Nuclear Science and Technology

From evaluated data sources like ENDF, JENDL and JEFF, neutronics data libraries forMCNP computer code have been produced, including neutron scattering cross section library S (β,α) in thermal energy range, by using NJOY computer code. The evaluation and validation of these neutronics data libraries have been carried out through calculation of some parameters such as effective multiplication factor and reaction cross sections of benchmark problems, VVR-M2 fuel type as well as the critical configurations of the Dalat Nuclear Research Reactor loaded with low enriched Uranium fuel. After implementing about analysis and evaluation of the calculated results with abovementioned libraries, the library provides results that consistent with experimental data can be used in core and fuel management calculation for the Dalat Nuclear Research Reactor (DNRR).

This paper presents the conceptual design of a 300 MWt small modular reactor (SMR)using fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system and the JENDL-4.0 data library. The analysis showed that Doppler, moderator temperature, void, and power reactivity coefficients are all negative over the core lifetime. Semi-analytical thermal hydraulics analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel andclad surface temperature. The minimum departure from nucleate boiling ratio (MDNBR) is also calculated. The results indicate that a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

In this work the non - destructive gamma spectroscopic method for determination of uranium enrichment is presented. In order to determine the uranium enrichment the activity ratios of 234U/235U and 238U/235Uwere measured. The activity ratios234U/235U and 238U/235U were determined by using intrinsic efficiency calibration. As a test of this method, low - enriched uranium standard was measured, the obtained result was in good agreement with the estimated value.

This paper presents a design of the neutron flux measurement channel that consists of a Boron-contained gamma-compensated ionization chamber (CIC) named KNK-3 and operates in current mode, a current to frequency (I to F) converter, and a neutron flux measurement and control module (FPGA-WR). The designed measuring channel allows to measure and control the neutron flux density from 1.0x106 to 1.2x1010 n/cm2.s corresponding to the range from 0.1 to 120% of the nominal power of 500 kW of the Dalat nuclear research reactor (DNRR). The measurement and control module uses FPGA Artix-7 and digital signal processing algorithms to measure and calculate the reactor power and period values and generate warning and emergency signals by the reactor power and period. The measurement channel was tested by using simulated signals and examining in the reactor to compare with the neutron flux measurement channel using the BPM-107R neutron flux controller of the existing complex ASUZ-14R for the DNRR control and protection system (CPS). The comparison results show that the measurement channel fully meets the requirements on the accuracy of the reactor power and period parameters as well as the ability to respond at once to the warning and emergency signals of the reactor power and period. Therefore, the measurement channel can be used for testing, research, and training. The FPGA-WR measurement and control module can replace the BPM-107R controller for the working range of the CPS.

Gamma-irradiation effects on graphene oxide (GO) in a monoglyceride/ethanol (MG/EtOH) solution was investigated.  GO was dispersed in MG/EtOH solution (GOM) with the GO: MG ratio of 1:10 (w/w). The prepared GOM was irradiated by g-ray under nitrogen atmosphere in a range of absorbed dose from 0 to 50 kGy. The characteristics and morphology of reduced GOM were analyzed by Ultraviolet-visible (UV-vis) spectroscopy, Fourier transform infrared (FT-IR) spectroscopy, X-ray diffraction (XRD) and Transmission Electron Microscopy (TEM). The results confirmed that the structure of reduced graphene oxide in monoglyceride solution was changed and exfoliated completely after g-ray irradiation at absorbed dose 50 kGy compared with non-irradiation sample.

The Dalat Nuclear Research Reactor (DNRR) with the nominal power of 500 kW was reconstructed and upgraded from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The renovated reactor was put into operation on 20th March 1984. It was designed for the purposes of radioisotope production (RI), neutron activation analysis (NAA), basic and applied researches, and nuclear education and training. During the last 30 years of operation, the DNRR was efficiently utilized for producing many kinds of radioisotopes and radiopharmaceuticals used in nuclear medicine centers and other users in industry, agriculture, hydrology and scientific research; developing a combination of nuclear analysis techniques (INAA, RNAA, PGNAA) and physic-chemical methods for quantitative analysis of about 70 elements and constituents in various samples; carrying out experiments on the reactor horizontal beam tubes for nuclear data measurement, neutron radiography and nuclear structure study; and establishing nuclear training and education programs for human resource development. This paper presents the results of operation and utilization of the DNRR. In addition, some main reactor renovation projects carried out during the last 10 years are also mentioned in the paper.

Alternating Current Field Measurement (ACFM) is a technique of the Electromagnetic method used to detect surface defects of metal materials. Currently, this technique is widely applied in the field of maintenance of Oil and Gas projects as an alternative to the Magnetic Particle Testing method. The establishment of ACFM training program according to Recommended Practice No. SNT-TC-1A of The American Society for Nondestructive Testing (ASNT) will increase the autonomy of the domestic testing human resources, especially advanced techniques. Based on documents and standards combined with the actual survey, training programs, training materials, question banks, examinations developed meet the requirements of international standards and in accordance with the conditions applied in Vietnam

The IAEA TRS483 and TRS398 Code of Practices (CoP) were used to calculate relative output factors for small photon beams of 6X, 6XFFF energies shaped by High Definition Multileaf Collimator (HDMLC), jaws and cones mounted on TrueBeam STx medical linear accelerator (Varian Medical Systems), respectively. A comparison between these results were made. The results show a large discrepancy in relative output factor curves found among different collimation systems of the same equivalent field sizes and between the CoPs. Therefore, the specific beam modelling in treatment planning system for each type of the collimation system to be used for small fields maybe required for better computational accuracy.

The paper presents calculated results of neutronics, steady state thermal hydraulics and transient/accidents analyses for full core conversion from High Enriched Uranium (HEU) to Low Enriched Uranium (LEU) of the Dalat Nuclear Research Reactor (DNRR). In this work, the characteristics of working core using 92 LEU fuel assemblies and 12 beryllium rods were investigated by using many computer codes including MCNP, REBUS, VARI3D for neutronics, PLTEMP3.8 for steady state thermal hydraulics, RELAP/MOD3.2 for transient analyses and ORIGEN, MACCS2 for maximum  hypothetical accident (MHA). Moreover, in neutronics calculation, neutron flux, power distribution, peaking factor, burn up distribution, feedback reactivity coefficients and kinetics parameters of the working core were calculated. In addition, cladding temperature, coolant temperature and ONB margin were estimated in steady state thermal hydraulics investigation. The working core was also analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and MHA. Obtained results show that DNRR loaded with LEU fuel has all safety features as HEU and mixed HEU-LEU fuel cores and meets requirements in utilization as well.

Embrittlement is one of the most important effects affecting reactor pressure vessel (RPV) aging. RPV is irradiated with neutrons and gammas, especially fast neutrons, which mainly lead to embrittlement of RPV during operation lifetime of nuclear reactors. Therefore, the radiation-induced embrittlement of the RPV should be carefully evaluated. In this paper, a preliminary calculation was performed using the MCNP5 code to identify the areas in the RPV of the VVER-1000/V320 reactor where the neutron and gamma fluxes are maximum. Also, the neutron and gamma fluence distributions on the RPV were investigated and evaluated along with their energy spectra. These calculations are the starting point for the evaluation of radiation damage to RPV of VVER reactors.