http://www.hindawi.comThe latest articles from Hindawi Publishing Corporation© 2012, Hindawi Publishing Corporation. All rights reserved.http://www.hindawi.com/journals/stni/2012/507921/The paper deals with the evaluation of the Fukushima-Daiichi Nuclear Power Plant (NPP) accident in Units 1 to 4: an attempt is made to discuss the scenario within a technological framework, considering precursory documented regulations and predictable system performance. An outline is given at first of the NPP layout and of the sequence of major events. Then, plausible time evolutions of relevant quantities in the different Units, is inferred based on results from the application of numerical codes. Scenarios happening in the primary circuit and containment (three Units involved) are distinguished from scenarios in spent fuel pool (four Units involved). Radiological releases to the environment and doses are approximately estimated. The event is originated by a natural catastrophe with almost simultaneous occurrence of earthquake and tsunami. These caused heavy destruction in a region in Japan much wider than the land around the NPP which was affected by the nuclear contamination. Key outcome from the work is the demonstration of strength for nuclear technology; looking at the past, misleading Probabilistic Safety Assessment (PSA) data and inadequacy in licensing processes have been found. Looking into the future keywords are Emergency Rescue Team (ERT), Enhanced Human Performance (EHP), and Robotics in Nuclear Safety and Security (RNSS).Francesco D'Auria, Giorgio Galassi, Patricia Pla, and Martina AdorniCopyright © 2012 Francesco D'Auria et al. All rights reserved.http://www.hindawi.com/journals/stni/2012/106759/In water-cooled nuclear power reactors, significant quantities of steam and hydrogen could be produced within the primary containment following the postulated design basis accidents (DBA) or beyond design basis accidents (BDBA). For accurate calculation of the temperature/pressure rise and hydrogen transport calculation in nuclear reactor containment due to such scenarios, wall condensation heat transfer coefficient (HTC) is used. In the present work, the adaptation of a commercial CFD code with the implementation of models for steam condensation on wall surfaces in presence of noncondensable gases is explained. Steam condensation has been modeled using the empirical average HTC, which was originally developed to be used for “lumped-parameter” (volume-averaged) modeling of steam condensation in the presence of noncondensable gases. The present paper suggests a generalized HTC based on curve fitting of most of the reported semiempirical condensation models, which are valid for specific wall conditions. The present methodology has been validated against limited reported experimental data from the COPAIN experimental facility. This is the first step towards the CFD-based generalized analysis procedure for condensation modeling applicable for containment wall surfaces that is being evolved further for specific wall surfaces within the multicompartment containment atmosphere.Pavan K. Sharma, B. Gera, R. K. Singh, and K. K. VazeCopyright © 2012 Pavan K. Sharma et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/897165/This paper addresses the problem of fission material reproduction increase in thermal reactors. Reproduction increase is achieved while decreasing the fission material content in the fuel. In that case, a decrease of neutron loss in construction materials and a neutron leakage decrease are required for obtaining reactor criticality. Effectiveness of the reactor functioning can be increased by the use of additional neutron sources, for example, n-2n reaction in beryllium framing. The possibility of additional neutron source development with the use of fast neutrons with an energy distribution close to the fission spectrum in the major part of thermal reactor core is researched in this paper.Vladimir M. Kotov, Anna S. Sergeeva, Ruslan A. Irkimbekov, and Vladislav I. SuprunovCopyright © 2011 Vladimir M. Kotov et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/252903/As a rule materials of small atomic weight (light and heavy water, graphite, and so on) are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium) and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative) reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium) ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.A. N. Shmelev, G. G. Kulikov, V. A. Apse, E. G. Kulikov, and V. V. ArtisyukCopyright © 2011 A. N. Shmelev et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/589747/Air ingress with graphite burning belongs to the accident scenarios in HTRs with potentially severe consequences. This paper gives an overview of basic phenomena of graphite burning like ignition conditions and moving reaction fronts. The pioneering graphite burning experiments of Don Schweitzer are successfully reevaluated. Ignition conditions are examined, and it is underlined that burning depends not only on graphite properties but also on the heat balance in the whole graphite arrangement. In graphite-moderated reactors, ignition occurs at about 650°C for small air flow rates: this means that normal operation temperatures in HTRs always allow for ignition. Fuel behaviour in air ingress, as determined in the KORA facility, is discussed: up to about 1300°C modern TRISO fuel is stable in air, but from 1500°C a complete, fast destruction is observed. Exemplary calculations on massive air ingress by chimney draught performed with REACT/THERMIX are outlined. For a hot bottom reflector there is a substantial time span before fuel is attacked. Because severe air ingress in well-designed HTRs belongs to beyond design basis accidents, the knowledge is fairly good. Concerning protecting measures, a more detailed examination of thick SiC layers is proposed.Rainer MoormannCopyright © 2011 Rainer Moormann. All rights reserved.http://www.hindawi.com/journals/stni/2011/948457/The author extends the previously proposed combined asymptotic method (CAM) of seismic SSI analysis for the multi-support systems and applies it to the transportation portal as a double-support system (together with the reactor building). The key issue is the development of the structural dynamic stiffness matrix condensed to the supports by the modal approach. Then the condensed structural matrix is combined with the soil dynamic stiffness matrix also condensed to the rigid basements. As a result, a very simple linear system 12∗12 is solved in the frequency domain. This gives the transfer functions from the free-field motion to the motion of the basements. The only important limitations are the linearity of the soil’s and structure’s properties and the rigidity of the basements. The results for the sample system are checked against the full SASSI solution. The results can be used to justify the further simplification of the system.Alexander TyapinCopyright © 2011 Alexander Tyapin. All rights reserved.http://www.hindawi.com/journals/stni/2011/175492/In a series of previous articles, one of the authors published designs of the AB Dome which can cover a city, important large installations or subregions by a transparent thin film supported by a small additional air overpressure. The AB Dome keeps the outside atmospheric conditions from the interior protecting a city from chemical, bacterial, and radioactive weapons (wastes). The design in this article differs from previous one as this design employs an inflatable columns which does not need an additional pressure (overpressure) inside the dome and is cheaper in construction (no powered air pumping station) and in operation (no special entrance airlock and permanent pumping expense). When dome is supported by columns, no overpressure is required inside the dome which is important when the dome covers a damaged nuclear reactor. The nuclear reactor may produce radioactive gases and dust, and, as inflatable domes are not typically hermetically sealed, the increased pressure inside the dome can leak out gas and dust into the atmosphere. The suggested design does not have this drawback. Positive pressure gradients expel dust particles—neutral pressure gradients will not. (Negative pressure gradients may even be possible in certain configurations.)Alexander Bolonkin, Shmuel Neumann, and Joseph FriedlanderCopyright © 2011 Alexander Bolonkin et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/262585/This paper makes use of the concept and theory of fuzzy number in fuzzy mathematics, to research for the response time of operator in accident of Chinese nuclear power plant. Through the quantitative analysis for the performance shape factors (PSFs) which influence the response time of operators, the formula of the operator response time is obtained based on the possibilistic fuzzy linear regression model which is used for the first time in this kind of research. The research result shows that the correct research method can be achieved through the analysis of the information from a small sample. This method breaks through the traditional research method and can be used not only for the reference to the safe operation of nuclear power plant, but also in other areas.Fang Xiang, Zhou Yangping, and Li FuCopyright © 2011 Fang Xiang et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/284274/The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating) of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.Vadim E. Seleznev, Vladimir V. Aleshin, and Sergey N. PryalovCopyright © 2011 Vadim E. Seleznev et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/519538/Nuclear islands, which are integrated power production sites, could effectively sequester and safeguard the US stockpile of plutonium. A nuclear island, an evolution of the integral fast reactor, utilizes all the Transuranics (Pu plus minor actinides) produced in power production, and it eliminates all spent fuel shipments to and from the site. This latter attribute requires that fuel reprocessing occur on each site and that fast reactors be built on-site to utilize the TRU. All commercial spent fuel shipments could be eliminated by converting all LWR nuclear power sites to nuclear islands. Existing LWR sites have the added advantage of already possessing a license to produce nuclear power. Each could contribute to an increase in the nuclear power production by adding one or more fast reactors. Both the TRU and the depleted uranium obtained in reprocessing would be used on-site for fast fuel manufacture. Only fission products would be shipped to a repository for storage. The nuclear island concept could be used to alleviate the strain of LWR plant sites currently approaching or exceeding their spent fuel pool storage capacity. Fast reactor breeding ratio could be designed to convert existing sites to all fast reactors, or keep the majority thermal.Charles W. SolbrigCopyright © 2011 Charles W. Solbrig. All rights reserved.http://www.hindawi.com/journals/stni/2011/560157/A specialized module VAPEX-M has been developed and implemented as a part of an integral code, SOCRAT, to enable the modeling of fuel-coolant interactions (FCIs) during severe accidents. The mathematical model and correlations for the main physical processes are described. Results of computational analysis of three experimental series reported in the literature are presented. The calculations were carried out by the combined SOCRAT/VAPEX code and were aimed at validation of the predictive capabilities of the code. The experiments chosen cover a wide range of physical parameters, which enables different aspects of the code to be verified, that is, drag correlations (MAGICO-2000), evaporation rate (QUEOS), fuel fragmentation, and interaction with the coolant in all complexity (FARO). Generally, reasonable agreement between the measured data and calculated results was obtained, which allows one to use the combined SOCRAT/VAPEX code for severe accidents analysis.Vladimir Melikhov, Oleg Melikhov, Sergey Yakush, and Nikita RtishchevCopyright © 2011 Vladimir Melikhov et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/862812/In water-cooled nuclear power reactors, significant quantities of hydrogen could be produced following a postulated loss-of-coolant accident (LOCA) along with nonavailability of emergency core cooling system (ECCS). Passive autocatalytic recombiners (PAR) are implemented in the containment of water-cooled power reactors to mitigate the risk of hydrogen combustion. In the presence of hydrogen with available oxygen, a catalytic reaction occurs spontaneously at the catalyst surfaces below conventional ignition concentration limits and temperature and even in presence of steam. Heat of reaction produces natural convection flow through the enclosure and promotes mixing in the containment. For the assessment of the PAR performance in terms of maximum temperature of catalyst surface and outlet hydrogen concentration an in-house 3D CFD model has been developed. The code has been used to study the mechanism of catalytic recombination and has been tested for two literature-quoted experiments.B. Gera, P. K. Sharma, R. K. Singh, and K. K. VazeCopyright © 2011 B. Gera et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/574514/A large amount of Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR), shortly after the hypothetical beginning of a severe accident leading to the melting of the core. According to local gas concentrations, the gaseous mixture of hydrogen, air and steam can reach the flammability limit, threatening the containment integrity. In order to prevent mechanical loads resulting from a possible conflagration of the gas mixture, French and German reactor containments are equipped with passive autocatalytic recombiners (PARs) which preventively oxidize hydrogen for concentrations lower than that of the flammability limit. The objective of the paper is to present numerical assessments of the recombiner models implemented in CFD solvers NEPTUNE_CFD and Code_Saturne. Under the EDF/EPRI agreement, CEA has been committed to perform 42 tests of PARs. The experimental program named KALI-H2, consists checking the performance and behaviour of PAR. Unrealistic values for the gas temperature are calculated if the conjugate heat transfer and the wall steam condensation are not taken into account. The combined effects of these models give a good agreement between computational results and experimental data.Stéphane Mimouni, Namane Mechitoua, and Mehdi OuraouCopyright © 2011 Stéphane Mimouni et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/181393/The horizontal steam generator (SG) is one of specific features of Russian-type pressurized water reactors (VVERs). The main advantages of horizontal steam generator are connected with low steam loads on evaporation surface, simple separation scheme and high circulation ratio. The complex three-dimensional steam-water flows in the steam generator vessel influence significantly the processes of the steam separation, distribution, and deposition of the soluble and nonsoluble impurities and determine the efficiency and reliability of the steam generator operation. The 3D code for simulation of the three-dimensional steam-water flows in the steam generator could be effective tool for design and optimization of the horizontal steam generator. The results of the code calculations are determined mainly by the set of the correlations describing interaction of the steam-water mixture with the inner constructions of the SG and interfacial friction. The results obtained by 3D code STEG with the usage of the different interfacial friction correlations are presented and discussed in the paper. These results are compared with the experimental ones obtained at the experimental test facility PGV-1500 constructed for investigation of the processes in the horizontal steam generator.Vladimir Melikhov, Oleg Melikhov, Yury Parfenov, and Alexey NerovnovCopyright © 2011 Vladimir Melikhov et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/941239/The present work is focused on the condensation heat transfer that plays a dominant role in many accident scenarios postulated to occur in the containment of nuclear reactors. The study compares a general multiphase approach implemented in NEPTUNE_CFD with a homogeneous model, of widespread use for engineering studies, implemented in Code_Saturne. The model implemented in NEPTUNE_CFD assumes that liquid droplets form along the wall within nucleation sites. Vapor condensation on droplets makes them grow. Once the droplet diameter reaches a critical value, gravitational forces compensate surface tension force and then droplets slide over the wall and form a liquid film. This approach allows taking into account simultaneously the mechanical drift between the droplet and the gas, the heat and mass transfer on droplets in the core of the flow and the condensation/evaporation phenomena on the walls. As concern the homogeneous approach, the motion of the liquid film due to the gravitational forces is neglected, as well as the volume occupied by the liquid. Both condensation models and compressible procedures are validated and compared to experimental data provided by the TOSQAN ISP47 experiment (IRSN Saclay). Computational results compare favorably with experimental data, particularly for the Helium and steam volume fractions.S. Mimouni, N. Mechitoua, A. Foissac, M. Hassanaly, and M. OuraouCopyright © 2011 S. Mimouni et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/584256/The studies on Accelerator-Driven Systems (ADSs) have renewed the interest in the theoretical and computational evaluation of the main integral parameters characterizing subcritical systems (e.g., reactivity, effective delayed neutron fraction βeff, and mean prompt neutron generation time). In particular, some kinetic parameters, as the effective delayed neutron fraction, are evaluated in Monte Carlo codes by formulations which do not require the calculation of the adjoint flux. This paper is focused on a theoretical and computational analysis about how the different βeff definitions are connected and which are the approximations inherent to the Monte Carlo definition with respect to the standard definition involving weighted integrals. By means of a refined transport computational analysis carried out in a coherent and consistent way, that is, using the same deterministic code and neutron data library for the βeff evaluation in different ways, the theoretical analysis is numerically confirmed. Both theoretical and numerical results confirm the effectiveness of the Monte Carlo βeff evaluation, at least in cases where spectral differences between total and prompt fluxes are negligible with respect to the value of the functionals entering the classical βeff formulation.M. Carta, S. Dulla, V. Peluso, P. Ravetto, and G. BianchiniCopyright © 2011 M. Carta et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/347320/Wind turbines are becoming widely used as they are an environmentally friendly way for energy production without emissions; however, they are exposed to a corrosive environment. In addition, as wind turbines typically are the tallest structures in the surrounding area of a wind farm, it is expected that they will attract direct lightning strikes several times during their operating life. The purpose of this paper is to show that the radiography with a transportable unit is a solution to find defects in the wind turbine blade and reduce the cost of inspection. A transportable neutron radiography system, incorporating an Sb–Be source, has been simulated using the MCNPX code. The simulated system has a wide range of radiography parameters.J. G. Fantidis, C. Potolias, and D. V. BandekasCopyright © 2011 J. G. Fantidis et al. All rights reserved.http://www.hindawi.com/journals/stni/2011/741396/An optimized single-channel phoswich well detector design has been proposed and assessed in order to improve beta-gamma coincidence measurement sensitivity of xenon radioisotopes. This newly designed phoswich well detector consists of a plastic beta counting cell (BC404) embedded in a CsI(Tl) crystal coupled to a photomultiplier tube. The BC404 is configured in a cylindrical pipe shape to minimise light collection deterioration. The CsI(Tl) crystal consists of a rectangular part and a semicylindrical scintillation part as a light reflector to increase light gathering. Compared with a PhosWatch detector, the final optimized detector geometry showed 15% improvement in the energy resolution of a 131mXe 129.4 keV conversion electron peak. The predicted beta-gamma coincidence efficiencies of xenon radioisotopes have also been improved accordingly.Weihua Zhang, Pawel Mekarski, Marc Bean, Jing Yi, and Kurt UngarCopyright © 2011 Weihua Zhang et al. All rights reserved.http://www.hindawi.com/journals/stni/2010/517471/We present experimental results on 9 reduced-scale fuel assemblies arranged in a network of 3 by 3, subjected to an axial flow. The objective is to analyse the fluid force induced by the motion of the central fuel assembly on the others fuel assemblies. The displacement of the central fuel assembly is imposed, while the others are fixed. Fluid forces acting on fuel assemblies are measured with force sensors. We observed that the coupling between fuel assemblies increases with the fluid velocity, and that the coupling in the transverse direction is not negligible compared to the coupling in the direction of excitation. We also observe that the fluid flow induces a stiffening of the central fuel assembly.Guillaume Ricciardi, Sergio Bellizzi, Bruno Collard, and Bruno CochelinCopyright © 2010 Guillaume Ricciardi et al. All rights reserved.http://www.hindawi.com/journals/stni/2010/574195/The behaviour of a parallel-channel natural circulation boiling water reactor under a low-pressure low-power startup condition has been studied numerically (using RELAP5) and compared with its scaled model. The parallel-channel RELAP5 model is an extension of a single-channel model developed and validated with experimental results. Existence of in-phase and out-of-phase flashing instabilities in the parallel-channel systems is investigated through simulations under equal and unequal power boundary conditions in the channels. The effect of flow resistance on Type-I oscillations is explored. For nonidentical condition in the channels, the flow fluctuations in the parallel-channel systems are found to be out-of-phase.S. P. Lakshmanan and Manmohan PandeyCopyright © 2010 S. P. Lakshmanan and Manmohan Pandey. All rights reserved.http://www.hindawi.com/journals/stni/2009/459139/Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in pool” delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand “SVYP-440” for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system “Sipping in pool,” the limit values of leak tightness were established.M. Mikloš and V. KršjakCopyright © 2009 M. Mikloš and V. Kršjak. All rights reserved.http://www.hindawi.com/journals/stni/2009/973605/A neutronic module for the solution of two-dimensional steady-state multigroup diffusion problems in nuclear reactor cores is developed. The module can produce both direct fluxes as well as adjoints, that is, neutron importances. Different numerical schemes are employed. A standard finite-difference approach is firstly implemented, mainly to serve as a reference for less computationally challenging schemes, such as nodal methods and boundary element methods, which are considered in the second part of the work. The validation of the methods proposed is carried out by comparisons of results for reference structures. In particular a critical problem for a homogeneous reactor for which an analytical solution exists is considered as a benchmark. The computational module is then applied to a fast spectrum system, having physical characteristics similar to the proposed lead-cooled ELSY project. The results show the effectiveness of the numerical techniques presented. The flexibility and the possibility to obtain neutron importances allow the use of the module for parametric studies, design assessments, and integral parameter evaluations as well as for future sensitivity and perturbation analyses and as a shape solver for time-dependent procedures.Song Han, Sandra Dulla, and Piero RavettoCopyright © 2009 Song Han et al. All rights reserved.http://www.hindawi.com/journals/stni/2009/835162/Plant measured data from VVER-1000 coolant mixing experiments were used within the OECD/NEA and AER coupled code benchmarks for light water reactors to test and validate computational fluid dynamic (CFD) codes. The task is to compare the various calculations with measured data, using specified boundary conditions and core power distributions. The experiments, which are provided for CFD validation, include single loop cooling down or heating-up by disturbing the heat transfer in the steam generator through the steam valves at low reactor power and with all main coolant pumps in operation. CFD calculations have been performed using a numerical grid model of 4.7 million tetrahedral elements. The Best Practice Guidelines in using CFD in nuclear reactor safety applications has been used. Different advanced turbulence models were utilized in the numerical simulation. The results show a clear sector formation of the affected loop at the downcomer, lower plenum and core inlet, which corresponds to the measured values. The maximum local values of the relative temperature rise in the calculation are in the same range of the experiment. Due to this result, it is now possible to improve the mixing models which are usually used in system codes.Thomas HöhneCopyright © 2009 Thomas Höhne. All rights reserved.http://www.hindawi.com/journals/stni/2008/597491/Fission products deposited in the coolant circuit outside of the active core play a dominant role in source term estimations for advanced small pebble bed HTRs, particularly in design basis accidents (DBA). The deposited fission products may be released in depressurization accidents because present pebble bed HTR concepts abstain from a gas tight containment. Contamination of the circuit also hinders maintenance work. Experiments, performed from 1972 to 88 on the AVR, an experimental pebble bed HTR, allow for a deeper insight into fission product transport behavior. The activity deposition per coolant pass was lower than expected and was influenced by fission product chemistry and by presence of carbonaceous dust. The latter lead also to inconsistencies between Cs plate out experiments in laboratory and in AVR. The deposition behavior of Ag was in line with present models. Dust as activity carrier is of safety relevance because of its mobility and of its sorption capability for fission products. All metal surfaces in pebble bed reactors were covered by a carbonaceous dust layer. Dust in AVR was produced by abrasion in amounts of about 5 kg/y. Additional dust sources in AVR were ours oil ingress and peeling of fuel element surfaces due to an air ingress. Dust has a size of about 1  μm, consists mainly of graphite, is partly remobilized by flow perturbations, and deposits with time constants of 1 to 2 h ours. In future reactors, an efficient filtering via a gas tight containment is required because accidents with fast depressurizations induce dust mobilization. Enhanced core temperatures in normal operation as in AVR and broken fuel pebbles have to be considered, as inflammable dust concentrations in the gas phase.Rainer MoormannCopyright © 2008 Rainer Moormann. All rights reserved.http://www.hindawi.com/journals/stni/2008/796268/Decay heat in fission reactors is almost equally subdivided into two parts, one part due to beta rays and the other due to gamma photons. Beta rays are absorbed practically where they are generated while gamma photons travel some distance in core before being absorbed. The decay power peaking factor is, in fact, affected by this phenomenon of gamma decay heat redistribution. Calculations have been performed by the Monte Carlo MCNP5 computer code on the experimental LOFT reactor and on a larger 1000 MWe PWR using various initial power distributions with variable power peak sharpness (midheight peak width). The results indicate that an average peak energy reduction ratio of 0.82 for gamma (18% peak reduction) can be used with tolerable error up to a midheight width of the produced energy peak (neutron flux shape during operation) of 120 cm. Beyond this value, no peak energy reduction is warranted. This phenomenon of absorbed γ power redistribution in core may be very significant (100 to 150∘K reduction in calculated PCT).Gianni Petrangeli and Calogero SollimaCopyright © 2008 Gianni Petrangeli and Calogero Sollima. All rights reserved.http://www.hindawi.com/journals/stni/2008/159083/This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR) technology with the supercritical (SC) steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR) reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency). Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.Shutang Zhu, Ying Tang, Kun Xiao, and Zuoyi ZhangCopyright © 2008 Shutang Zhu et al. All rights reserved.http://www.hindawi.com/journals/stni/2008/781824/The present paper deals with the description of the technical activities conducted within the TACIS Project R2.03/97, 2 EC Contract no. 30303, related to RBMK. The project activities are focused toward the setting-up of a chain of computational tools suitable for the analysis of transients expected in the RBMK nuclear power plant (NPP). The accident leading to the rupture of one pressure channel, with fuel melting or high temperature damage, creep and brittle failure of the pressure tube and of graphite bricks with possibility of rupture propagation, constitutes the reference scenario for the project. However, a series of expected scenarios has been selected to prove the capability of the individual codes or chains of code in simulating the envisaged phenomenology. The paper summarizes the activities performed at NIKIET in Moscow and at University of Pisa (UNIPI) in Pisa. A top-down approach is pursued in structuring the executive summary that includes the following sections: (i) the safety needed for the RBMK NPP, (ii) the roadmap, (iii) the adopted computational tools, (iv) key findings, (v) Emphasis is given to the multiple pressure tube rupture (MPTR) issue and the individual channel monitoring (ICM) proposal.F. D'Auria, S. Soloviev, D. Mazzini, and C. SollimaCopyright © 2008 F. D'Auria et al. All rights reserved.http://www.hindawi.com/journals/stni/2008/756181/The main challenge in nuclear fuel cycle closure is the reduction of the potential radiotoxicity, or of the time in which that possible hazard really exists. Probably, the transmutation of minor actinides with fast fission processes is the most effective answer. This work, performed in SCK⋅CEN (Belgium) and DIMNP Pisa University, is focused on preliminary evaluation of industrial scale ADS (400 MWth, 2.5 mA) burning capability. An inert matrix fuel of minor actinides, 50% vol. MgO and 50% vol. (Pu,Np,Am,Cm)O1.88, core content, with 150 GWd/ton discharge burn up, is used. The calculations were performed using ALEPH-1.1.2, MCNPX-2.5.0, and ORIGEN2.2. codes.Barbara Calgaro, Barbara Vezzoni, Nicola Cerullo, Giuseppe Forasassi, and Bernard VerboomenCopyright © 2008 Barbara Calgaro et al. All rights reserved.http://www.hindawi.com/journals/stni/2008/318956/The present paper deals with the investigation study on accident management in VVER-1000 reactor type conducted in the framework of a European Commission funded project. The mentioned study involved both experimental and computational fields. The purpose of this paper is to summarize the main findings from the execution of a wide-range analysis focused on AM in VVER-1000 with main regard to the qualification of computational tools and the proposal for an optimal AM strategy for this kind of NPP.F. D'Auria, A. Suslov, N. Muellner, G. Petrangeli, and M. CherubiniCopyright © 2008 F. D'Auria et al. All rights reserved.http://www.hindawi.com/journals/stni/2008/430768/The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.A. Kaliatka, E. Uspuras, and M. VaisnorasCopyright © 2008 A. Kaliatka et al. All rights reserved.