Gamma-ray Contamination Research Articles

A 3D diffusive and advective model of electron transport applied to the pulsar wind nebula HESS J1825 − 137

ABSTRACT HESS J1825 − 137 is one of the most powerful and luminous TeV gamma-ray pulsar wind nebulae, making it an excellent laboratory to study particle transportation around pulsars. We present a model of the (diffusive and advective) transport and radiative losses of electrons from the pulsar PSR J1826 − 1334 powering HESS J1825 − 137 using interstellar medium gas (ISM) data, soft photon fields, and a spatially varying magnetic field. We find that for the characteristic age of $21\, \mathrm{k}\mathrm{yr}$, PSR J1826 − 1334 is unable to meet the energy requirements to match the observed X-ray and gamma-ray emission. An older age of $40\, \mathrm{k}\mathrm{yr}$, together with an electron conversion efficiency of 0.14 and advective flow of v = 0.002c, can reproduce the observed multiwavelength emission towards HESS J1825 − 137. A turbulent ISM with magnetic field of $B=20 \,{\rm to}\, 60 \,\mathrm{\mu }{\rm G}$ to the north of HESS J1825 − 137 (as suggested by ISM observations) is required to prevent significant gamma-ray contamination towards the northern $\mathrm{T}\mathrm{e\mathrm{V}}$ source HESS J1826 − 130.

Feasibility study of optical imaging of the boron-dose distribution by a liquid scintillator in a clinical boron neutron capture therapy field.

Evaluation of the boron dose is essential for boron neutron capture therapy (BNCT). Nevertheless, a direct evaluation method for the boron-dose distribution has not yet been established in the clinical BNCT field. To date, even in quality assurance (QA) measurements, the boron dose has been indirectly evaluated from the thermal neutron flux measured using the activation method with gold foil or wire and an assumed boron concentration in the QA procedure. Recently, we successfully conducted optical imaging of the boron-dose distribution using a cooled charge-coupled device (CCD) camera and a boron-added liquid scintillator at the E-3 port facility of the Kyoto University Research Reactor (KUR), which supplies an almost pure thermal neutron beam with very low gamma-ray contamination. However, in a clinical accelerator-based BNCT facility, there is a concern that the boron-dose distribution may not be accurately extracted because the unwanted luminescence intensity, which is irrelevant to the boron dose is expected to increase owing to the contamination of fast neutrons and gamma rays. The purpose of this research was to study the validity of a newly proposed method using a boron-added liquid scintillator and a cooled CCD camera to directly observe the boron-dose distribution in a clinical accelerator-based BNCT field. A liquid scintillator phantom with 10 B was prepared by filling a small quartz glass container with a commercial liquid scintillator and boron-containing material (trimethyl borate); its natural boron concentration was 1wt%. Luminescence images of the boron-neutron capture reaction were obtained in a water tank at several different depths using a CCD camera. The contribution of background luminescence, mainly due to gamma rays, was removed by subtracting the luminescence images obtained using another sole liquid scintillator phantom (natural boron concentration of 0wt%) at each corresponding depth, and a depth profile of the boron dose with several discrete points was obtained. The obtained depth profile was compared with that of calculated boron dose, and those of thermal neutron flux which were experimentally measured or calculated using a Monte Carlo code. The depth profile evaluated from the subtracted images indicated reasonable agreement with the calculated boron-dose profile and thermal neutron flux profiles, except for the shallow region. This discrepancy is thought to be due to the contribution of light reflected from the tank wall. The simulation results also demonstrated that the thermal neutron flux would be severely perturbed by the 10 B-containing phantom if a relatively larger container was used to evaluate a wide range of boron-dose distributions in a single shot. This indicates a trade-off between the luminescence intensity of the 10 B-added phantom and its perturbation effect on the thermal neutron flux. Although a partial discrepancy was observed, the validity of the newly proposed boron-dose evaluation method using liquid-scintillator phantoms with and without 10 B was experimentally confirmed in the neutron field of an accelerator-based clinical BNCT facility. However, this study has some limitations, including the trade-off problem stated above. Therefore, further studies are required to address these limitations.

Photon hormesis deactivates alpha-particle induced bystander effects between zebrafish embryos

In the present work, we studied the effects of low-dose X-ray photons on the alpha-particle induced bystander effects between embryos of the zebrafish, Danio rerio. The effects on the naive whole embryos were studied through quantification of apoptotic signals (amounts of cells undergoing apoptosis) at 24h post fertilization (hpf) using vital dye acridine orange staining, followed by counting the stained cells under a fluorescent microscope. We report data showing that embryos at 5 hpf subjected to a 4.4mGy alpha-particle irradiation could release a stress signal into the medium, which could induce bystander effect in partnered naive embryos sharing the same medium. We also report that the bystander effect was deactivated when the irradiated embryos were subjected to a concomitant irradiation of 10 or 14mGy of X-rays, but no such deactivation was achieved if the concomitant X-ray dose dropped to 2.5 or 5mGy. In the present study, the significant drop in the amount of apoptotic signals on the embryos having received 4.4mGy alpha particles together X-rays irradiation from 2.5 or 5mGy to 10 or 14mGy, together with the deactivation of RIBE with concomitant irradiation of 10 or 14mGy of X-rays supported the participation of photon hormesis with an onset dose between 5 and 10mGy, which might lead to removal of aberrant cells through early apoptosis or induction of high-fidelity DNA repair. As we found that photons and alpha particles could have opposite biological effects when these were simultaneously irradiated onto living organisms, these ionizing radiations could be viewed as two different environmental stressors, and the resultant effects could be regarded as multiple stressor effects. The present work presented the first study on a multiple stressor effect which occurred on bystander organisms. In other words, this was a non-targeted multiple stressor effect. The photon hormesis could also explain some failed attempts to observe neutron-induced bystander effects in previous studies, as neutron sources invariably emit neutrons with concomitant gamma-ray photons, which is often referred to as gamma-ray contamination.

Neutron induced bystander effect among zebrafish embryos

Abstract The present paper reported the first-ever observation of neutron induced bystander effect (NIBE) using zebrafish ( Danio rerio ) embryos as the in vivo model. The neutron exposure in the present work was provided by the Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) facility at the National Institute of Radiological Sciences (NIRS), Chiba, Japan. Two different strategies were employed to induce NIBE, namely, through directly partnering and through medium transfer. Both results agreed with a neutron-dose window (20–50 mGy) which could induce NIBE. The lower dose limit corresponded to the threshold amount of neutron-induced damages to trigger significant bystander signals, while the upper limit corresponded to the onset of gamma-ray hormesis which could mitigate the neutron-induced damages and thereby suppress the bystander signals. Failures to observe NIBE in previous studies were due to using neutron doses outside the dose-window. Strategies to enhance the chance of observing NIBE included (1) use of a mono-energetic high-energy (e.g., between 100 keV and 2 MeV) neutron source, and (2) use of a neutron source with a small gamma-ray contamination. It appeared that the NASBEE facility used in the present study fulfilled both conditions, and was thus ideal for triggering NIBE.

Surrounding material effect on measurement of thunderstorm-related neutrons

Observations of strong flux of low-energy neutrons were made by 3He counters during thunderstorms (Gurevich et al., 2012) [11]. How the unprecedented enhancements were produced remains elusive. To better elucidate the mechanism, a simulation study of surrounding material impacts on measurement by 3He counters was performed with GEANT4. It was found that unlike previously thought, a 3He counter had a small sensitivity to high-energy gamma rays because of inelastic interaction with its cathode-tube materials (Al or stainless steel). A 3He counter with the intrinsic small sensitivity, if surrounded by thick materials, would largely detect thunderstorm-related gamma rays rather than those neutrons produced via photonuclear reaction in the atmosphere. On the other hand, the counter, if surrounded by thin materials and located away from a gamma-ray source, would observe neutron signals with little gamma-ray contamination. Compared with the Gurevich measurement, the present work allows us to deduce that the enhancements are attributable to gamma rays, if their observatory was very close to or inside a gamma-ray emitting region in thunderclouds.

Advances in the production, processing and microPET image quality of technetium-94m

This work involves the production, processing and imaging of the short-lived, rarely used positron emission tomography (PET) radionuclide technetium-94m (94mTc). Our procedures are an extension of methods reported in the literature and are detailed within. A key modification was the development of a single step that combines purification and concentration of an aqueous 94mTc-pertechnetate solution, which both reduces processing time and increases the final concentration of the solution. Additionally, a convenient method for the direct recovery of 94mTc into an organic solvent was developed, eliminating the solvent transfer step needed for organic syntheses using 94mTc. Each of these advances potentially extends the scope of syntheses possible with this short-lived radionuclide. To explore the imaging potential of 94mTc, we carried out phantom imaging studies on small-scale high-resolution PET scanners to estimate the limitations of detection associated with 94mTc and PET. Preliminary studies demonstrate that useful images can be obtained with modern image reconstruction algorithms when using a correction for the cascade gamma ray contamination.

Subcriticality Measurement of Pressurized Water Reactors by the Modified Neutron Source Multiplication Method

Applicability of the modified Neutron Source Multiplication (NSM) method with extraction of the fundamental mode to subcriticality measurement has been investigated. This concept can be applied to the control rod worth measurement that is performed during reactor physics tests at periodical inspection of commercial PWRs. Following the feasibility verification in the previous study based on numerical analyses, its applicability has been proven in a more realistic situation; Regarding a withdrawal sequence of control rod banks during the PWR startup as a mockup experiment, subcriticalities with various control rod insertion patterns were estimated based on the modified NSM method. The subcriticality could be evaluated with good accuracy even with the mockup experiment where any special treatments for accurate measurement were not taken into account and furthermore the insensitivity of measuring signals by reactivity changes and their large fluctuations were seen. Some vital treatments for accurate estimation were identified: The reference subcritical state must be as closer to zero as possible. It is necessary to wait long enough for the delayed neutron effect to attenuate after a reactivity change by control rod bank withdrawal. Gamma-ray contamination contained in the detector signals should be corrected.

Startup of the Fission Converter Epithermal Neutron Irradiation Facility at the MIT Reactor

A new epithermal neutron irradiation facility, based on a fission converter assembly placed in the thermal column outside the reactor core, has been put into operation at the Massachusetts Institute of Technology Research Reactor (MITR). This facility was constructed to provide a high-intensity, forward-directed beam for use in neutron capture therapy with an epithermal flux of ≈ 1010 n/cm2·s at the medical room entrance with negligible fast neutron and gamma-ray contamination. The fission converter assembly consists of 10 or 11 MITR fuel elements placed in an aluminum tank and cooled with D2O. Thermal-hydraulic criteria were established based on heat deposition calculations. Various startup tests were performed to verify expected neutronic and thermal-hydraulic behavior. Flow testing showed an almost flat flow distribution across the fuel elements with <5% bypass flow. The total reactivity change caused by operation of the facility was measured at 0.014 ± 0.002% ΔK/K. Thermal power produced by the facility was measured to be 83.1 ± 4.2 kW. All of these test results satisfied the thermal-hydraulic safety criteria. In addition, radiation shielding design measurements were made that verified design calculations for the neutronic performance.

Somatic Mutation Frequencies in Tradescantia Stamen Hairs Treated with Relatively Low Thermal Neutron Fluxes

Young inflorescence-bearing cuttings of Tradescantia clone BNL 02, a blue/pink heterozygote, were treated with thermal neutrons for 12-120 s in the Irradiation Tube set in the Heavy Water Facility of the KUR Reactor of the Research Reactor Institute, Kyoto University. In this facility, practically all components of the neutron flux were thermal neutrons. The thermal neutron fluxes measured simultaneously with gold foils were 2.90-33.3 x 10(10) nth cm-2, and the gamma-ray contamination per 10(10) nth cm-2 was 2.58 x 10(-4) C/kg (1.0 R). The induced somatic pink mutation frequencies in the stamen hairs increased linearly with thermal neutron flux at a rate of 1.69 pink mutant events per 10(4) hair-cell divisions per 10(10) nth cm-2 in Experiments 1 and 2, but at a higher rate of 3.87 pink mutant events per 10(4) hair-cell divisions per 10(10) nth cm-2 in Experiment 3. The relatively small thermal neutron fluxes applied were calculated to have resulted in absorbed doses of 16.8-193 mGy, and the absorbed doses of contaminating gamma rays were 28.0-321 mGy. Therefore, these mutation frequencies could be converted into 109 and 251 pink mutant events per 10(4) hair-cell divisions per gray, including the contaminating gamma rays, and were 2.14-4.92 times higher than the 51.0 pink mutant events per 10(4) hair-cell divisions per gray for relatively small doses of acute X rays which was obtained in earlier studies. The relative biological effectiveness of heavy particles (p, alpha and 7Li) produced by thermal neutrons was calculated to be between 11.0 and 35.4 compared with acute X rays rather than with 18-20-h exposure to gamma rays, the value of 11.0 being considered to be more reliable.

Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 109 n cm−2 s−1 and have sufficiently low fast-neutron and gamma-ray contamination.

A monoenergetic neutron beam facility for biological irradiations

This paper describes in detail the calibration methods and utilisation of a fast neutron source for biological irradiations, which is relatively free from gamma ray contamination. The project arose from the need for biologists to have access to a source of high energy neutrons (>1 MeV) for comparison of data with that already obtained using lower LET sources such as X-rays and γ-rays. This Standard Neutron Irradiation Facility (SNIF) was set up on a 3 MV Van de Graaf accelerator and produced well characterised monoenergetic beams of 3 to 6 MeV neutrons using the 2D(d,n) 3He nuclear reaction. Full width neutron energy spreads of 200 keV for 5.4 MeV neutrons were readily obtainable as were fluxes up to 6×10 8 n/cm 2 s producing dose rates up to 100 Gy/h. The γ-ray contamination was estimated to be less than 5% for most neutron irradiations. The major limitation to the maximum dose rates attainable was the maximum power dissipation in the thin Havar window separating the pressurised cell deuterium gas from the evacuated beamline.

The thermal neutron sensitivities of SEI and Toshiba FD-P10-7 radiophotoluminescent glass dosimeters

The thermal neutron sensitivities of SEI and Toshiba FD-P10-7 radiophotoluminescent silver activated metaphosphate glass dosimeters have been measured and studied theoretically. The graphite thermal neutron column attached to the AEEW reactor NESTOR was utilised as a source of thermal neutrons with minimal fast neutron and gamma-ray contamination. Both the neutron removal and gamma correction measurement techniques were applied and these gave consistent results. It was found that the thermal neutron sensitivity of the Toshiba dosimeters is comparable to the most neutron insensitive thermoluminescent materials available. Because of their larger size and higher silver concentration, however, SEI dosimeters are an order of magnitude more neutron sensitive. The neutron sensitivity of FD-7 glass has not been reported previously, whereas in the case of SEI glass, there exists a modest body of previous work on dosimeters of nominally identical composition. The present results are in reasonable agreement with these earlier determinations. A simple physical model was constructed in order to interpret the measured responses. The model takes into account neutron-capture recoil energy, neutron-capture gamma-rays and the beta- and gamma-ray emissions from neutron induced radionuclides created within the dosimeters. The calculated and experimental values are in excellent agreement indicating that the response of these dosimeters in mixed thermal neutron/gamma-ray fields is sufficiently well understood to be applied with confidence.

A Neutronic Study of an Accelerator-Based Neutron Irradiation Facility for Boron Neutron Capture Therapy

A neutronic study of an accelerator-based neutron irradiation facility (ANIF) for boron neutron capture therapy (BNCT) was performed using three-dimensional Monte Carlo transport calculations. The major components of the ANIF are a radio-frequency quadrupole proton accelerator, a /sup 7/Li target, and a moderator assembly. Neutrons are generated by bombarding the /sup 7/Li target with 2.5-MeV protons. The neutrons emerging from the /sup 7/Li target are too energetic to be used for BNCT and must therefore be moderated. Calculations show that, among all materials for the ANIF, beryllia (BeO) and heavy water (D/sub 2/O) are the best moderators. Between them, beryllia provides better neutron spectra, but D/sub 2/O gives higher neutron intensities. Adding alumina (Al/sub 2/O/sub 3/) to D/sub 2/O improves the neutron spectra, but it also increases gamma-ray contamination.

Microdosimetric Spectra Measurements of Janus Neutrons

Neutron radiation from the JANUS reactor at Argonne National Laboratory is being used with increasing frequency for major biological experiments. The fast neutron spectrum has a Kerma-weighted mean energy of 0.8 MeV and low gamma-ray contamination. In 1984 the JANUS fission converter plate of highly enriched uranium was replaced by one made of low-enriched uranium. We recorded microdosimetric spectra at several different positions in the high-flux irradiation room of JANUS before the change of the converter plate. Each set of measurements consisted of spectra taken at three different site diameters (0.5, 1.0, and 5.0 ..mu..m) and in both ''attenuator up'' and ''attenuator down'' configurations. At two conventional dosimetry reference positions, two sets of measurements were recorded. At three biological reference positions, measurements simulating several biological irradiation conditions, were taken. The dose rate at each position was estimated and compared with dose rates obtained previously by conventional dosimetry. Comparison of the different measurements showed no major change in spectra as a function of position or irradiation condition. First results from similar sets of measurements recorded after the installment of the new converter plate indicate no major change in the spectra. 11 refs., 4 figs., 5 tabs.

Microdosimetric Spectra Measurements of Janus Neutrons

Neutron radiation from the JANUS reactor at Argonne National Laboratory is being used with increasing frequency for major biological experiments. The fast neutron spectrum has a Kerma-weighted mean energy of 0.8 MeV and low gamma-ray contamination. In 1984 the JANUS fission converter plate of highly enriched uranium was replaced by one made of low-enriched uranium. We recorded microdosimetric spectra at several different positions in the high-flux irradiation room of JANUS before the change of the converter plate. Each set of measurements consisted of spectra taken at three different site diameters (0.5, 1.0, and 5.0 ..mu..m) and in both ''attenuator up'' and ''attenuator down'' configurations. At two conventional dosimetry reference positions, two sets of measurements were recorded. At three biological reference positions, measurements simulating several biological irradiation conditions, were taken. The dose rate at each position was estimated and compared with dose rates obtained previously by conventional dosimetry. Comparison of the different measurements showed no major change in spectra as a function of position or irradiation condition. First results from similar sets of measurements recorded after the installment of the new converter plate indicate no major change in the spectra. 11 refs., 4 figs., 5 tabs.

Intercomparison of personnel dosimetry for thermal neutron dose equivalent in neutron and gamma ray mixed fields.

In order to consider the problems concerned with personnel dosimetry, an Intercomparison of personnel dosimetry, especially dose equivalent responses of personnel dosimeters to thermal neutron, was carried out in neutron and gamma ray mixed fields.The dose equivalent responses and their standard deviations of individual dosimeter were almost independent on cadmium ratio and gamma ray contamination in the radiation fields. However, a difference of about 4 times was observed among the relative thermal neutron dose equivalents of individual dosimeter normalized to the ICRP recommended value.It is suggested that the standardization of dose evaluation method is required from the practical point of radiation protection and safety.

Experimental tests of the Bonner sphere spectrometer using filtered neutron beams

The operation of a Bonner Sphere neutron detector system has been tested using several unique neutron sources. Filtered neutron beams at beamport F at the University of Missouri Research Reactor (MURR) were used as a source of known quasi-monoenergetic neutrons for precise energy spectra analysis and calibration. A PuBe neutron source was used for absolute flux magnitude and spectral verification. Two computer codes, SWIFT and Least Squares Unfolding Techniques (LSUT), were used to unfold the experimental data. Several operational problems were encountered during these tests. First, many of the measurements involved neutron beam measurements in which the beams had a smaller diameter than the moderating spheres. This caused partial illumination of the spheres for which correction factors had to be developed. A partial illumination correction factor has been proposed and tested to account for this problem. Second, reactor core gamma-ray contamination in the neutron beams was of sufficient magnitude to interfere with some measurements. Gamma-ray background subtraction techniques using a multi-channel analyzer were used to alleviate this problem. After correcting for gamma-ray background and applying partial illumination correction factors, unfolded neutron spectra from the unfolding codes gave good results for most neutron sources. In particular the SWIFT results were quite good, exceeding expectations in terms of energy resolution and spectral accuracy.

Performance of Neutron Radiography Systems Using Particle Accelerators

The most crucial element of building a useful neutron radiography system is an intense source of neutrons. Among nuclear reactions employing accelerators, the 9Be(d,n)10B reaction is probably the best method for generating neutrons for radiography. This is due to its superior yield of neutrons that can be efficiently moderated and very low gamma-ray contamination. This has been demonstrated using a KN-4000 Van de Graaff accelerator coupled to an IRT Model ANR-100 moderator-collimator assembly, and excellent quality radiographs have been obtained. The Radio Frequency Quadrupole Linear Accelerator is found to be capable of providing an intense source of neutrons using the same reaction and is recommended for neutron radiography applications.

The correction for the gamma-ray component in neutron therapy.

Neutron beams for therapy always contain some gamma ray contamination that varies with depth and with distance from the beam axis. The problem therefore arises how the varying gamma ray contribution should be accounted for in dose specification. Not infrequently a "total effective dose" DE is quoted that is equal to the neutron dose plus the gamma ray dose divided by a constant weight factor tau. On general biophysical considerations this appears to be not a valid approach since it must be assumed that tau decreases with increasing dose. The nature and the magnitude of this dose dependence is derived in the present article. Application of the results to actual doses per fraction and to factual gamma ray to neutron ratios demonstrates that the dose dependence of tau has, in fact, very minor influence on the numerical values of DE. Utilization of a constant value tau is therefore satisfactory in practice.

Thermal neutron standard field with a Maxwellian distribution using the KUR heavy water facility

A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure themal neutrons with a Maxwellian distribution. The facility is placed next to the core of KUR and contains about 2 t of heavy water. The width of the heavy water layer is about 140 cm. The neutron spectrum was measured with the time-of-flight technique using a fast chopper. The measured spectrum was in good agreement with a Maxwellian distribution in the whole energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons from the γ-n reaction in heavy water is very small. The maximum intensity of thermal neutrons is 3 × 10 11 n/cm·s. When a bismuth scatterer is attached, the gamma ray contamination is decreased to a ratio of 0.05 of gamma rays to neutrons in Rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, diffusion length measurement, detector calibration, activation analysis and for biomedical purposes.