• Nuclear Engineering and Technology
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• v.5 no.3
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• pp.191-201
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• 1973

For fast neutron spectroscopy in MeV region, a recoil proton telescope detector was designed and constructed so as to increase in detection efficiency without appreciable deterioration in energy resolution by adopting a special type of recoil proton radiator which is a combination of a ring-shaped vertical radiator and a cone-shaped horizontal radiator at a certain geometry. A neutron stopper was built in the detector system to minimize the background due to direct exposure of the Si(Li) detectors to primary incident neutrons. The detection efficiency and the energy resolution calculated at various neutron energies and geometries are given and these characteristics of the detector system were tested by 14.1 MeV neutrons. As the calculation predicted, the relative detection efficiency in case of the combined radiator system is almost 2.2 times of that for a single, ring-shaped vertical radiator system. The calculated energy resolution is 3.7% FWHM, whereas the measured resolution was 3.9% which means resolution broadening of approximately. 30% was resulted by introducing a combined radiator system into the telescope. Increase in background less than 40% was also observed.

• Nuclear Engineering and Technology
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• v.18 no.2
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• pp.92-96
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• 1986

The $^{93}Nb(n,n\alpha)^{89m}Y$, $^{93}Nb(n,{\alpha})^{90m}Y$ and $^{93}Nb(n,2n)^{92m}Nb$ cross sections at a neutron energy of 14.6 MeV have been measured relative to the $^{27}Al(n,p)^{27}Mg$ and $^{27}Al(n,{\alpha})^{24}Na$ cross sections. A small accelerator utilizing $T(D,n)^4He$ reaction was used as a neutron source and the neutron energy spread is about 0.4MeV at the sample. All induced activities were measured with a 70cc HPGe detector in the same geometry.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3591-3598
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• 2023

Photoneutron background spectroscopy observations at linac are essential for directing accelerator shielding and subtracting background signals. Therefore, we constructed a Bonner Sphere Spectrometer (BSS) system based on an array of BF₃ gas proportional counter tubes. Initially, the response of the BSS system was simulated using the MCNP5 code. Next, the response of the system was calibrated by using neutrons with energies of 2.86 MeV and 14.84 MeV. Then, the system was employed to measure the spectrum of the ²⁴¹Am-Be neutron source, and the results were unfolded by using the Gravel and EM algorithms. Using the validated system, the undesirable neutron spectrum of the 120 MeV electron linac was finally measured and acquired. In addition, it is demonstrated that the equivalent undesirable neutron dose at a distance of 3.2 m from the linac is 19.7 mSv/h. The results measured by the above methods could provide guidance for linac-related research.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Journal of Radiation Protection and Research
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• v.37 no.1
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• pp.35-40
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• 2012

Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

• Nuclear Engineering and Technology
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• v.56 no.2
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• pp.402-409
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• 2024

High energy neutron irradiations impact on structural and electrical properties of alumina are studied with particular emphasis on real time in-situ radiation induced conductivity measurement in low flux region. Polycrystalline Al₂O₃ samples are subjected to high energy neutrons produced from D-T neutron generator and Am-Be neutron source. 14 MeV neutrons from D-T generator are chosen to study the role of fast neutron irradiation in the structural modification of samples. Real time in-situ electrical measurement is performed to investigate the change in insulation resistance of Al₂O₃ due to radiation induced conductivity at low flux regime. During neutron irradiation, a significant transient decrease in insulation resistance is observed which recovers relative higher value just after neutron exposure is switched off. XRD results of 14 MeV neutron irradiated samples suggest annealing effect. Impact of relatively low energy neutrons on the structural properties is also studied using Am-Be neutrons. In this case, clustering is observed on the sample surface after prolonged neutron exposure. The structural characterizations of pristine and irradiated Al₂O₃ samples are performed using XRD, SEM, and EDX. The results from these characterizations are analysed and interpreted in the manuscript.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2334-2337
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• 2022

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1355-1366
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• 2020

A review of the documentation and an interpretation of the NEA-1517/74 and NEA-1517/80 shielding benchmarks (measurements of photon leakage flux from a hollow sphere with a central 14 MeV neutron source) from the SINBAD database with the Monte Carlo code MCS and the most up-to-date ENDF/B-VIII.0 neutron data library are conducted. The two analyzed benchmarks describe satisfactorily the energy resolution of the photon detector and the geometry of the spherical samples with inner beam tube, tritium target and cooling water circuit, but lack information regarding the detector geometry and the distances of shields and collimators relatively to the neutron source and the detector. Calculations are therefore conducted for a sphere model only. A preliminary verification of MCS neutron-photon calculations against MCNP6.2 is first conducted, then the impact of modelling the inner beam tube, tritium target and cooling water circuit is assessed. Finally, a comparison of calculated results with the libraries ENDF/B-VII.1 and ENDF/B-VIII.0 against the measurements is conducted and shows reasonable agreement. The MCS and MCNP inputs used for the interpretation are available as supplementary material of this article.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1234-1239
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• 2017

Basic research on the accelerator-driven system is conducted by combining $^{235}U$-fueled and $^{232}Th$-loaded cores in the Kyoto University Critical Assembly with the pulsed neutron generator (14 MeV neutrons) and the proton beam accelerator (100 MeV protons with a heavy metal target). The results of experimental subcriticality are presented with a wide range of subcriticality level between near critical and 10,000 pcm, as obtained by the pulsed neutron source method, the Feynman-${\alpha}$ method, and the neutron source multiplication method.

• Transactions of Materials Processing
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• v.8 no.3
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• pp.241-241
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• 1999

With the aim of assessing the degradation of Zr-2.5Nb pressure tubes operating in the Wolsong unit-1 nuclear power plant, characterization tests are being conducted on irradiated Zr-2.5Nb tubes removed after 10-year operation. The examined tube had been exposed to temperatures ranging from 264 to 306℃ and a neutron fluence of 8.9×$10^{21}$ n/cm²(E>1 MeV) at the maximum. Tensile tests were carried out at temperatures ranging from RT to 300℃. The density of a-type and c-type dislocations was examined on the irradiated Zr-2.5Nb tube using a transmission electron microscope. Neutron irradiation up to 8.9×$10^{21}$ n/cm²(E>1 MeV) yielded an increase in a-type dislocation density of the Zr-2.5Nb pressure tube to 7.5×$10^{14} m^{-2}$, which was highest at the inlet of the tube exposed to the low temperature of 275℃. In contrast, the c-component dislocation density did not change with irradiation, keeping an initial dislocation density of 0.8×$10^{14} m^{-2}$ over the whole length of the tube. As expected, the neutron irradiation increased mechanical strength by about 17-26% in the transverse direction and by 34-39% in the longitudinal direction compared to that of the unirradiated tube at 300℃. The change in the mechanical properties with irradiation is discussed in association with the microstructural change as a function of temperature and neutron fluence.