Irradiation Rig Research Articles

Small punch test size reduction to enable high flux irradiation

Screening and qualification of new nuclear materials with superior irradiation resistance requires expensive irradiation exposure and post-irradiation examination before consideration into industrial nuclear structural components. The development of miniaturized test methods is needed for enhanced cost/time efficient testing, while enabling comparative and robust material property estimation in a timely and high throughput manner. The Small Punch Test (SPT) is drawing much interest, especially since the publication of an European standard. Motivated by this standard, a new SP test sample geometry is proposed in order to reduce the material volume to be tested, hences increasing the number of samples that can be introduced in an irradiation rig. The modified sample diameter has been reduced from 8 to 5.35 mm. The qualification of this new design is described here together with a detailed description of the SPT setup modification as compared to the reference setup of the EN standard. Finite element analyses and a series of experiments have been conducted using both standard and reduced size SP samples to demonstrate that the modified test gives estimates of the basic mechanical properties with the same magnitude of error and validates the modified sample geometry for usage in irradiated material testing.

Effect of temperature history on the irradiation behavior of vanadium alloy irradiated with the MARICO-II rig in a fast reactor, JOYO

The irradiation behavior of vanadium alloys was examined to evaluate the effect of temperature history during reactor operation. Irradiation experiments were performed in a fast reactor, Joyo, with an irradiation-temperature-controlled irradiation rig, MARICO-II. In-core irradiation rigs that control the irradiation temperature and are not influenced by reactor power were used. Irradiation conditions with or without the MARICO-II rig were performed. Significant differences in irradiation hardening and swelling behavior in undersized vanadium binary alloys were visible between temperature-controlled and conventional rigs. The temperature history affects the microstructural evolution, especially the nucleation and growth of voids during startup and shutdown procedures. No difference was observed in the vanadium alloys with oversized solute atom because the vacancy migration was suppressed by tight binding between oversized solute atoms and vacancies during irradiation. V–4Cr–4Ti–(Si,Al,Y) alloys showed significant irradiation hardening at high-temperature irradiation even though the contribution of dislocation and precipitates for irradiation hardening was insufficient to satisfy the measured irradiation hardening. The significant irradiation hardening in the V–4Cr–4Ti–(Si,Al,Y) alloys may be caused by transmission-electron-microscope-invisible clusters including minor elements (Si, Al, Y) that were formed at a high temperature during irradiation.

Design and first operation of the MACARON irradiation experiment of Sodium Fast reactor absorber pins in the BOR-60 reactor

The development of the next generation of Sodium-cooled Fast Reactors (SFRs) needs to integrate innovative options regarding the core design in order to comply with the objectives of the 4th generation reactors in terms of safety, cost and performance. In this framework, the aim to design fast core with economy improvements leads to limit the number of control and shutdown systems and to extend their lifetime. These objectives require choosing absorber material with maximal efficiency in terms of neutronic absorption.The French concept of absorber pin is based on the use of boron carbide (B4C) pellets with 10B various enrichments and the absorber element consists of a vented non-leaktight pin with a permanent sodium bond between the pellet column and the stainless steel clad.In order to test an innovative concept of absorber pin and with the aim to improve in one hand, the in pile behavior understanding and in the other hand, the absorber rod modelling codes, an experimental program named MACARON is undergone. This program consists in an irradiation test in the BOR-60 reactor following by a complete NDE and PIE program. Two short absorber pins are irradiated currently. This program is performed by both CEA and RIAR teams.A brief description of the experiment is first provided, in terms of objectives. The main characteristics of absorber pin with the main design criteria are detailed. The manufacturing steps and the characterization of the experimental absorber pins are given with as manufactured results.The design studies of the Irradiation Rig and the neutronic and thermal hydraulic calculations are described in details. The irradiation test has begun in September 2018 and the first experimental results are provided. The huge program of Non-Destructive and Destructive Examinations is detailed with experimental objectives and methods that will be used.

Confocal chromatic sensor for displacement monitoring in research reactor

Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision. Its operation principle is based on a wavelength encoding of the object position. Historically, the company STIL based in the south of France has first developed this class of sensors in the 90’s. Of course, this sensor can only operate in a sufficiently transparent medium in the used spectral domain. It presents the advantage of being contactless, which is a crucial advantage for some applications such as the fuel rod displacement measurement in a nuclear research reactor core and in particular for cladding-swelling measurements. The extreme environmental conditions encountered in such experiments i.e. high temperature, high pressure, high radiations flux, strong vibrations, surrounding turbulent flow can affect the performances of this optical system. We then need to implement mitigation techniques to optimize the sensor performance in this specific environment. Another constraint concerns the small volume available in the irradiation rig next to the rod to monitor, implying the challenge to conceive a miniaturized sensor able to operate under these constraints.

Experimental study of ISHTAR thermostatic irradiation device for the MARIA research reactor

Materials and core components for the next generation power reactors technologies require testing that can be performed in existing research reactors. Such experiments employ devices dedicated to reflect the relevant thermal and neutron parameters simulating conditions present in, for example, but not limited to, HTGR reactors. A novel thermostatic irradiation device named ISHTAR (Irradiation System for High-Temperature Reactors) has been designed and constructed in the MARIA research reactor. Its mission is to enable irradiation of samples in controlled, homogeneous temperature field reaching 1000°C and inert gas atmosphere. The high temperature is achieved by a combination of electric and gamma heating, together with carefully designed thermal insulation. Additionally, samples holder made of graphite with high thermal conductivity enables the temperature homogenization in all directions. Device will be placed inside the Beryllium matrix of MARIA core and cooled with forced circulation of water from the reactor pool loop. This paper presents the outcome of experiments conducted with the rig prototype in external hydraulic mock-up of the MARIA reactor irradiation channel. The results have proved that the desired conditions for irradiation of the samples were achieved and their comparison against computational data has shown the adequacy of the design process. Finally, the loss of flow scenario was tested in protected and unprotected conditions (meaning with and without the safety system based on temperature feedback), proving the operational safety of the ISHTAR design. Experimental results will be used in the future to validate the numerical models (two and three dimensional) of the irradiation rig, providing an improved understanding of free convection and radiation phenomena modeling.

Effect of temperature history on swelling behavior of V-Fe binary alloy irradiated in a fast reactor Joyo

Abstract An on-line instrumented material irradiation rig (MARICO-II) was used for investigating temperature history effect for void swelling behavior in V-Fe binary alloys irradiated at a fast rector, Joyo. Neutron irradiation were performed at 477C with 7.5 dpa in the temperature control capsule (TC capsule) and at 482C maximum temperature with 20.4 dpa in the conventional capsule with monitoring temperature. In the TC capsule, voids was formed and distributed homogeneously, on the other hand a lot of tiny void formation or bimodal void distribution were observed in the conventional capsule, The void swelling rate of V-Fe alloys irradiated in the TC capsule was larger than that in the conventional capsule and it is caused by the imbalance of point defect flux to sink due to void nucleation at the low temperature transient regime in start-up and shutdown period of each cycle in neutron irradiation, that is temperature history effect for swelling behavior of V-Fe binary alloys during neutron irradiation.

РАСЧЕТНО-ЭКСПЕРИМЕНТАЛЬНОЕ ОПРЕДЕЛЕНИЕ ПРОСТРАНСТВЕННО-ЭНЕРГЕТИЧЕСКОГО РАСПРЕДЕЛЕНИЯ НЕЙТРОНОВ В БОКОВОМ ЭКРАНЕ РЕАКТОРА БОР-60. ЧАСТЬ 2

The BOR-60 reactor is actively used for conducting irradiation experiments in support of nuclear reactors both under operation, construction and design. Most of these experiments are conducted in the reactor core using various types of irradiation rigs. However, in recent years there has been keen interest in conducting reactor tests in the BOR-60 blanket. This is associated with several factors: the availability of blanket cells, weak influence on the reactor neutronic characteristics and safety, wide range of irradiation conditions (temperature, neutron flux density, damage doses, neutron spectrum), and cost of irradiation. The experience in BOR-60 blanket computational and experimental studies showed an ill agree-ment between the calculated and experimental neutronic characteristics. In addition to no orientation of the samples under study relative to the core center, there is a significant radial non-uniformity in the neutronic characteristics distribution in the BOR-60 blanket that limits its application for experimental studies. Therefore, detailed computational and experimental studies of the spatial-energy distribution of neutrons in the BOR-60 steel blanket is an important and urgent task.

Design of an irradiation rig using screen method for silicon transmutation doping at the Dalat research reactor

The neutron transmutation doping of silicon (NTD-Si) at research reactors has beensuccessfully implemented in many countries to produce high-quality semiconductors. In the late 1980s, NTD-Si has been tested at the Dalat Nuclear Research Reactor (DNRR) but the results have been limited. Therefore, the design and testing of an irradiation rig for NTD-Si at the DNRR are necessary to have a better understanding in order to apply the NTD-Si in a new research reactor of the Research Centre for Nuclear Science and Technology (RCNEST), which has planned to be built in Viet Nam. This paper presents the design and testing of a new irradiation rig using screen method for testing NTD-Si at the DNRR. The important parameters in the rig such as neutron spectrum and thermal neutron flux distribution were determined by both calculation using MCNP5 computer code and experiment. The aluminum ingots, which have similar neutronic characteristics with silicon ingots, were irradiated in the rig to verify the appropriate design. The uniformity of thermal neutron flux in the rig is less than 5% in axial and 2% in radial directions, respectively. However, the thermal/fast flux ratio of the irradiation rig is 4.38/1 would affect target resistivity of testing Silicon ingots after irradiation.

TRACKING DOWN ALPHA-PARTICLES: THE DESIGN, CHARACTERISATION AND TESTING OF A SHALLOW-ANGLED ALPHA-PARTICLE IRRADIATOR.

Human exposure to α-particles from radon and other radionuclides is associated with carcinogenesis, but if well controlled and targeted to cancer cells, α-particles may be used in radiotherapy. Thus, it is important to understand the biological effects of α-particles to predict cancer risk and optimise radiotherapy. To enable studies of α-particles in cells, we developed and characterised an α-particle automated irradiation rig that allows exposures at a shallow angle (70° to the normal) of cell monolayers in a 30 mm diameter dish to complement standard perpendicular irradiations. The measured incident energy of the α-particles was 3.3 ± 0.5 MeV (LET in water = 120 keV μm−1), with a maximum incident dose rate of 1.28 ± 0.02 Gy min−1, which for a 5 μm cell monolayer corresponds to a mean dose rate of 1.57 ± 0.02 Gy min−1 and a mean LET in water of 154 keV μm−1. The feasibility of resolving radiation-induced DNA double-strand breaks (DSB) foci along the track of α-particles was demonstrated using immunofluorescent labelling with γH2AX and 53BP1 in normal MRC-5 human lung cells.

Feasibility studies for simultaneous irradiation of NBSR & MITR fuel elements in the BR2 reactor

The BR2 reactor physics group has performed extensive neutronic feasibility studies of four Design Demonstration Element (DDE) tests foreseen for the US High Performance Research Reactor (USHPRR) conversion program. The purpose of these tests would be to irradiate the future LEU fuel elements for MURR, MITR, NBSR and HFIR in the BR2 reactor under conditions that are similar to those of the original reactor. The present paper focuses on the preliminary feasibility study for the simultaneous irradiation of the MITR-DDE and the NBSR-DDE in the BR2 reactor. In order to mimic the irradiation conditions of the original reactors for both DDEs inside the BR2 reactor, a number of irradiation scenarios and irradiation rig designs were optimized.The results obtained in this paper show that the BR2 reactor is capable to accurately recreate the typical irradiation conditions experienced by the future fuel elements of the NBSR and MITR with high fidelity. The power evolution and final burnup distributions are close to the specified values. We have proposed and have shown the possibility to irradiate both DDEs at the same time. The irradiations, if conducted as specified at fully representative powers, are long (three to five years), due to the requested relatively low power in combination with targeted elevated burnups. It is possible to reduce the irradiation time if a higher power can be accepted, but in this feasibility study, we attempted to simulate the required conditions as accurately as possible.

HTR-PM fuel pebble irradiation qualification in the high flux reactor in Petten

NRG has performed a High Temperature Reactor fuel qualification test in the framework of the Chinese HTR development for INET – Tsinghua University, as a part of the qualification process of Chinese fuel for use in the HTR-PM (High Temperature Reactor-Pebble bed Module). Fuel qualification is an essential part of the licensing process of the HTR-PM, currently under construction in the Shandong province in the People’s Republic of China.The design of the INET irradiation setup for irradiation in the High Flux Reactor is performed by NRG. In the design of the irradiation rig, the pebbles are encased in graphite half shells to form a solid cylindrical stack. Within the graphite, thermocouples are located for temperature monitoring, and neutron fluence monitor sets for neutron fluence verification. Next to the temperature measurements, the online monitoring consisted of fission product release measurements. These measurements were performed with the Sweep Loop Facility, designed and built by JRC-IET.Post Irradiation Examinations are performed and consisted of gamma scanning, dimensional and weight measurements and visual inspection. In April 2016, the pebbles were transported from Petten to Karlsruhe, Germany where the second required step of fuel qualification will be performed, namely heating tests in the KÜFA facility.This paper describes the global lay-out and functionality of the irradiation facility. The online data generated and irradiation parameters are reported, with focus on the measured fission product release behavior of the fuel. The paper also presents non-destructive post irradiation examinations.

Neutronic analysis of the fuel loaded irradiation loop device of the RMB Multipurpose Brazilian Reactor

The structuring project of the Brazilian Multipurpose Reactor (RMB) is responsible for meeting the capacity to develop and test materials and nuclear fuel for the Brazilian Nuclear Program. An irradiation test device (Loop) capable of performing fuel test for power reactor rods is being conceived for RMB reflector. In this work neutronic calculations have been carried out in order to determine parameters to the Loop basic design. Safe position determined to be 52.7 cm displaced from the core, where linear heat rate (q’) < 1 W/cm fission power is released in the rig loading fissile sample. Since irradiation rig moves during normal operation of the reactor, displacement of the samples must be investigated in terms of perturbation in the core. Results confirmed efficacy of the 12 cm heavy water layer designed to suppress reactivity effects in the core. Linear heat rate in the fissile sample was calculated in order to estimate predefined power ramps of 50–350 W/cm, information necessary to the project of the displacement device. Ramp lengths of 9.59 cm–13.22 cm were found depending on the peak factor considered for the sample rod and the number of beryllium rows arranged in the reflector.

Steady state and transient thermal-hydraulic performance of the IFMIF-HFTM irradiation rigs and container with modeling of nuclear heating

During the EVEDA phase of the International Fusion Materials Irradiation Facility (IFMIF), the High Flux Test Module (HFTM) was developed as dedicated irradiation device for Small Scale Testing Technique (SSTT) material specimens in the intensive IFMIF neutron radiation field. The specimens are contained in temperature controlled irradiation rigs. Since one of the requirements for the HFTM is to provide a uniform temperature field for the irradiated specimens, thermal testing was a priority for the performed validation activities.In the HFTM “single-rig” (HFTM-SR) experiments a single rig of 1:1 scale was tested. The heater plates and the specimen region inside the rig were instrumented with thermocouples to monitor heat transfer and specimen temperature spread.In the High Flux Test Module “double compartment” (HFTM-DC) experiments a fully equipped prototype with three heated rigs was tested in the HELOKA-LP helium loop. Special heater cartridges are used to substitute the nuclear heating. These experiments show that the full range of operation temperatures (250–550°C) required for the IFMIF HFTM could be well achieved and well controlled with and without surrogate nuclear heater power. The temperature spread measured inside a capsule is in the range of +/−3K in the lateral direction and +2K/−8K in the vertical direction for the 350°C reference case, fulfilling very well the requirements. This compares well to an allowed +/−19K range according to the requirements.No unforeseen thermal hydraulic effects like oscillations, hysteresis etc. could be detected. To cool down from 350°C to 50°C it takes roughly 315seconds, heating up 135seconds.

A Concise Design for the Irradiation of U–10Zr Metallic Fuel at a Very Low Burnup

Abstract In order to investigate the swelling behavior and fuel–cladding interaction mechanism of U–10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel–cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal–hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

Innovation of Irradiation Equipment at the LVR-15 Nuclear Research Reactor

This paper describes the design and use of a new irradiation facility for the LVR-15 nuclear research reactor. The CHOUCA MT irradiation rig was produced in France according to a design of the ÚJV Group (ÚJV Řež and Research Center Řež). There are six heating sections situated along the rig, each instrumented and controlled by its own thermocouple. The rig’s insulation layers ensure a balanced temperature in irradiated specimens along its entire length. The specimen holder is 55.9 mm in diameter and 320 mm long. The CHOUCA MT rig can be repetitively irradiated in different positions within the reactor core, depending on irradiation condition requirements. The CHOUCA MT rig expands the possibilities of radiation research in the ÚJV Group.

Precise Control of Neutron Irradiation Fluence in the Neutron Transmutation Doping Process at HANARO Using SPND and Zirconium Foils

A neutron irradiation control method using a self-powered neutron detector (SPND) and zirconium foils was developed for neutron transmutation doping (NTD) application at HANARO. An SPND was installed at a fixed position of the upper part of the sleeve at the HANARO NTD hole for real-time monitoring of the neutron irradiation. The SPND signal output was significantly affected by the in-core conditions and surroundings of the facility. Furthermore, the SPND signal varied by $\sim 15\%$ throughout the reactor cycle according to the change of the control rod position. However, it was also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation rig was not significant for the control rod motion. In this procedure, the process for determining the thermal neutron flux by using zirconium foils was established for achieving efficient NTD irradiation. Accordingly, the relationship between the ratio of the neutron flux to the SPND output and the control rod position was deduced. Silicon irradiation experiments for transmutation doping were performed using the established relationship. The irradiated neutron fluence was controlled within ±1.3% of the target value. The mean irradiation/target ratio of the fluence was 0.9992, and the standard deviation was 0.0071. Thus, we confirmed that the produced irradiation was extremely accurate. This procedure can be useful in SPND applications requiring real-time high-accuracy neutron fluence measurements.

Computational and experimental study of an irradiation rig with a fuel heater for the BOR-60 reactor

Advanced core materials and components for nuclear reactors of a new generation are tested in the BOR-60 reactor. A novel irradiation rig (IR) with a fuel heater is used for high-temperature tests of material samples. The irradiation rig features a number of advantages over ampoule-type irradiation rigs commonly used nowadays. Computational and experimental studies on an IR with a fuel heater have been conducted in the BOR-60 reactor core. The results of a dedicated methodical experiment have proved that it is possible to provide the required temperature conditions for irradiation of tested samples. MCU-RR, a precision code, was used for neutronic calculations, and thermohydraulic calculations were performed using the ANSYS CFX software system. A comparison of calculated temperature values against experimental data has shown a fit in the experimental error limits which confirms the applicability of the selected codes, models and procedures. Computational and experimental studies have also been conducted for the temperature distribution in the IR with a fuel heater following the withdrawal of the IR from the reactor and its placement in a dry cooling channel. The decay power in the IR fuel pins were calculated using the AFPA code and the temperature fields were calculated based on ANSYS CFX. It has been shown that the permissible temperature value on the fuel cladding is not exceeded in the IR withdrawn from the reactor following two-day cooling after the reactor shutdown.

Расчетно-экспериментальные исследования облучательного устройства с твэльным подогревом в реакторе БОР-60

Расчетно-экспериментальные исследования облучательного устройства с твэльным подогревом в реакторе БОР-60

Hydraulic and thermal testing of different helium cooled irradiation rig models for the IFMIF High Flux Test Module

The hydraulic and thermal testing of two different irradiation rig models A and B, differing in the inlet nozzle design, bottom reflector length and steps inside a mock-up container is part of the HFTM validation activities which support the engineering design of the High Flux Test Module.The pressure drops for all models in the test section are measured for overall mass flow rates of 1–12g/s and different absolute pressures of 1500hPa and 2500hPa at the pressure port at the inlet section. The pressure drops in different sections of the experiment and in the single rig minichannels are also measured with additional pressure ports on the surfaces of the rig models.Predictability and reproducibility of the cooling effects of the main cooling channels in the HFTM irradiation zone can be shown. Rig model B with a backward facing step is for high mass flow rates >∼7.5g/s (this is the operation regime of the HFTM) superior to rig model A. Uniform perfusion of the multiple parallel minichannels of the irradiation rigs by helium gas is of importance to obtain uniform and predictable temperatures. Temperature fields under different heater and flow conditions have been measured.

Commissioning of the New Irradiation Device for Neutron Transmutation Doping

The LVR-15 reactor is a 10 MW tank type multi-purpose research reactor. The reactor is utilized for production of a wide scale of isotopes with the main focus on generators, testing of materials and chemical regimes of coolant in irradiation rigs and experimental loops (in-pile and out of pile), beam experiments and neutron transmutation doping. The current capacity for neutron transmutation doping was one irradiation facility dedicated to maximum 3 inch ingots. Recently, this capacity was increased by a new irradiation device. Although, the current market demand for neutron transmutation doping is for ingots with diameters up to 6 inch, occasionally even 10 inch, the new device was designed for smaller ingots. The main reason for having a 4 inch diameter maximum for this new device was the limited space in the reactor core. For commissioning purposes of the irradiation channel and for repetitive verification of irradiation parameters special dummy ingots were developed. These dummies can be equipped with three sets of activation detectors, so that the neutron fluence across the ingot can be determined. Within the commissioning, two sets of measurements using the dummy ingots were performed. The primary purpose of these measurements was to verify axial positioning of the irradiation device. The first measurement revealed some issues with accurate positioning of the irradiation device. After these issues were fixed, a new set of measurements using dummy ingots was performed. As a result of these measurements, an estimation of optimal irradiation position was refined. For this refined position a set of silicon ingots was irradiated. From the irradiation results of these ingots, a calibration of the online monitoring system could be done. The successful irradiation of this set of silicon ingots was the last test of the device commissioning; the positioning of the irradiation device and its performance was verified.