Potential Production of Theranostic Boron Nitride Nanotubes (64Cu-BNNTs) Radiolabeled by Neutron Capture

In this work, the radioisotope 64Cu was obtained from copper (II) chloride dihydrate in a nuclear research reactor by neutron capture, (63Cu(n,γ)64Cu), and incorporated into boron nitride nanotubes (BNNTs) using a solvothermal process. The produced 64Cu-BNNTs were analyzed by TEM, MEV, FTIR, XDR, XPS and gamma spectrometry, with which it was possible to observe the formation of64Cu nanoparticles, with sizes of up to 16 nm, distributed through nanotubes. The synthesized of 64Cu nanostructures showed a pure photoemission peak of 511 keV, which is characteristic of gamma radiation. This type of emission is desirable for Photon Emission Tomography (PET scan) image acquisition, as well as its use in several cancer treatments. Thus, 64Cu-BNNTs present an excellent alternative as theranostic nanomaterials that can be used in diagnosis and therapy by different techniques used in nuclear medicine.

Cyclical Patterns Affect Microbial Dynamics in the Water Basin of a Nuclear Research Reactor

The BR2 nuclear research reactor in Mol, Belgium, runs in successive phases of operation (cycles) and shutdown, whereby a water basin surrounding the reactor vessel undergoes periodic changes in physico-chemical parameters such as flow rate, temperature, and radiation. The aim of this study was to explore the microbial community in this unique environment and to investigate its long-term dynamics using a 16S rRNA amplicon sequencing approach. Results from two sampling campaigns spanning several months showed a clear shift in community profiles: cycles were mostly dominated by two Operational Taxonomic Units (OTUs) assigned to unclassified Gammaproteobacterium and Pelomonas, whereas shutdowns were dominated by an OTU assigned to Methylobacterium. Although 1 year apart, both campaigns showed similar results, indicating that the system remained stable over this 2-year period. The community shifts were linked with changes in physico-chemical parameters by Non-metric Multidimensional Scaling (NMDS) and correlation analyses. In addition, radiation was hypothesized to cause a decrease in cell number, whereas temperature had the opposite effect. Chemoautotrophic use of H2 and dead cell recycling are proposed to be used as a strategies for nutrient retrieval in this extremely oligotrophic environment.

Reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor

The leakage from the reactor pool back into the dry irradiation channels due to corrosion or mechanics based reason is a postulated event that could occur under operating conditions of the Dalat nuclear research reactor (DNRR), especially the channel 7-1 which has been installed more than 30 years. When it occurs, the air space in these channels will be occupied by the water, subsequently a water column will appear in fuel region. The appearance of water column considerably enhances medium of neutron moderation for its surrounding fuel assemblies. As a result, a positive reactivity is inserted in the core and this event is classified as an insertion of excess reactivity. This event needs to be addressed by analysis and assessment from safety point of view and the results of analysis are also important for updating the reactor operating procedures. This paper presents assumptions, computer models and the results of analysis for such event in the DNRR by using MCNP5 code (code for neutronics analysis) and EUREKA-2/RR code (code for transient analysis). The calculation results include value of reactivity insertion, change in power of reactor, as well as surface temperature of the hottest fuel assembly. This research contributes to updating the reactor operating procedure.

Calculation Results for Enhancing Ability of I-131 Radioisotope Production Using Tellurium Dioxide Target on the Dalat Nuclear Research Reactor

The paper presents the calculation results in re-design of neutron trap of the Dalat Nuclear Research Reactor (DNRR) for I-131 radioisotope production using TeO2 target. The new design permits for loading more TeO2 capsules from 9 to 12, 15 and 18 in the neutron trap. The enhancement of radioisotope production was implemented by re-arrangement of the neutron trap without changing the dimension or geometry of irradiation capsules. By using neutronics computer code as MCNP6, the obtained calculation results of I-131 activity in 6 investigated cases showed that the new design by the re-arrangement of the neutron trap can be used effectively for radioisotope production with thermal neutron flux in average range from 5.3×1012 to 1×1013 n/cm2.s and the total activity of I-131 isotope was increased from about 19.2% to 38.8% comparing with the original design using 9 capsules. The negative reactivity insertion was from 0.60 βeff to 0.96 βeff when loading capsules that also met the safety requirements of operational conditions of the DNRR.

Simulation of absorbed doses distribution in a polyethylene phantom for BNCT application at the Dalat research reactor

In this paper, the distribution of absorbed dose components in a polyethylene phantom for BNCT application at Dalat Nuclear Research Reactor (DNRR) were calculated using the MCNP5 code. The configuration of horizontal neutron channel No.2 of the DNRR, which contains a cylindrical collimator with neutron filters of 20-cm Si and 3-cm Bi, was simulated. The results show that the gamma dose along the central axis of the phantom has the maximum value of 1.82×10-6 Gy at the 0.5-cm depth, and reduces to 9.05×10-7 Gy at the 3-cm depth. The main contribution to gamma dose is due to the interaction of thermal neutron with hydrogen in the phantom via the 1H(n,γ)2H reaction, and its value is much smaller than thermal neutron dose. The total absorbed dose along the central axis of the phantom has the maximum value of 7.87×10-5 Gy at the 0.5-cm depth, and decreases rapidly to 1.52×10-5 Gy at the 3-cm position, and mainly depends on the boron and thermal neutron doses caused by the 10B(n, α)7Li and 14N(n, p)14C reactions, respectively.

Verification of a real-time interactive transient simulator for Dalat Nuclear Research Reactor

A PC-based real-time interactive transient simulator of Dalat Nuclear Research Reactor (DNRR), namely DalatSim, based on the best-estimate thermal-hydraulic code RELAP5/MOD3.3 has been currently building at Center for Nuclear Technologies (CNT). This paper presents the study on developing the physics core, control module, and human-machine interface (HMI) of DalatSim. The nodalization of DNRR used for DalatSim was based on the reported numerical model in the Safety Analysis Report (SAR) in 2012. DalatSim can simulate operational procedures and several hypothetical transient accidents of DNRR. A curve of real operational power of DNRR was used to compare with calculation power results from DalatSim to verify its capability. The verification results are presented and discussed.

Design of a FPGA-based controller for power and period measurement in the start range of Dalat Nuclear Research Reactor

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.

Human factors engineering and participatory design in the project of a nuclear research reactor control centre

The Human Factors Engineering (HFE) program is an essential aspect for the design of nuclear installations. The overall aim of the HFE program is the improvement of the operational reliability and safety of plant operation. The HFE program main purpose is to ensure that human factor practices are incorporated into the plant design, emphasizing man-machine interface issues and design improvement of the nuclear reactor Control Centre. The Control Centre of nuclear reactor is a combination of control rooms, control suites and local control stations, which are functionally connected and located on the reactor site. The objective of this paper is to present a participatory design approach, including human factor issues, for the Control Centre of a nuclear research reactor used to produce radioisotopes and for nuclear research. The design approach is based on participatory design principles, using human factor standards, ergonomic guidelines, and the participation of a multidisciplinary team during all design phases. Using the information gathered, an initial sketch 3D of the Control Centre was developed.

Using Fission-Track Radiography Coupled with Scanning Electron Microscopy for Efficient Identification of Solid-Phase Uranium Mineralogy at a Former Uranium Pilot Mill (Grand Junction, Colorado)

At a former uranium pilot mill in Grand Junction, Colorado, mine tailings and some subpile sediments were excavated to various depths to meet surface radiological standards, but residual solid-phase uranium below these excavation depths still occurs at concentrations above background. The combination of fission-track radiography and scanning electron microscope energy-dispersive X-ray spectroscopy (SEM-EDS) provides a uniquely efficient and quantitative way of determining mineralogic associations of uranium that can influence uranium mobility. After the creation of sample thin sections, a mica sheet is placed on those thin sections and irradiated in a nuclear research reactor. Decay of the irradiated uranium creates fission tracks that can be viewed with a microscope. The fission-track radiography images indicate thin section sample areas with elevated uranium that are focus areas for SEM-EDS work. EDS spectra provide quantitative elemental data that indicate the mineralogy of individual grains or grain coatings associated with the fission-track identification of elevated uranium. For the site in this study, the results indicated that uranium occurred (1) with coatings of aluminum–silicon (Al/Si) gel and gypsum, (2) dispersed in the unsaturated zone associated with evaporite-type salts, and (3) sorbed onto organic carbon. The Al/Si gel likely formed when low-pH waters were precipitated during calcite buffering, which in turn retained or precipitated trace amounts of Fe, As, U, V, Ca, and S. Understanding these mechanisms can help guide future laboratory and field-scale efforts in determining long-term uranium release rates to groundwater.