Understanding neutron radiography reading 2016 i nrt a

Understanding Neutron Radiography More on Neutron, Reading 2016-1–NRT My ASNT Level III, Pre-Exam Preparatory Self Study Notes 29 July 2016

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NRT Source

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NRT Source

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Neutron Source

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The Magical Book of Neutron Radiography

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数字签名者：Fion Zhang DN：cn=Fion Zhang, o=Technical, ou=Academic, email=fion_zhang@ qq.com, c=CN 日期：2016.07.31 17:23:18 +08'00' Charlie Chong/ Fion Zhang

ASNT Certification Guide NDT Level III / PdM Level III NR - Neutron Radiographic Testing Length: 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

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3. Techniques/Calibrations

• Electron emission radiography

• Blocking and filtering

• Micro-radiography

• Multifilm technique

• Laminography (tomography)

• Enlargement and projection

• Control of diffraction effects

• Stereoradiography

• Panoramic exposures

• Triangulation methods

• Gaging

• Autoradiography

• Real time imaging

• Flash Radiography

• Image analysis techniques

• In-motion radiography • Fluoroscopy

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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 Number NDT Handbook, Third Edition: Volume 4, Radiographic Testing 144 ASM Handbook Vol. 17, NDE and QC 105 Charlie Chong/ Fion Zhang

Fion Zhang at Copenhagen Harbor 30th July 2016

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SME- Subject Matter Expert http://cn.bing.com/videos/search?q=Walter+Lewin&FORM=HDRSC3 https://www.youtube.com/channel/UCiEHVhv0SBMpP75JbzJShqw

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闭门练功

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http://greekhouseoffonts.com/

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neutron

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http://www.nuclear-power.net/nuclear-power/reactor-physics/atomic-nuclear-physics/fundamental-particles/neutron/

1.0 A neutron is one of the subatomic particle that make up matter. In the universe, neutrons are abundant, making up more than half of all visible matter. It has no electric charge and a rest mass equal to 1.67493 × 10-27 kg— marginally greater than that of the proton but nearly 1839 times greater than that of the electron. The neutron has a mean square radius of about 0.8×10-15 m, or 0.8 fm, and it is a spin-½ fermion.

The neutrons exist in the nuclei of typical atoms, along with their positively charged counterparts, the protons. Neutrons and protons, commonly called nucleons, are bound together in the atomic nucleus, where they account for 99.9 percent of the atom’s mass. Research in high-energy particle physics in the 20th century revealed that neither the neutron nor the proton is not the smallest building block of matter. Protons and neutrons have also their structure. Inside the protons and neutrons, we find true elementary particles called quarks. Within the nucleus, protons and neutrons are bound together through the strong force, a fundamental interaction that governs the behaviour of the quarks that make up the individual protons and neutrons.

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The quark structure of the neutron. The color assignment of individual quarks is arbitrary, but all three colors must be present. Forces between quarks are mediated by gluons.

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A nuclear stability is determined by the competition between two fundamental interactions. Protons and neutrons are attracted each other via strong force.

On the other hand protons repel each other via the electric force due to their positive charge. Therefore neutrons within the nucleus act somewhat like nuclear glue, neutrons attract each other and protons , which helps offset the electrical repulsion between protons. There are only certain combinations of neutrons and protons, which forms stable nuclei. For example, the most common nuclide of the common chemical element lead (Pb) has 82 protons and 126 neutrons.

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Nuclear binding energy curve.

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Because of the strength of the nuclear force at short distances, the nuclear binding energy (the energy required to disassemble a nucleus of an atom into its component parts) of nucleons is more than seven orders of magnitude larger than the electromagnetic energy binding electrons in atoms. Nuclear reactions (such as nuclear fission or nuclear fusion) therefore have an energy density that is more than 10 000 000x that of chemical reactions.

Knowledge of the behaviour and properties of neutrons is essential to the production of nuclear power. Shortly after the neutron was discovered in 1932, it was quickly realized that neutrons might act to form a nuclear chain reaction.

When nuclear fission was discovered in 1938, it became clear that, if a fission reaction produced free neutrons, each of these neutrons might cause further fission reaction in a cascade known as a chain reaction. Knowledge of crosssections (the key parameter representing probability of interaction between a neutron and a nucleus) became crucial for design of reactor cores and the first nuclear weapon (Trinity, 1945).

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Trinity, 1945

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Trinity, 1945

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2.0 Discovery of the Neutron The story of the discovery of the neutron and its properties is central to the extraordinary developments in atomic physics that occurred in the first half of the 20th century. The neutron was discovered in 1932 by the English physicist James Chadwick, but since the time of Ernest Rutherford it had been known that the atomic mass number A of nuclei is a bit more than twice the atomic number Z for most atoms and that essentially all the mass of the atom is concentrated in the relatively tiny nucleus. The Rutherford’s model for the atom in 1911 claims that atoms have their mass and positive charge concentrated in a very small nucleus.

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The alpha particles emitted from polonium fell on certain light elements, specifically beryllium, an unusually penetrating radiation is produced.

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http://www.nuclear-power.net/nuclear-power/reactor-physics/atomic-nuclear-physics/fundamental-particles/neutron/

An experimental breakthrough came in 1930 with the observation by Bothe and Becker. They found that if the very energetic alpha particles emitted from polonium fell on certain light elements, specifically beryllium, boron, or lithium, an unusually penetrating radiation was produced. Since this radiation was not influenced by an electric field (neutrons have no charge), they presumed it was gamma rays (but much more penetrating). It was shown (Curie and Joliot) that when a paraffin target with this radiation is bombarded, it ejected protons with energy about 5.3 MeV. Paraffin is high in hydrogen content, hence offers a target dense with protons (since neutrons and protons have almost equal mass, protons scatter energetically from neutrons). These experimental results were difficult to interpret. James Chadwick was able to prove that the neutral particle could not be a photon by bombarding targets other than hydrogen, including nitrogen, oxygen, helium and argon. Not only were these inconsistent with photon emission on energy grounds, the cross-section for the interactions was orders of magnitude greater than that for Compton scattering by photons. In Rome, the young physicist Ettore Majorana suggested that the manner in which the new radiation interacted with protons required a new neutral particle.

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The task was that of determining the mass of this neutral particle. James Chadwick chose to bombard boron with alpha particles and analyze the interaction of the neutral particles with nitrogen. These particular targets were chosen partly because the masses of boron and nitrogen were well known.

Using kinematics, Chadwick was able to determine the velocity of the protons.

Then through conservation of momentum techniques, he was able to determine that the mass of the neutral radiation was almost exactly the same as that of a proton. In 1932, Chadwick proposed that the neutral particle was Rutherford’s neutron. In 1935, he was awarded the Nobel Prize for his discovery.

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Sir James Chadwick

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Handsome Ettore Majorana

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Chadwick’s neutron chamber containing parallel disks of radioactive polonium and beryllium. Radiation is emitted from an aluminum window at the chamber’s end.

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3.0 Structure of Neutron Neutrons and protons are classified as hadrons, subatomic particles that are subject to the strong force and as baryons since they are composed of three quarks. The neutron is a composite particle made of two down quarks with charge −⅓ e and one up quark with charge +⅔ e. Since the neutron has no net electric charge, it is not affected by eletric forces, but the neutron does have a slight distribution of electric charge within it. This results in non-zero magnetic moment (dipole moment) of the neutron. Therefore the neutron interacts also via electromagnetic interaction, but much weaker than the proton.

The mass of the neutron is 939.565 MeV/c2, whereas the mass of the three quarks is only about 12 MeV/c2 (only about 1% of the mass-energy of the neutron). Like the proton, most of mass (energy) of the neutron is in the form of the strong nuclear force energy (gluons). The quarks of the neutron are held together by gluons, the exchange particles for the strong nuclear force.

Gluons carry the color charge of the strong nuclear force.

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The quark structure of the neutron. The color assignment of individual quarks is arbitrary, but all three colors must be present. Forces between quarks are mediated by gluons

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Like the proton, most of mass (energy) of the neutron is in the form of the strong nuclear force energy (gluons). (99%)

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Two down quarks with charge −⅓ e and one up quark with charge +⅔ e.

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Properties Of Neutrons Key properties of neutrons are summarized below:  Mean square radius of a neutron is ~ 0.8 x 10-15m (0.8 fermi)  The mass of the neutron is 939.565 MeV/c2 (E=mc2)  Neutrons are ½ spin particles – fermionic statistics  Neutrons are neutral particles – no net electric charge.  Neutrons have non-zero magnetic moment.  Free neutrons (outside a nucleus) are unstable and decay via beta decay.

The decay of the neutron involves the weak interaction and is associated with a quark transformation (a down quark is converted to an up quark).  Mean lifetime of a free neutron is 882 seconds (i.e. half-life is 611 seconds ).  A natural neutron background of free neutrons exists everywhere on Earth and it is caused by muons produced in the atmosphere, where high energy cosmic rays collide with particles of Earth’s atmosphere.  Neutrons cannot directly cause ionization. Neutrons ionize matter only indirectly.

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 Neutrons can travel hundreds of feet in air without any interaction. Neutron radiation is highly penetrating.  Neutrons trigger the nuclear fission.  The fission process produces free neutrons (2 or 3).  Thermal or cold neutrons have the wavelengths similar to atomic spacing.

They can be used in neutron diffraction experiments to determine the atomic and/or magnetic structure of a material.

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It is known the fission neutrons are of importance in any chain-reacting system. Neutrons trigger the nuclear fission of some nuclei (235U, 238U or even 232Th). What is crucial the fission of such nuclei produces 2, 3 or more free neutrons.

But not all neutrons are released at the same time following fission. Even the nature of creation of these neutrons is different. From this point of view we usually divide the fission neutrons into two following groups:  Prompt Neutrons. Prompt neutrons are emitted directly from fission and they are emitted within very short time of about 10-14 second.  Delayed Neutrons. Delayed neutrons are emitted by neutron rich fission fragments that are called the delayed neutron precursors 先驱者 . These precursors usually undergo beta decay but a small fraction of them are excited enough to undergo neutron emission. The fact the neutron is produced via this type of decay and this happens orders of magnitude later compared to the emission of the prompt neutrons, plays an extremely important role in the control of the reactor.

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Beta Decay

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Table of key prompt and delayed neutrons characteristics. Thermal vs.

Fast Fission

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Neutron Production – Prompt Neutrons.

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Energy release from 235U fission.

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Delayed Neutrons While the most of the neutrons produced in fission are prompt neutrons, the delayed neutrons are of importance in the reactor control. In fact the presence of delayed neutrons is perhaps most important aspect of the fission process from the viewpoint of reactor control.

The term “delayed� in this context means, that the neutron is emitted with half-lifes, ranging from few milliseconds up to 55 s for the longest-lived precursor 87Br. These neutrons have to be distinguished from the prompt neutrons which are emitted immediately (on the order of 10-14 s) after a fission event from a neutron-rich nucleus. Despite the fact the amount of delayed neutrons is only on the order of tenths of percent of the total amount, the timescale in seconds plays the extremely important role.

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Delayed neutrons are traditionally represented by six delayed neutron groups, whose yields and decay constants (Îť) are obtained from nonlinear leastsquares fits to experimental measurements.

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Key Characteristics of Delay Neutron  The presence of delayed neutrons is perhaps most important aspect of the fission process from the viewpoint of reactor control.  Delayed neutrons are emitted by neutron rich fission fragments that are called the delayed neutron precursors.  These precursors usually undergo beta decay but a small fraction of them are excited enough to undergo neutron emission.  The emission of neutron happens orders of magnitude later compared to the emission of the prompt neutrons.  About 240 neutron emitters are known between 8He and 210Tl, about 75 of them are in the non-fission region.  In order to simplify reactor kinetic calculations it is suggested to group together the precursors based on their half-lives.  Therefore delayed neutrons are traditionally represented by six delayed neutron groups.

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 Neutrons can be produced also in (γ, n) reactions (especially in reactors with heavy water moderator) and therefore they are usually referred to as photoneutrons. Photoneutrons are usually treated no differently than regular delayed neutrons in the kinetic calculations.  The total yield of delayed neutrons per fission, vd, depends on:  Isotope, that is fissioned.  Energy of a neutron that induces fission.  Variation among individual group yields is much greater than variation among group periods.  In reactor kinetic calculations it is convenient to use relative units usually referred to as delayed neutron fraction (DNF).  At the steady state condition of criticality, with keff = 1, the delayed neutron fraction is equal to the precursor yield fraction β.  In LWRs the β decreases with fuel burnup. This is due to isotopic changes in the fuel.

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 Delayed neutrons have initial energy between 0.3 and 0.9 MeV with an average energy of 0.4 MeV.  Depending on the type of the reactor, and their spectrum, the delayed neutrons may be more (in thermal reactors) or less effective than prompt neutrons (in fast reactors). In order to include this effect into the reactor kinetic calculations the effective delayed neutron fraction – βeff must be defined.  The effective delayed neutron fraction is the product of the average delayed neutron fraction and the importance factor βeff = β . I.  The weighted delayed generation time is given by τ = ∑iτi . βi / β = 13.05 s, therefore the weighted decay constant λ = 1 / τ ≈ 0.08 s-1.  The mean generation time with delayed neutrons is about ~0.1 s, rather than ~10-5 as in section Prompt Neutron Lifetime, where the delayed neutrons were omitted.  Their presence completely changes the dynamic time response of a reactor to some reactivity change, making it controllable by control systems such as the control rods.

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Delay Neutron- Their presence completely changes the dynamic time response of a reactor to some reactivity change, making it controllable by control systems such as the control rods.

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Delay Neutron- Their presence completely changes the dynamic time response of a reactor to some reactivity change, making it controllable by control systems such as the control rods.

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Delayed neutrons originate from the radioactive decay of nuclei produced in fission and hence they are different for each fissile material. They are emitted by excited neutron rich fission fragments (so called the delayed neutron precursors) some appreciable time after the fission. How long afterward, is dependent on the half-life of the precursor, since the neutron emission itself occurs in a very short time.

The precursors usually undergo beta decay without any neutron emission but a small fraction of them (highly excited nuclei) can undergo the neutron emission instead of the gamma emission.

In addition current nuclear physics facilities can produce more neutron-rich isotopes that can emit multiple neutrons. Currently about 18 2n-emitters are experimentally confirmed [IAEA – INDC(NDS)-0599], but only two of them are also fission products.

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As an example, the isotope 87Br is the major component of the first group of precursor nuclei. This isotope has half-life of 55.6 seconds. It undergoes negative beta decay through its two main branches with emission of 2.6 MeV and 8 MeV beta particles. This decay leads to the formation of 87Kr* and 87Kr (ground state) and the 87Kr nucleus subsequently decays via two successive beta decays into the stable isotope 87Sr. But there is also one possible way for the 87Br nucleus to beta decay. The 87Br nucleus can beta decay into an excited state of the 87Kr nucleus at an energy of 5.5 MeV, which is larger than the binding energy of a neutron in the 87Kr nucleus. In this case, the 87Kr nucleus can undergo (with probability of 2.5%) a neutron emission leading to the formation of stable 87Kr isotope.

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Precursors of Delayed Neutrons

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According to the JEFF 3.1 database, about 240 n-emitters are known between 8He and 210Tl, about 75 of them are in the non-fission region.

Furthermore 18 2n-emitter, and only four 3n-emitters ( 11Li, 14Be, 17B, 31Na) are experimentally confirmed. These numbers are not certainly final. Since new IAEA Coordinated Research Project (CRP) on Beta-delayed neutron emission evaluation has been started in 2013, it is expected these numbers will change significantly.

As can be seen it was identified many precursor nuclei. Not all of them are fission products (about 75 of them in the non-fission region A<70), but there are also many precursor nuclei, that are in the fission region between A=70150. Their half-lives range between tenths of second (0.12 s) and tens of seconds (55.6 s), therefore their delayed neutrons appear with considerably differing delay times.

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Fission fragment yield for different nuclei. The most probable fragment masses are around mass 95 (Krypton) and 137 (Barium).

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Six Groups of Delayed Neutrons – Parameters Reactor-kinetic calculations with considering of such a number of initial conditions would be correct, but it also would be very complicated. Therefore G. R. Keepin and his co-workers suggested to group together the precursors based on their half-lives.

Therefore delayed neutrons are traditionally represented by six delayed neutron groups, whose yields and decay constants (λ) are obtained from nonlinear least-squares fits to experimental measurements. This model has following disadvantages:    

All constants for each group of precursors are empirical fits to the data.

They cannot be matched with decay constants of specific precursors.

These constants are different for each fissionable nuclide.

These constants change also with the neutron energy spectrum.

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Although this six group parameterization still satisfies the requirements of commercial organizations, a higher accuracy of the delayed neutron yields and a better energy resolution in the delayed neutron spectra is desired.

It was recognised that the half-lives in six-group structure do not accurately reproduce the asymptotic die-away time constants associated with the three longest-lived dominant precursors: 87Br, 137I and 88Br. This model may be insufficient especially in case of epithermal reactors, because virtually all delayed neutron activity measurements have been performed for fast or thermal-neutron-induced fission. In case of fast reactors, in which the nuclear fission of six fissionable isotopes of uranium and plutonium is important, the accuracy and energy resolution may play an important role.

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Source: Keepin, G. R., Physics of Nuclear Kinetics. Addison-Wesley, 1965.

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Source: Keepin, G. R., Physics of Nuclear Kinetics.

Addison-Wesley, 1965.

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Eight Groups of Delayed Neutrons â&#x20AC;&#x201C; Parameters In order to reduce discrepancies between measured and calculated values of the reactivity scale based on reactor kinetics (this discrepancy results in excessive conservatism in the design, because it must be covered by uncertainties during core design and safety analyses calculations), the NEA (Nuclear Energy Agency) and its NEANSC/WPEC Subgroup 6 recommends a new eight-group representation. It seems reasonable to increase the number of delayed neutron groups, because many studies have shown that most of delayed neutrons are produced approximately by twelve precursors that are common to all fissioning isotopes.

The eight-group representation uses a set of eight-group half-lives for all fissioning systems, with the half-lives adopted for the three longest-lived groups corresponding to the three dominant long-lived precursors: 87Br, 137I and 88Br (these precursors was separated into single groups).

See also: Delayed Neutron Data for the Major Actinidies, NEA/WPECâ&#x20AC;&#x201C;6.

OECD 2002.

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Delayed neutrons are traditionally represented by six delayed neutron groups, but a new eight-group representation is recommended. https://www.oecd-nea.org/science/wpec/volume6/volume6.pdf

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Photoneutrons In nuclear reactors the gamma radiation plays a significant role also in reactor kinetics and in a subcriticality control. Especially in nuclear reactors with D2O moderator (CANDU reactors) or with Be reflectors (some experimental reactors). Neutrons can be produced also in (Îł, n) reactions and therefore they are usually referred to as photoneutrons.

A high energy photon (gamma ray) can under certain conditions eject a neutron from a nucleus. It occurs when its energy exceeds the binding energy of the neutron in the nucleus. Most nuclei have binding energies in excess of 6 MeV, which is above the energy of most gamma rays from fission.

On the other hand there are few nuclei with sufficiently low binding energy to be of practical interest. These are: 2D, 9Be, 6Li, 7Li and 13C. As can be seen from the table the lowest threshold have 9Be with 1.666 MeV and 2D with 2.226 MeV.

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Nuclides with low photodisintegration threshold energies.

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In case of deuterium, neutrons can be produced by the interaction of gamma rays (with a minimum energy of 2.22 MeV) with deuterium:

Because gamma rays can be emitted by fission products with certain delays, and the process is very similar to that through which a “true” delayed neutron is emitted, photoneutrons are usually treated no differently than regular delayed neutrons in the kinetic calculations. Photoneutron precursors can be also grouped by their decay constant, similarly to “real” precursors. The table below shows the relative importance of source neutrons in CANDU reactors by showing the makeup of the full power flux.

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Despite the fact photoneutrons are of importance especially in CANDU reactors, deuterium nuclei are always present (~0.0156%) also in the light water of LWRs. Moreover the capture of neutrons in the hydrogen nucleus of the water molecules in the moderator yields small amounts of D2O. This enhances the heavy water concentration. Therefore also in LWRs kinetic calculations, photoneutrons from D2O are treated as additional groups of delayed neutrons having characteristic decay constants Îťj and effective group fractions.

After a nuclear reactor has been operated at full power for some time there will be a considerable build-up of gamma rays from the fission products. This high gamma flux from short-lived fission products will decrease rapidly after shutdown. In the long term the photoneutron source decreases with the decay of long-lived fission products that produce delayed high-energy gamma rays and the photoneutron source drops slowly, decreasing a little each day. The longest-lived fission product with gamma ray energy above the threshold is 140Ba with a half-life of 12.75 days.

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The amount of fission products present in the fuel elements depends on how long has been the reactor operated before shut-down and at which power level has been the reactor operated before shut-down. Photoneutrons are usually major source in a reactor and ensure sufficient neutron flux on source range detectors when reactor is subcritical in long term shutdown.

In comparison with fission neutrons, that make a self-sustaining chain reaction possible, delayed neutrons make reactor control possible and photoneutrons are of importance at low power operation.

http://www.nuclear-power.net/nuclear-power/fission/delayed-neutrons/#prettyPhoto

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Prompt Neutrons It is known the fission neutrons are of importance in any chain-reacting system. Neutrons trigger the nuclear fission of some nuclei (235U, 238U or even 232Th). What is crucial the fission of such nuclei produces 2, 3 or more free neutrons.

But not all neutrons are released at the same time following fission. Even the nature of creation of these neutrons is different. From this point of view we usually divide the fission neutrons into two following groups: Prompt Neutrons. Prompt neutrons are emitted directly from fission and they are emitted within very short time of about 10-14 second.

Prompt fission neutrons and prompt gamma rays are emitted when excited primary fission fragments release their energy to reach a more stable configuration. This happens shortly after the fission of the nucleus on two fission fragments. Studying prompt neutrons and gamma rays is of importance not only for deepening our understanding of the nuclear fission process, but also in core design calculations or in radiation shielding calculations.

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Most of the neutrons produced in fission are prompt neutrons. Usually more than 99 percent of the fission neutrons are the prompt neutrons, but the exact fraction is dependent on the nuclide to be fissioned and is also dependent on an incident neutron energy (usually increases with energy). For example a fission of 235U by thermal neutron yields 2.43 neutrons, of which 2.42 neutrons are the prompt neutrons and 0.01585 neutrons (0.01585/2.43=0.0065=Ă&#x;) are the delayed neutrons. The production of prompt neutrons slightly increase with incident neutron energy.

Prompt neutrons are emitted within 10-14 second. This is very important, because it is very very fast and causes very very fast response of the reactor power in case of prompt criticality. Therefore nuclear reactors must operate in a prompt subcritical, delayed critical condition. All power reactors are designed to operate in a delayed critical conditions and are provided with safety systems to prevent them from ever achieving prompt criticality.

It must be noted the response of the reactor on the reactivity insertion is determined by a neutron generation time (not by emission time), which is the average time from a prompt neutron emission to a capture that results only in fission.

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Table of key prompt and delayed neutrons characteristics.

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Key Characteristics of Prompt Neutrons  Prompt neutrons are emitted directly from fission and they are emitted within very short time of about 10-14 second.  Most of the neutrons produced in fission are prompt neutrons – about 99.9%.  For example a fission of 235U by thermal neutron yields 2.43 neutrons, of which 2.42 neutrons are prompt neutrons and 0.01585 neutrons are the delayed neutrons.  The production of prompt neutrons slightly increase with incident neutron energy.  Almost all prompt fission neutrons have energies between 0.1 MeV and 10 MeV.

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 The mean neutron energy is about 2 MeV. The most probable neutron energy is about 0.7 MeV.  In reactor design the prompt neutron lifetime (PNL) belongs to key neutron-physical characteristics of reactor core.  Its value depends especially on the type of the moderator and on the energy of the neutrons causing fission.  In an infinite reactor (without escape) prompt neutron lifetime is the sum of the slowing down time and the diffusion time.  In LWRs the PNL increases with the fuel burnup.  The typical prompt neutron lifetime in thermal reactors is on the order of 10-4 second.  The typical prompt neutron lifetime in fast reactors is on the order of 10-7 second.

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Prompt Gamma Rays With the prompt neutrons, prompt gamma rays are associated. Most of prompt gamma rays are emitted after prompt neutrons. The fission reaction releases approximately ~7 MeV in prompt gamma rays and additional ~7 MeV (for 235U) in delayed gamma rays. This is a significant portion of energy (~7 % of fission energy released) and it must be considered in many fields of reactor design or in the design of nuclear reactor shields. A particular challenge is the calculation of the gamma heat deposition in core baffles (reflectors), pressure vessels or excore detector channels.

In past the comparisons of various benchmarks experiments with calculated gamma heating showed a systematic under estimation for the main fuel isotopes 235U and 239Pu. Discrepancies observed for C/E ratios in various benchmarks range from 10 to 28%, while required accuracy is 7.5%.

Therefore requests for new measurements of prompt gamma rays in the reactions 235U(n,f) and 239Pu(n,f) have been formulated in the Nuclear Data High Priority Request List of the Nuclear Energy Agency.

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Energy Spectra of Prompt Gamma Rays – New Measurements. Source: Laborie, Bélier, Taieb. PROMPT FISSION GAMMA-RAY SPECTRA IN FAST FISSION. Presentation – NFS WORKSHOP 03/2014

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Prompt Neutron Lifetime Prompt neutron lifetime, l, is the average time from a prompt neutron emission to either its absorbtion (fission or radiative capture) or to its escape from the system. This parameter is defined in multiplying or also in Non multiplying systems. In both systems the prompt neutron lifetimes depend strongly on:  material composition of the system 1. multiplying – non multiplying system 2. system with or without thermalization 3. isotopic composition of the system  geometric configuration of the system 1. homogeneous or heterogeneous system 2. shape of entire system  size of the system

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In an infinite reactor (without escape) prompt neutron lifetime is the sum of the slowing down time and the diffusion time. l=ts + td In an infinite thermal reactor ts << td and therefore l â&#x2030;&#x2C6; td. The typical prompt neutron lifetime in thermal reactors is on the order of 10â&#x2C6;&#x2019;4 second. Generally, the longer neutron lifetimes take place in systems in which the neutrons must be thermalized in order to be absorbed.

Systems in which most of the neutrons are absorbed at higher energies and the neutron thermalization is suppressed (e.g. in fast reactors), have much shorter prompt neutron lifetimes . The typical prompt neutron lifetime in fast reactors is on the order of 10â&#x2C6;&#x2019;7 second.

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Slowing Down and Diffision Times for Thermal Neutrons in an Infinite Medium Source: Robert Reed Burn, Introduction to Nuclear Reactor Operation, 1988.

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4.0 Neutron Energy Free neutrons can be classified according to their kinetic energy. This energy is usually given in electron volts (eV). The term temperature can also describe this energy representing thermal equilibrium between a neutron and a medium with a certain temperature.

Classification of free neutrons according kinetic energies Cold Neutrons (0 eV; 0.025 eV). (0~0.01eV) Neutrons in thermal equilibrium with very cold surroundings such as liquid deuterium. This spectrum is used for neutron scattering experiments.

Thermal Neutrons. (0.01~0.3eV) Neutrons in thermal equilibrium with a surrounding medium. Most probable energy at 20°C (68°F) for Maxwellian distribution is 0.025 eV (~2 km/s). This part of neutron’s energy spectrum constitutes most important part of spectrum in thermal reactors.

Epithermal Neutrons (0.025 eV; 0.4 eV) (0.3~104 eV) . Neutrons of kinetic energy greater than thermal. Some of reactor designs operates with epithermal neutron’s spectrum. This design allows to reach higher fuel breeding ratio than in thermal reactors.

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Neutrons of kinetic energy below the cadmium cut-off energy (~0.5 eV) are strongly absorbed by 113-Cd. Source: JANIS (Java-based nuclear information software) www.oecd-nea.org/janis/

(~0.5 eV) (10-6 MeV) Cadmium Neutrons

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Epicadmium Neutrons

Cadmium Neutrons (0.4 eV; 0.5 eV). Neutrons of kinetic energy below the cadmium cut-off energy. One cadmium isotope,113Cd, absorbs neutrons strongly only if they are below ~0.5 eV (cadmium cut-off energy).

Epicadmium Neutrons (0.5 eV; 1 eV). Neutrons of kinetic energy above the cadmium cut-off energy. These neutrons are not absorbed by cadmium.

Slow Neutrons (1 eV; 10 eV). (0~103 eV) Resonance Neutrons (10 eV; 300 eV). The resonance neutrons are called resonance for their special bahavior. At resonance energies the crosssections can reach peaks more than 100x higher as the base value of crosssection. At this energies the neutron capture significantly exceeds a probability of fission. Therefore it is very important (for thermal reactors) to quickly overcome this range of energy and operate the reactor with thermal neutrons resulting in increase of probability of fission.

Intermediate Neutrons (300 eV; 1 MeV).

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Fast Neutrons (1 MeV; 20 MeV) (103 ~ 20Mev) . Neutrons of kinetic energy greater than 1 MeV (~15 000 km/s) are usually named fission neutrons.

These neutrons are produced by nuclear processes such as nuclear fission or (É&#x2018;,n) reactions. The fission neutrons have a Maxwell-Boltzmann distribution of energy with a mean energy (for 235U fission) 2 MeV. Inside a nuclear reactor the fast neutrons are slowed down to the thermal energies via a process called neutron moderation.

Relativistic Neutrons (>20 MeV)

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Distribution of kinetic energies of neutrons in the thermal reactor. The fission neutrons (fast flux) are immediately slowed down to the thermal energies via a process called neutron moderation.

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The reactor physics does not need this fine division of neutron energies. The neutrons can be roughly (for purposes of reactor physics) divided into three energy ranges: Thermal neutrons (0.025 eV – 1 eV). (0.01~ 0.3 eV) Resonance neutrons (1 eV – 1 keV).

Fast neutrons (1 keV – 10 MeV).

Even most of reactor computing codes use only two neutron energy groups: Slow neutrons group (0.025 eV – 1 keV). (0 ~ 103 eV) Fast neutrons group (1 keV – 10 MeV). (103 ~ 20Mev)

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5.0 Neutron interactions with Matters Neutrons are neutral particles, therefore they travel in straight lines, deviating from their path only when they actually collide with a nucleus to be scattered into a new direction or absorbed. Neither the electrons surrounding (atomic electron cloud) a nucleus nor the electric field caused by a positively charged nucleus affect a neutron’s flight. In short, neutrons collide with nuclei, not with atoms. A very descriptive feature of the transmission of neutrons through bulk matter is the mean free path length (λ – lambda), which is the mean distance a neutron travels between interactions. It can be calculated from following equation:

λ=1/Σ

(λ = 1/μn = 1/ σ·N)

λ mean free path length Σ Macroscopic cross section (Σ = σ·N) (μn= Σ = σ·N)

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Neutrons may interact with nuclei in one of following ways:

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Neutron cross-section The extent to which neutrons interact with nuclei is described in terms of quantities known as cross-sections. Cross-sections are used to express the likelihood of particular interaction between an incident neutron and a target nucleus. It must be noted this likelihood do not depend on real target dimensions. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the thermal power of a nuclear power plant. The standard unit for measuring the microscopic cross-section (σsigma) is the barn, which is equal to 10-28 m2. This unit is very small, therefore barns (abbreviated as “b”) are commonly used. The microscopic crosssection can be interpreted as the effective ‘target area’ that a nucleus interacts with an incident neutron.

Keywords: σ- microscopic cross-section barn 10-28 m2

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σ in barn, which is equal to 10-28 m2 the microscopic cross-section can be interpreted as the effective ‘target area’ that a nucleus interacts with an incident neutron.

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http://www.nuclear-power.net/nuclear-power-plant/nuclear-reactor/

Typical cross-sections of fission material. Slowing down neutrons results in increase of probability of interaction (e.g. fission reaction).

Resonance peaks?

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A macroscopic cross-section is derived from microscopic and the material density: Σ=σ.N Here σ, which has units of m2, is referred to as the microscopic cross-section.

Since the units of N (nuclei density) are nuclei/m3, the macroscopic crosssection Σ have units of m-1, thus in fact is an incorrect name, because it is not a correct unit of cross-sections. (haha!)

Neutron cross-sections constitute a key parameters of nuclear fuel. Neutron cross-sections must be calculated for fresh fuel assemblies usually in twoDimensional models of the fuel lattice.

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The neutron cross-section is variable and depends on:  Target nucleus (hydrogen, boron, uranium, etc.) Each isotope has its own set of cross-sections.  Type of the reaction (capture, fission, etc.). Cross-sections are different for each nuclear reaction.  Neutron energy (thermal neutron, resonance neutron, fast neutron). For a given target and reaction type, the cross-section is strongly dependent on the neutron energy. In the common case, the cross section is usually much larger at low energies than at high energies. This is why most nuclear reactors use a neutron moderator to reduce the energy of the neutron and thus increase the probability of fission, essential to produce energy and sustain the chain reaction.  Target energy (temperature of target material – Doppler broadening) This dependency is not so significant, but the target energy strongly influences inherent safety of nuclear reactors due to a Doppler broadening of resonances.

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Law 1/v For thermal neutrons (in 1/v region), absorption cross-sections increases as the velocity (kinetic energy) of the neutron decreases. Therefore the 1/v Law can be used to determine shift in absorbtion cross-section, if the neutron is in equilibrium with a surrounding medium. This phenomenon is due to the fact the nuclear force between the target nucleus and the neutron has a longer time to interact.

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For thermal neutrons (in 1/v region), absorption cross sections increases as the velocity (kinetic energy) of the neutron decreases.

Source: JANIS 4.0

1E-10

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Resonance neutron capture Absorption cross section is often highly dependent on neutron energy. Note that the nuclear fission produces neutrons with a mean energy of 2 MeV (200 TJ/kg, i.e. 20,000 km/s). The neutron can be roughly divided into three energy ranges: Fast neutron. (10MeV – 1keV) (103~ 20Mev) Resonance neutron (1keV – 1eV) (1~1000eV) Thermal neutron. (1eV – 0.025eV) (0.001-0.3eV) The resonance neutrons are called resonance for their special bahavior. At resonance energies the cross-section can reach peaks more than 100x higher as the base value of cross-section. At this energies the neutron capture significantly exceeds a probability of fission. Therefore it is very important (for thermal reactors) to quickly overcome this range of energy and operate the reactor with thermal neutrons resulting in increase of probability of fission.

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Resonance peaks for radiative capture of U238. At resonance energies the probability of capture can be more than 100x higher as the base value.

Source: JANIS program

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Chernobyl

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Doppler broadening A Doppler broadening of resonances is very important phenomenon, which improves reactor stability. The prompt temperature coefficient of most thermal reactors is negative, owing to an nuclear Doppler effect. Although the absorption cross-section depends significantly on incident neutron energy, the shape of the cross-section curve depends also on target temperature.

Nuclei are located in atoms which are themselves in continual motion owing to their thermal energy. As a result of these thermal motions neutrons impinging on a target appears to the nuclei in the target to have a continuous spread in energy. This, in turn, has an effect on the observed shape of resonance. The resonance becomes shorter and wider than when the nuclei are at rest.

Although the shape of a resonance changes with temperature, the total area under the resonance remains essentially constant. But this does not imply constant neutron absorbtion (?, absorption). Despite the constant area under resonance, a resonance integral, which determines the absorption, increases with increasing target temperature. This, of course, decreases coefficient k (negative reactivity is inserted).

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Doppler effect improves reactor stability. Broadened resonance (heating of a fuel) results in a higher probability of absorption, thus causes negative reactivity insertion (reduction of reactor power).

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Typical cross-sections of materials in the reactor - Following table shows neutron cross-sections of the most common isotopes of reactor core.

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6.0 Free Neutron A free neutron is a neutron that is not bounded in a nucleus. The free neutron is, unlike a bounded neutron, subject to radioactive beta decay.

It decays into a proton, an electron, and an antineutrino (the antimatter counterpart of the neutrino, a particle with no charge and little or no mass). A free neutron will decay with a half-life of about 611 seconds (10.3 minutes).

This decay involves the weak interaction and is associated with a quark transformation (a down quark is converted to an up quark). The decay of the neutron is a good example of the observations which led to the discovery of the neutrino. Because it decays in this manner, the neutron does not exist in nature in its free state, except among other highly energetic particles in cosmic rays. Since free neutrons are electrically neutral, they pass through the electrical fields within atoms without any interaction and they are interacting with matter almost exclusively through relatively rare collisions with atomic nuclei.

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The free neutron decays into a proton, an electron, and an antineutrino with a half-life of about 611 seconds (10.3 minutes).

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7.0 Application of Neutron Since their discovery in 1932 neutrons play an important role in many fields of modern science. The discovery of the neutron immediately gave scientists a new tool for probing the properties of atomic nuclei. In particular, discovery of neutrons and their properties has been important in the development of nuclear reactors and nuclear weapons. Main branches where the neutrons play key role are summarized below: Nuclear Reactors A nuclear reactor is a key device of nuclear power plants, nuclear research facilities or nuclear propelled ships. Main purpose of the nuclear reactor is to initiate and control a sustained nuclear chain reaction. The nuclear chain reaction is initiated, sustained and controlled just via the free neutrons. The term chain means that one single nuclear reaction (neutron induced fission) causes an average of one or more subsequent nuclear reactions, thus leading to the possibility of a self-propagating series of these reactions. The â&#x20AC;&#x153;one or moreâ&#x20AC;? is the key parameter of reactor physics. To raise or lower the power, the amount of reactions, respectively the amount of the free neutrons in the nuclear core must be changed (using the control rods).

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Nuclear fission is a nuclear reaction in which the nucleus of an atom splits into smaller parts (lighter nuclei).

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Neutron diffraction Neutron diffraction experiments use an elastic neutron scattering to determine the atomic (or magnetic) structure of a material. The neutron diffraction is based the fact that thermal or cold neutrons have the wavelengths similar to atomic spacings. An examined sample (crystalline solids, gasses, liquids or amorphous materials) must be placed in a neutron beam of thermal (0.025 eV) or cold (neutrons in thermal equilibrium with very cold surroundings such as liquid deuterium) neutrons to obtain a diffraction pattern that provides information about the structure of the examined material. The neutron diffraction experiments are similar to X-ray diffraction experiments, but neutrons interact with matter differently. Photons (X-rays) interact primarily with the electrons surrounding (atomic electron cloud) a nucleus, but neutrons interact only with nuclei. Neither the electrons surrounding (atomic electron cloud) a nucleus nor the electric field caused by a positively charged nucleus affect a neutronâ&#x20AC;&#x2122;s flight. Due to their different properties, both methods together (neutron diffraction and X-ray diffraction) can provide complementary information about the structure of the material.

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Simple scheme of neutron diffraction experiment.

When the neutron collide with the atom in the sample material, they change direction (scattered)-elastic scattering

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Neutron Source

Applications in Medicine Medical applications of neutrons began soon after the discovery of this particle in 1932. Neutrons are highly penetrating matter and ionizing, so they can be used in medical therapies such as radiation therapy or boron capture therapy.

Unfortunately neutrons, when they are absorbed in matter, active the matter and leave the matter (target area) radioactive.

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Neutron activation analysis Neutron activation analysis is a method for determining the composition of examined material. This method was discovered in 1936 and stands at the forefront of methods used for quantitative material analysis of major, minor, trace, and rare elements. This method is based on neutron activation, where an analyzed sample is first irradiated with neutrons to produce specific radionuclides. The radioactive decay of these produced radionuclides is specific for each element (nuclide). Each nuclide emits the characterictic gamma rays which are measured using gamma spectroscopy, where gamma rays detected at a particular energy are indicative of a specific radionuclide and determine concentrations of the elements. Main advantage of this method is that neutrons does not destroy the sample (?, making it radioactive and transmuted the target?) .

This method can be also used for determine an enrichment of nuclear material.

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An analyzed sample is first irradiated with neutrons to produce specific radionuclides. The radioactive decay of these produced radionuclides is specific for each element (nuclide).

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Macho Leader with nuclear ambition.

Science is for allâ&#x20AC;Ś..

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8.0 Detection of Neutron Since the neutrons are electrically neutral particles, they are mainly subject to strong nuclear forces but not to electric forces. Therefore neutrons are not directly ionizing and they have usually to be converted into charged particles before they can be detected. Generally every type of neutron detector must be equipped with converter (to convert neutron radiation to common detectable radiation) and one of the conventional radiation detectors (scintillation detector, gaseous detector, semiconductor detector, etc.).

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Neutron converters Two basic types of neutron interactions with matter are for this purpose available: Elastic scattering. The free neutron can be scattered by a nucleus, transferring some of its kinetic energy to the nucleus. If the neutron has enough energy to scatter off nuclei the recoiling nucleus ionizes the material surrounding the converter. In fact, only hydrogen and helium nuclei are light enough for practical application. Charge produced in this way can be collected by the conventional detector to produce a detected signal. Neutrons can transfer more energy to light nuclei. This method is appropriate for detecting fast neutrons (fast neutrons do not have high cross-section for absorption) allowing detection of fast neutrons without a moderator.

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Neutron absorption. This is a common method allowing detection of neutrons of entire energy spectrum. This method is is based on variety of absorption reactions (radiative capture, nuclear fission, rearrangement reactions (?) , etc.). The neutron is here absorbed by target material (converter) emitting secondary particles such as protons, alpha particles, beta particles, photons (gamma rays) or fission fragments.

Some reactions are threshold reactions (requiring a minimum energy of neutrons), but most of reactions occurs at epithermal and thermal energies.

That means the moderation of fast neutrons is required leading in poor energy information of the neutrons. Most common nuclei for the neutron converter material are:

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

10B(n,α).

Where the neutron capture cross-section for thermal neutrons is σ = 3820 barns and the natural boron has abundance of 10B 19,8%.

 3He(n,p). Where the neutron capture cross-section for thermal neutrons is σ = 5350 barns and the natural helium has abundance of 3He 0.014%.  6Li(n,α). Where the neutron capture cross-section for thermal neutrons is σ = 925 barns and the natural lithium has abundance of 6Li 7,4%. 

113Cd(n,ɣ).

Where the neutron capture cross-section for thermal neutrons is σ = 20820 barns and the natural cadmium has abundance of 113Cd 12,2%.

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ď Ž

Where the fission cross-section for thermal neutrons is Ď&#x192; = 585 barns and the natural uranium has abundance of 235U 0.711%.

235U(n,fission).

Uranium as a converter produces fission fragments which are heavy charged particles. This have significant advantage. The heavy charged particles (fission fragments) create a high output signal, because the fragments deposit a large amount of energy in a detector sensitive volume.

This allows an easy discrimination of the background radiation (i.e. gamma radiation). This important feature can be used for example in a nuclear reactor power measurement, where the neutron field is accompanied by a significant gamma background.

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9.0 Shielding of Neutron Radiation In radiation protection there are three ways how to protect people from In radiation protection there are three ways how to protect people from identified radiation sources: Limiting Time. The amount of radiation exposure depends directly (linearly) on the time people spend near the source of radiation. The dose can be reduced by limiting exposure time.

Distance. The amount of radiation exposure depends on the distance from the source of radiation. Similarly to a heat from a fire, if you are too close, the intensity of heat radiation is high and you can get burned. If you are at the right distance, you can withstand there without any problems and moreover it is comfortable. If you are too far from heat source, the insufficiency of heat can also hurt you. This analogy, in a certain sense, can be applied to radiation also from nuclear sources.

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Shielding. Finally, if the source is too intensive and time or distance do not provide sufficient radiation protection the shielding must be used. Radiation shielding usually consist of barriers of lead, concrete or water. Even depleted uranium can be used as a good protection from gamma radiation, but on the other hand uranium is absolutely inappropriate shielding of neutron radiation.

In short, it depends on type of radiation to be shielded, which shielding will be effective or not.

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Shielding of Neutrons There are three main features of neutrons, which are crucial in the shielding of neutrons.

1. Neutrons have no net electric charge, therefore they cannot be affected or stopped by electric forces. Neutrons ionize matter only indirectly, which makes neutrons highly penetrating type of radiation.

2. Neutrons scatter with heavy nuclei very elastically. Heavy nuclei very hard slow down a neutron let alone absorb a fast neutron.

3. An absorption of neutron (one would say shielding) causes initiation of certain nuclear reaction (e.g. radioactive capture or even fission), which is accompanied by a number of other types of radiation. In short, neutrons make matter radioactive, therefore with neutrons we have to shield also the other types of radiation.

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The best materials for shielding neutrons must be able to: ď Ž Slow down neutrons (the same principle as the neutron moderation).

First point can be fulfilled only by material containing light atoms (e.g. hydrogen atoms), such as water, polyethylene, and concrete. The nucleus of a hydrogen nucleus contains only a proton. Since a proton and a neutron have almost identical masses, a neutron scattering on a hydrogen nucleus can give up a great amount of its energy (even entire kinetic energy of a neutron can be transferred to a proton after one collision). This is similar to a billiard. Since a cue ball and another billiard ball have identical masses, the cue ball hitting another ball can be made to stop and the other ball will start moving with the same velocity. On the other hand, if a ping pong ball is thrown against a bowling ball (neutron vs. heavy nucleus), the ping pong ball will bounce off with very little change in velocity, only a change in direction. Therefore lead is quite ineffective for blocking neutron radiation, as neutrons are uncharged and can simply pass through dense materials.

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Table of cross-sections

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ď Ž Absorb this slow neutron. Thermal neutrons can be easily absorbed by capture in materials with high neutron capture cross sections (thousands of barns) like boron, lithium or cadmium (prompt) . Generally, only a thin layer of such absorbator is sufficient to shield thermal neutrons. Hydrogen (in the form of water), which can be used to slow down neutrons, have absorbtion cross-section 0.3 barns. This is not enough, but this insufficiency can be offset by sufficient thickness of water shield. ď Ž Shield the accompanying radiation. In the case of cadmium shield the absorption of neutrons is accompanied by strong emission of gamma rays.

Therefore additional shield is necessary to attenuate the gamma rays. This phenomenon practically does not exist for lithium and is much less important for boron as a neutron absorption material. For this reason, materials containing boron are used often in neutron shields. In addition, boron (in the form of boric acid) is well soluble in water making this combination very effective neutron shield.

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TABLE 6. Properties of Some Thermal Neutron Radiography Conversion Materials Material

Useful Reactions

Lithium

6Li(n,α) 3H

910

prompt

Boron

10B(n,α) 7Li

3,830

prompt

Rhodium

103Rh(n)104mRh

11

45 min

103Rh(n)104Rh

139

42 s

107Ag(n)108Ag

35

2.3 min

109Ag

91

24 s

Silver

(n)110 Ag

Cadmium

113Cd((n,γ)114Cd

Indium

115

In(n)116n

115 In(n)116mln

Samarium

149Sm(n,γ) 150Sm I52 Sm(n)153Sm

Cross Section for Thermal Neutrons (barns)

Life

20,000

prompt

157

54 min

42

14 s

41,000

prompt

210

47 h

Europium

151

Eu(n)152Eu

3,000

9.2 h

Gadolinium

155 Gd(n,γ) I56Gd

61,000

prompt

157 Gd(n.γ)158Gd

254,000

prompt

164 Dy(n)165mDy

2,200

1.25 min

800

140 min

99

2.7 days

Dyprosium

164 Dy(n)165Dy

Gold Charlie Chong/ Fion Zhang

197

Au(n)198Au

Water as a neutron shield Water due to the high hydrogen content and the availability is effective and common neutron shielding. However, due to the low atomic number of hydrogen and oxygen, water is not acceptable shield against the gamma rays.

On the other hand in some cases this disadvantage (low density) can be compensated by high thickness of the water shield. In case of neutrons, water perfectly moderates neutrons, but with absorption of neutrons by hydrogen nucleus secondary gamma rays with the high energy are produced.

These gamma rays highly penetrates matter and therefore it can increase requirements on the thickness of the water shield. Adding a boric acid can help with this problem (neutron absorption on boron nuclei without strong gamma emission), but results in another problems with corrosion of construction materials.

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Concrete as a neutron shield Most commonly used neutron shielding in many sectors of the nuclear science and engineering is shield of concrete. Concrete is also hydrogencontaining material, but unlike water concrete have higher density (suitable for secondary gamma shielding) and does not need any maintenance.

Because concrete is a mixture of several different materials its composition is not constant. So when referring to concrete as a neutron shielding material, the material used in its composition should be told correctly. Generally concrete are divided to “ordinary “ concrete and “heavy” concrete. Heavy concrete uses heavy natural aggregates such as barites (barium sulfate) or magnetite or manufactured aggregates such as iron, steel balls, steel punch or other additives. As a result of these additives, heavy concrete have higher density than ordinary concrete (~2300 kg/m3). Very heavy concrete can achieve density up to 5,900 kg/m3 with iron additives or up to 8900 kg/m3 with lead additives. Heavy concrete provide very effective protection against neutrons.

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Neutron cross sections Neutron cross sections are defined in Part 1 of this Section. Values for thermal neutrons for many materials (elements) are given in Table 9 (see Bibliography item 8 for a more extensive compilation). Generally, neutron cross sections decrease with increasing neutron energy; exceptions include resonances, as mentioned earlier. Cross section values can be used to calculate the attenuation coefficients and the neutron transmission as shown in 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: μn = ρ∙Nσ/ A

or

μn = Ṅσ

Eq.7

Where: μn 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.

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Ṅ is the number of atoms per cubic centimeter Ṅ=

ρ∙N A

μn = Ṅ∙σ Where: μn 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.

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

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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∙(ѵ1σ1 + ѵ2σ2 + ѵ3σ3 +..... ѵiσi )

Eq. 8

M 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).

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As an example, consider the calculations of the linear attenuation coefficient, p.., for the compound polyethylene (CH2)N : μ = ρ∙N (ѵ1σ1 + ѵ2σ2 + ѵ3σ3 +..... ѵiσi ) /M μ = ρ∙N (ѵCσC + ѵHσH) /M for:

ρ = 0.91 g/cm3 N = 6.023 X 1023 atoms/g-mol M= 14.0268 g ѵC = 1 σC = 4.803 X 10-24 cm2 ѵH= 2 σH = 38.332 X 10-24 cm2

μ = 0.91 x (6.023 x 1023) x (1x 4.803+ 2x 38.332) x 10-24 14.0268 μ = 3.18329 cm-1

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Eq. 8

σH

σC

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10.0 Neutron Source A neutron source is any device that emits neutrons. Neutron sources have many applications, they can be used in research, engineering, medicine, petroleum exploration, biology, chemistry and nuclear power. A neutron source is characterized by a number of factors: 1.

2.

3.

4.

5.

Significance of the source Intensity. The rate of neutrons emitted by the source.

Energy distribution of emitted neutrons.

Angular distribution of emitted neutrons.

Mode of emission. Continuous or pulsed operation.

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Classification by significance of the source â&#x2013; Large (Significant) neutron sources â&#x20AC;˘ Nuclear Reactors. There are nuclei that can undergo fission on their own spontaneously, but only certain nuclei, like uranium-235, uranium-233 and plutonium-239, can sustain a fission chain reaction. This is because these nuclei release neutrons when they break apart, and these neutrons can induce fission of other nuclei. Uranium-235 which exists as 0.7% of naturally occurring uranium undergoes nuclear fission with thermal neutrons with the production of, on average, 2.4 fast neutrons and the release of ~ 180 MeV of energy per fission. Free neutrons released by each fission play very important role as a trigger of the reaction, but they can be also used fo another purpose. For example: One neutron is required to trigger a further fission. Part of free neutrons (let say 0.5 neutrons/fission) is absorbed in other material, but an excess of neutrons (0.9 neutrons/fission) is able to leave the surface of the reactor core and can be used as a neutron source.

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• Fusion Systems. Nuclear fusion is a nuclear reaction in which two or more atomic nuclei (e.g. D+T) collide at a very high energy and fuse together.

Thy byproduct of DT fusion is a free neutron (see picture), therefore also nuclear fusion reaction has the potential to produces large quantities of neutrons.

• Spallation Sources. A spallation source is a high-flux neutron source in which protons that have been accelerated to high energies hit a heavy target material, causing the emission of neutrons. The reaction occurs above a certain energy threshold for the incident particle, which is typically 5 – 15 MeV.

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■ Medium neutron sources • Bremssstrahlung from Electron Accelerators / Photofission. Energetic electrons when slowed down rapidly in a heavy target emit intense gamma radiation during the deceleration process. This is known as Bremsstrahlung or braking radiation. The interaction of the gamma radiation with the target produces neutrons via the (γ,n) reaction, or the (γ,fission) reaction when a fissile target is used. e-→Pb → γ→ Pb →(γ,n) and (γ,fission). The Bremsstrahlung γ energy exceeds the binding energy of the “last” neutron in the target. A source strength of 1013 neutrons/second produced in short (i.e. < 5 μs) pulses can be readily realised.

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â&#x20AC;˘ Dense plasma focus. The dense plasma focus (DPF) is a device that is known as an efficient source of neutrons from fusion reactions.

Mechanism of dense plasma focus (DPF) is based on nuclear fusion of short-lived plasma of deuterium and/or tritium. This device produces a short-lived plasma by electromagnetic compression and acceleration that is called a pinch. This plasma is during the pinch hot and dense enough to cause nuclear fusion and the emission of neutrons. â&#x20AC;˘ Light ion accelerators. Neutrons can be also produced by particle accelerators using targets of deuterium, tritium, lithium, beryllium, and other low-Z materials. In this case the target must be bombarded with accelerated hydrogen (H), deuterium (D), or tritium (T) nuclei.

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Electromagnetic Compression This is a basic explanation of how a pinch works. (1) Pinches apply a huge voltage across a tube. This tube is filled with fusion fuel, typically deuterium gas. If the voltage times the charge is higher than the ionization energy of the gas the gas ionizes. (2) Current jumps across this gap. (3) The current makes a magnetic field which is perpendicular to the current. This magnetic field pulls the material together. (4) These atoms can get close enough to fuse.

http://www.snipview.com/q/pulsed_power

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https://en.wikipedia.org/wiki/Pinch_(plasma_physics)

Small neutron sources â&#x20AC;˘ Neutron Generators. Neutrons are produced in the fusion of deuterium and tritium in the following exothermic reaction.

2D

+ 3T â&#x2020;&#x2019; 4He + n + 17.6 MeV.

The neutron is produced with a kinetic energy of 14.1 MeV. This can be achieved on a small scale in the laboratory with a modest 100 kV accelerator for deuterium atoms bombarding a tritium target. Continuous neutron sources of ~1011 neutrons/second can be achieved relatively simply.

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• Radioisotope source – (α,n) reactions. In certain light isotopes the ‘last’ neutron in the nucleus is weakly bound and is released when the compound nucleus formed following α-particle bombardment decays. The bombardment of beryllium by α-particles leads to the production of neutrons by the following exothermic reaction: 4He + 9Be→12C + n + 5.7 MeV. This reaction yields a weak source of neutrons with an energy spectrum resembling that from a fission source and is used nowadays in portable neutron sources. Radium, plutonium or americium can be used as an αemitter.

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Radioisotope source – (γ,n) reactions. (γ,n) reactions can also be used for the same purpose. In this type of source, because of the greater range of the γ-ray, the two physical components of the source can be separated making it possible to ‘switch off’ the reaction if so required by removing the radioactive source from the beryllium. (γ,n) sources produce a monoenergetic neutrons unlike (α,n) sources. The (γ,n) source uses antimony-124 as the gamma emitter in the following endothermic reaction. (?)

124Sb→ 124Te + β− + γ γ + 9Be→ 8Be + n – 1.66 MeV

Charlie Chong/ Fion Zhang

Radioisotope source â&#x20AC;&#x201C; spontaneous fission. Certain isotopes undergo spontaneous fission with emission of neutrons. The most commonly used spontaneous fission source is the radioactive isotope californium-252. Cf252 and all other spontaneous fission neutron sources are produced by irradiating uranium or another transuranic element in a nuclear reactor, where neutrons are absorbed in the starting material and its subsequent reaction products, transmuting the starting material into the SF isotope.

Charlie Chong/ Fion Zhang

Peach – 我爱桃子

Charlie Chong/ Fion Zhang

Charlie Chong/ Fion Zhang

More Reading http://nmi3.eu/neutron-research/techniques-for-/imaging.html http://www.nuclear-power.net/nuclear-power-plant/nuclear-reactor/

Charlie Chong/ Fion Zhang

Good Luck

Charlie Chong/ Fion Zhang

Good Luck

Charlie Chong/ Fion Zhang

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