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In this paper, detailed verification and experimental validation of the formulations and algorithms of the Multi-stage Response-function Transport (MRT)–based Real-time Analysis for Particle-transp...

The TRIGA fuel of the Philippine Research Reactor-1 (PRR-1) will be used in a subcritical reactor assembly (SRA) to strengthen and advance nuclear science and engineering expertise in the Philippines. SRA offers a versatile and safe training and research facility since it can produce neutrons through nuclear fission reaction without achieving criticality. In this work, we used a geometrically detailed model of the PRR-1 TRIGA fuel to design a subcritical reactor assembly and calculate physical parameters of different fuel configurations. Based on extensive neutron transport simulations an SRA configuration is proposed, comprising 44 TRIGA fuel rods arranged in a 7 × 7 square lattice. This configuration is found to have a maximum k e f f value of 0.95001 ± 0.00009 at 4 cm pitch. The SRA is characterized by calculating the 3-dimensional neutron flux distribution and neutron spectrum. The effective delayed neutron fraction and mean neutron generation time of the system are calculated to be 748 pcm ± 7 pcm and 41 μs , respectively. Results obtained from this work will be the basis of the core design for the subcritical reactor facility that will be established in the Philippines.

Neutron detection is a relevant topic in the field of nuclear instrumentation. It is at the heart of the concerns for fusion applications (neutron diagnostics, measurements inside the Test Blanket Modules TBM) as well as for fission applications (in-core and ex-core monitoring, neutron mapping or safety applications in research reactors). Moreover, due to the even more harsh conditions of the future experimental reactors such as the Jules Horowitz Reactor (JHR) or International Thermonuclear Experimental Reactor (ITER), neutron detectors need to be adapted to high neutron and γ fluxes, high nuclear heating rates and high temperatures. Consequently, radiation and temperature hardened sensors with fast response, high energy resolution and stability in a mixed neutron and γ environment are required. All these requirements make wide-bandgap semiconductors and, more precisely, Silicon Carbide (SiC) serious candidates due to their intrinsic characteristics in such extreme environments. Thus, since the last decades, SiC-based detectors are developed and studied for neutron detection in various nuclear facilities. In this paper, a SiC-based neutron detector is 3-D designed and studied through thermal and radiation-matter interaction numerical simulations for a future irradiation campaign at the Jožef Stefan Institute TRIGA Mark II research reactor in Slovenia. Firstly, this paper presents the scientific background and issues of our SiC-based detectors. In a second part the 3-D geometry is shown. Thereafter, the 3-D numerical thermal simulation results are reported. Finally, the 3-D numerical radiation/matter interaction simulations results are presented.

In 2016, the “E-SiCure” project (standing for “Engineering Silicon Carbide for Border and Port Security”), funded by the NATO Science for Peace and Security Programme was launched. The main objective is to combine theoretical, experimental and applied research towards the development of radiation-hard SiC-based detectors of special nuclear materials (SNM), with the end goal to enhance border and port security barriers. Prototype neutron detectors, configured as 4H-SiC-based Schottky barrier diodes, were developed for the detection of secondary charged particles (tritons, alphas and lithium atoms) which are the result of thermal neutron reactions on 10B and 6LiF layers above the surface of the 4H-SiC diodes. We designed a stand-alone prototype detection system, consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer, for testing of neutron detector prototypes at the Jožef Stefan Institute (JSI) TRIGA reactor using a broad beam of reactor neutrons. The reverse bias for the detector diode and the power to electronic system were provided by a standalone battery-powered voltage source. The detector functionality was established through measurements using an 241Am alpha particle source. Two dedicated experimental campaigns were performed at the JSI TRIGA reactor. The registered pulse height spectra from the detectors, using both 10B and 6LiF neutron converting layers, clearly demonstrated the neutron detection abilities of the SiC detector prototypes. The computed neutron detection sensitivity of the single prototype detectors demonstrates that scaling SiC detectors into larger arrays, of dimensions relevant for border and port radiation detectors, could enable neutron sensitivity levels matching gas-based detector technology.

The design and test of a robot-compatible, radiation detection instrument providing simultaneous $\gamma $ -ray imaging and $\gamma $ -ray spectroscopy is described. The sensing system comprises a cerium bromide inorganic scintillation detector and a cylindrical, lead slot collimator that is configured with a robot-compatible, on-board data acquisition system. The mount for the sensor is a lightweight, bespoke 3-axis gimbal actuated by servos for pan, tilt and rotation of the collimator for imaging capability. This paper discusses the integration of this relatively low-cost radiation detection apparatus with a commercially available, Robot-Operating-System-controlled robotic platform (a Clearpath Robotics™ Jackal). The detection system is compliant with the power and mass payload constraints of the robot. Its performance has been evaluated by means of two practical examples: a) measurements in a laboratory environment to assess the ability of the system to resolve two caesium-137 point sources, and b) deployment at the Jožef Stefan Institute TRIGA Mark II research reactor to assess the ability of the system to characterise the $\gamma $ -ray emission at 1 kW from a horizontal tangential beam port in the reactor hall and from the reactor sample pool above the core.

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, allows for real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and subcritical facilities. This paper presents a novel methodology for modeling control rods movement within the RAPID’s Fission Matrix (FM) formulation. This methodology, referred to as FM Control Rods “deltas” (FM-CRd), allows for combination and interpolation of RAPID’s FM coefficients to take into account the effect of inserting the control rods into the reactor core, while simultaneously limiting the amount of pre-calculations required. The FM-CRd methodology is benchmarked against experimental data from the JSI TRIGA ICSBEP benchmark and reference Serpent continuous-energy Monte Carlo calculations. The results show that RAPID’s FM-CRd can accurately calculate 3-D fission source distribution and k eff in the JSI TRIGA in a matter of seconds, hence constituting a powerful analysis tool for online analysis of the reactor during experiments.

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.

The measurements of physical parameters of the TRIGA reactor and Nuclear power plant Krško (NEK) reactor cores have been in the past performed on hand of the neutron flux signal obtained from uncompensated ionization cells and by employment of the a digital meter of reactivity (DMR). At the TRIGA reactor only one ionization cell is currently used for flux measurements. During the insertion of one control rod the neutron flux distribution is significantly altered affecting the flux measurements of inserting different control rods. The problem is presently solved by assigning a correction factor to each control rod what introduces an additional uncertainty. In the present paper the implementation of four fission cells for reactivity measurements is presented. In this way determining the correct gamma background and its subtraction, performed by DMR algorithms, becomes less important as previously by using ionization chambers. The larger number of detectors also reduces the flux redistribution effects on the signal during individual control rod movements.

A pulse experiment model was validated in order to support future pulse experimental campaigns. All pulse experiments data was collected and are publicly available at http://trigapulse.ijs.si/. A comparison of the measured pulse physical parameters (maximal power, total released energy and full width at half maximum) and theoretical predictions (Fuchs-Hansen and the Nordheim-Fuchs models) was made.

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.

In 2016, the NATO Science for Peace and Security Programme funded research project ”Engineering Silicon Carbide for Border and Port Security” — E-SiCure was launched, its objective being the development of radiation-hard silicon carbide (SiC) based detectors of special nuclear materials (SNM), with the aim to enhance border and port security barriers. Detector prototypes based on SiC Schottky Barrier Diodes (SBDs) and neutron converter films were developed. This paper presents the results of a dedicated experimental testing campaign performed at the Jožef Stefan Institute (JSI) TRIGA reactor in which several SiC detector prototypes equipped with 10 B and 6 LiF converter films were irradiated in the Dry Chamber of the reactor. The obtained results demonstrate a clearly measurable neutron response, which varies linearly with the neutron flux. The measured particle spectra from the SiC detectors exhibit a clear structure, attributable to the nature and energy of secondary particles originating as reaction products from nuclear reactions involving 10 B and 6 Li isotopes. The determined sensitivity of the detectors, their active volume being 1 mm × 1 mm × 25 μ m , 1 mm × 1 mm × 69 μ m and 1 mm × 1 mm × 170 μ m , was of the order of 2 × 10−5 counts per second, per unit of neutron flux [counts s−1 per n cm−2s−1] (for neutron energies between 0 and 5 eV). Scaling the detection sensitivity by a factor of 1 0 5 , i.e. to an array with a surface of around 20 cm × 2 m, comparable to large B F 3 or 3 He detectors, would theoretically enable an overall sensitivity of around 2 counts s−1 per n cm−2s−1, which is already comparable to typical neutron sensitivity values of gas detectors, in the range from several to over 100 counts s−1 per n cm−2s−1. Due to its outstanding tolerance to harsh environments (including high temperatures and radiation fields) and superior electronic properties when compared to other semiconductors, SiC is a promising base material for the fabrication of solid-state detectors with stable and long life-time. Improvements in sensitivity combined with the capability of fabricating large modules (SiC arrays), could make SiC an important detection technology, applicable also in the context of border and port security barrier monitoring.

International audience; 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. On-line measurements were carried out using miniature ionization chambers 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, thermo-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. 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 ionization chambers from low to very high gamma-dose environment, such as in material testing reactors.

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.

A fission rate profile benchmark experiment has been performed at the Jožef Stefan Institute TRIGA Mark II reactor. The measurements were made using absolutely calibrated miniature fission chambers developed and manufactured by the Commissariat a l’Energie Atomique et aux Energies Alternatives. The aim of the paper is to describe the experimental set-up, fission rate measurements and to present the detailed Monte Carlo computational model of the TRIGA reactor, which was constructed with as used to compute absolute fission rate distributions in the core at a fixed control rod position, taking into account the detailed description of the experimental configuration. The paper focuses on the extensive evaluation of experimental and calculational uncertainties and biases following the International Reactor Physics Experiment Evaluation Project methodology. A comparison between the measured and computed absolute reaction rates concludes the paper, with the agreement being within one sigma standard uncertainty.

The main limitation of Computational Fluid Dynamics (CFD) lies in the unacceptable computational capability needed for performing accurate simulations for most real-time applications, and in the reduced accuracy of computationally more efficient low fidelity models for security-related applications. One example is the treatment of complex turbulent flows, where low-fidelity models introduce simplifications and source of uncertainties. A promising solution to improve accuracy is to use additional information about the actual flow field, such as experimental data taken on the system. The dynamic data-driven paradigm allows the direct incorporation of the knowledge coming from the measurements within the simulation, thus improving the model estimate itself by minimising its misfit with the data. In this work, the Kalman filter algorithm for data assimilation is combined with the segregated method for CFD modelling to get an integrated algorithm for resolving the incompressible Navier-Stokes equations along with the temperature one. The main novelty lies because such an integrated approach allows preserving mass conservation. This algorithm is herein validated regarding an instrumented cooling channel of the TRIGA Mark II reactor at the University of Pavia, using experimental data on temperature. The main takeaway of this validation is that, despite having only measurements on one quantity, also the prediction on velocity is improved regarding the standard segregated CFD algorithm. The prediction of the state also improves in domain locations where experimental data are not available. The increase in computational time is still lower than the one needed for a more accurate simulation.

The TRIGA research reactor ex-core nuclear instrumentation provides information on the reactor thermal power level, which is essential for the safe operation of the facility. In addition, exact knowledge of the reactor power level is necessary for the analysis of experiments and the normalization of reactor calculations. At the Jožef Stefan Institute TRIGA Mark II reactor the instrumentation is periodically calibrated using the calorimetric method. Though relatively simple and reliable it can be burdened with up to 30% uncertainty. In the paper a new calibration procedure using electrical heaters is presented, with which the heat capacity constant of the reactor pool is calculated to be C = 19.6 kW h / K ± 0.3 kW h / K and the uncertainty of the thermal power value is significantly reduced to approximately 2%.

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.

Monte Carlo N-Particle (MCNP) transport code accelerated by AutomateD VAriaNce reducTion Generator (ADVANTG) code was used to simulate the transport of delayed gamma particles emitted from irradiated TRIGA fuel. At the first step, this method was verified on a simple experiment where irradiated fuel was packed inside a transport cask. The calculated dose rate field in the vicinity of the transport cask was compared to the measurements. Due to an acceptable agreement, the method was later used to analyse Loss of Water Event (LOWE) for the JSI TRIGA research reactor main and spent fuel pool. Main results of the analysis show that in the case of LOWE for the reactor pool, the reactor building has to be evacuated since dose rates exceed 2 mSv/h. The highest dose rates are expected at the reactor platform. The radiation from the core is backscattered from the ceiling and affects the dose rates inside the whole reactor hall and control room. In the worst case scenario, the operating personnel can start with corrective actions 10 days after the last reactor operation, when dose rates at the reactor platform drop below 20 mSv/h. If LOWE occurs for the spent fuel pool, the reactor building has to be evacuated. However, the corrective actions by the operating personnel can be started immediately after the accident. This is due to the fact, that the fuel can only be transported into the spent fuel pool after a three-month cooling period inside the main reactor pool, resulting in a lower fuel activity.

For many practical applications, homogeneity of a field, whether neutron or electromagnetic, is of significance. Definitions of homogeneity or its calculation are diverse in the existing literature, however. This paper is an attempt to explore the concept of homogeneity and the ways it can be measured or calculated. The list of the measures that were considered includes: the ratio between the maximum and the minimum value, the ratio between the maximum difference and the mean value, the variance of the field and its generalization, and the mean and the variance of the field's gradient. These were applied to a realistic gamma-ray irradiation scenario, both by deriving a semi-analytical approximation and in a Monte Carlo numerical simulation. The irradiation arrangement included a hollow cylindrical irradiation channel and different numbers of spent fuel elements taken from a TRIGA reactor. By manipulating the geometry of this setup, fields of varying degrees of homogeneity were simulated and the homogeneity measures were calculated. By applying principal component analysis the maximum value to the weighted mean ratio was put forth as the most desirable measure. Finally, the application of this measure was demonstrated by choosing one of the irradiation configurations previously considered that produced a homogeneous gamma-ray field.

An overview of the TRIGLAV package, a tool for the computational analysis of TRIGA Mark II reactors based on lattice transport code WIMSD for unit cell cross section homogenization and four-group diffusion theory in 2D cylindrical coordinates, is presented. Possible unit cells include four types of fuel elements and different non-fuel elements, such as water, graphite, control rod, etc. Apart from the calculation of the main parameters, i.e. the multiplication factor and neutron flux distribution, secondary parameters, such as fuel power and temperature distribution, xenon defect, etc., are also calculated. The program has been compared with the Monte Carlo code MCNP5. Some possible applications of the TRIGLAV program are also outlined in the paper.

Modern Monte Carlo neutron transport codes offer many options for neutron flux and spectra calculations, however, they often lack the option to obtain the angular neutron flux in a region of the problem. The angular flux can also be obtained from deterministic programs, however, it includes biases due to discretization and other physical approximations. Therefore, a novel method for determining the angular neutron flux from the standard output of the MCNP is proposed in this paper. The method was also implemented as a set of Python libraries and tested in several examples. The results were then used to investigate the self-shielding effect in a realistic angular profile of the flux, i.e., the TRIGA research reactor.

The premise behind this research is the characterisation and integration of a unique detector system on board a submersible, remotely-operated vehicle (ROV) for the end purpose of fuel debris characterisation at Fukushima Daiichi. Currently, at Fukushima Daiichi. Whilst precise knowledge of the location of the core debris at Fukushima is not known it is commonly assumed that fuel has leaked through into the base of the pedestal below and it is suggested that it may have moved outside of the pedestal into the lower plenum. The flooding of the reactor floors immediately following the Fukushima accident adds an extra element of complexity for the detection system requiring it to be submersible and to hold any detector system in water-tight confinement. The research presented here focusses on the use of a CeBr 3 inorganic scintillator detector with a unique configuration of an in-built HV supply for ease of integration within an ROV in a submerged environment. The detector has been tested in several environments: small wave tank for source identification and a TRIGA reactor and a 60Co irradiator. It is hoped that the CeBr 3 detector will constitute one component of an on-board detector payload to determine the suitability for the localisation and identification of fuel debris inside the cores at Fukushima.

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.

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 currently no fusion reactors fusing deuterium and tritium are capable to perform water activation experiments with sufficient accuracy, fission research reactors present an opportunity to study the effects of activated water decay that can be extrapolated to fusion reactors. An irradiation system using activated cooling water as the source of energetic gamma rays is proposed at the Jožef Stefan Institute TRIGA Mark II research reactor. The conceptual designs utilizing radial piercing port for water activation is presented and analysed in the paper. At reactor full power and the maximum flow rate of 1 l/s the 16 N decay rate in the irradiation facility was calculated to be around 1.7 × 10 8 s−1 which will produce ambient dose equivalent rates up to 2 mSv/h. The proposed irradiation facility will allow shielding characterization at high gamma energies. An analysis of shielding experiment with fusion relevant materials, such as tungsten, Eurofer, SS 304L, etc., which can be implemented in the proposed water activation facility was performed. From all analysed material SS 304L and Eurofer are the best candidate materials for shielding around primary cooling water in fusion reactors to reduce additional nuclear heating to superconducting coils and other important tokamak components. From the analysis it can be concluded that around 30 cm of SS 304L or Eurofer is needed to reduce the high energy gamma ray flux by half.

In this study, analysis of the complete operational history of the “Jožef Stefan” Institute (JSI) TRIGA reactor was performed. Reactor power changes, core configurations and weekly excess reactivity measurements were analysed to obtain the needed data for fuel burnup calculations. More than 50 years of reactor operation was simulated using deterministic code TRIGLAV and stochastic code Serpent-2. The calculated core reactivities are in good agreement compared with the excess reactivity measurements. Code-to-code comparison is presented. Clear agreement is observed when comparing changes in core excess reactivity, and discrepancies are observed in the comparison of individual fuel element burnup and its isotopic composition. The Serpent-2 results are in better agreement with the measurements compared to the TRIGLAV code; nevertheless, a conclusion can be made that the TRIGLAV code is viable for TRIGA fuel management and burnup analysis. A three-dimensional (3D) burnup study was conducted, where individual fuel elements were further divided into multiple angular and axial depletion zones. Notable burnup effects were observed, and an explanation using surrounding water distance is presented.

Interest on development of computational tools for calculations of radiation field due to decay of radioactive activation and fission products (delayed radiation) has increased in recent years, mainly due to requirements on shutdown dose rate for the future ITER fusion reactor and material testing reactors (MTR), where gamma heating values of several 10 Wg − 1 can be reached in reactor structural components during reactor steady state operation and for characterization of research reactor irradiation facilities. Previous experimental work shows a roughly 30% contribution of delayed gamma rays to the total gamma ray flux. An extensive experimental campaign at the JSI TRIGA reactor has been carried out using multiple ionization and fission chamber detectors for gamma and neutron field characterization. A series of reactor power steps was performed with said detectors in different irradiation positions during steady reactor power and after shutdown. The reactor and the used detectors, along with irradiation power steps have been modelled in detail using the JSIR2S code system framework to compare measured and calculated fluxes and dose rates. Simulated results are in good agreement with measurements and with experimentally obtained values of delayed gamma fractions from previous work.

In 2016, the ”E-SiCure” project (standing for Engineering Silicon Carbide for Border and Port Security), funded by the NATO Science for Peace and Security Programme, was launched. The main objective is to combine theoretical, experimental and applied research towards the development of radiation-hard SiC-based detectors of special nuclear materials (SNM), and by that way, to enhance border and port security barriers. Along the plan, material modification processes are employed firstly to study, and secondly to manipulate the most severe electrically active defects (which trap or annihilate free charge carriers), by specific ion implantation and defect engineering. This paper gives an overview of the experimental activities performed at the JSI TRIGA reactor in the framework of the E-SiCure project. Initial activities were aimed at obtaining information on the radiation hardness of SiC and at the study of the energy levels of the defects induced by neutron irradiation. Several Schottky barrier diodes were fabricated out of nitrogen-doped epitaxial grown 4H-SiC, and irradiated under Cd filters in the PT irradiation channel in the JSI TRIGA reactor with varying neutron fluence levels. Neutron-induced defects in the material were studied using temperature dependent current-voltage (I-V), capacitance-voltage (C-V) and Deep-Level Transient Spectroscopy (DLTS) measurements. Our prototype neutron detectors are configured as 4H-SiC-based Schottky barrier diodes for detection of secondary charged particles (tritons, alphas and lithium atoms) which are result of thermal neutron conversion process in 10B and 6LiF layers above the surface of the 4H-SiC diodes. For field testing of neutron detectors using a broad beam of reactor neutrons we designed a standalone prototype detection system consisting of a preamplifier, shaping amplifier and a multichannel analyser operated by a laptop computer. The reverse bias for the detector diode and the power to electronic system are provided by a standalone battery-powered voltage source. The detector functionality was established through measurements using an 241Am alpha particle source. Two dedicated experimental campaigns were performed at the JSI TRIGA reactor. The registered pulse height spectra from the detectors, using both 10B and 6LiF neutron converting layers, clearly demonstrated the neutron detection abilities of the SiC detector prototypes.

Neutron and gamma dose rate calculations were carried out around horizontal beam tube no. 5 at the Jožef Stefan Institute (JSI) TRIGA Mark II research reactor. Results were compared to the experimental measurements in order to verify the computation model. In addition, another set of calculations and measurements was carried out, where an additional shield made out of concrete and paraffin was installed. With that configuration, we were able to study neutron and gamma scattering.

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.

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.

Within the bilateral project between the CEA Cadarache and the Jožef Stefan Institute (JSI) a wide variety of measurements using multiple fission chambers simultaneously inside the reactor core were performed. The fission rate axial profiles were measured at different positions in the reactor core and at different control rod configurations. A relative comparison of the calculated fission rates using the MCNP code and the measured fission rates was performed. In general the agreement between the measurements and calculations is good, with the deviations within the uncertainties. For better observation and understanding of the neutron flux redistribution due to the control rod movement, the neutron flux and fission rate had been calculated through the entire reactor core for different control rod configurations. The detector position with minimum signal variations due to the regulating and compensating control rod movement during normal operation was determined. The minimum variation is optimal in case we want to reliably determine the reactor power without influence of the regulating and compensating control rod positions.

The work presented in this thesis deals with the characterization of gamma field inside a nuclear reactor by experiments and computational modelling. In the first part of the thesis an outline of the nuclear with neutrons and neutron transport. A description of high energy photon and electron reactions and importance of their coupling for accurate calculations of energy deposition. Particle transport equations are presented with emphasis on deriving adjoint operators used for variance reduction of Monte Carlo particle transport codes. Characterization of gamma radiation field using Monte Carlo transport codes only takes into account prompt gamma generation from fission, inelastic scattering and prompt (n,gamma) reactions. Previous evaluations suggest a roughly 30 % underestimation compared to measurements. A JSIR2S code package for delayed radiation field calculations has been developed and validated by numerous experiments. Characterization of neutron and prompt gamma radiation field inside the JSI TRIGA reactor core irradiation facilities was performed using the kerma approximation. The computational model was later expanded and the criticality source term translated to a fixed source for calculations of variance reduction parameters. The methodology has been validated by experiments, showing good agreement for neutrons, while underestimating the gamma field due to neglecting delayed radiation field. Several experimental campaigns were performed at JSI TRIGA reactor using fission and ionization chamber and Thermoluminescent dosimeters. An experimental procedure for estimation of the delayed gamma fraction was developed. Validation of the JSIR2S was performed on the above mentioned measurements, showing agreement within the uncertainty. use case on using the JSIR2S for calibration of semiconductor detectors in the JSI TRIGA reactor after reactor shut-down is described. The JSIR2S code package is also applied to shut-down dose rate calculations in fusion problems showing good agreement with experiments and similar two-step and single-step methodology codes for delayed radiation field characterization.

A series of fission rate profile measurements with miniature fission chambers, developed by the Commisariat á l’énergie atomique et auxénergies alternatives, were performed at the Jožef Stefan Institute’s TRIGA research reactor. Two types of fission chambers with different fissionable coating (235U and 238U) were used to perform axial fission rate profile measurements at various radial positions and several control rod configurations. The experimental campaign was supported by an extensive set of computations, based on a validated Monte Carlo computational model of the TRIGA reactor. The computing effort included neutron transport calculations to support the planning and design of the experiments as well as calculations to aid the evaluation of experimental and computational uncertainties and major biases. The evaluation of uncertainties was performed by employing various types of sensitivity analyses such as experimental parameter perturbation and core reaction rate gradient calculations. It has been found that the experimental uncertainty of the measurements is sufficiently low, i.e. the total relative fission rate uncertainty being approximately 5 %, in order for the experiments to serve as benchmark experiments for validation of fission rate profiles. The effect of the neutron flux redistribution due to the control rod movement was studied by performing measurements and calculations of fission rates and fission chamber responses in different axial and radial positions at different control rod configurations. It was confirmed that the control rod movement affects the position of the maximum in the axial fission rate distribution, as well as the height of the local maxima. The optimal detector position, in which the redistributions would have minimum effect on its signal, was determined.

Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 1015 neutrons/cm2 in irradiation time of 20 h.

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 5.5×103Svh-1 to 6×106Svh-1, silicon dose equivalent from 6×102Gy/hsi to 3×105Gy/hsi, and neutron air kerma from 4.3×103Gyh-1 to 2×105Gyh-1. Ratio of fast (1MeV

The aim of the current bilateral project between CEA Cadarache and JSI is to improve the accuracy of the online thermal power monitoring at the JSI TRIGA reactor. Simultaneously, a new wide range multichannel acquisition system for fission chambers, recently developed by CEA, is tested. In the paper, calculational and experimental power calibration methods are described. The focus is on use of multiple detectors in combination with pre-calculated and pre-measured control-rod-position-dependent correction factors to improve the reactor power reading. The system will be implemented and tested at the JSI TRIGA reactor in 2014.

Neutron and gamma flux levels are key parameters in nuclear research reactors. In Material Testing Reactors, such as the future Jules Horowitz Reactor, under construction at the French Alternative Energies and Atomic Energy Commission (CEA Cadarache, France), the expected gamma flux levels are very high (nuclear heating is of the order of 20 W/g at 100 MWth). As gamma rays deposit their energy in the reactor structures and structural materials it is important to take them into account when designing irradiation devices. There are only a few sensors which allow measurements of the nuclear heating [12]; a recent development at the CEA Cadarache allows measurements of the gamma flux using a miniature ionization chamber (MIC) [3]. The measured MIC response is often compared with calculation using modern Monte Carlo (MC) neutron and photon transport codes, such as TRIPOLI-4 and MCNP6. In these calculations only the production of prompt gamma rays in the reactor is usually modelled thus neglecting the delayed gamma rays. Hence calculations and measurements are usually in better accordance for the neutron flux than for the gamma flux. In this paper we study the contribution of delayed gamma rays to the total MIC signal in order to estimate the systematic error in gamma flux MC calculations. In order to experimentally determine the delayed gamma flux contributions to the MIC response, we performed gamma flux measurements with CEA developed MIC at three different research reactors: the OSIRIS reactor (MTR — 70 MWth at CEA Saclay, France), the TRIGA MARK II reactor (TRIGA — 250 kWth at the Jozef Stefan Institute, Slovenia) and the MARIA reactor (MTR — 30 MWth at the National Center for Nuclear Research, Poland). In order to experimentally assess the delayed gamma flux contribution to the total gamma flux, several reactor shut down (scram) experiments were performed specifically for the purpose of the measurements. Results show that on average about 30 % of the MIC signal is due to the delayed gamma rays. In this paper we describe experiments in each of the three reactors and how we estimate delayed gamma rays with MIC measurements. The results and perspectives are discussed.

In this paper we investigate the influence of various up-to-date nuclear data libraries, such as ENDF/B-VI.6, ENDF/B-VII.0 and JEFF 3.1, on the multiplication factor of the TRIGA benchmark with fuel made of enriched uranium and zirconium hydride and SB light-water reactor benchmarks with fuel made of fissile material in zirconium matrix. The calculations are performed with the Monte Carlo computer code MCNP. Differences of ∼600 pcm in k eff are observed for the benchmark model of the TRIGA reactor, while there are practically no differences in the k inf of the fuel. Therefore, an investigation is performed also for hypothetical homogeneous and heterogeneous systems with different leakage. The uncertainty analysis shows that the most important contributors to the difference in k eff are the Zr isotopes (especially 90 Zr and 91 Zr) and thermal scattering data for H and Zr in ZrH. As the differences in k eff due to the use of different cross section libraries are relatively large, there is certainly a need for a review of the evaluated cross section data of the zirconium isotopes.

International audience; The characterization of experimental locations of a research nuclear reactor implies the determination of neutron and photon flux levels within, with the best achievable accuracy. This study focuses on the photon flux problematic. In nuclear reactors, photon fluxes are commonly calculated by Monte Carlo simulations but rarely measured on-line. In this context, experiments were conducted with a miniature gas ionization chamber (MIC) based on miniature fission chamber mechanical parts, recently developed by the CEA (French Atomic Energy and Alternative Energies Commission) irradiated in the core of the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. The aim of the study was to compare the measured MIC currents with calculated currents based on simulations with the MCNP6 code. A discrepancy of around 50 per cent was observed between the measured and the calculated currents; in the latter taking into consideration only the prompt photon field. Further experimental measurements of MIC currents following reactor SCRAMs (reactor shutdown with rapid insertions of control rods) provides evidence that over 30 per cent of the total measured signal is due to the delayed photon field, originating from fission and activation products, which are untreated in the calculations. In the comparison between the measured and calculated values, these findings imply an overall discrepancy of less than 20 per cent of the total signal which is still unexplained.

Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software.

In order to verify and validate the computational methods for neutron flux calculation in TRIGA research reactor calculations, a series of experiments has been performed. The neutron activation method was used to verify the calculated neutron flux distribution in the TRIGA reactor. Aluminium (99.9 wt%)–Gold (0.1 wt%) foils (disks of 5 mm diameter and 0.2 mm thick) were irradiated in 33 locations; 6 in the core and 27 in the carrousel facility in the reflector. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and experimental normalized reaction rates in the core are in very good agreement for both isotopes indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux and reaction rate distribution in the reactor core. In the reflector however, the accuracy of the epithermal and thermal neutron flux distribution and attenuation is lower, mainly due to lack of information about the material properties of the graphite reflector surrounding the core, but the differences between measurements and calculations are within 10%. Since our computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of research reactor utilization.

Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jozef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104 Rh were made and amount to 6.657 0.18 and (6.617 0.12) � 10 -2 , respectively.

Modern Monte Carlo transport codes in combination with fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, β eff , and mean neutron generation time, Λ . We calculate β eff and Λ for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It is observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. β eff varies for 12 wt.% uranium standard fuel with 20% enrichment from 0.0080 for a small core (43 fuel rods) to 0.0070 for a full core (90 fuel rods). It is found that calculated value of β eff strongly depends also on the nuclear data set used in calculations. The prompt neutron lifetime mainly depends on the amount (due to either content or enrichment) of 235 U in the fuel as it is approximately inversely proportional to the average absorption cross-section. It varies from 28 μs for 30 wt.% uranium content fuelled core to 48 μs for 8.5 wt.% uranium content LEU fuelled core. Description of the calculation method and detailed results are presented in the paper.

Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

Modern Monte–Carlo codes (e.g. MCNP) allow calculation of the power density distribution in 3D geometry assuming detailed geometry without unit-cell homogenization. The power density distribution (and its maximum value—the peaking) can be calculated ‘point-wise’ with the resolution of approximately 1 mm. Results of the detailed power density distribution calculated by MCNP are presented for 250 kW TRIGA Mark II reactor, assuming various realistic and hypothetical core-loading patterns with focus on the mixed cores. Combinations of 8.5 w/o, 12 w/o, 20 w/o and 30 w/o low-enriched (20%) fresh TRIGA fuel rods are systematically treated in the mixed cores. The power peaking factor value and position strongly depends on the core configuration. Power peakings are usually found in fuel rods with higher uranium content especially if they are inserted near the core centre. The results are conservative and can be applied in planning realistic mixed core-loading patterns.

The main aim of this paper is the development and validation of a 3D computational model of TRIGA research reactor using Serpent 2 code. The calculated parameters were compared to the experimental results and to calculations performed with the MCNP code. The results show that the calculated normalized reaction rates and flux distribution within the core are in good agreement with MCNP and experiment, while in the reflector the flux distribution differ up to 3% from the measurements.

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system.

International audience; In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results.

International audience; Significant efforts have been made over the lastfew years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g.miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. Thesedata are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations withsatisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm$^3$). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position,the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presentedin the paper, the injudicious use of special ptions aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from differenttypes of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome.The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values.