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Croatian islands, with strong natural wind and solar potential, are possible locations for extensive usage of renewable energy sources. This paper presents the results of the preliminary investigation of the applicability of renewable energy sources for the zero-energy island concept on Croatian islands, in particular for the central Dalmatian island of Vis.

Considering the necessity and future role of nuclear energy as relevant to the climate problem, we have focused on the period to the year 2065. For quantification of the required emission reduction we have used IEA WEO 2009 and WEO 2011 data as presented in their Reference strategies predicting emissions with business as usual practices, and WEO 450 Energy strategies which show the time development of allowed emissions consistent with a limit on the global temperature increase of 2 ºC and the peak CO2 concentration of 450 ppm. By extrapolating these data to the year 2065 we obtain 77.4 GtCO2-eq for Reference emission and 10 GtCO2-eq for WEO 450 strategy allowing emission, resulting in 67.4 GtCO2-eq reduction required to come down to sustainable WEO 450 trajectory. The large contributions to emission reduction from fusion energy and fossil fuel with carbon separation and storage are not likely. Main carbon non-emitting sources assumed in the years up to 2065 are proven technology nuclear fission and renewable sources. In our specified strategy aimed to achieve WEO 450 target we assumed an energy mix including nuclear power build-up in the period 2025-2065 to the level of 3300 GW in 2065. With the resulting nuclear contribution of 25.2 GtCO2 to the total required emission reduction of 67.4 GtCO2, what remains for renewable sources, energy saving and increased efficiency of energy use to contribute are prodigious 42.2 GtCO2-eq. Assuming that energy saving and more efficient energy use will by 2065 effect an annual reduction between 10 to 16 GtCO2-eq, remaining 26.2 to 32.2 GtCO2, respectively 27290 and 33540 TWh would be the task for renewable energy sources. Our estimates about contribution of renewable sources going as far as 2065 are based on EREC prediction for EU and on our extension to world total with EREC and GWEC prediction as a guide. Our high, but still credible estimates of predicted world renewable energy contribution by 2065 come to the similar figures between 29260 and 36180 TWh. However, without nuclear contribution in 2065, renewable energy contribution would have to be doubled, practically impossible task in the time period in consideration. Resulting contributions by renewable sources, probably their upper limits, allow some conclusions about the role of nuclear energy in future decades. By combining highest contributions from energy saving, efficiency increase and other measures to reduce emission, apart from energy production, with highest prediction for renewable sources contribution, we obtain the minimum nuclear energy requirement of about 2190 GW in 2065. This minimum nuclear strategy should be planned and prepared for, unless there is strong evidence that other carbon free energy sources (CCS or fusion) could be developed in time. Expansion of nuclear power by about 1800 GW by 2065 would come from different and already developed industrial sector, which can give its contribution to the energy mix, without obstructing the build-up of renewable sources. It would not be wise to forfeit nuclear contribution at least in the period to 2065, critical for the control of climate change.

A few years ago Westinghouse innovations in PWR technology resulted in a proposal for a new type of integral fuel burnable absorbers containing Gadolinium. Preliminary designs of loading patterns for NPP Krško were made for 18 month fuel cycles with standard VANTAGE+ fuel type containing Integral Fuel Burnable Absorbers (IFBA) with enriched Boron, as well as newly proposed Gadolinium based absorbers. In this paper we investigate the possibility to design 24 month cycle for the NPP Krško with VANTAGE+ fuel type and Gadolinium based integral absorbers. The key fuel cycle parameters are compared to the one based on IFBAs. The analysis is performed by the latest version of FUMACS code package capable of simulating Gadolinium based integral fuel burnable absorbers.

The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals as the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working environment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 with the results of the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analyzed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be useful technique to monitor both fission source convergence (stationarity) and core eigenvalue convergence (keff) to fundamental eigenmode with MCNP5. All calculations were done with ENDF/B-VII.0 library. The obtained results showed high similarity with reference results.

The Pool Critical Assembly Pressure Vessel (PCA) benchmark is a well known benchmark in the reactor shielding community which is described in the Shielding Integral Benchmark Archive and Database (SINBAD). It is based on the experiments performed at the PCA facility in the Oak Ridge National Laboratory (ORNL) and it can be used for the qualification of the pressure vessel fluence calculational methodology. The measured quantities to be compared against the calculated values are the equivalent fission fluxes at several experimental access tubes (A1 to A8) in front, behind, and inside the pressure-vessel wall simulator. This benchmark is particularly suitable to test the capabilities of the shielding calculational methodology and cross-section libraries to predict invessel flux gradients because only a few approximations are necessary in the overall analysis. This benchmark was analyzed using a modern hybrid stochastic-deterministic shielding methodology with ADVANTG3.0.1 and MCNP6.1.1b codes. ADVANTG3.0.1 is an automated tool for generating variance reduction (VR) parameters for Monte Carlo (MC) calculations with MCNP5v1.60 code (and higher versions). It is based on the multigroup, discrete ordinates solver Denovo, used for approximating the forward-adjoint transport fluxes to construct VR parameters for the final MC simulation. The VR parameters in form of the weight windows and the source biasing cards can be directly used with unmodified MCNP input. The underlining CADIS methodology in Denovo code was initially developed for biasing local MC results, such as point detector or a limited region detector. The FW-CADIS extension was developed for biasing MC results globally over a mesh tallies or multiple point/region detectors. Both CADIS and FW-CADIS are based on the concept of the neutron importance function, which is a solution of the adjoint Boltzmann transport equation. The equivalent fission fluxes calculated with MCNP are based on several highenergy threshold reactions from international dosimetry libraries IRDF-2002 and IRDFF-2014, distributed by the IAEA Nuclear Data Section. The obtained results show a good agreement with referenced PCA measurements. Visualization of the deterministic solution in 3D was done using the VisIt code from the Lawrence Livermore National Laboratory (LLNL).

The most serious problem facing humanity is that we have only a few decades in which to implement effective measures to stop global warming. For these years up to about 2065, fission energy from light water thermal reactors is relevant as an available, developed and proven non-carbon technology with the potential to make an essential contribution to the mitigation of global warming, in addition to renewable energy. Nuclear power is expected to have more economic advantages than intermittent renewable sources for generating base load electrical energy requirements. This would be especially important in the years from about 2025 up to 2065, during which one cannot expect a serious contribution from nuclear fusion and even less from fossil fuels with carbon capture and storage (CCS) facilities. In a strategy to eliminate all non-CCS coal power stations, some 1600 MW of nuclear power would be required and sufficient to cover the base load for the electrical energy supply system. This nuclear expansion should be accompanied by effective international safety assurances, including a mandate to stop construction of unsafe nuclear power plants. In the long term, after 2065, we expect inherently safe molten salt thorium reactors to compete with fusion reactors.

Radioactive waste management is a demanding problem that is raising tensions and misunderstanding amongst the general public, experts, and policy makers in a number of countries. An investigation into the public opinion, legal framework, policy, and expert involvement on the matter of radioactive waste management has been performed in Croatia. The analysis is based on the 2016 national public opinion survey on nuclear energy and radioactive waste, and intensive examination of the changes made in Croatian legislation on radioactive waste management in the last few years. The results of the analysis indicate that hasty political decisions accompanied by a poor information sharing process and an initial low level of public knowledge on radioactive waste in general, leads to strong tensions among stakeholders. It was also shown that the role of experts and the scientific community is not clearly defined resulting in a contradictory public position towards experts and scientists, declaring them as the most trustworthy on one hand, but ignoring their expertise on the other hand. The findings of the conducted research, although specific for Croatia, could be applied in other countries to avoid similar mistakes.

In recent years, most of the growth in electricity demand is covered by renewable and nuclear energy sources. However, electricity generation in fossil-fired power plants is also increasing resulting in the increase of CO2 emissions. Nuclear energy has to be considered as one of the available tools to accomplish CO2 emission reduction in electricity sector. Light water reactors (LWR) are currently the dominant nuclear technology but their intensive application in long-term period is constrained by available uranium fuel resources. Fast breeder reactors’ (FBR) technology is not used on a larger scale. Plutonium resources are limited, but do have the potential of stronger buildup if light water reactors, as the source of plutonium, are used on a larger scale. The appropriate dynamics for LWR/FBR buildup till the end of the 21st century is developed under assumptions of different LWR life times, and different uranium fuel resources available. The possible CO2 emission reduction is calculated with World Energy Outlook 2015 development scenarios being set as reference ones. It is shown that nuclear fuel resources do not represent an obstacle for strong nuclear buildup leading to significant CO2 emission reduction. However, the reduction is mostly achieved in the second half of the century.

The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the SN part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high SN/PN parameters. Detailed voxelization (homogenization) process in Denovo together with high SN/PN parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR facility model which included only reactor core and adjacent concrete structures with the steam generator and the pump. Hybrid FW-CADIS methodology appears to be very effective for shielding calculations of complicated MC geometries. It was demonstrated that definition of air as a global adjoint source gave lower uncertainty of photon dose rates. Activation of the primary loop coolant, which becomes additional gamma source in operating reactor, was considered. Compared with analog MC simulations, the hybrid methodology drastically improved Monaco distributions in quality as well in space coverage. The SCALE6.0 hybrid stochastic methodology was thoroughly investigated and its versatile ability for deep penetration models was proved. A careful selection of parameters can relax high CPU time and hardware requirements.

A threat of global warming should convince the public to accept a nuclear fission energy contribution to climate change mitigation, at least for the climate critical years up to 2065. The nuclear fission energy is available now, with proven reactors, such as advanced LWR (light water reactors). Nuclear strategy in this paper outlines a proposal to replace all coal power plants (without carbon and capture storage system) with nuclear power plants in the period 2025-2065. Assuming once through advanced LWR technology, one would need nuclear capacity of 1, 600 GW to replace coal power plants in that period. Corresponding reduction of emission would amount to 11.8 Gt of CO2. This energy strategy would reduce carbon emission by approximately 22% in the year 2065 and would be covered by projected uranium resources. An estimation of replacement costs showed that future carbon tax has a considerable potential to offset higher costs of nuclear replacement power.

Radiation induced damage to the reactor pressure vessel (RPV) by fast neutron fluence is a main factor when considering lifetime extension for nuclear power plants. Revised guidelines with the support of computational benchmarks are needed for the regulation of the allowed neutron irradiation to reactor structures during power plant lifetime. Currently, U.S. Nuclear Regulatory Commission (U.S.NRC) Regulatory Guide 1.190 is the effective guideline for reactor dosimetry calculations, i.e. for correlation of in-core and ex-core experimental measurements with the computational results from computer codes. A well known international shielding database SINBAD contains large selection of models for that purpose, for benchmarking neutron transport methods via stochastic and deterministic computer codes. In this paper a Pool Critical Assembly Pressure Vessel Facility Benchmark (PCA benchmark) has been chosen from SINBAD for qualification of our methodology for pressure vessel neutron fluence calculations, as required by the Regulatory Guide 1.190. The SCALE6.0 (Standardized Computer Analysis for Licensing Evaluation) code package, developed at Oak Ridge National Laboratory, was used for modeling of the PCA benchmark. The CSAS6 criticality sequence of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, as well as MAVRIC/Monaco hybrid shielding sequence, was used for calculation of equivalent fission fluxes. The shielding analysis was performed using multigroup shielding library v7_200n47g derived from general purpose ENDF/B-VII.0 library. As a source of response functions for reaction rate calculations with MAVRIC we used international reactor dosimetry libraries (IRDF-2002 and IRDF- 90.v2) and appropriate cross sections from transport library v7_200n47g. The comparison of calculational results and benchmark data showed a good agreement of the calculated and measured equivalent fission fluxes.

The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by a team lead by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typically reserved for Generation III+ small modular reactors. The main safety feature is based on an integral primary circuit configuration, bringing together a compact core design with 121 fuel assemblies (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on uranium silicide fuel (U3Si2) is selected to achieve high thermal power which is extracted with PHEs placed in the annular region between the barrel and the vessel. The compact and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find the neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,γ) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of PHEs with respect to MC statistics using the FW-CADIS methodology. For that purpose the SN module Denovo, based on forward-adjoint transport theory, was used to find VR parameters (importance map and biased source) to effectively bias MC simulation. Further research is required to account for other activation pathways, i.e. isotopes of iron which may generate 59Co through neutron activation and beta decay.

In recent research we outlined climate relevant and immediately available proven light water nuclear strategy with a potential to contribute essentially and timely to reduction of carbon dioxide emission to the year 2065. We consider in this paper what is the perspective of fission energy after that year should its contribution be needed. Singling out two technologies with long term perspective which need no or small amounts of uranium, fast breeders and molten salt thorium reactors, we consider their main technical and safety characteristics. In both of these technologies it is essential to have starter nuclides as neither U238 nor Th232 are fissile. We investigated whether plutonium from spent fuel of light water reactors generated to the year 2065 would be present in quantities sufficient to continue operation on the same or similar level in both technologies. However in operational safety, proliferation risks, waste production, in our judgement we must give preference to thorium technology, if it would be ready in second half of the century.

The capabilities of the SCALE6.1/MAVRIC hybrid shielding methodology (CADIS and FW-CADIS) were demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of a full-scale PWR containment model. Automatic preparation of variance reduction (VR) parameters is based on deterministic transport theory (SNmethod) providing the space-energy importance function. The aim of this paper was to determine the neutron-gamma dose rate distributions over large portions of PWR containment with uniformly small MC uncertainties. The sources of ionizing radiation included fission neutrons and photons from the reactor and photons from the activated primary coolant. We investigated benefits and differences of FW-CADIS over CADIS methodology for the objective of the uniform MC particle density in the desired tally regions. Memory intense deterministic module was used with broad group library “v7_27n19g” opposed to the fine group library “v7_200n47g” used for final MC simulation. Compared with CADIS and with the analog MC, FW-CADIS drastically improved MC dose rate distributions. Modern shielding problems with large spatial domains require not only extensive computational resources but also understanding of the underlying physics and numerical interdependence betweenSN-MC modules. The results of the dose rates throughout the containment are presented and discussed for different volumetric adjoint sources.

Journal of Energy special issue: Papers from 11th International Conference of the Croatian Nuclear Society “Nuclear Option in Countries with Small and Medium Electricity Grids” Welcome to this special issue, which is based on selected papers presented at the 11th International Conference of the Croatian Nuclear Society “Nuclear Option in Countries with Small and Medium Electricity Grids”, held in Zadar, Croatia, on June 5th–8th, 2016. This International Conference was organized by the Croatian Nuclear Society in cooperation with International Atomic Energy Agency (IAEA), Croatian State Office for Nuclear Safety and University of Zagreb, Faculty of Electrical Engineering and Computing. The goal of the Conference was to address the various aspects of the implementation of nuclear energy for electricity production in the countries with small and medium electricity grids and in power system in general. The conference also focuses on the exchange of experience and co-operation in the fields of the plant operation, nuclear fuel cycle, nuclear safety, radioactive waste management, regulatory practice and environment protection. The conference was organized in eight main topics covered in ten oral sessions and one poster session. In three Conference days authors presented 49 papers orally and 23 papers in poster session. 102 participants came from 16 countries representing equipment manufacturers and utilities, universities and research centres, and international and government institutions. Eight invited lectures were held and 72 papers were accepted by international programme committee. The importance of international cooperation for the assessment of the nuclear option has been recognized by everybody planning to introduce nuclear power plant to the grid. That is even more important for small and medium countries having limited resources and specific requirements due to limited grid size. The Conference topics reflect some current emphasis, such as country energy needs, new reactor technologies (especially small reactors), operation and safety of the current nuclear power plants, move of the focus in nuclear safety toward severe accidents and accident management strategies, improvement in nuclear safety, reactor physics and radiation shielding calculation tools and ever increasing requirements for minimization of environmental impact. From 72 papers presented at the Conference, 16 papers were accepted for publication in this number of Journal of Energy after having undergone the additional peer-review process. We would like to thank the authors for their contributions and the reviewers who dedicated their valuable time in selecting and reviewing these papers, both during the Conference and during the preparation of this special issue of Journal of Energy. It was very challenging to collect a balanced overview of the entire Conference. We decided to select 16 papers for this issue and additional 14 for the next one. We believe that the papers which were selected for this number represent some of the best research related to nuclear plant operation, energy planning, development of new reactors and technologies, reactor physics and radiation shielding. We hope this special issue will provide a valuable insight into different aspects of nuclear and electrical engineering and reactor physics, as well as a pleasant and inspiring reading.

Lattice codes are able to produce burnup dependent fuel assembly homogenized neutron cross sections and pin power form factors. These data are used by diffusion nodal codes to produce fuel assembly powers and intra assembly power distributions. 2D transport codes are usually used for this type of the calculation. Four lattice codes (DRAGON 4, FA2D, NEWT and PSG2/Serpent) were used to calculate pin power distribution for reflected single fuel assemblies during depletion. The NPP Krsko 16 × 16 fuel assemblies without and with IFBAs (Integral Fuel Burnable Absorber), for two U-235 enrichments 4.40% and 4.80%, as found in cycle 24 of the plant, were modeled. The heterogeneous depletion calculation was performed up to burnup of 60 GWd/tU. The prediction capabilities of the codes are compared for infinite multiplication factor, some group cross sections, and for 2D pin power form factors, in order to assess their applicability for preparing cross section data for diffusion nodal codes.

Nuclear fission is receiving new attention as a developed source of carbon-free energy. A much larger number of nuclear reactors would be needed for a major impact on carbon emission. The crucial question is whether it can be done without increasing the risk of nuclear proliferation. Specifically, can a larger nuclear share in world energy production, well above the present 6%, be achieved in the next few decades without adding the proliferation-sensitive technologies of reprocessing spent fuel and recycling plutonium to the problems of the unavoidable use of enrichment technology? The answer depends on the available uranium resources. We first looked for the maximum possible nuclear build-up in the 2025–2065 period under the constraints of the estimated uranium resources and the use of once-through nuclear fuel technology. Our results show that nuclear energy without reprocessing could reduce carbon emission by 39.6% of the total reduction needed to bring the WEO 2009 Reference Scenario prediction of total GHG emissions in 2065 to the level of the WEO 450 Scenario limiting global temperature increase to 2 °C. The less demanding strategy of the nuclear replacement of all non-CCS coal power plants retiring during the 2025–2065 period would reduce emission by 26.1%.

The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employed a recently introduced machine learning technique, Support Vector Regression (SVR), which has a strong theoretical background in statistical learning theory. Superior empirical performance of the method has been reported on difficult regression problems in different fields of science and technology. SVR is a data driven, kernel based, nonlinear modelling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modelling. The starting set of experimental data for training and testing of the machine learning algorithm was obtained using a two-dimensional diffusion theory reactor physics computer code. We illustrate the performance of the solution and discuss its applicability, i.e., complexity, speed and accuracy, with a projection to a more realistic scenario involving machine learning from the results of more accurate and time consuming three-dimensional core modelling code.

The accuracy of the point-kernel method, which is a widely used practical tool for γ-ray shielding calculations, strongly depends on the quality and accuracy of buildup factors used in the calculations. Although, buildup factors for single-layer shields comprised of a single material are well known, calculation of buildup factors for stratified shields, each layer comprised of different material or a combination of materials, represent a complex physical problem. Recently, a new compact mathematical model for multi-layer shield buildup factor representation has been suggested for embedding into point-kernel codes thus replacing traditionally generated complex mathematical expressions. The new regression model is based on support vector machines learning technique, which is an extension of Statistical Learning Theory. The paper gives complete description of the novel methodology with results pertaining to realistic engineering multi-layer shielding geometries. The results based on support vector regression machine learning confirm that this approach provides a framework for general, accurate and computationally acceptable multi-layer buildup factor model.

The PWR biological shield concrete activation is analyzed using SCALE6.1 hybrid deterministic–stochastic shielding methodology. Since accurate geometry representation is a paramount step for this type of shielding calculation involving heterogeneous three dimensional regions (reactor core, thermal shield, downcomer, pressure vessel, cavity, and shield), the Monte Carlo method is the method of choice. A detailed model of a combinatorial geometry, materials and characteristics of a typical PWR reactor was based on the best available data. The sources of ionizing radiation included fission neutrons and photons originating from the HBR-2 benchmark critical core. The activation reduction of reactor biological shield materials was investigated via concrete boration, which is applied when shielding from ionizing neutron radiation is especially important. The concrete boration is the common method for activation reduction in nuclear shields since the 10B has especially high thermal neutron cross section which effectively captures thermalized neutrons and decreases their (n,γ) reaction in structural materials. The reduction of secondary gamma emissions from radiative capture of thermalized neutrons in biological shield has been examined via concrete boration with natural boron (B) and boron carbide (B4C). The possibility of neutron flux mitigation in a biological shield as well as impurity isotopes (59Co, 151Eu, and 153Eu) activity decrease was especially explored. Satisfactory boron concentration which leads to saturation of neutron flux attenuation has been proposed. The activation levels of impurities in concrete, activated above the limit of IAEA clearance for the free release limit, have been estimated. The obtained results showed that activation over IAEA limits is mostly present in a thin layer of borated shield facing the critical reactor core in cavity, while the bulk of the shield is activated below IAEA threshold. The saturation of the neutron flux attenuation was clearly demonstrated for several boron concentrations so even small amounts of boron in biological shield drastically benefit the overall flux reduction.

The basic issue in considering the contribution of nuclear power to solving the world’s energy problem in the future is the availability of uranium resources and its adequacy in meeting the future needs of nuclear capacity. Increased interest in nuclear energy is evident, and a new look into nuclear fuel resources is relevant. In this chapter we address the issue of nuclear fuel resources long term sustainability in relation to the expected growth of the world nuclear power. Three main aspects have to be analyzed in order to estimate how long the world’s nuclear fuel supplies will last: nuclear fuel resources (uranium and thorium), technologies for nuclear fuel utilization, and energy requirements growth scenarios including different scenarios for nuclear share growth. Uranium nuclear fuel resources are analyzed based on joint OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency categorization which classifies resources into conventional resources and unconventional resources. Conventional resources are further divided into identified resources (reasonably assured and inferred) with four cost ranges and undiscovered resources (prognosticated and speculative) with three cost ranges. Analyzed unconventional resources include phosphate deposits and seawater. Total resources are estimated to about 6.3 million tons of uranium in identified resources and 10.4 million tons of uranium in undiscovered resources. The amount of uranium contained in phosphate deposits and seawater is estimated to 22 million tons and 4 billion tons, respectively. Thorium nuclear fuel resources are analyzed based on slightly different categorization than uranium one. Four categories are used: reasonably assured resources, estimated additional resources of type I and II, and prognosticated resources. Total thorium resources according to the latest OECD-NEA report are estimated to about 6 million tons. Detailed analysis of potential technologies for improved nuclear fuel utilization is required in order to assess long term sustainability of nuclear fuel resources. Nowadays, thermal converter reactor technology with once-through nuclear fuel cycle is used. The effectiveness of the technology can be improved in the area of enrichment process as well as by introducing reprocessing of the spent fuel on larger scale. Other technologies are also on the development stage that allows their implementation in short or medium period of time. These include: thermal and fast breeder reactors of different kind, thorium based fuel cycle, and conversion of uranium or thorium by particle accelerators or fusion devices. Very important aspect of long term sustainability of nuclear fuel resources are scenarios for energy requirements growth, and scenarios for growth of nuclear share in electricity production resulting in overall nuclear capacity growth. On one hand, low growth scenario has a moderate continuous growth strategy of 1.3% per year. On the other hand, high growth scenarios are required if nuclear energy is to give an essential contribution to carbon emission control. Finally, there is a scenario based on a compromise between low and high growth assumptions. These scenarios have been used to estimate exhaustion time periods of nuclear fuel resources for different fuel utilization technologies. The obtained exhaustion time periods range from 50 years for the once-through thermal converter fuel cycle with high increase in world nuclear capacity to many thousands of years for breeder reactor technology or for other methods of releasing the energy of U238. A development of these methods would require several decades, in return for practically unlimited amount of nuclear energy. Our analysis shows that nuclear fuel resources are not a limiting factor for a long term large scale nuclear power development. Our study is based on the most recent data on nuclear fuel resources, energy growth predictions, and technologies for the nuclear fuel utilization improvement, as well as our own developed code for calculation of nuclear fuel demand for different nuclear energy growth scenarios.

The calculations were performed at different power levels for selected IFBA and non-IFBA NPP Krsko fuel assemblies using 2D lattice codes (DRAGON, NEWT and FA2D) to show quantitative influence of xenon concentration on homogenized cross sections. In addition to usual influence on fast and, to larger extent, on thermal absorption cross sections, similar type of influence is found in case of fission cross sections, and to some extent in fast to thermal removal cross section. Based on relative change of cross sections from the values obtained during reference depletion, burnup and power level dependent correction values were calculated and included in separate cross section library. Limited dependence of correction values on IFBA presence and enrichment is found. When applied within nodal code the correction has potential to improve prediction of multiplication factors (boron concentration) and axial power distribution in Pressurized Water Reactor (PWR) reactors.Copyright © 2012 by ASME

A new method has been developed for the optimization of pressurized water reactor loading patterns. The optimization algorithm is built of the enhanced simulated annealing cooling schedule and adaptive trial loading pattern generator. Trial loading patterns are evaluated using one-and-a-half dimensonal two energy group diffusion code, which provides a good trade off between the speed and accuracy in scoping analysis. A simplified loading pattern optimization test problem with predefined global optimum of the given objective function is designed in order to analyze the efficiency of different optimization strategies. The results obtained using different cooling schedules and trial loading pattern generator modes illustrate their importance in the optimization algorithm and identify efficient strategies for simulated annealing based loading pattern optimization.

Glavne značajke reaktora IRIS (International Reactor Innovative and Secure) date su u ovom članku. IRIS je integralni, modularni, lakovodni reaktor srednje snage (335 MWe). Projekt reaktora IRIS razvija međunarodni konzorcij od dvadesetak organizacija iz deset zemalja predvođenih kompanijom Westinghouse. Primarni cilj projekta je razviti siguran, pouzdan i ekonomičan reaktor male do srednje snage. U članku je opisana integralna reaktorska posuda koja osim jezgre reaktora, moderatora, kontrolnih šipki i reflektora neutrona sadrži parogeneratore, reaktorske pumpe, mehanizme kontrolnih šipki i tlačnik. Jedinstvene sigurnosne karakteristike (,,safety-by-design) omogućile su manje i jednostavnije pasivne sigurnosne sustave. Poboljšana sigurnost i pouzdanost, ekonomičnost, te mogućnost postupne izgradnje novih kapaciteta uz relativno male investicije čine IRIS naročito privlačnim za manje elektroenergetske sustave.

The intention was to model a dry storage facility that could satisfy the needs of a medium nuclear power plant similar to the NPP Krsko. The attention has been focused on radiation dose rate analyses and criticality calculations. Using the SCALE 4.4 code package and modified QAD-CGGP code, we modeled a facility that satisfies the basic criteria for public radiation protection. The capacity of the storage is 1,400 spent fuel assemblies which is adequate for a forty years medium NPP lifetime.

In the electricity sector, market participants must make decisions about capacity choice in a situation of radical uncertainty about future market conditions. Sector is normally haracterised by non- storability and periodic and stochastic demand fluctuations. In these cases capacity determination is a decision for the long term, whereas production is adjusted in the short run. Capacities need to be installed well in advance (decision for investment even earlier because of long construction time and even longer in case of NPP to prepare all needed legal, financial and physical infrastructure), at times when companies face considerable demand and cost uncertainty when choosing their capacity. Paper looks on the main contributions in investment planning under uncertainty, in particular in the electricity market for capital intensive investments like NPP. The relationship between market and non-market factors (recent UK policy example) in determining investment signals in competitive electricity markets was analysed. Paper analyse the ability of competitive electricity markets to deliver the desired quantity and type of generation capacity and also investigates the variety of market imperfections operating in electricity generation and their impact on long- term dynamics for generation capacity, the most capital-intensive of the liberalised functions in the electricity supply industry. Paper analyses how price formation influences investment signals. Today, investment decisions are made by several operators that act independently. Number of factors (including market power, wholesale price volatility, lack of liquidity in the wholesale and financial market, policy and regulatory risks etc.) contribute to polluting the price signal and generating sub-optimal behaviour. Climate change policies can easily distort market signals, insulating renewables generation from market dynamics. This in turn reduces the proportion of the market that is effectively opened to competitive forces. When renewable support policies are undertaken, investments in conventional technologies suffer (especially in capital intensive investment like NPP). This produces distorting effects on the generation mix. Policy intervention, rather than market forces, is able to select artificially winners and losers, thus potentially undermining, in the long run, the necessary diversity of the energy mix.

If the TENORM is a byproduct originating from any energy gaining technological production process is to be reused as a product – raw material for other advanced product it has to meet the ultimate condition regulating the content of NORM or TENORM material in a final product. It means that the minimum requirements to be achieved are to mix the TENORM material with any other suitable material resulting in final Hex less than unity (Hex

Radiation dosimetry in protection against ionizing radiation involves research of all possible pathways through which natural or man-made radioactive materials can contaminate a habitat and actually harm its biota. It also takes into account natural and artificial (man-made) electromagnetic ionizing radiation (gama and X radiation). The purpose of conducted study was to validate a new active electronic dosimeter (AED) of type ALARA OD and to develop a new monitoring method by tracing the possible external occupational and/or public exposure over real time. This method is used to continuously measure and record external radiation doses. A novelty is establishing dose rates receiving pattern as a function of real time. This article presents a dosimetric study assessing exposure to man-made ionizing radiation of local environment and total occupational exposure of two professional workers involved in characterization, conditioning, and securing of unused radioactive sealed sources. Occupational whole body dosimetric results obtained with AED were compared with results obtained with passive dosimetry (film badge and thermoluminiscence). Air, dust, and silicon sand samples were analyzed by gamma-spectrometry to estimate internal exposure of the two workers to 222Rn due to inhalation or ingestion of dust and sand in indoor air. In order to establish total occupational exposure control radon measurement was performed in the immediate environment. The set of quantities describing possible hazard due to exposure of a public member to radionuclides with activity concentrations contributing the LBG and the external and internal Hazard index (Hex, Hint) were calculated and brought to discussion.

Urgency to stop further increase of greenhouse gases emissions and reverse the trends, as stated in the Fourth Intergovernmental Panel on Climate Change (IPPC) Report and in Copenhagen discussions, limits the realistic choice of energy technologies to those available now or in the near future of few decades. In the coming fifty years neither nuclear fusion nor carbon capture and storage (CCS) can be expected to give a significant contribution to world energy production. Two perspective intermittent sources such as wind and solar together with nuclear fission energy covering the base load consumption appears to be a combination with a potential to produce a large share of carbon free energy in the total world energy production. This contribution considers the issues, associated with required large scale deployment of nuclear fission energy. A serious question associated with nuclear energy is nuclear proliferation. Spread of uranium enrichment and spent fuel reprocessing installations in many new countries constructing nuclear reactors would be a major concern in present political environment. We investigate whether uranium resources would be sufficient to support nuclear build-up in next 50-60 years sufficiently large to significantly reduce carbon emission without reprocessing of spent nuclear fuel. A positive answer would mean that 50-60 years can be available to develop effective international control of nuclear fuel cycle installations. Our results show that a maximum nuclear build-up which would consume currently estimated uranium resources by 2065 without reprocessing could reduce by 2065 carbon emission by 39.6% of the total reduction needed to bring the WEO 2009 Reference Scenario prediction of total GHG emissions in 2065 to the level of the WEO 450 Scenario limiting global temperature increase to 2 ºC. The less demanding strategy of the nuclear replacement of all non- CCS coal power plants retiring during the 2025-2065 period would reduce emission by 26.1%.

Paper presents the results of on-line partial discharge (PD) measurements of large turbogenerator's armature winding. Measurements were performed for different load conditions using capacitive couplers and applying high frequency phase resolved method. The results show high PD levels where main source of PD moves from one phase to another and changes its intensity.

Uninterrupted auxiliary system power supply in large power plants is the key factor for safe standard operation, transient states, start-ups and shutdowns, particularly during fault conditions. Therefore, there are many challenges in designing the main electrical system as well as the auxiliary systems power supply. Depending upon the type of fuel used and the environmental control system required, a thermal power plant may consume as much as 10% of its total generation for auxiliary power, while a nuclear power plant may require only 4 – 6%. In general, the larger the power generating plant, the higher the voltage selected for the AC auxiliary electric system. Most stations in the 75 to 500 MW range utilize 4, 2 kV as the base auxiliary system voltage. Large generating stations 500 – 1000 MW and more use voltage levels of 6, 9 kV and more. Some single dedicated loads such as electric driven boiler feed pumps are supplied by a 13, 8 kV bus. While designing the auxiliary electric system, the following areas must be considered: motor starting requirements, voltage regulation requirements, short-circuit duty requirements, economic considerations, reliability and alternate sources. Auxiliary power supply cannot be completely generalized and each situation should be studied on its own merits to determine the optimal solution. Naturally, nuclear power plants have more reliability requirements and safety design criteria. Main coolant-pump power supply and continuity of service to other vital loads deserve special attention. This paper presents an overview of some up-to-date power plant auxiliary load system concepts. The main types of auxiliary loads are described and the electric diagrams of the modern auxiliary system supply concepts are given. Various alternative sources of auxiliary electrical supply are considered, the advantages and disadvantages of these are compared and proposals are made for high voltage distribution systems around the thermal and nuclear plant. Arrangements for DC supply systems and uninterruptable power supply (UPS) systems are also discussed.