Dalat Nuclear Research Reactor Research Articles

Self-characterization of irradiation facility using synthetic multi-element standard for determination of k0-factors of seven radionuclides of interest

In this study, the k0 factors of 7 short and medium-lived radionuclides (i.e. 66Cu, 52V, 38Cl, 134mCs, 128I, 140La and 56Mn) were experimentally determined at the Dalat nuclear research reactor (DNRR). The synthetic multi-element standard (SMELS) samples were irradiated at Channel 7-1 installed at the DNRR. SMELS Type III contains Au and Zr, thus also allowing the self-characterization of the irradiation facility (i.e. the direct determination of f and a with only this material). The k0 factors of 7 short and medium-lived radionuclides obtained in this work are a good agreement with the reference k0 factors. The ratio values between the experimental and reference k0 values were 66Cu of 1.004, 52V of 1.004, 38Cl of 0.961 and 0.976 (1642.7 keV and 2167.4keV, respectively), 134mCs of 0.831, 128I of 1.022, 140La of 0.936, 0.955 and 0.955 (328.8 keV, 487.0 keV and 1596.2keV, respectively), 56Mn of 0.985 and 0.973 (846.8 keV and 1810.7keV, respectively). The radionuclides range of uncertainty values were lower than 7%. The present result indicates that the k0 factors can be determined by using synthetic multi-element standards (SMELS).

Improved Genetic algorithm for fuel loading optimization of the DNRR with HEU fuel

This paper investigates the performance of genetic algorithm (GA) with improved selection techniques, i.e. Tournament and Roulette Wheel, applied to in-core fuel management of the Dalat nuclear research reactor (DNRR). Numerical calculations have been performed based on the DNRR core with 100 HEU fuel bundles. The optimal fitness function was chosen to maximize the keff and minimize the power peaking factor. The statistical analysis using Mann-Whitney test shows that the performance of GA with Tournament selection is advantageous over the Roulette Wheel selection in the ICFM problem of the DNRR. The optimal core configurations obtained with the improved GA methods have the keff values greater by about 500 pcm, and the PPF lower by about 4.0% compared to the reference core.

A Digital Controller for Reactivity Monitoring and Power Control

This paper introduces a controller unit for reactivity monitoring and automatic power control that was designed and constructed for the 500 kW Dalat Nuclear Research Reactor (DNRR). For power control and reactivity calculations, frequency signals from neutron measurement channels of starting and working ranges of the reactor are used. Two abovementioned independent functions were combined in an Artix-7 FPGA board for determining reactivity values by solving the point reactor kinetics equations with six delayed neutron groups and for stabilizing the reactor power at preset levels by determining the unbalance voltage signal to control the automatic control rod. With real-time calculations, the newly developed controller can monitor the reactor reactivity and control the reactor power online. The developed controller unit’s reactivity measuring and power stabilizing capabilities were assessed using the DNRR in normal operation and assumed emergency conditions and compared with those of the preexisting imported BNO-102R1 module of the DNRR control and protection system, known as ASUZ-14R. The results of the experiments show that the produced FPGA-based unit and the BNO-102R1 unit have the same technical characteristics and features, with the disparities being less than 5% and 1%, respectively, in reactivity measurement and power stabilization. The experimental data of reactivity measurements by the FPGA-based unit and the calculation results were also compared and found that the relative deviations between those are also less than 10%. The developed controller unit is capable of carrying out a variety of training and operational experiments on the DNRR.

Investigation of prompt [formula omitted]-ray neutron activation spectrometer at the Dalat research reactor using Geant4 simulation

The prompt γ-ray activation analysis (PGAA) detector, which has been developed at the Dalat nuclear research reactor, is investigated in this study. The Geant4 model is proposed for the PGAA detector to evaluate the Compton scattering background suppression. Based on this model, the detector efficiency in the energy range from 0.1 MeV to 11 MeV was determined using the thermal neutron capture of chlorine, titanium, nitrogen, and reference sources. The detector efficiencies in the unsuppressed and suppressed modes and the difference of efficiencies in both modes were calculated. Additionally, the suppression of additional peaks of the PGAA detector was also estimated in this work. Finally, the polynomial function for the detector efficiency and the ratio of additional peaks to corresponding full-energy peaks were determined. This work makes an essential contribution to the analysis of prompt γ-ray spectrum acquired from the PGAA detector at the Dalat nuclear research reactor.

Design and simulation of a thermal neutron beam for neutron capture studies at the Dalat research reactor

This paper presents the application of the MCNP5 code to conduct a modification design and simulation of a thermal neutron beam for neutron capture studies at the Dalat Nuclear Research Reactor (DNRR). The designed configuration of the horizontal neutron channel No.2 at the DNRR, which contains a conical collimator of 240.3 cm in length, and neutron filters of crystal Al2O3 and Bi with alternative thickness, were simulated. A pure thermal neutron beam can be obtained at the irradiation position when a composition of crystal filter of 20 cm Al2O3 and 6 cm Bi is employed. The thermal and epithermal neutron fluxes are 1.02´108 n/cm2/s (account for 97.92% of total neutron flux) and 0.22´107 n/cm2/s (account for 2.08% of total neutron flux), respectively.

Design of a FPGA-based digital reactivity meter for Dalat Nuclear Research Reactor

This paper describes a digital reactivity meter module based on solving the inverse kinetics equations and the time behavior of the output pulse frequency from the amplifier, which is proportional to the power of the Dalat Nuclear Research Reactor (DNRR). The designed reactivity meter, which operates online with real-time calculation, was tested using the PGT-17R simulator signals in comparison with the existing imported module PNO-121R5 of the DNRR’s control and protection system. The experimental results show that the technical characteristics and functions of the two modules are equivalent.

Determination of the Activity Inventory in the Structural Components of the Dalat Nuclear Research Reactor for Its Decommissioning Planning

This report presents the methods and calculated results of the activity inventory in the structural components of the Dalat Nuclear Research Reactor (DNRR). These components include the shielding concrete, the reactor tank, and its inside irradiated facilities; the thermal and thermalizing columns; and the horizontal channels. The MCNP5 code with a three-dimensional neutron transport model was used to calculate the neutron flux distribution, neutron energy spectrum at different locations, and activation cross sections of long-lived radioactive nuclides in activated major materials, including heavy concrete, reflection graphite, and aluminum of the reactor. The calculated results of the energy spectrum and activation cross sections of MCNP5 were used in the ORIGEN2.1 point depletion code to calculate the neutron-induced activity of activated materials at different time points by modeling the irradiation history and radioactive decay. Radioactivity of long-lived key activation products such as 41Ca, 60Co, 55Fe, 63Ni, and 152Eu was modeled, and volumes of radioactive waste mainly of ordinary concrete, graphite, and aluminum in the structural components of the reactor were estimated. Experimental results of neutron flux and specific activities of some typical nuclides such as 60Co, 152Eu, 55Fe, and 63Ni in activated aluminum samples showed good agreement with the calculated results. As part of the national regulation requirements, the obtained data have been used to develop the decommissioning plan for the operational DNRR, with an estimation of about 10 years before its permanent shutdown.

Sensitivity and Uncertainty Analysis on Reactivity for HEU and LEU Fuel Assemblies of Dalat Nuclear Research Reactor using Monte Carlo Code and ENDF/B-VII.0 and ENDF/B-VII.1 Nuclear Libraries

This paper presents the analysis of sensitivity and uncertainty for the infinite multiplication factor (kinf) for the VVR-M2 typed HEU and LEU fuel assemblies of the Dalat nuclear research reactor (DNRR) using the MCNP6.1-Whisper1.1 code. Sensitivity calculations were performed for the ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries. In Whisper, the keff uncertainty due to nuclear data was evaluated by the uncertainty propagation law using the sensitivities obtained by MCNP and the available covariance matrix. The most significant sensitivity coefficients in positive contribution are the coefficient total n and fission reaction of U-235, elastic scattering reaction of H-1 and inelastic scattering of thermal neutrons and in negative contribution are capture reactions of U-235, H-1, Al-27, U-238 and U-234 isotopes. The large discrepancy between sensitivities of elastic scattering reaction cross section of H-1 and inelastic scattering of thermal neutrons between two libraries are found because of the change in neutron spectra of HEU and LEU fuel assemblies using the two library versions. The uncertainty of the kinf from the ENDF/B-VII.0 nuclear library error for all isotopes was found to be significant (about 0.45% of reactivity effect) with the largest contribution from isotope U-235 contributions (about 0.3% to 0.40%).

Two-step gamma cascade decays from the compound state to the ground state of 182Ta nucleus

Two-step gamma transitions corresponding to the decays from the compound state (6,062.93 keV) to the ground state of 182Ta nucleus have been investigated via a 181Ta(n,γ)182Ta reaction using the thermal neutron beam from the No.3 neutron channel of the Dalat Nuclear Research Reactor and a gamma-gamma coincidence spectrometer. By comparing with data extracted from the Evaluated Nuclear Structure Data File (ENSDF) library, the authors have determined the order of 20 pairs of gamma cascade transitions, and then constructed a corresponding partial nuclear level scheme of 182Ta. Based on the detected gamma transitions and levels, 17 primary transitions, 1 secondary transition, and 20 intermediate levels are found to be the same as those in the ENSDF. The remaining 3 primary transitions and 19 secondary transitions are, thus, considered as the new data. In addition, a tentative spin range of (2, 3, 4)ℏ. has been assigned to all the intermediate levels detected within the present work.

Evaluation of neutron flux distribution in structural components and activation products in aluminum alloy at 13-2 channel of the DNRR

The estimation of radiological properties of activated structural components of a nuclear reactor due to irradiation of neutron produced by fission is a very important task for radiation safety and reasonable cost of dismantling and radioactive waste management in the decommissioning plan of the reactor. In this work, the calculation approach was carried out by using three-dimensional neutron transport model with the Monte Carlo code MCNP5 to evaluate neutron fluxes and reaction rates. The Bateman equation was solved with neutron absorption reactions (fission and capture) and disintegration by ORIGEN2 code to obtain the activity of materials in reactor structures. This paper presents the evaluation results of the neutron flux distribution and the radioactivity of long-lived key activation products such as 60Co, 55Fe, 59Ni, 63Ni, etc. isotopes in the structural components of the Dalat Nuclear Research Reactor (DNRR). The validation of calculation methodology of the two codes was implemented by comparing calculation results with measured neutron fluxes at irradiation positions in the reactor core as well as specific activities at the bottom part of the aluminum guiding tube at 13-2 channel, which has been removed from the reactor core about six years. The calculation results were in good agreement under 7% difference with the experimental neutron flux value of (6.05±0.52) × 1012 n/cm2.s, and under 33% difference with the experimental specific activities of 60Co isotope being 1.86×104, 9.99×104, and 1.28×105 Bq/g at the positions of -32.5, -17.5 and -2.1 cm (the centerline of the reactor core is at 0 cm), respectively, in the aluminum guiding tube of irradiation channel 13-2. The neutron flux distributions in other structural components such as the graphite reflector, thermal column, thermalizing column, concrete shielding, etc. of the reactor were also evaluated. The obtained calculation results and experimental data are very valuable for the development of a suitable decommissioning plan and a reasonable dismantling strategy for the DNRR.

Design of a neutron flux measurement channel using the ionization chamber KNK-3 at the Dalat Nuclear Research Reactor

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.

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.

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.

A discrete differential evolution for fuel loading optimization of the DNRR research reactor and comparison with genetic algorithm

A discrete differential evolution (DE) method has been applied to the problem of fuel loading pattern optimization of the Dalat Nuclear Research Reactor (DNRR). A classic strategy DE/rand/l/bin was chosen for the mutation of the DE method. Numerical calculations have been performed based on the core configuration of 100 highly enriched uranium (HEU) fuel bundles with various burnup levels. Comparison of the performance between the DE method and a genetic algorithm (GA) was also carried out. The optimal LPs obtained from the two methods are significantly better than the reference core. DE is more advantageous in exploring search space and approaching a global optimum than GA.

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Development of a PARCS/Serpent model for neutronics analysis of the Dalat nuclear research reactor

Cross-sectional homogenization for full-core calculations of small and complex reactor configurations, such as research reactors, has been recently recognized as an interesting and challenging topic. This paper presents the development of a PARCS/Serpent model for the neutronics analysis of a research reactor type TRIGA Mark-II loaded with Russian VVR-M2 fuel (known as the Dalat Nuclear Research Reactor or DNRR). The full-scale DNRR model and a supercell model for a shim/safety rod and its surrounding fuel bundles with the Monte Carlo code Serpent 2 were proposed to generate homogenized few-group cross sections for full-core diffusion calculations with PARCS. The full-scale DNRR model with Serpent 2 was also utilized as a reference to verify the PARCS/Serpent calculations. Comparison of the effective neutron multiplication factors, radial and axial core power distributions, and control rod worths showed a generally good agreement between PARCS and Serpent 2. In addition, the discrepancies between the PARCS and Serpent 2 results are also discussed. Consequently, the results indicate the applicability of the PARCS/Serpent model for further steady state and transient analyses of the DNRR.

Updated nuclear level density parameters of 153Sm nucleus within the back-shifted Fermi gas model

The present work re-evaluates the level density parameter a, asymptotic level density parameter aasy, and back-shifted energy parameter E1 within the back-shifted Fermi gas model (BSFG) for the 153Sm nucleus. This reevaluation is based on the experimental nuclear level scheme extracted from the ENSDF library, the average level spacing at the neutron binding energy (D0 value), and the latest updated nuclear level scheme obtained from an experimental gamma cascade experiment, which was performed at the Dalat Nuclear Research Reactor using the thermal neutron beam. The updated values of the BSFG level parameters are: (1) a=18.09±0.25 MeV-1and E1=-0.92±0.07 MeV for the energy-independent level density parameter; and (2) aasy=15.00±0.20 MeV-1and E1=-0.81±0.08 MeV for the energy-dependent level density parameter. It has been found that the total nuclear level densities calculated using these updated parameters agree with the experimental data better than those using parameters taken from the nuclear reference database RIPL-3. These updated parameters are more accurate and reliable than those extracted from RIPL-3 and are, therefore, highly recommended for all the applications hereafter.

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.

Interim Storage of the Dalat Nuclear Research Reactor: Radiation Safety Analysis

Radiation safety analysis of a new interim storage of the Dalat Nuclear Research Reactor (DNRR) for keeping spent high enriched uranium (HEU) fuel bundles during the core conversion to low enriched uranium (LEU) fuel had been performed and presented. The photon source and decay heat of the spent HEU fuel bundles were calculated using the ORIGEN2.1 code. Gamma dose rates of the spent fuel interim storage were evaluated using the MCNP5 code with various scenarios of water levels in the reactor tank and cooling time. The radiation safety analysis shows that the retention of 106 spent HEU fuel bundles at the interim storage together with a core of 92 LEU fuel bundles meets the requirements of radiation safety. The results indicate that in the most severe case, i.e., the complete loss of water in the reactor tank, the operators still can access the reactor hall to mitigate the accident within a limited time. Particularly, in the control room, the dose rate of about 1.4 μSv/his small enough for people to work normally.