage of requiring alead shield to prevent the gamma rays from causing a health hazard. Thereare a few radioisotopes which decay by spontaneous fission (a processdescribed in the next section) but of these only califomium-252 has sufficientneutron output to be considered here.

At the time of writing, the only available supply of this material is from nuclear reactors in the U.S.A.*), and because it takes a long time to produce usable quantities it is very expensive. Fortunately, the price is falling and so in future it may be an attractive neutron source. Table 1.2 gives' details of radioisotopic sources.


Table 1.2 Radioisotopic Neutron Sources

Source Half-life

124Sb-Be 60 d 210p 8 o- e 138 d 241A B m- e 458 a 226Ra-Be 1620 a 227 Ac-Be 21,8 a 22eTh- e 1,91 a 2s2Cf 2_,65 a

Be9

+ 0 ..,

Reaction

(y.n) (a.n) (a.n) (a.n} (a.n) (a.n) fissjon

8 Be + n

Neutron yield ( - 1 -1) n .. s g

2.7.1 0 9

1.28.1 0 10

1 .. 107

1 .. 3.1 0 7

1.1~10 9

1 .. 7 .. 1 0 10

2~34.1 012

Neutron energy (MeV)

0~024 4f3

"l.J.4 'l.l4

""-'4 ~4

2J3


1 .1 .3.3 Thermal Nuclear ReactorsAt the present stage of neutron radiographic development the nuclear reactorprovides the most intense neutron beams and therefore the highest qualityneutron radiographs. Whilst accelerators and isotopie sources are limited to aneutron flux at the detector foil of about 106 ncm-2s-1 the nuclear reactor canprovide a neutron flux of up to 108~109 ncm-2s-1 for a comparable collimatorarrangement.

The disadvantage of the nuclear reactor is its lack of mobility, high capital cost and the necessity to obtain a licence to operate. Its advantage lies in its intense neutron source strength, its low cost per neutron (about 20-25 times less than an accelerator) and its lower moderation factor (see below). Most of the reactors in use for neutron radiography are principally used for nuclear research, and their resulting high utilization justifies the capital cost. Theaverage neutron radiography facility could rarely make use of more than 20% of the neutrons available from typical nuclear reactors and so the use of reactor sources will probably be limited to organizations that can use the surplus neutrons for activation analysis, neutron physics studies, isotope production, nuclear research etc.


The nuclear reactor is an assembly in which a fissionable material, such asuranium, is dispersed in a moderating material, such as heavy water, and these are contained in a concrete radiation shield (see Fig. 1 .4). Some formof cooling is provided to remove the process heat and a number of controlelements are inserted into the assembly to regulate the nuclear reaction. Thefission process is induced by a neutron striking a uranium atom and therebycausing the nucleus to split into two roughly equal parts. These parts arecalled fission fragments and are accompanied by charged particles, gammarays, and other neutrons. These other neutrons are available to continue thereaction by striking other nuclei and so producing further fissions in a chainreaction.

One important condition must be achieved in order to maintain this state of self perpetuation: the liberated neutron must be slowed down in order to give it a high chance of causing further fission. This slowing down is achieved by making the fast neutron pass through an essentially non-absorbing moderating material before it hits another uranium atom.


Fig. 1 .4 Thermal Nuclear Reactor Source. (X,n) or ( γ,n )

THERMAL NEUTRON FLUX AT FOIL 108

- 101 n cm·2 s·1

BIOLOGICAL SHIELD

~~~~~~,1 Slow neutrons (produceel by scattering collisions)

OBJECT


or Cf 252 (spontaneous fission)

Cf 252

THERMAL NEUTRON FLUX AT FOIL

108- 101 n cm4 s.·1 //

/ BIOLOGICAL SHIELD

-~--~~~~,1 Slow neutrons (produc·ed by scattering coll isions,)

OBJECT


Neutron SourceFast neutron soutce e.g. 124Sb

Moderator e.g. 94Be + γ → 8

4Be + 10n

94Be

124Sb

http://large.stanford.edu/courses/2011/ph241/chenw2/


Neutron Source- Reactor

"J

' .. '

' ... "r \

' I~ ' ' i ... X-~

~

~- " l (I '

\

~ ~ RaGI'tlfJ~

.:,

l~::i ~-~II"

-.} ~ II

... .. .. (. ::.

!---~~~ 1rmipa#S~ r:- -I -.

~

-... ·- ::

·• .. '

"r_ N•IJ1i1u;l ' :.. I ..

~ I aa~~l'tf t~bt'&r ::.

'I ~rp j

- j ; ..: '.

\ _.1• ·- (.

' 1:": i . ' . ~

.; -Cor~

-.:! ~ . ~

c; .:: - o!l

r t /

p ' f. ... .. ;; ~

•c t I{. . _..,..... __ :.. ... (: &:'

"' ..,; .J .; ' I

,. -')I

IJ ·J 'I r_ ·-· I

~ 'J 1-. - c ._: ·-· IJ ·_.

I : I

CM!h:lat Cdi'rlllb' a-fi~~


Fig. 1 .4 Thermal Nuclear Reactor Source.the liberated neutron must be slowed down in order to give it a high chance of causing further fission.

non-absorbing moderating material to increase the σ


The Nuclear Chain Reaction

Neutron

Proton

Cesium (fission f ragment )

Urani um-236 /

initial neuton bombardment

I

Rubidium (fission fragment )

Nearby U-236 atom

!Energy

!Energy

Nearby U-236 atom

Cs

Nearby U-236 ato

Rb


Moderating materials contain light elements such as hydrogen, carbon and beryllium, and the neutron loses energy by a series of scattering collisions, in the manner of billiard balls striking each other. For efficient neutron production the number of neutrons lost during the moderation phase must be kept as low as possible, and the uranium and moderator 'mix' in a nuclear reactor is designed to achieve this.

Reactor neutrons are born at about 2 MeV and are slowed down by the moderating material to about 0.03 eV (the so-called thermal energy). This is the energy at which the neutron is in thermal equilibrium with its surrounding and when the fission process operates most effectively and it is also the energy most suitable for neutron radiography.

Accelerator and isotopie source neutrons are mostly born at higher energies, up to about 14 MeV, and so the moderation factor (neutrons lost in the energy-reduction process) for these sources is usually poorer than that for anuclear reactor.

Keywords: moderating factor


Neutron Energy (primary)Source

Reactor neutrons are born at about 2 MeV and are slowed down by the moderating material to about 0.03 eV (the so-called thermal energy).

Reactor

Accelerator and isotopie source neutrons are mostly born at higher energies, up to about 14 MeV, and so the moderation factor (neutrons lost in the energy-reduction process) for these sources is usually poorer than that for a nuclear reactor.

Radioisotopes, accelerator


Standard Californium-252 Sources: Model 10 Series

http://www.frontier-cf252.com/standard-californium-252-sources-model-10-series.html

Charlie Chong/ Fion Zhang http://www.frontier-cf252.com/standard-californium-252-sources-model-10-series.html

Cf 252 (spontaneous fission)

1.0 after Capture Atoms Required I Atom 252 Cf Produced and · Decay j

i173 1173 . 297 297 . 87 87 87 83 12.5 12.5 . 4.6 4.6 . 4.5 4.5 1.3 I I. . I l I I I I ., ,. I ' I I I . I I I I .' . I I

Fission · F1sston ·( I I

85°/o 71 °/o ~ l

,• . ·-. . Z49[f •.• .J50Cf ~251Cf

Fission Fission Fission t t · \. 4 °/o 85°/o 64 °/o 249Bk-•)50 Bk I ' . t t t t ~~r~s~~~neous 244cm~Z45cm..J46cm-.J47cm-..?48cm-249Cm . I Tlq= a5.5t0.5a

Fission . Fission t . + I 75°/o 71°/o 243Am-J44Am L _____________ j f t t Alpha Decay

zJ9pu_..240pu-..24lpu ...... z42pu ....J43pu T112 = 2.731 t0.007 a t . ..

zJau __.. 239u t . I I . I I I . I I I I

215 128 44.5 32.0 19.5 14.9 I t I. I" I . I I l I I I I I I I I · I

2156 983 689 389 30-2 10.3 5.8 1.3 0 0 Neutrons . Required I Atom 252 Cf Produced


Spontaneous fission (SF) is a form of radioactive decay that is found only in very heavy chemical elements. The nuclear binding energy of the elements reaches its maximum at an atomic mass number (A) of about 58; spontaneous breakdown into smaller nuclei and a few isolated nuclear particles becomes possible at greater atomic mass numbers.

Because of constraints in forming the daughter fission-product nuclei, spontaneous fission into known nuclides becomes theoretically possible (that is, energetically possible) for some atomic nuclei with atomic masses greater than 92 atomic mass units (amu), with the probability of spontaneous fission increasing as the atomic mass increases above this value.

https://en.wikipedia.org/wiki/Spontaneous_fission


Spontaneous fission (SF)

https://en.wikipedia.org/wiki/Spontaneous_fission

Fiss.ion prob. per Neutrons per Neutrons per gram-· Spontaneous half .Z21A NucUde Half-life

decay liission second Ufe·

235 7.04>c1 08

2J)x1 o-9 3_0x10-4 3_5x 1017 years u 1.a6 36.0 years

238 4_47x1 09

5_4x1 o-7 B.4x1015 years u 2.07 0.0136 35.6 years

239 2.411 x1 0 41

4_4x1 o-12 S_5x1015 years Pu 2.1 16 0.022 37.0 years

240 Pu 6S691 years 5J)x1 o-3 2.21 920 1_116x1011 years 36.8

250 Gm1 6~900 y~ears 0_61 3.31 1_6x1010 N/A 36.9

252 Gf 2.638 years 3J)9x1 o-2 3.73 2_3x1 012 N/A 38.1


Spontaneous fission (SF)Fission which does not require initiation by another particle is known as spontaneous fission. This is the mode by which 252Cf nucleus undergoesfission. The half life of 252Cf is 2.645 years. Being a very strong neutronemitter, it extremely radioactive and harmful.

■ 252Cf undergoes alpha decay 97% of the time to form 248Cm, and ■ undergoes spontaneous fission remaining 3% of the time.

One microgram of 252Cf produces 2.3 million neutrons per second, an average of 3.7 neutrons per spontaneous fission.

When 252Cf undergoes spontaneous session, it produces 140Xe, 108Ru, neutrons and gammas. (other various?)

Approximately, four gammas are produced per neutron during itsspontaneous fission. These neutrons produced by 252Cf have diferentenergies. The energy distribution of these fssion neutrons, as shown in Fig. 2,is known as the spontaneous fission spectrum.

http://www.physics.byu.edu/docs/thesis/328


Figure 2: Spontaneous Fission Spectrum of Neutrons from 252Cf

http://www.physics.byu.edu/docs/thesis/328

Charlie Chong/ Fion Zhang https://en.wikipedia.org/wiki/Isotopes_of_californium#Californium-252

nuclliide Z(p) N(n) isotopic mass (u) de,cay daughter nucl,ear half-life

mode(s)lsnn 11 symbol isotope(s·)' spin

excitation energy

249mc f . II 144 .. g,a (5) keV 45(5) ~s 512+

a ('99. 92°/01) 246c , m 2soCf 98 15,2 250.0764061 (.22} 13 ... 08(9) a 0+

SF (.077'0k) (variious)

251c -fn2J 98 15~3 251 .079687(5) 900(40) a a 247cm 1112+

a ('96. 9°/o) 248cm 252c -fn 11 98 154 252.08162,6(5) 2.1645(8) a

SF (3,.09°k )lin 41 ~I 0+

I~ (99.69%1) L::;l.:sEs 253Cf 98 15~5 253.085<13·3(7) 17 ... 81 (B) d (7/2.+)

a (.311'0/o) 249c , m


This fission process produces a considerable amount of heat and this is usually removed by a stream of coolant (water or gas) flowing in speciallyconstructed cooling passages in the reactor core. This coolant then losesheat in a conventional heat exchanger.

Control of a nuclear reactor is achieved by simply removing neutrons from theprocess and thereby stopping the chain reaction. The neutrons are removed by inserting a neutron absorbing material into the reactor core and regulating the extent of the insertion in order to maintain the reactor at a steady operating power.

Thus the nuclear reactor is a device that produces fast and slow neutrons,gamma rays and charged particles, all in prolific quantities and at present it isthe most widely used neutron source for neutron radiography.


1.1.3.4 Sub-Critical AssembliesThe output of a non-reactor neutron source can be boosted by incorporating itinto a sub-critical assembly. This is a small thermal neutron reactor which hasbeen-'under-designed' so that it is not capable of sustaining a chain reaction.

The neutron source is placed in the reactor core and provides the supply ofneutrons to keep the reaction going. The fission processes in the fuel cannow be regarded as a multiplication phenomena in that the neutron output ofthe source is enhanced by the reactor neutrons. Crosby et al [Ref. 1 7] haveinvestigated such a system for neutron radiography using a Ca 5 source witha water moderator. This assembly had a multiplication factor of about 30. Thefactor is dependent upon the value of the effective multiplication constant(Keff) for the system. This constant may be regarded as the ratio of thenumber of neutrons in one generation, compared to that of the nextgeneration. So if Keff = 1 then the neutron population is sustained from onegeneration to the next. This condition is called critical and is that whichapplies to the type of thermal neutron reactor described in the previoussection. The sub-critical assembly, however, has a Ketf of less" than one andthe neutron density will depend upon the neutron emission from the neutronsource.


Keywords: effective multiplication constant (Keff) for the system. This constant may be

regarded as the ratio of the number of neutrons in one generation, compared to that of the next generation.

The sub-critical assembly, however, has a Keff of less than one and the neutron density will depend upon the neutron emission from the neutronsource.


Figure 1 .5 shows the relationship between the multiplication factor and Keff and it can be seen that Keff must be greater than about 0,9 before a useful multiplication is obtained. In practice the choice of Keff is made fromconsiderations of safety and cost. A balance must be found whereby thesystem produces a useful multiplication but Keff is sufficiently far from unity toensure that the reactor will not go critical. If criticality is possible then stand-bycontrol absorbers must be added to the assembly. Clearly if this situationwere possible then it would be more practical to build a critical nuclear reactorin the first place.

So safety requirements will limit the Keff of a sub-critical assembly to about 0,99 although this will be dependent upon the safety philosophies followed bythe local reactor licensing authority and a corresponding multiplication factorof about 30 (?) . This would be a useful increase, but its cost effectiveness must be examined by comparing the cost of the assembly with other types ofneutron sources of the same intensity.


Fig. 1.5 Increase in Neutron Flux as Subcriticai Size is increased. (After Bouchey, Int. J. Non-Destr. Test. 2, 1971, p. 350).


1.1.3.5 A Comparison of Neutron SourcesTable 1 .3 shows a summary of the properties of some typical facilities for neutron radiography by Hawkesworth and thus provides a convenient format for comparison of the sources, in that it provides a practical criteria for thecomparison in the times required to produce a radiograph. It can be seen thatthe reactor sources are approximately an order of magnitude better thanaccelerators and 252Cf sources, on this basis. Moreover, other criteria such as cost and mobility may take precedence.


Table 1.3 Summary of the Properties of some typical Facilities for Neutron Radiography [Ref. 54].

a) The transfer technique has been used as the main example here because of its value to the nuclear industry. The neutron exposure times assume a film exposure time> 5 x T½

b) Films from the Agfa-Gevaert range. Other manufacturers offer a closely parallel series of films. c) See Section on Collimation. d) A small reactor designed to make neutron radiography as convenient and economic as possible.


1.1.4. Neutron BeamSo far we have discussed various ways of producing neutrons withoutconsidering whether these neutrons will be suitable for neutron radiography. Now, the radiographic process must use a radiation which has a highprobability of reacting with the material of the sample, and it is usual todescribe this probability of interaction as an effective target size called the'cross section'.

1 .1 .4.1 Nuclear Cross Sections The cross section quantifies the probability that a reaction will take place between the neutron (travelling at some effective velocity) and the target material; it can be considered as a target size and it is measured in cm2. There are several types of cross section but the two that are of principal interest to neutron radiographers are:

(1) the cross section σabs and (2) the scattering cross section σscat.

The total cross section σtotal is the sum of these two.


The unit of cross section is the 'barn', which is 10-24 cm2, and typical crosssections vary from a few millibarns to several thousand barns.

The cross section of the elements and their isotopes vary with the energy of the bombarding neutrons (see Fig. 1.6) in general the lower the energy thehigher the cross section.

This fact provides the opportunity to increase the transmission of the neutron through the sample by using a neutron energy that coincides with a region of low cross section in the sample (see Neutron Beam Filters). The transmission of a neutron through a sample may be expressed by considering the rate at which the neutron intensity reduces as it passes through the sample material.


Fig. 1.6 Total Cross-Section Curves for Boron, Cadmium, Indium and Dysprosium

t Vl1o3 z a: ~

10

O ABSO&PT!ON CROSS-SECTION •Rt..TI SCATTERING

1~--------~ .. o~-2--------i,o~-~-----L--~------ss 10 -'3 E NERG'f I ~v I -

....... UJ

'"G X

0 0 ~

Charlie C

hong/ Fion Zhang

Fig. 1.6 Total Cross-Section Curves for Boron, Cadmium, Indium and Dysprosium

104

10

• Rt.TIO ABSOf.sPT!ON CROSS-SECTION ~ SCATTERING

1~--------~~-------L~--~--~------~ 10 -'3 10-2 10-1 1 5

ENERGY (I?V l -


Cadmium Ration- ratio of the activity induced by the neutron beam in a bare gold foi l to that induced when the foil is covered with cadmium.

Cadmium ratio- the ratio of the response of two identical neutron detectors, usually activation types such as indium or gold, one exposed bare to thebeam and the other cadmium covered (the cadmium covered detectorrecords primarily neutrons having an energy above 0.5 eV and the ratio is ameasure of thermalization in the neutron spectrum).

(Cadmium act as high pass filter (E>0.5eV))


One isotope of cadmium, 113Cd, absorbs neutrons with very high probability if they have an energy below the cadmium cut-off and transmits them otherwise. The cadmium cut-off is about 0.5 eV. Neutrons with energy below the cut-off are deemed slow neutrons (thermal neutron 0.03~0.1 eV) , distinguishing them from intermediate (epithermal/ resonance) and fast neutrons.

Cadmium is created via the long s-process in low-medium mass stars with masses of 0.6 to 10 solar masses, which lasts thousands of years. It requires a silver atom to capture a neutron and then undergo beta decay.

http://schools-wikipedia.org/wp/c/Cadmium.htm


The cadmium cut-off is about 0.5 eV. Neutrons with energy below the cut-off are deemed slow neutrons (thermal neutron 0.03~0.1 eV) , distinguishing them from intermediate (epithermal/ resonance) and fast neutrons.

http://schools-wikipedia.org/wp/c/Cadmium.htm

0.5 eV


This is given by

- dФ /dx = ФσN (dФ/Ф = -σNdx) (1)

Where: Ф = neutron intensity, i.e., number of particles passing across unit area in

unit time, n cm-2 s-1

x = specimen thickness, cm σ = microscopic cross section, cm2

N = number of target nuclei per unit volume, n∙cm-3

(see Appendix 1.5 for a calculation of N)

Rearranging and integrating gives:

Ф = Фoe -Nσx (2)

Where:Ф = neutrons transmitted through the sample, n∙cm-2s-1

Фo= neutrons incident upon the sample, n∙cm-2 s-1


Neutron attenuation ProbabilityNeutron attenuation is a statistical process which depends upon the

interaction of the neutron with the nuclei, but which a sample is completely unpredictable. The attenuation can be predicted by assuming thatindividual neutron interactions with nuclei are purely random events. If there are N nuclei traversing distance x, then the number ΔФ which would be attenuated in any given Δx would be proportional to Ф :

ΔФ = NσФΔx, ΔФ/Ф = -NσΔx∫ΔФ/Ф = -∫NσΔxln Ф + C = - Nσxln Ф = - Nσx - CФ = e-C e –Nσx

when x =0, Ф = Ф0 = eC e –Nσx = e-C

Ф = e-C e–Nσx = Ф0e –Nσx

http://hyperphysics.phy-astr.gsu.edu/hbase/Nuclear/halfli2.html#c3


The ratio between these two neutron fluxes is called the transmission, i.e. transmission

(3a)

It should be noted that there are several ways of expressing the cross section of a material, i.e.:

a) microscopic cross section, cm2, σ (probability of interaction for each nuclei)b) macroscopic cross section, cm-1 (total target areas = Nσ, in each cm3)c) mass absorption coefficient, cm2 g-1

The first is the basic unit and, as stated earlier, is measured in barns 10-24 cm2.

The second is the product Nσ; this is given the symbol Σ and it is the total target area for a given neutron interaction presented by a cubic centimeter of material. (linear attenuation coefficient μn) Nσ= μn = ρN’σ/A, where N’ is Avogadro's number (6.023 X 1023 atoms/gram-molecular weight) ; σ is the total cross section in barns cm2 ) ; and A is the gram atomic weight of material.


Thus, for this case equation (3) can be rewritten:

(4)

and it can be seen that the use of the macroscopic cross section Σ simplifies the use of this equation. The third form, the mass absorption coefficient, is denoted by the symbol

where ρ = density


Neutron Attenuation

)

Incident lntensity--~)•r """' ........ ~ •• 1..,__~ Transmitted Intensity )

It = 10 exp ( -:l:t)

)

)

. spec1men


Thus the mass absorption coefficient is the total target area for a given neutron interaction per cubic centimeter of material per unit density. Ittherefore more conveniently expresses values for solids and gasses whosedensities normally differ by several orders of magnitude. In this case equation(3) is expressed as

(3b)

where the unit xp has the dimension of g cm-2

Values of Σ and μm are given in Appendix 1.2 and an example of the calculation of Σ for a compound is given in Appendix 1.5. It should be noted that there is also a cross section called the 'linear absorption coefficient' with the dimension cm-1. This is the same as the macroscopic cross section but is more usually used in describing alpha, beta and gamma


PART 5NEUTRON CROSS SECTIONS AND ATTENUATION


5.1 Neutron cross sectionsNeutron cross sections are defined in Part 1 of this Section. Values for thermal neutrons for many materials (elements) are given in Table 9 (seeBibliography item 8 for a more extensive compilation). Generally, neutron cross sections decrease with increasing neutron energy; exceptions includeresonances, as mentioned earlier. Cross section values can be used to calculate the attenuation coefficients and the neutron transmission as shownin eqs. 1 and 2. For compound inspection materials, the method for calculating the linear attenuation coeffici ent is shown following Table 9.

If the material under inspection contains only one element, then the linear attenuation coefficient is:

μ = ρ∙Nσ/ A Eq.7

Where:μ is the linear attenuation coefficient (cm-1 ) ;ρ is the material density (g/cm3); N is Avogadro's number (6.023 X 1023 atoms/gram-molecular weight) ; σ is the total cross section in barns (cm2 ) ; and A is the gram atomic weight of material.


For photons:

I = Ioe –μx t Eq.1For Neutron

I = Ioe –Nσt = Ioe –μn t Eq.2

Where: I is the transmitted beam; Io is the incident beam; μx is the linear attenuation coefficient for photons; t is the thickness of specimen in the beam path;

N is the number of atoms per cubic centimeter; σ is the neutron cross section of the particular material or isotope

(a probability or effective area); and, μn is the linear attenuation coefficient for neutrons (μn = Nσ).


TABLE 9. Thermal Neutron Linear Attenuation Coefficients Using Average Scattering and 2200 m/s Absorption Cross Sections for the Naturally Occurring Elements

Element Atomic No. Symbol

I H2 2 He 3 Li 4 Be s B 6 c 7 N2· 8 02 9 F2

10 Ne ll ll 13

Cross Sectfon fbarnsJ

S tterfng Absorption*

38.0 0..8 (..4 7.1 4.4 4.8 tO 4.2 3.9 2.4 4.0 3..6 1.4 17

0.332

71.0 0.010 755· O.OOJ 1.88 0 0.01 2.8 0.536 0.063 0.23 0.16 0.20 O.S2 33.6

Linear Atten·· tJon Co Jd nt fcm- 1 J

gas g 3.36 0.8.8 99 0.541 g gas gas gas O.llS 0.158 0.0984 0.0965 0.184 0.0.591 a as


If on the other hand, the material under inspection contains several elements,or is in the form of a compound, then the linear absorption coefficient is:

μ = ρ∙N/M (ѵ1σ1 + ѵ2σ2 + ѵ3σ3 +..... ѵiσi ) Eq. 8

Where:μ - is the linear attenuation coefficients of the compound (cm-1) ; ρ is the compound density (g/cm3 ) ; N is Avogradro's number (6.023 X 1023 atoms/gram-molecular weight) ; M is the gram molecular weight of the compound; ѵi is the number of absorbing atoms of ѵi kind per compound molecule; and, σ; is the total cross section of the ith atom (cm2).


As an example, consider the calculations of the linear attenuation coefficient,p.., for the compound polyethylene (CH2)N :

μ = ρ∙N/M (ѵ1σ1 + ѵ2σ2 + ѵ3σ3 +..... ѵiσi ) Eq. 8

μ = ρ∙N/M (ѵCσC + ѵHσH)

for: ρ = 0.91 g/cm3

N = 6.023 X 1023 atoms/g-molM= 14.0268 gѵC = 1σC = 4.803 X 10-24 cm2

ѵH= 2σH = 38.332 X 10-24 cm2

μ = 0.91 x 6.023 x 1023 x (14.0268)-1 (1x4.803+2x38.332) x 10-24

μ = 3.18329 cm-1


σC

σH


FIGURE 10. Half-Value Layers of Selected Materials for a Thermal Neutron Radiograph Beam.

p. (CM-1)

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5.2 Half-Value LayersAn important concept for radiography is the half value layer (HVL); that is, thethickness of material that will reduce the radiation intensity by a factor of two.A plot of half-value layers for a practical thermal neutron radiographic beam isgiven in Fig. 10. This information can be used to estimate the transmissionand detectability of various materials combined with others.

I = Ioe –μn t

at half value layer

I/Io = ½ = e –μn t½

Ln 0.5 = –μn t½

t ½ = Ln 0.5/ - μn = 0.693/ μn


5.3 Tenth-Value LayersThere will be a thickness of material that is sufficiently thick that little of theneutron beam penetrates. In Fig. 11, thicknesses of material that will transmitonly 10% of an incident thermal neutron radiographic beam are plotted. Thisthickness represents about the limit that should be attempted in normalthermal neutron radiography. Variations in neutron energy should beconsidered for thicknesses greater than those shown in Fig. 11.

Similarly:

t 1/10 = ln 0.1/ - μn = 2.302/ μn


5.4 CONCLUSIONNeutron radiography is a valuable method for nondestructive testing. Theattenuation differences between X-rays and neutrons make these tworadiographic methods, to a large degree, complementary. Figure 6 in thisSection is an illustration of how the two methods provide a more completeinspection when used together. The neutrons in this example show lightmaterials such as the explosive, plastic and epoxy components, while the X-radiograph shows the metallic components.

Neutrons offer sensitivity to different isotopes and can also be very useful for inspecting highly radioactive material. These two characteristics offeradvantages particularly to the nuclear industry. Other areas of application include aerospace, the military and transportation industries. The neutron radiographic technique is relatively expensive, but it can be used to perform inspections that present problems for other NOT methods. When used for these unique inspections, neutron radiography is a cost-effective nondestructive testing echnique.


1 .1 .4.2 ModerationThe neutrons from all available sources are born with high energies, up to about 14 MeV, whereas neutron radiography requires neutrons havingenergies of about 0.03 - 10 eV. Thus they must be slowed down to thethermal/epithermal neutron range, and so here we have the same problemthat exists in the nuclear reactor, that of moderation, and of course the samesolution can be used. A moderating material, usually water or beryllium, isplaced around the radioisotope neutron source or the target of the accelerator,to produce a neutron energy spectrum similar to that of the nuclear reactor.The inherent advantage of the nuclear reactor now becomes clear; it alreadyhas a predominantly low-energy spectrum, and furthermore reactor neutronsare born at a lower energy (approx. 2 MeV as compared to about 14 MeV foran accelerator) and so fewer neutrons are lost in the moderation process.

Keypoints:Reactor neutron source: 2 MeVAccelerator, isotopes sources: 14 MeV


A prime requirement for a moderator is that it shall slow down neutrons without absorbing them, and so good moderating materials will have a largescattering cross-section and a small absorption cross-section. Each time aneutron interacts with the moderator there is a probability that it may beabsorbed and so the fewer the number of collisions required to produce athermal neutron the better for moderation.

Once it has reached thermal energy the neutron will continue to bounce around until it is finally captured.

In order to reduce the energy of a neutron in the fewest number of collisionsthe amount of energy lost per collision must be as large as possible. It can beshown that the log of the average energy loss is, to a close approximation,inversely proportional to the atomic number, and so for the best 'energy loss‘ conditions light materials should be used. Hence moderators are thosematerials that have low atomic number, low absorption cross-section, and ahigh scattering cross-section and typical moderator materials are light water,heavy water, and carbon.


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1.1.4.3 CollimationHaving produced a source of low energy neutrons we now have to form theminto a usable beam. The neutrons will move about the moderator in a completely randow manner, and unlike electrons cannot be focused. The bestthat can be achieved is to contain the neutron source and the moderatorwithin a neutron absorbing shield and allow some of the neutrons to streamdown a hole in this shield.

To achieve this containment the walls of the collimator are lined with a neutron opaque material which will prevent stray neutrons entering the system via the collimator walls and will also reduce low angle scattering within the system.

This 'lining material must have a high cross section to neutrons, and thesecondary radiation produced by neutron absorption with this lining must have a low probability of being recorded.


These requirements lead to the use of boron, cadmium* (The use of cadmium for the direct method is limited by the high energy gamma emitted after neutron capture.), dysprosium, europium, gadolinium and indium as lining materials.

The angular spread of the emerging beam will be confined by the length-todiameter ratio (L/D ratio) of the collimator hole, and so to ensure a narrow beam-spread a collimator usually has a high L/D ratio. Near parallel neutron beams are achieved by using a bundle of small tubes or a stack of equispaced plates within the collimator hole. With these arrangements the L/D ratio of the collimator is that of an individual tube or the gap between the plates. This ratio can be made very large, but this type of collimator suffers a considerable loss in beam intensity and produces a pattern of circles or lines on the radiographic image.


An alternative method, widely used for neutron radiography, is the divergentcollimator. This has a relatively small inlet aperture, and the collimator holediverges uniformally along its length. With this arrangement the angularspread of the neutron beam reaching the object only depends on the sourcesize and the distance. These are shown in Figs. 1.7A and 1.7B **)

** In the diagrams in Fig. 1.7 the lengths L and Ls plus Lf are the radiographic lengths of the collimator, and these are assumed to be approximately equal tothe physical length of the collimator.


Fig. 1.7 Neutron-Beam and Collimator Geometry.

~~~-1.. L

A. MULTI-TUB!! COLLIMATOR

of p:l ==========: -I• L

8. DIVERGENT COLLIMATOR

~tiru;~~·~·:::L,~·~~ ~~L:'::·J~ =!ug = llf D INLET DL I ' APERTURE • JOBJECij-IMAGE (at

5

FOIL) DISTORTION PENUMBRA --

C. GEOMETRIC UNSHARPNESS

Lt«L~

D. GEOMETRIC ENLARGEMENT AND DIMINUTION

Charlie C

hong/ Fion Zhang

Fig. 1.7 Neutron-Beam and Collimator Geometry.

A. MULn-'ruBI! COLLIMATOR

APERIURE OISIORilON PENUMBRA

C. GEOMETRIC UNSHARPNESS

ls lt

hi tat IU

Lt«l.v

GECIME1fRIC ENLARGEMENT AND DfMINUTIOI\I


An estimate of the flux emerging from a collimator tube [Ref. 1 1] is given by:

(5)

Where:Фi = neutron intensity at entrance to the collimator, n∙cm-2 s-1

Ф = neutron emission at the exit from' the collimator, n∙s-1

A = collimator area at exit, cm2

D = diameter of inlet aperture, cmL = length of collimator, cmΣ = macroscopic total cross-section of moderator, cm-1

dФ/dz = flux gradient at inner face of collimator


As the flux gradient is usually small this can be further approximated by

(6)

and so

(7)

where Фo = neutron flux at outer face of collimator, n∙cm2s-1


ФiФ

intensity

fluxL

D


Now the fraction of neutrons lost due to collimation will be the ratio:

The collimator length to inlet diameter ratio (L/D) is called the collimator ratio.Expression (8) assumes that all of the neutrons in the collimator originate at the entrance aperture, but in practice some will come from the walls of thecollimator [Refs. 1 2,22] adjacent to the aperture. Hawkesworth [Ref. 1 2] hasshown that the total flux is given by

(8)

(9)


where I = length of collimator wall which emits neutrons. The lenght I varieswith different types of neutron radiography unit in that it is usually a section ofthe collimator that is not lined. A lining close to the source will depress theflux and so with low intensity sources it is necessary to leave an unlinedsection at the beginning of the collimator. High flux reactor sources are not soconcerned with this flux depression and in most designs the collimator is linedalong its full length.

For sources of low intensity the unlined length is usually about two diameters (D) long and so equation (9) can be modified to

(10)


for simplicity this can be expressed in the same form as equation (8) by a close approximation, i.e.

where the value of Fc is close to 12 (with short section (2D) of unlined absorber along the collimator).

(11)

12 (11a)


Keypoints: The 2 equations to remember

(L was fully lined with neutron absorbing material)

(2D of L was not lined with neutron absorbing material)

12

c/J = c/Ji D 2

16 L


Hawkesworth found reasonable agreement with equation (9) for small L/D ratios (<30) on accelerator units. Matfield examined published data relating toa large number of neutron radiography units of all types and found that theconstant F varied from about 1 to 100 although this may well be due touncertainties in the value of the source flux, as this is often difficult tomeasure accurately, and also the uncertainties in the assumption that thecollimator aperture is the true neutron aperture. The length to diameter (L/D)ratio for a collimator effects both the resolution and the collimator efficiencyand so it is widely used as a simple means of characterising a collimator. Theresolution of the collimator can be described by considering the effect of thewidth of the radiographic dimensions of the collimator on the unsharpness ofthe image,(as shown at Fig. 1 .7c: where the dimension Lf is shown grosslyoversize in order to ensure clarity of the diagram (also see footnote to page 15).


This unsharpness is expressed by

where Ug = geometric image unsharpness, cmD = source-aperture size, cmLf = image to object distance, cmLs = source to object distance. cm.Usually Lt < < Ls and so the geometric unsharpness is linearly dependent upon the inverse of the collimator ratio.

(12)

Ug = Ft/D ≡ D∙Lf / Ls


To minimise unsharpness the aperture size D and the distance Lf (Lfilm) must be small and Ls ( Lsource ) must be large.

These requirements lead to a direct conflict with the desire to achievemaximum intensity, for this requires D to be large and Ls to be small. So we must compromise, and the practical design is based on a judgement of the unsharpness that can be tolerated, in conjunction with a workable neutron flux. It is clear from the above discussion that the length to diameter L/D ratio of a collimator is an important characteristic, but we can achieve a particular L/D ratio by using an infinite variation of length or diameter, so is it better to have a long and wide collimator or a short and narrow one?

The effect of collimator size has been measured by Barton [Ref. 1 3] and others and it is clear that the answer cannot be given without a number of qualifications. for we are concerned with the interdependence of (1) resolution, (2) contrast, (3) intensity, (4) neutron/gamma ratio and the (5) attenuation produced by the specimen.


Nevertheless the following general statements can be made:the contrast will vary with the intensity of the gamma rays in the beam for any particular sample the neutron attenuation will bed ifferent to the gamma attenuation.

(a) Due to the transfer method (see 1.1.6.2) usually produces better contrast than the direct method,

(b) the presence of gamma rays in the beam will reduce the contrast for the direct method unless the gamma attenuation of the sample is high, say > 80%.

(c) The neutron flux increases more rapidly than the gamma flux as the aperture size increases,

(d) the neutron flux and the gamma flux decrease at about the same rate as the collimator length increases,

(e) the slow neutron intensity decreases more rapidly than the fast neutron intensity as the length of the collimator increases. Due to (c) and (d) long and wide collimators have better neutron to gamma ratios and hence better contrast for the direct method. But due to


(f) the resolution improves with an increasing L/D ratio. To fully achieve f) there must be no Joss of contrast and so if the available collimator is short and narrow with a poor neutron to gamma ratio the best resolution may be achieved by the direct method. Similarly the resolution may be effected when a sample with a large scattering cross section is radiographed in a beam with a high fast neutron component.

(g) the larger the inlet aperture the greater is the displacement of themoderator.

(h) the longer the collimator the greater is the attenuation loss due to thecollimator atmosphere.


Statements (g) and ( h) are mainly relevant to low intensity sources and for this type of source (g) often outweighs all others for it has a direct effect onthe neutron intensity at the entrance to the collimator. Furthermore the spatialneutron flux in a small source moderator can vary across the collimatoraperture and thus for this type of source the collimator is usually made shortand narrow.

For a collimator filled with air the loss is about 5% per metre, and where helium is used this is reduced to < 1% per metre. Hence the neutronattenuation due to the collimator atmosphere is only significant for air andonce again leads to the use of short coll imators for low intensity sources.

When we collect these statements together it is clear that the originalquestion on the shape of the collimator must include some information on thetype of sample to be radiographed, the acceptable results, etc. In practice theconstruction of the neutron source usually sets the overall limits to the widthand length of the collimator and where possible the best solution is to makethe aperture as wide as possible and stop it down with a range of insertableaperture plates.

Charlie Chong/ Fion Zhang Nares Chankow Department of Nuclear Engineering, Faculty of Engineering

Neutron generator

Water moderator

Neutron collimator

' I

I I

(

I

• r

I I


Neutrons Collimator

Nares Chankow Department of Nuclear Engineering, Faculty of Engineering

Water moderator


1 .1 .5 Neutrons Applied to RadiographySo far we have followed the progress of a neutron, born with high energy,progressively losing this energy by successive collisions within a moderatingmaterial, until it finally escapes a long a collimator tube. At the outer end ofthis tube there is a sample to be radiographed, and this is duly struck by theneutron. What happens now is well described by a comparison with X-rays.This comparison is shown in Figure 1 .8 where the mass attenuation coefficients of Xrays and thermal neutrons have been plotted against the atomicnumbers of most of the elements. X-rays show a continuous curve and so anytwo materials having a similar atomic number wil llie close to each other onthe curve and consequently have similar mass absorption coefficients. Bothmaterials will, therefore, attenuate an X-ray beam by about the same amountand so it will be difficult for a detector to discriminate one from the other.


The attenuation of neutrons however, is a function of the nucleus rather than the density of electrons in a material, and it is frequently found that adjacent-number elements, for example boron and carbon, show marked differences inneutron-attenuation coefficient and are therefore readily discriminated.

Hydrogen has a high neutron attenuation coefficient, and so it is possible todetect rubbers and plastics. Conversely dense materials such as lead andtungsten have low coefficients and a re readily penetrated by thermalneutrons. Thus we find that the two radiographic processes, X-ray andneutron, are often complementary. X-rays are stopped by dense materialsand pass through light ones, and in many instances neutrons have thereverse qualities.


Neutrons will penetrate the body of a large metal valve to give a good image of an internal rubber seal. For the X- rays to record this seal would require a long exposure which would probably obliterate most of the other detail on the radiograph. However, if the valve were inside a thick polythene case then the X- rays would penetrate this with negligible attenuation, whereas the neutrons would have difficulty in producing a radiograph.



Neutrons radiograph.


eu tro tl (Gd_ / fih11) r eutt·o1r1 (n11agii1g plate)




I eutt·on(Gd/ filn1) Neutron (itnaging plate)



Neutron (it11agii1g plate) X-RaJ


Charlie C

hong/ Fion Zhang


Neutrons radiograph - Left: X-ray does not show content. Right: neutrons can provide information about the embedded organic material.

http://www.psi.ch/niag/cultural-heritage


Neutrons radiograph

http://einrichtungen.ph.tum.de/E21/e21_boeni.site/antares/web_new/first_neutrons/first_neutrons.html


Fig. 1 .8 Neutron and X-Ray Mass Attenuation Coefficients for the Elements.

?

10 I

100

10 30 40

~ Cd

*

• PREDOMINANTLY SCA TIER (oA/Os > 10 . THERMAL e SCATTER AND ABSOAPTION }

• . · PREDOMINANT.L Y ABSORPTION (o A/og < 10 . NEUTRONS if ·ABSORPTION ONLY . iC COLD NEUTRONS 0,003 t!'V

~ -~------------ .~~,.PU

70 80 90 100 ATOM1IC NUMBER •

IS - --·X-RAYS ~t25 KV}1


100 -¥od

8 AGd

100 .. AB ~

I ~u Ill Cd N

E 10 i(·

u -I-z w

0· 0 1 o~----~1o~----~2o~----~~----~,~o----~~-----L----~7o~----~so~----~~----1~00 ATOMIC NUMBER ----

---X-RAYS 1125 KV) e SCATTER AND ABSORPTION } • PREDOMINANTLY SCATTER (oA/os> 10 THERMAL A PREDOMINANTLY ABSORPT ION (o A/OS< 10 NEUT RONS

if. ·ABSORPT ION ONLY • COLD NEUTRONS 0,003 eV

IS


FIG. X1.1 Approximate Mass Attenuation Coefficients as a Function of Atomic Number

ASTM E748

1 000~E~~~--~--~----~--~--------~------~---------f-~ ~ .... -- Gd •

Sm

• Cd e • Eu

10t~~--r---~--~~--~---+~--+---~--~~--~~~----~ r::eu ~ ' ..... -

0.001 0 10 20 30 40 50 60

• N ElJTRONS (A = 1.08 A) ..... -- X-RAYS (/.;;;; 0.098 A)

70 ao so 100


Attenuation Coefficients for Neutrons and X-rays

http://mnrc.ucdavis.edu/neutronimaging.html

- X-rays (lOOkeV) Gd

+ Thermal neutrons

• •

Cd •

•

•

•

• • •

•

0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80

Atomic Number


Mass attenuation coefficients for thermal neutrons (0.03~0.1MeV) and gamma-rays as a function of atomic number of elements (reproduced from [3] with some modifications)


1000

- Gd

~ • N 100 E \JI B - • ~ H Sm z • Cd • • Eu L&.l • y-rays (100 keV) u 10 . u u.. H20 u..

Pu L&.l • 0 l r • u Hg Ac • • • . z 1 0 r • Pa • • TmHf' I~ Nd . Ho• •, Re Au "{-rays [300 keV)

••Pm l ::::> • u .z Yb

t: 0.1 "{-ray.s 1 MeV I I<

TI ••Pb •• Ra 1h• u • • Ce Bi

0.01

0 10 20 30 40 50 60 70 80 90 100

AlOMIC NUMBER


1 . 1 .6 Neutron Image DetectorsUnlike X-rays, neutrons have very l ittle direct effect on photographic film, thequantity of -silver in the grains of a photographic emulsion being insufficient toproduce much neutron absorption. The thermal neutron half-life is relatively long, and it would take many hours to obtain a usable image. Thus for neutron radiography it is necessary to use a slightly different technique. The method normally employed uses an intermediate foil which converts theneutron image into (1) alpha, (2) beta, or (3) gamma radiation, and it is this secondary emission which is detected by a photographic film.

Keypoints:Secondary emission: alpha α, gamma γ, beta β


1.1.6.1 Direct TechniqueWith this method an atom in the foil a bsorbs a neutron and it promptly emitsother actinic 光化性 radiation (?). This is called the direct technique (see Figure 1 .9), because the foil is placed directly into the neutron beam in contact with the photographic film. When a metal foil, such as gadolinium, is used the induced radiation is an electron.

Alternatively a scintillator screen containing a mixture of lithium-6 and zinc sulphide can be used. On absorbing a neutron a lithium atom emits an alpha particle (& tritium) and this (either of these) then strikes the zinc sulphide,which in turn emits a l ight photon. (0

1n + 36Li ---> 13H + 24He + 4.78MeV) As the above processes are continuous reactions this type of foil and scintillator screen can be used with low neutron fluxes and long integrating exposures, and because the film is in contact with the converter during the neutron exposure, all of the forward emitted radiation takes part in the exposure of the film. Thus the direct technique is fast, the scintillator screen type of converter being 30~100 times faster than metal foils.

actinism: the property of radiation by which chemical effects are produced.


The particular a dvantage of the gadolinium metal foil is that the neutrons are absorbed in a very thin layer of the foil and the emitted electrons have a short range, and so a resolution of a bout 12 μm can be achieved. For nuclear applications the direct technique has a disadvantage. If the object isradioactive it will invariably emit gamma rays and these will mingle with theneutron beam and produce a second, auto-radiographic image on the film. Asthese gamma rays come from a different source from that of the neutrons theimages will be different and the film will be 'gamma-fogged'. Fortunately this can be avoided by using the (1) transfer technique or (2) track-etch recorders (see 1 .1 .7.2)


Lithium-6 and zinc sulfide In this application, the silver activated zinc sulfide phosphor powder is intimately mixed with a Lithium-6 isotope enriched lithium fluoride powder, employing predetermined ratios for detection optimization, in an appropriate binder matrix to form a thin layer on a substrate. Neutrons incident on the screen impact the Li-6 isotope and the ensuing transmutation generates an alpha particle (helium atom) and a triton particle according to the following equation:

01n + 36Li ---> 13H + 24He + 4.78MeV

Subsequent collisions with either of these charged particles cause the silver activated zinc sulfide to scintillate and emit broad band emission centered on the blue region of the visible spectrum at 450nm.

http://loradchemical.com/news/zinc-sulfide-phosphors-for-neutron-detection.html


Figure 2: (a) Silver Activated zinc sulfide powder, (b) Silver activated zinc sulfide powder viewed under UV illumination

http://loradchemical.com/news/zinc-sulfide-phosphors-for-neutron-detection.html


Lithium-6 reactions

http://www.sandia.gov/pcnsc/departments/iba/ibsphys/sigmabase/data/li6.html

L 6Li( d,o:0/ He 6Li(d,a0)4He

0 9=150 • B.Maurel, G.Amsel and D.Dieumegard, NIM, 191(1981), 349

BJioArAI 2· 6Li(p .)He)4He

6Li(pJHe)4He 0 9=60

• JB.Marion, G. Weber and F.S:Mozer, Ph.vs.Rev., I 04(1 956), 1402

BJioArAI 3· 6Li(p o:/He

6Li(p,a)3He 0 9=60

• JB.Marion, G.WeberandF.S.Mozer, Phys.Rev., 104(1956), 1402

BJioArAI

4- 6Li(3He p0) 8Be 6Li(3He,p0)8Be

0 9=165 • JP.Schiffer, T. W Bonner, R.HDavi.s and F.JV.Prosser, Phys.Rev., 104(1956), 10

BJioArAI 5- 6LitHe p1/ Be

6Li(3He,p1)8Be 0 9=165

• J P.Schiffer, T. W Bonner, R.HDavi.s andF.JV.Prosser, Phys.Rev., 104(1956),10

BJioArAI


1.1 .6.2 Transfer TechniqueThis method (see Fig. 1 .9) relies on the build-up of radioactivity in the foilproduced by neutron absorption. In this way an activation image is formed in the foil, and this is subsequently transferred to a photographic film in a dark-room by placing the foil and film in contact and allowing the decay radiationfrom the foil to produce the latent image in the film. With this method thedecay process starts during neutron exposure in the beam and so some ofthe emitted radiation is lost. This makes the transfer technique slower thanthe direct method but this disadvantage is more than compensated for by thefact that since the foil is insensitive to gamma rays the method can be used ingamma-ray fields of any intensity.


Fig. 1 .9 Direct and Transfer Method for producing a Neutron Image.

w 1- 2~ (J ..... w w ~ :!tn ~

=~ Ill 0 0 lL ILLU

\rr~ l I

NEUTRON J .. I BEAM I

l I

I I t:~

t; w .., m 0

NEUTRON .. BE~M

~

0 u.

' 1. =r-, I.

TRANSFER TO I .._._~- _.~ ..... J

DARKROOM I

I I r

B.. TRANSFER METHOD


With the transfer method only the foil is present in the beam, and the activitybuilds up exponentially according to

where sS = activity, disintegrations s-1

Ф = neutron flux, ncm-2s-1

σ = microscopic neutron cross-section, cm2

N = number of atoms in the sampleλ = 0.69/ττ = half life, s (T½)T = irradiation time, s

- 0.69T/T½


It can be seen from expression (13) that the activity will gradually reach amaximum, depending upon the neutron flux and the cross-section. The foil activity wil l eventually saturate, and since neutron fluxes below about104 n∙cm-2s-1 will not provide sufficient intensity to store enough energy in thefoil to produce an acceptable image on a photographic film, this restricts theuse of the transfer technique to moderate and high intensity sources.

The correct exposure for the transfer method is determined by both the lengthof time that the foil is in the neutron beam and the time that the image isallowed to transfer to the film (see Fig. 1.10). Both of these may be varied, asany practical product of the irradiation fraction (activity given to foil) andtransfer fraction ( part of this activity transfered to the film) will give the same result. Table 1.4 gives the characteristics of some converter foil materials.


Fig. 1.10 Irradiation /Transfer Curve.

-----------,.,. ,.

EXPOSURE TRANSFER

TIME


Fig. 1.10 Irradiation /Transfer Curve.

N = Noe -λT = Noe T½

-0.69T


Table 1.4 The Characteristics of Some Possible Neutron Radiography Converter Materials [Ref. 14]

Abundance Emission of Mode of Production Cross- Half-

Material Parent of Active Isotope Section lite Max. Isotope, barns Type Energy. % MeV

0 lithium 7.4 L~ (n,a )H3 935 STABLE a 4.7 0 Boron 19 .5 B

10 j n.a)li

7 3,837 STABLE a 2.3 0 Rhodium 100 Rh

1 3(n,y1Rh

104 144 43 s {3 2.41 Rh,oo(n,n )Rh103m 57 min X-ray 0 .02 Rh 103(n, 'Y) Rh 104m 11 4.4 min {3 0.5

0 Silver 51.4 Ag 101 (n,y)Ag 1oa 44 2.4 min {3 1.64 {3 0.43

48.7 Ag,09(n,y)Ag,,o 110 24.5 s {3 2.87 Ag 109( n, y)Ag 11Om 3 254 d {3 1.5

0 .66 0 Cadmium 12.3 Cd113(n,y)Cd,,4 20.000 STABLE 'Y 9 T Indium 95.7 In 11s( n.y)ln 11s 45 14 s {3 3.3.

0 .44 In 11s( n, y)ln 116m 154 .54 min {3 1.0

0 .42



0 Samarium 13.9 Sm 14~( n. )')Sm15o 41.500 STABLE )'

26.6 Sm1s2(n, "Y)Sm 153 210 46.7 h {3 0.8 0.1

0 Gadolinium 14.7 Gd,ss(n.e)Gd,se 58.000 STABLE e 0.14 15.7 Gd157 (n.e)Gd,sa 240.000 STABLE e 0.13

T Dysprosium 28.1 Oy164(n;y)Oy,ss 800 2.3 h {3 1.29 0.095

Dy164(n;y) Dy165m 2,000 1.26 min {3 1.04 1.108

T Gold 100 Au197(n,)')Au19B 98.8 2.69 d {3 0.962 0.412

0 =direct method T = transfer method



Abundance Emission of 1\/lode of Production Cross- Half-

Material Parent of Active Isotope Section lite Max. Isotope, barns Type Energy. % MeV

0 lithium 7.4 L~ (n.a)H3 935 STABLE a 4.7

0 Boron 19.5 B10 {n.a)li7 3,837 STABLE a 2.3

0 Rhodium 100 Rh 103(n,ylRh 104 144 43 s {3 2.41 Rh,oo(n,n )Rh103m 57 min X-ray 0.02 Rh 1~3(n, 'Y) Rh 104m 11 4 .4 min {3 0.5


48.7 Ag,09(n,y)Ag,,o 110 24.5 s {3 2.87 Ag 109( n, y)Ag 11Om 3 254 d {3 1.5

0.66

0 Cadmium 12.3 Cd113(n,y)Cd114 20.000 STABLE 'Y 9

T Indium 95.7 In 11s( n,y)ln 11s 45 14 s {3 3.3. 0.44

In 11s( n, y)ln 116m 154 .54 min {3 1.0 0.42



0 Samarium 13.9 Sm 14~( n. )')Sm15o 41.500 STABLE )'

26.6 Sm1s2(n, "Y)Sm 153 210 46.7 h {3 0.8 0 .1

0 Gadolinium 14.7 Gdlss(n.e)Gdlse 58.000 STABLE e 0.14 15.7 Gd157 (n.e)Gdlsa 240.000 STABLE e 0.13

T Dysprosium 28.1 Oy1e4(n;y)Oy1ss 800 2.3 h {3 1.29 0.095

Oy164(n,)')Oy165m 2.000 1.26 min {3 1.04 1.108

T Gold 100 Au197(n,)')Au19a 98.8 2.69 d {3 0.962 0.412

0 =direct method T = transfer method


1.1 .6.3 Dynamic Imaging MethodsThe exposure times for the methods described above, even assuming thatintense neutron beams from nuclear reactors are used, are usually greater than one second so they are essentially 'still' techniques. To produce high-speed, or flash, radiography requires a neutron source that will produce anultra- high intensity flash of neutrons lasting a few milliseconds, such as canbe achieved with a pulsed nuclear reactor. This is a reactor which is madesub-critical by removing part of the uranium fuel in the reactor core. Toproduce a pulse the missing fuel rod is passed rapidly through the core,causing the reactor to go critical for a very short time and producing a largepulse of neutrons. Pulse widths of a few milliseconds are possible, so rapidmotion may be arrested in the radiograph. A limitation of this system is that most pulse reactors are only capable of a few pulses per day. Examples ofthis technique have been reported by Mullender and Hart [Ref. 34].


A method of producing a moving picture is to observe the light output from ascintillator screen with a television system (see Figure 1 . 1 1 ). With most neutron radiography units the neutron beam strength is too low to givesufficient light intensity to be seen on a TV monitor, so an image intensifier isplaced between the scintillator screen and the camera. This boosts the lightfrom the screen by about 104 x , and object movement at up to about 3 m/scan be observed with good definition. Where a neutron beam strength ofabout 5 x 108 ncm-2s-1, or greater is available the image intensifier might be omitted, provided the spectral output of the screen is properly matched to the spectral response of the TV tube. A more sophisticate system [Ref. 35] uses an intensifier tube which has a front gadolinium screen, the secondaryelectrons from which are accelerated on to a scintillator screen (this assembly is often called a neutron camera). The tube is [Ref. 35] claimed to have a resolution of 30 line pair per cm and gains of 3~5x106 for cold and thermal neutrons. These tube can be used with a T.V. camera/monitor to provide a remote display and hard copies of the images can be provided.


Fig. 1.11 Television System for Neutron Radiography.SCINTILLA TOR

OBJECT

LIGHT

IMAGE INTENSIFIER LENS T.V. MONITOR

T.V. CAMERA

---------__ _J


A more sophisticate system [Ref. 35] uses an intensifier tube which has a front gadolinium screen, the secondary electrons from which are accelerated on to a scintillator screen (this assembly is often called a neutron camera).

ece

Po· t


1.1.7 Image RecordersAt present the recording materials in general use for neutron radiography are(1) photographic film and (2) track etch films, each of which is described below.


1.1.7.1 Photographic Film and its CharacteristicsNo special films are available for neutron radiography, and standard X-ray and photographic films are normally used. These consists of a base materialwith a gelatine coating in which are dispersed minute grains of silver halides,the grain size varying from 0.1 to 3 μm, depending on the type of film. Whenphotons or electrons fall onto such an emulsion electrons and positive silverions migrate to points of imperfection in the silver-halide crystals and onarrival some of the silver ions are reduced to metallic silver to form the latentimage. When developed with suitable agents the silver halide at the latentimage is further reduced to metallic silver, and the unaffected halide grainsare subsequently dissolved away by the fixing solution, leaving a blackmetallic-silver image. The latent image is distributed throughout the entireemulsion, and the density of activated silver grains increases almostuniformly with exposure.


Development, on the other hand, proceeds from the surface in a complexmanner, grains at different depths converting at different rates as the developer penetrates the emulsion.

After development the silver grains are viewed as a two-dimensional array, and the distribution in depth is not apparent. This distribution causes the grains to appear as groups, and contributes to the characteristic graininess seen in photographic films.


The speed of a film is essentially a measure of the blackening produced by agiven exposure. Blackening is better produced by large grain emulsions, where fewer developed grains per unit area are needed to give arecognisable change. But the larger" the grain the poorer the resolution, sowhilst large-grain emulsions, such as X-ray films, will give rapid andacceptable results for normal inspection purposes, for high resolution workfine-grain film should generally be used.

Contrast and resolution are dependent qualities, and in order to obtain goodresolution on a film the contrast. i.e. the density variations, must be adequate.This. is well demonstrated by exposing and developing a double- emulsion X-ray film and then scraping away part of one emulsion. It is possible for the image on the double emulsion to appear to be of higher resolution then that of the single emulsion owing to the greater contrast. So although the use of thin emulsions will lead to better resolution, sufficient image density must begenerated to achieve the best contrast for the type of film being used.


The relationship between exposure and film density is expressed by acharacteristic curve, where the density is plotted against the logarithm of therelative exposure (see Fig. 1 . 1 2) and density as

Where: Io = intensity of incidence light on the film, IT intensity of transmitterlight from the film.

The neutron radiographic technique produces an activation image of thespecimen and its background on a foil, and the radiation from these regions will determine the relative exposure of the adjacent areas of the film. Thecorresponding density difference will depend on the part of the characteristiccurve upon .which these exposures fall.

T


Fig. 1.12 Typical Characteristic Curve for a Radiographic Film.

Density H&D

Log relative exposure

Base density usually 0.2 H&D?

lOGE


Fig. 1.12 Typical Characteristic Curve for a Radiographic Film.

http://micro.magnet.fsu.edu/primer/photomicrography/filmexposure.html

F

Charlie C

hong/ Fion Zhanghttp://w

ww

.sprawls.org/ppm

i2/FILMC

ON

/

Film Contrast Characteristics

r l--~ 2.0 w 0 ....I <(

() 1.5 -t: 0

D ma.x

t Film Cont rast

50% Exposure Contrast

_.,.,...... -'

~

I

.J

I

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I Shoulder I

_J

Toe ---+-- ...J.Ba:se + Fog Density

-!..._.!.._ 1 1 1 1 '1 2 jJ.I 3:1' 16 -8- -'T -r- . 4 8 16 32 64

RELATIVE EXPOSURE


The steeper the slope of the curve in this region, the greater will be thedensity difference and hence the greater the contrast. As high contrast is abasic requirement for a good radiograph every endeavour is made to work onthis steep part of the curve. A comparative list of some of the photographicfilms that have been used for n eutron radiography is given in Table 1.5.Photographic films suffer from dimensional instability due to changes inhumidity, temperature, and processing [Ref. 49]; an average variation couldbe 0,002 in./in. for triacetate base film and 0,001 in./in. for polyester films.However most modern films are manufactured with polyester bases.


Table 1.5 Approximate Comparison of European and U.S.A. Films [ Ref. 32 ]

Note : Most of the above l ist is arranged in approximate order of speed (fast at the top). The precise relative speed data in the right-hand column was supplied byH.P. Leeflang.

Kodak Ltd. IIford Ltd Kodak Gevaert Eastman DuPont Ansoo Ref. Type (U.K.) (U.K.) Pathe Agfa N.V. Kodak Ltd (U.S.A.) (U.S.A.) Speed

(France) (Belgium) (U.S.A.

Royal Blue Gold Seal Curix RP 1 Royal Blue Type F Blue Brand Red Seal Curix A' 2 Blue Band Type508 lndusvex S Industrial A Structurix-5 Type504 Standard Standard

~ Kodirex· Auto Process llfex Osray T 4 Non-screen Me<ical

x Industrial G Kodirex TypeF 0.2 .. Kodirex Structurix - 010 TypeC 0.3 ·s .. lndustrex 0 Industrial B 0.5

"' Analyse 0.6 -o

-= Crystallex Industrial CX Structurix-0 7 TypeAA Type 504 Type A 1 -o Defioex 1.2 c: .. Type506 1.4 -.; Industrial C 1.5 u 'g TypeT TypeB 2 :iE Industrial F 3.1

Structurix-0 4 TypeM Type 510 3.8 Microtex 4

TypeM 5 TypeR 8

Structurix-02 15 ---- --------------------------------------------------------- -----Royai -X pan H.P.S. Aerial N lsopan rec<rd Royal-X Pan

u Panchro-Royal H.P.3sheet lsopan 27 :E Cl. Tri-X pan H.P.3 roll lsopan ultra Tr~X pan e Super-XX Selochrome Pan lsopan 24 Super-XX "' 0 Plus-X pan F.P. 3 -0 Super XX Ariel ·(?eXt p~~

.<: 0.. HR Aerial

Paratomic X (55) -------- ----------------- ----"' c: ·:;. R5.50 Cl. 0 u


Type

> .. ~ x .. ·s .. ::> "0 c "0 c .. .. u

1l ::.

.!! s:. c. e 8' -0

-&.

Kodak Ltd. (U.K.)

Royal Blue Blue Brand lndustrex S Standard Kodirex· Auto Process

Kodirex

lndustrex 0

Crystallex

Microtex

Royai·X pan Panctv-o-Royal Tri-X pan Super-XX Plus-X pan Super XX Ariel

Panatomic X (55)

IIford Ltd (U.K.)

Gold Seal Red Seal Industrial A Standard llfex Industrial G

Industrial B

Industrial CX

lnd us trial C

Industrial F

Kodak Pathe (France)

Kodirex

Analyse

Oefinex

TypeM

Gevaert Agfa N.V. (Belgium)

Curix RP 1 Curix If' 2 Structurix..S

Osray T4

Structurix · 010

Structurix· 0 7

Structurix · 04

Structurix-02

Eastman Kodak Ltd (U.S.A.

Royal Blue Type F Blue Band

Non-screen Medical

Type F

TypeAA

TypeT

TypeM

TypeR

Ou Pont (U.S.A.)

Type508 Type 504

Type 504

Type506

Type 510

Ansco Ref. (U.S.A.) Speed

0.2 TypeC 0.3

0.5 0.6

Type A 1 1.2 1.4 1.5

Type B 2 3.1 3.8 4 5 8

15

---------- ---------------------·----·--------------------------H. P .S. Aerial N H.P.3 sheet H.P.3 roll Seloctv'ome Pan F.P. 3

HR Aerial

lsopan record lsopan 27 lsopan ultra lsopan 24

Royal- X Pan

Tri-X pan Super-XX

--------------·------------------------------------------------------------------------"' .: > g.

R5.50

u


1.1.7.2 Track- etch RecordersPhotographic film has some disadvantages. It must be developed underdarkroom conditions, and the dimensions of the film can change withvariations in temperature, humidity, and development. Alternative materialsare feasible, and experiments with detectors for health physics applications inthe early 1960s showed that the damage tracks caused by bombarding micawith heavy particles could be 'fixed' by etching in an acid solution and thatthese could be observed with an electron microscope.

The technique was developed for direct viewing by using a combination of a boron (or lithium) foil with a nitrocellulose film, a combination which utilises the neutron/alpha- particle reaction in boron (10

5B(n,α)73Li) , the alphas

causing surface-damage pits in the cellulose and, unlike the beta particles emitted from the metal foils currently in use with the transfer process, the alpha particles take short and relatively straight paths through a material andso give good resolution.

Keypoints: alpha particles take short and relatively straight paths through a material and so give good resolution.



.Materia1l

D Lhhium

D Boron

D Rhodiu·m

D SJJver

D

T Indium

Abundance of fv'bde of Production Parent of Active Isotope sotope •.

'%

7.4

19.5

100

51.4

48.7

12.3

'95.7

Li6: (n~a) H3

e 1'0 -{n.a)Li'

A·. 1109( v)A· . .. 1110 Q 0,1, · IQ .Ag 109( n,.y).A

1

g 1111 Om

Cd1113(n,;y)Cd1114

Jn 1111S( n,.y )j nne

ln115(n y)Jin 11em

CrossSection bans

Half.. life

:EmJSS,JOn

M,ax. Type Energy ...

MeV

935 STABLE a 4.7

3,837 STABLE .cr

144 43 s f3 5'7

. .X-ray m11n

11 4 .4 . f3 m1n

44 2 .. 4 ·mJn [1 (3

110 24.5 s, f1 3 254 d (3

20.000 STABLE y

45 14 s [1

154 .54 . f1 mJn

2.3

2.411 0.02 0.5

11.164 0 .43 2.87 1.51 0.1616

'9

3 .. 3 ... 0.44 1.0 0 .42


In 1973 Kodak, France, marketed a nitrocellulose film for neutron radiographycoated on both sides with lithium borate. After irradiation the lithium borate isremoved by washing and then the film is etched. The contrast is very low butthis can be significantly improved by printing on to a copying film, using a point source enlarger.


Farny [Ref. 36] has made the following comparison between nitrocellulose and photographic film/foil techniques, based upon the use of type CA 80-1 5 8 and a gadolinium foil and type R film (single coated X-ray film):


Comparison between nitrocellulose and photographic film/foil techniques■ Direct method

(a) The performance differs little for a given fluence, but as cellulose nitrate is insensitive to γ's a bismuth filter is unnecessary, which leads to a flux gain for the same source and consequently higher overall efficiency.

(b) Using cellulose nitrate the definition is better if the examination can be made on the negative, it becomes comparable if copying is unavoidable due to lack of contrast.

(c) The contrast is weaker, but can be improved by copying. Moreover, the feasibility of stopping development at intermediate stages without fixing permits the production of many prints of differing contrast from a single exposure.

(d) As in X-radiography, cellulose nitrate film can take the shape of the object during irradiation, or be folded to reach into a cavity.

(e) The customary use of the film in industry usually involves routine printing from the cellulose nitrate, which is a disadvantage considering its simplicity when used in the direct method.


■ Indirect method

(a) The irradiation must be 3-5 times higher to obtain an image comparable to that with Dy and Type R film exposed to infinite

(b) The definition of cellulose nitrate is markedly better ( range of α’s less than range of = 1 MeV β's).

(c) There is no handling of an active converter.(d) This film lends itself much better to measurement with a profile projector.(e) Contrasts are weaker, but this is not always a drawback. For example,

when neutron radiographing irradiated fuels, cellulose nitrate is preferable for the examination of fuel cladding contact, as the difference in contrast between the two materials is very much less.

(f) Because of the linear flux response, there is no saturation of the converter, which can be advantageous when using low intensitysources. Moreover, the absence of decay and the reduction of exposure times considerably simplifies the whole process of neutron radiography.

(g) Nitrocellulose is a 1/v detector and so it cannot replace indium for epithermal flux work. (?)


Barbalat [Ref. 37] has investigated the effectiveness of several converters when used with Kodak CN 85 nitrocellulose film and found that the relativespeeds (in decreasing order) of these were:

a) 10B (75p.)b) 6LiF (50p.) / 11 B (50p.)c) 7LiF (50p.)/ 7LiF (35p.)d) 7LiF (30p.)

The developer used was 150 g/l of KOH at 40°C and the developing time was 30 mins. The etching bath was stirred before use and long etching times were avoided in order to prevent sediment formation in the bath due to the camfer removed from the nitrocellulose. Barbalat reported that agitation during the etching causes cloudiness on the fi lm, but not all workers have found the phenomenon. Close temperature control is important so the bath should be placed in a temperature controlled medium. The CN85 nitrocellulose can be directly examined by placing it between two polarised filters [ Ref. 46], or by enlargement with an optical projector. Measurements may also be made on a micro-densitometer using the polaroid filters. However the filter does not giveany improvement over the direct examination method.


1. 1 .8 Film and Foil RelationshipsFor a particular neutron radiographic facility the choise of film and foil willdetermine the information that can ben recorded in the image. Because theradiographer normally has to make a trade-off between speed and resolution the optimal film and foil combination will change with every type of object tobe radiographed. The following sections discuss some of the factors relatingto this trade-off.

1 . 1 .8.1 Film and Foil SpeedThe speed of films and foils used in neutron radiography have been measured by Berger [Ref. 4], Hawkesworth [Ref. 20] and many other workers,but the results are always relative to the neutron beam which was used. Thevariation in published values are not large but for the praticing radiographer itis important to have a reliable set of film characteristic curves and one of hisfirst tasks is to carry out a series of calibration measurements on a range offilms using d ifferent types and thicknesses of foil and screens. A typicalgroup of these curves for a number of films and foils is shown on Figure 1.11 (?) . The data used in this figure were obtained from the DIDO radiographyunit at Harwell.


1.1 .8.2 Film and Foil ResolutionMethods of recording i mages in neutron radiography depend on anintermediate detector foil" which emits radiation to which the film is sensitive.This introduces unsharpness into the recording process, for while theneutrons passing through the specimen and arriving at the foil may have anarrow angular spread, the radiation emitted from the foil has not, and mayenter the film obliquely and be recorded at some small lateral d istance fromits point of origin in the foil. The magnitude of this effect depends on the rangeof the particle and the thickness of the detecting foil and film, and isminimised by making these as thin as possible. Reducing the thickness of thefoil and film however leads to a reduction in sensitivity of the system toneutrons.

The dependence of film blackening on particle energy is an inverse function of the average energy of the particle, the most efficient energy for this process imately 100 keV [Ref. 27]. This is demonstrated by the superiorresolution of a gadolinium foil, which emits an internal-conversion electron of70 keV, compared to indium and dysprosium which both emit particles of about 1 MeV.



.Abundance of Nbde of Production P,arent of .ActJVe llsotope Isotope.,. %

.D lithium 7.4 ~· ( )H:a: L1 n.a

D Bo on 19.5

D Rhodium 100 'Rh,.la:a:.( .. ?J· R. •ht04 . 1 n~ . Rh103(n,n )Rh103m

Rh,~~(nt Y)Rh 104m

D SJJver 51.4 107 . . . ·· . 108 .Ag (nY).Ag ·

48.7

D Cadm1Ju·m 12.3

T Indium 95.7

Cr,oss- H.a1Jf.. SectiOn hfe bans

935 STABlE

3,,.837 ST.ABlE

144 43 s 57

. m1:1n

11 4 4 +

m n

44 2 .. 4 ·•· ·m•n

110 24.5 s,

.3 254 d

Emiss,ion

IMiax. Type Enelgry •.

.a

.a

(3 x-ray (3

(3 {3 (3 {3

MeV

4.7

2.3

2.411 0.02 0.5

11.164 0.43 2.87 1.51 0.616

20..000 STABlE y 9

45 14 S;

154 .54

p

f3

3.3 ..

1.0 0 .42



D

D

T

T

Samarjum 13 .. 9 216,16

G .ado'li~n.ium 114 .7 116.7

Oysp osiium1 28.. 1

Gold 1100

D -== dimct method

Sm1'4~(n. y)Sm1511

SmliSZ(n"' )1Sm 153

Dyl64(n

Dy,64(

165 y

Au197 (n. Y.)Aull·ss

T = tra sfer method

41 .500 STA:BlE y .210 46.7 h (1

58.000 STABlE e .240.000 STABlE e

23h {1

2.000 1.216 m1:in {1

98 . .8 (3

Q,8

0.11

0 .1'4 0 .13

1.29 0 .095

OBI6.2 OA12


The literature often describes methods of measuring the resolution capabilities of a film/foil combination by radiographing an indicator on whichthere are a number of objects of graduated size, such as wires or a number ofwebs between a row of closely spaced holes, and the smal lest size that canbe discerned by d irect viewing is taken as the resolution capabil ity. Suchmeasurements are not a determination of the resolution of the film/foilcombination alone for they include the unsharpnesses in the totalradiographic system. Also when the total system unsharpness is greater thanthe size of the wire of web (which can occur with the smallest sizes) there isan additional complication due to the resulting overlapping of the edges of theimage, causing the width of the image to increase and the i mage contrast tobe reduced. The method usually used is to place a thi n opaque knife edge incontact with the foil and assume that because it is in intimate contact with thefoil, the geometric u nsharpness wil l be negligible.


The measurement will, of course, include unsharpness due to film/foil contact, but if a vacuum cassette is used then this unsharpness can also be assumed to be negligible and the measurement taken as the fi lm/foil unsharpness. The unsharpness measurement is made by using a microdensitometer to record the density variations across the image of the edge. The method evolved by Klasens [ R ef. 48] is usually used to determine the unsharpness from themicrodensitometer curve, in which a straight line is drawn to cut the 'S'- hapeddensity curve at 0.1 6 x (density range of test object). The projection of thisline on the maximum and minimum plateau l ines of the density curve givesthe unsharpness value (see Figure 1 . 1 3) .


Fig. 1 . 1 3 Measu rement of Unsharpness . A. Exposur e distribution across th e edge of a specimen

NEUTRON BEAM

RECTANGULAR I FOIL

METAL BAR ~~--~ ----t-,r /

F ILM DENSITY

t:::• :::k::=~===:===:::=J I TRANSFER I TO FILM

IMAGE SCAN

EDGE OF SPECIMEN


Fig. 1 . 1 3 Measu rement of Unsharpness . B. Klasens' method

INTENSITY

u


There are three types of unsharpness in a radiography system: geometric, foila n d film. Geometric unsharpness was discussed earlier a nd is readily determined from the collimator dimensions and the size of the sample but theother two are difficult to seperate and must be assessed together by theKlasens technique. The combin'ed effect of different types of unsharpnesswas also considered by Klasens, who evolved the following form of empiricalrelationship:

Ut = (Ug3 + Uff

3)1/3

where Ut = total unsharpnessUg = geometric unsharpnessUff = foil and film unsharpness

15


Thus the source of unsharpness cannot be considered in isolation and theirinterdependence shows that there is little point in achieving a small geometricunsharpness if the foil/film unsharpn ess is much larger.Whilst the a bove methods of measuring unsharpness are widely used in allforms of radiography they fail to acknowledge that resolution is a function ofimage contrast for they only give the resolution at one particular contrast value. The important failing of this system is that is does not allow theradiographer to systematically evaluate his radiography system to determinethe conditions under which it will produce the maximum information. A systemwhich comes much closer to this ideal has been in use by designers ofcommunication and optical systems for many years and has been advocatedfor use in radiography for about a decade.


This is a frequence response method of evaluating the performance of a system, usually called the Modulation Transfer Function. It can be shown that any image, even a series of step changes in intensity, may be expressed as a series of sine waves of differing frequency, amplitude and phase by the use of Fourier analysis. It follows therefore that a test-object which will transmit an intensity pattern which varies sinusoidally with known frequency and amplitude would provide a means of evaluating the detail-recording capability of a radiographic system. Such a test-object is shown at Figure 1 .1 4 andideally the radiographic system should reproduce an image of such an objectwithout loss or change of information.


In practice such test objects are difficult to produce and so the requiredinformation is obtained by the use of a narrow slit, or more conveniently, asharp edge. This, of course, brings us back to the type of curve shown in Figure 1.13, and by analysing this curve into its component sine waves thetype of curves shown in Figure 1 . 1 5 can be produced to show how theresolving power varies with contrast. The resolving power is expressed as theobject spatial frequency and this can be conveniently regarded as themaximum number of lines per millimeter, that can be resolved. For example,a spatial frequency of 2 cycles per mm would be equivalent to a resolution oftwo l ines per mill imeter (two 14 mm wide lines seperated by a 14 mm gap).The figure shows that the gadolinium foil will give the best resolution over theentire contrast range that was measured, only being equalled by the slow X-ay film at the contrast of about 0. 1 .


This resolution is at the expense of speed, in general the higher the resolving power the slower the dfltector. Whilst this method has considerable advantages over the methods described earlier, it is time-consuming to apply and so it is mainly used by experimentalists and designers .who wish to analyse a complete radiographic system. Modulation transfer functions of in dividual components in a system i.e. the collimation, converter foil, photographic film etc. can be measured separately and then multiplied to give the overall modulation transfer function. Such a technique is a great aid to design, in that the weakest part of the system can be evaluated and improvements made where they will be most effective. For further information on the frequence responce method see Halmshaw [Ref. 42] and Halms [Ref. 43].


Fig. 1 . 1 4 Diagram of a Test Object whose Transmission varies Sinusoidallyalong its Length.


Fig. 1 . 1 5 Contrast Transfer Functions of N E 241 and N E 905 Scinti l lators,Gadolinium Foil and X-Ray Films.

.... .s~ ••• .0-1--'

·· ... ~-9. ...... :'/ r-, . . . . •• •

0.2 0.5 1.0 2 5

OBJ ECT SPATIAL FREQUENCY, CYCLES/MM

-J" DENOTES THE SPREAD OF SEV ERAL

J:- INDEPENDENT MEASUREMENTS

10


1 .1 .8.3 Some Observations on Resolution SensititvityRadiography is commonly used to detect voids and inclusions within a material and so it is often necessary to know the m inimum size of void thatcan be detected. Let us ask the question 'Can the detector/recordercombination discriminate between the beam intensity which will form theimage of the bulk of the sample and the intensity which will form the image ofthat part of the sample which contains the void.' The answer will depend uponthe response of the detector/recorder being used. With a metal foil incombination with a photographic film the photographic response is linear;when the combination is a light-emitting scintillator and a film then theresponse will be logarithmic. Consider these two cases. The relationshipbetween the incident neutron intensity and the transmitted neutron intensity(ignoring scattering) will be:


Ф = Фo e-Σx

where Ф = transmitted intensity, n∙cm-2s-1

Фo = incident intensity, n∙cm-2s-1

Σ = total macroscopic cross section, cm-1 (Σ = σN)x = thickness of sample, cm


Assuming we are using a metal foil then the film density ( De) will have a linear relationship with the neutron exposure, i.e.

De ~ Фand we can rewrite (1 6) as

Di = Dbe-Σx

Where:Di = image density of the filmDb = background density of the film

Differentiating and expressing as sma l l differences gives:

∆Di = Db∙Σ∙∆x Where: ∆Di = minimum detectable density change∆x =minimum detectable thickness change in the object cm


For a metal foil detector

De = G∙E (20)

Where:G = slope of density/ exposure curve for the filmE – exposure n∙m-2

So from (19) & (20)

∆X = ∆Di/(G∙E∙Σ) (21)

This equation shows that in order to detect the smal lest thickn ess cha nge, or detect the smallest void, then the film must have the highest contrast (G)and the exposure must be as high a possible.


Now the second case is where the metal screen is replaced by a light emittingscintillator, and for this there will be logarithmic relationship between filmdensity and neutron exposure, i.e.

De = G(log E) (22)De = G(0.434 ln E) (23)

Differentiating and expressing as small difference givesDe = (0.434∙G∙∆E)/E (24)Now as in equation (19) the fractional difference in exposure can be expressed:∆E/E = Σx, (25)and so (25) becomes

∆De = 0.434∙G∙Σ∙∆x, or (26)∆x = 2.3∙∆De/G∙Σ (27)

Where: ∆De = minimum detectable density change∆x, G, Σ = as above


Hawkesworth [Ref. 23] considered the a bove sensitivity equations and showed that for ionising radiation (i.e. metal foils) the background radiationincident on the film during the exposure has no effect on the thicknesssensitivity. For light emitting screens however the thickness sensitivity isdependent upon the background exposure and equation (27) becomes:

∆x = (2.3∙D/G∙Σ) (1+E+Db/E)

Where:Db = background exposureEo = intercept of the film characteristic curve on the abscissa-see figure 1.14


This leads to the desire to maximise the exposure to achieve the highestthickness sensitivity, but is must be said that this may be nullified by theconsequent increase in the background exposure caused by the gamma radiation in the beam. We now have two expressions which give a reasonabledescription of the minimum detectable void, or the minimum detectablethickness of a nonscattering material in a neutron beam which is free frominterferin g radiation such as fast neutrons and gamma rays. The a llowancewhich must be made when scattered neutrons and gamma rays are presentdepends upon the type of detector in use. If the transfer technique is beingused then the gamma rays arriving at the detector will not be recorded, andsimilarly some of the scintillators (direct technique) have a sufficie ntly lowgamma sensitivity for this factor to be ignored. If a gadolinium foil (directtechnique) is being used then all of these radiations wil l be present and wil lcontribute to a reduction in the contrast-sensitivity.


Furthermore thermal neutrons are attenuated by matter in two ways, they areabsorbed or scattered. In the first case they are removed from the beam and in the second they are scattered in all directions from the target nuclei. So tothe detector the first process in unequivical, the neutron is removed from thebeam and is simple not there to be recorded. The response to scatteringhowever depends upon the spacing between the scattering material and theobject. If they are widely spaced the most of the attenuated neutrons will bescattered out of the beam and once more will not be there to be recorded.This 'widely spaced' geometry is, in principle, similar to that used when crosssection measurements are made and so the equations which use crosssection values wil l only be accurate for scattering materials when the objectand the detector are spaced apart. However, the usual practice in neutronradiography is to place the object and the detector as close as possible, andso many of the scattered neutrons are recorded.


Gamma rays are also scattered and absorbed by the sample but their absorption is usually negligible at the gamma ray energies which predominates in the beam. The effect of gamma ray scattering on the radiograph is similar to that of scattered neutrons in that they reduce the contrast. The gamma ray component of the beam is always made as small as possible and so gamma ray scattering is not usually a serious problem (unless a gamma emitting specimen is being radiographed), although a direct technique neutron radiograph can usually be improved by placing a lead or bismuth filter in the beam. A more detai led examination on the detection of voids is given in Appendix 1 .4.


1. 1 .9 Neutron Beam FiltersNeutron radiography is usual ly performed with a neutron beam containing arange of energies, but the predominant value will be the thermal energy (-0,025 eV). N ow although the converter foils are generally most sensitive atthe thermal energy, and the response fa lls stead ily as the energy increases,superimposed upon this effect are the large resonances which occur atwelldefind energies. At these energies the cross-section of the foilincreases/decreases rapidly, often by several orders of magnitude, and thengenerally returns to about its previous value. So if a filter is placed in theneutron beam which will only pass neutrons whose energy coincides with theresonance in the converter foil then the system will be more sensitive to theremaining neutrons. This ideal situation can be approached by using forexample, a cadmium filter and an indium converter foil. The cadmium willabsorb all neutrons of energy below about 0.4 eV and so leave the beamrelatively rich in higher energies.


One of these, 1 .4 eV, is the energy of a resonance in indium at wh ich the cross-section increases from 2 x 1 02 barns to 3 x 1 04, and a n example of the use of this particular filtered neutron energy is found in the radiography of fuel pins where the cross-section of uranium is lower at 1 .4 eV than at thermal energy and so a fuel pin is more readily penetrated by neutrons at this h igher energy. Spowart [Ref. 33] investigated this effect and found that the penetration of a 0.6 em diameter (20% Pu, U235-enriched, mixed-oxide pellet) was improved by about 40 x by using 1 .4 eV neutrons. It must be recognised that this technique works because the thermal neutrons, which are normally responsible for most of the film blackening, have been removed from the beam. With a non-filtered beam these would fully expose the image before the more penetrating neutron could make a perceptible contribution. The filter therefore usually reduces the beam intensity and a longer exposure may be required. This effect will vary with the neutron energy, and thematerials of the foi l and the sample, but taking the fuel pin example g iven above we can say that the increase in exposure time wi ll be a function of:


(a) the ratio of 1 .4 eV neutrons to thermal neutrons in the unperturbed eutron beam, the decrease in neutron intensity produced by the filter,

(b) the change in transmission through the sample, and(c) the increased sensitivity of the foil.


The effect of these factors will, of course, vary with each beam and samplecombination. Further examples of this technique are discussed by Miller andWatanabe [Ref. 38] who used a combined cadmium/gadolini um/indium filter to detect the 4.3 eV resonance in tantalum. F;g. 1 . 1 6 shows thetransmission of the filter, the resonance in the gold detector and the object.They also discussed the use of a high purity silicon filter to penetratehydrogen. Cold neutrons [Ref. 1 9] have a lso been used to increase theneutron penetration of crystalline materials and Fig. 1 .8 shows the fall inneutron attenuation coefficient for aluminium, iron, zirconium, tin and l ead,compared to say hydrogen and gadolinium, where the attenuation coefficientincreases. A Jist of potential filter materials is given in Table 1 .6.


Fig. 1 . 1 6 Resonance Curves for a Cd/ln/Gd Filter, a Gold Foi l and a Tantalum Sample.

~

tft -z 0

~ :E (/)

z c( a: 1-

Gt

60 ' ' ' (~ tr NEUTRON FIL TEA TRANSMISSION

40 f- 0 .500- Cd (1 0.050- In 0.008- Gd

20

0

-f

' 1 1

104

10 3 Au DETECTOR

102

10 4 ~====~==~==~~~~~ 103

10

1 1.0

Ta OBJECT!

10

NEUTRON ENERGY (EV)


Table 1 .6 Neutron Filters 1 J KEY TO RESONANCE TYPES

Element Minimum Absorption Maximum Absorpt ion Temp. Resonance Energy. eV Cross-section Energy. eV Cross-section OK Type 1)

Barns Barns

Beryllium 4 5x1 o-3 0.45 7x1 o-3 6 300 A Beryllium 4 3.5x1 0-3 0.05 7x1 o-3 6 100 A Beryllium Oxide 2x1 ()3 2 6x1 o-3 10 A Boron 5 6.5x10-4 4.5x103 3x103 6 c Carbon 6 1.5x1 0-a 0 .55 2x1 o-3 8 A Sodium 11 5x1d 3.2 3x103 380 0 Silicon 14 1.4x105 0.45 1.9x1 05 12 E Sulphur 16 7x1 04 0.45 1.2x1 05

20 E Scandium 21 30 20 4x103 40 G Iron 26 2.6x104 0.45 2.9x1 04 3{av.) E Rhodium 45 1.3 5x103 20 4.5 0 Rhodium 45 0.5 100 1.0 10 0 Cadmium 48 0.18 8x1rl 10 4.5 B Xenon 54 0 .1 3x106 1.0 104

B Promet hium 3.5 40 5.5 2.5x104 F Gadolinium 64 2x1 ()3 1.4x1 05 1.5 35 B Bismuth 83 6x10-4 0.55 2x1 o-3 8 300 A Bismuth 83 8x10-4 0 .3 8x1 o-3 1 100 A Proactinium 91 0.4xi 05 0.05 1.5x106 1 A Plutonium 94 0.3 HY 6.0 20 F


1) J KEY TO RESONANCE TYPES

A 8 c 0 E F

G


1.1.10 TomographyTomography is a radiographic technique first developed for medical Xradiography in which a series of exposures are taken at uniform intervalsaround a specimen and, by the use of a computer programme, the data istransposed to give a picture of a cross section a right angles to the plane ofthe radiographs.

Barton [Ref. 1 7] et al has used epicadmium neutron radiography with thismethod in order to examine nuclear reactor fuel bundles having 2 1 7 pins in a hexagon array, requiring the penetration of 9 fue l pins having a total attenuation of 47 cm-1.

The computer output is in the form of a 'dot-picture' in which the sensivity islimited by the dot size, but at the moment the system resolution is in the orderof a few millimeters and so this is not a limitation. However, a resolution of afew millimeters is, by normal standards, very poor, but even this is much better than no information at all from samples that are normally extremely difficult to penetrate.


1.2 THE DESIGN OF NEUTRON RADIOGRAPHY EQUIPMENTHaving considered the principles relating to neutron radiography we can nowexamine some of the topics discussed earlier in greater depth in order to design or select a neutron radiography unit suitable for the radiography of aparticular range of samples.


1 .2.1 The Choice of Neutron SourceThe starting point is to ask a few questions, viz 就是1. What is the object radioactive,2. what resolution do you require,3. can the object be taken to the equipment or must the equipment go to the

object,

https://en.wikipedia.org/wiki/Neutron_source


in order to determine whether reactor or an accelerator/isotopic source is to be used. Reactor sources are undoubtedly the most powerful of those discussed earlier in this text and if high resolution is required then a reactor source must be used. Its disadvantages are that it is immobile and expensive. A small reactor would cost $ 500,000 or more, it requires a license to operate and a small staff to out those operations. On the plus side of the argument can be added the fact that it would provide a larger number of neutron beams than the accelerator and isotopic assemblies, the flux is more stable and the cost per neutron is much less. Typically a nuclear reactor will provide neutron fluxes at the collimator inlet that are some 103 to 106 (one thousand to one million) times higher than those from the alternative sources.


Let us assume that your resolution requirements are sufficiently exacting torequire the use of metal foil converter and that you require to locate the neutron radiography equipment in your works to examine a variety of components. Your company has radiography department who's staff are familiar with high energy X-ray equipment and so you advise the management to obtain a 3 MeV Van der Graaff accelerator with a beryllium target of the type shown in Figure 17.

This arrangement will produce neutrons with an energy of about 5 MeV and this can be moderated by inserting the accelerator tube into the centre of a block of polythene or a tank of water. If you now ask for the assistance of the Physics Department in your organization to determine the magnitude and position of the peak thermal flux they will take the details of the flux spectrum from the accelerator target and perform the necessary diffusion theory calculations. The peak thermal flux will probably be……….


The peak thermal flux will probably be about 20 cm from the target and the thermalisation factor i.e.

(fast neutron flux at the target, n•cm-2s-1)(peak thermal neutron flux in the moderator, n•cm-2s-1)

will be about 100. If we assume that your accelerator produces a sourceintensity of about 4 x 1011 ns-1 this will give a peak thermal flux in themoderator of 4 x 109 n∙cm-2s-1 .


Fig. 1.17 Van Der Graaff Accelerator for Neutron Radiography.

VAN DER GRAAFF ACCELERATOR

VACUUM PUMP

WATER MODERATOR TANK

F1LM-F01L CASSETTE


Van Der Graaff Accelerator for Neutron Radiography

http://neutron.ujf.cas.cz/vdg/graaff-principle.html


The Van De Graaff Accelerator.

https://www.helmholtz-berlin.de/zentrum/locations/historie/lise-meitner-campus/index_en.html


The van de Graaff accelerator

https://www.helmholtz-berlin.de/zentrum/locations/historie/lise-meitner-campus/index_en.html


Neutron Source: Linac


Neutron Source: Accelerator



. I ' r


1 .2.2 The Collimator1 .2.2.1 Collimator DesignThe next stage will be to consider the collimation, and the neutron flux required at the foil. Your objects will not be radioactive and so the directtechnique can be used, and, as the Inspection Department require aresolution of (say) 0.01 cm you are going to use a gadolinium converter foil.As the incident neutrons wil l all be absorbed in about a 10 μm th ickness ofgadolinium then the foil can be thin and a 25 μm thickness mounted on analuminium backing plate would be a practical choice. As the photographic filmwill be placed in the neutron beam with the foil it will respond to the gamma-rays in the neutron beam. The gamma exposure will reduce the contrast soyou will probably need to place a lead or bismuth filter, say 0.5 cm thick,across the entrance to the collimator to absorb the unwanted gamma's.

Keywords:mounted on an aluminium backing plate


gadolinium converter foilTable 1.4 The Characteristics of Some Possible Neutron Radiography Converter Materials [Ref. 14]

D

D

T

T

Materia1l

Abundance of Parent :Isotope.,. %

1.3 :9 216. 16

GadoJinium 14.7 15.7

Dyspros~um1 28. '1

Gold 1100

rv'bde of Production of ,Active lsot,ope

D 1164( . v,,0 .. _1ss, Y n,, Y

Dy164,(n,.i')rDy165

Au 11917 (n, y)Au 1ss

T =transfer method

Ha1lf.. ]ife Max.

CrossSection ba·ns Type EneJgry ...

4'1.500 STABLE y 210 46.7 h f1

58~000 e 240,000 STABlE e

aoo 23 h f1

98.8

MeV

0.8 0.11

0 .114 0.113

11.29 0 .095 11.04 11.108

0.9162 OA12


The L/D ratio of the collimator can be determined from the equation (29) i. e.

Ug = D∙Lf / Ls (29)

Where:Ug = geometric image unsharpness = 0.01 cmD = source-aperture sizeLf = image-to-object distance = 0.5 cm (film to object distance)Ls = source-to-object distance

if we make the assumption thatLs = L = collimator lengththe equation now becomes:

L/D =Lf / Ug (30)


and so knowing that Lf (film to object distance) is normally equal to the thickness of the object and given that Ug is known, then L/D can be determined.

The next problem to be considered is the selection of a material to line thewalls of the collimators choosing from the l ist given earlier, i. e. boron (in theform of boral), cadmium, dysprosium, europium, gadolinium or in dium. The nuclear effectiveness of these materials can be assessed from Fig. 1.6. Thisshows the total cross-section plotted against the neutron energy, and theabsorption-to-scattering cross-section ratio for thermal neutrons. The effectiveness of an absorbing material will vary with the neutron-energyspectrum of the neutron beam (compare the cross-section of Cd and In at 10-2

and 1.4 eV), and this spectrum can be broadly characterised by the cadmiumratio* of the beam. (The ratio of two neutron flux measurements made by irradiating a foil or a wire with and without a cadmium cover. Such a cover is taken as giving a cut-off at 0.4 eV (0.5 eV?) .)


Cadmium ratio - the ratio of the response of two identical neutron detectors, usually activation types such as indium or gold, one exposed bare to thebeam and the other cadmium covered (the cadmium covered detectorrecords primarily neutrons having an energy above 0.5 eV and the ratio is ameasure of thermalization in the neutron spectrum).

PRACTICALNEUTRONRADIOGRAPHVJ. C. DomanusEditor


Fig. 1.6 shows that on the left-hand side of the cadmium cutoff line the mosteffective materials are europium, gadolinium and cadmium. All of thesematerials have a high absorption-to-scattering cross-section ratio, whichmeans that there is a high probability that a neutron wil l be absorbed in thelining rather than being scattered into the beam.


Fig. 1.16 Resonance Curves for a Cd/ln/Gd Filter, a Gold Foil and a Tantalum Sample.

........ ~ 0 ~

z 0 -C/) C/) -:a C/)

z < a: 1-

O't

O't

60 I I I I' NEUTRON FILTER TRANSMISSION

40 ~ o.soo- Cd n o.oso- In o.ooe- Gd

20 -

0 'I I I _j_ I I

104

10 3 Au DETECTOR

10 2

10 4 r--------.----.---,n,-,--.-.-~~

10 3

102

10

1 1.0

Ta OBJ ECT

10

NEUTRON ENERGY (EV)


Table 1.6 Neutron Filters 1 J KEY TO RESONANCE TYPES

Element Minimum Absorption Maximum Absorpt ion Temp. Resonance Energy. eV Cross-section Energy. eV Cross-section OK Type 1)

Barns Barns

Beryllium 4 5x1 o-3 0.45 7x1 o-3 6 300 A Beryllium 4 3.5x1 0-3 0.05 7x1 o-3 6 100 A Beryllium Oxide 2x1 ()3 2 6x1 o-3 10 A Boron 5 6.5x10-4 4.5x103 3x103 6 c Carbon 6 1.5x1 0-a 0 .55 2x1 o-3 8 A Sodium 11 5x1d 3.2 3x103 380 0 Silicon 14 1.4x105 0.45 1.9x1 05 12 E Sulphur 16 7x1 04 0.45 1.2x1 05

20 E Scandium 21 30 20 4x103 40 G Iron 26 2.6x104 0.45 2.9x1 04 3{av.) E Rhodium 45 1.3 5x103 20 4.5 0 Rhodium 45 0.5 100 1.0 10 0 Cadmium 48 0.18 8x1rl 10 4.5 B Xenon 54 0 .1 3x106 1.0 104

B Promet hium 3.5 40 5.5 2.5x104 F Gadolinium 64 2x1 ()3 1.4x1 05 1.5 35 B Bismuth 83 6x10-4 0.55 2x1 o-3 8 300 A Bismuth 83 8x10-4 0 .3 8x1 o-3 1 100 A Proactinium 91 0.4xi 05 0.05 1.5x106 1 A Plutonium 94 0.3 HY 6.0 20 F


1) J KEY TO RESONANCE TYPES

A B

c 0

E F

G


Cadmium has a disadvantage in that it emits a high energy gamma ray when a neutron reaction occurs and this will add to the unwanted radiation in thebeam. The information on the-right hand side of the cadmium cut-off lineindicates that, for the higher-energy neutrons, indium is generally the mosteffective, closely followed by all of the other materials except cadmium.

It is thus clear that there are no outstanding materials from the neutronicviewpoint and so the cost and the mechanical properties of these materialsmust also be considered when making a selection. The cost of each of thesematerials is clearly something which could vary considerably with time sothese must be determined at the time of need. The properties of interest forboron, cadmium and europium are given in the Tables 1.7, 1.8 and 1.9 of the following section and those for indium, dysprosium and gadolinium will befound in the Section of Characteristics of Foil Materials (see 1.2.3.1 ), Tables1.10, 1.11 and 1.12


1 .2.2.2 Characteristics of Lining materialsBoron is a light metal of high hardness and melting point which is normallymade into solid shapes by powder metallurgy techniques. It can be obtained in the forms of boron carbide, boron oxide, boron nitride and boral. This latterform is a mixture of boron carbide and aluminium which is clad in aluminium.

Boral is somewhat difficult to machine but can be sawn, sheared or sparkeroded. A 6 mm thick sheet can be rolled to a minimum diameter of about 200mm. When used as a converter foil the material is in the form of enriched 10B vacuum deposited upon aluminium, or as a boron powder in a plastic matrix.

Table 1 .7 shows the natural material consists of 20% B10 and 80% B11 andthat under neutron irradiation the B10 is converted to Li7 with a production of a2.3 MeV alpha particle which is responsible for the damage tracks innitrocellulose when the track etch method is used.

* ) The ratio of two neutron flux measurements made by irradiating a foil or a wire with and without a cadmium cover. Such a cover is taken as giving a cut-off at 0.4 eV.


under neutron irradiation

n, α ?


Cadmium is readily available in foil and sheet form from about 0.04cm thick. It is a very soft metal with a dull mottled appearance and the surface usually has imperfections. When used as a converter foil the surface should be polished until it is smooth and flat. It is readily fabricated by all of thecommercial techniques. It oxidises very slightly in air and does not react with boiling water.

Table 1.8 shows that 98% of the natural material. consists of six isotopes, Cd110, Cd 111 , Cd 112, Cd 113, Cd 114 and Cd 116 of which only the Cd113 ( n;γ ) and Cd114 has a large cross section.

This reaction produced a gamma ray of 9 MeV which is e radiation mainly responsible for the blackening of film when cadmium is used as a converter foil. When cadmium is used as a lining then this gamma ray will enter. The neutron beam and cause a reduction in contrast or fogging.


Table 1.8 Data on Cadmium [Refs. 24. 25. 25]

General

Appearance Density Atomic Nunaber Atomic Weight Atomic Density Natural Cross Sections:

Absorption Scattering Total

Isotopic Composition

Soft metal. dull surface 8.65 gem-a 48 112.41 4.64 x 1022 atomic cm-3

Microscopic 2,450 X 1 0-24cm2

7 x 10-24cm2

2.457 x 1 o-24cm2

Macroscopic 114 cm- 1

0.325 cm-1

114 cm-1


which only the Cd 113 ( n;γ ) and Cd114 has a large cross section.(?)

106 '1017 108 10'9 11110

1 11 1 111'1m 111.2.

11 .3 113 114

115m 116

0.:9

1.2.4

'1.2 .. 8

.24.0

7 .. 6

Sta'ble n, ·r~ 6.5h Stab:fe n. 'li',

'-"11 453d n. o Sta'ble n ·-;r

·y; Stab:fe n, ~

4'9m gamma 10 ... 245 Stab:l ~e n. ¥"

0.6

53.3Sh Beta ·1 . .,1 Stable n,.·t j

11 ,;1 65'0 111 '0' '1 ·. ,.,

2.4~3

Cd11:0s Cd11:o :Cd111 :Cd1111.m

·c· .· ·d11112 1 .1 ,

Cd117 I '

:Cd117.m


Europium is a rare earth material that is of interest as a reactor control material because, under neutron irradiation, it produces a series of high crosssection daughter-products which cause the initial cross section of 4300 barnsfor the natural material to fall to about 700 barns and then stay relativelyconstant even at high irradiation densities. It is available as europium oxideand as a dispersion (probably about 12%) in titanium. It oxidises rapidly in airat room temperature and should therefore be handled and stored under aninert atmosphere. Finely divided europium can ignite spontaneously in air. Italso reacts vigorously with cold water.


Table 1.9 shows that the natura l material consists of 47.8% Eu151 and52.12% Eu153 with two large cross sections for the Eu151 ( n,γ) Eu152 reactions.


1.

1 1


Isotopic Composition

Emission Transmutation Isotopic Abundance Half Number % Life Type Energy Cross Isotope

MeV Section Formed Barns

1 51 47.8 Stable n. r 3300 } Gd1s2

5900 11 4.0

152 9 .3h beta 1.9 Gd1s2 152 12.4y 152m 96m gamma 0.04,0 .09 Eu1SJ

153 52.12 n. G 390 Eu154 154 8.5y Beta 0.6,1.8 Gdls4

• Gd1ss 155 4 .96y Beta 0 .1,0.2 156 5 .2d Beta 0 .5,2.4 Gd1ss

157 1 5.1 5 h Beta 1.3 Gd1s1

158 46m Beta 2.4,3.4 Gd1sa

159 18.7m Beta 2 .6 Gd1ss

Gamma 0.045,1 .76 160 42s Beta 3.9 Gdlso

11 The three cross section values belong to different states of the 151 Eu nucleus.


1 .2.2.3 Defining the Inlet ApertureAs the resolution of the coll imator is a function of the inlet-aperture size it isimportant that this aperture should be well defined. This can be achieved byconstructing the inlet face of the collimator from a material which is opaque toneutrons, and especially those neutrons to which the converter foils are mostsensitive. This leads to the conclusion that the in let face should be madefrom layers of the converter foil materials that wil l be used. The mostcommon of these are dysprosium, gadolinium and indium, but furtherexamination of Fig. 6 shows that gadolinium has the highest cross sectionvalues on the left of the cadmium cut-off line and that indium predominates onthe right hand side, so that a combination of these two materials only wouldprobably provide an acceptable front face for the collimator.

The appropriate thicknesses can be calculated from:

A = 1- e -Σd (31 ) N=Noe-0.693t/t½ , I =Ioe-Σt

WhereA = attenuation factorΣ = macroscopic cross-section, cm-1 (Σ = σN)d = thickness, cm

The thickness should be chosen to make the attenuation factor equal to at least 0.95 and the cross-section values appropriate to about 5 eV should beused in order to ensure that all of the neutrons of energies to which the foilsare sensitive are absorbed in the inlet face.



1 .2.2.4 Divergence AngleIn travelling from the source aperture to the foil the neutrons in a divergentcollimator will follow a shorter path at the collimator centreline than at thewalls, assuming the target is a plane surface. As the neutron flux will vary withthe square of the collimator length then clearly the dose at the centre of thefoil will be greater than that at the edges. However, this only a problem for lowflux neutron radiography units with short and wide collimators, for it can beshown that it requires a divergence angle of 35º to produce a 10% exposuredifference between the centre and the edge of a foil.

I1

I2


1 .2.2.5 Geometric Enlargement and DiminutionThe divergence of the beam will cause the radiographic image to be generallylarger or smaller than the object size, depending on the relative sizes of theobject and the size of the inlet aperture to the collimator. The geometry isshown in Fig. 1 .7 D and the percentage change in the height of the object asrecorded at the image is

(32)


1 .2.3 The Converter Foils1 .2.3.1 Characteristics of Foil MaterialsThere are a considerable number of materials which have been tested asconverter-screen Materials [Ref. 4), but the foils that are now in general use are:

1. indium In and dysprosium Dy (& gold Au?) for the transfer method using X-ray films and

2. gadolinium Gd, for the direct method also using X-ray films and 3. boron B, and lithium Li foils for the track etch method.

The characteristics of these materials are as follows.



Abundance Em ission of Mode of Production Cross- Half-

Material Parent of Ac tive Isotope Section life M ax. Isotope, barns Type Energy. % MeV

0 lithium 7.4 L~ (n,a )H3 935 STABLE a 4.7 0 Boron 19 .5 B

10 j n.a)li

7 3,837 STABLE a 2.3 0 Rhodium 100 Rh

1 3(n,y1Rh

104 144 43 s {3 2.41 Rh,OJ(n,n )Rh103m 57 min X-ray 0 .02 Rh 103(n, 'Y) Rh 104m 11 4 .4 min {3 0.5


48.7 Ag, 09(n,y)Ag,,o 110 24.5 s {3 2.87 Ag 10:9( n, y)Ag 11Om 3 254 d {3 1.5

0 .66 0 Cadmium 12.3 Cd11J(n,y)Cd114 20,000 STABLE 'Y 9 T Indium 95.7 In 11s( n,y)ln 11s 45 14 s {3 3.3.

0 .44 In 115(n;y)ln116m 154 .54 min fj 1.0


1

1 1

1

Emrs::m..~ ..

1

1

1 1.1


- Characteristics of IndiumIndium metal is readily available as a foil from about 0.05 to 0.1 cm thick. It is a soft metal and has a dull mottled appearance when received, and thesurface usually has slight undulations. This can cause small densityvariations on the radiographs and so it i s advisable to polish the surface untilit is smooth and flat. Table 1.1 0 shows that the natural material consists of4.3% 113In and 95.7% 115In. The cross-section of the 113In isotope forconversion into 114 In is small, so very little 114 In is produced (and this only has a half life of 72 seconds) and only one of its two alternative states of 115Inhas a significant cross-section for conversion into 116In .

The 116In isotope is the most important for the transfer method because it emits a 1.0 MeV (maximum) beta ray when it decays to 116Sn. This decay has a halflife of 54 minutes, and so allows reasonable neutron exposure and transfer times.


An alternative mode of decay for 116In has a 14-second half-life and emits a 3.3 MeV beta. This will contribute to the exposure of the photographic film,but in practice the short half- life makes it difficult to use this radiation since itis emitted during the time while the foil is being transported from the exposureposition to the darkroom. Anyway, owing to its high energy it will be a lowresolution contribution [Ref. 27]. The isotopes of indium beyond 116In do not have appreciable cross-sections for neutron absorption and so they do not make any practical contribution to the process of neutron-radiographic image formation.


Table 1.10 Data on Indium [Refs. 24, 25, 26]

General Appearance Density Atomic Number Atomic Weight Atomic Density Natural Cross-Sections:

Absorption 1)

Scattering Total

Soft metal, dull surf ace 7.28 gem-a 49 114.82 3.82 x 1022 atoms em-a Microscopic 196 x 10-24 cm2

2.2 X 1 0-24 cm2

1 98.2 X 1 0-24 cm2

Macroscopic 7.75 cm-1

0.084 cm-1

7.564 cm-1


lso10pic Composition

Emission Transmutation Isotopic Abundance, Half-Number % Life Type Energy, Cross-Section, Isotope

MeV Barns formed 2)

113 4.3 Stable n, "f 3.9 ln114

7.5 In 114m

114 71.9 s beta 2.0 Sn114

114m 49.5 d gamma 0.192 115 95.7 Stable n.r 155 lnlls

115 6x101\ beta 0.5 65 lnlls

45 lnl1s

92 In 115m

115m 4.5h beta 0.8 116 54m beta, gamma 3) In 111

116 14s beta 3.3 Sn 116

116m 2.2s gamma 1.64 117 38m in beta 0.7 Sn117

117m 1.95n beta 1.8

1) This is a reaction. which determines the actiivity of the fo~ after irradiation and hence the neutron image. 2) m = metastable state

3) (J: 0.34 (1 .5°/o), 0.59 (11 °/ol. 0.87 (40°/o). 1.0 (49 °/o)

7: 0.138 (3 °/o), 0.147 (36°/o), 0.819 ('17 %). 1.09 (53 °/o), 1.293 (80 °/o).

1.508 (11 °/ol. 2.111 (20 °/o).


- Characteristics of DysprosiumDysprosium metal is obtainable as foil up to about 0.025 cm thick. It has a semi bright, smooth appearance and is sufficiently hard to withstand normal handling without incurring scratches or abrasions which will show on a radiographic image. Table 1.11 shows that naturally occurring dysprosium has seven stable isotopes, of which Dy156 and Dy158 can be neglected owing to their small abundance. The isotopes Dy160, Dy161 , Dy162 and Dy163 are not important for the transfer method since they do not form radioactive isotopes. The important isotope is Dy164 since this has a large cross-section for the formation of Dy165 , which is formed with a half life of 2.35 hrs and which decays into stable Ho 165 with the emission of 1 .3 MeV (maximum) beta rays. This transition has a 2.35 hr half-life. Dy 165 also has a metastable state, and this has an associated decay emission of 1 .0 MeV (maximum) beta rays, but the half-life is only 1.3 minutes and in practice this isotope does not contribute much to the photographic exposure. The isotopes of dysprosium beyond Dy165 do not have any great cross-section for neutron absorption and so do not make any practical contribution to the process of neutron-radiographic image formation.


Table 1.11 Data on Dysprosium. [Refs. 24, 25, 26, 50]

General

Appearance Density Atomic Number Atomic Weight Atomic Density

Natural Cross-Section: Absorption Scattering Total

Hard metal, semi-bright surface 8.56 g cm-3

66 162.51 3.17 x 1 022 atoms em -J

Microscopic 950 X 1 0-24 cm2

1 ()() X 1 0-24 cm2

1 050 X 1 0-24 cm2


3.17cm-1

33.4 cm-1


Isoto pi~c Compo:sition

l.sotopi1c 1

) .Abundance Number 1%

1516 15~7

158 1 !519 1160 116'1 116,2

0.09

.21.29 18 .. 88

.2~a.3

28.18

1:) ·m = ·me~tasta1 Je~ sta1te~.

·oecay Emission

:HalfUfe Type .Energy,

Stabl1e n, l 8 .. 1 h Stabl~e n~l' r 144.4d

.2 .361 h 1.3 m

""''t n, r n,, r

eV

2) 10~9

Tr.ansmutafon

Oross·Section, barns

33

96

!5:51

6 :'0 160 '.2151

2:) beta1 : 10.22 ~0. 1 ° /oJ1, :[t254 ~0 .. 103 ° / ol, 10~3 ~1 .3 '0/o), ·1.2 ~C1S 0 /o),, ·1 :3 (83 °/~o). gam·ma1 : 0.094 ('10 ·0/o.) .• 04279 ·(1 OJo), 0..3~61 (40 °/~o .).., 0.,7 ·1 (2 OJ~oJI, t.Or2 (8 0/~o.) .

1 sotope ) fo ·med


- Characteristics of GadoliniumGadolinium metal is obtained as a foil up to abciut 0.025 cm thick. It has a bright smooth appearance and it is strong enough, as a foil, to withstandnormal handling without incurring scratches or abrasions which will show on aradiographic image. Table 1.12 shows that naturally occurring gadolinium has six stable, and one very long lived, isotopes of which Gd152 a n d Gd154 can be neglected owing to their small abundance. The important isotopes are Gd1 55 and Gd157 since these have large cross sections.


Table1 .12 Data on Gadolinium [Refs. 24. 25. 26]

General

Appearance Density Atomic Number Atomic Weight Atomic Density

Natural Cross-Sections: Absorption Scattering Total

Hard met al. bright surface 7.95 g. cm-3

64 157.26 3.05 x 1022 atoms cm-3

Microscopic 46,000 X 1 0-'4cm2

-



Isot opic Composition

Emission Transmutation Isotopic Abundance Half Number % Life Type Energy Cross Isotope

MeV Section Formed Barns

152 0.2 1.1 x1 014y alpha 2.14 1100 Eu 149

153 241 .6d gamma 0.097. 0.103

154 2.2 Stable n.'f 85 Gd1ss

155 14.9 Stable n.~ 61.000 Gd1ss

156 20.6 Stable n, 'I' 1.5 Gd1s1

157 15.7 Stable n, r 254.000 Gd15B

158 24.7 Stable n. 't 2.5 Gd1s9

15 9 18.56h Beta 0.9 Tb159

160 21.7 Stable n, lS' 0 .77 Gd1s1

161 3 .6m Beta 1.6 Tb1s1

162 8.2m Beta 1 .0 Tb1s2


1 .2.3.2 Converter- Foil Thickness and Speed.The choice of the converter-foil thickness is a compromise between:

(a)a thin foil for high resolution and(b)a thick foil for short exposure times and sufficient rigidity for handling.

The choice is l i m ited by the th ickn ess of foi ls available and by the difficulties of handling extra-thin foils.


Fig. 1.18 Film Foil Geometry.

EMULSION

I (

BETA RAY EMITTED HERE

BETA RAY EMERGES

RANGE OF FROM FOIL PARTICLE HERE

,t--~

BETA RAY FINALLY ABSORBED HERE

/ Ui / ~~~~-

b

PHOTOGRAPH IC FILM

\ \ d

CONVERTER FOIL


Fig. 1.18 Film Foil Geometry.

μ

EMULSION

I (

BETA RAY EMITTED HERE

BETA RAY EMERGES

RANGE OF FROM FOIL PARTICLE HERE

,t--~

BETA RAY FINALLY ABSORBED HERE

/ Ui / ,...._~~......._-

b

PHOTOGRAPH IC FILM

\ \ d

CONVERTER FOIL


Fig. 1.18 is a sketch of the film and foil combination and shows the path of aparticle which has been emitted from an atom in the centre of the foil. For anyparticular direction of emission towards the film, a simple theoretical relationship would be:

Where:Ui = inherent unsharpness, emμ = angle of emission, degreesb = distance from film surface to emulsion, cmd =foil thickne ss, em.

μ


So theoretically the unsharpness is directly proportional to the film and foildimensions, and the smaller these are the smaller the unsharpness becomes.The activity on the foil per unit thickness is given by

S/d = ΣФ(1- e -λT) (34) S/d =σNФ - σNФe -λT

Where:S = activity, disintegrations s-1cm-3

Σ = macroscopic cross-secti on, cm-1, (Σ = σN)Ф = neutron flux, n∙cm-2 s-1

d = foil thickness, cmλ = decay constant 0.693/ττ = half-life of foil material, s.

I = Ioe-σN∙t, macroscopic cross section = σN N = ρ∙N’/A, Where: ρ=density, N’ Avogadro’s number (6.023 X 1023 atoms/gram-molecular weight) ; a is the total cross section in barns (cm2 ); and A is the gram atomic weight of material., A= gram atomic weight.


More on:

I = Io e-σN∙t, macroscopic cross section Σ = σN N = ρ ∙N’/A,Where: ρ=density, N’ Avogadro’s number (6.023 X 1023 atoms/gram-molecular weight) ; a is the total cross section in barns (cm2 ); and A is the gram atomic weight of material., A= gram atomic weight.

I = Ioe-σρ ∙N’∙t/A

Mass a bsorption coefficient, is denoted by the symbol μmAnd is related to the macroscopic cross section by the μm relationship:

μm = Σ/ρ

I = Ioe


Now, ignoring self-shielding, equation (34)

S/d = ΣФ(1- e -λT)S = d∙ΣФ(1- e -λT)

shows that for a particular irradiation time and neutron flux. the activity will increase in direct proportion to the foil thickness, so the thicker the foil the greater will be the exposure of the film for any total neutron dose to the foil.

Unfortunately these simple theoretical concepts only give a limited explanation of the observed phenomenaa, and some of the reasons are as follows. Taking dysprosium and indium as examples we can say that as the foil thickness is increased it approaches the maximum range of the β decay radiation and so it becomes more difficult for this β radiation to escape from the foil. Conversely as the escape path gets longer (i.e. the foil gets thicker) the particle loses energy and is more likely to be at the most effective energy for film blackening (~100 keV [Ref. 27] for beta particles). So the optimum foil thickness for film blackening is dependent upon the range (?) and energy of the emitted particles.


Berger [Ref. 23] gave a constant neutron exposure to converter foils of varying thickness placed in front and behind the film (direct technique). Theseresults were plotted and he found that initia lly the density increased as thefoil thickness increased, but then in levelled out or fell. The optimum speedcombination was taken to be the foil thickness and combination (back, frontetc.) that gave the highest density. The results, expressed as relativeconverter foil speeds, are g iven in Table 1.13. Because the highest speeds a re given by the double converter foil technique, Berger assumed that singlefoils would be used where improved resolution was required. Thus therelative speed g iven for single foils in Table 1.13 is based on a compromisebetween speed and resolution. Berger's resolution data is given in Table 1.14.It should be noted that the relative speeds of the direct and transfertechniques given in Table 1.13 cannot be compared as there is insufficientdata in Berger's paper to make a reliable comparison.


For the transfer data the foils were exposed to give a constant film density after a 3 half-life transfer time. The effects of varying the irradiation and transfer time is d iscussed later i n this section. All of the data in Table 1.13was obtained with a monochromatic thermal neutron beam, but Berger states that this shows a reasonable correlation with similar data taken from a reactor beam containing significant intensities of neutrons outside the thermal energy region.

When considering the relative speed of the foils used in the t ra nsfer process it m ust be remembered t h at the process of expos i n g a meta l foil to a neutron beam and then using the decay activity of the foil to expose aphotographic film is conceptually one of indirectly util ising the mass energy ofthe neutrons via the secondary radiation in order to convert silver halide tometallic silver with in the photographic emulsion of the film. The foil may beregarded as a container for the mass energy of the neutron - a container withan exponential profile that fills less a nd less rapidly as it approaches its stateof maximum capacity. This container is then 'emptied into the photographicfilm', again at a n exponential rate, with the rate of transfer getting less andless as the container becomes empty.


Table 1.13 Relative Speed of Converter Foils 1)

Foil Thickness Converter Technique JJ.m Relative Foil Speed

Front Back

Rh/Gd Direct 250 50 5.3 Rh/Rh Dtrect 250 250 4.7 Gd/Gd Direct 25 50 3.7 In/In Direct 500 750 3.7 Dy/Dy Direct 150 250 3.7 Cd/Cd Direct 250 500 3.3 Ag/Ag Direct 450 450 2.7 Dy Direct 250 2.5 Gd Direct 25 2.4 Cd Direct 250 2.2 Rh Direct 250 2.1 In Direct 500 1. 7 Au Direct 375 1.2 Au/Au Direct 150 250 1.0

Dy Trans fer 250 16.4 In Transfer 50 11.2 Au Transfer 75 1

t ) The relative speed for the direct and transfer methods are not comparable.


Table 1.14 Resolut ion Characterist ics of Con•!erter Foils

Converter Technique Foil Thickness Resolution • Exposure for Foil (J.(m) (J.tm) AA Film D=1.5

Front Back -2 n em

Gd Direct 12.5 30 2 X 108

Cd Direct 75 30 2. 7 X 108

Rd Direct 75 50 3.1 X 108

In Direct 125 50 4.1 X 1 08

Dy Direct 250 ~0 1.4 X 108

Ag Direct 125 50 - 90 3.9 X 108

Rh/Rh Direct 250 250 50 - 90 7. 7 X 1 07

In/In Direct 500 750 90 108

Ag/ Ag Direct 450 450 90 1.4 X 108

Cd/Cd Direct 250 500 500 1. 1 X 1 08

Gd/Gd Direct 6.25 50 30 8.6 X 107

Rh/Gd Direct 250 50 30 7 X 107

Au Transfer 75 30 4.3 X 109

In Transfer 50 50 2.9 X 10 9

Dy Transfer 250 50 2.6 X 1 08

* ) The minimum resolvable separation between 500 JJ1Tl diameter holes in a 500 JJ1Tl thick cadmium test piece.


T he question of exposure and transfer time is therefore one of determining what fraction of the maximum foil activity is to be induced onto the foil andwhat fraction of this activity is to be transferred to the film. The product of these two ractions will determine the total fraction that is transferred. and thus the total exposure. Figure 1.19 shows that when indium and dysprosium foils of the same thickness are irradiated in a thermal- neutron beam then: (a) when irradiated to saturation and transferred to infinity ( i. e. > 3 half- lives ineach case) the dysprosium is about 5 times faster than the indium (b) whenirradiated for up to 3 hours and transferred to infinity the dysprosiumis 2-4 times faster than indium. But when the irradiation and transfer times areequal it can be shown that because of the more rapid energy transfer duringdecay, the indium is faster for exposures below about 0.5 hours, after whichthe dysprosium is progressively faster, rising to about x3 at about a 3-hourexposure and transfer time. A method of determining the exposure a ndtransfer times is given in Appendix 1 .3.


Fig. 1.19 Build-Up of Activity i n Indium and Dysprosium of the same Thickness.

20

... 'e " !., 18

.!!! .., )(

3 LL

~ z 12 ::I cr ~ >~

> ;::: Q I <

SATURATION

DYSPROSIUM

SATURATION

OL-------~~------~---------i---------L--~ 8 12 18 4

IRRADIATION TIME,h

Charlie C

hong/ Fion Zhang

Fig. 1.19 Build-Up of Activity i n Indium and Dysprosium of the same Thickness.

I

I


1 .2.3.3 Film and Foil ResolutionBerger [Ref. 29] a lso studied the resolution capabilities of foils by judging thesmallest observable space between closely pitched holes, and the valuesgiven in Table 1.1 4 are those foil thicknesses below which this methodshows little or no gain in resolution. The resolution test piece used was 0.05em thick cadmium plate with a line of 0.05 cm diameter holes at varyingseparation. The use of this data to make comparisons between theresolutions effectiveness of various foils should be made with care becauseresolution is dependent upon contrast and object size [Ref. 44] and the hole-pacing method is a practical way of defining resolution rather than anabsolute method. It should be noted that Berger [Ref. 29] also used agadolinium test piece of 55 μm thickness with which a hole spacing of 10μm was resolved by a 12.5 μm thick gadolinium converter foil. Berger also tested the dependance of the results on film grain size and concluded that it was not influencing the results obtained.


1 .2.3.4 The Mounting of Converter FoilsMost converter foils used in neutron radiography a re between 0.0025 cm and 0.05 cm in thickness, and are, typically, of th e ordero f 200 to 600 cm2 in area . It is important that these foils remain flat and undamaged so that good contact between the film and the foil is achieved over the whole surface of the foil. The thicker foils will withstand normal handling, but, whilst methods of foil stiffening should be avoided if possible, when very thin metal foils re used they will require the support of a backing plate in order to withstand day-to-day handling.

Experience with indium and dysprosium indicates that these foils need to beabout 0.08 cm and 0.012 cm thick respectively for use without such backing.Below these thicknesses the foils should be attached to 0.15 cm thickaluminium plate of at least 99.5% purity. The adhesive used should be as thin as possible as it will usually contain hydrogen, which will scatter the neutrons. A suitable adhesive is photographic mounting tissue. This is applied with a hot iron and so produces a flat, wrinklefree, mounting. For details of the mounting technique see the manufacturer's literature.


Hot Charcoal Iron


1 .2.3.5 Enrichment of Converter FoilsFor high resolution and short exposure times the h igher the foil activity thebetter will be the results. One possible means of effectively increasing thisactivity is to make the foil more sensitive to the neutrons by enriching thoseisotopes which are the most effective for absorbing the neutrons and converting them to film-blackening radiation. Unfortunately the enrichmentprocess requires special equipment and is likely to be expensive, but it hasbeen reported [ Ref. 31 ] that boron, dysprosium and gadolinium foils havebeen enriched.


1 .3 A PPLICATIONS OF NEUTRON RADIOGRAPHYThe listing of reports which describe the applications found up to 1977 hasbeen admirably carried out by John Barton in his edited a nd indexedcompilation of Neutron Radiography Newsletters, Numbers 1 - 15, (available form the American Society for Non-Destructive Testing, 3200 R iverside Drive, Columbus, Ohio, 43221 ), and by the contributers to ( Ref. 5 1 ] . The following survey gives a general overview of the present situation .


1 .3.1 Nuclear ApplicationsProbably the biggest nuclear use is the examination of experimental fuel pins.The transfer and track-etch methods make such radiography possible, andconsiderable data can be obtained on cracking, slumping, swel ling, etc. By very careful techniques the dimensional changes of the fuel can be measuredand then translated into volume changes. The use of this type of application isclosely followed by general examination of all types of irradiation experimentsfor any type of failure that can be detected by visual observation. A ‘marker’technique has been developed whereby the swelling of a pressurised tube in an irradiation experiment (see Figure 1.20 ) is followed by a pair of plungers which are marked by small washers of dysprosium. This material has a large neutron cross-section and the marker consequently shows on a radiograph as a fine, high contrast line. The distance between two such lines is directlyrelated to the diameter of the tube, and by comparing this with the distancebetween two other fixed markers of known separation, also within theexperiment, the seperation of the measuring markers can be gaugedprecisely. This method measures the growth of the tube to + 25 μm.


Fig. 1.20 Cross-Section of Pressure-Tube Rig.

STA INLESS STEEL TIE ROD

Zr THIMBLE

0

SPRING

Zr PLUNGER

Zr TUBE

0

0

0.010 . DISC OF DYSPROSIUM


Nuclear reactors a re controlled by inserting highly neutron-absorbing materials into the pile. As irradiation proceeds the atoms of this material undergo transmutation, causing a marked change in neutron-attenuation cross-section. The rate of depletion of such control materials is of considerable interest to reactor operators as this determines the life of the control absorber and hence its planned replacement. The 'burn-up' (used up of initial material that undergo transmutation, 113Cd (n,γ) 114Cd, where the 2 isotopes have markedly different σ) of control absorbers can be detected by taking regular neutron radiographs and measuring the size of the depleted areas as is shown by Figure 1.21 .


Fig. 1.21 Neutron Radiographs showing Burn-up of Cadmium in Vertical Control Rods.

STAINLESS STEEL I

TUBES I

I l . I

WELD

I

I

TOP EDGE OF NOSE SECTION

LOWER EDGE OF CADMIUM

l I

STAINLESS STEEL NOSE SECTION

BEFORE IRRADIATION

DEPLETED CADMIUM

AFTER IRRADIATION


When neutron shields a re built around nuclear installations it is necessary tocheck their integrity. A typical example is the inspection of a resin filled shieldplug where neutron radiography is used to check that the resin has flowed into all the extremities of the volume to be filled. A neutron radiograph will distinguish between the isotopes of many materials since these often have very different neutron cross-sections. For example U235 has a thermal-neutron cross-section of 100.5 barns whilst the cross-section of U238 is 2.7barns. Such d ifferences are readily detectable and allow experimental fuelelements to be checked for rogue fuel pellets.

Neutron radiography has been in use by many workers for the quantitivemeasurement of hydrides in zirconium -hydride. This is a nuclear problemassociated with water reactors in which a corrosion reaction occurs between the water and the zirconium to produce zirconium hydride.


The detection technique is non-destructive and provides a two dimensional survey of the hydride concentration in the a rea under examination. A commercial neutron radiography service offers the detection of hydrogen in zircaloy to a sensivity of 3 ppm-cm.

It must be made clear that this method only detects a material of high neutronattenuation cross section, and it is not able to label an individual element. When detecting hydrogen is zirconium-hydride the experimenter knows that the parent material is pure zirconium and that only high neutron-attenuation cross-section material is present, namely hydrogen.


Fig. 1.22 Characteristic Curves for some Film - Foil Combinations.

(!) 0 Ll..

w

4.0

~3.0 Ill 4(

~ en ffi 2 .0 c :E ~

Ll..

1.0

I I I I

CAYSTALEX/0.0025 em Oy

10- 10-1

EXPOSURE UNIT, (EU)


1.3.2 Industrial ApplicationsHydrogen has a large thermal-neutron cross-section and many of the mostwidely used applications of neutron radiography involve its detection. Rubberand plastic materials have many hydrogen atoms in their molecular structureand so rubber seals, plastic insulation, etc. are easily detected in sealedassemblies. Explosives are also rich in hydrogen and the presence of voids,blockages etc. In ordnance components 军械部件 can be seen (see Figure 1 .23). Quantitative measurements of hydrogen have been made to determineabsorbed hydrogen in getters and a commercial neutron radiography serviceoffers the detection of hydrogen in zircaloy to a sensivity of 3 ppm.cm.


Fig. 1.23 Neutron Radiograph of Explosive Detonators (Magn. ca. 5 x)


Brazing and soldering meterials are good subjects. The braze usually contains silver and boron and the flux also conta ins boron, so both of thosecan be detected by neutrons. This often makes it possible to detect a dry jointby the presence of excess flux, and the correct flow and penetration of thebraze into the joint can be seen from the presence of the boron (see Fig.1.24).Turbine blades contain small cooling passages through the length of theblade and neutron radiography has been used to establish the thickness ofmetal round the passag es prior to machining to outer surface of the bladeand to identify materials causing blockages in the passages. Aircraft engineparts have been inspected for the presence of solidified oil and grease inlubrication holes and passages.


Fig. 1.24 A B razed Joint between Two Concentric Cylinders.


Racing-car wheels are made of magnesium alloy, and epoxy resins are used in their construction. A combination of ultrasonic methods and neutron radiography has been used to inspect the resin joints. Helicopter blades have been constructed by bonding carbon fibers to steel, and the lay-up of the fibers has been inspected through the steel by imaging the resin used in the bond. Printed circuits have been constructed with epoxy resins between layers of copper and neutron radiography has been used to detect voids in the resin.

Soldered joints sometimes exhibit poor electrical characteristics due to contamination within the joint. Several such joints have been neutron radiographed and a contaminant, probably boron, has been detected. High pressure hose has a metal braid wrapped about a rubber tube and NR has been used to examine the rubber through the metal (steel) braid. Laminations of various forms are widely used throughout industry and many of these use epoxy resins as the adhesive. Neutron radiography has been used to examine bonded wooden aircraft floors, aluminium honey comb sections for aircraft structures etc. Friction welding has been used to join stainless steel and aluminium tubes in which the weld is formed at a conical joint. Inclusions and poor bonding have been detected in such joints.


Run-out on deep drilled holes occur, and sometimes this can only be detected by neutron radiography. Run-out has been measured on deep-drilled molybdenum bars using water as a contrast agent in the hole. Other contrast agents which have been used are parafin, alcohol, gadolinium oxide, and boron fluoride.

Undoubtable the most impressive industrial application has been the coldneutron radiography of a running aircraft gas turbine engine in order to establish the dynamic distribution of the lubricationg oil throughout the oil-passages within the engine. This type of examination is claimed to lead to significant reductions in the time to develop new aircraft engines.


1.3.3 Biomedical ApplicationsWhilst a number of experiments have been performed in the field of application there have not been any that have shown significant advantageover other methods. This is principal ly because the neutron has a greaterbiological effect (to a patient) than photons per unit of absorbed dose andbecause the required exposures are too high. Figure 1 .26 shows thestructure of grass and leaves and is a simple i llustration of a biologicalspecimen.


F ig. 1. 25 X- Radiograph and Neutron Radiograph of a Cigarette Lighter.

Neutron RadiographX-Radiograph

PETROL IN COTTON WOOL

BRAZE

FLINT

WICK


Fig. 1.26 Neutron Radiograph of Grass and Leaves.


1 .3.4 Other ApplicationsFigure 1.25 shows a cigarette lighter and illustrates how the hydrogen in thepetrol is more readily detected by neutron than by X-rays. The flint and thebraze metal show-up well and the fibre sea ling washer is clearly seen, again due to the hydrogen content. Perhaps of greater interest are thearchaeological applications in which a saxon shield boss was examined andinformation obtained on the metal joining techniques that had been used.Examination of a Roman spearhead showed that a type of wiped lead jointhad been used between the head and the shaft.


Neutron Radiography in Archeology.

Kugel

Charlie C

hong/ Fion Zhang http://w

ww

.usatoday.com/story/new

s/2015/02/23/mum

mified-m

onk-inside-buddha-statue/23908879/


APPENDIX 1.1NEUTRON RADIOGRAPHIC TERMINOLOGY


absorption coefficient: related to the rate of change in the intensity of beamof radiation as it passes through matter.

absorption cross the probability expressed in barns, that a neutron willsection: be totally a bsorbed by the atomic nucleus.

activation: the process of causing a substance to become artificially radioactive by subjecting it to bombardment by neutrons or other particles.

attenuation: the loss of power suffered by radiation as it passes through matter.

attenuation coefficient (μn) : the interaction probability of neutrons per unit path length (cm-1) It is the same quantity as the macroscopic cross- ection.(μn = Σ = σN where σ = microscopic cross section in barns 10-24 cm2)

Barns: unit of area for measuring the cross-section of nuclei (with probability of interaction with nuetron σtotal = σs +σa) (1 barn = 10-24 cm2)


BPI: Beam Purity Indicator, a device for measuring the composition of the beam used in neutron radiography.

Cadmium ratio: ratio of the activity induced by the neutron beam in a bare gold foil to that induced when the foil is covered with cadmium.

Cold neutron: see thermal neutrons.

Collimator: device for obtaining a neutron beam of small angular spread.

Collimator ratio: also called L/D-ratio, where L is the collimator length and D is the characteristic entrance diameter (≠ source size) of the collimator.


contrast agent: a material added to a component to enhance details byselective absorption of the incident radiation. (In, Cd, Boron, Dy, Cd?)

contrast capability:the smallest inclusion, thickness change or densitychange that can be perceived on the radiographic film, expressed as apercentage. (analogous to photon radiography IQI sensitivity?)

conversion screen: a lso called converter, a material placed in contact withthe radiographic film, that absorbs neutrons and emits ionising radiationthereby exposing the film. (categorized into direct and transfer screens)

cassette: a light-tight device for holding film or conversion screens and film in close contact during exposure.

cross section: the apparent cross sectional area of the nucleus ascalculated on the basis of the probability of occurence of a reaction by collision with a particle (total σtotal; scattering σs and or absorption σabs) . Itdoes not necessarily coincide with the geometrical cross-sectional arearπR2 It is given in units of area (barns).


direct exposure imaging: in the direct exposure imaging method the conversion screen and image recorder are simultaneously exposed to the neutron beam.

direct imaging method: method by which the neutron radiation is recordedimmediately after passing through the material being tested.

electron volt: the kinetic energy gained by an electron after passingthrough a potential difference of one volt.

epithermal neutrons: neutrons which have energies in excess of the energy associated with thermal agitation. Neutrons which have speeds and energies intermediate between fast and thermal neutrons (i.e.between about 0.1 and 100 eV). (0.1~10KeV?)

Charlie Chong/ Fion Zhang ASNT NON DESTRUCTIVE TESTING HANDBOOK Neutron Radiography

TABLE I. NeLib'ons Crasslrfled AA;cordl g teO En rmr·

1i m

CoJd

Fas

RelarJVlS u;

tCommenu

· ate11a s. possess high uos!!.-sec.t100S a1111~se en gl , wf"li h ecrease It! ra sparency of most 'Ci cenal.s our a 'SO incre,ase ffi iet1 . or de ernon A part1~~;:ular ad~ rrti'lge IS lh duced scatte in ma all\ r1 n r ·es ~ I w th~ Bragg ruta

or fa~ neutrons unW the a\lefdge energy o the ne~ ron m- helll'l J 11e.utrons. p!D\11de gDOll d lS:Ctl lnatory .a il ity

e11: m.Jclej exhibit strong absorp~ion ar eri:ltlcs at well-defined enetgies called reSDllam: absorptions.. N uons irl hese spec:~:ffr: energy rt~nges are ~ rerre!l m as.

sconaflce 11'€Utrons and rovu::J~ excFtr t df'){:rimina 1(1(1 or parillcurar materi Is b 1

vvork1 g ar e-nergies or reiD"l¢lnc; _ Grea er uansm 1 ssro ilnd les~ scatter ocr r in s.ample.S conra1 n lng ma(er.ia.l~ su n .:. r1ch d r c'l'!Ct.ai ruel m;;!l ria s.

Fasr neu rons provide gocd pen r 011. Good p nr so rc:: s of ras ne:u lions ,;:ue ailable_ the aw~ en end of t specnum fa rlfU ron red og r.cphy l'l'I.3Y be

~ble o pwfOil'm m y lm:p c:no pefformed wit:n herrn:::~l ru!t.J~rons. cult w 1 l

1c ec:hnrque. or mat :rial d s n m~ on ocrur:s. however, t.Jecause rhe cross-_ i{lfls terld to b@ small

o.oo ev ro 1 o1 eV

Les.s than 0.01 ev

0.0 V o OJ . V

>20M V

Charlie Chong/ Fion Zhang https://en.wikipedia.org/wiki/Neutron_temperature

Ne tron1 energy· range name·s[1]

!Neutron energy En~ergy range

O.D-0.025 eV ~Co lld neutrons

0.025 eV The rtmal ne·utro ns

0.025-D.4 eV

0.4-0.6· ev· ~Cadmiium neutrons

1- 110 eV Slow neutrons

10-300 ev· Resonance neutrons

300 eV- ·1 MeV llntermed'ate neutrons

1- 20 MeV Fast neutrons

> 20 MleV Ultrafast neutrons


More Reading: (epithermal neutron)General Properties of Key Fissile and Breeder NucleiKey nuclear data for the nuclides 232Th, 233U, 235U, 238U, 239Pu and 241Pu is provided in the table below. To complement the table, a brief description of practice for defining and quantifying thermal and epithermal neutron fluxes is first given.

In the literature, the ‘cadmium cut-off’ energy defines the boundary between terming a neutron to be in the thermal or epithermal energy regime. The definition arises from 113Cd, which has a particularly large neutron absorption coefficient below neutron energies of 0.55 eV, above this energy the probability of neutron absorption rapidly reduces, see the Figure to the right. The probability of capturing thermal neutrons is sometimes quoted at the velocity 2200 ms-1. This corresponds to the mode neutron velocity for a Maxwellian energy distribution at 20 °C (E = 0.0253 eV).

(0.55 eV or 0.0253 eV?)

http://thorea.wikia.com/wiki/Thermal,_Epithermal_and_Fast_Neutron_Spectra


Keypoints: The definition arises from 113Cd, which has a particularly large neutron

absorption coefficient below neutron energies of 0.55 eV, above this energy the probability of neutron absorption rapidly reduces, see the Figure to the right.

The probability of capturing thermal neutrons is sometimes quoted at the velocity 2200 ms-1. This corresponds to the mode neutron velocity for a Maxwellian energy distribution at 20 °C (E = 0.0253 eV).

The 'Cadmium cut-off' energy is defined to be at 0.0253 eV.



Neutron capture cross section for 113Cd for a range of neutron energies. The 'Cadmium cut-off' energy is defined to be at 0.0253 eV.


106

-(/) c 11.....

ctS

104 ..0 -c 0

+=i 0 Q) (/)

I

102 (/) (/)

0 11..... 0 Q) 11.....

:J ....... 10° a.

ctS 0 c 0 11..... ....... :::J Q)

10-2 z

10-4 ~--~--------~--------~--------~--------~--------~----~ 10-10 10-8 10-6 10-4 10-2 10°

Neutron energy (MeV)


The (n,γ) reaction rate can be described as,

r = Фthσo + Фepi Io (α)for thermal and epithermal neutrons combined.

Where the rate of neutron absorption is, r (s-1); Фth and Фepi are the conventional thermal and epithermal neutron fluxes (cm-2), respectively; σo is the neutron capture cross-section at 2200 ms-1 (barns); Io is the infinite dilution resonance integral (cm2); and α is the epithermal flux distribution parameter [Verhijke 2000].

………………….read further (http://thorea.wikia.com/wiki/Thermal,_Epithermal_and_Fast_Neutron_Spectra)



Maxwell–Boltzmann distributionIn physics, particularly statistical mechanics, the Maxwell–Boltzmann distribution or Maxwell speed distribution describes particle speeds in idealized gases where the particles move freely inside a stationary container without interacting with one another, except for very brief collisions in which they exchange energy and momentum with each other or with their thermal environment. Particle in this context refers to gaseous atoms or molecules, and the system of particles is assumed to have reached thermodynamic equilibrium.[1]

The distribution is a probability distribution for the speed of a particle within the gas - the magnitude of its velocity. This probability distribution indicates which speeds are more likely: a particle will have a speed selected randomly from the distribution, and is more likely to be within one range of speeds than another. The distribution depends on the temperature of the system and the mass of the particle.[2]

https://en.wikipedia.org/wiki/Maxwell%E2%80%93Boltzmann_distribution


The Maxwell–Boltzmann distribution applies to the classical ideal gas, whichis an idealization of real gases. In real gases, there are various effects (e.g., van der Waals interactions, vortical flow, relativistic speed limits, and quantum exchange interactions) that make their speed distribution sometimes very different from the Maxwell–Boltzmann form. However, rarefied gases at ordinary temperatures behave very nearly like an ideal gas and the Maxwell speed distribution is an excellent approximation for such gases. Thus, it forms the basis of the kinetic theory of gases, which provides a simplified explanation of many fundamental gaseous properties, including pressure and diffusion.[3]

The distribution is named after James Clerk Maxwell and Ludwig Boltzmann. While the distribution was first derived by Maxwell in 1860 on basic grounds,[4] Boltzmann later carried out significant investigations into the physical origins of this distribution.



Probability density function


0"5 . ooooooooo•~·-••o••••••ooo••-~ooo••••••ooo

a=l a~2

a=S

0.4 . ' '

~ ~ ~ R ~~~a a a- 0 a .... a- a 0 a a a a a-- 0 a a a- ... - 0 - a a a--- 0 a a a--- .. a a a a-- 0 0 - a a--- 0 a

• 0 0

•• a I I •••• It I I I W'" •• t: I I I :. ••• a I I I ••• '1: I I I ••••• I I I ••• I> I • I •••• a I I I ••• a I I I

0 2 • • • • • I • • • • • • ~- • • • • • • • • • • • • • • • • o.. • • • • • • • • • • • • • • • • "' • • • • • • • • • • • • • • •-i ..

0.1 l!'!r I I I •• "' II I!' f I I ... I!' I!' I ol I •• "' .. I!' I I I." I!' PI It

' '

5 10 15 20 v


Cumulative distribution function of Maxwell-Boltzmann distribution


0.8

0 .. 6

0 104 - · -- -- - ... ·

0.2

5

-.-

10 X

::0 -

.,. -

a=l a=2 a=5

20


The Maxwell–Boltzmann distribution The speed distribution for the molecules of an ideal gas is given by From this function can be calculated several characteristic molecular speeds

and such things as what fraction of the molecules have speeds over a certain value at a given temperature. It is involved in many rates of phenomena.

http://hyperphysics.phy-astr.gsu.edu/hbase/kinetic/kintem.html


cadmium cutoff 镉截止能(nucleonics) The neutron energy, approximately 0.3 electronvolt, below which cadmium has a high neutron absorption cross section but above which this cross section falls off sharply.

http://encyclopedia2.thefreedictionary.com/cadmium+cutoff


filtered neutron beam: neutron beam after passing through a uniform layer of material for the purpose of absorbing specific parts of the neutron spectrum. (neutron beam tailoring?)

flash neutron radiography: technique using a neutron source that yields a very high flux neutron beam during a very short time.

gamma ray: electromagnetic radiation having its origin in an atomicnucleus. (as compares with X-ray with origin from the orbiting electrons)

gamma ray fogging: increase in the optical density of a radiograph causedby the gamma radiation emitted by the neutron source, by the facility itself, by the object being tested or by a combination of them.


geometric resolution: smallest theoretical size of discontinuity that can bedetected according to the geometry of the neutron radiography facility. Itdepends on the L/D ratio of the collimator, the neutron source to objectdistance and the object to converter distance.

half value layer: the thickness of an absorbing material required to reduce the intensity of a beam of incident radiation to one-half of its original intensity.

imaging method: method by which the neutron radiation passing through amaterial is recorded.

image quality indicator: a device or combination of devices whose image or images give a measurement of the neutron radiographic image quality.


indirect exposure imaging: in the indirect exposure methode, only a gamma insensitive conversion screen is exposed to the neutron beam. After exposure the conversion screen is placed in contact with the image recorder.

Indirect imaging method: also called “activation transfer method”. Method by which the neutron radiation passing through the material being tested is used to activate a foil of a suitable material. This activated foil is subsequently placed in contact with a medium capable of recording the radiation emitted as the activity of the foil decays.

in-motion neutron radiography: neutron radiography on moving objects by means of techniques allowing multiple exposures of short duration. (with the utilization of flash neutron radiography?)

IQI: Image Quality Indicator.


L/D ratio: one measure of the resolution capability of a neutronradiographic system. It is the ratio of the distance from the entrance aperture to the image plane (L) to the diameter of the entrance aperture (D).

linear absorption coefficient: the fractional decrease in radiation beam intensity per unit of distance (cm-1 ). (μn = Σ = σN ?)

mass absorption coefficient: the fractional decrease in radiation beam intensity per unit of surface density (cm2 g-1 ) (Σ /ρ ?) – to check!

moderator: a material used to slow down fast neutrons. Neutrons are slowed down when they collide with atoms of light element such as hydrogen deuterium, beryllium and carbon.

neutron: a neutral elementary particle having an atomic mass of 1 . In the free state outside of the nucleus, the neutrons is unstable having a half-life of approximately 12 minutes.


More reading onMass absorption coefficientIt should be noted that there are several ways of expressing the cross section of a material, i.e.:a) microscopic cross section, cm2

b) macroscopic cross section, cm-1

c) mass absorption coefficient, cm2 g-1

The first is the basic unit and, as stated earlier, is measured in barns. Thesecond is the product Νσ; this is given the symbol Σ and it is the total targetarea for a given neutron interaction presented by a cubic centimeter of material.

Thus, for this case equation (3) (The ratio between these two neutron fluxes is called the transmission, i.e. Transmission) can be rewritten:

I/Io = e –σNx = e –Σx


and it can be seen that the use of the macroscopic cross section Σ simplifiesthe use of this equation. The third form, the mass absorption coefficient, is denoted by the symbol μm and is related to the macroscopic cross section by the relationship: (dimension check)

μm (cm2∙g-1) = Σ (cm-1) / ρ (g∙cm-3)

Note also:μ = N σ = σ ∙(ρN’/A)μm = N σ / ρ = (ρN’/A)∙σ / ρ = ρ ∙ σ ∙(ρN’/A) = σ∙(N’/A)


neutron radiography: a process of making a picture of the internal details of an object by the selective absorption of a neutron beam by the object.

Neutron to gamma ratio: ratio of neutron fiux and gamma dose rate at theimage plane of a neutron radiography facility (n∙cm-2∙mR-1).

Recording medium: a film or detector that converts radiation into a visibleimage.

saturation effect: occurring at the activity transfer technique where theactivity of the converter, induced by the neutron radiation, increases exponentially to a saturation value where activation and decay are in equilibrium. (saturation effect of converter ≠ film)


scatter factor: ratio of scattered and non-scattered neutrons that contribute to the resulting visible image.

scattered neutrons: neutrons that have undergone a scattering collision but still contribute to the resulting visible image. These neutrons may be (1) facility scattered or (2) object scattered neutrons.

Sensitivity indicator: (SI) = a device for indicating the sensitivity of detailvisible on a neutron radiography. It is determined by the smallest observable hole and thickness of the corresponding absorber in the indicator. (IQI?)

sensitivity level: the level determined by the smallest standarddiscontinuity in any given sensitivity indicator observable in the radiographic film. Levels are defined by identification of type of indicator, size of defect and the absorber thickness on which the discontinuity isobserved.


sub-thermal neutrons:neutrons having energies below 0.01 eV.

thermal neutrons: neutrons of very slow speed and consequently of lowenergy. Their energy is of the same order as the thermal energy of the atoms or molecules of the substance through which they are passing; i.e. About 0.025 electron-volts which is equivalent to an average velocity of about 2200 metres per second. Thermal neutrons are responsible for numerous types of nuclear reactions, including nuclear fission.(0.01~0.3eV)

total cross section: the sum of the absorption and scattering cross sections.

track-etch imaging: method by which neutron radiation passing through thematerial being tested is used to cause damage tracks in a dielectric medium. The damage tracks are made visble by chemical etching.

vacuum cassette: a light-tight device having a flexible entrance windowwhich operated under a vacuum, holds the film and conversion screen in intimate contact during exposure.


APPENDIX 1.2 THERMAL CROSS SECTIONS OF THE ELEMENTS AND SOME MATERIALS.

NucleN or Macroscopic Cross Section. cm'"'1 Absorption Coefficient c~ g-1

Atomic Element or DensitV Molecules Numbe·r Materials gm cm-3 cm-l • •o2• Absorption Scattering Total Absorption Scattering Total

t H 8.99 X t tr 5.37 X t a-l 1.7 X t iJ• 2 X t 0-<l 2 X t 1J"3 O. t 96 23 23.2 2 He t .78x l o-" 2.68 x w·• 2 X 10-7 2 .t X t a-l 2.t x t cr 0 O.t O.t 3 u 0.534 0.046 3.29 0.065 3.85 6. t O.t 6.2 4 Be 1.84 O.t 23 1.24 X 1 o-3 0 .865 0.865 0.0067 0 .47 0.48 5 B 2.45 O.t 36 t 03 0.549 t 04.6 42. t 0 .2 42.3 6 c 1.60 0.080 2.6 X t o-' 0.385 0.385 0.000t 8 0 .24 0.24 7 N 1.25 X t o-3 5.38x to-6 9.9 X to-' 5 X to-' 6 X to-' 0.08t 0.43 0 .5t 8 0 1.43 X t o-3 5.38 X t tr 2.t X 1 !T" 2 .t X t o-" <7xttr O.t 6 O.t 6 9 F t .7 X t iJ-3 5.39 X t tr 1<T1 2 X t o-" 2x t o-' <3xt o-' O.t 2 O.t 2

t O Ne 9.0x t o-' 2.69 X t tr 2.6 X t iJ' 6.2 x 1 cr 8.9 X t iJ' 0.0009 0 .072 0.073 1 t Na 0 .97 t 0.025 0.0 13 O. t 02 O.t 15 O.Ot 39 O.t O O.tt t 2 Mg 1.74 0.043 3 X t a-l O.t 55 O.t 58 0.00t 6 0.089 0.09t t 3 AI 2.7 0.060 1.4 X t a-l 8.4x 11J2 9.8 X t a-2 0.0053 0.031 0.036 14 Si 2.35 0 .050 7 X t iJol 8.9 X 1 IJ"2 9.6 X t a-2 0.0034 0.036 0.039 t 5 F 1.83 0.036 7 X t 0-<l O.t 77 O.t 84 0.0039 0.10 O.t O t 6 s 2.t 0.040 1.9 X t 1J2 4.3 X 1 IJ"2 6.2x t a-l 0.0098 0.021 0.03t t 7 Cl 3.21 X t a-l 5.45 X t tr 2 X t o-3 8 X t o-" 2.8x t o-3 0 .58 0 .27 0.85 t 8 A 1.78 X t a-l 2.68 X ttr 2.6 X t iJ' 3.9 X t tr 5.5x t tr 0.0099 0.02 O.Q3 19 K 0 .87 O.Ot 3 2.6 X t o-3 2 X t o-3 4 .7 X t o-3 0.03 t 8 0.023 0.053 20 Ca 1.54 0 .020 ta4 7 X t o-2 8x t a-l 0.0066 0.046 0.053 2 t Sc 2.5 0.034 0 .787 0.804 1.59 0 .32 0.4 0.72 22 T J 4.5 0 .057 0.328 0.226 0.555 0.072 0.05 O.t 2 23 v 5.96 0 .07 t 0.359 0 .352 0.7 1t 0.059 0.059 O. tt8 24 Cr 6.92 0.080 0.238 0.247 0.485 0.036 0.035 0.07t 25 Mn 7.42 0 .08t t .04 O.t 8t t .22 O.t 4 t 0.025 0.166 26 Fe 7.86 0.085 0.2 15 0.933 1.t 5 0.0282 0 .12 0.15 27 Co 8.7t 0.089 3.37 0 .637 4.00 0.38 0 .07t 0.45 28 Ni 8.75 0.090 0.42 1 .6 2.02 0.047 O.t 8 0 .23 29 Cu 8.94 0.085 0.3 t 3 0.6t 1 0.924 0.036 0.068 O.t 04 30 Zn 7. t 4 0.066 7 X 1()"2 0.237 0.307 O.O t02 0.033 0.043 3t Ga 5.90 0.051 O.t 42 0.204 0.346 0 .024t 0.036 0.060


Nuclei or Macroscopic Cross Section. cm-1 Absorption Coefficient an2 g050::1 A tome Elemenr or Density Molecules Number Materials gm cm-3 cm-.3 c 1024 Absorption Scauering Total Absorption Scaltering Total

32 Ge 5.46 0.045 0.105 0.134 0.239 0.02030 0.025 0.045 33 As 3.70 0.030 0.189 0.277 0.466 0.035 0.048 0.083 34 Se 4.5 0.034 0.431 0.403 0.835 0.089 0.084 0.173 35 Br 3.12 0.024 0.155 0.41 0.296 0 .052 0.045 0.097 36 Kr 3. 71 X 1()"3 2.67 X 10"3 7.3 X 1 0""' 1 x 1 o-" 1.4x1<J3 0.22 0.052 0.27 37 Rb 1.53 0 .011 8 X 104 0 .13 0.138 0.0051 0.039 0.044 38 Sr 2.54 0 .018 2 X 1()2 0 .175 0.195 0.0083 0.069 0 .077 39 v 5.51 0.037 4.8 X 1 ()"0 0.112 0.160 0.0089 0.0203 0.0292 40 Zr 6.44 0.043 8 X 1()3 0338 0.346 0.0012 0.053 0.054 41 Nb 8.4 0.055 6 t 1()2 0.272 0.333 0.0075 0.032 0.039 42 Mo 10.2 0.064 0 .16 0.448 0.608 0.017 0.044 0.061 43 Tc 0.13 44 Ru 12.1 0.072 0 .179 0.436 0.615 0.016 0.035 0.05 45 Rh 12.4 0.073 11 0 .36 11.4 0 .88 0.029 0.91 46 Pd 12.2 0.069 0.551 0.248 0 .799 0.045 0.020 0.065 47 Ag 10.5 0.059 3.63 0.325 3.98 0.349 0.033 0382 48 Cd 8.65 0.046 154 0.325 154.3 13.5 0.037 13.5 49 In 7.28 0.038 7.26 8.4 X 1 <J2 7.34 1.02 0.011 1.03 50 Sn 7.29 0.037 2 X 10"2 0 .132 0 .152 0.0031 0.020 0.023 51 Sb 6.22 0 .031 0.182 0.142 0.324 0.0282 0.0213 0.0495 52 Te 6.02 0.028 0.133 0.1 48 0.281 0.0221 0.024 0.046 53 I 4.94 0.024 0 .157 0.084 0.241 0.034 O.DI7 0.051 54 Ke 5.85 X 1()3 2.68 X~~ 1()3 1.2 X 1 0""' 1.1 x1<r' 0.34 0.020 0.36 55 Cs 1.87 8.5 X 1 <J' 0.246 0.1 70 0.416 0.128 0 .032 0 .160

56 Ba 3.5 0.015 1.8 X 1()2 0.123 0 .141 0.0052 0 .035 0.040

57 La 6.15 0.027 0.239 0.403 0.642 0.0386 0.040 0.079

58 Ge 6.9 0 .030 2.1 x10"" 0.262 0.283 0.0031 0 .012 0 .015

59 Pr 6.48 0 .028 0.324 0.1 16 0.44 0.048 0 .017 0.065 60 Nd 6.96 0.029 1.33 0.464 1.79 0.208 0.056 0.274 61 Pm 0.249 0.249

62 Sm 7.75 0.032 255 0 .155 255.2 23 23 63 Eu 5.22 0.021 90.5 0.166 90 .6 1 7 0.033 17 64 Cd 7.95 0.031 18.83 18.83 1 78 1 78 65 Tb 8.33 0.032 1.39 1.39 0 .18 0.18


Nuclei or Macroscopic Cross Section. cm-1 Absorption Coefficient cm2 g-1

Atomic Element or Density MolecLies Number Materials gm cm-3 cm-3 a 1024 Absorption Scatterilg Total Absorption Scattering Total

66 Dy 8.56 0 .032 34.9 3.17 38.1 3.5 0 .37 3.9 67 Ho 8.76 0 .032 2.05 2.05 0.23 0.23 68 Er 4 .77 0.0 17 5.44 0 .495 5.98 0.58 0.58 69 Tm 9.35 0.033 3.93 3.93 0.39 0.025 0.42 70 Yb 7.0 1 0.024 0.878 0.293 1.1 7 0 .128 0.04 0.17 7 1 Lu 9.74 0 .036 3.62 3.62 0.38 0.38 72 HI 13.3 0.045 4.71 0.359 5.07 0.341 0 .27 0.6 1 73 T a 16.6 0.055 1. 18 0 .277 1.46 0.070 0.017 0.087 74 w 18.9 0.062 2.21 0.3 16 1.53 0.063 0.016 0.079 75 Re 29.15 0 .095 5.58 0 .93 6.51 0.28 0.045 0.33 76 Cs 22.5 0.073 1.05 0 .783 1.83 0.049 0.049 0.098 77 lr 22.4 0.078 30.2 30.2 1.37 1.37 78 Pt 2 1.4 0.066 0 .535 0 .660 1.1 9 0.027 0.031 0.058 79 Au 19.3 0 .060 5.79 0 .55 6.34 0.302 0.028 0.330 80 Hg 13.6 0.041 14. 7 0.814 15.5 1 .1 2 0 .06 1 .1 8 8 1 Te 11 .9 0.035 0.115 0.489 0.604 0.0 10 0.041 0.051 82 Pb 11. 1 0.033 6 X 1!1" 0 .363 0.369 4.9x1~ 0.032 0.032 83 8j 9.7 0 .028 1!1" 0.264 0.265 9.8x1 tr5 0.026 0.026 84 Po 9.24 0 .027

86 Rv 9.73 x w-3 2.64 x ur• 87 Fr 88 Ra 5 0.0 13 0.266 0.266 0.053 0.053 89 Ac 1.35 1.35 90 Th 11.5 0.205 0 .366 0.571 0.019 0 .032 0.051 91 p 1 5.4 10.4 10.4 0.675 0.675 92 u 1 8. 7 0.047 0.364 0.397 0. 761 0.0193 0.0209 0.0402 93 Np 0.432 0.432 94 Pu 19.74 0.049 57 0.478 57.5 2.593 0.024 2.6 17 95 Am 96 Cm

97 8k 98 C1 99 E

100 Fu


Nuclei or Maeto6CGpiC Cr066 Section. em- • Absorption Coefficient cm:f g-1

At ernie Element or Den6ity Molecules Number Materials gm cm- 3 cm-3 x 1024 Absorption Scattering Total Absorption Scattering Total

8eo 2.96 0 .071 7.3 X 10"' 0 .501 0.501 2.47xlo-' 0.169 0.169 co, 1.98 X 1CJ"'' 2 .71 X Ia-" 2.4x1 o-' 2.4xHT .. 0. 121 0 .12 1 o,o 1.1 0 0.033 3.3 x 1 crs 0.449 0.449 3x1 a-" 0.408 0.408

Dvl03 7.81 0.0 13 27.7 2.7 30.4 3.547 0.346 3.893 EulO. 7.42 0 .0 13 111 0.383 111 .4 14.96 0.052 15.012 H:!D 0.997 0 .033 0.022 3.45 3.47 0.022 3.46 3.482 l if 2.29 0 .053 3.75 0.282 4.032 1.638 0.123 1. 761 uc 13.63 0.033 0.227 0.491 0.718 0.167 0 .36 0.527 uc2 11.68 0.027 0.185 0 .537 0.7 17 0 .158 0.46 0.6 18 UN 14.22 0.035 0.237 0.692 0.929 0 .017 0.49 0.507 uo. 10.8 0.024 0.169 0.372 0.542 0.157 0.034 0.191 ZrH 5.61 0.036 0.026 2.50 2.526 0.005 0.446 0.451

Alumnlum (99,5%) 2. 71 0.0 15 0.084 0.099 5.5x1 0"'1 0.031 0.036 8utyi Rubber 0.92 0.01 1.45 1.46 0.01 1.58 1.59 Concrete (8aoytes) 3.3 0.02 0.1 0.2 6x1 o-3 0.03 0.036 Concrete (Standard) 2.33 6xlo-' 0.14 0.15 2.6x1 o-> 0.06 0.064 lnconel 8.43 0.366 1.225 1.591 0.04 0.15 0.19 lead ag de 4.8 0.8 0.1 7 Monel 8.83 0.398 1.258 1 .656 0.045 0.142 0.187 Nylon 1 .11 0.04 2.56 2.6 0.036 2.31 2.34 Oil 0 .88 0.03 3.07 3.1 0.034 3.49 3.52 Pil'affin 0.9 0 .05 3 .1 3.15 0.056 3,44 3.5 Perspex 1.18 2.8 2.37 Polythene 0.94 4.0 1.25 Polystytane 1.07 0.02 2.1 2.12 0 .0 19 1.96 1.98 Rc,c cxpJoGlvc t .82 2.9 1.59

Stainless Staal 316 7.92 0 .27 0.85 \ .\2 0 .034 0.107 0.\4 TeHon 2.17 lo-' 0.3 0.3 4.6x1 o-' 0.138 0.138 Water 1.0 3.45 3.45


APPENDIX 1 .3Irradiation and Transfer Times for the Indirect MethodThe formation of a neutron radiographic image on a film, using the transfer

method will depend upon :a) the neutron fluxb) the rate at which the foil material becomes activated and its saturated

activityc) the rate at which this activity decaysd) the fraction of the activation particles that escape from the foil, ande) the sensitivity of the film to these activation particles.

Assuming that a beta-emitting foil is being used, then these factors may be expressed in an equation for the exposure ( Σ ) of the film, as follows [ Ref. 32] :


where N =number density of stable isotopes, atoms cm-3

u = m icroscopic cross-section, cm2

.A = 0 ,69/-r, where -r is the half-l ife of foil material, s ¢ =thermal neutron flux, n cm-2 s-1

. T = exposure time, s t1 = interval from end of irradiation time to commencement of f ilm

exposure, s t

2 = interva1 from end of irradiation to completion of film exposure, s

P =film density, mg cm-2

P: =density of material required to stop beta particles, mg cm-2

d =foil thickness, em (assuming d < R) •

R = range of betas in foil material, em.


The constants in this equation are the foil material, the foil th ickness, and thespeed of the photographic film. The variable factors are:

Expression(2) is called the integrated exposure, and is the product of the neutron flux, the irradiation unit, and the transfer unit. The product of the lasttwo terms (the quantities in the brackets) is called the exposure unit, or EU,and is the fraction of the maximum possible integrated exposure that isachieved during a particular combination of irradiation and transfer times. Fig.1.22 shows the characteristic curves for several X-ray films, in which thedensity above fog* is plotted against the exposure unit (EU) for a neutron fluxof 1 06 n∙cm-2 s-1.

*) Density above fog = log10 lo/l , where for a beam of light falling on a film Io = incident intensity and I = transmitted intensity. Fog is extraneous density or 'noise'.

(2)


The required film density can be determined from a knowledge of the imagedensity required and the neutron attenuation through the sample. This isdescribed by the transmission equation

( 3)

where

B = e - l:x I

8 = fraction of neutrons transmitted I = macroscopic cross-section of sample, cm-1

x = sample thickness, em.


APPENDIX 1 .5Calculation of the Cross Section of CompoundsAssuming that the property of the nuclear species is unaffected byconsiderations of the molecular or crystal structure in which it resides (thisassumption is usua lly acceptable for scoping calculation of the type below)then the macroscopic cross section for the compound can be calculated fromthe summation of the macroscopic cross sections of each nuclear species:

Σc = Σ Niσi ( 1 ) (Σ = Nσ, for pure element)

where Σc = macroscopic cross section of compound, cm-1

σi = microscopic cross section for ith nuclear species, cm2


N (number of nuclei cm-3) = ρNa/A

Where:ρ = density Na = Avogadro's number 6.023 x 1023

A = gram atomic weight or molecular weight

Σc = μn = Σ (ρNa/A)i ∙σi

Σc = μn = ρ Na/A (v1σ1+…..Vi σi)


Example of cellulose acetate polyethylene (CH2)N:

p = 0.91 g/cm3

N = 6.023 X 1023 atoms/g-molM= 14.0268 gvC = 1uC = 4.803 X 10-24 cm2

vH= 2uH = 38.332 X 10-24 cm2


μn = 0.91x 6.023x 1023 / 14.0268 (1 x 4.803 + 2 x 38.332) x 10-24

= 3.183 cm-1


Example of cellulose acetateC6H7O2(OH)(C2H3O2)2

ρ = 1.3 g cm-3

NA = 6.023 X 1023

A = 246.22vC = 10, σC = 4.803vH = 14, σH = 38.332vO = 7, σO = 4.2

i = 3, C’ = 10, H’= 14, O’ = 7


μn = (1.3 x 6.023 x 1023)/246.22 x [(10x4.803 + 14x38.332 + 7x 4.2) x 10-24]

= 1.3 x 6.023 x1023/246.22(614)x10-24 = 1.953 cm-1

Charlie C

hong/ Fion Zhang

TABL£ 9. Thermal ~n un.,ar ~nuatlon C""'"d""a Using A~ag" Sc.attrrlng and 2200 m/s Absorption Cross s.ctlom tor th• Naturally Oocurrlng El.,•nts

Element Cross s..ctlon !barns) Unear Attenuation Atomic No. Symbol Sc.att•rlng Absorption • Coi!Mclent lcm- •)

I H, J.I!.O 0.332 gu 2 He 0.8 gu 3 ~· 1.< 71.0 ).)6 4 Be 7.1 0.010 0.88 l 8 ••• 7ll 99 6 c 4.8 0.003 O.S41 7 No 10 1.88 , .. a 0, 4.2 0 , ... 9 F, 3.9 0.01 aas

10 Ne 2.4 2.8 , .. 11 Na 4.0 O.S:l6 O.lll 12 Mg 3.6 0.063 O.ll8 13 AI 1.4 0.23 0.098< 14 Si 1.7 0. 16 0.096S IS p s.o 0.20 0.184 16 s 1.1 O.S2 O.OS91 17 Cl, 16.0 33.6 , .. 18 A L5 0.66 gu 19 It l.l 2.07 0.047 20 Ca 3.2 0.4< 0.08<9 21 Sc 24.0 24.0 1.609 22 Ti <.0 5.8 O.SSl 23 v 5.0 4.9 0.698 24 Cr 3.0 ).I O.S09 2l MD 23 1).2 1.224 26 Fe 11.0 2.53 1.149 27 c. 7.0 )7.0 4.01 28 Ni 17.S 4.8 2.04 29 Cu 7.2 3.8 0.931 30 Zn 3.6 1.1 0.309 31 Oa 4.0 2.8 0.347 32 Ge 3.0 2.45 0.242 33 AJ 6.0 <.3 0.475 )4 s. 11.0 12.3 0.856 )~ Br 6.0 6.1 0.26) 36 Itt 7.2 31 gu 37 Rb B 0.73 0.0613 38 Sr 10.0 1.21 0.201 39 y 8.0 1.31 0.3 .. 7 40 Zr 8.0 0.180 0.346 41 Nb s.o LIS 0.341 42 Mo 7.0 2.7 0.621 4) Tc s.o 22.0 denshy u.nknowD

Charlie C

hong/ Fion Zhang

TABLE 9. Continued

E:le~nt Crou section (l>llmsJ Unear Attenuation

Atomic No. 5ymbol Scattering Absorpdon• Coefnclrnl (em-• I

44 Ru 6.0 2.56 D.61S 45 llh 5.0 156.0 11 .70 46 Pd 3.6 8.0 0.746 ., ... , 6.0 ~).0 •.05 48 Cd 1.0 2450 11).5 49 In 2.2 196 1.60 50 Sn 4.0 0.625 0.111 51 Sb 4.3 5.7 0.370 52 Te M 4.7 0.2116 53 I 3.6 1.0 0.248 54 Xe 4.3 pa 55 c. 7.0 29.0 0.306 56 Ba 8.1 1.2 0.143 57 La 9.3 9.3 0.496 58 c. 2.8 0.73 0.102 59 Pr 4.0 11.6 0.434 60 Nd 16.0 46.0 1.785 61 Pm 60 denlily wWtOWll 62 Sm 5600 173.0 63 Eu 8.0 4300 89.1 64 Od 46 000 1405.0 65 Th 46.0 1.455 66 Dy 100 950 33.3 67 Ho 65 2.08 68 Er 7.8 (cob) 11) 5.94 69 Tm 1 127 4.46 70 Yb 12 )1 1.195 71 Lu 112 3.75 72 Hf 8 105 5.o7 73 Ta 5 O.oJ 0.278 74 w 5 19.2 1.53 75 Re 14 86 6.64 76 Os 15.2 (cob) 15.3 2.17 11 lr 440 30.9 78 .. 10 u I.:IAA 19 Au 9.3 98.8 6.39 w H& 20 JW 16.3 81 n 14 3.4 O.Wl 12 Pb II 0.17 0.368 8) Bi 9 0.034 0.258 84 Po 85 AI density unknown 116 Rll pa 17 Fr density unknown 88 Ra 130 1.69 89 Ac 510 dcuity u..o.k:nown. 90 Tb 1500 (fission) 44.1 91 Pa 1500 (r-.ssion) 60.4 92 u 7.68 (includes fLSslon) 0.788 93 Np 900 (t'i$$i0ft) deOJh)' u.ak.oowa 94 Pu tW (ttS:SIOO) 7.96

• All cross-stedorl values ¥t most proe»olo \IOJ'-'rl witn no Implied ~

FROM lit. A. MOitR:tS. LO$ AlAMOS NATK>NAL LA80RAfOFf"f. C AMERICAN SOOm f'Off TUnNGNID MATIIItW.S. RIPRWTf0\IIT1')f PERMISSION.


Further Reading: Interesting Formhttp://www.ncnr.nist.gov/instruments/bt1/neutron.html


Other Readings:

http://novascientific.com/neutron.html https://en.wikipedia.org/wiki/Neutron_source


Peach – 我爱桃子


Good Luck

Charlie Chong/ Fion Zhanghttps://www.yumpu.com/en/browse/user/charliechong

understanding neutron radiography reading vii nrhb part 1 of 2

Engineering

charlie chong fion zhang

motion radiography fluoroscopy

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