Cracking Driving Force at the Tip of SCC under Heterogeneous Material Mechanics Model of Safe-End Dissimilar Metal-Welded Joints in PWR

The complicated driving force at the stress corrosion cracking (SCC) tip of the safe-end dissimilar metal-welded joints (DMWJs) in the pressurized water reactor (PWR) is mainly caused by the heterogeneous material mechanical properties. In this research, to accurately evaluate the crack driving force at the SCC in DMWJs, the stress-strain condition, stress triaxiality, and J-integral of the crack tip at different positions are analyzed based on the heterogeneous material properties model. The results indicate that the larger driving force will be provided for the I-type crack when the crack is in the SA508 zone and the interface between the 316L region and base metal. In addition, the heterogeneous material properties inhibit the J-integral of the crack in the 316L region, which has a promoting effect when the crack is in the SA508 zone and weld metal. It provides a new idea for analyzing driving force at the crack tip and safety evaluation of DMWJs in PWRs.

Mechanical Properties Evaluation and Crack Propagation Behavior in Dissimilar Metal Welded Joints of 304 L Austenitic Stainless Steel and SA508 Low-Alloy Steel

The material mechanical properties and crack propagation behavior of dissimilar metal welded joint (DMWJ) of pressurized water reactor (PWR) was investigated. In this research, the mechanical parameters of the cladding layer materials (304L-SA508) of the DMWJ in PWRs were obtained by the continuous indentation test. Simultaneously, the user-defined (USDFLD) subroutine in ABAQUS was used to establish the heterogeneous materials model of the welded joint. On this basis, the local crack propagation path of DMWJs has been discussed based on the extended finite element method (XFEM). The result indicated that the strength value at the fusion boundary line (FB line) is the largest, and the yield strength reaches 689 MPa. The yield stress values of the cladding metal (304 L) and base metal (SA508) are 371 MPa and 501 MPa, respectively. Affected by the material constraint effect of the DMWJ, the crack will propagate through the FB line when the initial crack is perpendicular to the FB line. And when the initial crack parallels the FB line, the crack will deviate from it. Meanwhile, the crack propagation length is smaller as the initial crack tip is closer to the FB line when the load condition is constant.

Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

Calculation of the Neutron Parameters for Accelerator-Driven Subcritical Reactors

In this paper, the accelerator-driven subcritical reactor (ADSR) is simulated based on structure of the TRIGA-Mark II reactor. A proton beam is accelerated and interacts on the lead target. Two cases of using lead are considered here: firstly, solid lead is referred to as spallation neutron target and water as the coolant; secondly, molten lead is considered both as a target and as a coolant. The proton beam in the energy range from 115 MeV to 2000 MeV interacts with the lead to create neutrons. The neutron parameters as neutron yield Yn/p, neutron multiplication factor k, the radial and axial distributions of the neutron flux in the core have been calculated by using MCNPX program. The results show that the neutron yield increases as the energies of the proton beam increases. When using the lead target, the differences between the neutron yield are from 4.2% to 14.2% depending on the energies of the proton beam. The proportion of uranium in the mixtures should be around 24% to produce an effective neutron multiplier factor greater than 0.9. The neutron fluxes are much higher than the same calculations for the TRIGA-Mark II reactor model using tungsten target and light water coolant.

Solution of the Static Deflection Mode Shape Function of the Cantilever Beam under Transverse Flow Based on the Boundary Shooting Method

According to the characteristics of the reactor internal structure of nuclear power plants, the vibration of the secondary core support pillar in water can be modeled as the vibration of the cantilever beam structure under the action of transverse flow, and its first beam mode is highly likely to be activated. It is thus necessary to dedicate a separate study on the first-order beam mode. In this work, we study the secondary core support pillar in nuclear reactor AP1000 under the action of transverse flow and focus on the derivation of its static cantilever deflection mode shape function in order to lay a foundation for the calculation of hydrodynamic added mass and frequency for the nuclear reactor internal components and their structural integrity evaluation. First, we proposed a set of nonlinear differential equations for the analysis of the single cantilever beam. Second, to solve the nonlinear differential equations, we used a boundary shooting framework in combination with the Runge–Kutta method. The results of the numerical simulation agree with the analytical solution to a very high degree, which demonstrates the effectiveness of the simulation method. Finally, we solved the static deflection mode shape function of the secondary core support pillar under the normal operating conditions. The nonlinear differential model and simulation method proposed in this paper can be used to solve the static cantilever deflection mode shape function of the equipment support tube.

Study on the Applicability of Neutron Radiation Damage Method Used for High-Temperature Superconducting Tape Based on Geant4 and SRIM

High-temperature superconducting material is a promising candidate to fabricate superconducting magnet for magnetic confinement fusion reactors. The DPA number of the 1 µm thick superconducting layer in a high temperature superconducting tape under neutron irradiation needs to be calculated to predict the property changes. The DPA cross sections, which ignore the spatial distribution of vacancies caused by PKAs, are commonly used to obtain the results of the damage energy and DPA. However, for geometric models with the thickness as small as 1 µm, the energy and angular distribution of PKAs reveal that a significant number of PKAs with relatively high energy tend to scatter forward and cross the boundary of model, so the thickness of model has the potential to affect the number of displaced atoms. In this paper, we developed a method based on Geant4 and SRIM to evaluate the deviation of the traditional analytic method caused by the thickness. Geant4 is used to obtain the location, direction, and energy of PKAs, while SRIM is used to track every PKA and obtain damage energy and the number of displaced atoms. The radiation damage calculation of simple thin plate models with different thicknesses and the tape model are conducted with the neutron energies from 1 to 14 MeV. The results show that PKAs need to be tracked continuously for models with thickness less than 10 µm and the deviation of the analytic formulas increases rapidly with the decrease of thickness. For the superconducting layer composed of four different elements in the tape, the deviation also depends on the proportion of each atomic species and the neutron-atom interaction cross sections under different incident neutron energy.

Xe Ion Irradiation-Induced Microstructural Evolution and Hardening Effect of Nickel-Base Superalloy

The nickel-base superalloy Hastelloy N was irradiated using 1 MeV Xe20+ and 7 MeV Xe26+ ions with displacement damage ranging from 0.5 dPa to 10 dPa at room temperature (RT). The irradiated Ni-based superalloy was characterized with transmission electron microscopy (TEM), XRD, and nanoindenter to determine the changes in microstructural evolution and nanoindentation hardness. The TEM results showed that ion irradiation induced a large number of defects such as black spots and corrugated structures and the second phase was rapidly amorphized after being irradiated to a fluence of 0.5 dPa. The XRD results showed that the Hastelloy N alloy sample did not undergo lattice distortion after ion irradiation. An obvious irradiation hardening phenomenon was observed in this study, and the hardness increased with Xe ion fluence. The pinning effect in which the defects can become obstacles to the free movement of dislocation may be responsible for the irradiation-induced hardening.

Technical Options and Cost Estimates for Spent Nuclear Fuel Management at the Barakah Nuclear Power Plants

In the planning and management of the interim storage of spent nuclear fuel, the technical and economic parameters that are involved have a significant role in increasing the efficiency of the storage system. Optimal parameters will reduce the total economic costs for countries embarking on nuclear energy, such as the UAE. This study evaluated the design performance and economic feasibility of various structures and schedules, to determine an optimal combination of parameters for the management of spent nuclear fuel. With the introduction of various storage technology arrangements and expected costs per unit for the storage system design, we evaluated eight major scenarios, each with a cost analysis based on technological and economic issues. We executed a number of calculations based on the use of these storage technologies, and considered their investment costs. These calculations, which were aligned with the net present value approach and conducted using MS Project and MATLAB software programs, considered the capacities of the spent fuel pools and the amount of spent nuclear fuel (SNF) that will be transferred to dry storage facilities. As soon as they sufficiently cool, the spent nuclear fuel is to be stored in a pool storage facility. The results show that applying a centralized dry storage (CDS) system strategy is not an economically feasible solution, compared with using a permanent disposal facility (PDF) (unless the variable investment cost is reduced or changed). The optimal strategy involves operating a spent fuel pool island (SFPI) storage after the first 20 years of the start of the permanent shutdown of the reactor. After 20 years, the spent fuel is then transferred to a PDF. This strategy also results in a 20.9% to 26.1% reduction in the total cost compared with those of the other strategies. The total cost of the proposed strategy is approximately 4,307 million USD. The duration of the fuel storage and the investment cost, particularly the variable investment cost, directly affect the choice of facility storage.

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

Electrochemical Deposition of U and RE Elements Using the Stirred Liquid Cadmium Cathode in LiCl-KCl Molten Salts

This study was conducted in an attempt to understand the effect of a stirred liquid cadmium cathode (LCC) on the electrodeposition of U and U/RE on Cd. For this purpose, a series of electrowinning tests were performed using an LCC equipped with a Cd stirrer. Initially, three runs of the U electrodeposition tests were conducted using LiCl-KCl-UCl3 at 500°C under a constant current. From the results obtained from the initial three runs, it was found that the maximum deposited amount of U was 7.4 wt% U/Cd. U dendrite formation on the LCC crucible was not observed across each of the three runs. Three additional runs were conducted using LiCl-KCl-UCl3-RECl3 to determine the extent of U/RE electrodeposition. The maximum number of moles of U + RE metals deposited was 0.07, a value estimated to be 2.14 times higher than the solubility limits exhibited by these metals in Cd. The results of this study show that the use of a Cd stirrer significantly improves the extent of U deposition.