A fractional PID controller based on fractional point kinetic model and particle swarm optimization for power regulation of SMART reactor

The small modular reactor tends to drive down the price of electricity and heat, installed far from the national power grid. In order to provide the active power balance in small networks in fluctuating power demand, the output power of the reactor should be regulated. In this paper, a reactor core model for Korean integral-type small reactor, SMART, is proposed and verified in a rod ejection accident. A fractional controller intended to regulate the reactor power to chase the power demand. The particle swarm optimization has been carried out to minimize a certain cost function for step response of the original nonlinear plant. Simulation results show the excellent tracking of the desired output with practical control rod velocity and reactivity. The framework provided for the design of the FOPID shows the robust stability in the Nichols chart

Simulation of hydrogen distribution due to in-vessel severe accident in WWER-1000 NPP containment: a comparison of CONTAIN and MELCOR codes results

During a severe accident or Beyond Design Basis Accident (BDBA), the reaction of water with zirconium alloy as fuel clad, radiolysis of water, molten corium-concrete interaction (MCCI) and post-accident corrosion can generate a source of hydrogen. In the present work, hydrogen distribution due to in-vessel reaction (between zircaloy and steam) has been simulated inside a WWER-1000 reactor containment. In the first step, the thermal hydraulic parameters of containment have been simulated for a DECL (Double Ended Cold Leg) accident (DBA phase) in both short and long time and the effects of spray as Engineering Safety Features (ESFs) on mitigating the parameters have been studied. In the second step, it has been assumed that the accident developed into an in-vessel core melting accident. While in pre-phase of core melting (severe accident phase), hydrogen will be produced as a result of zircaloy and steam reaction (BDBA phase), the hydrogen distribution has been simulated for 23 cells inside the reactor containment by using CONTAIN 2.0 (Best estimate code) and MELCOR 1.8.6 codes. Finally, the results have been compared to FSAR results. As it can be seen from the comparisons, both CONTAIN and MELCOR codes can predict the results in good agreement with FSAR (ANGAR code) results. CONTAIN shows peak pressure around 0.36 MPa in short-term and this amount is about 0.38 and 0.4 MPa for MELCOR and ANGAR (FSAR) results respectively. All these values are under design pressure that is around 0.46 MPa. Cell 20 has the maximum mole fraction of hydrogen in long-term about 9.5% while the maximum amount of hydrogen takes place in cell 22. The differences between the results of codes are because of different equations, Models, Numerical methods and assumptions that have been considered by the codes. The simulated Hydrogen Distribution Map (HDM) can be used for upgrading the location of HCAV systems and Hydrogen Mitigator features (like the recombiners and ignitors) inside the containment to reduce the risk of hydrogen explosion

Micromechanical modeling of neutron irradiation induced changes in yield stress and electrical conductivity of zircaloy

A micromechanical model of the changes in electrical conductivity of zircaloy induced by neutron irradiation is developed. The approach is illustrated on example of High Flux Isotope Reactor (HFIR). First, HFIR reactor core has been simulated by MCNPX v2.4 code and neutron flux-energy spectrum has been extracted. The spectrum is used as the main input in SPECTER code to calculate the displacement per atom (dpa), primary knock out atoms (pka) distribution, spectral averaged kerma (MACKLIB), etc. in dependence on radiation time. These results are used to estimate variation of the yield stress of zirconium as function of radiation time applying experimental dependence of the yield stress on dpa (available in literature). Finally, the cross-property connection between yield stress and electrical conductivity is applied to estimate the latter as a function of the irradiation time

Calculation of control rod worth and temperature reactivity coefficient of fuel and coolant with burn-up changes for VVRS-2MWth nuclear reactor

One of the main issues of a nuclear reactor is safety and controlling system. This system is designed to control the power of reactor and to prevent accidents. Control rods worth calculation is used to specify safety margin of reactor. Temperature reactivity coefficients of fuel and coolant are one of the inherent factors that can control reactor power. This study has been done about control rod worth calculation and temperature reactivity coefficients of a VVRS-2 MWth reactor. The reactor core has been simulated by using WIMSDB5 and CITATION-LDI2 codes to perform neutronic calculations. The WIMSDB5 code solves neutron transport equation and obtains macroscopic cross-sections (Winfrith, 1982). The CITATION-LDI2 code solves diffusion equation for the reactor core (Winfrith, 1972). These two codes were linked by an interface that has been programmed with DELPHI (Winfrith, 1998). The rod worth and reactivity coefficients are calculated considering burn-up calculations. The burn-up calculation is performed with WIMSDB5 code. The new macroscopic cross-sections were used in a new condition. Control rods worth and temperature coefficients are calculated with these models. The obtained results of control rods worth are compared with some experimental results; and the reliability of the method was confirmed. Reactivity coefficients results are compared with the results that were obtained from the simulation model. The simulation model has used Point Kinetic equations and single heated channel model respectively for neutronic and thermo hydraulic calculations (Kazimi et al., 1990)

Radiological dose assessment for the hypothetical severe accident of the Tehran Research Reactor and corresponding emergency response

In this study, the radiological dose assessment for the most severe hypothetical accident of Tehran Research Reactor was carried out. According to the calculation results, corresponding protective actions for emergency response were obtained. The results showed that the largest distance from the reactor requires sheltering, or evacuation response is 2000 m for the E and D atmospheric stability class and rainy weather condition. Also, maximum distance requiring iodine prophylaxis is 400 m, for the A atmospheric stability class and calm weather condition

A proposed improvement for the design of safety injection system in VVER-1000/V446 reactor

If the Total Loss of Feedwater (TLFW) accident occurs in a VVER reactor, the primary side Feed & Bleed (F&B) recovery strategy prevents core damage by reducing the reactor's pressure by opening the pressurizer's safety valves and establishing core cooling by the Emergency Core Cooling System (ECCS). Although the strategy saves the core, the leaked coolant together with the containment spray system pollutes the containment environment, deactivates electrical devices, and jeopardizes the containment integrity by the release of the generated hydrogen during the core re-flooding. Further, Steam Generators (SGs) would be exposed to excessive thermal stress because of drying out, making use of recovered feedwater too risky because of the primary to secondary side leakage possibility. ECCS in VVER-1000/V446 has low-pressure Hydro accumulators (HAs) that could carry out secondary side F&B through SGs. This feasibility study uses Deterministic Safety Analysis (DSA) to show that the secondary side F&B recovery strategy can provide 6 h after losing Emergency Feedwater (EFW) which is a significant time to recover the failed safety systems. This strategy is more significant, especially if the TLFW accident is followed by the loss of A.C. power, which makes the primary side F&B ineffective, leading to the core meltdown within 2 h. The Standardized Plant Analysis Risk-Human Reliability Analysis (SPAR-H) is used to calculate Human Error Probability (HEP) for the primary and secondary side F&B recoveries. Probabilistic Safety Assessment (PSA) reveals that the proposed recovery strategy reduces the Core Damage Frequency (CDF) by one order of magnitude

Evaluation of single heated channel and subchannel modelling on a nuclear once through steam generator (OTSG)

Steam generator is one of the most important components of pressurized-water reactor. This component plays the role of heat transfer and pressure boundary between primary and secondary side fluids. The Once Through Steam Generator (OTSG) is an essential component of the integrated nuclear power system. In this paper steady state analysis of primary and secondary fluids in the Integral Economizer Once Through Steam Generator (IEOTSG) have been presented by Single Heated Channel (SHC) and subchannel modelling. Models have been programmed by MATLAB and FORTRAN. First, SHC model has been used for this purpose (changes are considered only in the axial direction in this model). Second, subchannel approach that considers changes in the axial and also radial directions has been applied. Results have been compared with Babcock and Wilcox (B&W) 19-tube once through steam generator experimental data. Thermal-hydraulic profiles have been presented for steam generator using both of models. Accuracy and simplicity of SHC model and importance of localization of thermal-hydraulic profiles in subchannel approach has been proved