[en] Through the Poland Hungary Aid for Reconstruction of the Economy (PHARE) programme, the European Commission (EC) supported the transition of the Eastern European states to the European market economy. PHARE was a pre-accession financial assistance programme which involved countries from Central and Eastern Europe that applied to become members of the European Union. The paper presents a synthesis of the projects carried out in Romania for enhancing nuclear safety by consolidating key areas such as Regulatory Activities, Radioactive Waste Management and On-Site assistance, in order to fulfil the requirements for accession to the European Union. Statistical considerations on the impact of the projects are also proposed and an analysis of the methodology of intervention is made.

[en] This paper presents an analysis of the way in which the CANDU 6 and ACR reactors behave in case of a design basis accident (LOCA) taking into account the differences between the fuel, coolant and moderator, but the fact as well that CANDU 6 has a positive Void Reactivity Coefficient and ACR has a negative one. The code used to calculate the data needed is DRAGON, a program developed at Ecole Polytechnique de Montreal which makes use of WIMS nuclear data bases. All the data used in this program are public and the cells models are made as close to the reality as possible. One uses only these data and some approximations. We attempt to prove that the full - core coolant void reactivity may be reduced at a desired value by an accurate determination of fuel, coolant agent and moderator type and parameters. Also we will show that the alteration of full - core coolant void reactivity from a positive value (about +15 mk in CANDU 600 reactor) to a small negative value (-3 mk in Advanced CANDU Reactor (ACR)) results in an important improvement of nuclear reactor response to sudden (and unwished) increments of temperature. To solve the transit equations for different void fractions we used Dragon 3.05 code. In order to compare the way the neutron flux and the reactivity evolve in CANDU 6 and ACR cells at different void fractions a method of graphic data processing was developed. The following conclusions can be highlighted: - by using CANFLEX SEU Fuel in ACR we have the following favorable effects: - one can use H₂O as coolant agent; - it allows the reduction of moderator to reduce the coolant void reactivity (by using H₂O coolant, less than 24% of D₂O used in CANDU 600 is required); - allows the use of neutron absorber in the central fuel pin to further reduce coolant void reactivity to target of -3mk ; - one can reach to higher fuel burnup (from 7.5 MWd/kg(U) at CANDU 600 to 20.5 MWd/kg (U) at ACR, average values); - one reduces the loading rate of fuel. We also can notice unique Lost Of Cooling Agent (LOCA) features in ACR: - power in reactor core region drops upon LOCA due to negative void reactivity; - steep rise in thermal neutron flux in the reflector region due to migration and subsequent thermalization of fast neutrons from the core region

[en] This work aims to compare the effects, on thermal flux and reactivity, of introducing a Mechanical Control Absorber inside a CANDU 600 super-cell respectively inside an ACR supercell. The Mechanical Control Absorber is a reactivity control device by means of which the reactor neutrons flux is adapted to the needs of the operator. In order to determine these effects we used the DRAGON code which solved the 3-D differential transport equations giving information on neutron fluxes, absorption ratio and neutron dispersal and transport ratio. A model of supercell form of two elementary cells with a MCA bar between them was used. The study of this effect will show the way that differences concerning the fuel, the cooling agent and the moderator, affect the Mechanical Control Absorber impact into the active lattice. For studying these effects we considered three-dimensional CANDU 600 and ACR supercells, with an ensemble of 5 blocks. The first block is moderator, the second block is fuel, the third block is the absorber in case the Mechanical Control Absorber is in the supercell, while if not, its place is occupied by the moderator, the next block is fuel and the last block is moderator. The supercell model is presented. This study showed that comparing the results obtained in the case of the super-cell without absorber with the results obtained in the case of the super-cell with the absorber inside, one can conclude: - The Mechanical Control Absorber effects inside a CANDU 600 super-cell are more evident on the thermal neutron flux then on the fast neutron one; - The introduction of an absorbent rod inside a CANDU 600 super-cell has the effect of reducing the thermal neutron flux leading to a diminution of super-cell's reactivity; - The absorption ratio at a CANDU 600 super-cell with an absorbent rod inside is higher then at the super-cell without the absorber. (authors)

[en] Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study