• Nuclear Engineering and Technology
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• 제50권3호
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• pp.504-511
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• 2018

In Lithuania, all spent nuclear fuel (SNF) resulted from the operation of the Ignalina Nuclear Power Plant (INPP), which had two Russian Acronym for "Channelized Large Power Reactor"-type reactors. After the final shutdown, the total amount of SNF at the INPP was approximately 22,000 fuel assemblies. All these assemblies will be stored for about 50 years and disposed of after that. The decision to shut down and decommission both reactors in Lithuania before termination of design period raises a significant challenge for the treatment of accumulated SNF. Therefore, various techniques and technologies for SNF management were developed and justified for that specific case, and a set of special equipment was installed at the INPP, the effectiveness of which was demonstrated during its operation. This article presents unique techniques related to the management of SNF adopted and commissioned at the INPP after its operation shutdown, namely fuel rod cladding leak tightness control system and special equipment for collection of possible spillage during handling of SNF assembly in the hot cell. The operational experience and measurement results of fuel rod cladding leak tightness control system are presented.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1869-1877
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• 2021

This paper presents the neutron dose rate analysis of the new CONSTOR® RBMK-1500/M2 storage cask intended for the spent nuclear fuel storage at Ignalina Nuclear Power Plant in Lithuania. These casks are designed to be stored in a new "closed" type interim storage facility, with the capacity to store up to 202 CONSTOR® RBMK-1500/M2 casks. In 2016 y, the "hot trials" of this new facility were conducted and 10 CONSTOR® RBMK-1500/M2 casks loaded with the spent nuclear fuel were transported to the dedicated storage places in this facility. During "hot trials", the dose rate measurements of the CONSTOR® RBMK-1500/M2 casks were performed as the dose rate is one of the critical parameter to control and it must be below design (and safety) criteria. Therefore, having the actual data of the spent nuclear fuel characteristics, the neutron dose rate modeling of the CONSTOR® RBMK-1500/M2 cask loaded with this particular fuel was also performed. Neutron dose rate modeling was performed using MCNP 5 computer code with very detailed geometrical representation of the cask and the fuel. The obtained modeling results were compared with the measurement results and it was revealed, that modeling results are generally in good agreement with the measurements.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.234-243
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• 2022

This study aims to characterize the irradiated RBMK-1500 nuclear graphite in terms of both structural and radiological properties. The experimental results of morphological and structural analysis of the irradiated graphite samples by using SEM, Raman spectroscopy as well as the theoretical evaluation of primary displacement damage are presented. Moreover, the experimental and theoretical evaluation of the neutron flux is provided and the presence of several γ emitters in the analyzed graphite samples is assessed. Furthermore, the improved version of rapid analysis method for ¹⁴C activity determination is applied and the experimentally obtained results are compared with calculated ones. Results indicate that structural changes are uniform enough in all the analyzed samples. However, the distribution of radionuclides is non-homogeneous in the irradiated RBMK-1500 reactor graphite matrix. The comprehensive understanding of both structural and radiological characteristics of nuclear graphite is very important when dealing with decision about irradiated graphite waste management strategy or treatment options prior to its final disposal.