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Experience Gained during the Decommissioning of the SSEC from the VVR-S Nuclear Research Reactor

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Deju R., Iorga I., Gurau D.
Acta Physica Polonica A
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2017
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vol. 131
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issue 1
82-88
EN
The VVR-S nuclear research reactor owned by Horia Hubulei National Institute of Physics and Nuclear Engineering was designed for research and radioisotope production being permanently shut-down in 2002 after 40 years of operation. The decommissioning of the reactor started in 2010 and is planned to be finalized in 2020, being the first nuclear research reactor that has started the decommissioning project from the South-East of Europe. This paper presents the lessons learned from the decommissioning of systems, structures, equipment and components from pumps room. The experiences gained cover a diverse range of topics from planning, approval, radiological characterization process made, dismantling, equipment used, resources, etc. Radiation protection and economics are the key factors for the selection of technology for decommissioning. The evaluation of the available technology has been made and areas for improving the technology have been identified. Dedicated and advanced technologies were chosen to perform specific tasks. The project was planned well in advance. Safety aspects have been taken into consideration. Resource constraints have also been an important issue considered in defining the decommissioning strategy. Experiences and knowledge were presented and shared with the community.
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Role of Operational Radioprotection Simulation in the Dismantling of the Protection and Control Rods of VVR-S Reactor

100%
Iorga I., Deju R., Gurau D., Mincu M.
Acta Physica Polonica A
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2017
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vol. 131
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issue 3
514-518
EN
The VVR-S nuclear research reactor owned by Horia Hubulei National Institute of Physics and Nuclear Engineering, has functioned between 1957 and 1997 at a nominal thermal power of 2 MW, using less-enriched nuclear fuel (10%), type EK-10, and highly enriched fuel (36%), type S-36. The reactor control was carried out by means of nine rods placed in the central part of the core. Four rods were used for the manual control of power, one for automatic control, three rods for reactor safety (emergency rods) and one for the fine control of power. One of the high hazard operations was the dismantling of the activated-contaminated control rods. Because of a higher radiological risk, simulation was performed to determine the expected maximum dose. The maximum gamma dose rate obtained using simulated results was 3.5 Sv/h, for the highly activated rod. In the real measurement, when the work was performed, the maximum gamma dose rate was 3 Sv/h. ⁹⁰Sr+⁹⁰Y are expected to be present in the control rods radioactivity.
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