This paper focuses on the development of a new computational model of the CNESTEN’s TRIGA Mark II researchreactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was ...developed toassess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor,control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used toaccurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivityand uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were usedto build the T4 model. In this regard, the differences between the two models are mainly due to mathematicalapproaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part dealswith the development and the validation of the T4 model. The results obtained with the T4 model were comparedto the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report(FSAR). Different core configurations were investigated via simulations to test the computational model reliabilityin predicting the physical parameters of the reactor. As a fairly good agreement among the results wasdeduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values.The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out toquantify the nuclear data uncertainty in the multiplication factor keff . For that purpose, the T4 model was used tocalculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to theresults obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools anddifferences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities.Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matriceslibrary and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff isaround 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopesto the nuclear data uncertainty in the keff is not negligible and should be taken into account when performingS/U analysis.
This paper presents a comparative analysis when transitioning from fuel meat based on uranium zirconium hydride (U-ZrH1.65) to one based on thorium-uranium-233 zirconium hydride ((Th-233U)-ZrH1.65) ...as an alternative option for TRIGA reactors. The aim of this study was to investigate the neutronic characteristics of (Th-233U)-ZrH1.65 fuels with two different Th-233U weight contents (i.e., 8.5 wt% and 12 wt%) and various U-233 enrichment levels. Using MCNP6.2 computer code based on the Monte Carlo method, a three-dimensional (3D) hypothetical TRIGA model was developed to investigate various neutronic parameters, such as the effects of U-233 enrichment and different (Th-233U) weight contents on the effective multiplication factor, burnup behavior, spent fuel composition, neutron flux, and fission product poison concentration. Since fuel temperature coefficients of reactivity and effective delayed neutron fractions play important roles in the reactor’s kinetics, these were also investigated, and the results were compared to the model with U-ZrH1.65 fuel. The obtained results revealed that the presence of Th-232 in the fuels did not drastically reduce the fuel temperature coefficients due to the zirconium hydride within the fuels providing most of the feedback in a TRIGA reactor. In contrast, the presence of U-233 in the (Th-233U)-ZrH1.65 fuels did reduce the effective delayed neutron fraction. Despite this, the lower transuranic generation in cores fueled by (Th-233U)-ZrH1.65 make this fuel advantageous in terms of achieving longer cycle lengths and less-hazardous nuclear waste for disposal.
Julius Federico M. Jecong1,3,4*, Alvie A. Astronomo2, Frederick C. Hila1, Neil Raymund D. Guillermo1, and Sweng Woong Woo4 1Applied Physics Research Section; 2Nuclear Reactor Operations Section ...Department of Science and Technology–Philippine Nuclear Research Institute Quezon City, Metro Manila 1101 Philippines 3Nuclear and Quantum Engineering Department Korea Advanced Institute of Science and Technology 291 Daehak-ro, Yuseong-gu, Daejeon 34141 Republic of Korea 4Reactor and Safety Evaluation Department Korea Institute of Nuclear Safety 62 Gwahak-ro, Yuseong-gu, Daejeon 34142, Republic of Korea Safety Analysis of 250-kW Philippine Research Reactor-1 Thermal-hydraulics under Steady-state Operations Using MARS-KS Code Keywords: 250-kW TRIGA reactor, MARS-KS, safety analysis, steady-state, thermal-hydraulics The Philippine Nuclear Research Institute (PNRI) of the Department of Science and Technology (DOST) is implementing a project to use the Philippine Research Reactor-1 (PRR-1) TRIGA nuclear fuel in a subcritical reactor (SCR) for training, education, and research in nuclear science and technology. However, although an SCR is a valuable training and research facility, it offers limited industrial applications. To potentially reap more benefits, DOST-PNRI is also investigating the feasibility of converting the zero-power SCR to a low power critical reactor in the future. In this work, the thermal-hydraulic behavior of a hypothetical 250-kW low power configuration of the PRR-1 TRIGA reactor is investigated using the MARS-KS code. At this power level, the fuel centerline or the hottest region of the fuel was determined to have 236.85- °C temperature during steady-state conditions. This temperature is well below the 749.85-°C temperature limit of TRIGA fuel that will result in excessive swelling. The peak heat flux occurs at the axial center of the fuel rod while the calculated departure from nucleate boiling ratio (DNBR) of 6.18 is minimum at this location. Moreover, the maximum fuel temperature is found to be insensitive to variations in coolant velocity, which ranges from 0.20–0.26 m/s within the natural circulation conditions of this facility. Based on the Bernath critical heat flux (CHF) correlation, coolant velocities below 0.2 m/s will still yield a DNBR above 5.0. These results demonstrate that a 250-kW PRR-1 TRIGA reactor operating under steady-state conditions is capable of safely removing the heat generated in the fuel
Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate neutron and prompt gamma particles emitted from TRIGA research ...reactor during operation. Firstly, the method was validated by measuring dose rates around open beam port number 5 was unplugged. Neutron and gamma dose rates inside the reactor hall in the vicinity of the beam port were calculated and compared to the measurements. Due to the satisfactory agreement, the method was later used to design external shielding for the same beam port when it was upgraded – special mechanism was installed that allows irradiation of larger samples. Computational analysis of the proposed shielding configuration provided acceptable dose rate levels inside the reactor hall. When the shield was constructed, calculated dose rates were confirmed by the actual measurements. No modifications were needed.
•For the first time neutron and gamma ray transport outside JSI TRIGA shielding body was simulated using Monte Carlo Method•Method was validated by comparing calculated and measured dose rates around beam port no. 5 during reactor operation.•Removable shield was placed in front of the beam tube opening in order to study backscattered radiation.•It is demonstrated that developed method can be used in designing appropriate shielding for the future experiments.
The 2021 IAEA software intercomparison for k0-INAA Blaauw, Menno; D’Agostino, Giancarlo; di Luzio, Marco ...
Journal of radioanalytical and nuclear chemistry,
08/2023, Letnik:
332, Številka:
8
Journal Article
Recenzirano
Odprti dostop
In order to establish the variation between results in mass fractions due to software implementation, as measured by the
k
0
-method for INAA, the IAEA has organized a software intercomparison. A ...complete set of test spectra and associated information was assembled. Efficiency curves, neutron spectrum parameters, correction factors and mass fractions were calculated with the participating programs (k0-IPEN, k0-INRIM, k0-DALAT, k0-IAEA and KayWin) using identical peak areas. In this paper, we report on the observed discrepancies, causes, remedies and future software developments. The test data, as well as intermediate results and observed mass fractions of the certified reference material BCR-320R “channel sediment” are available through the IAEA on request. The variations in concentrations attributed to differences between the programs were initially found to be 5.6 and 7.9%, for certified and uncertified concentrations, respectively. After the certified concentrations had been made available to the participants and they had been allowed to improve their programs, the variations found were 2.7 and 3.4%, respectively. The main identified remaining causes of variation are differences in the procedures used for detector efficiency characterisation and neutron spectrum parameter determination.
Two series of measurements were performed in the JSI TRIGA research reactor in 2014 and 2017 to validate the 55Mn(n,γ)56Mn cross-sections and experimentally investigate the relationship between the ...55Mn(n,γ)56Mn reaction and the rate of tritium production through the 6Li(n,t)4He reaction. Indeed, previously observed similarities between the sensitivity profiles of the neutron reaction of tritium production on lithium, 6Li(n,t)4He, and those of the 55Mn(n,γ)56Mn reaction in tritium breeder modules indicated that the latter reaction could be used as an effective monitor of tritium production, at least for short-term monitoring (the half-life of 56Mn being 2.579 h). However, experimental verification, improvements and validation of the 55Mn(n,γ)56Mn cross-sections are needed in order to meet the required accuracy. Foils of certified reference material Al-1%Mn, as well as LiF thermoluminescent detectors and Li2O samples were irradiated, both bare and under cadmium, to study the potential use of the 55Mn(n,γ)56Mn reaction for monitoring tritium production in fusion devices. Additionally, Al-0.1%Au was also irradiated for comparison, the 197Au(n,γ)198Au reaction cross-section being a standard. In order to obtain complementary information for data validation purposes, the irradiations were performed in positions within the JSI TRIGA reactor with different neutron spectra, i.e. in the central channel, the pneumatic tube and the F19 position, both in the outer 'F' ring of the reactor core and in the IC-40 irradiation channel located in the graphite reflector surrounding the reactor core. Bare and cadmium-covered irradiations were needed to subtract the contribution of epithermal neutrons to the 55Mn(n,γ)56Mn reaction. Calculations of the reaction rates were performed using the Monte Carlo code MCNP6.1 with a detailed model of the JSI TRIGA reactor, with the samples, the irradiation capsules and covers being modelled explicitly. The uncertainties involved in the measurements and the calculations were carefully evaluated. The principal objective was to study the energy response and correlations between the 55Mn(n,γ)56Mn reaction in irradiated Al-1%Mn and the 6Li(n,t)4He reaction in irradiated LiF and Li2O. Good consistency between the measured and calculated 55Mn(n,γ)56Mn and 197Au(n,γ)198Au reaction rates, in most cases within the uncertainty bars, was observed.
An evaluation of validity of theoretical pulse models (Fuchs-Hansen and Nordheim-Fuchs model) in reactor pulse modelling was performed by developing so called Improved Pulse Model. The effect of each ...of the five assumptions on the most important pulse physical parameters, maximal power, total released energy and full width at half maximum was studied. In the Improved Pulse Model the assumptions are disposed out with the improvements, where to account the delayed neutrons the six point kinetic equations are solved, the temperature dependences of the temperature reactivity coefficient of fuel and specific heat are taken into account, also the final ejected time of transient control rod from reactor core, whose value of reactivity varies in height and the heat dissipation from the fuel are considered. It is found that the theoretical models predict a higher maximum power, lower total released energy and full width at half maximum than the Improved Pulse Model.
•Chemical shim is used for the control of pressurized water reactors.•Effectiveness of boron as a neutron absorber in controlling the reactivity.•Samples were placed with various concentrations of ...boric acid in the IPR-R1 reactor.•Simulate B-10 consumption during the reactor operation.•Effect on reactor reactivity with increasing boric acid concentrations.
The isotope boron-10 (B-10) has a high thermal neutron absorption cross-section and is used to poison the chain reaction in some light-water reactors (chemical shim). The purpose of this work was to show the effectiveness of boron as a neutron absorber in controlling the reactivity of nuclear reactors cooled by light water. In this way, samples were placed in sealed containers with various concentrations of boric acid in the IPR-R1 Triga nuclear research reactor core. Thus, the reactivity variation of this reactor was determined. The sample characterization was performed before and after the experiments by the measurement of pH and electrical conductivity. The IPR-R1 reactor is located at the Nuclear Technology Development Center (CDTN) in Belo Horizonte, Brazil. Variations of reactivities were evaluated using the static reactivity null method and the dynamic method. The results obtained made it possible to simulate B-10 consumption during the operation of a reactor and its effect on reactivity with increasing boric acid concentrations. The pH values showed small increases after irradiation and their conductivities showed minor changes. As a result of this experiment, a correlation was drawn between various boric acid concentrations and reactor reactivity.
New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to ...check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations.
•TRIGA Benchmark keff calculated with the TRIPOLI code.•Reaction rate profiles in TRIGA calculated with TRIPOLI code.•TRIPOLI model of the JSI TRIGA was validated.•TRIGA Kinetic parameters were calculated with TRIPOLI code.•All results are in good agreement, largest discrepancies due to nuclear data.
•New multigroup Monte Carlo code specifically designed for TRIGA reactor core design.•Complement other MC codes for functionalities that are less optimal.•Burnup calculations based on homogenized ...group cross sections.•Improved simulation performance in keff and fission source convergence.
This paper outlines the details of the development of TRIMON (TRIGA Monte Carlo Code), a next-generation reactor code that integrates diffusion-theory-type group cross sections into Monte Carlo method for TRIGA reactors. Such hybrid combination speeds up stochastic simulations via homogenization of complex local core regions. TRIGA reactors are currently installed in 24 different countries, thus, a robust core management code must be developed for their safety analysis. TRIMON includes important features that improve certain functionalities that are less optimal in most state-of-the-art Monte Carlo codes such as direct integration of local fuel burnup in core calculation, sophisticated reactor core design considerations and simulation time improvements in complex core configurations. TRIMON also eliminates the complicated jargon related to the core geometry and tally specifications imposed by most general multi-purpose Monte Carlo codes during the process of defining TRIGA core problems.
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