Results 1 - 10 of 11
Results 1 - 10 of 11. Search took: 0.017 seconds
|Sort by: date | relevance|
[en] Reliability represents one of the most important attributes of software quality. Assessing the reliability of software embedded in the safety of highly critical systems is essential. Unfortunately, there are many factors influencing software reliability that cannot be measured directly. Furthermore, the existing models and approaches for assessing software reliability have assumptions and limitations which are not directly acceptable for all systems, such as reactor protection systems. This paper presents the result of a study which aims to conduct quantitative assessment of the software reliability at the reactor protection system (RPS) of RSG-GAS based on software development life cycle. A Bayesian network (BN) is applied in this research and used to predict the software defect in the operation which represents the software reliability. The availability of operation failure data, characteristics of the RPS components and their operation features, prior knowledge on the software development and system reliability, as well as relevant finding from references were considered in the assessment and the construction of nodes on causal network model. The structure of causal model consists of eight nodes including design quality, problem complexity, and defect inserted in the software. The calculation result using Agenarisk software revealed that software defect in the operation of RPS follows binomial statistic distribution with the mean of 1.393. This number indicated the high software maturity level and high capability of the organization. The improvement of software defect concentration range on the posterior distribution compared with the prior's is also identified. The result achieved is valuable for further reliability estimation by introducing new evidence and experience data, and by setting up an appropriate plan in order to enhance software reliability in the RPS. (author)
[en] To fulfill the present needs in the Slovak nuclear industry detailed and precise KENO 3D model of the VVER-440/V213 reactor has been developed for criticality, shielding and detector response calculations. The model was partially validated by the criticality calculation of the real operational conditions reached on the 310th effective day in Slovak NPP Bohunice unit 4 during cycle 30. This paper investigates several modelling issues associated with VVER-440 criticality and shielding calculations using the SCALE computational system. The detailed model of the VVER-440/213 reactor was developed for criticality and shielding analyses including reactor core, core basket, core barrel, pressure vessel with all internals in an appropriate level of accuracy. To minimize the costs of criticality and fission source calculations, several truncation planes were introduced to the geometry model. Based on the keff values and associated neutron spectra, the truncation planes closest to the FAs were chosen as acceptable geometry simplification. The applied simplifications resulted in low computational bias of keff which did not exceed 0.8% in computational cases. Special attention was given to the methodology applied to the determination of the fuel isotopic vectors modelled in one-sixth symmetry core configuration what is one of the reasons of small calculation bias. The results of gamma and spontaneous fission neutron source calculations revealed that the gamma source strengths and the spectra are for different burnups almost identical. In contrary, spontaneous fission neutron source strengths decreased as function of burnups. To demonstrate the capabilities of the SCALE system and the developed VVER-440 model and by using CADIS-FW variance reduction technique, detector responses placed in different locations were successfully calculated. From the statistical point of view, the analog calculation without advanced variance reduction techniques gave no answer to asked questions, therefore the use and understanding of variance reduction techniques in Monte Carlo codes is a must and cannot be avoided. The results of the detector responses confirm the effectiveness of reactor biological protection and revealed that neutrons travel through the concentric channel between the reactor pressure vessel and the thermal insulation to the room A004 situated under the core barrel bottom. We can consider that any construction change of the thermal shield and the serpentine concrete should be carefully investigated in advance, due to the possible impact to the radiation situation under the core. (authors)
[en] Fast reactor technology aimed on improved safety and sustain ability as in accord to the goals of Generation IV candidates , places specific requirements on the validating the codes which are used for evaluation of the performance and behaviour of such systems. A new calculation benchmark has been proposed for the startup core of the Superphenix reactor. The paper gives an overview of the benchmark content, provides the core model definition for static neutronics calculations and presents preliminary results on core performance and comparisons with some experimental data, obtained during startup trials. The calculation results allowed concluding that the developed model can serve as an appropriate basis for benchmark activity. (Author)
[en] As the results show the SB-LOCA event causes strip cooling of the RPV as expected. However, the cooling strip was shown to be unstable and to oscillate over time. This oscillation could result in cyclical loading of the RPV wall and its fractures. Continued investigation into the cause of this instability and into its effects is needed. (authors)
[en] Knowledge of adjoint-weighted kinetics parameters plays an importance role in analyzing the reactor safety and in describing the transient behavior of nuclear reactors during normal operation or accidents. In this work, the capability of calculating adjoint-weighted kinetic parameters, including the effective neutron generation time, the effective delayed neutron fraction, and the ratio of these two quantities was implemented in OpenMC code based on two different methods of interpreting the physical meaning of the adjoint weighting, namely, the iterated fission probability (IFP) method and the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization (Clutch) method. The results of the implementation were benchmarked by comparison to MCNP5 calculations, Serpent2 calculations and experimental data in 10 critical configurations. (Author)
[en] There is a significant need to understand, analyse and assess moisture transport in cementitious materials exposed to elevated temperatures in order to confidently predict the behaviour and ultimately the development of damage in safety critical applications such as nuclear reactor vessels, structures exposed to fire and well bore grouts. In view of this need a rigorous and robust formulation to describe water retention curves (sorption isotherms) as a function of temperature based on the evolution of physical parameters is presented. The model formulation is successfully validated against independent sets of experimental data up to temperatures of 80 °C. It is then further validated under isothermal drying conditions and then high temperature conditions through the numerical reproduction of laboratory experiments following implementation in a fully coupled hygro-thermo-mechanical finite element model. The new formulation is found to work well under a variety of conditions in a variety of cementitious material types.
[en] Highlights: • Based on the experimental data, empirical gradient change was set as the quantified ONB criterion. • The effect of mass flux and inlet temperature on ONB was provided. • Existing ONB prediction correlations were evaluated based on the experimental data. • A new non-dimensional empirical correlation to predict ONB was developed based on the experimental data. - Abstract: Bubble nucleation itself is less important safety issue for nuclear reactor, but it can easily lead to critical thermal-hydraulic events such as OFI (Onset of Fluid Instability) or CHF (Critical Heat Flux) when a research reactor operates under atmospheric conditions. Thus, the ONB (Onset of Nucleate Boiling) margin for normal operation in research reactor is recommended. In the IAEA-TECDOC-233 report (IAEA, 1980), the ONB margin for a research reactor is recommended as well. Although the ONB margin in a research reactor is emphasized for such reasons, only a few experiments have been performed for downward flow direction in a narrow, rectangular channel. In addition, several existing ONB prediction correlations are arguably applicable to the flow boiling condition in the narrow rectangular channel because most of them are developed based on Hsu’s model, which was developed in the pool boiling cases. In the study, ONB experiments for various inlet temperature conditions and mass flux conditions were performed with increasing heat flux step by step. Based on experimental data, the effect of inlet temperature and mass flux on the wall superheat and heat flux at ONB was investigated. In addition, existing ONB prediction correlations were evaluated for predicting wall superheat and heat flux at ONB based on the experimental data. A new ONB prediction correlation was then developed for better-evaluation and was compared with other correlations.
[en] The topic of this presentation is the realization of the cold collector heterogenous weld correction of the DN 1100 pipe line connection of steam generator PGV-213 on Unit 4 of NPP Bohunice from the radiation protection point of view. The correction consisted of the complete removal of the original weld and completely new heterogenous weld was made. Measures have been taken to minimize the collective effective dose and the maximum individual effective doses in accordance with the principle of optimization in radiation protection. The collective effective dose was 91.5 man mSv, maximum single personal effective dose was 0.632 mSv, average individual cumulative effective dose was 0.726 mSv, a maximum cumulative personal effective dose was 3.921 mSv, and the number of persons working on the specified radiation work permit was 126. (authors)
[en] For the validation of the nuclide inventory determination of the burn-up code MOTIVE, which is currently under development in GRS, extensive calculations of radiochemical assay data from spent nuclear fuel of commercial pressurized water reactors and boiling water reactors have been performed in the framework of the reactor safety re-search projects RS1542 „Further Development of Modern Methods in the Field of Burn-Up Calculation“. A set of publicly available experimental data taken from the SFCOMPO 2.0 data base of OECD/NEA was chosen for this purpose. The present report documents these efforts. The different series of samples chosen for the validation calculations are described in detail and the parameters used for the calculational models as well as the assumptions made in the modelling process are documented. The calculated nuclide inventories are compared to the measured data and the deviations between these are analysed and assessed. The overall results are collected, analysed and compared to corresponding data from own validation calculations with the predecessor code KENOREST and from calculations performed by Oak Ridge National Laboratory with the code package SCALE.
[en] The present report documents the work performed in the reactor safety research project RS1542 „Further Development of Modern Methods in the Field of Burn-up Calculation, and reports on the research and development goals reached. The general aim of this project was the further development of the burn-up code MOTIVE and its validation by means of a check against radiochemical assay data. Moreover, investigations in the domain of uncertainty and sensitivity analysis for burn-up calculations particularly for systems with fast neutron spectrum were performed. Within this work, MOTIVE models for a selection of openly available post irradiation experiments (PIE) taken from the SFCOMPO 2.0 database of OECD/NEA were developed. The results of the calculation of 74 of these models were compared with the corresponding experimental data with the purpose of validating MOTIVE. Several measures were implemented in MOTIVE to allow for an independent verification of calculation results within the code. This includes the coupling of an additional neutron transport code and the provision of additional cross-section libraries based on different data evaluations. The coupling between neutronics and nuclide inventory calculation was improved by implementing different types of so-called predictor/corrector methods. Additionally, a number of additional features have been implemented in the code, including methods for calculating material properties, e. g. the calculation of moderator density or fuel temperature. In the field of uncertainty and sensitivity analysis, systematic investigations of two fuel element types for fast spectrum reactors have been performed. Moreover, an analysis of the influence of the correlations in the uncertainties of fission yields on the results of burn-up calculations and an investigation of the usability of the TENDL cross-section library for burn-up calculations have been conducted. In the frame of the Coordinated Research Project (CRP) on High Temperature Gas-Cooled Reactor (HTGR) Uncertainty Analysis in Modelling (UAM) of IAEA, neutron transport calculations and uncertainty analyses for different types of HTR concepts have been performed.