**1. Introduction**

16 Gamma Radiation

Hubbell, J.H. & Seltzer, S.M. (2001). *Tables of X-Ray Mass Attenuation Coefficient and Mass Ene-*

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Parker, P.R.. Smith, S.H. & Taylor, M.D. (1978), *Basic Science of Nuclear Medicine,* London:

Sabol, J. & Weng, P.S. (1995). *Introduction to Radiation Protection Dosimetry,* Singapore: World

Shalek, R.S. & Stoval M.(1969). in: *Radiation Dosimetry Vol.III: Sources, Fields, Measurements* 

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Wasserman, H. & Groenwald, W. (1988). *Air kerma rate constants for radionuclides,* Eur. J.

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*rgy Absorption Coefficient from 1 keV to 20 MeV for Elements Z = 1 to 92 and 48 Additional Substances of Dosimetric Interest,* NISTIR 5632, Available at: http://physics.nist. gov./PhysRefData/XrayMassCoef/cover.htmlICRU (1963). *Handbook 86,* Natl. Bur.

> Neutron interrogation based methods of non-destructive analysis are well established techniques employed in the field of bulk material analysis. These methods utilize a source of neutrons (a neutron probe) to irradiate objects under scrutiny. Nuclear reactions initiated by neutrons in the volume of the irradiated sample include the following: inelastic neutron scattering, thermal neutron capture, and neutron activation. As a result of nuclear reactions with the material inside the object, the "fingerprint" -rays are emitted with characteristic energies. These characteristic gamma rays are used for the elemental identification. By measuring and counting the number of -rays emitted with a specific energy, one can deduce the amount of the associated chemical element in the sample. The amounts of chemical elements measured allow specifying the chemical composition of the analyzed sample.

> Neutron technique is an excellent choice to rapidly determine elemental content of the sample *in situ* in non-intrusive manner. It is a great fit for in situ applications that involve samples that are hard to reach or unsafe to handle, and that require the analysis to be performed rapidly, in real time.

> Accelerator based neutron sources such as deuterium – deuterium (d-d) and deuterium – tritium (d-t) fusion neutron generators provide the electronic control of neutron emission including its time structure. The pulse mode of neutron production allows the use of coincidence methods to segregate prompt and delayed gamma ray signatures emitted from neutron induced nuclear reactions. The kinematics of fusion reactions allows "tagging" of outgoing neutrons using the associated particles.

> The pulse neutron systems are used in industry for analysis of coal (Dep et al., 1998; Sowerby, 2009), cement (Womble et al., 2005), metal alloys (James & Fuerst, 2000), in geological and soil analysis (Wielopolski et al., 2008), and oil well logging (Nikitin & Bliven, 2010). Security applications of neutron based systems are for chemical and explosive threats detection (Vourvopoulos & Womble, 2001; Aleksandrov et al., 2005; Lanza, 2006), including the search for threats in cargo containers (Barzilov & Womble, 2003) and vehicles (Reber et al., 2005; Koltick et al., 2007), humanitarian demining and confirmation of unexploded ordinance (Womble et al., 2002; Holslin et al., 2006). Such technologies are considered in astrochemistry applications for in situ analysis of planetary samples (Parsons et al., 2011).

Material Analysis Using Characteristic Gamma Rays Induced by Neutrons 19

Pulse structure of neutron emission from isotopic source or reactor is usually controlled with a chopper system. Some reactors provide the pulse periodic operation mode. The neutron beam's energy follows 235U fission spectrum distribution, or depends on the moderator type used inside the core or in neutron beam optics. Nuclear reactors are bulky, expensive, and require significant radiation shielding. That makes them impossible for use

Accelerator-based neutron sources are widely used in material analysis. These sources utilize charged particle beams to create fast neutrons in nuclear reactions induced in various targets. Some examples of such neutron producing reactions are the following: T(d,n)4H, D(d,n)3He, 9Be(d,n)10Be, 7Li(d,n)8Be, 7Li(p,n)7Be, 7Be(p,n)7B. The pulse neutron emission scheme allowing high repetition rates is provided by controlling acceleration parameters electronically. Such sources can be turned off thus simplifying radiation shielding requirements. The neutron sources based on fusion reactions are compact systems due to a large reaction resonance at low deuteron energy (approximately 3.4 barn for 100 keV for d-t fusion). The d-t fusion neutron generators are widely used as sources of neutrons for portable probe and industrial applications. These isotropic neutron sources are rugged, low

Neutron generators utilize d-t and d-d fusion reactions that produce mono energetic neutrons. The d-t reaction shows greater energy release. At the incident particle's small energies, 4He and neutron share 17.59 MeV with conservation of linear momentum, and

In 50% of events, d-d reaction produces mono energetic 2.45 MeV neutrons and 3He that

With 50% probability, 3T and proton may be also produced in d-d reaction (Q=4.03 MeV,

Neutrons produced in the d-t reaction are emitted isotropically. Neutron emission in d-d reaction is slightly peaked forward along the direction of ion beam. The yield of d-d reaction at low energies of deuterons (reaction cross section is 3.3×10-2 barn at 100 keV) is approximately two orders of magnitude lower than in d-t fusion. Therefore, higher deuteron current is required to achieve d-d neutron yields comparable to a d-t source. Because of that d-t neutron generators are more common in applications requiring small size neutron sources of higher energy neutrons. The d-d systems may be preferable in applications where only the lower energy neutrons are required and where 14.1-MeV neutrons may cause unnecessary interference in the analysis due to many reactions channels open for high

Another reaction which can be used for neutron production is the t-t fusion

2H + 3H 4He + n (Q=17.59 MeV, En=14.1 MeV, EHe=3.49 MeV) (1)

2H + 2H 3He + n (Q=3.27 MeV, En=2.45 MeV, EHe=0.82 MeV) (2)

3H + 3H 4He + n + n (Q=11.3 MeV) (3)

as a neutron source for portable material analysis systems.

maintenance, and relatively inexpensive systems.

share 3.27 MeV:

energy neutrons.

Ep=3.02 MeV, ET=1.01 MeV).

**2.1 Accelerator based fusion neutron generators** 

mono energetic 14.1 MeV neutrons are emitted out of reaction

The use of the pulse neutron based analysis of nitrogen and oxygen content *in vivo* is discussed in nutrition research (Shypailo & Ellis, 2005) and in cancer diagnostics (Maglich & Nalcioglu, 2010).

In the presented chapter we discuss the components of pulse neutron based material analysis systems, nuclear reactions induced by neutrons, characteristic gamma radiation emitted in these nuclear reactions, gamma ray spectral analysis methods for elemental characterization, and "neutrons in – photons out" methods that utilize the characteristic gamma radiation.

### **2. Pulse neutron sources and system components**

A pulse neutron based material analysis system consists of a neutron source, gamma and particle radiation detector(s), a shadow radiation shielding to cover detectors from direct source neutrons, and associated hardware and software for system control, data acquisition and processing. Fig.1 shows the scheme of a typical system. The system operates as follows. Emitted by a source neutrons induce nuclear reactions in the irradiated object and excite nuclei. Excited nuclei emit photons due to various de-excitation processes that are measured by a gamma ray detector. The gamma ray spectrum is analyzed providing information on the chemical composition of the irradiated sample.

Fig. 1. Pulse neutron based elemental analysis scheme

Various radioisotopes, neutron beams from a nuclear reactor core, or accelerator-based devices are used as neutron sources. Radioisotopes used in neutron sources include 252Cf, 239Pu, 241Am, and others. Californium-252 undergoes spontaneous fission with emission of neutrons with the average energy 2.5 MeV. The 252Cf neutron emission spectral distribution is described by the semi-empirical Watt formula. Plutonium and americium based sources emit neutrons using alpha decay of 239Pu or 241Am and (,n) reactions in the matrix of light elements such as beryllium or lithium. The neutron energy spectrum of these sources is wide (up to ~11 MeV) with the average neutron energy ~4.5 MeV. Radioisotope sources may require radiation shielding while not in use.

The use of the pulse neutron based analysis of nitrogen and oxygen content *in vivo* is discussed in nutrition research (Shypailo & Ellis, 2005) and in cancer diagnostics (Maglich &

In the presented chapter we discuss the components of pulse neutron based material analysis systems, nuclear reactions induced by neutrons, characteristic gamma radiation emitted in these nuclear reactions, gamma ray spectral analysis methods for elemental characterization, and "neutrons in – photons out" methods that utilize the characteristic

A pulse neutron based material analysis system consists of a neutron source, gamma and particle radiation detector(s), a shadow radiation shielding to cover detectors from direct source neutrons, and associated hardware and software for system control, data acquisition and processing. Fig.1 shows the scheme of a typical system. The system operates as follows. Emitted by a source neutrons induce nuclear reactions in the irradiated object and excite nuclei. Excited nuclei emit photons due to various de-excitation processes that are measured by a gamma ray detector. The gamma ray spectrum is analyzed providing information on

Various radioisotopes, neutron beams from a nuclear reactor core, or accelerator-based devices are used as neutron sources. Radioisotopes used in neutron sources include 252Cf, 239Pu, 241Am, and others. Californium-252 undergoes spontaneous fission with emission of neutrons with the average energy 2.5 MeV. The 252Cf neutron emission spectral distribution is described by the semi-empirical Watt formula. Plutonium and americium based sources emit neutrons using alpha decay of 239Pu or 241Am and (,n) reactions in the matrix of light elements such as beryllium or lithium. The neutron energy spectrum of these sources is wide (up to ~11 MeV) with the average neutron energy ~4.5 MeV. Radioisotope sources

**2. Pulse neutron sources and system components** 

the chemical composition of the irradiated sample.

Fig. 1. Pulse neutron based elemental analysis scheme

may require radiation shielding while not in use.

Nalcioglu, 2010).

gamma radiation.

Pulse structure of neutron emission from isotopic source or reactor is usually controlled with a chopper system. Some reactors provide the pulse periodic operation mode. The neutron beam's energy follows 235U fission spectrum distribution, or depends on the moderator type used inside the core or in neutron beam optics. Nuclear reactors are bulky, expensive, and require significant radiation shielding. That makes them impossible for use as a neutron source for portable material analysis systems.

Accelerator-based neutron sources are widely used in material analysis. These sources utilize charged particle beams to create fast neutrons in nuclear reactions induced in various targets. Some examples of such neutron producing reactions are the following: T(d,n)4H, D(d,n)3He, 9Be(d,n)10Be, 7Li(d,n)8Be, 7Li(p,n)7Be, 7Be(p,n)7B. The pulse neutron emission scheme allowing high repetition rates is provided by controlling acceleration parameters electronically. Such sources can be turned off thus simplifying radiation shielding requirements. The neutron sources based on fusion reactions are compact systems due to a large reaction resonance at low deuteron energy (approximately 3.4 barn for 100 keV for d-t fusion). The d-t fusion neutron generators are widely used as sources of neutrons for portable probe and industrial applications. These isotropic neutron sources are rugged, low maintenance, and relatively inexpensive systems.

#### **2.1 Accelerator based fusion neutron generators**

Neutron generators utilize d-t and d-d fusion reactions that produce mono energetic neutrons. The d-t reaction shows greater energy release. At the incident particle's small energies, 4He and neutron share 17.59 MeV with conservation of linear momentum, and mono energetic 14.1 MeV neutrons are emitted out of reaction

$$\text{12H} + \text{3H} \rightarrow \text{4Fe} + \text{n (Q=17.59 MeV, E}\_{\text{n}} = 14.1 \text{ MeV, E}\_{\text{Hle}} = 3.49 \text{ MeV)} \tag{1}$$

In 50% of events, d-d reaction produces mono energetic 2.45 MeV neutrons and 3He that share 3.27 MeV:

$$\text{\textbullet H} + \text{\textbullet H} \rightarrow \text{\textbullet He} + \text{\textbullet (Q=3.27 MeV)}, \text{E}\_{\text{n}} \equiv \text{2.45 MeV}, \text{E}\_{\text{He}} \equiv \text{0.82 MeV} \tag{2}$$

With 50% probability, 3T and proton may be also produced in d-d reaction (Q=4.03 MeV, Ep=3.02 MeV, ET=1.01 MeV).

Neutrons produced in the d-t reaction are emitted isotropically. Neutron emission in d-d reaction is slightly peaked forward along the direction of ion beam. The yield of d-d reaction at low energies of deuterons (reaction cross section is 3.3×10-2 barn at 100 keV) is approximately two orders of magnitude lower than in d-t fusion. Therefore, higher deuteron current is required to achieve d-d neutron yields comparable to a d-t source. Because of that d-t neutron generators are more common in applications requiring small size neutron sources of higher energy neutrons. The d-d systems may be preferable in applications where only the lower energy neutrons are required and where 14.1-MeV neutrons may cause unnecessary interference in the analysis due to many reactions channels open for high energy neutrons.

Another reaction which can be used for neutron production is the t-t fusion

$$\text{H}\_3\text{H} + 3\text{H} \rightarrow 4\text{Fe} + \text{n} + \text{n} \text{ (Q=11.3 MeV)}\tag{3}$$

Material Analysis Using Characteristic Gamma Rays Induced by Neutrons 21

example, the beta decay) may produce photons or charged particles that interfere with the characteristic gamma ray spectra adding the noise and overloading the data acquisition electronics. It is a complicated task to satisfy all these requirements, especially with added cost limitations. Usually, the trade-off between various detector parameters including its

Standard gamma ray detector solutions for spectroscopy are high purity germanium detectors (HPGe) with liquid nitrogen dewar or mechanical cooling subsystems (Sangsingkeow et al., 2003), and scintillation detectors such as NaI(Tl), Bi4Ge3O12, LaBr3(Ce) (van Loef et al., 2001), etc. Noble gas scintillation or ionization detectors and gamma ray telescopes can also be used for neutron induced photon measurements in the MeV energy

The NaI(Tl) scintillator material has the light yield 38 photons/keV, 1/e decay time 250 ns, and density 3.67 g/cc. Atomic numbers are 53 and 11 for iodine and sodium, respectively. Under neutron irradiation, the NaI(Tl) scintillator is activated by neutrons showing the

The BGO scintillator has the light yield 9 photons/keV, and 1/e decay time 300 ns. Due to the high atomic number of bismuth, 83, and the crystal's high density of 7.13 g/cc, the BGO scintillator is very effective for detection of high energy photons. Its energy resolution is lower than NaI(Tl) resolution: ~10% FWHM versus ~7% FWHM for 662-keV -rays. The BGO demonstrates excellent behaviour under neutron irradiation without delayed decay issues. Significant downside of the BGO detector is its sensitivity to the environmental

The LaBr3(Ce) scintillator has the ~3%-resolution of the 662-keV peak, and density 5.08 g/cc. The lanthanum atomic number is 57. This scintillator has high light yield 63 photons/keV, fast 1/e decay time 16 ns, and better timing properties than NaI(Tl). The LaBr3(Ce) material contains small quantities of the radioactive lanthanum-138 isotope (t1/2=1.02×1011 years) producing the 1.47-MeV gamma ray peak that is always visible in the spectrum; it can be used for calibration purposes. The LaBr3(Ce) is affected by neutrons showing the delayed beta decay spectral continuum with endpoint energy ~3 MeV when irradiated with a d-t neutron source. The measured β-decay curve exhibits cumulative nature: two isotopes decay at the same time. The 80Br decays with a half-life 17.68 minutes. The 82Br isotope decays with the half-life approximately 35 hours. Lanthanum halide demonstrates stable gamma ray spectrum parameters in the mixed field under d-t neutron irradiation, when properly shielded. The good energy resolution under the room temperature, the high brightness, and the high scintillation decay speed pose this material as a promising candidate for active neutron interrogation applications, if the crystal's neutron activation issues are properly

The HPGe detector has superior energy resolution comparing to scintillation detectors. The atomic number of germanium is 32. The HPGe crystal density is 5.35 g/cc. HPGe crystal is sensitive to the high energy neutrons, which cause detector damage (Tsoulfanidis & Landsberger, 2010). High energy neutrons produce charges in the germanium crystal which are adding noise to the collected gamma ray spectrum (Ljungvall & Nyberg, 2005). Neutron collisions with the crystal cause atom displacements into interstitial positions creating a vacancy pair. These crystal defects behave as trapping

cost is considered for a particular application (Barzilov & Womble, 2006).

delayed beta decay spectral continuum with the endpoint energy ~2 MeV.

range (April et al., 2006).

temperature (Womble et al., 2002).

addressed.

The reaction cross section is 3.4×10-2 barn at 100 keV, which is similar to the cross section of the d-d reaction. The distribution of energy between reaction products varies producing the wide neutron spectrum with maximum energy up to ~11 MeV. Wide neutron spectrum may be useful for material analysis applications that require both low and high energy neutrons.

The compact "sealed tube" neutron generator design includes an ion source, a positive ion accelerator, and a target. Commonly used source to generate positive ions is the cold cathode Penning source. This source has a cylindrical anode under ~1-2 kV potential applied to it, and grounded cathodes on the ends of the anode. The magnet surrounds the anode cylinder setting up the coaxial magnetic field inside it. The tritium or deuterium gas is introduced into the volume of the anode cylinder. The electric field between anode and cathodes causes ionization of gas molecules creating the cold plasma. Trapped inside the anode electrons are moving in the volume and ionizing gas molecules which helps to maintain the plasma quality. Ions are transferred into the acceleration region through the exit port of the cathode. This region supplies the electric field (up to ~100-120 kV) to accelerate the positive ions. Neutron generator target is a metal hydride loaded with deuterium or tritium or the mixture of both. The ions interact with a target, producing neutrons in fusion reactions. Typical neutron output levels of sealed tube neutron generators are ~108-109 n/s (d-t) and ~106 n/s (d-d). Higher output usually shortens the sealed tube's life time. Sealed tube neutron generators are produced by Thermo Fisher Scientific, Schlumberger, Baker Hughes, EADS SODERN in France, and VNIIA in Russia.

Other designs of neutron generators utilize ion sources such as hot cathode source, radiofrequency ion source, or inertial electrostatic confinement (IEC) based source. For example, the pulse d-d system produced by Adelphi Technology Inc. (Williams et al., 2008) using the microwave driven plasma source technique provides the 2.45-MeV neutron output up to 8×109 n/s. The similar technique using tritium provides the yield of 14.1-MeV neutrons ~1011 n/s. The d-t neutron generator developed by NSD-Fusion GmbH for irradiation of extended samples uses the IEC technique producing ~1010 n/s. These neutron generator designs have the longer life time compared to sealed tube sources.

#### **2.2 Gamma ray detectors**

Physical parameters and limitations of the gamma ray detectors used in the system govern parameters of the entire system. The choice of gamma ray detectors is important for neutron based system to be effective. Improper solution can generate both false positive and false negative results.

The gamma ray detectors must be suitable for operation in mixed radiation fields where neutrons and gamma rays present. The detector material must have a high Z value to effectively detect characteristic photons with energies up to 10.8 MeV. The detection medium must also provide the energy resolution that allows resolving peaks of interest. Ideally, the detector should provide minimum interference with the signal emitted from a sample when the detector material is irradiated with neutrons. Thus, if possible the detector material should avoid isotopes that are anticipated in the analyzed samples. The neutron induced gamma ray peaks for elements of the detector material should not interfere with the sample's spectral signatures. In addition, neutrons may produce radioactive activation products with the time delayed decay inside the detector volume. These decays (for

The reaction cross section is 3.4×10-2 barn at 100 keV, which is similar to the cross section of the d-d reaction. The distribution of energy between reaction products varies producing the wide neutron spectrum with maximum energy up to ~11 MeV. Wide neutron spectrum may be useful for material analysis applications that require both low and high energy neutrons. The compact "sealed tube" neutron generator design includes an ion source, a positive ion accelerator, and a target. Commonly used source to generate positive ions is the cold cathode Penning source. This source has a cylindrical anode under ~1-2 kV potential applied to it, and grounded cathodes on the ends of the anode. The magnet surrounds the anode cylinder setting up the coaxial magnetic field inside it. The tritium or deuterium gas is introduced into the volume of the anode cylinder. The electric field between anode and cathodes causes ionization of gas molecules creating the cold plasma. Trapped inside the anode electrons are moving in the volume and ionizing gas molecules which helps to maintain the plasma quality. Ions are transferred into the acceleration region through the exit port of the cathode. This region supplies the electric field (up to ~100-120 kV) to accelerate the positive ions. Neutron generator target is a metal hydride loaded with deuterium or tritium or the mixture of both. The ions interact with a target, producing neutrons in fusion reactions. Typical neutron output levels of sealed tube neutron generators are ~108-109 n/s (d-t) and ~106 n/s (d-d). Higher output usually shortens the sealed tube's life time. Sealed tube neutron generators are produced by Thermo Fisher Scientific, Schlumberger, Baker Hughes, EADS SODERN in France, and VNIIA in Russia.

Other designs of neutron generators utilize ion sources such as hot cathode source, radiofrequency ion source, or inertial electrostatic confinement (IEC) based source. For example, the pulse d-d system produced by Adelphi Technology Inc. (Williams et al., 2008) using the microwave driven plasma source technique provides the 2.45-MeV neutron output up to 8×109 n/s. The similar technique using tritium provides the yield of 14.1-MeV neutrons ~1011 n/s. The d-t neutron generator developed by NSD-Fusion GmbH for irradiation of extended samples uses the IEC technique producing ~1010 n/s. These neutron

Physical parameters and limitations of the gamma ray detectors used in the system govern parameters of the entire system. The choice of gamma ray detectors is important for neutron based system to be effective. Improper solution can generate both false positive and false

The gamma ray detectors must be suitable for operation in mixed radiation fields where neutrons and gamma rays present. The detector material must have a high Z value to effectively detect characteristic photons with energies up to 10.8 MeV. The detection medium must also provide the energy resolution that allows resolving peaks of interest. Ideally, the detector should provide minimum interference with the signal emitted from a sample when the detector material is irradiated with neutrons. Thus, if possible the detector material should avoid isotopes that are anticipated in the analyzed samples. The neutron induced gamma ray peaks for elements of the detector material should not interfere with the sample's spectral signatures. In addition, neutrons may produce radioactive activation products with the time delayed decay inside the detector volume. These decays (for

generator designs have the longer life time compared to sealed tube sources.

**2.2 Gamma ray detectors** 

negative results.

example, the beta decay) may produce photons or charged particles that interfere with the characteristic gamma ray spectra adding the noise and overloading the data acquisition electronics. It is a complicated task to satisfy all these requirements, especially with added cost limitations. Usually, the trade-off between various detector parameters including its cost is considered for a particular application (Barzilov & Womble, 2006).

Standard gamma ray detector solutions for spectroscopy are high purity germanium detectors (HPGe) with liquid nitrogen dewar or mechanical cooling subsystems (Sangsingkeow et al., 2003), and scintillation detectors such as NaI(Tl), Bi4Ge3O12, LaBr3(Ce) (van Loef et al., 2001), etc. Noble gas scintillation or ionization detectors and gamma ray telescopes can also be used for neutron induced photon measurements in the MeV energy range (April et al., 2006).

The NaI(Tl) scintillator material has the light yield 38 photons/keV, 1/e decay time 250 ns, and density 3.67 g/cc. Atomic numbers are 53 and 11 for iodine and sodium, respectively. Under neutron irradiation, the NaI(Tl) scintillator is activated by neutrons showing the delayed beta decay spectral continuum with the endpoint energy ~2 MeV.

The BGO scintillator has the light yield 9 photons/keV, and 1/e decay time 300 ns. Due to the high atomic number of bismuth, 83, and the crystal's high density of 7.13 g/cc, the BGO scintillator is very effective for detection of high energy photons. Its energy resolution is lower than NaI(Tl) resolution: ~10% FWHM versus ~7% FWHM for 662-keV -rays. The BGO demonstrates excellent behaviour under neutron irradiation without delayed decay issues. Significant downside of the BGO detector is its sensitivity to the environmental temperature (Womble et al., 2002).

The LaBr3(Ce) scintillator has the ~3%-resolution of the 662-keV peak, and density 5.08 g/cc. The lanthanum atomic number is 57. This scintillator has high light yield 63 photons/keV, fast 1/e decay time 16 ns, and better timing properties than NaI(Tl). The LaBr3(Ce) material contains small quantities of the radioactive lanthanum-138 isotope (t1/2=1.02×1011 years) producing the 1.47-MeV gamma ray peak that is always visible in the spectrum; it can be used for calibration purposes. The LaBr3(Ce) is affected by neutrons showing the delayed beta decay spectral continuum with endpoint energy ~3 MeV when irradiated with a d-t neutron source. The measured β-decay curve exhibits cumulative nature: two isotopes decay at the same time. The 80Br decays with a half-life 17.68 minutes. The 82Br isotope decays with the half-life approximately 35 hours. Lanthanum halide demonstrates stable gamma ray spectrum parameters in the mixed field under d-t neutron irradiation, when properly shielded. The good energy resolution under the room temperature, the high brightness, and the high scintillation decay speed pose this material as a promising candidate for active neutron interrogation applications, if the crystal's neutron activation issues are properly addressed.

The HPGe detector has superior energy resolution comparing to scintillation detectors. The atomic number of germanium is 32. The HPGe crystal density is 5.35 g/cc. HPGe crystal is sensitive to the high energy neutrons, which cause detector damage (Tsoulfanidis & Landsberger, 2010). High energy neutrons produce charges in the germanium crystal which are adding noise to the collected gamma ray spectrum (Ljungvall & Nyberg, 2005). Neutron collisions with the crystal cause atom displacements into interstitial positions creating a vacancy pair. These crystal defects behave as trapping

Material Analysis Using Characteristic Gamma Rays Induced by Neutrons 23

Neutrons emitted in d-d (En=2.45 MeV) and d-t (En=14.1 MeV) fusion reactions are highly penetrating particles. The typical range is several feet into materials commonly utilized in industry and commerce. Nuclear reactions energetically possible under 14.1-MeV fusion neutron's action in the volume of the irradiated object are the following: (n,n'), (n,), (n,), (n,p), (n,d), (n,t), (n,2p), (n,n'p), (n,n'), (n,3He), and (n,2n). If the sample contains heavy nuclei, (n,3n) and nuclear fission reactions may be induced with the low probability. Production of charged particles is prevailing for light nuclei; neutron production is favourable for heavier nuclei. The reactions (n,d) and (n,t) have noticeable cross-section for light mass isotopes, but products produced in such reactions are stable. The (n,d) and (n,t)

Widely used in material analysis neutron induced nuclear reactions are inelastic neutron scattering (n,n'), thermal neutron capture (n,), and neutron activation (n,) and (n,p). The only source of fast neutrons is a fusion neutron source. Thermal neutrons are created by slowing down the fast source neutrons in collisions with low Z materials within the sample itself or within the environment around the sample, or by using neutron moderating materials.

Isotope total inl n-n' 1st n-n' 2nd n-n' 3rd n, n,p 1H 692.0 0.0 0.0 0.0 0.0 0.0 0.0 12C 1303.2 426.9 184.7 0.9 9.9 72.7 0.2 14N 1628.6 399.3 14.9 26.7 15.3 60.1 54.0 16O 1611.1 508.5 27.0 82.5 43.0 109.0 43.7 19F 1740.7 164.2 0.3 36.8 0.3 21.3 14.7 31P 1831.7 53.9 0.2 0.2 0.1 126.9 91.9 32S 1829.7 378.9 99.3 10.3 18.0 159.6 247.4 35Cl 2100.0 820.0 5.8 5.2 12.0 137.3 98.0 75As 3456.2 685.1 0.8 0.5 7.3 10.1 19.0

Table 2. 14.1-MeV neutron induced nuclear reaction cross sections (in millibarns):

section; and n,p – the (n,p) cross-section

tot – the total neutron cross-section; inl – the inelastic neutron cross-section; n-n' 1st level – the (n,n') cross-section which excites the nucleons to the first nuclear level; n-n' 2nd level – the (n,n') cross-section which excites the nucleons to the second nuclear level; n-n' 3rd level – the (n,n') cross-section which excites the nucleons to the third nuclear level; n,– the (n,) cross-

Isotope total inl n-n' 1st n-n' 2nd n-n' 3rd n, n,p 1H 2683.6 0.0 0.0 0.0 0.0 0.0 0.0 12C 1595.3 0.0 0.0 0.0 0.0 0.0 0.0 14N 1512.6 0.0 0.0 0.0 0.0 70.2 22.4 16O 561.4 0.0 0.0 0.0 0.0 0.0 0.0 19F 2763.5 995.3 246.3 346.8 99.4 0.01 0.0 31P 3036.1 448.3 448.3 0.0 0.0 0.0 30.8 32S 3422.6 6.9 0.0 0.0 0.0 129.9 58.2 35Cl 3050.4 428.3 124.4 243.9 0.0 4.1 32.0 75As 3238.3 1728.5 37.0 60.0 78.4 0.0 0.02

Table 3. 2.45-MeV neutron induced nuclear reaction cross sections (in millibarns)

**3. Nuclear reactions induced by neutrons** 

reaction cross sections for medium and heavier mass nuclei are low.

centers for holes and electrons, and may create new donor and acceptor states, thus gradually changing the charge collection efficiency, the resolution, and the pulse timing characteristics of the detector. The n-type HPGe detectors are preferable in applications that involve neutron irradiation. They have been shown to be more resistant to damage by fast neutrons (Pehl et al., 1979). The neutron damage problem requires special attention and treatment (Fourches et al., 1991). The speed of the HPGe charge collection is another parameter to be considered in high count rate conditions and applications that require good timing resolution (Cooper & Koltick, 2001).

The comparison of selected gamma ray detectors used in neutron-based material analysis applications is shown in Table 1.

The shielding is required to protect the gamma ray detector from direct hit by the neutrons. Shielding size defines the geometry of the system since a neutron source and a gamma ray detector are separated by the shielding column. The combination of materials with large scattering cross sections for fast neutrons and large low energy neutron capture cross sections, and high Z materials with high stopping power for gamma rays is used. The goal is to keep fast neutrons away from the detector volume either by redirecting their path or moderating them with the subsequent capture. The d-d or d-t targets are in general of the "point source" type, thus the shielding may have a conical shape to minimize the weight. For 108-n/s d-t source, the simplest "shadow" shielding is a layered conical structure of ~50 cm length; the 30-40 cm borated polyethylene layer near the source, and the 10-20 cm lead layer near the gamma ray detector (Womble et al., 2003). The more complex shielding designs are possible using layers of other materials, but the size / weight / cost considerations add design limitations. In addition, the detector may be also shielded from lower energy neutrons scattered from surrounding materials. The two-layer shielding can reduce spectral noise due to low energy neutron interactions with the detector crystal. The outer layer of borated resin is effective as a thermal neutron shielding; the inner lead layer attenuates photons emitted from thermal neutron capture reactions in the outer layer. The lead also attenuates low energy photons that are not of interest in material analysis thus helping to reduce dead time of the gamma ray spectroscopy system.


Table 1. Gamma ray detectors used in neutron-based material analysis applications

Data acquisition electronics used with the gamma ray detectors in such systems should be appropriate for the detector's signal processing and count rates attainable in neutron interrogation. Standard analog and digital spectroscopy solutions are typically used.
