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Ensuring the reliable use of particle transport computer codes necessitates verification against benchmark experiments. This study aims to verify the Monte Carlo code OpenMC using the criticality benchmark model IEUCOMP-THERM-003 from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The analysis focuses on the TRIGA Mark II reactor cores 132 and 133, employing nuclear cross-section libraries ENDF/B-VIII.0, ENDF/B-VII.1, ENDF/B-VII.0, and ENDF/BVI. 2. Results show that OpenMC provides KEFF values in close agreement with benchmark values, demonstrating its robustness in neutronic simulations. Comparison with MVP code results obtained previously, particularly with JENDL-3.3, shows similar accuracy. [ABSTRACT FROM AUTHOR]

AbstractHere we present a new method of irradiating 132Xe capsules with neutrons to produce 133mXe gas standards that are used for radiation detector calibration at radioxenon measurement laboratories in support of the Comprehensive Test Ban Treaty (CTBT). This method is designed to maximize the production of 133mXe compared to 133Xe, both of which are competing products from the 132Xe(n, g) reaction. The 133mXe is produced at a much higher fraction for high-energy neutron absorptions in 132Xe (~50% for fast neutrons versus ~11% for thermal neutrons).We performed “spectral tuning” of the Washington State University (WSU) TRIGA reactor neutron spectrum inside the 132Xe ampules to maximize the number of fast neutrons and minimize the number of thermal neutrons available for 132Xe absorption. Spectral tuning analysis, done with Monte Carlo simulations, provided valuable insights into a future final design for a 132Xe irradiation capsule. With no spectral tuning, the fractional yield of 133mXe in the WSU reactor was ~11.7%. By surrounding the 132Xe capsule with a 0.5-cm-thick layer of tungsten and a 2.83-cm layer of europium (III) oxide and placing it in the reactor’s cadmium rotator tube next to the fuel elements, the fractional yield of 133mXe can be increased to 24.6%, a 111% increase in yield. Thus, by improving the fractional yield of 133mXe through spectral tuning, the CTBT will have better quality gas standards to use for radioxenon detector calibration to assist in the CTBT’s mission. [ABSTRACT FROM AUTHOR]

This work is a contribution to the assessment of the radiological consequences of radioactivity and radiation dose for TRIGA Mark II research reactor during continuous operation. The potential release to the atmosphere of 131I, 137Cs, and 90Sr are computed in the South direction is calculated using CROM software. We have attempted to evaluate the daily concentration of radioactivity and its impact on foodstuff. The annual average dose received from internal and external irradiation by age group through the different pathways were also considered. The simulation results showed that the highest air concentration was found at 225 m distance from the source and the calculated doses were found to be significantly very low. The contribution of Iodine 131I is significantly higher in fruit vegetables, while the 137Cs and 90Sr are dominant in animal products. Furthermore, inhalation and ingestion of contaminated food were found to be the most likely routes of entry into the human body.

This work summarizes the radiation transport–based design for a new D2O-moderated ex-core irradiation facility in the Washington State University (WSU) TRIGA reactor for optimization of 135Xe sources used for calibration and quality control testing of Xe gas detection equipment in support of the Comprehensive Test Ban Treaty (CTBT). Three-dimensional (3-D) particle transport analysis characterizing the WSU reactor core using MCNP6.2 (3-D Monte Carlo) and PENTRAN (3-D deterministic parallel SN) form the basis for the computational optimization. Excellent agreement between MCNP6.2 and PENTRAN predictions is observed. A fundamental fuel bundle depletion analysis is applied to enable a more accurate prediction of neutron flux and neutron spectrum distribution, which drives production rates of 135Xe and 133Xe. The results of various model simulations were used to inform recommendations for the final irradiation chamber design, which has been optimized for safe placement in the reactor tank prior to startup and will allow for insertion and rotation of xenon "bean" samples using existing WSU irradiation equipment, while remaining within operational parameters. The irradiation chamber is expected to produce samples that will remain viable for use in CTBT standards applications for durations 70% to 80% longer than samples produced using current procedures. Thus, this design is expected to improve CTBT-related calibrations and performance testing and to support the continued stability of the CTBT monitoring network. [ABSTRACT FROM AUTHOR]

In order to enhance the safety of new advanced reactors, reliability based approach to the design of thermal hydraulic system becomes necessary. In this work, “load exceeds capacity” based approach of structural reliability analysis is employed and probability of failure of the system was then assessed in terms of a limit state function while probabilistic measure of limit state function violation is performed through different methods of reliability assessment. Here, we have focused on TRIGA core subjected to reactivity initiated fast transient. Initially, response surface design method has been used for approximating true failure surface, and then FORM-SORM analysis has been carried out. But, due to non linearity involved with failure surface, there have been noticed instability in FORM-SORM implementation. Later, directional simulation approach of Monte Carlo variance reduction techniques has been employed to illustrate such fast transient. In the investigation, there have been several aspects considered and in each case directional simulation method has shown its ability to give valid results. Hence, the method could be recommended as a viable and efficient scheme to solve even fast transient problem in design and analysis of any reactor.

Compared to conventional engineering systems, nuclear reactors present unique physical and safety features that make their high-fidelity modelling both necessary and complex. Indeed, modelling nuclear reactors is intrinsically a multi-scale and multi-physics task. Up to recent years, the modelling approach for nuclear reactors saw the use of highly performing computer codes and hardware to retrieve a model as close as possible to reality; whereas this remains true, especially from the point of view of regulators and safety assessments, an alternative modelling approach has appeared in the nuclear reactor world. Indeed, whereas high-fidelity models are invaluable for providing in-depth insights into the system, especially when experimental data are not available, their computational cost is such that they are not suited for all applications that involve multiple simulations over a parametric domain (such as in the design and optimisation phase, known as multi-query scenarios). Thus, this novel approach aims at reducing the computational complexity of high-fidelity models whilst preserving high enough accuracy to satisfy the regulatory requirements of the nuclear world, and modelling techniques with this scope fall into the broad category of Model Order Reduction (MOR) techniques. MOR methodologies offer a trade-off between computational cost and solution accuracy; they can also jointly work with Data Assimilation techniques, which deal with the dynamic integration of experimental data and numerical estimates, thus surpassing the logic of using experimental data only as a posteriori validation tool. As the use of MOR and Data Assimilation (DA) for nuclear reactor analysis and for the development of integrated tools and digital twins for the system is still in the first stages, this work overviews some MOR and DA methodologies developed by the authors applied to an existing nuclear system, the TRIGA Mark II research reactor at the University of Pavia, which represents a benchmark test case of a complete nuclear reactor with experimental data available. The choice of using different MOR techniques to tackle various problems follows the logic of developing specific algorithms for specific issues and then merging them into a single MOR and DA-based digital twin, thus reducing the complexity and the cost of the single algorithm modules compared to a single general one: the three MOR techniques considered in this work (Dynamic Mode Decomposition, Proper Orthogonal Decomposition with Kalman Filtering and Generalised Empirical Interpolation Method) follows this logic. Indeed, the results show the potentiality of this approach for complex engineering problems, showing how these techniques can offer significant insights into the system without the computational cost associated with high-fidelity models. • Definition of the MOR and DA frameworks for reactor dynamics analysis. • Three different MOR techniques have been considered: DMD, POD with KF nd GEIM. • Application to the TRIGA Mark II reactor of the University of Pavia as a test case. • Good results in state prediction and model improvement. [ABSTRACT FROM AUTHOR]

[Display omitted] • Estimation of radionuclides formed in core components of TRIGA 2000 Bandung reactor. • This study focuses on the predominant aluminum components in the reactor core. • The numerical calculation uses MCNP 6.1 and ORIGEN 2.1 software. • Aluminum-27 is found to be the most prevalent stable isotope for all components. • Zinc-65 is the most significant radioactivity up to 6 years after decommissioning. This paper presents estimates of the radionuclides formed in the core components of the TRIGA 2000 Bandung reactor. This research is crucial for updating decommissioning programs and predicting the quantity and activity levels of radioactive wastes. This study primarily focused on the predominant aluminum components in the reactor core. MCNP 6.1 was utilized to calculate the neutron flux during the reactor operation, whereas ORIGEN 2.1 determined the radionuclides resulting from neutron activation. The calculations span over six years, with 2021 as the base year. The nine modelled components include the reflector wall, Lazy Susan, thermal column, thermalizing column, core wall and safety plate, shim wall, top grid plate, bottom grid plate, and beam port. The findings indicate that Aluminum-27 (Al-27) is the most prevalent stable isotope across all components, with mass percentages ranging between 97.89% and 97.93%. Aluminum exhibits low activation, and the radionuclides produced primarily stem from other elements present in 6061-type aluminum. Among the radionuclides, Zinc-65 (Zn-65) displayed the most significant radioactivity, with percentages varying from 81.6% to 99.82% in the initial year and diminishing over six years due to radioactive decay. By the sixth year, a notable reduction in the radioactivity was observed. Decommissioning nuclear facilities requires a holistic approach that encompasses cost, risk, potential environmental impacts, human safety, regulatory compliance, and the chosen technical strategy. Such planning is vital for adhering to the safety and security standards of the national regulatory body, BAPETEN. [ABSTRACT FROM AUTHOR]

FiR 1 is a TRIGA Mark II-type research reactor in Finland. It was in operation between 1962 to 2015 and will be dismantled in 2022 to 2023. Preliminary calculations of the activities in the reactor main structures were performed in an earlier stage of the decommissioning project. Samples of the activated parts of the reactor biological shield concrete were drilled in December 2018 to validate these estimates. This paper describes the calculations and gamma activity measurements performed for the activated concrete samples to determine the boundary between radioactive parts and concrete that can plausibly be free-released from regulatory control. The activities have been estimated with a two-step calculation process using the MCNP and ORIGEN-S calculation codes and measurements using an ISOCS gamma spectrometer with a high-purity germanium detector. [ABSTRACT FROM AUTHOR]

A closed-water activation loop is being built at the Jožef Stefan Institute TRIGA reactor in Slovenia to serve as a well-defined and stable source of high-energy gamma rays and neutrons. The radial piercing port, which penetrates the graphite reflector and touching the reactor core was chosen for the installation of the closed-water loop due to the high neutron flux and favourable shielding conditions of the surrounding concrete bioshield. The main objective of this work is to assess the neutron and gamma dose field outside the port to obtain important details for the final design of the inner part of the irradiation facility and to assess the background noise for the detectors. The main part of the work consists of the design of the shielding plugs and the construction of a detailed MCNP model of the whole irradiation facility. The dose field calculations were performed with a two-step hybrid transport approach using ADVANTG for variance reduction and MCNP for particle transport. Such deep penetration and shielding calculations are challenging and computationally intensive. The results showed that the dose rate using shielding plugs is more than 7 orders of magnitude lower compared to an empty open port. To reduce the computational uncertainty, further optimisation of the ADVANTG input is essential. The final design of the shielding plugs is described. Additional lead shielding blocks will be added outside the port afterwards if needed.

Monte Carlo methods are useful for simulating new reactor designs, but even with advances in computing, these methods still require a significant amount of time to perform transient or multiphysics calculations coupled with thermal modeling. This work demonstrates a hybrid reactor physics method that uses Monte Carlo to precalculate an initial database of fission matrix parameters, then combines these results for fast calculations on arbitrary system states. This paper extends previous work that demonstrated these methods on the Penn State Breazeale Reactor (PSBR). Approaches for reducing time and memory cost and increasing the accuracy in reproducing Monte Carlo output are considered. For modeling fuel temperature, a representative temperature distribution is used while tallying the initial fission matrix database. Different approaches for modeling the coupling between individual control rod insertions as well as control and fuel temperature effects are presented as well. Individual solutions are completed in less than 1 s on a single core, and error with respect to Monte Carlo is within 35 pcm for multiplication factor, 0.6% root-mean square, and 2.8% maximum for the normalized three-dimensional fission source distribution on critical, steady-state configurations. Further qualification on different reactor types is needed, but the simplicity and flexibility of this method make further development promising. [ABSTRACT FROM AUTHOR]

A novel polar angular quadrature set called the Kim-Jevremovic polar angular quadrature set is derived for the method of characteristics. It is based on neutron anisotropic scattering cross-sections in the Evaluated Nuclear Data File. This new set is implemented within the state-of-the-art neutron transport code AGENT and tested in comparison to MCNP6 as well as to other known quadrature sets for the UO2 unit cells, the well-known C5G7 benchmark, unreflected cylinders of uranyl-fluoride solutions in heavy water, and the University of Utah 100 kWth TRIGA MARK-I reactor core. These comparisons show that the newly proposed polar angular quadrature set provides better agreements than other quadrature sets for the lower order of anisotropic scattering expansions. This paper presents a complete derivation of the Kim-Jevremovic polar angular quadrature set and the analysis for the mentioned bench-mark examples.

TRIGA type research reactors, even relatively new ones, are originally equipped with rather obsolete irradiation pneumatic transfer systems. Therefore, the irradiation system of Slovenian TRIGA Mark II system was renewed in 2015 to improve the overall quality of irradiations. This year, the system was upgraded to allow for automated short irradiations under more thermalized neutron flux. The modernization includes also a so-called "triangular" channel, allowing for in-core irradiation of samples up to 5 cm in diameter and a horizontal channel allowing for irradiations of objects under homogeneous neutron flux in the length of over 60 cm. [ABSTRACT FROM AUTHOR]

The neutron field of various irradiation positions of the TRIGA Mark II reactor of the Jožef Stefan Institute (JSI) has been thoroughly characterized by neutron activation dosimetry and miniature fission chambers techniques. In order to have a fully validated calculation scheme to analyze and plan experiments, the gamma field also has to be experimentally validated. The 10-year long collaboration between CEA and JSI is a perfect framework to carry out such a study, and measurements of the gamma field started in late 2016. Several measurement techniques were investigated in in-core and ex-core positions. Online measurements were carried out using miniature ionization chambers (ICs) manufactured by the CEA and PTW Farmer. Positional dependence was studied, showing a decrease in the delayed gamma contribution to the total gamma flux with increasing distance from the reactor core center. To characterize the gamma dose in the core, as well as in the periphery, thermoluminescent and optically stimulated luminescent detectors (TLDs and OSLDs, respectively) were tested. These detectors are commonly used at CEA to measure the gamma dose in a given material, in order to study the nuclear heating in various core elements (control rod, baffle, structural material). Different filters were used in order to assess an integrated dose ranging from a few Gy up to several kGy. The comparisons of experimental results against calculations performed with the JSIR2S code package show a very good agreement. The feasibility of such measurements demonstrates the complementarity between measurements with dosimetry and ICs from low to very high gamma-dose environment, such as in material testing reactors. [ABSTRACT FROM AUTHOR]

An RIA (reactivity initiated accident) analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1) to 2.0 % dk/k (>$2). The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip) was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57) for step reactivity and 1.99 % dk/k ($2.84) for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes. In this paper, the applicability of the RELAP/SCDAPSIM/MOD3.4 thermal-hydraulic code to one Indonesian research reactor, which is named TRIGA-2000, is performed. The aim is to validate the model and use the model to analyze the thermal-hydraulic characteristics of TRIGA-2000 for main transient events considered in the Safety Analysis Report. The validation was done by comparing the calculation results with experimental data mainly in steady state conditions. The comparison of calculation results with the measurement data showed good agreement with little discrepancies. Based on these results, simulations for thermal-hydraulic analyses were performed for loss of coolant transients. The calculation results also properly depicted the physic of the thermal-hydraulic phenomena following the loss of coolant transients. These results showed the adequacy of the model. It could be shown that the engineered safety features of TRIGA-2000 play an important role in keeping the reactor safe from the risk of postulated loss of a coolant accident.

A novel polar angular quadrature set called the Kim-Jevremovic polar angular quadrature set is derived for the method of characteristics. It is based on neutron anisotropic scattering cross-sections in the Evaluated Nuclear Data File. This new set is implemented within the state-of-the-art neutron transport code AGENT and tested in comparison to MCNP6 as well as to other known quadrature sets for the UO2 unit cells, the well-known C5G7 benchmark, unreflected cylinders of uranyl-fluoride solutions in heavy water, and the University of Utah 100 kWth TRIGA MARK-I reactor core. These comparisons show that the newly proposed polar angular quadrature set provides better agreements than other quadrature sets for the lower order of anisotropic scattering expansions. This paper presents a complete derivation of the Kim-Jevremovic polar angular quadrature set and the analysis for the mentioned bench-mark examples. [ABSTRACT FROM AUTHOR]

Sensitivity analysis studies the effect of a change in a given parameter to a response function of the system under investigation. In reactor physics, this usually translates into the study of how cross sections and fission spectrum modifications affect the value of the multiplication factor, the delayed neutron fraction or the void coefficient for example. Generalized Perturbation Theory provides a useful tool for the assessment of adjoint weighed functions such as keff and void coefficient sensitivities. In this work, the capability of SERPENT code to perform sensitivity calculation based on GPT is used to study the TRIGA Mark II research reactor installed at L.E.N.A. of University of Pavia. A general sensitivity analysis to the most important reactor’s cross sections has been performed in order to highlight the biggest reactivity contributions. Two numerically challenging tasks related to GPT calculation have been performed thanks to the relatively quick Monte Carlo approach allowed by this reactor: investigating the linearity of the reactivity injection caused by the flooding of the central channel, and calculating the fuel void coefficient sensitivity to the coolant density.

Fuel burnup of the JSI TRIGA was calculated by simulating complete operational history consisting of 240 different core configurations from 1966 to 2020. At the moment we are unable to perform burnup measurements, e.g. gamma spectroscopy on burned fuel elements, hence we used weekly measured excess reactivity as a reference point of different core configurations to verify the calculated core reactivity. Changes in reactivity due to burnup were assumed to be linear and this assumption was verified for burnup intervals smaller than 3 MWd/kg(HM). The comparison was performed on 46 different core configurations with different type of fuel elements. The Serpent-2 calculations decently predict the rate of reactivity change on different cases, as 52 % of calculations are withing 1σ and 86.9 % within 2σ of the measurements for total number of 46 cases. Additional analysis was performed by comparing unit cell calculations of different fuel types. Four different types of TRIGA fuel were used to analyse burnup changes in LEU and HEU fuel, where positive reactivity feedback on burnup was observed for HEU fuel due to burnable absorbers. Serpent-2 and WIMSD-5B were compared on unit-cell basis where good agreement within 200 pcm of reactivity change for large burnup was observed. In addition neutron spectrum changes due to burnup were investigated using unit-cell calculations where 4 % increase of the thermal peak and 1 % decrease of fast peak of the spectrum was observed for typical fuel burnups of 20 MWd/kg(HM), which approximately represents JSI TRIGA burnup at this moment.

Constant improvements of the computational power and methods as well as demands of accurate and reliable measurements for reactor operation and safety require a continuous upgrade of the instrumentation. In particular, nuclear sensors used in nuclear fission reactors (research or power reactors) or in nuclear fusion facilities are operated under intense mixed neutron and gamma-ray fields, and need to be calibrated and modeled to provide selective and accurate neutron and gamma-ray measurements. The French Atomic Energy and Alternative Energies Commission (CEA) and the Jožef Stefan Institute (JSI) have started an experimental program dedicated to a detailed experimental benchmark with analysis using Monte Carlo particle transport calculations and a series of neutron and gamma-ray sensor types used in the JSI TRIGA Mark II reactor. CEA has setup a simplified TRIPOLI-4® modeling scheme of the JSI TRIGA reactor based on the information available in the IRPhEP benchmark in order to facilitate analysis of future neutron and gamma-ray measurements. These allow the CEA to perform a TRIPOLI-4 instrumentation calculation scheme benchmarked with the JSI MCNP model. This paper presents the main results of this CEA calculation scheme application and the analysis of their comparison to the JSI results obtained in 2012 with the MCNP5 & ENDF/B-VII.0 calculation scheme. This paper will conclude with some information about the new experimental program to be carried out in 2022 in the TRIGA reactor core.

Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

Abstract Well characterized neutron and gamma fields inside a nuclear reactor are of key importance for its safe operation and for successful utilization of various research reactor irradiation facilities. In case of high-flux research reactors such as BR2 in Belgium, Maria in Poland and the future Jules Horowitz Reactor in France, the gamma energy deposition rate in reactor structural components and irradiated samples reaches values well over 10 W/g. To assure safe reactor operation, the gamma field and associated heating must therefore be thoroughly characterized in order to provide adequate component and sample cooling. The gamma field can be divided into two contributions: prompt gamma rays are emitted almost instantly after neutron interaction with nuclei, while the delayed gamma rays are emitted from nuclei, which become radioactive by neutron absorption, generated from fission and other processes. Most modern Monte-Carlo particle transport codes enable the transport of prompt gamma rays; a few support delayed gamma ray generation and transport as well. The latter have mostly been applied to fusion devices, where detailed shutdown dose-rate measurements have been performed. Although the delayed gamma field can also be simulated in fission devices, significant inaccuracy in the result is to be expected due to the computational complexity arising from the large number of radioactive fission products and incompleteness of nuclear data. Furthermore, the unavailability of experimental delayed gamma measurements in fission systems presents an important challenge for the validation of the experimental results. Previous measurements in several research reactors show that the delayed gamma flux amounts to around 30 % of the total gamma flux. However, these evaluations were performed with measurement data obtained during rapid reactor shutdowns (SCRAMs), using a single measurement point per SCRAM. In this paper we propose a new technique to accurately determine the magnitude of the delayed gamma component and its time evolution, based on synchronous acquisition of fission and ionization chamber signals. The measurements were performed at the JSI TRIGA reactor, using fission and ionization chambers placed in several in-core measurement positions. Their signal was acquired synchronously and at the highest possible acquisition rate in order to distinguish between measurement noise and reactor transients. Using the novel delayed gamma extraction technique we were able to estimate the magnitude of the delayed gamma contribution to be: 18.9 % ± 2.0 % at the reactor core periphery, linearly increasing towards the reactor core center to 31.4 % ± 2.8 % of the total measured gamma flux signal after 10 min of reactor operation. [ABSTRACT FROM AUTHOR]

In this paper, Nodal dynamic method is introduced to model the TRIGA Mark II reactor of the University of Pavia as a dynamic system for the total operative power range (i.e. 0–250 kW) using a zero dimensional thermal hydraulic method. Neutronic and thermal hydraulic models are coupled in order to demonstrate reactor dynamic behavior. The reactor is divided into bi-dimensional zones and simulated in Serpent 2 to obtain two group cross sections for each specified zone as well as kinetic parameters. And they are applied in nodal method for the analysis of reactor dynamic behavior. Point kinetic and nodal dynamic methods are utilized for the second core configuration as the neutronic model. Different reactivities (i.e. 190, 104, 78, 40 pcm) are inserted by control rods in different power levels (i.e. 1, 50, 100, 150 kW) for the reactor dynamic analysis. A program is written in MATLAB to couple neutronic and thermal hydraulic models. A system of ordinary differential equations are produced and solved in space state model. The calculated and experimental power excursion results are in good agreement with less than 1% difference. The results between nodal method and point kinetic method are very consistent with less than 0.05% of difference. [ABSTRACT FROM AUTHOR]

The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using free open source software is presented. The proposed contributions go in two different directions: one, is the focus on the open software approach development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code milonga . This concept was applied to model the behavior of a TRIGA-IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using the Serpent code. The results show that this coupled system gives consistent results, encouraging system further development and its use for complex geometries simulations. [ABSTRACT FROM AUTHOR]

The computational model of the JSI TRIGA Mark II, coupling Monte-Carlo neutron transport code TRIPOLI and computational fluid dynamics code CFX was used to reproduce the behaviour of the reactor after extraction of a control rod. To tackle the time dependent Boltzmann equation, a quasistatic approach has been used and compared with point kinetic. Qualitative assessment of the model was performed by comparison with measured fuel temperature and power. Time evolutions of power and fuel temperature were reproduced. The quasistatic approximation was justified by updating the shape function at different time intervals. The quasistatic approach successfully reproduces the experimental results obtained with the TRIGA reactor. It was shown that most of the local effects (temperature, power density) were due to the control rod and that local effects of coupling were small. [ABSTRACT FROM AUTHOR]

The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications. [ABSTRACT FROM AUTHOR]

Proficiency testing is an important way of evaluating the analytical method used in the laboratory. In recent years, neutron activation analysis studies performed in ITU TRIGA Mark II reactor comprised five proficiency tests organized by Wageningen evaluating programs for analytical laboratories. In this study, the results obtained by relative INAA and k0-INAA method for 16 elements in soil and plant samples are presented. Since both methods have some advantages compared to each other, the possible approach for the laboratory should be to combine relative INAA and k0-INAA results. [ABSTRACT FROM AUTHOR]

The neutron field of various irradiation positions of the TRIGA Mark II reactor of the Jožef Stefan Institute has been thoroughly characterized by neutron activation dosimetry and miniature fission chambers techniques. In order to have a fully validated calculation scheme to analyze and plan experiments, the gamma field also has to be experimentally validated. The 10-year long collaboration between CEA and JSI is a perfect framework to carry out such a study, and measurements of the gamma field started in late 2016. Several measurement techniques were investigated in in-core and ex-core positions. On-line measurements were carried out using miniature ionization chambers manufactured by the CEA and PTW Farmer ionization chambers. Positional dependence was studied, showing a decrease in the delayed gamma contribution to the total gamma flux with increasing distance from the reactor core center. To characterize the gamma dose in the core, as well as in the periphery, thermo- and optically stimulated luminescent detectors were tested. These detectors are commonly used at CEA to measure the gamma dose in a given material in order to study the nuclear heating in various core elements (control rod, baffle, structural material). Different filters were used in order to assess an integrated dose ranging from a few Gy up to several kGy. The feasibility of such measurements demonstrates the complementarity between measurements with dosimetry and ionization chambers from low to very high gamma-dose environment, such as in material testing reactors.

The JSI TRIGA reactor features several in-core and ex-core irradiation facilities, each having different properties, such as neutron/gamma flux intensity, spectra and irradiation volume. A series of experiments and calculations was performed in order to characterise radiation fields in irradiation channel thus allowing users to perform irradiations in a well characterised environment. Since 2001 the reactor has been heavily used for radiation hardness studies for components used at accelerators such as the Large Hadron Collider (LHC) at CERN. Since 2010 it has been extensively used for testing of new detectors and innovative data acquisition systems and methods developed and used by the CEA. Recently, several campaigns were initiated to characterise the gamma field in the reactor and use the experimental data for improvement of the treatment of delayed gammas in Monte Carlo particle transport codes. In the future it is planned to extend the testing options by employing pulse mode operation, installation of a high energy gamma ray irradiation facility and allow irradiation of larger samples at elevated temperature.

The CNESTEN (National Center for Energy Sciences and Nuclear Technology, Morocco) operates a TRIGA Mark II reactor, which can reach a thermal maximum power at steady state of 2 MW. In reactors devoted to research and experiments, it is mandatory to characterize the neutron and photon fields in the irradiation positions. Together with a computational model of the core, it ensures the ability to reach the requested uncertainties when performing experiments, such as detectors testing, irradiation for hardening or nuclear data measurements. The neutron field of different irradiation positions has been characterized by dosimetry techniques and compared to the MCNP full model of the reactor. Preliminary photon propagation calculations are also performed with this model, but up to now, no experimental validation of the results exists. The aim of the newly set collaboration between CEA and CNESTEN is to characterize the gamma field of these positions. The first position investigated is the part of the NB1 tangential channel closest to the core. Among gamma measurements techniques, and according to the constraints arising from using this channel, it was chosen to use thermos- and optically stimulated luminescent detectors. This paper presents the experiments carried out in September 2018 as well as their results. Three detectors types were used: TLD400 (CaF2:Mn), TLD700 (7LiF:Mg,Ti) and OSLD (Al2O3:C). Measurements were performed in several steps: background measurements, transient measurements (divergence phase + SCRAM), and irradiation at steady state. In the end, these measurements will provide a dose as well as a gamma flux value for this position.

Activated cooling water in nuclear facilities can present a significant radiation source around primary cooling system causing radiation damage to electrical components, increasing doses to personnel and in the case of fusion facilities additional heating to superconducting coils. As there are only few sources of gamma rays with energies in the range of 6 MeV and 7 MeV an irradiation system using activated cooling water as the source of energetic gamma rays is proposed at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Two different conceptual designs, one utilizing central irradiation channel and one utilizing radial piercing port for water activation, are presented and analysed in the paper. Despite an order of magnitude higher water activation in central channel compared to radial piercing port the 16N decay rate in the irradiation facility is comparable between both design (order of 108 decays per second) due to longer transient time from central channel to irradiation facility. In the irradiation facility the expected biological dose rates due to the 16N decay rate are in order of several mSv/h. From the results he conceptual design utilizing the radial piercing port currently presents the best option for the irradiation facility due to the simpler design of the irradiation loop, already present shielding of the loop and comparable number of 16N decay rates to central channel.

Heating due to energy deposition of intense ionizing radiation in samples and structural materials of nuclear reactors poses severe limitations in terms of cooling requirements for safe reactor operation, especially in high neutron and gamma flux environments of material testing fission reactors (MTRs) and novel fusion devices. A bilateral CEA-JSI research project was launched in 2018 with the objective to measure the gamma heating rates in standard reactor-related materials (graphite, aluminium, stainless steel and tungsten) as well as fusionrelevant materials (low-activation steel Eurofer-97 and Nb3Sn superconductor) in the JSI TRIGA reactor my means of gamma calorimeters. The calorimeter design will be based on the the CALMOS-2 calorimeter developed at the CEA and used to perform gamma heating measurements in the OSIRIS MTR in Saclay. In order to optimize the detector response inside the JSI TRIGA reactor field and not to perturb the measurement field, a detailed computational analysis was performed in terms of energy deposition assessment and measurement field perturbation using the MCNP v6.1 code, and in terms of heat transfer using the COMSOL Multiphysics code. The abovementioned activities enabled us to finalize the detector design with the experimental campaign planned for the end of year 2019.

A well characterized radiation field inside a research nuclear reactor irradiation facilities enables precise qualification of radiation effects to the irradiated samples such as nuclear heating or changes in their electrical or material properties. To support the increased utilization of the JSI TRIGA reactor irradiation facilities in the past few years mainly on account of testing novel detector designs, electronic components and material samples, we are working on increasing the neutron and gamma field characterization accuracy using various modeling and measurement techniques. In this paper we present the dose field measurements using thermo-luminescent detectors (TLD’s) with different sensitivities neutron and gamma sensitivities, along with multiple ionization and fission chamber. Experiment was performed in several steps from reactor start-up, steady operation and a rapid shutdown, during which the ionization and fission chamber signals were acquires continuously, while the TLD’s were being irradiated at different stages during reactor operation and after shutdown, to also capture response to delayed neutron and gamma field. The results presented in this paper serve for validation of JSI designed JSIR2S code for delayed radiation field determination, initial results of its application on the JSI TRIGA TLD measurements will also be presented.

The prospect of operation of the Indonesian TRIGA reactors may be jeopardizes in the future due to the lack of fuel and control rods. Both fuel and control rods may not longer be imported and should be developed domestically. The most specific technology to fabricate TRIGA fuel rod is the production of UZrH1.6 pellet. The steps include converting the massive U metal into powder in by hydriding-dehydriding technique and mixing the U and Zr powders. A research has been planned to conducted by the National Nuclear Energy Agency (BATAN) in Indonesia. Fixed amount of U-Zr mixed powders at the ratio of U/Zr = 10 wt% was pressed into a pellet with a diameter of 1.41 in and a thickness of 1 or 1.5 in, sintered at a temperature of 1200oC, followed by hydriding at 800oC to obtained UZrH1.6. The pellets, cladding, and other components were then fabricated into a fuel rod. A detailed discussion of the TRIGA fuel fabrication is presented in the paper.

The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ -rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ -ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ -ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5 × 10 3 Sv h − 1 to 6 × 10 6 Svh − 1 , silicon dose equivalent from 6 × 10 2 Gy / h si to 3 × 10 5 Gy / h si , and neutron air kerma from 4.3 × 10 3 Gy h − 1 to 2 × 10 5 Gy h − 1 . Ratio of fast ( 1 MeV < E ) vs. thermal neutrons ( E < 0.625 eV ) ranges from 1.5 to 8.4 × 10 − 2 , γ ray ambient dose equivalent at full reactor power from 3.4 × 10 3 Sv h − 1 to 3.6 × 10 5 Sv h − 1 and γ air kerma range 3.1 × 10 3 Gy h − 1 to 2.9 × 10 5 Gy h − 1 . [ABSTRACT FROM AUTHOR]

A new computational model of the JSI TRIGA Mark II, coupling Monte Carlo neutron transport code TRIPOLI and fluid dynamics code CFX was built and verified with a set of new experimental data. A set of subroutines was developed to allow the communication between the Monte-Carlo transport code and CFD code. First, test of the coupling scheme is presented: for a given thermal power of the reactor, the coupled model numerically reproduced fuel temperature monitored during reactor operation and axial water temperature profile measured in the coolant channels. Then axial temperature profiles in the coolant channels were measured with a newly developed sensor during steady-state operation. Predictions of the coupled model are in expected agreement with experimental data recorded during reactor operations. Influence of the coupling has been investigated. [ABSTRACT FROM AUTHOR]

The impact of spatial discretization on reactivity biases in TRIGA fuel models was analyzed. In particular, unit-cell analyses in 2-D and 3-D were performed using Serpent to understand how spatial discretization affects the accuracy with which the effects of material evolution and temperature feedback are resolved. For temperature-dependent cases, a simple, single-channel model was employed. Analysis of 2-D models showed that essentially no radial discretization resolution is needed to eliminate biases (i.e., to reduce biases to the level of stochastic uncertainties) due to material evolution but that more than eight, equal-area, radial regions are needed to resolve temperature-feedback effects. Analysis of 3-D models showed that at least seven, equal-volume, axial regions may be required to resolve material evolution with temperature feedback leading to insignificant additional bias. Because of memory constraints, a full-core model with radially- and axially-resolved fuel elements may be impractical, especially for production-level analyses. Consequently, an “effective Doppler temperature” was determined empirically as a function of the radially-averaged temperature and may be used for future, full-core analyses. [ABSTRACT FROM AUTHOR]

Any events presumed to risk the safety of a nuclear reactor should be analyzed. In a research reactor, the applicability of best estimate thermal-hydraulic codes has been assessed for safety analysis purposes. In this paper, the applicability of the RELAP/SCDAPSIM/MOD3.4 thermalhydraulic code to one Indonesian research reactor, which is named TRIGA-2000, is performed. The aim is to validate the model and use the model to analyze the thermal-hydraulic characteristics of TRIGA-2000 for main transient events considered in the Safety Analysis Report. The validation was done by comparing the calculation results with experimental data mainly in steady state conditions. The comparison of calculation results with the measurement data showed good agreement with little discrepancies. Based on these results, simulations for thermal-hydraulic analyses were performed for loss of coolant transients. The calculation results also properly depicted the physic of the thermal-hydraulic phenomena following the loss of coolant transients. These results showed the adequacy of the model. It could be shown that the engineered safety features of TRIGA-2000 play an important role in keeping the reactor safe from the risk of postulated loss of a coolant accident. [ABSTRACT FROM AUTHOR]

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10 MW with engineering hot channel factors included. Thus, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0. [ABSTRACT FROM AUTHOR]

• Setting up a closed-water loop for water activation based experiments at JSI, Slovenia. • Considering ITER relevant conditions. • Effective water volume of the activation part is essential parameter for obtaining high activity values along the loop. • A pipe radius of 0.6 cm and a water flow rate range between 0.4 l/s and 0.5 l/s were chosen as the optimal configuration. • Estimated activity values of 16N, 17N and 19O within the radiation part are 1.7·108 Bq, 1.2·104 Bq and 1.4·107 Bq. A closed-water activation loop is being built at the research reactor TRIGA at the Jožef Stefan Institute, Slovenia, which will serve as a well-defined and stable high energy gamma-ray and neutron source. The main focus of this work is to analyse and optimise the main components of the water activation loop to achieve the highest activity at reasonably low pressure drop. Both neutronic and hydraulic aspects were considered. The results show that the desired configuration consists of a high effective water volume of the inner activation part, outer observation part and narrow transport pipes connecting both parts. The complex snail-shape design of the inner part systematically outperformed other simplified designs by more than a factor of two for the main water activated isotopes (16N, 17N and 19O). The activity values increase strongly with increasing flow rate, but reach saturation at about 0.5 l/s. On this basis, the snail configuration was selected as the main candidate for both the inner activation and outer observation part. To not exceed the pressure drop of 3 bar within the closed-water loop, the nominal water flow rate and the inner diameter of the transport pipes were set to 0.5 l/s and 1.2 cm, respectively. [ABSTRACT FROM AUTHOR]

The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination with experimental procedures so as to assess the potential maximum University of Utah TRIGA power with the currently available control rod system and have created feasibility studies for assessing new core configurations that could provide higher core power levels. For the maximum determined power of a new University of Utah TRIGA core arrangement, a new control rod system was proposed.

Neutron anisotropic scattering method of the AGENT (Arbitrary GEometry Neutron Transport) reactor neutronics modeling system is verified with a few representative benchmark examples in comparison to MCNP6. Benchmark examples include various unit cell types, and the University of Utah 100 kWth TRIGA MARK-I reactor (UUTR). The effect of neutron anisotropic scattering is analyzed and visually assessed in showing that the AGENT anisotropic scattering of up to P5 order of Legendre expansion provide the expected agreements with MCNP6 and with the gained computational times. [ABSTRACT FROM AUTHOR]

The nuclear reactor has two types of power reactors based on the function that is used as electrical energy and research reactors and radioisotope used as a producer of nuclear science and technology research. Indonesia has three research reactors are two types of TRIGA research reactor in Bandung with a power of 2 MW and in Yogyakarta with a power of 100 kW and a research reactor in PUSPIPTEK Serpong with nominal power of 30 MW. For the second fuel elements TRIGA type reactor that is currently using elements of cylindrical material, while the fuel elements RSG-GAS reactor in Serpong-shaped plate. Reactor TRIGA Bandung is a reactor that can be used to predict the buffer reactor. However, this reactor has a problem because of the limited number of existing fuel element. Meanwhile, production of TRIGA fuel elements abroad already closed. Given that Indonesia has the ability to produce nuclear fuel elements for research reactors fueled plate it is proposed to modify the reactor core of TRIGA Bandung of the terrace patio made from a cylinder into fuel plates. In this research will be studied more deeply about aspects thermal-hydraulics TRIGA research reactor using fuel elements plate replacement to cylinder fuel element. The method used is the modeling and simulation of the fuel element plate using porous media and nonporous media with the program of CFD Code. Results of the simulation show that the phenomenon of flow and temperature distribution closer to the comparison of the design elements used fuel plate. So later this plate fuel elements can be used in a nuclear reactor core TRIGA research Bandung, Indonesia. [ABSTRACT FROM AUTHOR]

This work presents the modeling and simulation of a TRIGA-Mark II pool-type reactor with Zirchonium-Hydryde and Uranium fuel immersed in light water, with Modelica object-oriented language, in Dymola simulation environment. The model encompasses the integrated plant system including the reactor pool and cooling circuits. The reactor pool plays a fundamental role in the system dynamics, through a thermal feedback effect on the reactor core neutronics. The pool model is tested against three experimental transients: simulation results are in good accordance with experimental data and provide useful information about the inertial effect of the water inventory on the reactor cooling. [ABSTRACT FROM AUTHOR]