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.Papers presented to the 12th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Costa de Sol, Spain on 11-13 July 2016

Simulation of hydrogen distribution and effect of Engineering Safety Features (ESFs) on its mitigation in a WWER-1000 containment

In this study, thermal-hydraulic parameters inside the containment of a WWER-1000/v446 nuclear power plant are simulated in a double-ended cold leg accident for short and long times (by using CONTAIN 2.0 and MELCOR 1.8.6 codes), and the effect of the spray system as an engineering safety feature on parameters mitigation is analyzed with the former code. Along with the development of the accident from design basis accident to beyond design basis accident, the Zircaloy-steam reaction becomes the source of in-vessel hydrogen generation. Hydrogen distribution inside the containment is simulated for a long time (using CONTAIN and MELCOR), and the effect of recombiners on its mitigation is analyzed (using MELCOR). Thermal-hydraulic parameters and hydrogen distribution profiles are presented as the outcome of the investigation. By activating the spray system, the peak points of pressure and temperature occur in the short time and remain below the maximum design values along the accident time. It is also shown that recombiners have a reliable effect on reducing the hydrogen concentration below flame propagation limit in the accident localization area. The parameters predicted by CONTAIN and MELCOR are in good agreement with the final safety analysis report. The noted discrepancies are discussed and explained

On the use of boundary conditions and thermophysical properties of nanoparticles for application of nanofluids as coolant in nuclear power plants; a numerical study

In the first part of the present study, a thermal-hydraulic subchannel code hereafter called \u2018SUBTHAC\u2019 is developed to evaluate the enhancement effects of nanoparticles in core heat transfer. The first version of SUBTHAC (V1.0) can analyze the steady state flow of coolant with Al2O3, TiO2 or CuO as nanoparticles (other types of nanoparticles can be added by the user). Different output profiles can be selected such as fluid temperature, pressure and velocity for each subchannel, clad outside temperature for each fuel rod, axial and lateral mass flow, etc. SUBTHAC uses a dedicated algorithm to solve the subchannel equations and, unlike many other codes, allows for thermophysical parameters of nanoparticles to be a function of the temperature, leading to improvement the accuracy of results. Results computed by SUBTHAC for base fluid (pure water) are validated against those obtained by COBRA-EN code. In the next step, with the aim of validating the capability of nanofluid analysis of SUBTHAC code, its nanofluids results have been validated against reference CFD simulations. After the validation, comprehensive numerical comparisons are conducted to assess the enhancement of thermal-hydraulic parameters by using nanofluids. It is shown that, among Al2O3, TiO2 and CuO nanofluids with volumetric concentration in the range of 1\u20135%, TiO2-3% and CuO-3% are the best choices to increase fluid outlet temperature and decrease clad temperature, respectively. Using nanofluids with a concentration higher than 3% volumetric is not justifiable as the core pressure drop increases up to more than 20%. In the second part of the manuscript, some relevant remarks are put forward on the assignment of boundary conditions (BC, i.e. inlet velocity/inlet mass flux/inlet Reynolds number) and the adoption of reliable values for specific heat capacity of nanoparticles in operational temperature of NPPs. The effects of using the above boundary conditions and incorrect values of the specific heat (as adopted in the literature so far) are depicted by presenting some profiles of coolant and clad temperature. Selecting different BCs and incorrect values of specific heat for nanoparticles can jeopardize the results of calculations

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

Steam generators are one of the most important components of pressurized-water reactors. 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, the 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 have been proved

Dose assessment for emergency workers in early phase of Fukushima Daiichi nuclear power plant accident

In the case of Fukushima Daiichi nuclear power plant (FNP) accident, the radioactive material was released from reactor units 1–3 and transported to short and long distances due to the atmospheric pathways-motions. Power sources for monitoring posts were lost due to earthquake and tsunami. Based on air dose rates and other data measured by monitoring cars, the amount of radioactive material released to the atmosphere from the power station was obtained. The atmospheric dispersion and the transport model used in the RASCAL code, estimate the radionuclide concentrations downwind, both in the air and on the ground due to deposition. The calculated concentrations are then used to estimate the projected doses for workers in vicinity of the accident area in the first minutes of accident time. For dose modeling, we assumed that each worker was 15 min in vicinity of FNP in accident situation, once without and once with protective clothes or respirator. According to Tokyo Electric Power Company (TEPCO) report six workers had received doses over 250 mSv (309 to 678 mSv) apparently due to inhaling Iodine-131 fume. In this paper the calculated dose results using RASCAL code shows that, if emergency workers who work in early phase of accident had not used protective equipment, for 15 min, inhalation doses from iodine in their thyroid gland up to 12 March afternoon would have been 520 mSv. A comparison between calculation results and TEPCO report shows that dose calculated virtually is nearly equal to TEPCO measurement results

Full scope simulation of VVER-1000 blowdown source and containment pressurization in a LBLOCA by parallel coupling of TRACE and CONTAIN

Nuclear power plants containment plays an important role as last-defined barrier in defense in depth approach against the release of radioactive material to the environment. In this study, a parallel processing couple has been developed to full scope analysis of blowdown source and containment pressurization parameters in a LBLOCA accident. To achieve this goal, primary and secondary loops of a VVER-1000/V446 were first simulated in TRACE V5.0 and steady-state results have been validated against reference data. The second step deals with containment simulation in CONTAIN 2.0 with new modified 30-cells models. A parallel processing interface was developed in MATLAB to couple TRACE and CONTAIN in the break point. Containment average pressure has been fed back to TRACE as forcing function of blowdown source in each time step during pressurization phase (coupling point). Finally, results of blowdown and containment pressurization have been validated against final safety analysis report (FSAR). Results of simulation confirm that the maximum containment pressure can reach 0.36 MPa and 0.395 MPa for this study and FSAR respectively that are lower than the maximum design absolute pressure of 0.46 MPa, so containment maintains its integrity during this accident. Temperature profiles of different control volumes inside containment during accident follow the FSAR profiles in terms of shape and value that show the ability of developed parallel coupling to full scope simulation of accidents accurately