The Development of Radiation-Resistant RF Tags for Use at Nuclear Power Plants

RF tags based on RFID (Radio-frequency Identification) technology have been widely used in various fields including power plant construction and maintenance for the purpose of improving the identification and traceability of the many components in the facility. To date, various types of tags have been developed, including tags that are resistant to chemicals or high-temperature environments, which are used in specialized fields. When considering widespread use of RF tags in nuclear power plants, there is a concern about the effects of radiation on the RF tags, because the data stored in the tag may receive radiation damage, resulting in corruption of data. Here, we describe a newly designed RF tag that achieves resistance to radiation damage by attaching a radiation shield layer and incorporating automatic data-correction software. This radiation-resistant RF tag has been tested under real radiation exposure fields to verify the intended radiation-resistant functions. It is expected that the use of these radiation-resistant RF tags with a data reader and database system will increase the capabilities of RF tags applied to nuclear power plants and it is also expected to lead to reductions in worker radiation exposure doses.

Corrosion Inhibitor on Copper and Stainless Steel for Component Cooling Water System of NPP

Based on the corrosion issues of component cooling water system (CCWs) in nuclear power plant (NPP), the corrosion inhibition properties and protection mechanism on copper and stainless steel was studied by using tests such as electrochemical method, immersion test and dynamic water simulation. Results show that the optimum inhibitor is the compound of tolyltriazole (TTA) and phosphate, which has an excellent corrosion inhibition efficiency on copper in either pure water or abnormal water. The inhibitor also elevated the pitting potential of stainless steel and contributed to the corrosion resistance.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

Effect of Bi Addition on the Corrosion Behavior of Zirconium Alloys

In order to investigate systematically the effect of Bi addition on the corrosion resistance of zirconium alloys, different zirconium-based alloys, including Zr-4 (Zr-1.5Sn-0.2Fe-0.1Cr), S5 (Zr-0.8Sn-0.35Nb-0.4Fe-0.1Cr), T5 (Zr-0.7Sn-1.0Nb-0.3Fe-0.1Cr) and Zr-1Nb, were adopted to prepare the zirconium alloys containing Bi of 0∼0.5% in mass fraction. These alloys were denoted as Zr-4+xBi, S5+xBi, T5+xBi and Zr-1Nb+xBi, respectively. The corrosion behavior of these specimens was investigated by autoclave testing in lithiated water with 0.01 M LiOH or deionized water at 360°C/18.6 MPa and in superheated steam at 400 °C/10.3 MPa. The micro structure of the alloys was examined by TEM and the second phase particles (SPPs) were analyzed by EDS. Micro structure observation shows that the addition of Bi promotes the precipitation of Sn as second phase particles (SPPs) because Sn is in solid solution in α-Zr matrix in Zr-4, S5 and T5 alloys. The concentration of Bi dissolved in α-Zr matrix increase with the increase of Nb in the alloys, and the excess Bi precipitates as Bi-containing SPPs. The corrosion results show that the effect of Bi addition on the corrosion behavior of different zirconium-based alloys is very complicated, depending on their compositions and corrosion conditions. In the case of higher Bi concentration in α-Zr, the zirconium alloys exhibit better corrosion resistance. However, in the case of precipitation of Bi-containing SPPs, the corrosion resistance gets worse. This indicates that the solid solution of Bi in α-Zr matrix can improve the corrosion resistance, while the precipitation of the Bi-containing SPPs is harmful to the corrosion resistance.

Helium Permeability Measurement of SiC Ceramic Composites for Advanced Reactor Application

Permeability of helium gas of Silicon carbide ceramic composites material, which is one of the most important properties in application of SiC composite for advanced reactors, is studied by using a simple, low-cost test system. The test system can not only qualitatively determine whether the sample is permeable or not, but also quantitatively measure the permeability for the permeable ones by water displacement. The tests are conducted with low pressure in room temperature. The permeability of the SiSiC composite depends on the preparation method. In four flat-plate materials prepared by different processes for the test, the splint based SiSiC and cordierite coated fiber reinforced SiSiC are hermetic, the permeability of uncoated fiber reinforced SiSiC and CVD carbon coated fiber reinforced SiSiC are 0.216cm2/s and 0.109cm2/s, which imply that the permeability is cut in half with the coating. The samples are scanned under SEM to analyze their microscopic structures and verified that the difference of permeability is related to their coatings as well as the pores and cracks.

Primary Shielding Design for an Optimized Molten Salt Reactor

Molten salt reactor (MSR) system, a candidate of the Generation IV reactors, has inherent safety, on-line refueling and good neutron economy as typical advantages. An optimized MSR is developed by changing the size of fuel channel and the graphite-to-molten salt volume radio, based on the Molten-Salt Reactor Experiment (MSRE), which was originally developed at the Oak Ridge National Laboratory (ORNL). In this paper, shielding calculations for the optimized MSR are presented. The goal of this study is to determine the necessary shielding to decrease the neutron and gamma dose rate to the acceptable level according to national regulations. The operating temperature of the optimized MSR is designed in the range of 500 °C–700 °C, heat removal is also considered in the shielding design. The shielding calculations are carried out by using Monte Carlo method. The shielding system of the optimized MSR consists of 7 zones: the core, the core can, the reactor vessel, the thermal shield, the reactor cell containment, the shield tank and the concrete wall. The combinations of shielding materials in the thermal shield were evaluated. The thermal shield filled with carbon steel balls and circulating water gets an excellent shielding performance and heat removing effects. The neutron spectra and dose distributions, as well as the energy deposition over different shields have been analyzed. The total neutron dose rate outside the thermal shield is attenuated by a factor of about 104, and the gamma dose rate by a factor of about 103. These results show that the shielding design could low dose rate to an acceptable level outside the shielding and far below dose limit required.

Electronic Structure and Mechanical Properties of Zircaloy-2 and Zircaloy-4: A First Principle Study

The local-density approximation (LDA) coupled with the virtual crystal approximation (VCA) method electronic structure is applied to evaluate elastic constants, bulk modulus, shear modulus, Young’s modulus and Poisson’s ratio mechanic properties of metal zirconium, Zircaloy-2 and Zircaloy-4. The results show that there is no obvious difference in band structure and total density of state (DOS) between metal zirconium and zirconium alloy. However, p and d electron partial density of state (PDOS) presents the slight difference between metal zirconium and zirconium alloy. Zircaloy-2 and Zircaloy-4 present better elastic mechanical properties than metal zirconium. The metal zirconium and zirconium alloy show the anisotropic mechanical properties.

Evolution of Crystal Orientations in Plastically Deformed Steels: Role of Constitutive Models Used in Finite Element Simulations

Stainless steel is a commonly used material in safety-important components of nuclear power plants. In order to study degradation mechanisms in stainless steels, like crack initiation and propagation, it is important to characterize the degree of plastic strain on microstructural level. One way to estimate local plastic strain is by measuring local crystal orientations of the scanned surfaces: the electron backscatter diffraction (EBSD) measurements on stainless steel revealed a strong correlation between the spread of crystal orientations within the individual grains and the imposed macroscopic plastic strain. Similar behavior was also reproduced by finite element simulations where stainless steel was modeled by an anisotropic elasto-plastic constitutive model. In that model the anisotropic Hill’s plasticity function for yield criteria was used and calibrated against the EBSD measurements and macroscopic tensile curve. In this work the Hill’s phenomenological model is upgraded to a more sophisticated crystal plasticity model where plastic deformation is assumed to be a sum of crystalline slips in all activated slip systems. The hardening laws of Peirce, Asaro and Needleman and of Bassani and Wu are applied in crystal plasticity theory and implemented numerically within the user subroutine in ABAQUS. The corresponding material parameters are taken from literature for 316L stainless steel. Finite element simulations are conducted on the analytical Voronoi tessellation with 100 grains and initial random crystallographic orientations. From the simulations, crystal and modified crystal deformation parameters are calculated, which quantify mean and median spread of crystal orientations within individual grains with respect to central grain orientation. The results are compared to EBSD measurements and previous simulations performed with Hill’s plasticity model.

Research on Negative Pressure Control Technology in NPP Reactor Building and its Application

The paper mainly elaborates the negative pressure control technology and commissioning approaches for double-wall containment of Russian WWER-1000 nuclear power units. It also carries out an analysis and research on the layered negative pressure technology in the containment. It mainly includes the following three parts: A Russian WWER-1000 nuclear power unit adopts the structure of double-wall containment for its Reactor building, with independent negative pressure systems for the containment and the annular space between the two walls. The paper mainly elaborates the control methods and limits requirements for the negative pressure in the containment and the annular space under the normal operation condition and in case of design basis accidents, with analysis and argumentation on the design function and operation requirements of the negative pressure system for the containment and the annular space. In the paper, the design philosophy of layered negative pressure and its feasibility study are analyzed from the aspects of radiated partition, air distribution of the negative pressure system and containment separator for layered negative pressure. The commissioning methods and technical requirements of negative pressure system in the Reactor building are described in the paper. Problems encountered during commissioning are also addressed and analyzed. Operations practices prove that the negative pressure control technology for double-wall containment of WWER-1000 nuclear power unit is advanced and reliable, which meets the requirements on nuclear air decontamination emission and radiation protection, and is worthy of study, research and reference.

Proton Dose Conversion Coefficients Based on Chinese Reference Adult Woman Voxel Phantom

In order to calculate the dose conversion coefficients for proton, the voxel model of Chinese Reference Adult Woman (CRAW) was established by the Monte Carlo transport code FLUKA according to the Chinese reference data and the Asian reference data. Compared with the reference data, the deviations of the mass for organs or tissues of CRAW is less than ±5%. Calculations have been performed for 14 incident monoenergetic protons energies from 0.02GeV to 10TeV at the irradiation incident of anterior-posterior (AP) and posterior-anterior (PA). The results of fluence-to-effective dose conversion coefficients are compared with data from the different models such as an anthropomorphic mathematical model, ICRP reference adult voxel model, the voxel-based visible Chinese human (VCH). Anatomical differences among various computational phantoms and the spatial geometric positions of the organs or tissues lead to the discrepancies of the effective dose conversion coefficients in the ranging from a negligible level to 107% at proton energies below 0.2GeV. The deviations of the coefficients, above 0.2GeV, are mostly within 10%. The results of fluence-to-organ absorbed dose conversion coefficients are compared with the data of VCH. The deviations of the coefficients, below and above 0.2GeV, are within 150% and 20%, respectively. The primary factors of the deviations for the coefficients should be due to the differences of the organ mass and the size of the body shape.