Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.     

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C,JENDL-3.3 and ENDF/B-Vl data libraries

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for ^natZr, ^natMo, ^natCr, ^natFe, ^natNi, ^natSi, and ^natMg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ²⁸, δ²⁵, ρ²⁵, and C¹ were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Experimental study of neutronic parameters in Tehran research reactor mixed-core

In this work, general characteristics of a typical mixed core, including HEU & LEU fuel is studied. The study is performed in the Tehran research reactor (TRR). In this study the neutronic parameters, reactivity feedback coefficients and kinetic parameters are investigated. The reference core designated for such study is the equilibrium core (No. 61) with an average bun-up of 27% & 36% for SFE's & CFE's, respectively. The MTR_PC package is used for neutronic analysis. In this research, experimental and computational results for the reference and mixed core are compared. Meantime, the obtained values for neutronic parameters are mostly below the adopted safety criteria and they are in good agreement with the experimental results. However β_eff and ℓ_p are a little bit higher in the mixed core with respect to the reference core, but in practice, these small changes will not cause substantial impacts on the dynamic behaviour of the reactor core. The absolute values of the fuel temperature, moderator density and void coefficients of reactivity, are less in the mixed core and only the moderator temperature coefficient is higher. The calculated values of power defect, based on the reactivity coefficients; in both core configurations are in good agreement with the experimental values.     

Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.     

Validation of CENDL and JEFF evaluated nuclear data files for TRIGA calculations through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors

The aim of this paper is to present the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through the analysis of the integral parameters of TRX and BAPL benchmark lattices of thermal reactors for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. In this process, the 69-group cross-section library for lattice code WIMS was generated using the basic evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 with the help of nuclear data processing code NJOY99.0. Integral measurements on the thermal reactor lattices TRX-1, TRX-2, BAPL-UO₂-1, BAPL-UO₂-2 and BAPL-UO₂-3 served as standard benchmarks for testing nuclear data files and have also been selected for this analysis. The integral parameters of the said lattices were calculated using the lattice transport code WIMSD-5B based on the generated 69-group cross-section library. The calculated integral parameters were compared to the measured values as well as the results of Monte Carlo Code MCNP. It was found that in most cases, the values of integral parameters show a good agreement with the experiment and MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation of evaluated nuclear data files CENDL-2.2 and JEFF-3.1.1 through benchmarking the integral parameters of TRX and BAPL lattices and can also be essential to implement further neutronic analysis of TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh.     

Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.     

Uncertainty quantification of neutronic parameters of light water reactor fuel cells with JENDL-4.0 covariance data

Neutronic parameter uncertainty induced by nuclear data uncertainty is quantified for several light water reactor fuel cells composed of different combinations of fissile/fertile nuclides. The covariance data given in JENDL-4.0 are used as the nuclear data uncertainty, and uncertainty propagation calculations are carried out using sensitivity coefficients calculated with the generalized perturbation theory for burnup-related neutronic parameters.

It is found that main contributors of nuclear data uncertainty to the neutronic parameter uncertainty are the uranium-238 capture cross section in a uranium-oxide fuel cell, and the plutonium-240 and plutonium-241 capture cross sections and fission spectrum of fissile plutonium isotopes in a uranium–plutonium mixed-oxide fuel cell. It is also found that thorium-232 capture cross section uncertainty is a dominant source of neutronic parameter uncertainty in thorium–uranium and thorium–plutonium mixed-oxide fuel cells. It should be emphasized that precise and detail information of component-wise uncertainties can be obtained by virtue of the adjoint-based sensitivity calculation methodology. Furthermore, cross-correlations are evaluated for each fuel cell, and strong correlations among the same parameters at the beginning of cycle and at the end of cycle and among different parameters are observed.     

Investigation of thermohydraulic parameters during natural convection cooling of TRIGA reactor

Important steady-state thermohydraulic parameters of the TRIGA research reactor operating under natural convection mode of coolant flow were investigated using NCTRIGA computer code. Neutronic parameters used in preparing the input of NCTRIGA were taken from the analysis performed by 3-D Monte Carlo code MCNP4C. Benchmarking of the NCTRIGA calculated results were performed against the experimental data measured by the thermocouples in the instrumented fuel element (IFE) during the steady state operation of the reactor under natural convection mode of coolant flow. Various thermohydraulic parameters like the coolant velocity, flow rate and mass flow rate were generated for the hot channel as well as for the two channels comprising instrumented fuels. Calculated peak fuel temperatures at different power levels were compared with the measured values and also with the calculations performed by PARET code. Axial temperature profile at the fuel centreline, fuel surface and coolant in the hot channel were generated. Fuel surface heat flux, heat transfer coefficient and Reynolds’s number for the hot channel were also calculated. The effect of fuel-cladding gap and the influence of fuel rod spacing were investigated to validate the performance of NCTRIGA code. The investigated results were found to be in good agreement with the experimental values, which indicates that the NCTRIGA code can be used with confidence for TRIGA reactor analysis.     

Wavelet approach for analysis of neutronic power using data of Ringhals stability benchmark

We have studied neutronic power oscillation in a boiling water nuclear reactor for three different scenarios of the Ringhals stability benchmark with a proposed wavelets-based method: the first scenario is a stable operating state which was considered as a base case in this study, and the last two correspond to unstable operating conditions of in-phase and out-of-phase events. The results obtained with the methodology presented here suggest that a wavelet-based method can help the understanding and monitoring of the power dynamics in boiling water nuclear reactors. The stability parameters frequency and decay ratio were calculated as a function of time, based on the theory of wavelet ridges. This method allows us to analyze both stationary and highly non-stationary signals. The resonant frequencies of the oscillation are consistent with previous measurements or calculated values.     

Evaluating the core damage frequency of a TRIGA research reactor using risk assessment tool software

After all preventive and mitigative measures considered in the design of a nuclear reactor, the installation still represents a residual risk to the outside world. Probabilistic safety assessment (PSA) is a powerful method to survey the safety of nuclear reactors. In this study the occurrence frequency of different types of core damage states (CDS) which may potentially arise in Tehran Research Reactor (TRR) is evaluated by use of the recently developed risk assessment tool (RAT) software which has been designed and represented in the Safety Research Center of Shiraz University. RAT uses event trees and fault trees to evaluate the total final core damage frequency (CDF) through studying the frequencies of initiation events, and following their consequences has resulted in one type of the CDS. The criterion must be of the order of smaller than 1E−04 through IAEA standards for research reactors (IAEA-TECDOC-400, 1986). Results show that the total final CDF for TRR is of the order of 10⁻⁶, which meets the criterion of nuclear research reactor.     

Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.     

Steady-state neutronic analysis of converting the UK CONSORT reactor for ADS experiments

CONSORT is the UK’s last remaining civilian research reactor, and its present core is soon to be removed. This study examines the feasibility of re-using the reactor facility for accelerator-driven systems research by replacing the fuel and installing a spallation neutron target driven by an external proton accelerator. MCNP5/MCNPX were used to model alternative, high-density fuels and their coupling to the neutrons generated by 230 MeV protons from a cyclotron striking a solid tungsten spallation target side-on to the core. Low-enriched U₃Si₂ and U–9Mo were considered as candidates, with only U–9Mo found to be feasible in the compact core; fuel element size and arrangement were kept the same as the original core layout to minimise thermal hydraulic and other changes. Reactor thermal power up to 2.5 kW is predicted for a k_eff of 0.995, large enough to carry out reactor kinetic experiments.     

Contributions to the neutronic analysis of a gas-cooled fast reactor

In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained results are very similar.     

Sensitivity analysis for AHWR fuel cluster parameters using different WIMS libraries

India is presently engaged in the design of an advanced heavy water reactor (AHWR) which utilises thorium as fuel. The AHWR is a boiling light water cooled heavy water moderated reactor where the heat is removed through natural convection. Dysprosium is used as burnable absorber to get a reduction in void reactivity. The design needs to be well validated. The 69 group old WIMS library distributed by NEA in 1980′s is presently being used for the design and analysis of AHWR. We have now undertaken an exercise to study the sensitivity of the design parameters, such as k-infinity and void reactivity with respect to the various datasets which have been made available as part of the IAEA CRP on the final stage of the WIMS library update project (WLUP). The k-infinity variations are within 1% both at the beginning of cycle (BOC) and at the end of cycle (EOC). The results for the coolant void reactivity, however, show significant differences between the different datasets at BOC itself which increases further with burnup. In comparison, the differences for natural uranium fuelled pressurised heavy water reactor (PHWR) lattice are relatively lower. Major source of variations in AHWR lattice are probably coming from Th-²³³U data.     

Stability analysis in nuclear reactor using Lyapunov exponent

A simple model for nuclear reactor is proposed. With increasing the fuel concentration, our minimal model shows two successive phases; subcritical and supercritical. In subcritical regime, the neutron population grows with increasing the fuel concentration. In the supercritical state, the Lyapunov exponent is positive implying that the neutron diffusion phenomena are spatiotemporal chaos. In the present paper, the infinite multiplication factor curve is qualitatively reproduced for fuel concentration. We have derived the critical fuel concentration based on the Lyapunov exponents. The basic objective of the work is to improve the prediction of the critical neutron population with respect to the fuel concentration.     

Neutronics analysis of the initial core of the TRIGA Mark II reactor

The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and −60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections.For the first experiment, the model predicts an effective multiplication factor (κ_eff) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.