Low Power Research Reactor Research Articles

A comprehensive framework for identifying safety-related and ageing sensitive sscs in low power nuclear research reactor

This work aims to establish a novel approach for the selection of structures, systems, and components (SSCs) important to safety and vulnerable to ageing degradation for low-power research reactors. For this purpose, the framework developed combines the Preliminary Risk Analysis (PRA) technique to categorize SSCs into four different safety classes, and Ageing Failure Mode and Effect Analysis (AFMEA) technique to find out ageing failure modes related to a specific SSC and evaluate them according to the Risk Priory Number (RPN). The intersection of the PRA and AFMEA outcomes then yields a list of SSCs that are both important to safety and strongly ageing sensitive, making it possible to optimize the selection process of SSCs to be prioritized in further preventive maintenance, periodic testing, and inspection activities. The present paper also conducts a case study to apply the proposed methodology to the Brazilian IPR-R1 TRIGA nuclear research reactor. The selected SSCs were then both AL and SS fuel rods from the reactor core, stainless steel pipe, thermocouples, pressure indicators, and water level indicators from the primary circuit, and power monitoring channels, radiation monitors, and cooling water temperature channels from the controlling system. Moreover, the approach was also able to identify SSCs that are important from a safety perspective but are not ageing sensitive (reactor tank and control rods) and components that are not important to safety but are strongly ageing sensitive (pipe filters). These findings may further serve as input for the comprehensive safety assessment conducted during the Periodic Safety Review of the facility.

A computational- experimental model of reactor kinetic for investigating the linearity response of in-core neutron detectors of a low power research reactor

In this study, the linear response of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) is investigated through experiments and calculations at different powers of the reactor. The linearity of the response of the main neutron detector in the MNSR is important due to its safety functions in reactor power control. Additionally, determining the linear response region of this detector presents a linear neutron dosimetry field even outside of the reactor core, i.e. at the outlet of external beam tubes. Moreover, it is very difficult to remove and calibrate this neutron detector that has been installed in the reactor core for a long time. For this purpose, an alternative method for measuring the linearity of the response of a neutron detector is extended by investigating the kinetic behavior of reactor. The aim of this method is to calculate the delayed neutrons population in the reactor core and compare with the FC detector response. In this method, the numerical solution method is used to solve the point kinetic equation with 15 delayed neutron groups (6 delayed neutron groups and 9 photo-neutron groups) after inserting a large negative reactivity into the MNSR core by considering temperature and xenon reactivity feedbacks during a coupling procedure. To obtain the xenon reactivity feedback, xenon concentration is calculated by coupling the 135Xe and 135I production equations and the εpPFnlPTnl parameter is obtained by an experimental method. Furthermore, the value of fission cross section (Σf) is calculated using a validated model in the Dragon code. The large reactivities are inserted at two initial power of 150 W and 30 kW to investigate the response of the main FC neutron detector in a cold xenon-free condition and by considering temperature and xenon reactivity feedbacks, respectively. The results demonstrate that the response of the MNSR FC detector is linear in the neutron flux range between 5 × 107 to 1 × 1012 n.cm−2.s−1, which is equivalent to reactor power of 1.5 W to 30 kW. This innovative method can be applied to calibrate the neutron detectors in other research and power reactors.

Temperature feedback reactivity analysis for LEU-Fuelled SLOWPOKE-2 research reactor using DRAGON5 and DONJON5 codes

The purpose of this study is to provide a new deterministic model for the SLOWPOKE-2 nuclear research reactor at École Polytechnique de Montréal (EPM) with LEU (Low Enriched Uranium) core. Using the latest release of the code system DRAGON5 and DONJON5 and the cross-section data library ENDFB.VII rel.1 evaluation, the developed model is applied to simulate the neutronic behavior of the SLOWPOKE-2 research reactor. We studied the separate temperature effects of the main components of the core (i.e., fuel, coolant/moderator, beryllium reflector, and water reflector). The contribution of different physical phenomena to the RTC was assessed. The temperature reactivity feedback calculated using the deterministic approach based on the DRAGON5 and DONJON5 code system using the ENDF/B-VII.1 evaluated nuclear data library produced in the WIMD-D4 format is in good agreement. Therefore, this work proves the capability of DRAGON5 and DONJON5 codes, normally used for power reactors, to reliably simulate a low-power research reactor.

Investigation of Buddhist and Bon votive statues at the Very low power reactor VR-1

The nuclear analytical techniques are excellent tools for the investigation in various research disciplines, i.e. cultural heritage preservation, archaeology or anthropology. Most of these radioanalytical techniques require sufficient sources of radiation. So far, neutron imaging has mostly been used at high-power neutron sources, i.e. high-power research reactors. On the other hand, small nuclear research reactors can provide more options for investigations in multidisciplinary research. Since the development of the NIFFLER neutron imaging facility in 2020, neutron imaging has been one of the nuclear analytical methods used at the very low-power research reactor VR-1 of the Czech Technical University in Prague. This paper deals with the investigation of the internal structures of Buddhist and Bon votive statues from Central Asia at the Training reactor VR-1. The experimental investigation was conducted to explore the possibilities of using neutron imaging for cultural heritage research at very low-power research reactors such as the VR-1 reactor. The results of these unique experiments clearly showed the potential of the VR-1 reactor or possibly other very low-power research reactors in neutron imaging for multidisciplinary research, which can provide important data for anthropologists, archaeologists or experts in cultural heritage preservation.

High-Resolution Powder Diffractometer Facility (HRPDF) for Low and Medium Power Research Reactor

Abstract: The present paper suggests a design of a high-resolution powder diffractometer facility that fits the low and medium power research reactors. The design choice guarantees the acceptable cost of a high-standard facility. Some components can be created in-house, while the design also includes detailed information about the commercially available components. It is highly recommended to carry out Monte Carlo simulations of realistic expected luminosity and resolution parameters of the finally designed and approved instrument in advance. Monte Carlo simulations can eventually show which components should be modified in order to achieve the expected resolution and luminosity parameters of the facility. Keywords: High-resolution powder diffractometer, Design, Construction, Low and medium research reactors.

Long-time irradiation effect on corrosion behavior of aluminum alloy in pool water of low-power research reactor

This study conducted an evaluation of the corrosion behavior of an aluminum alloy utilized in the Isfahan Miniature Neutron Source Reactor (MNSR). The component analyzed, dry channel (DC), had been exposed to radiation for 12 years in a water environment within the reactor pool since its installation. To determine the effect of radiation on the corrosion of the LT-21 aluminum alloy used in the DC, different parts of the pipe were sampled and various tests were performed. These tests included mechanical strengths (impact, and micro-hardening), XRD, TEM, SEM–EDS, and potentiodynamic polarization (PDP). The parameters measured included corrosion potential, corrosion rate, changes in microscopic structure, and mechanical properties of the aluminum alloy along the entire length of the DC. The neutron and gamma dose distribution along the height of the DC, which was 540 cm, was calculated to determine the correlation between the dose distribution and observed corrosion. The study found that the corrosion mechanisms were complex and resulted from the simultaneous presence of the DC in the pool water and radiation from the reactor core. The observed results are presented and discussed in this study.

A freely distributable professional computed tomography system using NICOS

Based on the control system of the ANTARES neutron imaging system at FRM II, we have designed a portable, but professional neutron computed tomography system using the same TANGO device servers and the instrument control system NICOS. We have created a minimalistic setup to control a computed tomography setup using a single control computer that directly connects to a camera detector and interfaces a Raspberry Pi computer with motor control board via a local network. The setup has the full capability of a large instrument controller and can control arbitrary hardware (like professional motor controllers) for which hundreds of free drivers are available. Full schematics for the controller as well as print files for 3D printers for the detector housing and disk or virtual machine images are available for free to allow for the quick implementation of a low-cost, but professional neutron CT system even at very low power research reactors.

Development of a Neutron Imaging Facility at the Very Low Power Reactor VR-1

The paper describes the design and construction of a new neutron imaging facility at the very low-power research reactor VR-1. Neutron imaging has been under development at the VR-1 reactor as a new possible field of utilisation since 2020. The paper describes the main parts of the NIFFLER – “Neutron Imaging Facility for Learning and Research”. The detector system was designed in collaboration with the Technical University of Munich. The radial beam line modifications include different collimation ratios L/D, additional moderator configurations and the basic concept and design of the instrumentation. The paper also includes calculations of thermal neutron flux at the detector position for different radial beamline configurations. The results achieved at the training reactor VR-1 showed the potential of using neutron imaging at other low and very-low-power research reactors worldwide.

Design, construction, and characterization of a Prompt Gamma Neutron Activation Analysis (PGNAA) system at Isfahan MNSR

In this research, a prompt gamma neutron activation analysis (PGNAA) system is designed and constructed based on the use of a low power research reactor. For this purpose, despite the fact that this reactor did not include beam tubes, a thermal neutron beam line is installed inside the reactor tank. The extraction of the beam line from inside the tank made it possible to provide the neutron flux from the order of 106 n.cm-2.s-1. Also, because the beam line is installed in a tangential position to the reactor core, its gamma level has been minimized. Also, a suitable radiation shield is considered for the detector to minimize the background radiation and prevent radiation damage to the detector. Calculations and measurements are done in order to characterize this system, as well as spectrometry of several samples. The results of evaluations and experiments show that this system is suitable for performing PGNAA.

Neutron capture-induced nuclear recoils as background for CEνNS measurements at reactors

Nuclear reactors represent a promising neutrino source for $\mathrm{CE}\ensuremath{\nu}\mathrm{NS}$ (coherent-elastic neutrino-nucleus scattering) searches. However, reactor sites also come with high ambient neutron flux. Neutron capture-induced nuclear recoils can create a spectrum that strongly overlaps the $\mathrm{CE}\ensuremath{\nu}\mathrm{NS}$ signal for recoils $\ensuremath{\lesssim}100\text{ }\text{ }\mathrm{eV}$ for nuclear reactor measurements in silicon or germanium detectors. This background can be particularly critical for low-power research reactors providing a moderate neutrino flux. In this work we quantify the impact of this background and show that, for a measurement 10 m from a 1 MW reactor, the effective thermal neutron flux should be kept below $\ensuremath{\sim}7\ifmmode\times\else\texttimes\fi{}{10}^{\ensuremath{-}4}\text{ }\text{ }\mathrm{n}/{\mathrm{cm}}^{2}\text{ }\mathrm{s}$ so that the $\mathrm{CE}\ensuremath{\nu}\mathrm{NS}$ events can be measured at least at a $5\ensuremath{\sigma}$ level with germanium detectors in 100 kg yr exposure time. This flux corresponds to 60% of the sea-level flux but needs to be achieved in a nominally high-flux (reactor) environment. Improved detector resolution can help the measurements, but the thermal flux is the key parameter for the sensitivity of the experiment. For silicon detectors, the constraint is even stronger and thermal neutron fluxes must be near an order of magnitude lower. This constraint highlights the need of an effective thermal neutron mitigation strategy for future low threshold $\mathrm{CE}\ensuremath{\nu}\mathrm{NS}$ searches. In particular, the neutron capture-induced background can be efficiently reduced by active veto systems tagging the deexcitation gamma following the capture.

A pile noise experiment in the reference core of the nuclear research reactor LR-0

For safety reasons, it is important to simulate precisely the kinetic behavior of light water nuclear reactors (LWR) in nominal or incidental situations. In the point kinetic approximation, a few integral parameters allow describing the reactor’s behavior. It is of interest to measure those parameters, such as the delayed neutron fraction and delayed neutron abundances, in low power research reactors loaded with benchmarked fuel configurations to compare with simulations results. In this context, Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA) and Centrum Výzkumu Řež (CVŘ) have been collaborating towards measuring the delayed neutron fraction of LR-0, a VVER-type research reactor operated by CVŘ. This paper presents the first estimation of the delayed neutron fraction of the reactor loaded with six fuel assemblies, arranged in a small hexagonal lattice with a center cavity accessible for instrumentation. This configuration was chosen because it was very close to the so-called reference configuration with seven fuel assemblies, the one that was benchmarked in the NEA IRPheP Handbook. The measurement was performed in July 2021 using the CEA-developed acquisition system SPECTRON dedicated to pile noise experiments. Pile noise refer to a set of techniques which focus on the correlations induced by the fission process in the signals of neutron detectors. The experimental campaign was designed to meet two objectives: first to provide an estimation of the delayed neutron fraction; second to study the impact of the location of the neutron detectors on the quality of the results. Two fission chambers were used, one was fixed and located in the core center and the second was placed in various locations in the reflector. Several configurations were tested with the mobile detector progressively moved up to 20 cm from the closest fuel element. The reactor fission rate was obtained thanks to gamma spectrometry on metal foils (gold, nickel and iron). Simulations with TRIPOLI-4® and JEFF33 libraries were performed to convert the saturated activities into fission rates. The delayed neutron fraction was estimated at 782 ± 10 pcm and the generation time was measured at 39.1 ± 0.5 μs. The uncertainty balance is as follows: 2 % associated to the Diven factor and 1.5 % associated to the core’s integral fission rate and 0.5% for the statistical uncertainty. The reproducibility of the experiments was found very good, with a negligible dispersion amongst experiments (0.2 %). These results are valuable complements to the characterization of LR-0 reference neutron benchmark field.

Development and characterization of a large thermal neutron beam for neutron radiography at Isfahan MNSR

The feasibility study of performing neutron radiography (NR) at the Isfahan Miniature Neutron Source Reactor (MNSR) was carried out in the previous work that was including calculations and experiments in support of the implementation of a prototype NR beam tube. Results were shown the obtained neutron beam from that prototype beam tube had appropriate conditions for NR and it is possible to obtain a large NR beam at this low power research reactor. The main disadvantage of the prototype beam tube was the small size of the beam port with diameter of 5 cm. In this study, a large diameter beam tube is designed and installed at the reactor. This beam tube has a telescopic divergent shape and is installed in the west side of the reactor pool. The neutron beam parameters and image quality of the obtained beam from this beam tube are measured based on ASTM-E545. Results showed that this neutron beam has good quality for NR experiments over a beam diameter of about 25 cm. Also, a simple and practical system, for fill and drain water to the beam tube, is designed and incorporated as a beam shutter in order to maintain the radiation safety.

New Neutron Imaging Facility NIFFLER at Very Low Power Reactor VR-1.

The paper describes the construction of the neutron imaging facility at the very low-power research reactor VR-1. The training reactor VR-1 is operated by the Czech Technical University in Prague, Czech Republic. It is mainly used for the education of students in the field of nuclear engineering as well as for the training of professionals. Neutron imaging is the new field of VR-1 reactor utilisation currently under development. Extremely low reactor power at the level of 100 W brought many challenges that were necessary to overcome to build and commission a sustainable neutron radiography facility. The paper describes the reactor's neutron flux verification and the basic concept and design of the neutron imaging instrumentation. The first experimental results were mainly dedicated to testing the detection system for different radial beam port configurations, different L/D ratios, and different exposure times. Preliminary results of neutron radiography and tomography measurements at VR-1 clearly showed the potential of using neutron imaging in low-power reactors such as the VR-1 reactor.

Experimental and analytical determination of irradiated control rod worth of a low power research reactor

The NIRR-1 has a single control rod that performs the function of shutting down, shimming and bringing the reactor to power. Assessment of the efficiency of control rod is very important especially after modification. In this work, the efficiency of the control rod prior to the removal of cadmium rabbit in one of the irradiation channels of NIRR-1 was assessed using experimental and analytical methods. The measured and calculated safety parameters of the control rod such as control rod worth, shutdown margin and safety reactivity factor were found to be (6.41 mk and 6.37 mk), (2.52 mk and 2.48 mk) and (1.65 and 1.64) respectively, which meet safety requirements of NIRR-1 in the SAR. The result from analytical methods was found to be within maximum of 3σ of the measured value, indicating that the analytical method can be used to obtain a good approximation of control rod worth. Results obtained in both methods indicate that it was safe to continue to use the control rod after removal of the cadmium rabbit.

Determination of urgent protective action zone in Tehran research reactors-associated accidents

In accidents leading to the release of radioactive materials into the environment, it is of great importance to determine the urgent protective action zone (UPZ) in the emergency preparedness phase. The present study was conducted to determine the radius of this area for a low-power research reactor, taking into account the worst conditions due to a hypothetical partial core melting accident. The UPZ was calculated based on the IAEA recommended generic criteria for the public and emergency workers during an emergency exposure. Furthermore, according to the simulated predictable operational intervention level (OIL) in the emergency response phase, a new criterion for exposure via ground shine (OIL1γ) was proposed in UPZ determination for the individuals in the research reactor-associated accidents.

Neutronic performance of the hybrid-low power research reactor for education and neutron applications

Korea Atomic Energy Research Institute (KAERI) has designed a second-generation Hybrid-Low Power Research Reactor (H-LPRR), which can be used as a critical assembly and conventional research reactor simultaneously. Particularly, the fuel assemblies and the reflectors of the H-LPRR are designed to stand alone without any supporting equipment and to replace each other. Its eight irradiation holes are located on the edge of the core and are applicable to neutron activation analysis and radioisotope production. In this study, some of the important nuclear characteristics are provided such as the control absorber rod worth, shutdown margin, safety reactivity factor, which are calculated by using the MCNP6.1 code with the ENDF/B-VII.0 library. Following a reactivity insertion accident, the power trend as a function of the time was also analyzed using the RELAP5/MOD3.3 code. The main safety parameters of the H-LPRR and MNSR were compared to quantitatively evaluate the neutronics performance of the H-LPRR. From these results, it can be confirmed that the H-LPRR possesses high performance in terms of nuclear design and safety analysis, and it can be a suitable option for countries considering replacement of existing and aging low power research reactors.

Neutronic feasibility study of using a multipurpose MNSR for BNCT, NR, and NAA

In this study, the conceptual design of a multipurpose research reactor for boron neutron capture therapy (BNCT), neutron radiography (NR) and neutron activation analysis (NAA) applications has been performed. Specifically, a suitable epithermal neutron flux (φEpi) of about 1×109 (n.cm−2.s−1) for BNCT and high quality thermal neutron flux (φTh) of above 1×106 (n.cm−2.s−1)for NR are carried out based on our LEU reactor core designing. The UO2 fuel with density of 10.5 g/cm3 and enrichment of 12.4% is applied as an appropriate LEU fuel. The reactor safety is assured by designing the safety control rod system with two banks. A fission converter facility (FCF) consists of 19 fuel rods of UO2 is used to increase φEpi for BNCT treatment. Furthermore, a unique thermal column of heavy water is used to increase φTh for NR purpose. Four internal irradiation sites, two external irradiation sites, and one large thermal column irradiation site are considered at the reactor which can be used for NAA application. The results of neutronic calculations show that the reactor meets the neutronic design limits for a low power research reactor.

Neutron activation analysis of Tibetan traditional medicinal pills at the VR-1 training reactor

The nuclear analytical methods provide the research techniques usually used in hard sciences such as physics, chemistry, engineering and technology, or natural sciences and technology, but it is very rare to use them in social sciences and humanities. The Czech Technical University (CTU) in Prague operates the VR-1 training reactor primarily focused on training students, nuclear engineering, reactor physics, and research; however, the utilization of this nuclear reactor has also been extended to humanities in recent years. In this direction, the research activities carried out at the reactor include the investigation of historical artefacts and testing of pharmaceutical and food industry samples. This contribution deals with the detailed study of traditional Tibetan medicinal pills by means of instrumental neutron activation analysis (NAA) at the VR-1 nuclear reactor. Three various kinds of Tibetan pills were irradiated by thermal neutrons (ϕ=2×109cm−2s−1) in the experimental vertical channel at maximum reactor power (80 W), and irradiated samples were analysed using the semiconductor HPGe detector; saturated activities were obtained by means of the nuclear gamma-ray spectrometry, and subsequently the composition of the pills was determined (qualitative and quantitative analysis). The presence of Au, K, Fe, Na, As, Sr, Sb, and Hg was revealed in studied samples. The results presented in this paper show clearly that the low-power research reactor VR-1 is excellent tool for the neutron activation analysis experiments.

Neutron activation analysis of Tibetan traditional medicinal pills at the VR-1 training reactor

The nuclear analytical methods provide the research techniques usually used in hard sciences such as physics, chemistry, engineering and technology, or natural sciences and technology, but it is very rare to use them in social sciences and humanities. The Czech Technical University (CTU) in Prague operates the VR-1 training reactor primarily focused on training students, nuclear engineering, reactor physics, and research; however, the utilization of this nuclear reactor has also been extended to humanities in recent years. In this direction, the research activities carried out at the reactor include the investigation of historical artefacts and testing of pharmaceutical and food industry samples. This contribution deals with the detailed study of traditional Tibetan medicinal pills by means of instrumental neutron activation analysis (NAA) at the VR-1 nuclear reactor. Three various kinds of Tibetan pills were irradiated by thermal neutrons (ϕ=2×109cm−2s−1) in the experimental vertical channel at maximum reactor power (80 W), and irradiated samples were analysed using the semiconductor HPGe detector; saturated activities were obtained by means of the nuclear gamma-ray spectrometry, and subsequently the composition of the pills was determined (qualitative and quantitative analysis). The presence of Au, K, Fe, Na, As, Sr, Sb, and Hg was revealed in studied samples. The results presented in this paper show clearly that the low-power research reactor VR-1 is excellent tool for the neutron activation analysis experiments.

A neutron radiography beamline relying on the Isfahan Miniature Neutron Source Reactor

Abstract Concerning to the costs of the establishment of new research reactors and their waste issues, the world’s quest is to expand the capabilities of existing research reactors and to develop their application further. In this way, attempts have been made to extract an appropriate thermal neutron beam for neutron radiography purposes at the 30 kW Isfahan MNSR reactor. In this study, the result of calculations and experiments in support of the implementation of a thermal neutron radiography beamline at this reactor has been demonstrated. While many research reactors have beam tube equipment, the original design of the MNSR did not include beam tubes. Therefore, at first stage of our feasibility study, a vertical aluminum tube with a diameter of about 5 cm has been used as an external beam tube to validate calculated values of MCNP code with experimental measurements. The image quality has been determined using ASTM image quality indicators. It can conclude that the Isfahan MNSR is an appropriate reactor with suitable neutron flux for a neutron radiography system. Results show that it is possible to obtain an appropriate neutron radiography beam at this low power research reactor. Also, according to the physical restrictions of the MNSR and considering the least possible changes in the current reactor status, some beam tubes with larger beam diameters have been designed and the related beam parameters have been calculated and demonstrated.