Preliminary Study on the Remaining Life Estimation Method for Research Reactor Tank Liner

The reactor tank liner is one of the most crucial safety barriers in a research reactor as it retains the radioactive material released from the fuel during the accident condition. It also contains the primary coolant for fission heat removal. The integrity of the tank liner determines the service life of the research reactor. So far, the remaining life estimation of pressure vessels in nuclear power plants is more widely applied and established than that of the research reactor tank liner. Therefore, a study on the remaining life estimation method of the research reactor tank liner is needed to ensure the research reactor operation safety. This paper aims to preliminarily study several methods applied to estimate the remaining life of a research reactor tank liner. The preliminary study consists of a qualitative assessment and a quantitative assessment. The qualitative assessment aims to propose several techniques or methods applied in estimating the remaining life of the reactor tank liner. The quantitative assessment applies one of the remaining life estimation methods discussed in the previous assessment. Generally, the remaining life of the research reactor tank liner can be estimated using the theoretical method and the experimental method. The theoretical methods are applied by calculating the neutron fluence received by the tank liner or by analyzing the fracture mechanics using numerical modeling if the cracks or other defects exist. The calculation of atom displacement number (dpa), as a standard measure of the neutron-induced radiation damage of the materials, can support the neutron fluence calculation. The experimental method is conducted by measuring several parameters of the tank liner material, such as the corrosion rate or the mechanical properties. In the quantitative assessment, the remaining life estimation of the Kartini Reactor tank liner was performed by neutron fluence calculation method using MCNP6 computer code. The result shows that the maximum neutron fluence received by the tank wall is 2.950E+17 n/cm2 for 40 years operating period. By comparing the cumulative neutron fluence received for 40 years to the thermal neutron fluence limit value of 1.18E+23 n/cm2, the Kartini Reactor tank liner can still be used for the next 1.6E+07 operation years. The result of the quantitative assessment implicitly shows that the remaining life estimation of the tank liner needs to: 1) consider all defects experienced by the tank liner and all factors (e.g., thermal, radiation, chemical, cyclic loading) which affect the tank liner material condition, and 2) perform the combination of theoretical and experimental methods. For an open-pool type reactor, corrosion monitoring and corrosion rate measurement are essential to perform the remaining life assessment of the tank liner.

Radiological characterization of a German pressurized water reactor based on a highly resolved method for activity analysis and dose rate calculation

The amendment to the atomic act in 2011 results to phase out nuclear energy in Germany until the end of 2022. Subsequently, the licensee of the nuclear power plant is responsible for decommissioning and dismantling. During operation, activation of structures near the core of the reactor occur which govern the amount of radioactive waste, the dose rate distribution and dismantling strategies. Thus, a detailed radiological characterization of in-core and out-core structures is required to optimize decommissioning processes regarding the quantification and minimization of radioactive waste, radiation protection and reducing radiation exposure. These objectives are achieved using an innovative and efficient method developed and applied at the Chair of Repository Safety (Lehrstuhl für Endlagersicherheit, ELS) RWTH Aachen University. Within the framework of the joint project „Development of a methodology for activity analysis and dose rate estimation“, funded by the Federal ministry of Education and Research, approaches the objective to develop a standardized and highly resolved method to calculate time-dependent activity of components and structures near the reactor core based on operating history of the nuclear power plant and neutron fluence distribution. The approach requires the development of a detailed model for Monte-Carlo simulations which provides the basis to neutron fluence, neutron spectra and radiation transport simulations. To calculate the nuclide specific 3-Dimensional (3D) activity distribution of the entire facility, a facility-dependent activation cross section library is produced which focuses on recent nuclear databases (ENDF/B-VIII.0). A highly resolved and space-dependent 3D activity distribution of the entire facility is obtained using a modular program package, developed at ELS, including the activation code ORIGEN2. The results are produced in the form of detailed 3D activity maps. The source terms are generated on the basis of the space-dependent 3D activity distribution using an additional module of the program package. The combination of recent nuclear databases focusing on ENDF/B-VII.1 and complemented by JEFF-3.3 ensures a comprehensive characterisation of source terms. Subsequently, source terms are prepared for 3D radiation transport simulation using the Monte-Carlo method and the computer code MCNP. The simulations are conducted separately for each individual component obtaining the partial contribution of all in-core and out-core structures as well as the dose rate distribution of the entire facility. Similar to the activity calculation, the simulation results are used to generate 3D gamma flux and dose rate maps using the graphic module of the whole program system. On the basis of the radiological characterisation and in view of a high-level radiation protection these maps allow the optimum planning and realisation of the decommissioning and dismantling process of the nuclear power plant.

Investigation of the Impact of Neutron Irradiation on SiC Power MOSFETs lifetime by Reliability Tests

High temperature reverse-bias (HTRB), High temperature gate-bias (HTGB) tests and electrical DC characterization were performed on planar-SiC power MOSFETs which survived to accelerated neutron irradiation tests carried out at ChipIr-ISIS (Didcot, UK) facility, with terrestrial neutrons. The neutron test campaigns on the SiC power MOSFETs (manufactered by ST) were conducted on the same wafer lot devices by STMicroelectronics and Airbus, with different neutron tester systems. HTGB and HTRB tests, which characterise gate-oxide integrity and junction robustness, show no difference between the non irradiated devices and those which survived to the neutron irradiation tests, with neutron fluence up to 2× 1011 (n/cm2). Electrical characterization performed pre and post-irradiation on different part number of power devices (Si, SiC MOSFETs and IGBTs) which survived to neutron irradiation tests does not show alteration of the data-sheet electrical parameters due to neutron interaction with the device.

Development of a neutron detector for radiation protection monitoring

The paper presents the results of the development of a neutron detector for radiation protection purposes. Monte Carlo simulations, using MCNP5 code, were performed to optimize the configuration of the neutron detector. The developed detector consists of a 3He proportional counter embedded in a multi-layer moderator made of high-density polyethylene (HDPE) and Cadmium. The characteristics of the developed neutron detector including neutron fluence response and ambient dose equivalent response were calculated, analyzed and compared with those from other neutron survey meters. The simulation model and computed results were assessed through experimental measurements at the Secondary Standards Dosimetry Laboratory of the Institute for Nuclear Science and Technology (INST). A good agreement between the simulated and experimental results was observed within 9.3% for 241Am-Be source and four simulated workplace neutron fields.

Assessment of neutrons from secondary cosmic rays at mountain altitudes – Geant4 simulations of environmental parameters including soil moisture and snow cover

Ground based measurements of neutrons from secondary cosmic rays are affected by environmental parameters, particularly hydrogen content in soil. To investigate the impact of these parameters, Geant4 Monte Carlo simulations were carried out. In a previous study the model used for the Geant4 Monte Carlo simulations was already validated by measurements performed with an Extended Range Bonner Sphere Spectrometer (ERBSS) at Zugspitze, Germany, and at Jungfraujoch, Switzerland. In the present study a sensitivity analysis including different environmental parameters (i.e., slope of mountain, snow height, soil moisture, and range of albedo neutrons) and their influence on the flux of neutrons from secondary cosmic rays was performed with Geant4. The results are compared with ERBSS measurements performed in 2018 at the Environmental Research Station “Schneefernerhaus” located at the Zugspitze, Germany. It is shown that the slope of the Zugspitze mountain reduces the neutron flux from secondary cosmic rays between about 25 % and 50 % as compared to a horizontal surface, depending on neutron energy and snow cover. An increasing height of snow cover, simulated as snow water equivalent (SWE), reduces the total neutron flux exponentially down to a factor of about 2.5 as compared to soil without any snow cover, with a saturation for snow heights greater than 10 cm to 15 cm SWE, depending on neutron energy. Based on count rates measured with the individual spheres of the ERBSS, SWE values were deduced for the whole year 2018. Specifically, mean SWE values deduced for the winter months (January to March) are between 6.7 and 10.1 cm or more, while those for the summer months (July to September) are between 2.1 and 3.6 cm. Soil moisture of 5 % water mass fraction in limestone leads to a decrease of the total neutron flux by about 35 % compared to dry limestone. At a height of 1.5 m above ground, 86 % of the total albedo neutron fluence at the detector position are from a ground area with a radius of about 75 m. It is concluded that measurement of neutrons from secondary cosmic radiation can be used to gain information on height of snow cover and its seasonal changes, soil moisture, but also information on local geometry such as mountain topography. Because the influence of such parameters on neutron fluence from secondary cosmic rays depends on neutron energy, analysis of the whole neutron energy spectrum is beneficial.