Numerical Simulation and Validation for Early Core Degradation Phase under Severe AccidentsRead the full article
Science and Technology of Nuclear Installations publishes research on issues related to the nuclear industry, particularly the installations of nuclear technology, and aims to promote development in the area of nuclear sciences and technologies.
Professor Michael I. Ojovan is the Chief Editor of the journal, and is currently based at the University of Sheffield, UK. He is known for many innovations in nuclear research, including metallic and glass-composite materials for nuclear waste immobilisation.
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The Influence of Radiation on Confinement Properties of Nuclear Waste Glasses
Self-irradiation can affect durability of glasses used to immobilize high-level nuclear waste (HLW). The stability of glasses can also be indirectly affected by the radiolytic changes in contact water leading to decrease in its pH although this is unlikely to occur for disposal systems where the interaction of groundwater with glass is delayed to times when radiation dose rates are decreased to levels insignificant to cause such effects. Besides, interaction of the water influenced by radiation with other repository protective elements (container and bentonite) will suppress the radiolytic changes. Literature analysis shows practical absence or very weak effect of self-irradiation on structure and characteristics of borosilicate glasses with typical content of nuclear waste. Data for aluminophosphate glass used in Russia have showed that, after γ-irradiation with a dose of 6.2·107 Gy, the leaching rates of elements were decreased approximately twice relatively to pristine samples.
Best-Estimate Plus Uncertainty Framework for Multiscale, Multiphysics Light Water Reactor Core Analysis
Tremendous work has been done in the Light Water Reactor (LWR) Modelling and Simulation (M&S) uncertainty quantification (UQ) within the framework of the Organization for Economic Cooperation and Development (OECD)/Nuclear Energy Agency (NEA) LWR Uncertainty Analysis in Modelling (UAM) benchmark, which aims to investigate the uncertainty propagation in all M&S stages of the LWRs and to guide uncertainty and sensitivity analysis methodology development. The Best-Estimate Plus Uncertainty (BEPU) methodologies have been developed and implemented within the framework of the LWR UAM benchmark to provide a realistic predictive simulation capability without compromising the safety margins. This paper describes the current status of the methodological development, assessment, and integration of the BEPU methodology to facilitate the multiscale, multiphysics LWR core analysis. The comparative analysis of the results in the stand-alone multiscale neutronics phase (Phase I) is first reported for understanding the general trend of the uncertainty of core parameters due to the nuclear data uncertainty. It was found that the predicted uncertainty of the system eigenvalue is highly dependent on the choice of the covariance libraries used in the UQ process and is less sensitive to the solution method, nuclear data library, and UQ method. High-to-Low (Hi2Lo) model information approaches for multiscale M&S are introduced for the core single physics phase (Phase II). In this phase, the other physics (fuel and moderator), providing feedback to neutronics M&S in a LWR core, and time-dependent phenomena are considered. Phase II is focused on uncertainty propagation in single physics models which are components of the LWR core coupled multiphysics calculations. The paper discusses the link and interactions between Phase II to the multiphysics core and system phase (Phase III), that is, the link between uncertainty propagation in single physics on local scale and multiphysics uncertainty propagation on the core scale. Particularly, the consistency in uncertainty assessment between higher-fidelity models implemented in fuel performance codes and the rather simplified models implemented in thermal-hydraulics codes, to be used for coupling with neutronics in Phase III is presented. Similarly, the uncertainty quantification on thermal-hydraulic models is established on a relatively small scale, while these results will be used in Phase III at the core scale, sometimes with different codes or models. Lastly, the up-to-date UQ method for the coupled multiphysics core calculation in Phase III is presented, focusing on the core equilibrium cycle depletion calculation with associated uncertainties.
Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations
Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.
Source Term Derivation and Radioactive Release Evaluation for JRTR Research Reactor under Severe Accident
The source term for the JRTR research reactor is derived under an assumed hypothetical severe accident resulting in generation of the most severe consequences. The reactor core is modeled based on the reactor technical design specifications, and the fission products inventory is calculated by using the SCALE/TRITON depletion sequence to perform burnup and decay analyses via coupling the NEWT 2-D transport lattice code to the ORIGEN-S fuel depletion code. Fifty radioisotopes contributed to the evaluation, resulting in a source term of 3.7 × 1014 Bq. Atmospheric dispersion was evaluated using the Gaussian plume model via the HOTSPOT code. The plume centerline total effective dose (TED) was found to exceed the IAEA limits for occupational exposure of 0.02 Sv; the results showed that the maximum dose is 200 Sv within 200 m from the reactor, under all the weather stability classes, after which it starts to decrease with distance, reaching 0.1 Sv at 1 km from the reactor. The radiation dose plume centerlines continue to the exceed international basic safety standards annual limit of 1 mSv for public exposure, up to 80 km from the reactor.
Effect of Deposit on the Evaporation Rate of Adhered Salt in Uranium Dendrite
Electrorefining is a key step in pyroprocessing. The solid cathode processing is necessary to separate the salt from the cathode of the electrorefiner since the uranium deposit in a solid cathode contains electrolyte salt. Moreover, it is very important to increase the throughput of the salt separation system due to the high uranium content of the spent nuclear fuel and high salt fraction of uranium dendrites. Therefore, in this study, the effect of deposit on the evaporation of the adhered salt in a uranium deposit was investigated by using the samples of salt in the uranium deposit and salt in the deposit of the surrogate material for the effective separation of the salt. It was found that the salt evaporation rate is dependent on the deposit type and bulk density in the crucible. Additionally, the evaporation rate was found to be lower when the deposit structure is complex; the rate also decreases as the bulk density of the deposit is increased owing to the retardation of the salt vapour transport process. It was concluded that the mass transfer of the salt vapour is an important parameter for the achievement of a high throughput performance in the salt distillation process.
Numerical Investigation of Water Film Evaporation with the Countercurrent Air in the Asymmetric Heating Rectangular Channel for Passive Containment Cooling System
Passive containment cooling system (PCCS) is an important passive safety facility in the large advanced pressurized water reactor. Using the physical laws, such as gravity and buoyancy, the water film/air countercurrent flow is formed in the external annular channel to keep inside temperature and pressure below the maximum design values. Due to the large curvature radius of the annular channel, one of the short arc segments is taken out, as a rectangular channel, to analyze the main water film evaporation heat transfer characteristics. Two numerical methods are used to predict the water film evaporative mass flow rate during the heat transfer process in the large-scale rectangular channel with asymmetric heating when the water film temperature is not saturated. At the same time, these numerical simulation results are validated by the experiment which is set up to study water film/air countercurrent flow heat transfer on a vertical back heating plate with 5 m in length and 1.2 m in width. It is shown that the maximum deviation between numerical simulation and experiment is 30%. In addition, the influences on these parameters, such as heat flux, evaporative mass flow rate, and water film thickness, are evaluated under the different tilted angles of the rectangular channel and horizontal plane, water/air inlet flow rates, water/air inlet temperatures, heating surface temperatures, and air inlet relative humidities. All these results can provide a good guidance for the design of PCCS in the future.