INTRODUCTION

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

Characterization of Fast Neutron Transmission Through an Iron Shield

In this paper we give an analysis of the neutron transmission through an iron sphere using Monte Carlo and transport theory methods based on ENDF/B-VII.1 general purpose library. The motivation for this investigation comes from a well-known deficiency in the iron inelastic data from the older library evaluation (ENDF/B-V), giving a concern for a fast neutron flux underestimation within the reactor pressure vessels. In order to benchmark the next-generation ENDF/B-VI iron data, the U.S. Nuclear Regulatory Commission and the former Czechoslovakian National Research Institute have jointly preformed several experiments in 1990s, addressing neutron leakage spectra obtained for a 252Cf fission source in a centre of an iron sphere. It was shown that the ENDF/B-VI iron cross section, containing several improvements over previous evaluations, will not entirely resolve the neutron spectrum discrepancies observed at high neutron energies. Since safety analyses of reactor pressure vessel embrittlement are often based on neutron transport calculations using specific multigroup cross section libraries, simulation of this benchmark was performed using a hybrid shielding methodology of ADVANTG3.0.3 and MCNP6.1.1b codes. Comparison of calculated and referenced dosimeter activation rates are presented for several "standard" nuclear reactions, often used in reactor pressure vessel dosimetry. For that purpose, the new IRDFF-II special library from the IAEA Nuclear Data Services was used as a reference source of dosimetry cross sections. The MCNP6.1.1b code was used for calculation of reaction rates, which were also compared with previous IRDFF-1.05 special library and general purpose ENDF/B-VII.1 library

Future without Nuclear Energy; is it Feasible, is it Sensible?

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

Monte Carlo Codes for Neutron Buildup Factors

The point-kernel method is a widely used practical tool for gamma-ray shielding calculations. However, application of that method for neutron transport simulations is very limited. The accuracy of the method strongly depends on the accuracy of buildup factors used in the calculations. Buildup factors are usually obtained using appropriate computer codes, either based on discrete ordinates transport method or Monte Carlo approach. Since these codes put strong demands on computer resources, they are applied on a limited number of shielding configurations and an attempt is made to use these results and formulate an empirical expression for buildup factors estimation. Due to high physical complexity of neutron transport through shielding material it is very hard to perform parameterisation in order to establish adequate empirical formula. Existing formulas are very limited and are usually applicable to a narrow neutron energy range for few commonly used shielding materials, mostly in monolayer configuration. Recently, a new approach has been proposed for determination of gamma ray buildup factors for mono-layer, as well as multi-layer shielding configurations covering a wide gamma ray energy range. The new regression model is based on support vector machines learning technique, which has theoretical background in statistical learning theory. Development of named regression model required a large number of experimental data obtained by Monte Carlo computer code. More than 7000 Monte Carlo runs were required. Due to physical complexity neutron transport is likely to require even more experimental data in order to generate a model of reasonable accuracy. Therefore, the choice of appropriate Monte Carlo code is a very important question. One has to take into account the accuracy as well as the time required for input preparation and running the code. What also has to be considered is the possibility of the code to be incorporated in an algorithm for automated generation of experimental data. In this paper three Monte Carlo codes are analysed, namely SCALE4.4 code package (SAS3 sequence), SCALE6.0 code package (MAVRIC sequence), and MCNP5. Two simple experimental setups based on a point isotropic source in spherical and slab-like shield are modelled, and the codes are examined on previously mentioned issues. The comparison results show that each one of the examined codes has potential to be used for neutron buildup factor model generation. However, some aspects of their utilization require further analysis prior to final selection

Full Core Criticality Modeling of Gas-Cooled Fast Reactor using the SCALE6.0 and MCNP5 Code Packages

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

PCA Benchmark Analysis with ADVANTG3.0.1. and MCNP6.1.1b Codes

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)