[en] Genetic Algorithm (GA) based systems are used for search and optimization problems. There are several applications of GAs in literature successfully applied for loading pattern optimization problems. In this study, we have selected loading pattern optimization problem of Pressurised Water Reactor (PWR). The main objective of this work is to evaluate the performance of Genetic Algorithm operators such as regional crossover, crossover and mutation, and selection and construction of initial population and its size for PWR loading pattern optimization problems. The performance of GA with antithetic variates is compared to traditional GA. Antithetic variates are used to generate the initial population and its use with GA operators are also discussed. Finally, the results of multi-cycle optimization problems are discussed for objective function taking into account cycle burn-up and discharge burn-up.(author)

[en] In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1^st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

[en] A passive system should be theoretically more reliable than an active one. In fact its operation is independent by any external input or energy and is relied only upon natural physical laws (e.g., gravity, natural circulation, etc.) and/or 'intelligent' use of the energy inherently available in the system (e.g., chemical reaction, decay heat, etc.). Nevertheless the passive system may fail its mission as consequences of component failures, deviation of physical phenomena, boundary and/or initial conditions from the expectation. This document describes at first the methodology developed by ENEA, in collaboration of University of Pisa and Polytechnic of Milano, allowing the evaluation of the reliability for a passive system, which operation is based on moving working fluids (type B and C, cf. IAEA). It reports the results of an exercise performed on a system, which operation is based on Natural Circulation.(author)

[en] This paper presents risk and sensitivity analysis of external events impacts on the safe operation in general and in particular the Ignalina Nuclear Power Plant safety systems. Analysis is based on the deterministic and probabilistic assumptions and assessment of the external hazards. The real statistic data are used as well as initial external event simulation. The preliminary screening criteria are applied. The analysis of external event impact on the NPP safe operation, assessment of the event occurrence, sensitivity analysis, and recommendations for safety improvements are performed for investigated external hazards. Such events as aircraft crash, extreme rains and winds, forest fire and flying parts of the turbine are analysed. The models are developed and probabilities are calculated. As an example for sensitivity analysis the model of aircraft impact is presented. The sensitivity analysis takes into account the uncertainty features raised by external event and its model. Even in case when the external events analysis show rather limited danger, the sensitivity analysis can determine the highest influence causes. These possible variations in future can be significant for safety level and risk based decisions. Calculations show that external events cannot significantly influence the safety level of the Ignalina NPP operation, however the events occurrence and propagation can be sufficiently uncertain.(author)

[en] The gradual reconstruction of the Bohunice V1 nuclear power plant (Slovakia) represents the most extensive reconstruction of a nuclear power plant in operation as implemented worldwide up to now. Extensive reconstruction works in both civil construction and process parts, in instrumentation and control part, and in electric part enhanced both nuclear safety and operational reliability of Bohunice V1 in a significant way.(author)

[en] The paper describes the measurement of the delayed signal of a Rh emitter Self Powered Neutron Detector (SPND) separately from other signal components originating from (n-gamma-e), (background gamma-e) and other effects. In order to separate the delayed signal, the detector was removed from the reactor core and placed to an adequately distant location during the measurement, where the radiation from the core was negligible. The experiment was carried out on the 100kW light water tank-type reactor of Technical University of Budapest and the results of the measurement were compared with the results of Monte Carlo calculations.(author)

[en] Quenching is always present in water samples and the degree of quenching can vary from one sample to another even within the same batch. This means that quench correction should be carried out for each sample in order to determine the activity so that comparisons can be made between samples and other batches. In this paper, a comparative study of tritium measurements between two methods used to correct for quenching is presented. The methods used to determine counting efficiency in the presence of quenching are as follow: Spectral Quench Parameter of External standard method (SQP(E)) and Internal Standard Method. In this work, a low background liquid scintillation system detector (Quantulus 1220) is used to determine tritium activity concentration in heavy water with different concentration from 99,66 D/H+D% to 1,65 D/D+H %. A standard calibration curve for the SQP(E) technique has been carried out with ³H low level quenched PACKARD standard set that had an assayed value of 29,240 dpm/std + 1,6%. Quench correction for Internal Standard Method has been made for each sample of heavy water with Tritiated Water Internal Standard that had a tritium concentration of 2,51 x 10⁶ dpm/g + 3,0%. A comparison between dilution factors, calculated both for D/(D+ H)% concentration and for tritium activity, measured by the two methods, is discussed in this paper. Internal Standard Method provides accurate results especially for lower D/(D+H)% concentration, which is the environmental samples case. Commercial standards set don't fulfill the exigency of an accurate environmental tritium measurement. It must consider the following problems: type of vial, type of scintillant, filled volume and counting geometry. Even if you make your own standard set for quenching calibration you must exceed another problem. The Compton electrons produced by external standard are energetic beta particles and do not have the same behavior of weak beta-particles in the sample itself. Hence a small amount quenching in tritium sample may remain undetected.(author)

[en] By Governmental decree of May 1999, the Agency for radwaste management (ARAO) was authorized as a state public service for managing radioactive waste from small producers. By this decree the ARAO also became the operator of the Central Interim Storage intended for radioactive waste from industry, medicine and research, located in Brinje near Ljubljana. In this paper the current situation will be presented, together with plans for improving public service and the necessary refurbishment and modernization of the storage facility. Execution of the proposed measures, modifications and a modernization will ensure proper and safe storing of all radioactive waste from small producers produced in Slovenia, thus fulfilling the requirements for full operation of the public service of radioactive waste management.(author)

[en] The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be ⁶⁰Co and ¹⁵²Eu and in barytes concrete samples ⁶⁰Co, ¹⁵²Eu and ¹³³Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

[en] Transport calculations can often provide useful information about irradiation field characteristics in addition to the measurements. For this reason Monte Carlo simulations were used to characterise irradiation facilities in the TRIGA Mark II (Ljubljana, Slovenia) reactor core. Several locations in two different core configurations were investigated. Some of the Monte Carlo results were also compared to the measurements. It was found out that the flux as well as its spectral characteristics depend very much on the position of the irradiation channel in the core and can vary significantly between different core configurations. The highest flux is achieved in the central channel of the core, and then decreases to about 10% in the rotary groove. The fast-to-thermal neutron flux ratio also varies significantly and achieves values of about 1.4 in the central channel but only 0.2 in the rotary groove. The core configuration can introduce up to 40% change in the fast-to-thermal neutron ratio, which enables optimisation of the core configuration for different irradiation channels and different irradiation goals. Comparison to the measurements showed that our Monte Carlo model can successfully predict absolute values of the flux as well as its spectral characteristics to within a few percent.(author)