Guglielmo Lomonaco

Generation IV reactor designs, operation and fuel cycle

This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

The use of Th in HTR: State of the art and implementation in Th/Pu fuel cycles

Nowadays nuclear is the only greenhouse-free source that can appreciably respond to the increasing worldwide energy demand. The use of Thorium in the nuclear energy production may offer some advantages to accomplish this task. Extensive R&D on the thorium fuel cycle has been conducted in many countries around the world. Starting from the current nuclear waste policy, the EU-PUMA project focuses on the potential benefits of using the HTR core as a Pu/MA transmuter. In this paper the following aspects have been analysed: (1) the state-of-the-art of the studies on the use of Th in different reactors, (2) the use of Th in HTRs, with a particular emphasis on Th-Pu fuel cycles, (3) an original assessment of Th-Pu fuel cycles in HTR. Some aspects related to Thorium exploitation were outlined, particularly its suitability for working in pebble-bed HTR in a Th-Pu fuel cycle. The influence of the Th/Pu weight fraction at BOC in a typical HTR pebble was analysed as far as the reactivity trend versus burn-up, the energy produced per Pu mass, and the Pu isotopic composition at EOC are concerned. Although deeper investigations need to be performed in order to draw final conclusions, it is possible to state that some optimized Th percentage in the initial Pu/Th fuel could be suggested on the basis of the aim we are trying to reach.

Preliminary Evaluation of a Nuclear Scenario Involving Innovative Gas Cooled Reactors

In order to guarantee a sustainable supply of future energy demand without compromising the environment, some actions for a substantial reduction of CO2 emissions are nowadays deeply analysed. One of them is the improvement of the nuclear energy use. In this framework, innovative gas-cooled reactors (both thermal and fast) seem to be very attractive from the electricity production point of view and for the potential industrial use along the high temperature processes (e.g., H2 production by steam reforming or I-S process). This work focuses on a preliminary (and conservative) evaluation of possible advantages that a symbiotic cycle (EPR-PBMR-GCFR) could entail, with special regard to the reduction of the HLW inventory and the optimization of the exploitation of the fuel resources. The comparison between the symbiotic cycle chosen and the reference one (once-through scenario, i.e., EPR-SNF directly disposed) shows a reduction of the time needed to reach a fixed reference level from ∼170000 years to ∼1550 years (comparable with typical human times and for this reason more acceptable by the public opinion). In addition, this cycle enables to have a more efficient use of resources involved: the total electric energy produced becomes equal to ∼630 TWh/year (instead of only ∼530 TWh/year using only EPR) without consuming additional raw materials.

Assessment of LWR-HTR-GCFR Integrated Cycle

Preliminary analyses already performed showed that innovative GCRs, both thermal and fast, are very promising candidate to reach the Gen-IV sustainability goal. The integrated LWR-HTR-GCFR basically aims at closing the current nuclear fuel cycle: in principle, thanks to the unique characteristics of Helium coolant reactors, LWR SNF along with DU become valuable material to produce energy. Additionally, burning HMs of LWR SNF means not only a drastic reduction in the Unat demand but also a remarkable decrease in the long-term radiotoxic component of nuclear waste to be geologically stored. This paper focuses on the analyses of the LWR-HTR-GCFR cycle performed by the University of Pisa in the frame of the EU PUMA project (6th FP). Starting from a brief outline of the main characteristics of HTR and GCFR concepts and of the advantages of linking LWR, HTR and GCFR in a symbiotic way, this paper shows the integrated cycle involving a typical LWR (1000 MWe), a PBMR (400 MWth) and a GCFR-“E” (2400 MWth). Additionally, a brief overview of the main technological constraints concerning (Pu

Analysis of Pebble-Fueled Zone Modeling Influence on High-Temperature Reactor Core Calculations

Abstract The pebble bed gas-cooled reactor is one of the most promising concepts among the Generation III+ and Generation IV reactors. Currently, the pebble bed modular reactor (PBMR) design, both U and Pu and minor actinide fueled, is being developed. Modeling the arrangement of coated particles (CPs) inside a spherical region like a pebble seems to be an important issue in the frame of calculations. To use the (relatively) old Monte Carlo codes without any correction, some approximations are often introduced. Recent Monte Carlo codes like MCNP5 and some new original subroutines that we have developed allow the possibility of obtaining more detailed and more physically correct geometrical descriptions of this kind of system. Some studies on modeling pebbles and pebble bed cores have already been carried out by other researchers, but these works are substantially limited to AVR-type UO2-fueled pebbles. However, the impact of approximated models on fuel mass, reactivity, and reactor life prediction has not yet been investigated for new PBMR-type pebbles. At the same time, an assessment of introducing a stochastic CP arrangement is not so widespread. Analyzing two PBMR pebbles, one Pu- and the other U-fueled, this paper focuses on quantifying errors due to the different approximations generally used to describe the CP lattice inside a high-temperature reactor pebble bed core, as far as mass of fuel, reactivity, and burnup simulation are concerned. This aim was reached also through a new feature implemented in the MCNP5 code, i.e., capability to treat (pseudo) stochastic geometries. Later, we compared the initial mass of fuel, keff, and isotopic evolution versus burnup of some approximated pebble models with the reference model, built by means of this new MCNP5 feature.

The Capabilities of HTRs to Burn Actinides and to Optimize Plutonium Exploitation

At present, the 125 GWe of nuclear power in the European Union produce about 3000 tons of spent fuel annually, containing about 25 tons of plutonium, 2.5 tons of minor actinides (MA) and about 100 tons of fission products, of which 3.1 tons are long-lived fission products. Actual reprocessing of LWR fuel and a first recycling as mixed plutonium and depleted uranium oxide fuel (MOX) in LWR already contribute to a significant reduction of waste volumes and radiotoxicity. However HTRs have some characteristics which make them particularly attractive: intrinsic safety, cost-effectiveness, reduced thermal pollution, capability of increasing energy availability (with the use of Pu-Th cycle) and of minimizing actinides radiotoxicity and volume of actinides. In this paper particularly the last item is investigated. Symbiotic fuel cycles of LWR and HTR can reach much better waste minimization performances. It happens because of the specific features of HTRs cores that leads to an ultra-high burnup and, last but not least, the ability to accommodate a wide variety of mixtures of fissile and fertile materials without any significant modification of the core design. This property is due to a decoupling between the parameters of cooling geometry and of neutronic optimization. In our calculations we considered a pebble-bed HTR using a Pu-based fuel (deriving from reprocessing of classical LWR fuel and/or weapons grade plutonium) at the maximum technological discharge burnup. As results, we find, at EOL (End Of Life), a relatively small amounts of residual Pu and MA produced in terms of quantities and of radiotoxicities. Furthermore we used in our calculations a different type of fuel based on a mixture of Pu and Th to try to optimize the previous results and to increase energy availability. Calculations have been done using MCNP-based burnup codes, capable of treating 3-D complex geometry and ultra-high burnup.Copyright

GCFR Coupled Neutronic and Thermal-Fluid-Dynamics Analyses for a Core Containing Minor Actinides

Problems about future energy availability, climate changes, and air quality seem to play an important role in energy production. While current reactor generations provide a guaranteed and economical energy production, new nuclear power plant generation would increase the ways and purposes in which nuclear energy can be used. To explore these new technological applications, several governments, industries, and research communities decided to contribute to the next reactor generation, called “Generation IV.” Among the six Gen-IV reactor designs, the Gas Cooled Fast Reactor (GCFR) uses a direct-cycle helium turbine for electricity generation and for a CO2-free thermochemical production of hydrogen. Additionally, the use of a fast spectrum allows actinides transmutation, minimizing the production of long-lived radioactive waste in an integrated fuel cycle. This paper presents an analysis of GCFR fuel cycle optimization and of a thermal-hydraulic of a GCFR-prototype under steady-state and transient conditions. The fuel cycle optimization was performed to assess the capability of the GCFR to transmute MAs, while the thermal-hydraulic analysis was performed to investigate the reactor and the safety systems behavior during a LOFA. Preliminary results show that limited quantities of MA are not affecting significantly the thermal-fluid-dynamics behavior of a GCFR core.

The Influence of the Packing Factor on the Fuel Temperature Hot Spots in a Particle-Bed GCFR

In the recent past the so-called GCFR has been again a subject of study by the international scientific community. This type of reactors, although still in a preliminary stage of development, is a very interesting perspective because combines the positive characteristics common to all the fast reactors with those of the reactors cooled by helium. Up to now, almost all the analyses on the GCFR thermodynamic aspects have been performed starting from a “global” point of view: generally the core has been modelled as a porous medium and only the global parameters have been taken into account. The local effects have been included in adhoc corrective peak factors. The analyses carried out in the present research will be devoted to the characterization of the local effects, on a microscopic scale. In order to have reliable “global” nuclear and thermal-fluid-dynamic data, the performed analyses will be based on simulations previously performed using the RELAP5-3D code, assuming as input parameters the ETDR core ones. For each considered case, the variation ranges of the evaluated parameters have been estimated on the basis of the “best” and the “worst” cases. To summarize the obtained results, even in transient conditions, the variations of the considered input parameters are less significant for the local output values if compared to those due to the assumed packing factor. As a consequence, in a more general core calculation, the obtained local temperature (and velocity) values will have to be corrected by a proper factor that would have to take into account the results of this research.

Corrosion issues in high temperature gas-cooled reactor (HTR) systems

Abstract: The corrosion issues in high temperature reactors (HTRs) are probably less critical compared to such issues in other nuclear reactors. Nevertheless it is important to take them into account adequately. In this chapter we consider the following items: structural metallic material corrosion in high temperature conditions; graphite corrosion (mainly oxidation) behaviour in high temperature environments; the corrosion of the PBMR spent fuel tank in severely corrosive environments; the interaction between UO2 and C layer into the CP. The chapter ends with a short description of future trends related to the corrosion aspects specific to the HTR.

Preliminary CFD Assessment of an Experimental Test Facility Operating with Heavy Liquid Metals

The CFD analysis of a Venturi nozzle operating in LBE (key component of the CIRCE facility, owned by ENEA) is presented in this paper. CIRCE is a facility developed to investigate in detail the fluid-dynamic behavior of ADS and/or LFR reactor plants. The initial CFD simulations have been developed hand in hand with the comparison with experimental data: the test results were used to confirm the reliability of the CFD model, which, in turn, was used to improve the interpretation of the experimental data. The Venturi nozzle is modeled with a 3D CFD code (STAR-CCM