Fast neutron irradiation causes embrittlement of the reactor pressure vessel (RPV) material; therefore, it may end operation life before design lifetime. Well-known method to recuperate crystal lattice dislocations is annealing. In the current version of thorium fueled supercritical water-cooled reactor (SCWR) design proposed by the Institute of Nuclear Technology at Budapest University of Technology and Economics (BME NTI), the supercritical fluid flows upward between the core barrel and the inner surface of the RPV thereby, the coolant would keep the RPV's temperature at similar to 500 degrees C. This reverse coolant flow direction would decrease the embrittlement of RPV by constant annealing. To minimize the fast neutron flux increase, a relatively thin shielding connected to the inner surface of the barrel could be used. This presents fast neutron irradiation analysis, performed for different settings of the shielding to reduce fast neutron flux reaching the inner surface of RPV.

The closure of the nuclear fuel cycle is currently envisaged with the deployment of Generation IV fast reactors which are able to generate their fuel from fertile U-238 or Th-232, and also burn minor actinides arising from legacy wastes and thermal reactors in the nuclear park. The optimization of such fuel cycle strategies requires detailed models, capable of simulating the transition from initial state to equilibrium. Due to the high computational cost of detailed burn-up calculations, most scenario codes use parametrized few group cross-sections to calculate fuel depletion in the reactors. This paper presents a fast and flexible burn-up scheme called FITXS which we applied on the GFR2400 Gas-cooled Fast Reactor and a European Pressurized Reactor MOX fuel assembly. Based on the fitting of one-group cross sections as functions of the detailed fuel composition, the developed models are able to calculate spent fuel compositions with high accuracy for a wide range of initial compositions in less than one second computational time. In order to demonstrate applicability, the models were integrated into a fuel cycle model containing GFR2400 and European Pressurized Reactors, as well as conventional Light Water Reactors, and the fuel utilization and transmutational properties of the system were analyzed. It was found that the GFR2400 can operate in a closed fuel cycle, with positive breeding gain in the equilibrium. Increased minor actinide feed improves breeding in the GFR2400, but the accumulation of plutonium can be prevented by recycling the excess plutonium as MOX fuel in thermal reactors. (C) 2017 Elsevier Ltd. All rights reserved.

The service life of the Hungarian Paks Nuclear Power Plant (NPP) will be extended from the originally planned 30 years to 50 years. To improve the reliability of the results obtained in frame of the old reactor pressure vessel (RPV) surveillance programme, new methods have been developed, and based on them, the old exposition data have been re-evaluated for all the four reactor units. At the same time, a new RPV surveillance programme has been developed and introduced, and long term irradiations have been performed to determine the radiation damage of the surveillance specimens due to the high fast neutron exposition. Neutron transport calculations have been performed with a validated neutron transport code system to determine the fast neutron exposition of the RPVs during the extended service life. The cavity dosimetry is in the introductory phase. This paper presents the new developments in the field of the RPV surveillance dosimetry and summarises the results obtained. According to the results the service life of the NPP can safely be extended for the planned 50 years.

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.

Activation calculations for two nuclear power plants of WWER-440 type have been performed by the authors in order to assist the decommissioning planning by assessing the radioactive inventory present at the time of and at different times after the final shutdown. According to related international literature and studies performed earlier by the authors, considering the activity more than 99% of this inventory is concentrated in the materials directly surrounding the reactor core, where the predominant evolution of radionuclides is generated by neutron induced nuclear reactions. In order to obtain the highest possible accuracy in modelling, three-dimensional Monte Carlo neutron transport calculations were performed. Besides the methods and models applied to these analyses, the paper also summarizes the results that can be generally applied to such nuclear power plant types. At the time of shutdown, the total activity of the stainless steel components is about 6 x 10(16) Bq and 1.3 x 10(17) Bq for the two NPPs considered. The biological shielding concrete constitutes approximately 7 x 10(13) Bq and -1.1 x 10(14) Bq. (C) 2014 Elsevier B.V. All rights reserved.

The Generation IV initiative was launched with the goal of developing nuclear reactors which surpass current designs in safety, sustainability, economics and non-proliferation. From the six most promising concepts the Gas Cooled Fast Reactor (GFR) represents a challenging and innovative idea that is prominent in the sustainability aspect with the ability to have a closed fuel cycle and the potential to burn minor actinides (MAs). The European FP7 GoFastR project was one of the latest steps in the development and further optimization of GFRs.This paper presents a comprehensive overview of the neutronic performance of GFR2400 which was considered as a conceptual design for a large scale GFR within the collaboration. This reactor is the newest on the evolutionary path of fully ceramic GFRs featuring ceramic fuel and structural materials allowing high temperatures and efficiency using helium coolant. An important innovation of the current design is the application of refractory metallic liners to enhance the fission product retention of the cladding, resulting in a significant neutronic penalty during normal operation, at the same time being advantageous under transient conditions involving spectrum softening.Using the ERANOS and SCALE code systems several parameters were determined for beginning of life (BOL) conditions, including excess reactivity, various reactivity effects such as depressurization, Doppler or thermal expansion effects, as well as kinetic parameters. An extensive sensitivity and uncertainty analysis of these parameters was also done with the 15 group BOLNA and 44 group SCALE covariance libraries. Open and closed fuel cycle operations were investigated and the transmutational capabilities were studied with the GFR connected to traditional light water reactors in a symbiotic system.The presented analysis shows that the GFR2400 design is a major improvement compared to previous concepts. All preliminary constraints are respected resulting in a manageable initial Pu inventory of 10 t/GWel at 45% plant efficiency, a low MA mass fraction of 1% by self-recycling and a near zero breeding gain without the use of fertile blankets. At the same time the reactor has acceptable safety features precluding super-prompt-criticality in depressurized conditions at BOL and in open cycle equilibrium. Either of the two planned control devices is sufficient to shut down the reactor independently of the other and the refractory liners introduce significant negative reactivity in case of water ingress. However the occurrence of hot spots when all control rods are inserted needs further analysis.The design also shows promising closed fuel cycle and transmutational performance. However as is the case in other fast reactors the fuel cycle closure causes safety related parameters to degrade, most importantly the depressurization reactivity effect to exceed the effective delayed neutron fraction in the current design. To assess the acceptability of this deterioration further analysis is needed.Finally, it can be concluded that current commercial codes are satisfactory for such analysis; however there is a need for better covariance data. Several parameters exceed their target uncertainty value, most notably the k-effective by a factor of 6, the main source of the uncertainty being the inelastic scattering of U-238. (C) 2014 Elsevier Ltd. All rights reserved.

This paper presents high accuracy tritium production rate measurement results compared with calculations using the MCNPX Monte Carlo particle transport code. The experimental results are regarded as reference values for a new passive technique based on the secondary charged particle activation method developed for measuring the tritium production rate in the test blanket modules of the ITER Tokamak. The O-16(t,n)F-18 reaction, which is one of the possible tritium monitor reactions, was also extensively investigated, and the experimentally determined reaction rates were compared with simulations.Li2CO3 solution was filled and sealed into quartz ampoules which were irradiated in the Training Reactor of the Budapest University of Technology and Economics. The amount of F-18 was determined using gamma-spectroscopy. Then the precise tritium measurements were carried out in the Hertelendi Laboratory of Environmental Studies using the H-3-He-3 ingrowth method, where the He-3 produced during the storage time is measured by a static noble gas mass spectrometer (VG-5400).The HT/HTO ratio in the irradiated aqueous solutions was found to be 0.1323 +/- 0.0034. Based on the comparison of the measurements and the simulations it was pointed out that the model calculations underestimate the reaction rate of both the Li-6(n,t)alpha and the O-16(t,n)F-18 reactions by 5-10% and 15%, respectively. (c) 2013 Elsevier B.V. All rights reserved.

Within the Generation IV initiative, the gas-cooled fast reactor (GFR) is one of the reactors dedicated to minor actinide (MA) transmutation. This paper summarizes the research performed with the GFR600 reference design in order to assess its MA burning capabilities. For the study, modules of the SCALE program system were used.Single-cycle parametric studies were performed with cores having different MA content and spatial distribution. It was shown that the addition of MAs to the fuel greatly reduced the reactivity loss during burnup. Moreover, the higher the MA content of the core, the higher the fraction of it that was fissioned; however, the more the delayed neutron fraction and the fuel temperature coefficient degraded. Significant reduction can be achieved in the amounts of neptunium and americium, while curium isotopes accumulate.The study of multiple consecutive cycles showed that by adding only depleted uranium (DU) to the reprocessed actinides in fuel fabrication (pure DU feed strategy), up to 70% of the initially loaded MAs can be fissioned in the first five cycles. Moreover, the reactor can he made critical during that time if the initial MA content is higher than 3%. By feeding MAs as well (constant MA content strategy), the reactivity has a steady increase from cycle to cycle, predominantly due to Pu-238 breeding from Np-237.The effects of the isotopic composition of the plutonium and MAs were also examined by performing calculations with data specific to the spent fuel of traditional western pressure water reactors and Russian type VVER440 reactors. Despite the considerably different MA vectors, no significant deviation was found in their overall transmutation. However, the Pu composition had a strong effect on the reactivity and the delayed neutron fraction in the first cycles.Finally, cores having nonuniform MA content were investigated. It was found that though the MA destruction efficiency was significantly higher in the middle of the core than at the edge, moving some of the MAs from the outer regions to the center resulted in only minor improvement in their destruction. However, the spectral changes caused by the rearrangement increased the k-effective, which allowed higher burn ups and increased MA destruction. Unfortunately, some of the safety parameters of the reactor degraded.

In this work, a new passive technique has been developed for measuring the tritium production rate in ITER (International Thermonuclear Experimental Reactor) test blanket modules. This method is based on the secondary charged particle activation, in which the irradiated sample contains two main components: a tritium producing target (6Li or 7Li) and an indicator nuclide, which has a relatively high cross-section for an incoming tritium particle (triton). During the neutron irradiation, the target produces a triton, which has sufficiently high energy to cause the so-called secondary charged particle activation on an indicator nuclide. If the product of this reaction is a radioactive nuclide, its activity must be proportional to the amount of generated tritium. A comprehensive set of irradiations were performed at the Training Reactor of the Budapest University of Technology and Economics. The following charged particle reactions were observed and investigated: 27Al(t,p)29Al; 26Mg(t,p)28Mg; 26Mg(t,n)28Al; 32S(t,n)34mCl; 16O(t,n)18F; and O(t,α)18N17. The optimal atomic ratio of the indicator elements and 6Li was also investigated. The reaction rates were estimated using calculations with the MCNPX Monte Carlo particle transport code. The trend of the measured and the simulated data are in good agreement, although accurate data for triton induced reaction cross-sections cannot be found in the literature. Once the technique is calibrated with a reference LSC (Liquid Scintillation Counting) measurement, a new passive method becomes available for tritium production rate measurements.

The neutronic measurements in the ITER Test Blanket Modules (TBMs) aim at the determination of several physical quantities such as tritium production rate, heat load, and neutron spectrum, which are difficult to measure inside the hermetic zone by active detectors. For this reason activation detectors (foils) are planned to be used. The measurement circumstances, such as the irradiation time and the accessibility of the foils, highly depend on the detailed design of the TBMs and are not yet completely specified. Therefore, the flexible design requires an objective qualification method to select the optimal foil set for an actual experiment.The qualification method presented in this paper assigns a quality factor to each foil set. This factor depends on the orthogonality of the cross sections, the energy range covered by the foils and the discreteness of the energy range. A computer code has been developed to calculate the quality factor based on a calculated neutron spectrum and cross section data available in the nuclear data libraries. This study has been made in the framework of the Work Package No. 5 (measurement techniques development for breeder blankets) of the Breeding Blanket Developments for Fusion Reactors (EUROBREED) Goal Oriented Training Project supported by EFDA. (C) 2011 Elsevier B.V. All rights reserved.

This paper presents the development of a 3D time dependent calculation scheme for graphite moderated molten salt reactors (MSRs). The neutronics was modelled by diffusion and the delayed neutron precursor equations were extended with a convection term to take into account the drift of precursors. To account for the heat transfer in the fuel, a 1D heat convection equation was applied. All fuel channels were calculated individually. Heat transfer in the moderator was described by the 3D heat conduction equation, thus thermally connecting the fuel channels. The computational scheme was applied for the Molten Salt Reactor Experiment (MSRE). Steady-state and time dependent simulations were carried out, including the investigation of a local fuel channel blocking event where the space dependent effects are especially relevant.

The molten-salt reactor (MSR) concept is one of the most promising systems for the realisation of transmutation. The objective is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures. The procedure is the multi-step transmutation, in which the transformation is carried out in several consecutive steps of different neutron flux and spectrum. In order to implement this, a multi-region transmutational device, i.e. nuclear reactor or sub-critical system is proposed, in which several separate flow-through irradiation rooms are formed with various neutron spectra and fluxes. The paper presents calculations that were performed for a special 5-region version of the multi-region molten-salt reactor.

Nowadays the molten salt reactor (MSR) concept seems to revive as one of the most promising systems for the realization of transmutation. In the molten salt reactors and subcritical systems the fuel and material to be transmuted circulate dissolved in some molten salt. The main advantage of this reactor type is the possibility of the continuous feed and reprocessing of the fuel. In the present paper a novel molten salt reactor concept is introduced and its transmutational capabilities are studied. The goal is the development of a transmutational technique along with a device implementing it, which yield higher transmutational efficiencies than that of the known procedures and thus results in radioactive waste whose load on the environment is reduced both in magnitude and time length. The procedure is the multi-step time-scheduled transmutation, in which transformation is done in several consecutive steps of different neutron flux and spectrum. In the new MSR concept, named “multi-region” MSR (MRMSR), the primary circuit is made up of a few separate loops, in which salt-fuel mixtures of different compositions are circulated. The loop sections constituting the core region are only neutronically and thermally coupled. This new concept makes possible the utilization of the spatial dependence of spectrum as well as the advantageous features of liquid fuel such as the possibility of continuous chemical processing etc. In order to compare a “conventional” MSR and a proposed MRMSR in terms of efficiency, preliminary calculational results are shown. Further calculations in order to find the optimal implementation of this new concept and to emphasize its other advantageous features are going on.

In molten salt reactors, the delayed neutron precursors continuously change their position in the reactor and primary loop due to the circulation of the fuel. Therefore, the influence of the delayed neutrons on the reactivity differs from that of solid fuel reactors. The paper describes a method, which is developed in order to modify the program MCNP so that it can take into account the transport of the delayed neutron precursors, and therefore it is capable to calculate the reactivity loss of reactors with circulating fuel. With the aid of the modified version of MCNP, calculations were performed on a simple homogeneous reactor with cylindrical core. The method was also applied to perform calculations on the reactivity loss of the MSRE (Molten-Salt Reactor Experiment, Oak Ridge National Laboratories, 1960s). The results obtained are in good agreement with the theoretical predictions and the measured reactivity loss values.