Preliminary analysis of tritium fuel cycle in Z-pinch-driven fusion–fission hybrid reactor

Attaining tritium self-sufficiency is indispensable in a Z-pinch-driven fusion–fission hybrid reactor(ZFFR).In this paper,a conceptual design is presented in which the Z-FFR tritium cycle system was divided into eight subsystems.A theoretical analysis of tritium inventory based on the mean residence time was performed to quantitatively obtain the tritium distribution in each subsystem.Tritium self-sufficiency judgment criteria were established using a tritium mass flow analysis method.The dependency relationships between the burning rate,tritium breeding ratio,extraction efficiency,and tritium self-sufficiency were also specified for the steady state.     

Tests in an MR reactor of experimental fuel elements of the VVÉR-1000 reactor

Conclusions In an MR reactor performance tests of 16 fuel assemblies, with elements having essentially the same structure as standard VVÉR-100 fuel elements, were carried out. Tests of five more fuel assemblies are continuing. Of the 16 assemblies, 13 were studied in a hot laboratory.The tests in the MR, carried out at high loads and with a large number of transition processes, as well as the postreactor studies, indicated that fuel elements of the specified design (with initial helium pressures of 1.96–2.45 MPa) have a high reliability. None of the elements of the fuel assemblies studied malfunctioned due to design defects or faults in their fabrication. During the tests the jackets were subject to a little oxidation and hydrogenation (zirconium-oxide film<3 m thick, hydrogen content less than 0.008% by mass), and their plasticity remained high (the relative elongation at the working temperature remained at the 20% level).Translated from Atomnaya Énergiya, Vol. 62, No. 5, pp. 312–317, May, 1987.     

Oxidation kinetics of Zry-4 fuel cladding in mixed steam–air atmospheres at temperatures of 1273–1473 K

This paper deals with the oxidation behavior of Zry-4 nuclear fuel cladding tubes in mixed steam–air atmospheres at temperatures of 1273 and 1473 K. The main goal is to study the oxidation kinetics of Zry-4 fuel cladding in dependence on the air fraction in steam in the range from 0% up to 100%. The purpose of this study is to provide experimental data suitable for an oxidation correlation, applicable for severe accident computer codes of nuclear power reactors. The influence of the air addition in steam on parameters of Zry-4 kinetic equation has been quantified using the results of weight gain measurements. At 1273 K, both pre-transition and post-transition regimes were treated. The results of weight gain measurements showed a strong dependence of the Zry-4 oxidation kinetics on the air fraction in steam, especially at 1473 and at 1273 K in the post-transition regime.     

Burnup characteristics and fuel cycle economics of mixed uranium–thorium fuel in a simplified small pebble bed reactor

The limitation of natural uranium resources and the improvement of economic values of nuclear reactors are important issues to be solved in the future development of these reactors. In our previous study, we presented an innovative design for simplifying a pebble bed reactor, and the optimization of this design showed that burnup values could be increased and natural uranium uses could be reduced. The purposes of the current study were to design a simplified pebble bed reactor by removing the unloading device from the reactor system and to further optimize the burnup characteristics of this reactor with a peu à peu fuel-loading scheme by introducing thorium in the fuel configuration as a fertile material. Another goal was to optimize the fuel composition so that the system could achieve even better burnup characteristics and use scarce uranium resources more efficiently. Using a specially developed computer code, we analyzed and optimized the performance of a 110-MWt simplified pebble bed reactor using a peu à peu fuel-loading scheme. An optimized design using 30% of fertile thorium mixed with uranium fuel with 15% ²³⁵U enrichment and a 7% packing fraction calculated to achieve a high burnup of 140 GWD/T for more than 21 years' operation time that could save 13 to 33% of natural uranium use compared with the savings noted in our previous study. Neutronic, burnup and fuel economic analysis for this optimized design are discussed in this study.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible.     

Startup and burnup strategy for Th–U/U–Pu fuel cycles in an EM2 reactor

General Atomics (GA) is developing the Energy Multiplier Module (EM²) which is a compact gas-cooled fast reactor as one of candidates of the Generation-IV nuclear energy systems. In the EM² core, low enriched uranium is used as igniting fuel and depleted uranium is used for converting and burning. It indicates that EM² can maintain critical operation for more than 30 years without refueling. To further study the Th–U fuel cycle performance in the EM², two kinds of start-up strategies with Th–U (Th + ²³³U) and semi Th–U (Th + enriched ²³⁵U) are evaluated. Neutronics characteristics, such as the effective multiplicity factor (k_eff) and conversion ratio (CR) are analyzed from neutron usage point of view. The simulated results for the two kinds of fuels are compared with the U–Pu fuel from the design of GA. The analysis gives an insight into the pros and cons of U–Pu and Th–U fuel cycles in terms of the breeding capability and the discharged radio-toxicity. The breeding performance of the second generation EM² is also presented and compared with that of the first generation EM². It indicates that the multi-generation EM² can deepen the burnup and reduce the waste management pressure for each kind of fuel loading strategy.     

Potential of a fusion–fission hybrid reactor using uranium for various coolants to breed fissile fuel for LWRs

In this study, neutronic performances of the (D,T) driven hybrid blankets, fuelled with UC₂ and UF₄, are investigated under first wall load of 5 MW/m². The fissile fuel zone is considered to be cooled with three coolants: gas (He or CO₂), flibe (Li₂BeF₄), and natural lithium. The behaviour of the UC₂ and UF₄ fuels are observed during 48 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. At the end of the operation time, calculations have shown that Cumulative Fissile Fuel Enrichment (CFFE) values varied between 5 and 8.5% depending on the fuel and coolant type. The best enrichment performance is obtained in UF₄ fuelled blanket with flibe coolant, followed by gas and natural lithium coolant. CFFE reaches maximum value (8.51%) in UF₄ fuelled blanket (in row #1) and flibe coolant mode after 48 months. The lowest CFFE value (4.71%) is in UC₂ fuelled blanket (in row #8) and natural lithium coolant at the end of the operation period. This enrichment would be sufficient for LWR reactor. At the beginning of the operation, tritium breeding ratio (TBR) values were 1.090, 1.3301 and 1.2489 in UC₂ fuelled blanket and 1.0772, 1.2433 and 1.1533 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. At the end of the operation, TBR reach 1.1820, 1.3983 and 1.3138 in UC₂ fuelled blanket and 1.2041,1.3266 and 1.2407 in UF₄ fuelled blanket for flibe, natural lithium and gas coolant, respectively. Nuclear quality of the plutonium increases linearly during the operation period. The isotopic percentage of ²⁴⁰Pu is higher than 5% in UF₄ and UC₂ fuel with flibe coolant, so that the plutonium component in these modes can never reach a nuclear weapon grade quality during the operation period. This is very important factor for safeguarding. The isotopic percentage of ²⁴⁰Pu is lower than 5% in UC₂ fuel with gas and natural lithium coolant. In these modes, operation period must be increased to safeguarding.     

A BWR fuel assembly design for efficient use of plutonium in thorium–plutonium fuel

The objective of this study is to develop an optimized BWR fuel assembly design for thorium–plutonium fuel. In this work, the optimization goal is to maximize the amount of energy that can be extracted from a certain amount of plutonium, while maintaining acceptable values of the neutronic safety parameters such as reactivity coefficients, shutdown margins and power distribution. The factors having the most significant influence on the neutronic properties are the hydrogen-to-heavy-metal ratio, the distribution of the moderator within the fuel assembly, the initial plutonium fraction in the fuel and the radial distribution of the plutonium in the fuel assembly. The study begins with an investigation of how these factors affect the plutonium requirements and the safety parameters. The gathered knowledge is then used to develop and evaluate a fuel assembly design. The main characteristics of this fuel design are improved Pu efficiency, very high fractional Pu burning and neutronic safety parameters compliant with current demands on UOX fuel.     

Compatibility of fuel-element cladding materials with gallium-containing uranium and uranium–plutonium oxide fuel

Powders of uranium oxide powder and mixed fuel containing 5 and 20 mass % plutonium and 0.4 and 5 mass % gallium were prepared by coprecipitation from nitrate solutions. Pelleted samples for testing were made by cold pressing and sintering. The compatibility of uranium oxide fuel and mixed uranium–plutonium fuel, containing 0.4 and 5 mass % gallium, with the zirconium alloy E-110 at 400 and 500°C and ChS-68 corrosion-resistant steel at 650 and 750°C over periods of 1000, 2000, and 3000 h was investigated. Metallographic and x-ray spectral microprobe analyses of diffusion samples established that there was no interaction and penetration of gallium into the zirconium alloy and steel. In addition, the diffusion coefficient of metallic gallium in zirconium alloy and the distribution of the elements on interaction of gallium with ChS-68 steel were evaluated.     

Advanced U–Np–Pu fuel to achieve long-life core in heavy water reactor

The objective of this paper is to look at the possibility of approaching the long-life core comparable with reactor life-time. The main issues are centered on U–Np–Pu fuel in a tight lattice design with heavy water as a coolant. It is found that in a hard neutron spectrum thus obtained, a large fraction of ²³⁸Pu produced by neutron capture in ²³⁷Np not only protects plutonium against uncontrolled proliferation, but substantially contributes in keeping criticality due to improved fissile properties (its capture-to-fission ratio drops below unit). Equilibrium fuel composition demonstrates excellent conversion properties that yield the burn-up value as high as 200 GWd/t at extremely small reactivity swings.     

Effects of molybdenum on microstructural evolution and mechanical properties in Zr–Nb alloys as nuclear fuel cladding materials

The Zr–Nb alloys were modified by doping of Mo as a minor alloying element to seek for the nuclear fuel cladding materials with better characteristics. The effects of Mo on microstructural evolution and mechanical properties in Zr–Nb alloys were systematically investigated and elucidated. Results showed that the martensitic microstructure, a mixture of lath martensites and lens martensites with internal twins, was observed in the alloys quenched from β-phase. Width of the lath martensite reduced with the increasing Mo concentration, and the volume fraction of lens martensite increased with increase in the Mo concentration. After final annealing, a new kind of precipitate, namely β-(Nb, Mo, Zr), was identified in the Mo-containing alloys. It was also found that Mo reduced the growth of the precipitates but increased their number density. Furthermore, Mo addition retarded the recrystallization process strongly and reduced the grain size significantly. In terms of the mechanical properties, Mo addition enhanced the yield strength and the ultimate tensile strength at room temperature, however decreased the ductility. The grain size strengthening was presumed as the greatest contributor in this system.     

Study for 228Th reduction in thermal reactor with Th—U fuel cycls

By using computer code WIMS/CENDL,the effects of some parameters,core configuration such as fuel element structure,neutron flox and burn-up,are discussed in this paper.It is shown that high neutron flux,small fuel rod diameter,large volume ratio of coolant to fuel,seed-blank heterogeneous core arrangement and 231 Pa chemical separation are necessary for reducing 228 Th production in reactor.     

Repatriation of Vinča RA reactor spent fuel

Abstract

The RA research reactor is located at the Vin?a Institute of Nuclear Sciences near Belgrade, Serbia. The reactor is a 6·5 MW, tank-type, heavy water moderated and cooled reactor of Russian design which commenced operation in 1959. After being temporarily shut down in 1984 for refurbishment, a final shutdown decision was made in 2002. Operations are underway to safely remove and repatriate the spent nuclear fuel (SNF) to the Russian Federation (RF), as well as to improve waste management throughout the Vin?a site and prepare a plan for reactor decommissioning. As a major activity within the Vin?a Institute Nuclear Decommissioning (VIND) Programme, the repatriation of over 8000 SNF elements containing 2·5 tons of uranium metal will significantly reduce nuclear proliferation and environmental safety risks confronting the current facility. Poor water quality in the SNF storage basins and degraded fuel integrity significantly challenge efforts to repackage and transport the SNF. This paper will focus on the activities related to SNF repackaging and shipment, report on progress, detail significant challenges and provide an overview of the fully integrated VIND project.     

Molten salt fast reactor with U–Pu fuel cycle

In the last two years it was discovered that solubility of PuF₃, UF₄ and AmF₃ in the eutectics LiF–NaF–KF are unexpectedly high (30 mol %, 45 mol % and 43 mol % correspondingly at 700 °C). This result opens the way for the development of the molten salt fast reactor with U–Pu nuclear fuel cycle (UPu-MSFR). The first calculations of the critical UPu-MSFR and subcritical MSR-burner of Am are presented.     

Characterization of molten 2LiF–BeF_2 salt impregnated into graphite matrix of fuel elements for thorium molten salt reactor

The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.     

Dissolution of spent nuclear fuel in carbonate–peroxide solution

This study shows that spent UO₂ fuel can be completely dissolved in a room temperature carbonate–peroxide solution apparently without attacking the metallic Mo–Tc–Ru–Rh–Pd fission product phase. In parallel tests, identical samples of spent nuclear fuel were dissolved in nitric acid and in an ammonium carbonate, hydrogen peroxide solution. The resulting solutions were analyzed for strontium-90, technetium-99, cesium-137, europium-154, plutonium, and americium-241. The results were identical for all analytes except technetium, where the carbonate–peroxide dissolution had only about 25% of the technetium that the nitric acid dissolution had.     

Fabrication of Si–C–N compounds in silicon carbide by ion implantation

The chemical variation and depth profile of silicon carbide implanted with nitrogen and overgrown with epitaxial layer has been studied using X-ray photoelectron spectroscopy (XPS). The results of this study have been supplemented by transmission electron microscopy (TEM) imaging and electron energy loss-spectroscopy (EELS) in an attempt to correlate the chemical and structural information. Our results indicate that the nitrogen implantation into silicon carbide results in the formation of the Si–C–N layer. XPS revealed significant change in the bonding structure and chemical states in the implanted region. XPS results can be interpreted in terms of the silicon nitride and silicon carbonitride nanocrystals formation in the implanted region which is supported by the electron microscopy and spectroscopy results.