On-site Refueling Research Articles

On-site solar powered refueling stations for green hydrogen production and distribution: performances and costs

Today, the hydrogen is considered an essential element in speeding up the energy transition and generate important environmental benefits. Not all hydrogen is the same, though. The “green hydrogen”, which is produced using renewable energy and electrolysis to split water, is really and completely sustainable for stationary and mobile applications. This paper is focused on the techno-economic analysis of an on-site hydrogen refueling station (HRS) in which the green hydrogen production is assured by a PV plant that supplies electricity to an alkaline electrolyzer. The hydrogen is stored in low pressure tanks (200 bar) and then is compressed at 900 bar for refueling FCHVs by using the innovative technology of the ionic compressor. From technical point of view, the components of the HRS have been sized for assuring a maximum capacity of 450 kg/day. In particular, the PV plant (installed in the south of Italy) has a size of 8MWp and supplies an alkaline electrolyzer of 2.1 MW. A Li-ion battery system (size 3.5 MWh) is used to store the electricity surplus and the grid-connection of the PV plant allows to export the electricity excess that cannot be stored in the battery system. The economic analysis has been performed by estimating the levelized cost of hydrogen (LCOH) that is an important economic indicator based on the evaluation of investment, operational & maintenance and replacement costs. Results highlighted that the proposed on-site configuration in which the green hydrogen production is assured, is characterized by a LCOH of 10.71 €/kg.

3D Analysis of a multipurpose modular reactor for small energy production

The Fixed Bed Nuclear Reactor (FBNR) is a modular reactor for small energy needs (70-120 MWel). FBNR is pressurized water reactor, but uses GEN-IV reactors fuel. UO2 kernels of 500 μm diameter with 25 μm thick SiC cladding are embedded in graphite matrix form the spherical fuel elements with a diameter of 15 mm and covered by 300 μm thick Zircaloy-4 cladding. FBNR has high level safety features and operates without On-site Refueling. Approximately 1.62 million spherical fuel elements are suspended in a core height of 200 cm and inner and outer diameters of 32 cm and 224 cm by upwards flowing light water coolant with 280°C entry and 340°C exit temperatures. The unit lattice geometry is dodecahedron with fuel element in center and surrounded with light water coolant-moderator.In the present work, 9% enriched UO2 fuel is used. Neutronic analysis has been performed with the MCNPX-2.7.0 code. As, it is not practical to tackle millions of fuel elements separately, at first run unit fuel cell calculations are performed in spherical geometry for one single lattice consisting of separate fuel, cladding and volume-equivalent moderator regions. In the second run, the fuel cell is homogenized. Infinite cell calculations have been executed for variable moderator/fuel (H/235U) ratios in order to determine under- and over-moderated criticality values. The criticality values for the homogenized cell geometry are slightly lower and remains <5% range, that is, on the conservative side, which is acceptable for the purpose of this study. This allows us to homogenize the entire core for the full 3-D reactor with core and reflector regions. Reactor criticality increases with increasing H/235U ratios in the, under moderated region, where H/235U = 18 is selected for further investigations. Temporal variation of the reactor criticality is pursued for full reactor power of 210 MWth (~70 MWel net) with keff = 1.36 at the beginning of life down to keff = 1.02 over 1260 days with a plant factor of 1.0. The final fuel burn up was found as 57.3 GWd/ton.

Ammonia as hydrogen carrier for realizing distributed on-site refueling stations implementing PEMFC technology

Ammonia is a particularly promising hydrogen carrier due to its relatively low cost, high energy density, its liquid storage and to its production from renewable sources. Thus, in recent years, great attention is devoted to this fuel for realizing next generation refueling stations according to a carbon-free energy economy. In this paper a distributed onsite refueling station (200 kg/day of hydrogen filling 700-bar HFCEVs (Hybrid Fuel Cell Electric Vehicles) with about 5 kg of hydrogen in 5 min), based on ammonia feeding, is studied from the energy and economic point of views. The station is designed with a modular configuration consisting of more sections: i) the hydrogen production section, ii) the electric energy supplier section, iii) the compression and storage section and the refrigeration/dispenser section. The core of the station is the hydrogen production section that is based on an ammonia cracking reactor and its auxiliaries; the electric energy demand necessary for the station operation (i.e. the hydrogen compression and refrigeration) is satisfied by a PEMFC (Proton-Exchange Membrane Fuel Cell) power module. Energy performance, according to the hydrogen daily demand, has been evaluated and the estimation of the levelized cost of hydrogen (LCOH) has been carried out in order to establish the cost of the hydrogen at the pump that can assure the feasibility of this novel refueling station.

Techno-economic calculations of small-scale hydrogen supply systems for zero emission transport in Norway

In Norway, where nearly 100% of the power is hydroelectric, it is natural to consider water electrolysis as the main production method of hydrogen for zero-emission transport. In a startup market with low demand for hydrogen, one may find that small-scale WE-based hydrogen production is more cost-efficient than large-scale production because of the potential to reach a high number of operating hours at rated capacity and high overall system utilization rate. Two case studies addressing the levelized costs of hydrogen in local supply systems have been evaluated in the present work: (1) Hydrogen production at a small-scale hydroelectric power plant (with and without on-site refueling) and (2) Small hydrogen refueling station for trucks (with and without on-site hydrogen production). The techno-economic calculations of the two case studies show that the levelized hydrogen refueling cost at the small-scale hydroelectric power plant (with a local station) will be 141 NOK/kg, while a fleet of 5 fuel cell trucks will be able to refuel hydrogen at a cost of 58 NOK/kg at a station with on-site production or 71 NOK/kg at a station based on delivered hydrogen. The study shows that there is a relatively good business case for local water electrolysis and supply of hydrogen to captive fleets of trucks in Norway, particularly if the size of the fleet is sufficiently large to justify the installation of a relatively large water electrolyzer system (economies of scale). The ideal concept would be a large fleet of heavy-duty vehicles (with a high total hydrogen demand) and a refueling station with nearly 100% utilization of the installed hydrogen production capacity.

Evaluating the Cost, Safety, and Proliferation Risks of Small Floating Nuclear Reactors.

It is hard to see how our energy system can be decarbonized if the world abandons nuclear power, but equally hard to introduce the technology in nonnuclear energy states. This is especially true in countries with limited technical, institutional, and regulatory capabilities, where safety and proliferation concerns are acute. Given the need to achieve serious emissions mitigation by mid-century, and the multidecadal effort required to develop robust nuclear governance institutions, we must look to other models that might facilitate nuclear plant deployment while mitigating the technology's risks. One such deployment paradigm is the build-own-operate-return model. Because returning small land-based reactors containing spent fuel is infeasible, we evaluate the cost, safety, and proliferation risks of a system in which small modular reactors are manufactured in a factory, and then deployed to a customer nation on a floating platform. This floating small modular reactor would be owned and operated by a single entity and returned unopened to the developed state for refueling. We developed a decision model that allows for a comparison of floating and land-based alternatives considering key International Atomic Energy Agency plant-siting criteria. Abandoning onsite refueling is beneficial, and floating reactors built in a central facility can potentially reduce the risk of cost overruns and the consequences of accidents. However, if the floating platform must be built to military-grade specifications, then the cost would be much higher than a land-based system. The analysis tool presented is flexible, and can assist planners in determining the scope of risks and uncertainty associated with different deployment options.

Power-to-Gas Hydrogen: Techno-economic Assessment of Processes towards a Multi-purpose Energy Carrier

The present work investigates Power-to-Gas (PtG) options for variable Renewable Electricity storage into hydrogen through low-temperature (alkaline and PEM) and high-temperature (SOEC) water electrolysis technologies. The study provides the assessment of the cost of the final product when hydrogen is employed for mobility (on-site refueling stations), electricity generation (by fuel cells in Power-to-Power systems) and grid injection in the natural gas network. Costs estimations are performed for 2013-2030 scenarios. A case study on the impact of variable Renewable Electricity storage by hydrogen generation on the Italian electricity and mobility sectors is presented.

Physics characteristics of the fixed bed nuclear reactor

This article is about calculation of fuel burn-up and some physics parameters of the Fixed Bed Nuclear Reactor (FBNR), using spherical fuel elements similar to TRISO fuel, which is investigated by SRAC code. The fuel that was used is a uranium dioxide sphere (UO2) with U235 enrichment of 5%. The main results of this study can be summarized as follows:(1)Variation of nuclear isotope densities and effective neutron multiplication coefficient in process of fuel burn-up is determined and the reactor can operate during the time of 2.73 years, without on-site refueling. The U235 burn-up reaches value 71.84% of its initial density at the end of the reactor fuel cycle.(2)When the core height is over 120 cm, the reactor control can be carried out by the movement of a core height level limiter, without the fine control rods in the core center.(3)The kinetic parameters of the reactor (the prompt-neutron lifetime and effective delayed neutron fraction) change in process of reactor operation.

Burnup characteristic of CERMET and TRISO fuel particles used in Boiling Water Reactor with Pebble-Bed Coated Particles

The reactor concept of boiling water reactor with pebble-bed coated particles (BWR-PB), as a 890MW(th) without the need of on-site refueling, is originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by boiling light water. At the lower coolant temperature, the TRISO fuel particles demonstrate an unacceptable irradiation swelling in the silicon carbide coating outer layer during a fuel cycle. This integrity problem, under low temperature irradiation, related to the accumulation of atomic displacements in the graphite lattice structure (Winger Energy) is due to insufficient annealing at temperatures below 260C. Instead, an alternative Spherical Cermet Fuel (SCF) concept is proposed. It consists of coated UO2 kernels embedded in a Zirconium matrix which is then over-coated with a protective outer layer of Zirconium. This new SCF concept may offer the following advantages: high thermal conductivity, low fuel temperatures, good fission product retention and good performance at high burnup. The SCF concept aims at long life operation capability and neutronic feasibility in BWR-PB reactor. This paper is focused on the evaluation of some nuclear parameters of a SCF concept, for a BWR-PB reactor, and the comparison with the previous TRISO fuel calculation results. The calculated burnup characteristic parameters are k-infinity, Instantaneous conversion ratio, Fuel isotopic and Neutron energy spectra. The investigation of this latter has found out that burnup characteristic parameters are changed slightly between the two fuel concepts. In fact, using SCF reduces the lifetime operation by 2 equivalent full power years in each BWR-PB assembly’s part, mainly due to the larger thermal component in the neutron spectrum for TRISO fuelling compared to the SCF one. The concentration of Uranium in the SCF spent fuel is similar to that contained in TRISO. However, the plutonium isotopic composition shows a significant deviation between the two concepts. Besides, in both fuel concepts, the fuel fertile matter is more exploited from the bottom to the top of BWR-PB fuel assembly.

Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

The Atoms For Peace Reactor (AFPR-100), as a 100MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY.

Neutronic simulation of a research reactor core of (232Th, 235U)O 2 fuel using MCNPX2.6 code

The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of 233U converted from 232Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.

Design Study of Small Pb-Bi Cooled Non-Refueling Nuclear Power Reactors (SPINNORs)

Small Pb-Bi Cooled Non-refueling Nuclear Power Reactors(SPINNORs) are small lead-bismuth cooled nuclear power reactors with fast neutron spectrum that could be operated for 20 years without on-site refuelling. In this study the results of neutronic design for 62.5-250MWt NPPs are discussed. The reactor cores are designed to have near zero (less then one effective delayed neutron fraction) burn-up reactivity swing during the burn-up period to avoid a possibility of super-prompt critical accident. The basic concept is that the reactor core is adjusted so that its internal conversion ratio is nearly one. During the reactor operation fissile material accumulates in this central region, which helps to compensate fissile material loss in the peripheral core region and also contributes to negative coolant loss reactivity effect. In general higher fuel volume fraction is needed for smaller core.

PASCAR: Long burning small modular reactor based on natural circulation

PASCAR is a 100 MWt/35 MWe lead–bismuth-cooled small modular reactor which requires no on-site refueling and well suits to be used as a distributed power source in either a single unit or a cluster for electricity, heat supply, and desalination. This paper includes both steady-state and transient performance evaluations for neutronics and thermal–hydraulics. Through design optimization studies for minimizing a burn-up reactivity loss, the metallic fuels-loaded core was designed with less than 1$ reactivity swing over 20-year cycle. A radial peaking power location shows the slow inward migration from outer enrichment zones while maintaining peaking factor within 1.35, reducing radiation damage and corrosion duty of high temperature environments. Equipped with coolant flow path large enough to ensure low pressure drop, this reactor is intended to operate by only natural circulation of chemically inert coolant within relatively low temperature range, 320–420 °C. Peak outlet temperature is nearly 450 °C where an Al-containing duplex cladding has sufficient corrosion resistance. Despite of 50% decrease of fuel thermal conductivity after swelling, inherent negative reactivity feedback and passive decay heat removal capability could secure an ample safety margin of peak fuel centerline temperature in tow safety analyses, unprotected transient overpower and unprotected loss of heat sink. The likelihood of loss of coolant, loss of flow, and local blockage is virtually eliminated by employing respectively a double-walled vessel, pump-less cooling, and cross-flow allowed open square assemblies. Simple fabrication, modular construction, and long burning cycle would compensate for economic disadvantages over smaller power and lower temperature than those of conventional fast reactors.

An innovative nuclear reactor for electricity and desalination

A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are being developed by the INPRO program of the International Atomic Energy Agency (IAEA). It is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The fixed bed nuclear reactor (FBNR) is a small nuclear reactor that meets all the requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve in a dual purpose plant to produce simultaneously both electricity and desalinated water, thus making it especially suitable to the needs of the Middle-East Countries. FBNR is being developed with the support of the IAEA under its program of small reactors without on-site refueling. The reactor uses the pressurized water reactor technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability, and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. Copyright © 2010 John Wiley & Sons, Ltd.

Benchmark results for the “BWR-PB” fuel assembly based on pebble-bed coated particles with uranium dioxide and gadolinia poison

This paper presents an innovative concept for a long life, boiling water reactor with pebble-bed coated particles (BWR-PB), small nuclear reactor without open-vessel refuelling. Small nuclear reactors without the need for on-site refuelling have a greater simplicity, better compliance with passive safety systems and are more adequate for countries with small electric grids and limited investment capabilities. Besides, it will help the developing countries to develop their economy, but at the same time these reactors must not lead to the proliferation of a nuclear material, which could be used to fabricate weapons. This study contributes to the evaluation of some neutronic parameters with the aim of predicting both their static and dynamic behaviour before starting on concrete processes, an optimization of fuel assembly parameters for fuel cycle characteristics improvement and a possibility of assembly lifetime extension. In this work, all the parameters as well as infinite multiplication factor, spectrum index, neutron spectra and instantaneous conversion ratio were calculated for the fuel assembly with gadolinia poison and without any poison. The evaluation of this parameters shows that the fuel assembly life time can be extended for a reasonably period without outside refuelling. So, the aim of BWR-PB concept was verified.

The utilization of thorium for long-life small thermal reactors without on-site refueling

Thorium cycle has many advantages over uranium cycle in thermal and intermediate spectrum nuclear reactors. In addition to large amount of resources in the world which up to now still not utilized optimally, thorium based thermal reactors may have high internal conversion ratio so that they are very potential to be designed as long-life reactors without on-site refueling based on thermal spectrum cores. In this study preliminary study for application of thorium cycle in some of thermal reactors has been performed. We applied thorium cycle for small long-life high temperature gas reactors without on-site refueling. Calculation results using SRAC code show that 10 years lifetime without on-site refueling can be achieved with excess reactivity of about 10% dk/k. The next application of thorium cycle has been employed in long-life small and medium PWR cores without on-site refueling. Relatively high fuel volume fraction is also applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 20–300 MWth PWR with maximum excess reactivity of a few %dk/k. The last application of thorium cycle has been employed in long-life BWR cores without on-site refueling. Relatively high fuel volume fraction is applied to get relatively hard spectrum, small size, and high internal conversion ratio. In the current study we have been able to reach more than 10 years lifetime without on-site refueling for 100–600 MWth BWR with maximum excess reactivity of a few %dk/k.

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.

The fixed bed nuclear reactor concept

In the present work, the basic features of a new reactor type, the so-called Fixed Bed Nuclear Reactor (FBNR) is presented. FBNR is a small reactor (40 MWe) without the need of on-site refueling. It utilizes the PWR technology but uses the HTGR type fuel elements. It has the characteristics of being simple in design, modular, inherent safety, passive cooling, proliferation resistant, and reduced environmental impact. The study comprises reactor description, fuel element description, criticality calculations. A series of one dimensional criticality calculations are conducted with SCALE5 using S N methods. S N calculations with SCALE5 have resulted for the cold reactor (20 °C, 1 bar) k ∞ = 1.4408 and for the hot reactor (308 °C, 160 bar), based on the average inlet–outlet temperatures, k ∞ = 1.40003 for the spherical fuel element cell. Time calculations have been pursued for 12 years. Three dimensional criticality calculations benchmarking are conducted with MCNP5-1.4 using Monte Carlo methods and have yielded k ∞ = 1.45673 for the hot FBNR unit cell.

Neutronic performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three type of fuels: MOX, Nitride and Metal are compared and discussed. In general MOX fuel (UO 2–PuO 2) has lower atomic density compared to the nitride or metal fuel, but MOX fuel has some advantages such as higher Doppler coefficient, high melting point and availability. Nitride fuel has advantages such as high density, high thermal conductivity, and high melting point, but need N-15 to avoid C-14 problems. The results show that nitride fuel as well as MOX fuel can be used to develop 25–100 MWe (75–300 MWth) Pb–Bi cooled long life reactors without on-site fuelling. The results show that nitride fuels have more superior neutronic characteristics compared to that of MOX fuel due to higher density. However, in the large power level both fuels can be easily applied. In lower power level the MOX fuel need higher fuel volume fraction to reach the comparable target of nitride fuelled cores.

Options for small and medium sized reactors (SMRs) to overcome loss of economies of scale and incorporate increased proliferation resistance and energy security

The designers of innovative small and medium sized reactors pursue new design and deployment strategies making use of certain advantages provided by smaller reactor size and capacity to achieve reduced design complexity and simplified operation and maintenance requirements, and to provide for incremental capacity increase through multi-module plant clustering. Competitiveness of SMRs depends on the incorporated strategies to overcome loss of economies of scale but equally it depends on finding appropriate market niches for such reactors. For many less developed countries, these are the features of enhanced proliferation resistance and increased robustness of barriers for sabotage protection that may ensure the progress of nuclear power. For such countries, small reactors without on-site refuelling, designed for infrequent replacement of well-contained fuel cassette(s) in a manner that impedes clandestine diversion of nuclear fuel material, may provide a solution. Based on the outputs of recent IAEA activities for innovative SMRs, the paper provides a summary of the state-of-the-art in approaches to improve SMR competitiveness and incorporate enhanced proliferation resistance and energy security.

Small PWR Using Coated Particle Fuel of Thorium and Plutonium

An innovative concept of PFPWR50 for district heating has been studied, which is a small PWR of 50MWt capability using coated particle fuels with conventional zircaloy cladding. This concept takes advantages of fuel integrity against fission products release of coated particle fuels and a high reliability of PWR technology based on the long history of a successful operation. We have investigated burnup characteristics of fuel rods, assemblies, and reactor cores by the calculation code SRAC95 in order to establish a core concept of long life without on-site refueling. The loading pattern of assemblies with various concentrations of burnable poison is optimized to obtain a flat excess reactivity during the core life in order to eliminate a soluble boron control system. The core life of a cycle is about 8.9 equivalent full power years. And we have also studied the applicability of SiC/SiC composite cladding in place of zircaloy cladding, which is now under development for gas cooled fast reactor fuels. It could be applicable to high burnup fuel rods for a long term operation. From the calculation results, it is found out that the burnup characteristics do not change significantly with SiC cladding and contribute to elongate the core life to 9.2 equivalent full power years.