Thermal characteristics of container for on-site irradiated nuclear fuel transportation

Paper presented at the 8th International Conference on Heat Transfer, Fluid Mechanics and Thermodynamics, Mauritius, 11-13 July, 2011.An object of analysis in this paper is the container, which was developed for transportation of irradiated RBMK-1500 nuclear fuel assemblies at the Ignalina Nuclear Power Plant (NPP). Ignalina NPP (Lithuania) comprises two Units with RBMK-1500 reactors. After the Unit 1 of the Ignalina Nuclear Power Plant was shut down in 2004, approximately 1000 fuel assemblies from Unit 1 were safely transported and reused in the reactor of Unit 2 before final shutdown of the Unit 2 reactor in 2009. The RBMK reactor is continuously reloaded at power. Therefore the reactor core contains fuel assemblies with different burn-up level. After permanent reactor shutdown hundreds of fuel assemblies in the reactor core have considerably less burn-up than their design value. Such fuel assemblies have high energetic potential and can be reused. The fuel-transportation container, vehicle, protection shaft and other necessary equipment were designed in order to implement the process for on-site transportation of Unit 1 Fuel Assemblies for reuse in the Unit 2. The developed equipment can be used also in decommissioning phase for fuel transportation to fuel storage facilities. The set of this equipment can be applied for NPP-s with RBMK type reactors. The structural integrity, thermal, radiological and nuclear criticality safety calculations were performed to assess the acceptance of the proposed set of equipment. The purpose of this paper is to present the results of thermal analysis of new developed container, which was used for transportation of irradiated RBMK-1500 nuclear fuel assemblies. Using finite element code the irradiated fuel transportation container model was developed and influence of an environment temperature and influence of different axial fuel power density profiles over container temperatures field was determined. Performed analysis demonstrated that the temperatures in proposed nuclear fuel transportation container do not exceed acceptance limits for both normal operation and accident conditions.mp201

Evaluation and benchmarking of gamma dose rate employing different nuclear data libraries for MCNP code at the decommissioning stage of Ignalina NPP

A comparative study was performed to reveal the differences of three nuclear data libraries for gamma dose rate calculations when applied to heterogeneous environment in the case of decommission of the Ignalina Nuclear Power Plant (INPP). The following libraries were investigated by employing the Monte Carlo n-particle transport code (MCNP): ENDF/B-VII, JEFF-3.1 and JENDL-3.3, based on the experiments performed for gamma radiation dose rate measurements inside the emergency core cooling system (ECCS) tank with surface radioactive contamination up to 54 Bq/cm2. MCNP precise simulation and the benchmark between the libraries highlighted the differences of results for the selected case of this investigation. The results revealed that the ENDF library is trustworthy for various dose and shielding calculations and similar applications since it showed a statistically satisfied agreement between the simulation results and experimental data