Conference Agenda

The IAEA has been a key supporter of the development of advanced power-reactor technologies and nuclear fuel technologies for many decades. Its efforts include providing platforms for information exchange, organizing meetings, issuing publications, coordinating research activities and maintaining databases (for advanced reactor designs, fuels, fuel cycle and post irradiation examination facilities).

This presentation will highlight the IAEA’s on-going programmes and near-term plans to support the development of new reactor technologies and advanced fuels for both operating and innovative power reactors. This includes IAEA’s efforts in developing accident tolerant and advanced technology fuels (ATF), fuels for recycling/multi-recycling of nuclear materials, and advanced fuels for GEN-IV and small modular reactors, as well as advanced reactor designs.

Special emphasis will be paid to key Coordinated Research Projects (CRPs) including “Testing and Simulation of Advanced Technology and Accident Tolerant Fuels (ATF-TS)”, “Fuel Materials for Fast Reactors”, “Standardization of Subsized Specimens for PIE and Advanced Characterization for SMR and Advanced Applications”, “Fuel Modelling Exercises for Coated Particle Fuel for Advanced Reactors Including Small Modular Reactors”, “Developing a Phenomena Identification and Ranking Table and a Validation Matrix, and Performing a Benchmark for In-Vessel Melt Retention”, and “Advancing Thermal-Hydraulic Models and Predictive Tools for Design of SCWR Prototypes”.

IAEA Member States are strongly encouraged to participate in the IAEA’s topical meetings and new CRPs, which provide unique opportunities to engage with cutting-edge reactor and fuel technologies critical to the future of nuclear energy.

CRUD is one of the major considerations from the perspective of operation and safety, especially in PWR nuclear power plants. CRUD, which consists of corrosion products in the reactor coolant system, is known to induce thermal resistance, distortion of power distribution, local corrosion, boron hide-out, etc. Some of these adverse effects imply that the CRUD can affect the plant economics and core integrity hindering the thermal limit of nuclear power plants. These kinds of challenges significantly highlight that the CRUD effect should be investigated to calculate accurate operational margin of nuclear power plants in operation as well as in development. Various academic efforts have been made to figure out the mechanical and chemistry characteristics related to the deposition mechanisms and implications, trying to reflect them on the nuclear power plants. A reliable database of CRUD is necessary for solutions since deposit experiments are too hard to generate quantitative results under high-pressure/high-temperature PWR operational conditions. Hence, this paper investigates the CRUD effects utilizing the thermal properties obtained from experiments under actual PWR conditions. In this method, the CRUD layer is arranged on the surface of the fuel clad composing an active core in a simulated PWR plant model. Based on the results, the guideline can be made to calculate the local heat flux on the nuclear fuel considering the high burn-up rate of the reactor core. Further research will be conducted for the better quality of the database, expanding the test conditions and results.

Passive safety systems are widely used in the advance nuclear reactors. As an important component in the passive safety system, the core makeup tank (CMT) plays a key role in the safety injection and the core decay heat removal during the transient process of small-break loss of coolant accident (SBLOCA) . In this study, the thermal hydraulic behaviors of the CMT injection process were investigated by simulating different accident scenarios of the ACME test facility with different break sizes which includes the 1inch，2inch，4inch and 8inch breaks. Through the comparative analysis of the transient simulation under different break conditions along with visualization results, the switching process between the two working modes of CMT and its interaction with other safety injection component (such as ACC) were studied. Moreover, the mechanism of thermal stratification in CMT and the related influencing factors of CMT safety injection were analyzed. This work can provide guidance for the safety design and performance qualification of advanced passive nuclear reactors.

Abstract – In the framework of the OECD/NEA experimental projects, like ATLAS-3, a set of DEC-A experiments were carried out aiming at assessing the nuclear power plants capabilities to deal with complex accidental scenarios and evaluating the design provisions of the safety systems and the adequacy of the accident management measures. DEC-A scenarios are generally based on events and combinations of events which may lead to severe fuel damage in the core. The safety assessments are normally carried out using best estimate tools with the objective to demonstrate the fulfilment of the safety criteria and the design robustness. In this paper, a DEC-A experimental scenario carried out in the ATLAS test facility is considered. The ATLAS C3.2 test concerns a steam line break scenario relying on the passive auxiliary feedwater and operator action to cooldown the primary system. The transient involves complex interacting natural circulation phenomena including natural circulation flow interruption, steam condensation and heat exchange in large pool. In this framework the CATHARE code is used to simulate the course of the transient and the related natural circulation phenomena. It is shown, on the one hand, that the safety features of the design together with the operator actions are capable to bring the primary system to a safe end state and on the other hand, the CATHARE code prediction capabilities, for such complex scenario, are generally good. Nevertheless, additional efforts should be carried out to enhance the simulation under passive natural circulation cooling conditions.

During a Loss of Coolant Accident (LOCA) with failure of Emergency Coolant Injection (ECI) in a Pressurized Heavy Water Reactor (PHWR), the convective cooling is compromised, leading to an increase in the fuel channel temperature. Initially, the fuel temperature rises due to the decay heat and the energy stored in the fuel. The chemical reaction between the cladding and steam further escalates the temperature, causing the cladding to embrittle from oxygen diffusion. This can result in cladding rupture and the release of fissile materials. Additionally, this reaction produces hydrogen gas, which threatens the structural integrity of the containment. The heat from the fuel bundle is transferred to the Pressure Tube (PT) and the rising temperature of the PT leads to deformation, such as ballooning, sagging, or both, due to the rapid degradation of its thermo-mechanical properties, influenced by internal pressure. Given the significant risks associated with such accidents, it is crucial to study the behavior of fuel channels under LOCA conditions. This paper investigates the thermal performance of Indian PHWR under an oxidizing environment that simulates a late-phase accident scenario. The temperature profiles of the fuel element simulators, PT, and CT under steady-state conditions are obtained. A 37-element fuel bundle simulator is used, with the PT mounted eccentrically inside the CT, maintaining a 4 mm eccentricity.