• Title/Summary/Keyword: neutrons

Search Result 320, Processing Time 0.019 seconds

Computational design and characterization of a subcritical reactor assembly with TRIGA fuel

  • Asuncion-Astronomo, Alvie;Stancar, Ziga;Goricanec, Tanja;Snoj, Luka
    • Nuclear Engineering and Technology
    • /
    • v.51 no.2
    • /
    • pp.337-344
    • /
    • 2019
  • The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a $7{\times}7$ square lattice. This configuration is found to have a maximum $k_{eff}$ value of $0.95001{\pm}0.00009$ at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be $748pcm{\pm}7pcm$ and $41{\mu}s$, respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines.

Experimental setup for elemental analysis using prompt gamma rays at research reactor IBR-2

  • Hramco, C.;Turlybekuly, K.;Borzakov, S.B.;Gundorin, N.A.;Lychagin, E.V.;Nehaev, G.V.;Muzychka, A. Yu;Strelkov, A.V.;Teymurov, E.
    • Nuclear Engineering and Technology
    • /
    • v.54 no.8
    • /
    • pp.2999-3005
    • /
    • 2022
  • The new experimental setup has been built at the 11b channel of the IBR-2 research reactor at FLNP, JINR, to study the elemental composition of samples by registration of prompt gamma emission during thermal neutron capture. The setup consists of a curved mirror neutron guide and a radiation-resistant HPGe high-purity germanium detector. The detector is surrounded by lead shielding to suppress the natural background gamma level. The sample is placed in a vacuum channel and surrounded by a LiF shield to suppress the gamma background generated by scattered neutrons. This work presents characteristics of the experimental setup. An example of hydrogen concentration determining in a diamond powder made by detonation synthesis is given and on its basis, the sensitivity of the setup is calculated being ~4 ㎍.

Radiation effect on the polymer-based capacitive relative humidity sensors

  • Shchemerov, I.V.;Legotin, S.A.;Lagov, P.B.;Pavlov, Y.S.;Tapero, K.I.;Petrov, A.S.;Sidelev, A.V.;Stolbunov, V.S.;Kulevoy, T.V.;Letovaltseva, M.E.;Murashev, V.N.;Konovalov, M.P.;Kirilov, V.N.
    • Nuclear Engineering and Technology
    • /
    • v.54 no.8
    • /
    • pp.2871-2876
    • /
    • 2022
  • The sensitivity of polymer-based capacitive relative humidity (RH) sensors after irradiation with neutrons, electrons and protons was measured. Degradation consists of the decreasing of the upper RH limit that can be measured. At the same time, low RH-level sensitivity is almost stable. After 30 krad of absorption dose, RH cut off is equal to 85% of max value, after 60 krad-40%. Degradation reduces after annealing which indicates high radiation sensitivity of the internal circuit in comparison to RH-sensing polymer film.

Determining PGAA collimator plug design using Monte Carlo simulation

  • Jalil, A.;Chetaine, A.;Amsil, H.;Embarch, K.;Benchrif, A.;Laraki, K.;Marah, H.
    • Nuclear Engineering and Technology
    • /
    • v.53 no.3
    • /
    • pp.942-948
    • /
    • 2021
  • The aim of this work is to help inform the decision for choosing a convenient material for the PGAA (Prompt Gamma Activation Analysis) collimator plug to be installed at the tangential channel of the Moroccan Triga Mark II Research Reactor. Two families of materials are usually used for collimator construction: a mixture of high-density polyethylene (HDPE) with boron, which is commonly used to moderate and absorb neutrons, and heavy materials, either for gamma absorption or for fast neutron absorption. An investigation of two different collimator designs was performed using N-Particle Monte Carlo MCNP6.2 code with the ENDF/B-VII.1 and MCLIP84 libraries. For each design, carbon steel and lead materials were used separately as collimator heavy materials. The performed study focused on both the impact on neutron beam quality and the neutron-gamma background at the exit of the collimator beam tube. An analysis and assessment of the principal findings is presented in this paper, as well as recommendations.

Simulation, design optimization, and experimental validation of a silver SPND for neutron flux mapping in the Tehran MTR

  • Saghafi, Mahdi;Ayyoubzadeh, Seyed Mohsen;Terman, Mohammad Sadegh
    • Nuclear Engineering and Technology
    • /
    • v.52 no.12
    • /
    • pp.2852-2859
    • /
    • 2020
  • This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

MRI-Guided Gadolinium Neutron Capture Therapy

  • Ji-Ae Park;Jung Young Kim;Hee-Kyung Kim
    • Journal of Radiopharmaceuticals and Molecular Probes
    • /
    • v.8 no.2
    • /
    • pp.113-118
    • /
    • 2022
  • Gadolinium neutron capture therapy (Gd-NCT) is a precision radiation therapy that kills cancer cells using the neutron capture reaction that occurs when 157Gd hits thermal neutrons. 157Gd has the highest thermal neutron capture cross section of 254,000 barns among stable isotopes in the periodic table. Another stable isotope, 155Gd, also has a high thermal neutron trapping area (~ 60,700 barns), so gadolinium that exists in nature can be used as a Gd-NCT drug. Gd-NCT is a mixed kinetic energy of low-energy and high-energy ionizing particles, which can be uniformly distributed throughout the tumor tissue, thereby solving the disadvantage of heterogeneous dose distribution in tumor tissue. The Gd complexes of small-sized molecule are widely used as contrast agents for magnetic resonance imaging (MRI) in clinical practice. Therefore, these compounds can be used not only for diagnosis but also therapy when considering the concept of Gd-NCT. This multifunctional trial can look forward to new medical advance into NCT clinical practices. In this review, we introduce gadolinium compounds suitable for Gd-NCT and describe the necessity of image guided Gd-NCT.

Discrimination of neutrons and gamma-rays in plastic scintillator based on spiking cortical model

  • Bing-Qi Liu;Hao-Ran Liu;Lan Chang;Yu-Xin Cheng;Zhuo Zuo;Peng Li
    • Nuclear Engineering and Technology
    • /
    • v.55 no.9
    • /
    • pp.3359-3366
    • /
    • 2023
  • In this study, a spiking cortical model (SCM) based n-g discrimination method is proposed. The SCM-based algorithm is compared with three other methods, namely: (i) the pulse-coupled neural network (PCNN), (ii) the charge comparison, and (iii) the zero-crossing. The objective evaluation criteria used for the comparison are the FoM-value and the time consumption of discrimination. Experimental results demonstrated that our proposed method outperforms the other methods significantly with the highest FoM-value. Specifically, the proposed method exhibits a 34.81% improvement compared with the PCNN, a 50.29% improvement compared with the charge comparison, and a 110.02% improvement compared with the zero-crossing. Additionally, the proposed method features the second-fastest discrimination time, where it is 75.67% faster than the PCNN, 70.65% faster than the charge comparison and 38.4% slower than the zero-crossing. Our study also discusses the role and change pattern of each parameter of the SCM to guide the selection process. It concludes that the SCM's outstanding ability to recognize the dynamic information in the pulse signal, improved accuracy when compared to the PCNN, and better computational complexity enables the SCM to exhibit excellent n-γ discrimination performance while consuming less time.

Thermal-annealing behavior of in-core neutron-irradiated epitaxial 4H-SiC

  • Junesic Park ;Byung-Gun Park;Gwang-Min Sun
    • Nuclear Engineering and Technology
    • /
    • v.55 no.1
    • /
    • pp.209-214
    • /
    • 2023
  • The effect of thermal annealing on defect recovery of in-core neutron-irradiated 4H-SiC was investigated. Au/SiC Schottky diodes were manufactured using a 4H-SiC epitaxial wafer that was neutron-irradiated at the HANARO research reactor. The electrical characteristics of their epitaxial layers were analyzed under various conditions, including different neutron fluences (1.3 × 1017 and 2.7 × 1017 neutrons/cm2) and annealing times (up to 2 h at 1700 ℃). Capacity-voltage measurements showed high carrier compensation in the neutron-irradiated samples and a recovery tendency that increased with annealing time. The carrier density could be recovered up to 77% of the bare sample. Deep-level-transient spectroscopy revealed intrinsic defects of 4H-SiC with energy levels 0.47 and 0.68 eV below the conduction-band edge, which were significantly increased by in-core neutron irradiation. A previously unknown defect with a high electron-capture cross-section was discovered at 0.36 eV below the conduction-band edge. All defect concentrations decreased with 1700 ℃ annealing; the decrease was faster when the defect level was shallow.

On-the-fly energy release per fission model in STREAM with explicit neutron and photon heating

  • Nhan Nguyen Trong Mai;Woonghee Lee;Kyeongwon Kim;Bamidele Ebiwonjumi;Wonkyeong Kim;Deokjung Lee
    • Nuclear Engineering and Technology
    • /
    • v.55 no.3
    • /
    • pp.1071-1083
    • /
    • 2023
  • The on-the-fly energy release per fission (OTFK) model is implemented in STREAM to continuously update the Kappa values during the depletion calculation. The explicit neutron and photon energy distribution, which has not been considered in previous STREAM versions, is incorporated into the existing on-the-fly model. The impacts of the modified OTFK model with explicit neutron and photon heating in STREAM on the power distribution, fuel temperature, and other core parameters during depletion with feedback calculations are studied using several problems from the VERA benchmark suit. Overall, the explicit heating calculation provides a better power map for the feedback calculations particularly when strong gamma emitters are present. Generally, the fuel temperature decreases when neutron and photon heating is employed because fission neutrons and gamma rays are transported away from their points of generation. This energy release model in STREAM indicates that gamma energy accounts for approximately 9.5%-10% of the total energy released, and approximately 2.4%-2.6% of the total energy released will be deposited in the coolant for the VERA 5, NuScale, and Yonggwang Unit 3 2D cores.

Nano Yttrium-90 and Rhenium-188 production through medium medical cyclotron and research reactor for therapeutic usages: A Simulation study

  • Abdollah Khorshidi
    • Nuclear Engineering and Technology
    • /
    • v.55 no.5
    • /
    • pp.1871-1877
    • /
    • 2023
  • The main goal of the coordinated project development of therapeutic radiopharmaceuticals of Y-90 and Re-188 is to exploit advancements in radionuclide production technology. Here, direct and indirect production methods with medium reactor and cyclotron are compared to evaluate derived neutron flux and production yield. First, nano-sized 186W and 89Y specimens are suspended in water in a quartz vial by FLUKA simulation. Then, the solution is irradiated for 4 days under 9E+14 n/cm2/s neutron flux of reactor. Also, a neutron activator including three layers-lead moderator, graphite reflector, and polyethylene absorbent- is simulated and tungsten target is irradiated by 60 MeV protons of cyclotron to generate induced neutrons for 188W and 90Sr production via neutron capture. As the neutron energy reduced, the flux gradually increased towards epithermal range to satisfy (n/2n,γ) reactions. The obtained specific activities at saturation were higher than the reported experimental values because the accumulated epithermal flux and nano-sized specimens influence the outcomes. The beta emitters, which are widely utilized in brachytherapy, appeal an alternative route to locally achieve a rational yield. Therefore, the proposed method via neutron activator may ascertain these broad requirements.