Secondary Gamma Rays Research Articles

A cargo inspection system based on pulsed fast neutron analysis (PFNA™)

A cargo inspection system based on pulsed fast neutron analysis (PFNA) is to be used at a border crossing to detect explosives and contraband hidden in trucks and cargo containers. Neutrons are produced by the interaction of deuterons in a deuterium target mounted on a moveable scan arm. The collimated pulsed fast neutron beam is used to determine the location and composition of objects in a cargo container. The neutrons produce secondary gamma rays that are characteristic of the object's elemental composition. The cargo inspection system building consists of an accelerator room and an inspection tunnel. The accelerator room is shielded and houses the injector, accelerator and the neutron production gas target. The inspection tunnel is partially shielded. The truck or container to be inspected will be moved through the inspection tunnel by a conveyor system. The facility and radiation source terms considered in the shielding design are described.

12C + 12C cross-section measurements at low energies

In order to study the 12C + 12C reaction at low energies, a combination of an intense heavy ion beam on a thick target and the detection of secondary gamma rays on a well-shielded germanium detector has been used. Preliminary cross-section results are reported for E c.m. = 2.63−3.45 MeV.

GeV and Higher Energy Photon Interactions in Gamma‐Ray Burst Fireballs and Surroundings

We have calculated the opacities and secondary production mechanisms of high energy photons arising in gamma-ray burst internal shocks, using exact cross-sections for the relevant processes. We find that for reasonable choices of parameters, photons in the range of 10's to 100's of GeV may be emitted in the prompt phase. Photons above this range are subject to electron-positron pair production with fireball photons and would be absent from the spectrum escaping the gamma-ray burst. We find that, in such cases, the fireball becomes optically thin again at ultra-high energies ($\gtrsim$ PeV). On the other hand, for sufficiently large fireball bulk Lorentz factors, the fireball is optically thin at all energies. Both for $\gamma\gamma$ self-absorbed and optically thin cases, the escaping high energy photons can interact with infra-red and microwave background photons to produce delayed secondary photons in the GeV-TeV range. These may be observable with GLAST, or at low redshifts with ground-based air Cherenkov telescopes. Detection of the primary prompt spectrum constrains the bulk Lorentz factor, while detection of delayed secondary gamma-rays would provide a consistency check for the primary spectrum and the bulk Lorentz factor as well as constraints on the intergalactic magnetic field strength.

Enhancement of sodium borocaptate (BSH) uptake by tumor cells induced by glutathione depletion and its radiobiological effect

Sodium borocaptate (BSH) is widely used for boron neutron capture therapy (BNCT) of brain tumors. However, the mechanism of uptake by the tumor remains unclear. We investigated the sulfhydryl moiety of this compound. Down regulation of glutathione (GSH) by buthionine sulfoximine in cultured cells resulted in increase of BSH uptake (7.9–36.5%) compared to the control group and consequently the cytocidal effect of neutron irradiation also increased. On the other hand, the radiation caused damage by gamma-ray irradiation was suppressed when BSH uptake increased. These findings suggested that modulation of GSH enhanced the effect of B ( n, α) reaction and the protective effect of secondary gamma-ray in BNCT.

Studies on depth-dose-distribution controls by deuteration and void formation in boron neutron capture therapy

Physical studies on (i) replacement of heavy water for body water (deuteration), and (ii) formation of a void in human body (void formation) were performed as control techniques for dose distribution in a human head under neutron capture therapy. Simulation calculations were performed for a human-head-size cylindrical phantom using a two-dimensional transport calculation code for mono-energetic incidences of higher-energy epi-thermal neutrons (1.2–10 keV), lower-energy epi-thermal neutrons (3.1–23 eV) and thermal neutrons (1 meV to 0.5 eV). The deuteration was confirmed to be effective both in thermal neutron incidence and in epi-thermal neutron incidence from the viewpoints of improvement of the thermal neutron flux distribution and elimination of the secondary gamma rays. For the void formation, a void was assumed to be 4 cm in diameter and 3 cm in depth at the surface part in this study. It was confirmed that the treatable depth was improved almost 2 cm for any incident neutron energy in the case of the 10 cm irradiation field diameter. It was made clear that the improvement effect was larger in isotropic incidence than in parallel incidence, in the case that an irradiation field size was delimited fitting into a void diameter.

Development of neutron shielding material using metathesis-polymer matrix

A neutron shielding material using a metathesis-polymer matrix, which is a thermosetting resin, was developed. This shielding material has characteristics that can be controlled for different mixing ratios of neutron absorbers and for formation in the laboratory. Additionally, the elastic modulus can be changed at the hardening process, from a flexible elastoma to a mechanically tough solid. Experiments were performed at the Kyoto University Research Reactor in order to determine the important characteristics of this metathesis-polymer shielding material, such as neutron shielding performance, secondary gamma-ray generation and activation. The metathesis-polymer shielding material was shown to be practical and as effective as the other available shielding materials, which mainly consist of thermoplastic resin.

モンテカルロ法によるコンクリート中間貯蔵キャスクの遮蔽解析(NP4 核燃料サイクル・バックエンド技術)

This paper describes a neutron and gamma-ray shielding analysis of a concrete interim storage cask with the Monte Carlo code MCNP 5, and evaluated the dose-equivalent date distributions around the cask. The following two remarks are obtained by the Monte Carlo analysis. One is that the radiation streaming from the exhaust ducts is not so noticeable as expected for the ducts are tow-legged ones and the outlets are located away from the upper part of spent fuels and also from the lower part of them, respectively, ant the other is that the secondary gamma-ray dose-equivalent rates on the cask surface are rather large than the neutrons, because a large number of secondary gamma rays are produced by H(n, γ) reactions of thermal neutrons in the thick concrete shield.

Validity of the Four-Parameter Empirical Formula in Approximating the Response Functions for Gamma-ray, Neutron, and Secondary Gamma-ray Skyshine Analyses

A four-parameter approximating formula, R=A e a x b f(x), accurately represents the skyshine line beam response function (LBRF) as a function of the distance (x) of the source-to-detector separation. Here, A is a constant for a given source energy and f(x)=e cx x dx is a damping factor. The four parameters are obtained as follows. 1. The value of parameter a corresponds to that of the LBRF at x=1 meter, which is the result of integrating the basic dose spectrum due to a single scattering particle from an emitted beam for a specified angle and a specified source energy. 2. The value of parameter b corresponds to the slope of a straight line of the response function, log R vs. log x, in the range of small distance from a source, where a single scattering particle dominates. 3. The damping factor (f (x)) represents the attenuation trend of the LBRF at distances far from the source; the values of parameters c and d control the quantity of attenuation. The necessary reference LBRF data for point mono-directional photon source energies ranging from 0.1 to 10MeV were generated using the EGS4 Monte Carlo code at 20 emission angles from 0.0 to 180° for 24 source-detector distances up to 2,000 m. The validity of using the four-parameter formula to interpolate the LBRF in the source-to-detector distance, in the emitted angle, and in the energy was also ascertained. Furthermore, this formula was applied to the skyshine conical beam response function (CBRF) for a neutron and an associated secondary gamma-ray with the source energy ranged from thermal to 3 GeV. It was ascertained that the CBRF could be accurately approximated by an interpolation of the fitting parameters at an arbitrary distance and emitted cosine angle.

Measurement of radiation skyshine with D–T neutron source

The D–T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×10 11 n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9×0.9 m 2 was open during the experimental period. The radiation dose rate outside the target room was measured a maximum distance of 550 m from the D–T target point with a spherical rem-counter. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation counters. The highest neutron dose was about 9×10 −22 Sv/(source neutron) at a distance of 30 m from the D–T target point and the dose rate was attenuated to 4×10 −24 Sv/(source neutron) at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 230 m. The line source model agrees well with the experimental results within the distance of 350 m.

Photonuclear Physics in Radiation Transport—I: Cross Sections and Spectra

This paper describes model calculations and nuclear data evaluations of photonuclear reactions on isotopes of C, O, Al, Si, Ca, Fe, Cu, Ta, W, and Pb for incident photon energies up to 150 MeV. The calculations, using the GNASH code, include giant-dipole resonance and quasi-deuteron models for photoabsorption. The emission of secondary particles and gamma rays is computed using preequilibrium theory, together with an open-ended sequence of compound nucleus decays using the Hauser-Feshbach theory. The accuracy of the calculated and evaluated cross sections is assessed through extensive comparisons with measured cross sections, average neutron multiplicities, and energy-dependent emission spectra. The evaluated nuclear data files (ENDF) facilitate radiation transport studies of the importance of photonuclear reactions in a number of technologies including photoneutrons produced in electron/photon accelerators, shielding studies, and nondestructive detection of nuclear materials. A companion paper describes developments in the MCNP and MCNPX codes to utilize these data in transport simulations.

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

The Monte Carlo coupling technique with the coordinate transformation is used to evalulate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the “SSW-SSR/CRT calculation system.”In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house.

D-T Neutron Skyshine Experiments at JAERI/FNS.

The D-T neutron skyshine experiments have been carried out at the Fusion Neutronics Source (FNS) of JAERI with the neutron yield of ∼1.7×1011n/s. The concrete thickness of the roof and the wall of a FNS target room are 1.15 and 2 m, respectively. The FNS skyshine port with a size of 0.9 × 0.9 m2 was open during the experimental period.The radiation dose rate outside the target room was measured as far as about 550 m away from the D-T target point with a spherical rem-counter. The highest neutron dose was about 0.5 μSv/hr at a distance of 30 m from the D-T target point and the dose rate was attenuated to 0.002 μSv/hr at a distance of 550 m. The measured neutron dose distribution was analyzed with Monte Carlo code MCNP-4B and a simple line source model. The MCNP calculation overestimates the neutron dose in the distance range larger than 250 m. The neutron spectra were evaluated with a 3He detector with different thickness of polyethylene neutron moderators. Secondary gamma-rays were measured with high purity Ge detectors and NaI scintillation detectors.

Validity of the Four-Parameter Empirical Formula in Approximating the Response Functions for Gamma-ray, Neutron, and Secondary Gamma-ray Skyshine Analyses

A four-parameter approximating formula, R=A e a x b f(x), accurately represents the skyshine line beam response function (LBRF) as a function of the distance (x) of the source-to-detector separation. Here, A is a constant for a given source energy and f(x)=e cx x dx is a damping factor. The four parameters are obtained as follows. 1. The value of parameter a corresponds to that of the LBRF at x=1 meter, which is the result of integrating the basic dose spectrum due to a single scattering particle from an emitted beam for a specified angle and a specified source energy.2. The value of parameter b corresponds to the slope of a straight line of the response function, log R vs. log x, in the range of small distance from a source, where a single scattering particle dominates.3. The damping factor (f (x)) represents the attenuation trend of the LBRF at distances far from the source; the values of parameters c and d control the quantity of attenuation. The necessary reference LBRF data for point mono-directional photon source energies ranging from 0.1 to 10MeV were generated using the EGS4 Monte Carlo code at 20 emission angles from 0.0 to 180° for 24 source-detector distances up to 2,000 m. The validity of using the four-parameter formula to interpolate the LBRF in the source-to-detector distance, in the emitted angle, and in the energy was also ascertained. Furthermore, this formula was applied to the skyshine conical beam response function (CBRF) for a neutron and an associated secondary gamma-ray with the source energy ranged from thermal to 3 GeV. It was ascertained that the CBRF could be accurately approximated by an interpolation of the fitting parameters at an arbitrary distance and emitted cosine angle.

Experimental Study on Cross Section Data of Mercury with D-T and 252Cf Neutron Sources

Mercury is a promising candidate material for spallation neutron targets driven by high intensity proton accelerators. Validity of cross section data of mercury up to 20 MeV evaluated for JENDL-3.3 was investigated experimentally by means of both differential and integral approaches. Double differential neutron emission cross section (DDX) for incident neutron energy at 15 MeV was measured. Integral experiment for testing neutron transport and secondary gamma-ray production was conducted by using D-T and 252Cf neutron sources. As a result, general agreement between quantities derived from JENDL and the experimental results was confirmed. Some problems were still pointed out in the DDX data, cross section data in an incident neutron energy range approximately from 0.3 to 3 MeV and secondary gamma-ray production cross section.

Measurements of Secondary Gamma-Ray Production Cross Sections for natFe, 51V, natMo, natZr, natNi and 181Ta with Hp-Ge Detector Induced by DT Neutrons

Secondary gamma-ray production cross sections for natFe, 51V, natMo, nalZr, natNi and 181Ta induced by DT neutrons have been measured with an Hp-Ge detector at OKTAVIAN , the Intense 14MeV Neutron Source Facility of Osaka University, Japan. A new method to analyze the measured spectrum was proposed to yield the continuum component of the gamma-ray production cross section. The measured data were compared with the theoretical calculation results with SINCROS- II and evaluated nuclear data.

An experimental study on radiation streaming through a labyrinth in a proton accelerator facility of intermediate energy.

A radiation streaming experiment has been carried out at the Takasaki Ion Accelerator Facility for Advanced Radiation Application at the Japan Atomic Energy Research Institute in a room housing a Cu target irradiated with 68 MeV protons and in a labyrinth of three-legs having a total length of 29 m. In the experiment, neutron and gamma ray energy spectra, neutron reaction rates, and neutron and gamma ray dose equivalent rates were measured using various counters and dosimeters. The experimental data show the applicability of some empirical formulas for estimating the thermal neutron flux in a room and neutrons streaming in a labyrinth designed for a proton accelerator operating in the intermediate energy region. The data suggest that it is mandatory to estimate the gamma ray dose equivalent rate in a labyrinth, which is dominated by the secondary gamma rays due to the neutron capture reaction.

Inelasticity fluctuations and the efficiency of emulsion chambers for hadrons

Emulsion chamber experiments register high energy hadrons in cosmic rays indirectly, via gamma rays produced by nuclear reactions in the chamber materials (usually carbon and lead). If hadrons fail to produce any gamma rays above the detection threshold of typically a few TeV they will escape detection even in an infinitely thick chamber. The hadron energy is determined by dividing the energy sum of the secondary gamma rays by a mean inelasticity. Here the questions arise whether this conversion factor is the same for mesons and baryons and how large the fluctuations of this inelasticity are. These problems are investigated by simulations of the interactions of charged pions and protons with carbon and lead nuclei using the VENUS interaction model. Our calculations indicate considerable deviations of the detection probability from 1 in the energy region one to two orders of magnitude above the detection threshold for gamma rays, and clear differences between mesons and baryons, also as far as inelasticities and their fluctuations are concerned.

Measurement of Secondary Gamma-ray Skyshine and Groundshine from Intense 14 MeV Neutron Source Facility

Secondary gamma-ray skyshine and groundshine, including the direct contribution from the facility building, have been measured with an Hp-Ge detector and an Nal(Tl) detector at the Intense 14 MeV Neutron Source Facility OKTAVIAN of Osaka University, Japan. The mechanism of secondary gamma-rays propagation were analyzed with the measured spectrum with the Hp-Ge detector. The contribution of the skyshine was shown to be a continuum spectrum that was composed of mainly Compton scattered high energy secondary gamma-rays generated in the facility building created by (n,γ) reaction. The contribution of the groundshine considerably contained secondary gamma-rays generated by natSi (n,γ) reaction in soil, including the albedo contribution from the ground. And the total contribution contained capture gamma-rays from iron (Fe) and other nuclides. The measurements with the Nal(Tl) detector as well as the Hp-Ge detector were carried out to investigate the dependence of gamma-ray dose as a function of distance from the neutron source up to hundreds meters. Consequently, it was found that the dependence could be fitted with the function of const. exp(-r/λ)/rn, where n values were around 2 except for the skyshine (n~1). It was thus indicated that the contribution of the skyshine could be propagated farther downfield than the direct contribution from the facility. The measured ratios of the three contributions (skyshine, groundshine, and direct contributions) and the distance dependence in each path were shown to be in good agreement with calculated results by the Monte Carlo transport code MCNP-4A. And the total contributions for the two detectors of Nal(Tl) and Hp-Ge agree excellently with each other.

Analyses of Neutron and Gamma Ray Measurements in a Target Room of Several Tens MeV Proton Facility

In order to estimate the accuracy of calculation methods such as an empirical formula and Monte Carlo codes for neutron and gamma ray flux in a room with a Cu target irradiated by 68-MeV-proton, experimental analyses are carried out. The analyses reveals that (1) the NMTC/JAERI code is applicable to source neutron calculation, (2) the Patterson’s equation with Ishikawa’s parameter is useful and estimates thermal neutron flux conservatively and (3) the MCNP calculation underestimates the fast neutron scattered back from the wall of the room, consequently the secondary gamma ray from the wall.

Decision-making Methodology of Optimal Shielding Materials by Using Fuzzy Linear Programming

The main purpose of our studies are to select materials and determine the ratio of constituent materials as the first stage of optimum shielding design to suit the individual requirements of nuclear reactors, reprocessing facilities, casks for shipping spent fuel, etc. The parameters of the shield optimization are cost, space, weight and some shielding properties such as activation rates of individual irradiation and cooling time, and total dose rate for neutrons (including secondary gamma ray) and for primary gamma ray. Using conventional two-valued logic (i.e. crisp) approaches, huge combination calculations are needed to identify suitable materials for optimum shielding design. Also, re-computation is required for minor changes, as the approach does not react sensitively to the computation result. Present approach using a fuzzy linear programming method is much of the decisionmaking toward the satisfying solution might take place in fuzzy environment. And it can quickly and easily provide a guiding principle of optimal selection of shielding materials under the above-mentioned conditions. The possibility of reducing radiation effects by optimizing the ratio of constituent materials is investigated.