Neutron Attenuation Characteristics Research Articles

Flexible, chemically bonded bismuth/tungsten‐based polyvinyl alcohol‐polyvinyl pyrrolidone composite for gamma and neutron shielding application

AbstractIn this paper, the gamma and neutron shielding behavior of chemically bonded high atomic number (Z), tungsten trioxide‐polyvinyl alcohol‐polyvinylpyrrolidone (PW), and bismuth oxide‐polyvinyl alcohol‐polyvinylpyrrolidone (PB) composites were reported at varying thickness. The PW and PB composites were prepared through the dissolution technique rather than the commonly used dispersion technique. The gamma and neutron attenuation characteristics of the composites were studied using 22Na (0.511 and 1.274 MeV energy) and 252Cf (0–10 MeV energy) sources, respectively. The half value layer of PW and PB composite was found to be 61.22 ± 0.914 mm and 48.29 ± 1.52 mm and the linear attenuation coefficient was found to be 0.115 and 0.144 cm−1 at 0.511 MeV, respectively. Further, the experimental gamma‐ray attenuation results were validated theoretically using Phy‐X software and found to be well in agreement with the experimental data. Both the composites also showed similar neutron shielding behavior when compared to the commercially used high‐density polyethylene (HDPE). The developed composites possess good mechanical properties as they did not show any cracks or wrinkles after undergoing 5400 bending cycles during the flex test, showing their high durability of flexibility. Thus, these composites are found to be suitable for shielding both, the gamma‐rays and neutrons.

Comprehensive investigation on physical, structural, mechanical and nuclear shielding features against X/gamma-rays, neutron, proton and alpha particles of various binary alloys

In this study, four alloys composed of highly pure elements (lead, zinc, tin and bismuth), named Pb–Zn50, Pb–Sn50, Sn–Zn50 and Sn–Bi50 were prepared by rapid solidification process (RSP) using melt-spun technique. The structural, mechanical properties (elastic and plastic deformations), and microhardness have been investigated and analyzed using X-ray diffraction, tensile test machine, and vickers microhardness respectively. According to results, the yield strength, ultimate tensile strength (UTS), toughness, and Vickers hardness (Hv) were improved due to the advantages of rapid solidification technology including elimination of grain boundary segregation, improvement in alloy homogeneity, refinement of grain size, formation of novel metastable crystalline, rises in the equilibrium solid solubility's of solute elements and reduction in the degree of order. This can be ascribed to reinforcement of Zn and Bi particles, refined β-Sn, Zn and Bi grains that could obstruct the dislocation slipping. In addition, the shielding performance of these alloys for gamma radiation were examined and analyzed utilizing WinXCom software over energy range 0.015–15 MeV. The alloys under study found to be good shielding materials against γ-ray relative to common shielding materials and newly studied materials. The results indicate that the Sn–Bi50 alloy has the highest values of μm, Zeff and Zeq ranging from 0.039 to 81.3 cm2/g, 60.41–73.83, and 64.02–70.793 respectively. While it has the smallest values of HVL (ranging from 8.73E-4 to 1.86 cm) which means that it is the best choice as a γ-ray shielding material. In aspects of neutron attenuation characteristics, the results also indicate that the Pb–Zn50 alloy has the highest value of ΣR (0.133 cm−1) amongst all other alloys, frequently used neutron shielding materials, and recently studied materials. Furthermore, proton (H+1) and alpha particles (He+2) projected range (PR) and mass stopping power (MSP) values have been calculated. It can be concluded that Zn–Pb alloy among the examined samples is superior in terms of shielding against proton and alpha particles. This paper provides a comprehensive examination of the use of lead and the development of new safety lead-free shielding materials in many applications including nuclear medicine.

Exploring the radiation shielding properties of B2O3-PbO-TeO2-CeO2-WO3 glasses: A comprehensive study on structural, mechanical, gamma, and neutron attenuation characteristics

The objective of this study is to develop radiation-shielding glass materials with excellent physical properties by incorporating varying amounts of WO3 (x = 1, 3, 5, 10, 20 mol%) into 45B2O3-(20-x)PbO–30TeO2–5CeO2-xWO3 glasses. The glasses were synthesized using the melt casting method with B2O3–PbO–TeO2–CeO2 powders. Experimental measurements and theoretical calculations were performed to determine the mass attenuation coefficients (MAC), mean free path (MFP), half value layer (HVL), and effective atomic numbers (Zeff) of the glasses within the energy range of 59.54–661.62 keV. Additionally, the structural and mechanical properties of the synthesized glasses were evaluated using XRD, UV–Vis, FT-IR, and Micro-Vickers hardness techniques. XRD analysis indicated that all glass compositions formed amorphous structure. The addition of WO3 caused a decrease in the optical band gap energy values of the glasses. Moreover, the incorporation of WO3 enhanced the elasticity of the glasses by 17–21%. The hardness and fracture toughness values were also improved with the addition of 5% WO3, increasing from 3.53 to 4.14 GPa and from 1.07 to 1.14 MPa m0.5, respectively. The radiation shielding properties assessed through proton and alpha shielding characteristics, revealed that the addition of 20% WO3 to the glass composition allowed approximations of 82.8% of the MAC values of the glass doped with 20% PbO. Overall, the findings indicate that incorporating WO3 as a dopant improves the glasses' elastic and mechanical properties enhancing their radiation shielding ability.

A novel bismuth-nickel-aluminium nanocomposite as a protective barrier against nuclear radiation

In the present work, Alovera extract is used as a reducing agent during the versatile solution combustion synthesis of bismuth nickel aluminum nanocomposites (BiNiAl NCs). Standard techniques are used to characterize the obtained BiNiAl NCs. The formation of multi-phase BiNiAl NCs with crystal diameters of 18.20 nm and 22.06 nm was confirmed by Bragg’s reflections. The agglomeration of spherical-shaped nanoparticles was confirmed by the surface morphology of BiNiAl NCs. The presence of Bi, Ni, and Al was analyzed using the EDAX spectrum. The direct energy gap is calculated to be 3.06 eV. Then, using a NaI(Tl) detector coupled with MCA, the shielding characteristics against X-rays and gamma radiation are measured. Shielding standards are a little higher than lead standards. The characteristics of neutron shielding are compared to those of traditional shielding materials. When compared to lead, concrete, and steel, the neutron attenuation characteristics and scattering length of BiNiAl NCs are determined to be effective. The scattering length is found to be 6.6 fm. At 1.511 MeV, the shielding parameters for bremsstrahlung radiation are calculated. While bremsstrahlung dose rate and specific bremsstrahlung constant are found to be greater in comparison with lead, bremsstrahlung efficiency is equivalent to lead. As a result, BiNiAl NCs are very effective in absorbing neutrons, bremsstrahlung radiation, and X/gamma rays.

Radiation shielding and mechanical properties of mullite-zirconia composites fabricated from investment-casting shell waste

Low-cost zirconia-toughened mullite composites were prepared from investment casting shell waste and alumina. The influence of sintering temperature on the composites' properties, including crystalline phase composition, microstructure, degrees of densification, mechanical properties, and radiation shielding characteristics, was investigated. The results show that the higher the sintering temperature, the higher the degree of densification is, improving the mechanical and radiation shielding properties of prepared composites. The mullite-zirconia composite sintered at 1600 °C presents a good mechanical strength, with flexural and compressive strength values of 190 and 308 MPa, respectively. These values are comparable to or even better than mullite ceramics prepared from other waste materials. Furthermore, the composites' gamma-ray and neutron attenuation characteristics suggest that they can be promising as radiation shields.

Ionizing radiation attenuation features of Cu2-II-Sn-VI4 (II=Mn, Fe; VI = S, Se, Te) quaternary semiconducting compounds

The research's essential aim is to use Phy-X/PSD software to analyze photon attenuation properties of Cu2-II-Sn-VI4 (II=Mn, Fe; VI =S, Se, Te) quaternary semiconducting compounds at photons energies between 15 keV and 15 MeV. The parameters of photon interaction, including MAC, LAC, HVL, TVL, MFP, ACS, ECS, Zf, Nf, and Cf were determined in the chosen range of energies. The software was performed at the selected photon energies to compute each sample's R, Zeq, and build-up factors. The HVL and MFP values of the samples were compared to those mentioned in the previous studies for different concretes and commercial glasses. The HVL and MFP values for Cu2FeSnSe4, Cu2FeSnTe4, Cu2MnSnSe4, and Cu2MnSnTe4, samples are lower than those of the compared concrete types commonly utilized for protection. According to data acquired from the Phy-X/PSD software, Cu2FeSnTe4 is the optimum composite for gamma-ray attenuation among the studied samples. Also, in comparison to the other examined samples, the Cu2FeSnTe4 sample showed the greatest FNRCS (fast neutron removal cross-section) value (0.117), indicating good neutron attenuation characteristics.

Synergistic effect of boron nitride and graphene nanosheets on behavioural attitudes of polyester matrix: Synthesis, experimental and Monte Carlo simulation studies

We report the synergistic effects of boron nitride and graphene nanosheets on physical, structural, and nuclear radiation attenuation properties of polyester matrix-incorporated nanocomposites. Some critical material properties are thoroughly evaluated for several types of synthesized samples. Polyester is employed to strengthen graphene and boron nitride nanolayers, and their characteristics are investigated in detail. Additionally, we report the gamma-ray and fast neutron attenuation characteristics of synthesized nanocomposites to get a better understanding of the reinforcing effect as a function of material type and weight percentage. Thermal analysis findings indicate that adding graphene lowers the decomposition temperature but co-adding graphene and BNNS enhances thermal decomposition in comparison to graphene itself. Tensile tests showed that the inclusion of both GRP and GRP/BNNS strengthens the material. Among the polyester composite samples analyzed, the G3 sample with the most GNP reinforcement had the lowest HVL values throughout the broadest range of energy levels investigated. The recent findings may be beneficial to the scientific community in terms of incorporating these reinforcing types and ratios into polyester materials for a variety of applications, including industrial and research purposes.

Gamma and neutron attenuation characteristics of bricks containing zinc extraction residue as a novel shielding material

In recent years, people are exposed to radiation depending on the technology developed which may cause serious incurable health problems. To protect from radiation exposure, various radiation shielding materials are used in different areas. In this study, zinc extraction residue (ZER), which is released during zinc production containing various heavy metals, was evaluated in clay-based brick manufacturing. The fired bricks were characterized in terms of radiation shielding, bulk density and compressive strength. Clay brick with 50% ZER was fired at 1000 °C. Bulk density and compressive strength of fired brick were found as 2.16 g/cm3 and 12.6 MPa, respectively. The addition of ZER increased the density while reducing the compressive strength. Radiation shielding potential for the samples were evaluated both experimentally (using an Am–Be neutron source and a Cs-137 gamma source) and simulation-wise using the Monte Carlo technique. For this purpose, mass attenuation coefficients of gammas and total absorption cross-section of neutrons were determined through measurements. Additionally, Monte Carlo simulations were carried out under similar irradiation conditions. The MCNP simulation results were checked against those produced from XCOM database for photons and Phy-X/PSD for neutrons. Inclusion of ZER was found to elevate shielding capabilities especially at low photon energies because of the lead content of ZER. In addition, neutron attenuation characteristics of ZER bricks were observed to be relatively lower than that of fired clay brick because of the heavier ingredients existing in ZER extracts.

Investigations on neutron attenuation properties of poly-boron materials using Am-Be and 252Cf sources neutron spectra

The neutron attenuation characteristics of polyboron materials was studied using Am-Be (α,n) and 252Cf (Spontaneous Fission) sources. Neutron Attenuation Factor (NAF) is obtained for poly boron materials with addition of different concentrations of boron by neutron transmission experiments. Neutron transport calculations are carried by simulating the transmission of neutron from Am-Be and 252Cf sources through polyboron materials using Monte Carlo based simulations. The experimental results are fairly in good agreement with transport calculations. The results showed that Neutron Attenuation Factor (NAF) mainly depend on the incident neutron energy and boron number density present in HDPE shielding material. The effect of boron in HDPE on neutron shielding is only marginal (2%) because of low value of neutron cross section at fast energy region. However, in the low neutron energy region, it varies with wt % of boron and 20% boron as additive in HPDE is an optimum value for neutron attenuation. The details of experimental observations and theoretical investigations are discussed in this work.

Gamma ray and neutron shielding characteristics of polypropylene fiber-reinforced heavyweight concrete exposed to high temperatures

The residual properties of shielding structures need to be redefined after exposure to elevated temperatures due to their probable radiological hazards on biodiversity. In this study, the effect of high temperatures on mechanical, gamma ray and neutron attenuation characteristics was determined for limestone, barite and siderite concrete shields reinforced with polypropylene fiber. The increase in temperature up to 600 °C reduced the ultrasonic pulse velocity and compressive strength values of these shields by 59% and 62%, respectively. A good linear correlation (R2 ≥ 0.97) was found between these values. While the gamma ray linear attenuation coefficients increased by 9% for barite concrete, and reduced by 15% for normal concrete and by 17% for siderite concrete at 600 °C, trivial fluctuations (from −1% to +7%) were observed at 300 °C. The neutron attenuation factors of these shields gradually reduced up to 31% with the increase of temperature. The performance of siderite concrete was found to be slightly better than that of barite concrete in terms of strength and neutron shielding characteristics. With the satisfactory linear attenuation coefficients at all energy levels of gamma rays, barite concrete seems to be the most suitable concrete type under high temperature risks.

TeO2–B2O3–ZnO–La2O3 glasses: γ-ray and neutron attenuation characteristics analysis by WinXCOM program, MCNP5, Geant4, and Penelope simulation codes

γ-ray, fast and thermal neutrons shielding features for five Pb-free {[(TeO2)0.70(B2O3)0.30]0.7 (ZnO)0.30}(1-x)(La2O3)x (x = 0.01, 0.02, 0.03, 0.04, and 0.05 mol%) glasses have been inspected in this work. For all samples, within the photon energy range of 0.015–15 MeV, linear attenuation coefficients (μ) and mass attenuation coefficients (μ/ρ) were estimated using WinXCOM program (theoretical approach), and obtained μ/ρ values are in fine agreement with computed μ/ρ results of respective MCNP5, Geant4, and Penelope codes. By utilizing μ/ρ values of WinXCOM, γ-ray attenuation parameters like effective atomic number (Zeff), effective electron density (Neff), half-value layer (HVL), and mean free path (MFP) were assessed within 0.015–15 MeV photon energy range, for all samples. Additionally, exposure buildup factors (EBFs) and energy absorption buildup factors (EABFs) were calculated for all samples by applying a five parameter geometric progression (G‒P) fitting method as a function of various penetration depths (1, 5, 10, 15, 20, 25, 30, 35, and 40 mfp) within 0.015–15 MeV photon energy range. The derived radiation protection efficiency (RPE) results indicate all samples' effective attenuation competence for lower energy γ-rays. Among all glasses, 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) sample, by having relatively bigger μ/ρ and Zeff, minimal HVL and MFP, and lowest EBF and EABF, endorse its' excellent γ-ray shielding effectiveness. Further, neutron shielding parameters such as macroscopic removal cross-section for fast neutrons (ΣR), coherent and incoherent scattering cross-sections (σcs & σics), absorption cross-section (σA), and total cross-section (σT) for thermal neutrons were explored. By using the Geant4 code, ‘σT’ values have been deduced within 10−4‒10−8 MeV neutron energy range. The 46.55TeO2-19.95B2O3-28.5ZnO–5La2O3 (mol%) glass owns relatively larger ΣR (0.1474 cm−1) for fast neutrons attenuation and ‘σT’ (10.1444 cm−1 at 0.0253 eV neutron energy, 15.9824 cm−1 → 0.551 cm−1 from 1 × 10−8 MeV →1 × 10−4 MeV neutron energy) for thermal neutrons absorption, respectively, revealing its' superior neutrons shielding ability.

Er2O3 effects on photon and neutron shielding properties of TeO2-Li2O-ZnO-Nb2O5 glass system

Abstract In this study, a series of 75TeO2-5Li2O-10ZnO-(10-x)Nb2O5-(x)Er2O3 (where x = 0, 0.5, 1.0, 1.5, 2.0 and 2.8 mol%) glasses have been surveyed in terms of photon and neutron shielding characteristics. For this aim, 3 × 3 inch NaI(Tl) detector has been simulated in order to detect photons. Afterwards, the mass attenuation coefficients (μ/ρ) and some shielding quantities such as half-value layer (HVL), tenth value layer (TVL), mean free path (MFP), effective atomic number (Zeff), effective electron density (Nel), equivalent atomic number (Zeq) and exposure build-up factor (EBF) have been calculated. The obtained MCNPX results of all the glasses have been compared and approved with those of XCOM program. Also, their gamma ray buildup factors have been determined in a wide energy range of 0.02–20 MeV for penetration depths up to 15 mfp. Moreover, neutron attenuation abilities of glasses have been evaluated by estimating neutron total effective removal cross section. The results showed that Er2O3 partial replacement of Nb2O5 in 75TeO2-5Li2O-10ZnO-(10- x)Nb2O5-(x)Er2O3 glass system enhances photon and neutron attenuation characteristics. It can be deduced that obtained results from the present investigation can be useful to understanding of influence of Er2O3 on nuclear radiation shielding properties of tellurite glasses.

Evaluation of gamma-ray and neutron attenuation properties of some polymers

Abstract In the present work, we determined the gamma-ray attenuation characteristics of eight different polymers(Polyamide (Nylon 6) (PA-6), polyacrylonitrile (PAN), polyvinylidenechloride (PVDC), polyaniline (PANI), polyethyleneterephthalate (PET), polyphenylenesulfide (PPS), polypyrrole (PPy) and polytetrafluoroethylene (PTFE)) using transmission geometry utilizing the high resolution HPGe detector and different radioactive sources in the energy range 81–1333 keV. The experimental linear attenuation coefficient values are compared with theoretical data (WinXCOM data). The linear attenuation coefficient of all polymers reduced quickly with the increase in energy, at the beginning, while decrease more slowly in the region from 267 keV to 835 keV. The effective atomic number of PVDC and PTFE are comparatively higher than the Zeff of the remaining polymers, while PA-6 possesses the lowest effective atomic number. The half value layer results showed that PTFE (C2F4, highest density) is more effective to attenuate the gamma photons. Also, the theoretical results of macroscopic effective removal cross section for fast neutrons (∑R) were computed to investigate the neutron attenuation characteristics. It is found that the ∑R values of the eight investigated polymers are close and ranged from 0.07058 cm−1 for PVDC to 0.11510 cm−1 for PA-6.

Development of Gadolinium (neutron poison) monitoring system for fuel reprocessing facilities: Computational model and validation with experiments

Gadolinium, a strong neutron absorber is a good choice of material for use as soluble neutron poison for nuclear criticality control in nuclear fuel cycle facilities. During operation, it's availability has to be ensured in order to maintain the sub-critical state of the system. Towards this, periodic monitoring of presence of Gadolinium, a monitoring system was developed. The technique used is based on thermal neutron attenuation characteristics of the medium considered under study. As the design involves, choice of neutron source, moderator assembly and neutron detector sensitivity, Monte Carlo simulation was employed to optimize the design parameters. The details of simulation and experiments carried out to study the feasibility of the method is described. Results of measurements and comparison with computational model developed are also presented in detail. The results indicate that the proposed system can be successfully deployed for remote online monitoring of neutron poison in Fuel Reprocessing.

Effect of heat treatment on neutron attenuation characteristics of high density concretes (HDC)

In nuclear reactor and fuel cycle facilities, apart from ordinary concrete (2300 kg/m3) various special types of concrete of density varying in range 3600–4600 kg/m3have been effectively employed to achieve less shielding thickness. Radiation attenuation in concrete mainly depend on types of aggregate used; water to cement ratio, elemental composition and moisture content apart from density of concrete. In addition to this, operating temperature also plays a vital role in deciding its shielding properties due to loss of water content and microcracking. The effect of heat on neutron attenuation properties of HDC upon heat exposure is the scope of the study in this paper. Four types of concrete made of granite and having hematite and steel shots as aggregate were prepared with density varying from 2300 to 4300 kg/m3and exposed to 120 °C for durations (14, 28 and 56 days). Neutron Attenuation Factor (NAF) was obtained from experimental study and compared with results obtained with neutron transport calculations. The results of this study clearly indicate that the sustained and cyclic heat treatment at 120 °C up to 56 days reduced the neutron attenuation factor by a factor of two.

Visualization of the boiling phenomenon inside a heat pipe using neutron radiography

Heat pipes are effective heat exchangers that have a wide range of applications because of their ability to passively transfer large amounts of heat. Research into heat pipe technology has dramatically increased over the last decade and, more recently, has incorporated the use of visualization to help researchers gain a better understanding of the boiling phenomenon and heat transfer occurring inside a heat pipe. Neutron radiography is one method of visualization suitable for use in heat pipe investigations due to unique attenuation characteristics of neutrons attaching to various materials. In this study, an aluminum-based heat pipe was tested using working fluid filling ratios from a 10% to 90% capacity. Visualization using neutron radiography was conducted at a neutron radiography facility, RN1, under the supervision of the Centre of Science and Technology of Advanced Materials (PSTBM), National Nuclear Energy Agency of Indonesia (BATAN). Using temperature and pressure sensors, this study revealed that the optimum value of working fluid filling ratios directly correlates to the pressure inside a heat pipe and the size of vapor space available. The neutron radiography facility maintains high neutron flux at 106–107n/cm2s; high quality images were captured utilizing this radiography visualization technology. The captured images demonstrate that the boiling phenomenon inside a pressure-reduced heat pipe varies when compared with the boiling phenomenon at atmospheric pressure. The visualization result also shows the importance of wick structure in pumping return condensate from the condenser to the evaporator.

Comparison of neutron attenuation properties of ferro boron slabs containing 5% natural boron with other high density materials

Modeling and designing cost-effective neutron attenuation along with shield volume reduction is a challenging task in fast reactors. It involves reducing the neutron energy and absorbing them with suitable materials. A series of experiments were conducted in the South beam end of Kalpakkam Mini reactor with powders of ferro boron (FeB), ferrotungsten (FeW), boron carbide, slabs of FeB, and mild steel plates to study their neutron attenuation characteristics. In one of the experiments, FeB slab cast with 5% natural boron was used, and neutron attenuation measurements were carried out. The attenuation factors were found over a thickness of 28 cm for the measured reaction rates of 195 Pt (n, n') 195m Pt, 111 Cd (n, n') 111m Cd, 103 Rh (n, n') 103m Rh, 115 In (n, n') 115m In, 180 Hf (n, n') 180m Hf, 63 Cu (n,g) 64 Cu, 23 Na (n,g) 24 Na, 55 Mn (n,g) 56 Mn, and 197 Au (n,g) 198 Au reactions representative of fast, epithermal, and thermal neutron fluxes. A comparative analysis of the neutron attenuation behavior measured with various materials is presented. In case of attenuation of both thermal and fast fluxes, FeB is better than other high density materials such as mild steel and FeW. The outcome of the experimental study is that FeB slab cast with 5% natural boron can be utilized as cost-effective neutron shield in streaming paths in nuclear reactors.

Determination and calculation of gamma and neutron shielding characteristics of concretes containing different hematite proportions

To obtain gamma and neutron attenuation characteristics experimental measurements and Monte Carlo simulations were carried out for hematite-loaded concrete samples. Also, the mechanical strength of samples was studied. However, the results for neutron attenuation reveal that there is no effect of hematite inclusion in concrete with respect to the neutron absorption capability. Gamma-ray attenuation capability and the mechanical strength both increased with increasing the hematite percent.

Neutron radiography of porous rocks and iron ore

Neutron radiography can provide images of the internal structure of rocks, through virtue of the differing neutron attenuation characteristics of hydrogen versus other elements commonly found in nature. Thus, pores filled with water, or oil, will attenuate neutrons to a greater extent than the surrounding rock matrix. Similarly, the internal structure of iron ore, which contains mixtures of limonite and goethite, which contain hydrogen in their crystalline lattice, and magnetite (no hydrogen), can also be imaged using neutrons. The ability of the technique to distinguish “effective porosity” versus “total porosity” is derived from the neutron radiographs. Images of two iron ore samples have also been analyzed to demonstrate how both different mineral assemblages and porosity can be determined.

Visualization and measurement of two-phase flow by using neutron radiography

To apply neutron radiography (NR) technique to fluid research, high frame-rate NR with a steady thermal neutron beam has been developed in the present research program by assembling up-to-date technologies for neutron source, scintillator, high-speed video and image intensifier. This imaging system has many advantages such as a long recording time (up to 21 min), high-frame-rate (up to 1000 frames s −1) imaging and no need for triggering signal. Visualization of air-water two-phase flow in a metallic duct was performed at the recording speeds of 250, 500 and 1000 frames s −1. The qualities of those consecutive images were good enough to observe and measure the flow structure and the characteristics. It was demonstrated also that some characteristics of two-phase flow could be measured by using the present imaging system. Image processing technique enabled measurements of various flow characteristics in two ways. By utilizing geometrical information extracted from NR images, data on flow regime, rising velocity of bubbles, and wave height and interfacial area in annular flow were obtained. By utilizing attenuation characteristics of neutrons in materials, measurements of void profile and average void fraction were performed. It was confirmed that this new technique may have significant advantages both in visualizing and measuring high-speed fluid phenomena when the other methods such as an optical method and X-ray radiography cannot be applicable.