Burnup

This paper shows some results obtained through simulating a VVER-1000 mock-up installed within the LR-0 research reactor in the Research Centre Řež (Czech Republic): the simulations have been performed by using the Serpent 2 and MCNP6 Monte Carlo codes. The mock-up is analysed in 6 different critical configurations, obtained with variation of coolant level and boron concentration. Some important nuclear parameters are evaluated in different positions of the mock-up, in order to know how the level of coolant influences the flux, the fission density, the reaction rate and the neutron spectrum changes, although the reactor remains with keff around 1.0 to less than 300 pcm of reactivity.

The fusion-fission hybrid reactor is a promising technology that is likely to assume an increasingly important role in the global energy scene in the coming years. This kind of reactor can use both the nuclear fusion and fission processes to produce energy: neutrons from fusion reactions are used to sustain the fission of a sub-critical system. This method allows to have an intrinsically safe facility, with higher efficiency than a fusion reactor itself and with a harder neutron energy spectrum than a fission reactor, which could be suitable for nuclear waste transmutation. This paper, in particular, analyzes a type of hybrid reactor for the transmutation of Minor Actinides (MA). Nuclear waste, in the oxide form, is inserted as an element of the First Wall (FW) of an ITER-like fusion reactor. The aim is to demonstrate the feasibility of the transmutation of the MA characterized by higher long term radiotoxicity into shorter lived nuclides. The neutron transport in a detailed 3D geometrical model of the ITER reactor (B-lite) was performed by MCNP6 code, while the transmutation of the MA loaded in a single element of the FW was performed by SERPENT2 code. A pulsed ITER-like irradiation scenario was used. The analysis, which must be considered as a preliminary feasibility study, lead to very promising results, which could be further improved with a longer DEMO-like irradiation scenario and a larger number of MA loaded (" fission waste ") elements loaded in the FW.

The use of passive gamma and neutron signatures from fission indicators is a common means 6 of estimating used fuel burnup, enrichment, and cooling time. However, while characteristic fission 7 product signatures such as 134 Cs, 137 Cs, 154 Eu, and others are generally reliable estimators for used 8 fuel burnup within the context where the assembly initial enrichment and the discharge time are 9 known, in the absence of initial enrichment and/or cooling time information (such as when applying 10

BREEDING CAPABILITY OF URANIUM AND THORIUM FUEL CYCLES FOR WATER COOLED REACTORS. Nuclear energy has contributed to fulfill the world energy demand especially in relation to the sustainable development of the world without any greenhouse effect to the environment for more than 50 years. The breeder reactors seem to have a similar trend with the renewable energies as a sustainable energy source. A fuel breeding is very essential for extending the sustainability of nuclear fuel resource and furthermore, it can be used to perform the sustainable development of the world. The present study intends to find the feasible region of design parameters for light or heavy water cooled reactors using thorium and uranium fuels which fulfill the required design characteristics such as breeding, negative void reactivity coefficient, comparable burn up with standard PWR, homogeneous core and large pin gap. The basic reactor design parameters of investigated systems are basically based on the water c...

Abstract The core concept of simultaneous transmutation of six long-lived fission products (LLFP: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) was investigated using MONJU fast reactor by maintaining all 6 LLFP nuclides support ratios at 1 or greater. By loading the six LLFP nuclides with the partial use of moderator into the blanket, the shield, the axial blanket, a part of gas plenum region and the axial shield, the effective half-lives of the LLFP nuclides were greatly reduced while maintaining the support ratios at 1 or greater. A new transmutation system for the six LLFPs based on a fast reactor core was constructed. It was also found that the basic core characteristics of the LLFP loaded core: the sodium void coefficient, Doppler coefficient and power distribution, did not differ greatly from those of normal conventional fast reactors.

Fuel burnup analysis requires a high computational cost for full core calculations, due to the amount of the information processed for the total reaction rates in many burnup regions. Indeed, they reach the order of millions or more by a subdivision into radial and axial regions in a pin-by-pin description. In addition, if multi-physics approaches are adopted to consider the effects of temperature and density fields on fuel consumption, the computational load grows further. In this way, the need to find a compromise between computational cost and solution accuracy is a crucial issue in burnup analysis. To overcome this problem, the present work aims to develop a methodological approach to implement a Reduced Order Model (ROM), based on Proper Orthogonal Decomposition (POD), in fuel burnup analysis. We verify the approach on 4 years of burnup of the TMI-1 unit cell benchmark, by reconstructing fuel materials and burnup matrices over time with different levels of approximation. The re...

In this research work pure urania and urania-gadolonia (5 and 10%) fuels were first produced by sol-gel technique. The effects of some operational parameters such as grinding, use of binder, compaction pressure, and reduction temperature on the porosity of the sintered fuels were investigated. The sintered fuel pellets were then coated with boron nitride (BN). This is achieved through chemcial vapor depositon (CVD) using boron tricholoride and ammonia. The coated samples were sintered at 1600 K. The burn up calculations by using WİMS-D/4 showed that boron nitride coated and gadolinia containing fuels had larger burnups than other fuels. The calculations were repeated at different pitch distances. The change of radius of the fuel pellet ot the moderator/fuel ratio did not introduce an improvement in the performance of fuel and all fuel gives the highest burnup at the present design values of a PWR.