Neutron detection efficiency

Figure 2. The neutron spectrum at die NIST BT2 neutron imaging facility. The spectrum is not corrected for die neutron detection efficiency, which increases as die square root of die energy.

The water thickness measurement uncertainty due to neutron counting statistics can be calculated from Poisson counting statistics. For a random process, the standard deviation, Ah in the observed counts I is A/ = y7. The number of neutrons in the incident, or open beam, I0, is the product of the neutron fluence rate (cur2 s 1), integration time T (s), integration area A (cm2), and neutron detection efficiency, ip... 

The drawer neutron counter (DRNC) has been designed to perform the assay of plutonium in fast critical assembly fuel drawers (Krick and Menlove 1980). Eight tubes (2.5 cm diameter by 91 cm active length) were used in the system. The principal feature of the neutron coincidence detector is a 7 cm by 7 cm by 97 cm detector channel, which provides a uniform neutron detection efficiency of 16% along the central 40 cm of the channel. 

Fast neutron detection sometimes uses a hydrogenous moderator to slow down the neutrons and then employs a low-energy neutron detector as described above. One common fast neutron detector is a Bonner sphere. In this detector, a scintillator is placed in the center of a polyethylene sphere. Radiation transport calculations are used to produce efficiency curves that depend on the energy of the incident neutron. Another common fast neutron detector is a long counter. This detector uses a slow neutron detector (originally a BF3 chamber) at the center of a cylindrical moderator designed so that the detector is only sensitive to neutrons incident from one side. 

Indirect geometry spectrometers have no requirement (within the limitations implied by the use of S (Q,a>), 2.5.1) to calibrate detector efficiencies, on either continuous or pulsed sources (compare 3.4.3). Since the final energy of the neutrons never varies the detection efficiency is constant. Variations arising from differing discrimination levels ( 3.3.2) could play a significant role, except that (on low final energy instruments) all detectors follow almost the same path in Q,o ) space ( 3.4.2.3). Occasionally there is a need to calibrate the detected intensity in respect of the sample mass and standard analytical chemical techniques can be readily adapted to this circumstance. 

Efficiencies of neutron detectors are calculated by methods similar to those used for gammas. Neutrons are detected indirectly through gammas or charged particles produced by reactions of nuclei with neutrons. Thus, the neutron detector efficiency is essentially the product of the probability of a neutron interaction, with the probability to detect the products of that interaction (see Chap. 14). 

Effect of carbon recoils. Neutrons detected by methane-filled counters collide not only with hydrogen nuclei but also with carbon atoms. The ionization produced by carbon recoils is indistinguishable from that produced by protons. However, carbon recoils produce pulses that are smaller than those from protons because of differences in both kinematics and ionization ability. The maximum fraction of neutron energy that can be imparted to a carbon nucleus in one collision is 0.28 (versus 1 for a hydrogen nucleus), and the relative ionization efficiency of a carbon to a proton recoil is about 0.5. Thus, the effect of carbon recoils is to add pulses at the low-energy region of the response... 

In practice, the instruments are properly calibrated to read directly Sv (or rem), or Gy (or rad). For some neutron detection instruments, the neutron flux is recorded. Then the dose equivalent is obtained after multiplying the flux by the conversion factor given in Table 16.4. Since different detectors do not have the same efficiency or sensitivity for all types of radiation and for all energies, there is no single instrument that can be used for all particles (a, y, n) and all energies. 

The electric conductivity of various samples of germanium and silicon doped with °B during the neutron irradiation was measured by Kervalishvili et al. (1993). In particular, measurements of neutron fields of various nuclear plants with sensors made of these materials have been carried out. The selection of the base material and the concentration of °B impurity allowed the efficiency of neutron detection and flux measurement to be varied. Such neutron dosimeters possess memory and can be used as witnesses of unusual situations in nuclear plants. 

It was clearly demonstrated that the composite BN semiconductor polycrystalline bulk detectors with BN grains embedded in a polymer matrix operate as an effective detector of thermal neutrons even if they contain natural boron only (Uher et al. 2007). A reasonable signal-to-noise ratio was achieved with detector thickness of about 1 mm. A Monte Carlo simulation of neutron thermal reactions in the BN detector was done to estimate the detection efficiency and compare with widely used He-based detectors to prove advantages of BN detectors. They are found to be promising for neutron imaging and for large area sensors. 

PPO and POPOP are among the most commonly used organic primary and secondary fluors, respectively. Despite their high efficiencies in both liquid and plastic scintillators, one limitation is their incompatibility with hydrophilic reagents. In liquid scintillation techniques, an efficient extraction of radioactive nuclides into the organic phase where PPO and POPOP dissolve is often mandatory. In neutron detection by plastic scintillators, the use of an efficient, yet hydrophilic, neutron absorber, Li, is strictly limited if not prohibited. 

In the relative method, the mass fraction of the analyte element in the unknown sample is obtained using the mass fraction of the same element in the standard. The equation takes into account the peak areas obtained in the unknown and in the standard the sample weight (mass), W the decay factors S, D, and C (see Eq. (30.26)) differences in neutron flux and differences in detection efficiency due to slightly different counting geometries. 

Users of the kg method must determine the thermal neutron flux, the ratio of thermal flux to epithermal flux for the irradiation channel used (/), the shape parameter (a), and detection efficiencies for all counting geometries used. Since the calculations for the derivation and the use of these parameters are extensive, a software package is needed. Some new users begin a collaboration with one of the many laboratories using the Icq method and adopt their methods of calculation and their software. Two complete software packages are now readily available to all potential users. The KAYZERO/SOLCOI Software (Van Sluijs et al. 1997), now called KayWin, performs the calculations in the well-understood classical manner. It is widely used... 

Many methods have been developed to measure /and a. They all involve the activation and counting of a number of nuclides having a range of Qo values and mean resonance energies. The most accurate measurements of a use irradiations under cadmium cover to activate only with epithermal neutrons. However, many laboratories may not require such high accuracy or may not be permitted to irradiate under cadmium cover. Bare irradiations have been shown to give sufficient accuracy if carefully done. Since the parameters are determined by subtraction of the thermal neutron-produced activity from the total activity- two possibly similar quantities -accurate element masses, peak areas, and detection efficiencies are needed. The minimum number of monitors that need to be irradiated for the simultaneous determination of the thermal neutron flux, the factor /, and a. is three. 

BF3 detectors tree occasionally used based on the B(n, a) Li reaction. BF3 detectors are less sensitive to gamma radiation fields but are less efficient. Recently, solid-state neutron radiation devices with boron carbide diodes have been developed, which demonstrate very promising potential for future applications such as miniaturized handheld neutron detection devices. 

The portable neutron coincidence counter (PNCC) consists of four individual slab detectors with four He tubes each that can operate in multiple modes and configurations (Thornton et al. 2006). The detector is lightweight and portable ( 15 kg) to address flexibility of measurement requirements for various field environments. PNCC has about 12% detection efficiency. 

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