Neutron Detection

  • Alfred KlettEmail author
Living reference work entry


Neutrons are electrically neutral particles and therefore they are mainly subject to hadronic but not to electric forces. As neutrons are not directly ionizing, they have usually to be converted into charged particles before they can be detected. The basic physical principles for neutron detection are the neutron’s characteristic properties and several important nuclear reactions and processes.

The most important active neutron detector types are gas-filled, scintillation, and semiconducting detectors and the most important passive neutron detector types are thermoluminescent, etched-track, and nuclear-emulsion detectors. Specials techniques like the superheated emulsion detectors are unique in neutron detection.

Neutron detection covers a wide variety of applications in nuclear physics, in neutron scattering for biological, chemical, medical, and material analysis research, in metrology, in radiation protection, in nuclear energy and in the nuclear fuel cycle, in reactor instrumentation, in nuclear decommissioning and nuclear waste, in homeland security, in safeguards, in fusion monitoring, and in industrial measurements. Several important concepts and techniques as 3He proportional counters, rem counters, Bonner sphere spectrometers, tissue-equivalent proportional counters, time-of-flight measurement, and neutron activation analysis are described and discussed.

A few examples and results of neutron detection in aircrew dosimetry at flight altitudes, measurements in accelerator environments, and industrial measurements illustrate the diversity of neutron detection applications. As neutron detection and measurement requires calibration facilities and procedures, neutron reference fields are also discussed.


Neutron detection was earlier only a small niche in radiation detection but since about one decade its importance has considerably grown. Several fields dealing directly or indirectly with neutrons have been newly developed or extended. There is now a large number of new research facilities like synchrotron radiation sources, free electron lasers, fusion test facilities, neutron spallation sources, and other neutron sources. For medical applications in many countries, the installation of hadron therapy facilities with proton or heavy-ion accelerators has been started. Nuclear decommissioning and nuclear waste are growing, intermediate storage facilities have been commissioned, and spent-fuel transports require more neutron measurements. There is also more environmental monitoring and for instance an increased awareness about exposures at aircraft altitudes. Nuclear safety issues have triggered a large amount of homeland security activities and neutron detection techniques are utilized in the search for illicit trafficking nuclear materials in border control, in gate monitoring, and in the detection of special nuclear materials. Thus, recent developments and new technologies together with sometimes even reanimated classical nuclear knowledge are making neutron detection an exciting field.

Fundamental Neutron Physics

The Neutron

The neutron was discovered relatively late in 1932 by Chadwick, as neutron detection is not very easy (Chadwick 1932a, b). The neutron has total electric charge zero, while the proton carries one positive elementary unit of electric charge. As a consequence, the neutron is not subject to electrical force. Neutrons are primarily interacting by hadronic interactions with other particles. The hadronic interactions are much stronger than electromagnetic interactions, but they have only a short range of a few fermi. Neutrons are as neutral particles not directly ionizing when traversing through matter. They are only ionizing indirectly by generating charged particles or photons in hadronic scattering process or reactions. In the quark model the neutron is just like the proton described as a composition of three quarks glued together with gluons. Both nucleons are extended particles with electrical charge density distributions which can be characterized by mean electric and magnetic radii. The mass of the neutron is slightly exceeding the proton’s mass. Both nucleons carry spin, an internal angular momentum (see Table 1). The rotating charge density distributions are generating the neutron’s and the proton’s magnetic momenta. The neutron exhibits in the absence of electrical forces the weaker magnetic interaction and is therefore also useful as a magnetic probe.
Table 1

Physical data neutron–proton (Amsler et al. 2008)





939.5653 MeV

938.2720 MeV

Atomic mass unit [u]



Electric charge






Magnetic moment

−1. 913,042 μ N

+2. 792,775 μ N

Mean lifetime τ

885.7 s

>2. 1 × 10 29years

Mean electric charge radius

0.12 fm

0.88 fm

Neutrons are remarkable objects. They are stable as bound particles in nuclei, but they are unstable as free particles. They have a beta decay mode with a half-life of approximately 10 min. As neutrons are not directly ionizing, their detection is more complicated than charged-particle or photon detection. They have usually first to be converted into charged particles. Effective conversion via nuclear reactions requires in many cases slowing down of fast neutrons, because the cross sections are only sufficiently high at low neutron energies.

The particular interactions with other particles and the huge energy range over more than 15 orders of magnitude from ultracold neutrons below meV energies up to exceeding the TeV domain at large accelerators or in cosmic radiation are making neutron physics so diversified and cause a variety of phenomena and different detection techniques. In Table 2, there are a few common designations of neutron energy ranges. Neutrons are conventionally called slow neutrons if they are below the so-called cadmium cutoff energy at 0.5 eV and fast neutrons if they are above this value. In the terminology of the ICRP and the ICRU, neutrons of all energies are considered to be strongly penetrating radiation. Thermal neutrons are very important, because there are several efficient nuclear reactions which allow for thermal-neutron detection and because thermal neutrons induce some of the frequently used fission processes.
Table 2

Neutron energy ranges


Neutron energy

Wavelength λ [pm]

Ultracold neutrons

<0.2 meV


Cold neutrons

0.2 meV–2 meV


Thermal neutrons

2 meV–100 meV


Epithermal neutrons

100 meV–1 eV


Intermediate neutrons

1 eV–10 keV


Fast neutrons

10 keV–20 MeV


High-energy neutrons

>20 MeV


Basic Neutron Interactions

The interactions of neutrons with other nuclei (Evans 1982; Wirtz and Beckurts 1964) can be grouped into the following processes:
  • Elastic scattering

  • Inelastic scattering

  • Radiative neutron capture

  • Other nuclear reactions

  • Fission (inclusive of spontaneous fission)

  • Spallation

These processes are not only the base for most neutron detection mechanisms but also for neutron shielding and for the generation of any radiation hazard in bio-systems. Besides elastic and inelastic scattering neutron capture and several neutron-induced nuclear reactions are of particular importance in neutron detection. The most important nuclear processes with thermal neutrons are listed in Tables 3, 4, 5, and 6.
Table 3

Important thermal-neutron-induced nuclear processes (Knoll 2010)

Nuclear reaction

Q-value [MeV]

Cross section [barn]

Natural abundance

3He (n, p) 3H



Not applicable

6Li (n, α)3H




10B (n, α)7Li







Natural Cd

157Gd(n, γ)158Gd




235U fission

≈ 210



197Au(n, γ)198Au




Table 4

Examples: technical data of scintillators for neutron detection



6Li glass

Scintillating fiber

Density [g∕cm 3]




Index of refraction




λmax [nm]




Decay time [ns]






St. Gobain


Table 5

Materials for neutron threshold activation (Wirtz and Beckurts 1964; Knoll 2010; IAEA 1974)



Isotopic abundance [%]

Half-life [min]

γ Energy [MeV]

Yield [%]

Thresh [MeV]


24Mg(n, p)24Na







27Al(n, α)24Nas







27Al(n, p)27Mg







56Fe(n, p)56Mn







59Co(n, α)56Mn







58Ni(n, 2n)57Ni







58Ni(n, p)58Co







63Cu(n, 2n)62Cu







65Cu(n, 2n)64Cu







197Au(n, 2n)196Au






Table 6

Materials for slow-neutron activation (Wirtz and Beckurts 1964; Knoll 2010; IAEA 1974)



Isotopic abundance [%]

Half-life [min]

Thermal cross section [barn]


55Mn (n, γ)56Mn





59Co (n, γ)60mCo





63Cu (n, γ)64Cu





65Cu (n, γ)66Cu





107Ag (n, γ)108Ag





109Ag(n, γ)110Ag





113In(n, γ)114mln





197Au(n, γ)198Au




Because of the extremely short range of the hadronic forces, neutrons have to come very close within ≈10−15 m of a nucleus before any interaction can take place. For a neutron in normal matter, there is a lot of empty space and therefore interactions have relatively low probability and the neutron is a very penetrating particle. The probabilities of neutron interactions are characterized by their cross sections and are usually strongly depending on the neutron energy. The Q-value is a measure for the energy released to the reaction products. The higher the Q-value, the easier is detection and discrimination against gamma radiation. The most important process for fast-neutron detection and also for neutron moderation is elastic scattering on light target nuclei especially elastic n–p scattering.

Spallation is an inelastic interaction of a projectile, for instance, a proton or a neutron with high kinetic energy exceeding 100 MeV with a heavy nucleus. In the first fast stage, the projectile interacts with individual nucleons of the target nucleus and several nucleons are leaving the nucleus with high energies preferably in the forward direction. In the second slower stage, the energy in the residual nucleus is distributed across the other nucleons and neutrons and other particles at a typical energy scale of several MeV are evaporated with isotropic angular distributions. Spallation is used for neutron generation in spallation sources and is also used for the detection of high-energy neutrons. Spallation target materials are for instance tungsten or lead.

Neutron Generation

The main physical processes for neutron generation are fission, fusion, and nuclear reactions. The most important neutron sources or neutron-generating facilities are as follows:
  • Reactors

  • Accelerators

  • (α–n) Radionuclide sources like 241Am–Be, 239Pu–Be, 238Pu–Be, and 226Ra–Be

  • Spontaneous-fission radionuclide sources like 252Cf

  • Plasma neutron generators

  • Fusion facilities

  • Nuclear weapons

Reactors are very common as neutron sources and can deliver very high intensities. As neutrons cannot be accelerated directly by accelerators, they are generated by bombardment of appropriate target materials with charged projectiles. Nuclear (α–n) reactions are also important in neutron production. Especially if alpha particles are hitting 9Be nuclei there is a large probability for the generation of a neutron and 12C. The mixture of an alpha-emitting radionuclide with beryllium is therefore an excellent neutron source. Fusion processes are also efficient in neutron generation. Relatively new are plasma neutron generators which utilize the d–d or the d–t fusion reaction in a gas discharge tube. The former generates neutrons at energies around 2.5 MeV and the latter at about 14 MeV. Details of several of these processes and reactions are summarized below in section “Reference Neutron Radiation Fields”. Another neutron source is spent nuclear fuel. There are also significant amounts of neutrons in the secondary cosmic radiation in the atmosphere.

Neutron Moderation

As there are no electric forces acting between neutrons and matter, energy can only be transferred from neutrons to other particles by hadronic interactions. As a consequence of the conservation of energy and momentum, the maximum energy transfer in elastic collisions of projectiles with target nuclei depends on the particles’ masses. The maximum energy transfer in elastic neutron scattering occurs in collisions with protons or other light nuclei. This is efficient up to the 12C nucleus. In order to transfer kinetic energy from neutrons to other particles elastic neutron scattering on heavier target nuclei is relatively inefficient. Multiple elastic neutron scattering in a material containing light nuclei reduces the kinetic energy of incoming neutrons considerably and is called moderation.

If moderated neutrons are in thermal equilibrium with the surrounding materials, they are called thermal neutrons. The mean kinetic energy of thermal neutrons is about 0.025 eV. Neutron moderation is important because thermal neutrons can easily be detected or absorbed and because thermal neutrons can induce some of the important fission processes. Good moderators are materials with a large amount of hydrogen, for instance, polyethylene or water. Deuterium is also a suitable moderator because it has a lower neutron absorption cross section than hydrogen. Of course the number of collisions required for thermalization is dependent on neutron energy. About 20 interactions are sufficient to thermalize a 1 MeV neutron in hydrogen. Beckurts and Wirtz (1964) gave a comprehensive description of neutron moderation.

Neutron Absorption and Shielding

A neutron absorber is a material with which neutrons interact significantly by nuclear reactions resulting in their disappearance as free particles. As direct absorption of fast neutrons has low probabilities, efficient neutron absorbers are only existing for thermal neutrons. Therefore, neutron shielding is usually a combination of moderation and successive absorption of thermal neutrons. The 10B reaction or absorption on natural cadmium, which is a mixture of several isotopes, is providing excellent thermal-neutron shielding. Cadmium of a few millimeters thickness is absorbing basically all neutrons below a cutoff energy of about 0.5 eV. The cross section of 10B is decreasing reciprocally to neutron velocity and boron is strongly absorbing slow neutrons. Other possible reactions are listed in Table 3.

Highly efficient neutron shielding for fast neutrons is either moderator material followed by layers of slow-neutron absorber or a mixture of moderating material with slow-neutron-absorbing material. Boronated polyethylene is a very useful neutron shielding material. Boron–silicone is a heat- and fire-resistant elastomer, which can be used as castable neutron shielding. Polycast is a dry mix material designed to be cast into closed containers. It is field castable, providing excellent, low-cost neutron shielding, with a hydrogen content 6% greater than that of water. Neutron putty is a nonhardening boron-loaded putty with a high hydrogen content. Neutron shielding is available as sheets, plates, rods, or pellets. Other neutron shielding materials in use are water, concrete, soil, and steel.

Metrology and Dosimetric Quantities

Radiation measurements and investigations of radiation effects require the definition of radiometric quantities (ICRU 1998a). Radiation fields are characterized by radiometric quantities which apply in free space as well as in matter. The particle number Nis the number of particles that are emitted, transferred, or received. The flux is the quotient of d N/d t where d Nis the increment of the particle number in the time interval d t. One of the most important quantities in neutron detection is the fluence ϕ, which is the ratio of the number d N of particles incident on a sphere of cross-sectional area d a:
$$ \dot{\phi}=\frac{\mathrm{d}N}{\mathrm{d}a}. $$
The fluence is measured in unit m−2. The fluence rate is defined as
$$ \dot{\phi}=\frac{\mathrm{d}\phi }{\mathrm{d}t} $$
and has unit m−2 s−1. The distribution ϕ E represents the fluence with respect to energy where dϕ is the fluence of particles of energy between E and E+ d E:
$$ {\dot{\phi}}_E=\frac{\mathrm{d}\phi }{\mathrm{d}E}. $$

Dosimetric quantities should provide physical measures which are correlated with effects of ionizing radiation. The basic dose definitions are given in chapter “Radiation Protection”. Operational quantities for practical measurements, both for area and for individual monitoring, were introduced and further explained in ICRU Reports 39, 43, 47, and 51 (ICRU 1985, 1988, 1992, 1993). The International Commission on Radiation Protection has recommended their use in radiation protection measurements (ICRP 1991, 2007). They are based on the quantity dose equivalent and have the unit Sievert (Sv). The ICRU has provided definitions of the operational quantities at points at a depth din phantoms made out of tissue-like materials. For strongly penetrating radiation as neutrons, the depth din the phantom is 10 mm. The operational quantities for strongly penetrating radiation are for area monitoring the ambient dose equivalent H(10)and for individual monitoring the personal dose equivalent Hp(10). The ICRU sphere with diameter 30 cm is the phantom for H(10), while a slab phantom with dimensions 30 cm × 30 cm × 15 cmis used for the calibration to Hp(10). The ICRU material has a mass density of 1 g cm−3 and a mass composition of 76.2% oxygen, 11.1% carbon, 10.1% hydrogen, and 2.6% nitrogen (ICRU 2001).

The relation between radiometric quantities and the operational quantities is established by fluence-to-dose-equivalent conversion factors. The operational dose-equivalent quantities H(10) and Hp(10) for neutrons are determined from the equations
$$ {H}^{\ast }(10)=\int {h}_{\phi}^{\ast }(E){\phi}_E\mathrm{d}E, $$
$$ {H}_{\mathrm{p}}(10)=\int {h}_{p,\phi }(E){\phi}_E\;\mathrm{d}E, $$
where ϕE is the energy distribution of the neutron fluence and hp, ϕ(E) and \( {h}_{\phi}^{\ast }(E) \) are the corresponding energy-dependent fluence-to-dose-equivalent conversion coefficients. These coefficients were calculated by several groups and at an international level it was agreed upon the numerical values that can be found in ICRP Report 74 (ICRP 1996) and in ICRU Report 57 (ICRU 1998b). The conversion factor \( {h}_{\phi}^{\ast }(E) \) as a function of neutron energy is displayed in Fig. 1.
Fig. 1

Fluence-to-ambient-dose-equivalent H(10) conversion coefficients for neutrons (ICRP 1996; Ferrari and Pelliccioni 1998; Sannikov and Savitskaya 1997)

The fluence response Rϕ of a radiation detector is a useful quantity specifying its sensitivity for detection. For an irradiation in a homogeneous radiation field with fluence ϕ it is defined as (ISO 8529-1 2001)
$$ {R}_{\phi }=\frac{n}{\phi }, $$
where nis the total count of detected events. Fluence responses are measured in units of area, usually in cm2. The fluence response corresponds to the area of a hypothetical detector with 100% efficiency.
A radiation detector’s response RH to dose equivalent H is defined as
$$ {R}_H=\frac{R_{\phi }}{h_{\phi }}. $$

Materials and Detector Types for Neutron Detection

Neutron Detection Principles

Ionizing radiation is a radiation consisting of directly or indirectly ionizing particles. A radiation detector is a device which in the presence of radiation provides a signal for use in measuring one or several quantities of the incident radiation. The detection of ionizing radiation usually utilizes ionization effects in detector materials. As neutrons are not directly ionizing they have to be converted into charged particles, which are then transferring their energy in direct ionization processes to the detector’s sensitive volume. The ions are subject to charge collection and sometimes to internal amplification processes. Proportional counters, Geiger–Müller counters, GEM detectors, and photomultipliers are using avalanche charge amplification sometimes at very large gains. Pulse analysis and discrimination is easier with large signals. Other detector types like ionization chambers or semiconductors are only collecting the charge. Glenn Knoll gave an excellent and very detailed overview on radiation detection (Knoll 2010).

Active Neutron Detection Methods

Gas-Filled Detectors

Gas is a well-suited medium for the detection of ionizing radiation. Free electric charges are mobile in gases and their recombination probability is relatively low. The conductivity of gases is low enough that high voltages can be applied to produce sufficiently strong electric fields. Especially noble gases are well-suited components of counting gases. Quenching gas admixtures are required if avalanche gas amplification is used. This holds for the detection of all types of ionizing radiation.

For neutron detection an efficient conversion process of neutrons into charged particles has to be added. There are a few gases where nuclear reactions in Table 3 provide efficient neutron conversion. The most important counting gases in neutron detection are 3He, BF3, methane, and hydrogen. The reactions with 3He and 10B have sufficiently large cross sections and high Q-values to convert slow neutrons with high probability into charged particles with enough kinetic energy to exceed detection thresholds. In hydrogen or in methane, neutrons are producing recoil protons. The neutron energy has to be not too low, because the recoil energy has also to be above detection threshold and in elastic scattering there is no energy contribution from a Q-value. In neutron detection, the most common gas-operated detector types are proportional counters, ionization chambers, and fission chambers.

The working horse in neutron detection is certainly the 3He proportional counter. Cylindrical tubes are available with diameters from a fraction of inch to several inches. There are also spherical counters and detectors with rectangular cross sections for time-of-flight spectroscopy. Tube lengths are ranging between a couple of centimeters up to several meters and 3He filling pressures are up to 20 atm with small amounts of quenching gas. 3He counters are rigid and can be operated at temperatures up to 200 °C. The efficiencies for thermal-neutron detection are high. The 3He(n, p) 3H reaction releases a total of 764 keV kinetic energy as indicated by the Q-value in Table 3. According to the conservation of energy and momentum, the triton carries 191 keV and the proton 573 keV. Both particles are directly ionizing along their tracks through the gas volume and their total energy deposit is 764 keV. A typical pulse-height spectrum of a 3He counter tube is shown in Fig. 2. with the full-energy-deposit peak at 764 keV. The tails at lower pulse heights correspond to events where one of the two decay particles has hit the counter tube’s wall and only a fraction of the full energy release is detected. Only one particle can hit the wall because the angle between both decay particles is 180°. The triton escape and the proton escape can clearly be seen in the pulse-height spectrum. All detected thermal-neutron signals are exceeding a minimum energy deposit. Below this minimum, there is a gap. At very small pulse heights there are signals generated by gamma radiation. The discrimination threshold for the electronics is usually set above the gamma pulse heights and below the minimum neutron energy deposit.
Fig. 2

Pulse-height spectrum of a 3He counter tube with diameter 1 and filling pressure 3.5 atm

Detectors filled with BF3 have lower efficiencies than 3He detectors due to the lower cross section and because the filling pressure is usually lower. But they have a better gamma discrimination because of the larger Q-value. Drawbacks of BF3 are that the gas is toxic and corrosive. Most BF3 counters are filled with pure boron tri-fluoride enriched to about 96% in 10B.

Boron-lined proportional counters have a similar construction to BF3 and 3He proportional counters. However, the neutron detection is by means of a boron coating rather than boron or 3He in a gaseous form, resulting in a higher neutron sensitivity. Typically, boron-lined proportional counters are used where the temperature limitations of BF 3counters prevent their use. Detectors filled with hydrogen or with methane are used as recoil proton counters.


The following semiconducting materials have been used in neutron detection:
  • Silicon with 10B coating or with 6LiF film

  • Gallium arsenide with 10B coating

  • Boron–carbide semiconductor diodes

The advantages of semiconductors for neutron detection are mainly compact size, relatively fast timing characteristics, and an effective thickness that can be varied to match the requirements of the application. Drawbacks may be the limitation to small sizes and the relative high susceptibility to performance degradation from radiation-induced damage (Knoll 2010).


The following scintillators have been used in slow-neutron detection:
  • Boron-loaded plastic and liquid scintillators

  • 6Li scintillators (LiF, LiI, LiFZnS(Ag), glass, scintillating fibers)

  • Gadolinium-loaded liquid scintillators

Advantages of scintillators are that they can be sensitive to the amount of energy deposit and that they are fast detectors which can be used for time-of-flight measurement. Scintillators are robust, easy to be operated, and relatively cheap. Their disadvantages are aging effects and radiation damage, sometimes difficulties in the light detection, for instance, with photomultipliers in the presence of magnetic fields and some scintillators are hygroscopic.

The dominant fast-neutron interaction in plastic or in liquid scintillators is the generation of recoil protons. Plastic scintillators consist out of a solid solution of organic scintillating molecules in a polymerized solvent. They are very popular because of the ease with which they can be fabricated and shaped. Typical emission is at 400 nm. They have a large light output and short decay time and are well suited for timing measurement. Many scintillator designations are following the Saint-Gobain-type designation. A plastic scintillator for fast neutrons would be BC-720. Plastic scintillators do not allow for a good neutron–gamma separation.

The BC-501A (formerly called NE 213) is a widely used liquid scintillator with good pulse-shape discrimination properties intended for neutron detection in the presence of gamma radiation. It has extremely good timing properties and is well suited for coincidence measurements. Then there are for slow-neutron detection boron-loaded plastic scintillators (BC-454) and gadolinium (BC-521) and natural (BC-454) or enriched boron-loaded liquid scintillators (BC-523A). Also, scintillators with lithium are quite common. The lithium-iodide crystal is chemically similar to sodium iodide and also hygroscopic. Liquid scintillator with lithium is also commercially available. There are as well lithium-containing glass scintillators. A new type of a scintillation detector for neutrons is scintillating glass fibers loaded with lithium. Anthracene and stilbene also have been used for neutron detection. For neutron radiography there are scintillators with phosphor screen based on ZnS(Ag) and 6Li (BC-704 earlier NE-426).

Superheated Emulsion Detectors

Robert E. Apfel proposed in 1979 the superheated emulsion detectors, the so-called bubble technology as a new method for radiation detection (Apfel 1979). Superheated emulsion detectors are based on superheated droplets suspended in a viscoelastic gel medium, which vaporizes upon exposure to the high-LET recoils from neutron interactions. Bubbles evolved from the radiation-induced nucleation of drops give an integrated measure of the total neutron exposure. There are several different techniques to record and count the bubbles. In active devices, they can be detected acoustically, by optical bubble counting, or by vapor volume measurement. Neutron spectrometry can be performed by measuring responses at different temperatures or pressures. Bubble detectors are insensitive to low-LET radiation like gammas or X-rays. Acoustical recording has the issue with discrimination of bubble pulses against noise. Francesco d’Errico published overviews concerning superheated emulsion detectors (d’Errico et al. 1995, 2002; d’Errico 2001).

Passive Neutron Detection Methods

Track Detectors

Passive neutron detection with nuclear track emulsions is the oldest and was once the most common method for neutron personal dosimetry (Knoll 2010; ICRU 2001; d’Errico and Bos 2004). The emulsions are relatively inexpensive, but track analysis under a microscope is laborious. New developments are focusing on automated track scanning methods. Passive radiation detectors have the advantage of being able to measure also in pulsed radiation fields where active devices may suffer from dead-time losses or pulse pile-up.

Thermoluminescent Dosimeters

In thermoluminescent dosimeters (TLDs) (Knoll 2010; ICRU 2001; d’Errico and Bos 2004), electrons are elevated by a radiation from the valence to the conduction band and captured in trapping centers. Holes can also be trapped in analogous processes. The captured states are stable for longer periods. If a TLD is heated, the trapped electrons or holes are re-excited and emit visible photons, which can be detected by a photomultiplier. The number of photons is a measure of the dose deposit. An exposed TLD material is thus an integrating detector for ionizing radiation. There are many TLD materials. LiF has been the most widely exploited (Knoll 2010). TLDs for neutron measurement are primarily used as albedo dosimeters, which is a dosimeter capable of measuring the fraction of neutrons reflected by a human body. Thermoluminescence detectors for neutron detection utilize typically 6LiF and 7LiF crystals. 6LiF is sensitive to neutrons and to photons while 7LiF is only sensitive to photons. The neutron contribution is calculated by determining the difference of both readings. TLDs are simple, rugged, and cheap. They have a good linearity and low detection limits. TLDs are usually processed by automatic readout.

Etched-Track Detectors

Etched-track detectors (Knoll 2010; ICRU 2001; d’Errico and Bos 2004) are together with TLDs the most commonly used passive neutron detectors. Charged particles, like alpha particles or protons, damage the material along their tracks, which can be made visible by chemical or electrochemical etching. These secondary particles can originate from nuclear reactions both in materials adjacent to an etched-track detector and those created inside the bulk of it. Etched-track detectors are usually processed by imaging systems which analyze and count the tracks and determine the dose. This method is insensitive to photons. In particularly over the last few tens of years, polymer etched-track detectors have been used. The most popular detector material is polyallyl-diglycol carbonate (PADC), commercially available as CR-39. 6Li or 10B are mainly used as converters for slow neutrons. Their response characteristics are generally sufficiently well known for neutrons with energies up to several hundred MeV. These dosimeters are generally able to determine neutron ambient dose equivalent down to a few tenths of a mSv.

Passive Superheated Emulsion Detectors

In passive superheated emulsion detectors or bubble detectors (ICRU 2001), the most immediate readout method is the visual inspection, a process that can be automated using video cameras and image analysis techniques. In the past decade superheated emulsions have achieved acceptance among the passive systems for personal neutron dosimetry. The detectors are considered to be the passive devices with the most accurate energy dependence of the response and the lowest detection threshold (d’Errico and Bos 2004).

Direct Ion Storage

Direct ion storage (DIS) is a relatively new technology. A small ionization chamber filled with air is in contact with the floating gate of a MOSFET transistor. The charge of the gate is initially set to a predetermined value. The charge generated by the radiation in the ionization chamber discharges partially the gate. The stored charge at the gate can be measured as a voltage without modifying its value and it is proportional to the dose. Application of DIS to neutron dosimetry is possible using pairs of detectors for the separate determination of the photon and neutron dose contribution (d’Errico and Bos 2004; Fiechtner et al. 2004).

Other Passive Detectors

Radioluminescent glass detectors have found limited application in neutron dosimetry. A more recent technique is optically stimulated luminescence (OSL) (McKeever 2001). This method is based on laser stimulation and does not need heating of the detector material. It seems to be a breakthrough in passive radiation detection (d’Errico and Bos 2004).

Applications of Neutron Detection

Neutron Dose Measurement


As a consequence of the 1990 recommendations of the International Commission on Radiological Protection, the operational quantities were newly defined (ICRP 1991). The relations between the neutron fluence and the two operational quantities, the ambient dose equivalent H(10) and the personal dose equivalent Hp(10), vary widely with neutron energy. The fluence response of a well-designed instrument, which will give a reading sufficiently proportional to the operational quantities, regardless of the neutron energy spectrum should have a fluence response as a function of energy that is inversely proportional to the fluence-to-dose conversion coefficients (Knoll 2010) provided by the International Commission on Radiological Protection (ICRP 1996). This is the entire secret of the art of accurate neutron dose measurement.

Rem Counters

The first neutron dose-rate meters based on the concept of an active thermal-neutron detector centered in an appropriate moderator with internal neutron absorbers – so-called rem counters – have been already designed in the 1960s. The Andersson–Braun counter and the Leake counter, became very popular and have been used all over the world for decades. While the Andersson–Braun counter has a cylindrical moderator and a BF3 proportional counter tube as neutron detector (Andersson and Braun 1963, 1964), the Leake counter has a spherical moderator with a reduced weight of about 5 kg and had a LiI(Eu) crystal at the center (Leake 1966). The Leake design was later improved by replacing the crystal by a small spherical proportional counter filled with 3He gas to achieve better gamma rejection properties and increased neutron sensitivity (Leake 1968).

After the publication of the ICRP60 recommendations of the International Commission on Radiological Protection (ICRP 1991), the Research Center Karlsruhe and Berthold Technologies designed a new neutron survey meter – the Berthold LB 6411 – with an energy-dependent response optimized to the then new operational quantity ambient dose equivalent H(10) (Klett and Burgkhardt 1997). The rem counter utilizes a cylindrical 3He proportional counter tube centered in a moderating polyethylene sphere with 25 cm diameter. The energy-dependent response to H(10) tuned with internal perforated-cadmium neutron absorbers and with boreholes is within ±30% for neutron energies between 50 keV and 10 MeV (Knoll 2010). This is standard in gamma dosimetry but it is excellent in neutron dosimetry. The H(10) response to neutrons emitted by a bare 252Cf source is approximately 3 counts/nSv, which is very high. Figure 3 shows a drawing of the geometrical setup.
Fig. 3

Schematic drawing of the Berthold rem counter LB 6411

Figure 4 shows the responses of several widely used rem counters to ambient dose equivalent H(10) which were calculated from the fluence responses given in a technical report of the IAEA (IAEA 2001). In the United Kingdom, the radiation protection group of the Health Protection Agency together with the neutron metrology group of the National Physical Laboratory have carefully assessed the performances of several instruments (Tanner et al. 2006).
Fig. 4

Responses of several neutron survey meters to ambient dose equivalent H(10)

For radiation protection purposes the response functions of conventional rem counters are considered to be acceptable at energies below 20 MeV. At higher energies, the responses are decreasing and the instruments are underestimating ambient dose equivalent. A growing number of accelerators with high or even very high energies and enhanced interest in dose monitoring at flight altitudes triggered novel designs of instruments for extended energy ranges (Birattari et al. 1998; Fehrenbacher et al. 2007; Klett et al. 2007). Extended-range rem counters utilize layers of lead, tungsten, or other high-Z materials to convert in spallation processes high-energy neutrons into lower-energy neutrons. Response functions of several extended-range rem counters were calculated by Mares et al. (2002).

Tissue-Equivalent Proportional Counters

Tissue-equivalent proportional counters (TEPCs) allow the measurement of the probability distribution of the absorbed dose d(y) in terms of lineal energy yin radiation fields. The lineal energy is defined as the ratio of the energy imparted to the matter in a volume by a single-deposition event to the mean chord length in that volume. The lineal energy can be used as an approximation of the linear energy transfer LET, and the dose equivalent can be evaluated through a function Q(y) which relates the quality factor to the lineal energy.

The TEPC is an important tool in microdosimetry and in some cases, the only one to provide directly dose and radiation quality information in complex radiation fields. The TEPC can be used to distinguish photon and neutron contributions with good accuracy.

A tissue-equivalent proportional counter is a spherical or cylindrical detector with walls made out of tissue-equivalent material and filled with tissue-equivalent gas and operated in proportional mode. When the TEPC’s gas density and its diameter is maintained at about 10−4 g cm−2, it can simultaneously determine the absorbed dose to the tissue and of the spectrum of the pulse heights which corresponds to the energy deposition (ICRU 2001). By far the most commonly used TEPC has been the spherical counter. It has the advantage of being isotropic with respect to external radiation, but the electric field close to the wire has unfortunately not a cylindrical geometry. In order to obtain a cylindrical electric field geometry in the volume of gas amplification, Rossi proposed the so-called Rossi counter with an auxiliary helix electrode close to the wire (Rossi and Staub 1949; Rossi and Rosenzweig 1955). The sensitive volumes of TEPCs range from a few millimeters up to several centimeters in diameter. TEPCs are operated in pulse mode to record each individual event’s energy deposit. The pulse height is proportional to the charge released in the sensitive detection volume. The measurement of lineal energies from below 100 eV/μm up to more than 1 MeV/μm requires low-noise analogue electronics with a linearity over 4–5 orders of magnitude and an appropriate ADC system. TEPCs have not only been used in dose measurements at aviation altitudes and in mixed photon–neutron fields in accelerator environments, but also in investigations in radiotherapy and radiobiology (Kliauga et al. 1995; Gerdung et al. 1995).

Active Personal Dosimeters

A relatively new development now competing with and partially replacing the passive methods in individual dose monitoring are active personal dosimeters (APDs). In comparison to passive dosimeters, they have the advantages of instant reading, audible alarm, lower detection limits, data memory, and communication capabilities with other hosts. There are several neutron APDs on the market, which have one or several silicon diodes or silicon strip detectors for neutron detection. The semiconducting detectors are combined with neutron converters or with layers of material for neutron activation, for instance, silver. Some use separate photon channels for the subtraction of gamma doses. The main issue with neutron APDs is their poor energy-dependent response. There were comparisons of instruments on the market and summaries published (Bolognese-Milsztajn et al. 2004; Luszik-Bhadra 2007). Recent measurements with electronic personal neutron dosimeters for high neutron energies were reported by Luszik-Bhadra (2007).

Passive Dose Measurement

There is a large amount of passive dose measurement in individual monitoring of occupational exposure. The passive neutron detectors that are used are described above. A good overview about the state of art of the available techniques was given by F. d’Errico and A.J.J. Bos (2004).

Dose Measurement in Pulsed Radiation Fields

Many accelerators or other radiation generators operate in pulsed mode. It is well known that active radiation detectors are subject dead to time effects and exhibit limitations in pulsed radiation fields (Knoll 2010). These limitations cannot easily be overcome without the development of new active detection technologies. Measurement of pulsed radiation is usually done with passive detectors.

There are now a few developments of new technologies based on the activation of radionuclides by pulsed radiation fields. One of these designs utilizes the neutron-induced activation of the nuclides 8Li, 9Li, and 12B with short half-lives below 200 ms on the target nucleus 12C in the detector materials. The decay products are detected in a time-resolved measurement. The instrument is mainly intended for radiation protection at accelerators with high energies and accomplishes the measurement of even very short and intense pulsed neutron fields (Klett and Leuschner 2007; Klett et al. 2010).

Luszik-Bhadra published another design of a new monitor for pulsed fields based on the activation of silver. The device comprises four silicon diodes in a 12 polyethylene moderator sphere, two diodes covered on both sides with Ag, and two diodes covered with tin. The decay products of the activation products 109Ag and 110Ag are beta particles which are detected by the semiconductors. The detectors covered with silver are sensitive to neutrons and photons, while the detectors covered with tin are only sensitive to photons. The neutron dose is determined by subtraction (Luszik-Bhadra 2010; Leake et al. 2010).

Examples of Neutron Dose Measurements

The intensities of radiation levels at flight altitudes are exceeding-ground-level intensities by two orders of magnitude. The exposure of aircrews is comparable with or even larger than the exposure of workers classified as occupationally exposed. Primary galactic and solar particles – mainly protons – are interacting with the atmosphere and are generating secondary particles with a complicated composition. At flight altitudes of civil aircrafts about 50% of the ambient-dose-equivalent contribution is from neutrons, about 35% is from photons, electrons, and muons, and about 15% is from protons. Accurate dose measurements in these mixed fields with energies ranging from keV up to even exceeding the TeV domain is difficult. The recommendation by the International Commission on Radiological Protection (ICRP) in 1990, that exposure to cosmic radiation in the operation of jet aircraft should be recognized as occupational exposure, initiated a large number of new dose measurements onboard aircraft. A EURADOS working group has brought together all recent, available, preferably published, experimental data and results of calculations, mainly from laboratories in Europe (Lindborg et al. 2004). The reported results have been obtained using a variety of instrument types like rem counters, TEPCs, and Bonner sphere spectrometers. The results obtained are in good agreement almost all within ±25% of the mean values. During the time period 1995–1998 at temperate northern latitudes in 10 km altitude measured ambient-dose-equivalent rates for neutrons were about 3 μSv/h and the total about 5 μSv/h. The total exposure on a typical trans-Atlantic flight is about 50 μSv (Luszik-Bhadra 2007).

Another interesting example of neutron dose measurement was an international project investigating complex workplace radiation fields at European high-energy accelerators and thermonuclear fusion facilities. This study included all common types of existing neutron detection techniques in an environment with mixed radiation fields and high energies. The relevant techniques and instrumentation employed for monitoring neutron and photon fields around high-energy accelerators were reviewed with some emphasis on recent developments to improve the response of neutron-measuring devices beyond 20 MeV. It was investigated which type of area monitors to be employed (active and/or passive) and how they should be calibrated. The influence of the pulsed structure of the beam on the instruments and the needs and problems arising for the calibration of devices for high-energy radiation are addressed. The major high-energy European accelerator facilities are reviewed along with the way workplace monitoring is organized at each of them. The facilities taken into consideration are research accelerators, hospital-based hadron therapy centers, and thermonuclear fusion facilities. The issues of calibration are discussed and an overview of the existing neutron calibration facilities was provided (Bilski et al. 2006; Rollet et al. 2009; Silari et al. 2009).



Neutrons appear in nature, in laboratories, or in nuclear facilities covering a very large energy range from ultracold up to ultrahigh energies at accelerators up to the TeV domain. As the interaction of neutrons with matter usually strongly depends on their energy, spectral information is needed in order to describe the occurring process. Spectrometry measurements are needed to characterize neutron fields. Commonly used measurement methods are Bonner sphere measurement, time-of-flight measurement, nuclear recoil measurement, neutron-induced nuclear reactions, methods based on activation and on threshold effects, and neutron diffraction. An excellent overview about neutron spectrometry for radiation protection was provided by Thomas (2004).

Bonner Spheres

In 1960, the Bonner sphere spectrometer (BSS) was first described by Bramblett, Ewing, and Bonner (Bramblett et al. 1960). Of the many types of neutron spectrometers that have been developed this multisphere system has been used by more laboratories than any other. It is easy to operate, it has an almost isotropic response, it covers energies from thermal up to GeV neutrons, and it can be used with active or with passive detectors. A BSS is consisting of several moderating spheres with different diameters and a thermal-neutron detector which is assembled in the centers of these spheres. The spheres are usually made out of polyethylene and each sphere with the thermal-neutron detector has a sensitivity to neutrons over a broad energy range. However, the sensitivity for each sphere peaks at a particular neutron energy depending on the sphere diameter. From the measured readings of a set of spheres, information can be derived about the spectrum of a neutron field (Thomas and Alevra 2002; Thomas and Klein 2003a).

Several types of thermal-neutron detectors have been used. In the original Bonner sphere spectrometer, a small 6LiI(Eu) scintillator was used. Various cylindrical and spherical proportional counter tubes filled with BF 3or 3He are obvious alternatives. Several groups investigated the use of the SP9 spherical 3He proportional counter produced by Centronic Ltd. UK. It has a diameter of 32 mm and a gas pressure of about 2 atm. The characteristics of BSSs with this detector are well established. Typical moderator sphere diameters are between 3 and 18 with the number of spheres in a set ranging between 6 and 12 spheres (Thomas and Alevra 2002).

If sphere i has response function R i (E) and is exposed in a neutron field with the spectral fluence ϕ(E), then the sphere reading M i is obtained mathematically by folding Ri(E) with ϕ(E):
$$ {M}_i=\int {R}_i(E)\phi (E)\;\mathrm{d}E. $$

This integral extends over the range of neutron energies present in the field. Good approximations of Ri (E) can be obtained by simulation calculations supported by measurements in well-characterized reference neutron fields. Information about the spectrum ϕ(E) can be extracted by unfolding. However, because the total number of spheres is limited, the solution may provide a poor representation of the spectrum with important features smeared out. Additional a priori information on the spectrum is useful (Thomas and Alevra 2002).

The PTB NEMUS system, the INFN Frascati BSS, and the NPL BSS are a few examples where detailed descriptions, measurements, and intercomparison data were published (Wiegel and Alevra 2002; Bedogni and Esposito 2009). Figure 5 shows the components of the PTB NEMUS Bonner sphere spectrometer. BSSs have an excellent energy range, good sensitivity, isotropy, and photon discrimination, simple but time-consuming operation, and poor energy resolution. According to a neutron field’s intensity and time structure, various types of active or passive thermal-neutron detectors can be selected. The data analysis requires complex unfolding of the measured data. For instance, FRUIT (Frascati Unfolding Interactive Tool), an unfolding code for Bonner sphere spectrometers, was developed under the Labview environment at the INFN-Frascati National Laboratory and is available from the authors upon request (Bedogni et al. 2007). An excellent overview on Bonner sphere neutron spectrometry was provided by Thomas and Alevra (2002).
Fig. 5

Bonner sphere spectrometer NEMUS of the PTB with five of the ten polyethylene spheres in the back, a spherical 3He proportional counter in the center, and parts of the modified spheres in the foreground left and right

Time-of-Flight Spectroscopy

Time-of-flight spectroscopy is based on measurement of the time it takes for a neutron to travel a known distance. From this the neutron’s velocity and the energy can be calculated. The measurement needs a start and a stop signal. The former can, for instance, be derived from a pulsed neutron generation process while the latter is generated when the neutron arrives at a distant neutron detector. The indication of the start can be generated by time-pick-up signals from an accelerator, by an appropriate detector, or by a beam chopper. To minimize uncertainties time-of-flight spectroscopy requires precise time measurement. Therefore, the detectors from which the time information is derived have to be very fast. Excellent timing characteristics have for instance plastic scintillators. The flight paths have to be very long and can be as long as the order of magnitude of 100 m (Fig. 6).
Fig. 6

Neutron spectra of bare 252Cf, moderated 252Cf, and 241Am–Be (IAEA 2001)

Recoil Spectroscopy

Another neutron detection technique with spectral sensitivity is based on elastic neutron scattering on light target nuclei. During the interaction a fraction of the neutron’s energy is transferred to the recoil nucleus. As the recoil nucleus is directly ionizing, it deposits its energy in the detector materials. The maximum energy Emax which can be transferred from a neutron with kinetic energy En to a recoil nucleus with mass A in units of neutron mass is (Knoll 2010):
$$ {E}_{\mathrm{max}}=\frac{4A}{{\left(1+A\right)}^2}{E}_{\mathrm{n}}. $$

The maximum fractional energy transfer in elastic neutron scattering is 1 for hydrogen, 0.64 for 4He, and 0.22 for 16O. Therefore, only light nuclei are of primary interest with hydrogen being the best choice. Recoil proton spectroscopy can be performed with detectors with a substantial amount of hydrogen in the detector material. The easiest detector would be a scintillator containing hydrogen as organic crystals, plastic scintillators, or liquid scintillators. Liquid scintillators have the advantage of the possibility for neutron–gamma discrimination. Another detector type for recoil spectroscopy would be gas recoil proportional counters filled with hydrogen, methane, or with helium. The energy distribution for all scattering angles in elastic neutron–proton scattering is approximately a rectangular function. This is the detector’s response function. The neutron spectrum has to be determined by deconvolution (Knoll 2010).

Neutron Activation Analysis

Neutron activation analysis (NAA) is one of the most sensitive analytical techniques to determine concentrations of many elements in a variety of materials. It is based on neutron activation and requires irradiation of the specimen with neutrons. This creates artificial radioisotopes of the elements of interest. Usually reactors are used but other types of neutron sources as previously discussed can also be used, if the energy and intensity requirements are met. The decay products of the artificial radioisotopes are then measured and analyzed. Preferably gamma spectroscopy allows for the identification of nuclides and for quantitative measurement of concentrations. As it is a nuclear method, NAA does not depend on the chemical form of the sample. NAA is a nondestructive technique and requires usually very small amounts of sample. Many elements can be determined at the same time (Alfassi 1990).

There are a few important experimental conditions for NAA, first of all the kinetic energy of the neutrons used for irradiation. Generally, neutron activation is performed with thermal neutrons, but there are also nuclear reactions used, where higher neutron energies are required. The intensity of the neutron field and the cross sections of the selected activation processes are important parameters. The nuclear decay products can be measured during or after neutron irradiation.

Neutron Scattering

There is a lot of research utilizing neutron scattering in biology, biotechnology, medicine, nanotechnology, and in research on catalysis, drugs, energy, magnetism, molecular structure, polymers, and superconductors all over the world. Some of the leading institutes are the Institut Laue–Langevin in Grenoble, the Rutherford–Appleton Laboratories in Oxford with ISIS, the FRM II reactor in Munich, the Research Center Jülich, the Paul Scherrer Institute in Würenlingen with SINQ, the KENS Neutron Scattering Facility at KEK in Japan, the Oak Ridge National Laboratory with their new spallation neutron source, and the Los Alamos Neutron Science Center LANSCE. These research centers are using special techniques like elastic and inelastic scattering, diffractometry, time-of-flight measurement, small-angle neutron scattering, reflectometry, or measurements of polarized neutrons. The international scientific community benefits from applying these sophisticated large installed detector systems, which employ physical principles and detection techniques that were discussed here. The details of these spectrometers and techniques are very elaborate and are not subject of this article. More information about these detector systems and research disciplines can be obtained from the Web sites of the neutron scattering laboratories.

Nuclear Medicine

In radiotherapy with neutron beams, the estimation of the neutron doses to the organs surrounding the target volume is particularly challenging. For instance, at the Louvain-la-Neuve (LLN) facility, these doses were investigated. The transport of a 10 cm × 10 cm beam through a water phantom was simulated with the Monte-Carlo code MCNPX and measurements of the absorbed dose and of dose equivalent using an ionization chamber and superheated-drop detectors were performed (Benck et al. 2002).

Boron neutron capture therapy (BNCT) is a cancer treatment method where after the delivery of a suitable boron compound to tumor cells, the tumor is irradiated with slow neutrons. The boron concentration in the tumor has to exceed the boron concentration in normal tissue considerably, which can be achieved by a number of compounds. A reactor or an accelerator has to deliver large thermal-neutron fluences of the order of magnitude of 1012n/cm2to get sufficient results of these irradiations (Gahbauer et al. 1997). Accurate measurements of neutron fluences and dose distributions as well as Monte-Carlo calculations are the base of treatment planning. Neutron fluence and absorbed-dose measurements were for instance performed with activation foils and paired ionization chambers (Binns et al. 2005).

Search for Illicit Trafficking Nuclear Materials

Since the demise of the former Soviet Union and even more since the new terrorism, the search for illicit trafficking or hidden nuclear material became an important new application in radiation detection. In the beginning, the main focus was in gamma detection, but soon neutron detection was also included, because plutonium, a material used for nuclear weapons, is a significant source of fission neutrons (Kouzes 2005).

Neutron detection is very selective for the indication of dangerous nuclear materials. Plutonium is extremely hazardous and hard to be detected, because it is not very difficult to shield the alpha, beta, and photon emissions. But the even-numbered plutonium isotopes exhibit significant spontaneous fission yields. For instance, 1 g of 238Pu is emitting 2,660 fission neutrons per second. Therefore, a neutron detector for the search for illicit trafficking or hidden nuclear material needs a maximum of sensitivity in the fission-neutron energy region. Rem counters are for these applications not sensitive enough and perform poorly, because their “dose tuning” is based on a tremendous amount of neutron filtering and absorption. A well-designed detector’s energy-dependent response should be optimized to fission neutrons at a maximum of sensitivity. This can be achieved by a moderator of reasonable size in which a large thermal-neutron detector is located. An example of a highly sensitive handheld detector is described in Klett (1999).

Neutron detection is now widely used in security applications like access and exit control of nuclear facilities, vehicle monitoring, border control, monitoring in harbors and airports, in waste storage management, and in safeguard activities. The Austrian Research Center Seibersdorf in cooperation with a team of International Atomic Energy Agency IAEA experts and supported by the World Custom Organization (WCO) and by INTERPOL has performed the Illicit Trafficking Radiation Detection Assessment Program (ITRAP). The aim of the study was to work out the technical requirements and the practicability of useful monitoring systems. International suppliers and manufacturers of radiation detection equipment from nine different countries have participated. The study covered fixed-installed monitoring instruments, pocket-type instruments, and handheld instruments. Neutron monitoring should be included in the fixed-installed systems; it was desirable for the handhelds and not necessary for the pocket-type instruments (Beck 2000).

For the detection of special nuclear materials (SNM), there are neutron coincidence counters in use. They have arrays of neutron detectors – usually large 3He proportional counters in moderators optimized for fission-neutron detection – covering a container from several sides. There is active interrogation and passive measurement. For active interrogation, a neutron source is used to induce fission in a fissile material under investigation. Passive measurement measures neutrons emitted by the sample without external irradiation.

Homeland security applications consumed large amounts of 3He since a decade. Since about 2008, there is now a worldwide shortage in 3He supply and many groups are now developing neutron detection alternatives (Kouzes et al. 2010)

Reactor Instrumentation

Reactor instrumentation requires mostly slow-neutron detection at high intensities and under extreme conditions of reactor operation. Neutron intensities have to be measured in core up to 1014 cm−2 s−1 and out of core up to 1010 cm−2 s−1. There are high pressures and temperatures which can be as high as 300 °C. Because of a lower gamma sensitivity, gas-filled detectors are a preferable choice.

Boron ionization chambers can be tailored to measure the required range of neutron flux. Uncompensated boron ion chambers are generally used in regions of high neutron flux where the gamma flux is only a small share of the total radiation level. Fission chambers can be used in pulse or in direct-current mode. Fission chambers include a fissile material normally 235U. The fission fragment’s large energy deposits are generating the detector signals. Fission chambers in pulse mode are ideal in mixed fields because gamma discrimination is easy in pulse mode. So-called self-powered detectors utilize a material with a high cross section for neutron capture with subsequent beta decay. The beta decay current is measured without external bias voltage. Overviews and more details about reactor instrumentation can be found in Knoll (2010) and Boland (1970).

Fusion Monitoring

Neutron spectrometry is a tool for obtaining fusion plasma information such as ion temperature and fusion power. Neutron spectrometry measurements for diagnostics at the Joint European Torus (JET) between 1983 and 1999 were reported by Jarvis (2002). A wide variety of spectrometer types with nuclear emulsions, NE213 liquid scintillators, hydrogen ionization chambers, recoil proton counters, 3He ionization chambers, silicon detectors, and diamond detectors have been tested with varying degrees of success. Magnetic proton recoil spectrometers are successful in monitoring the d–d and the d–t reactions at 2.5 MeV and 14 MeV, respectively. Investigations about the time resolutions of several different neutron spectrometry techniques and an upgraded magnetic recoil spectrometer for ITER were recently described by Andersson (2010).

Industrial Applications

Neutron Imaging and Radiography

As neutron interactions with atoms and molecules are very different from X-ray interactions, the neutron is sensitive to other aspects of matter. For instance, in investigating in automotive industries with imaging techniques, X-rays would illuminate the metallic structure of an engine while neutrons would rather take a picture from the oil. Neutron imaging requires position-sensitive neutron detectors.

Neutron transmission radiography (NR) is based on the attenuation of radiation passing through a sample. Details of samples can be made visible, if the attenuation is different in different materials. As neutron detectors track etch foils, a combination of a neutron converter layer (Gd, Dy) and X-ray film, or a combination of a neutron-sensitive scintillator and a CCD camera or position-sensitive 3He detectors have been used. A new development is amorphous-silicon flat panels. They contain Gd as neutron absorber and BaFBr:Eu2+ as the agent which provides the photoluminescence. An imaging plate scanner is extracting the digitized image information from the plates by de-excitation caused by a laser signal.

Humidity Measurement

Because water is an excellent neutron moderator, it is possible to measure humidity with neutrons. If fast neutrons emitted by a neutron source are penetrating humid matter, there is thermalization depending on the amount of water. A typical neutron humidity measurement setup comprises a fast-neutron source and a detector for thermal neutrons close to each other. The probe is positioned in or close to the sample material, which could be coal, coke, sand, sinter, soil, or lime sand bricks. The measurement is online and continuously without direct contact with the sample. The measurement is not affected by temperature, pressure, pH value, or optical characteristics of the material and determines of the amount of water molecules, irrespective of their physical or chemical binding. Humidity measurement with neutrons is mainly used by chemical, cement, ceramics, coal, iron, and steel industries.

Reference Neutron Radiation Fields

Reference neutron radiation fields are very important for the calibration of neutron detectors. The following types are commonly used:
  • Neutrons from radionuclide sources, including sources in a moderator

  • Neutrons generated by nuclear reactions with charged particles from accelerators

  • Neutrons from reactors

These fields have usually unidirectional beams and cover neutron energies in the range from thermal up to several hundred MeV. There are quasi-monoenergetic neutron radiations for determining the response of neutron-measuring devices as a function of energy, and there are neutron fields with wide spectra for calibration of instruments.

An instrument under calibration is placed in a free-in-air radiation field of known fluence rate and the reading is recorded. Neutron scattering from the air, by the walls, floor, and ceiling should be minimized and corrected for. The room used for irradiation should be as large as possible and measurements with a shadow cone help to take into account the scattered neutrons’ contributions. The international standard ISO 8529 about reference neutron radiations covers in its three parts the general principles, the commonly used radiation fields which are listed in Table 7, and the calibration procedures (ISO 8529-1 2001; ISO 8529-2 2000; ISO 8529-3 1998).
Table 7

Commonly used ISO reference neutron radiations (ISO 8529-1 2001; ISO 8529-2 2000; ISO 8529-3 1998) with fluence-averaged energies E and fluence-to-dose conversion factors \( {h}_{\phi}^{\ast }(E) \) according to ICRP74 (ICRP 1996)


Half-life [years]


Energy [MeV]

h∗ϕ(E) [pSv cm 2]

252 Cf (D2O moderated)


Spontaneous fission





Spontaneous fission





(α, n)



Reactor or accelerator


9Be(d, n)X/thermal column

2. 5 × 10−8


Sc-filtered reactor beam


2 × 10−3




45Sc(p, n)45Ti

24 × 10−3




T(p, n) 3He and 7Li(p, n)7Be





T(p, n) 3He and 7Li(p, n)7Be





T(p, n) 3He and 7Li(p, n)7Be





T(p, n) 3He





T(p, n) 3He





D(d, n) 3He





T(d, n) 4He





T(d, n) 4He





Université Catholique de Louvain (UCL)

33 and 50




TSL Uppsala





CERN/CERF (Mitaroff and Silari 2002)

<1 × 103


The facilities providing neutron radiations traceable to national standards are, for instance, the Physikalisch-Technische Bundesanstalt PTB in Germany, the National Physical Laboratory NPL in the United Kingdom, and the National Institute of Standards and Technology NIST in the United States.


As neutrons are neutral and not directly ionizing particles; neutron detection is more difficult than photon or charged-particle detection. Neutrons can only be detected after conversion into charged particles. There are several efficient nuclear processes to convert neutrons into charged particles, among them especially elastic n–p scattering and nuclear reactions on 3He, 6Li, and 10B target nuclei. These processes are utilized in a variety of neutron detection techniques which are used in research, nuclear medicine, industry, and in many other fields.



Thanks to Dr. Burkhard Wiegel/PTB Braunschweig for the photo and print permission of the NEMUS spectrometer.


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Further Reading

  1. Anderson IS, McGreevy RL, Bilheux HZ (eds) (2009) Neutron imaging and applications a reference for the imaging community. Springer, New YorkGoogle Scholar
  2. Blomgren J, Lindborg L (eds) (2007) Tenth International Symposium on Neutron Dosimetry: progress in dosimetry of neutrons and light nuclei light nuclei, Uppsala, 1216 June 2006, Radiation Protection Dosimetry v 126, nos 14. Oxford University Press, OxfordGoogle Scholar
  3. Bottolier-Depois J-F, Beck P, Reitz G, Rhm W, Wissmann F (eds) (2009) Cosmic radiation and aircrew exposure, Radiation Protection Dosimetry v 136, no 4. Oxford University Press, OxfordGoogle Scholar
  4. Grupen C, Shwartz B (2008) Particle detectors, 2nd edn. Cambridge University Press, CambridgeGoogle Scholar
  5. Kiefer H, Maushart R (1972) Radiation protection measurement. Pergamon, OxfordGoogle Scholar
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  7. Kleinknecht K (1998) Detectors for particle radiation, 2nd edn. Cambridge University Press, CambridgeGoogle Scholar
  8. Leo RW (1994) Techniques for nuclear and particle physics experiments, 2nd revised edn. Springer, BerlinGoogle Scholar
  9. Menzel HG, Chartier JL, Jahr R, Rannou A (eds) (1997) Neutron dosimetry: proceedings of the Eighth Symposium, Paris, 1317 November 1995, Radiation Protection Dosimetry v 70, nos 1–4. Nuclear Technology, AshfordGoogle Scholar
  10. Rzsa S (1987) Radiometrische Messungen in der Industrie Grundlagen und Memethoden. Franzis-Verlag GmbH, MnchenGoogle Scholar
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  12. Thomas DJ, Klein H (eds) (2003b) Neutron and photon spectrometry and techniques for radiation protection, Radiation Protection Dosimetry v 107, nos 1–3. Nuclear Technology, KentGoogle Scholar

Suppliers of Neutron Detectors

  1. ALOKA Co. Ltd., 6-22-1 Mure, Mitaka-shi, Tokyo, 181-8622, JapanGoogle Scholar
  2. Berthold Technologies GmbH & Co KG, Calmbacherstrasse 22, Bad Wildbad, 75323, GermanyGoogle Scholar
  3. BTI Bubble Technology Inc., Chalk River, CanadaGoogle Scholar
  4. Canberra Canberra Industries Inc., 800 Research Parkway, Meriden, CT, 06450, USAGoogle Scholar
  5. Centronic Limited, Centronic House, King Henrys Drive Croydon CR9 0BG, United KingdomGoogle Scholar
  6. Framework Scientific, LLCGoogle Scholar
  7. General Electric Company GE-Reuter-Stokes, 3135 Easton Turnpike Fairfield, CT, 06828-0001, USA8Google Scholar
  8. Innovative American Technology (IAT), 4800 Lyons Technology Park Drive, Coconut Creek, FL, 33073, USAGoogle Scholar
  9. John Caunt Scientific Ltd., PO Box 1052, Oxford, OX2 6YE, United KingdomGoogle Scholar
  10. Laboratory Impex Systems Ltd, 15 Riverside Park, Wimborne, Dorset BH21 1 QU, United KingdomGoogle Scholar
  11. Landauer Inc., 2 Science Road, Glenwood, Illinois, 60425-1586, USAGoogle Scholar
  12. LND, INC, Nuclear Radiation Detectors, 3230 Lawson Boulevard, Oceanside, NY, 11572, USAGoogle Scholar
  13. Ludlum Measurements Inc., 501 Oak Street, Sweetwater, TX, 79556, USAGoogle Scholar
  14. MIRION, Lieu-dit Calés Route dEguis, F-13113 Lamanon, FranceGoogle Scholar
  15. NucSafe Inc., 601 Oak Ridge Turnpike, Oak Ridge, TN 37830, USA
  16. Polimaster International., 112, M. Bogdanovich St., Minsk, 220040, Republic of BelarusGoogle Scholar
  17. ROTEM, Rotem Industrial Park, Mishor Yamin, D.N Arava 86800, IsraelGoogle Scholar
  18. SAIC Inc., 1710 SAIC Drive, McLean, VA 22102, USAGoogle Scholar
  19. Saint-Gobain Crystals, 7900 Great Lakes Pkwy, Hiram, OH, 44234-9681, USAGoogle Scholar
  20. Scionix Holland B.V., Regulierenring 3, Bunnik, 3981 LA, The Netherlands,
  21. TA Technical Associates, 7051 Eton Avenue Canoga Park, CA, 912303, USAGoogle Scholar
  22. Thermo Fisher Scientific Inc., 81 Wyman Street, Waltham, MA, 02454, USA
  23. Toshiba Electron Tubes & Devices Co., Ltd. 1385 Shimoishigami, Otawara, Tochigi, JapanGoogle Scholar
  24. TSA Systems Ltd., 14000 Mead Street, Longmont, CO, 80504-9698 USAGoogle Scholar

Copyright information

© Springer Nature Switzerland AG 2020

Authors and Affiliations

  1. 1.Berthold Technologies GmbH & Co KGBad WildbadGermany

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