A three-dimensional thermal and fluid dynamics analysis of a gas cooled subcritical fast reactor driven by a D–T fusion neutron source

The entire nuclear fuel cycle involves partitioning classification and transmutation recycling. The usage of a tokamak as neutron sources to burn spend fuel in a gas cooled subcritical fast reactor (GCSFR) reduces the amount of long-lived radionuclide, thus increasing the repository capacity.     

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.     

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.     

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.     

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.     

Thermoelastic analyses of cladding on beam transient of lead–bismuth cooled accelerator-driven system

The thermoelastic analyses of cladding for lead–bismuth cooled accelerator-driven system (ADS) are conducted for the beam transients. The beam transients are considered to be caused by the abnormal behavior of the accelerator and are peculiar to ADS. The program of the thermoelastic analyses is developed for the evaluation of the stresses of the cladding. This program is intended to analyze a fuel pin of a cylindrical model, and solves the thermoelastic problem by the use of the finite-element-method. The beam transients are analyzed by employing the ADS dynamic calculation code and the program of the thermoelastic analyses for the ADS designed by Japan Atomic Energy Agency. As a result, the transformation of the beam shape does not cause the cladding failure. However, the ductile failure is caused by the beam incident position change in several seconds. These results are also compared with those of the creep analyses conducted in the previous study, and both the creep and the ductile failure are revealed to be caused by the beam incident position change. Consequently, the beam incident position change is concluded to have a high risk of cladding failure.     

Study of a self-regulated nuclear burn wave regime in a fast reactor based on a thorium–uranium cycle

The possibility of a wave of slow nuclear burning in a fast reactor in thorium–uranium fuel cycle is investigated. The calculations were performed using a model based on the solution of a nonstationary nonlinear diffusion equation for a cylindrical homogeneous reactor using the concept of a radial geometric factor (buckling) and the effective multigroup approximation taking account of the nuclear kinetics of the precursors of delay neutrons and burnup and production of the main nuclides of the thorium–uranium fuel cycle. The calculations showed that the generation and propagation of a wave of nuclear burning traveling with velocity approximately 2 cm/yr are possible in a thorium–uranium medium. However, the addition of even small quantities of a construction material and coolant to the composition of the reactor makes it impossible to obtain the burn wave regime. A self-maintained nuclear burn regime is also established in this case and exists for a long time (∼5 yr), but the system does not transition into a regime with a nuclear burn wave propagating along the axis of the reactor.     

Use of the alloy Na−K−Cs as a coolant in a fast reactor: Preliminary results of safety characteristics optimization

A possible variant of a safe reactor operating on fast neutrons and cooled by the alloy Na−K−Cs is examined. The results of optimization investigations of the layout of a fast reactor with constraints on the functionals characterizing the internal self-shielding from ATWS-type accidents are presented. 2 tables, 5 references. Moscow Engineering-Physics Institute. Translated from Atomnaya énergiya. Vol. 88, No. 3, pp. 169–176, March, 2000.     

Transient and stability analysis of a BWR core with thorium–uranium fuel

The kinetic response of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material, in an integrated blanket–seed assembly, is presented in this work. Additionally an in-house code was developed to evaluate this core under steady state and transient conditions including a stability analysis. The code has two modules: (a) the time domain module for transient analysis and (b) the frequency domain module for stability analysis. The thermal–hydraulic process is modeled by a set of five equations, considering no homogeneous flow with drift-flux approximation and non-equilibrium thermodynamic. The neutronic process is calculated with a point kinetics model. Typical BWR reactivity effects are considered: void fraction, fuel temperature, moderator temperature and control rod density. Collapsed parameters were included in the code to represent the core using an average fuel channel. For the stability analysis, in the frequency domain, the transfer function is determined by applying Laplace-transforming to the calculated pressure drop perturbations in each of the considered regions where a constant total pressure drop was considered. The transfer function was used to study the system response in the frequency domain when an inlet flow perturbation is applied. The results show that the neutronic behavior of the core with thorium uranium fuel is similar to a UO₂ core, even during transient conditions. The stability and transient analysis show that the thorium–uranium fuel can be operated safely in current BWRs.     

Development and validation of a fast reactor core burnup code – FARCOB

A large fast breeder reactor requires the accurate estimation of power produced in different parts of the reactor core and blanket during any operating condition for a safe and economic operation through out reactor life time. A fast reactor core simulation code FARCOB based on multigroup diffusion theory has been developed in IGCAR for core simulation of PFBR reactor under construction. FARCOB uses centre mesh differencing scheme with triangular meshes in the X–Y plane. Steady state solution results match exactly with those of other reputed codes DIF3D and VENTURE for SNR-300 benchmarks. For burnup simulation, core is divided into radial and axial burnup zones and burnup equations are solved at constant power. Burnable fuel and blanket number densities are found and stored for each mesh, so that the user can alter burnup zones and core geometry after a burnup step. For validation, results of FARCOB has been compared with results of other institutes in two burnup benchmarks (ANL 1000 MWe benchmark and BN-600 hybrid core benchmark). It is found that FARCOB results match well with those of the other institutes.     

Fabrication of SiC–SiC composites for fuel cladding in advanced reactor designs

Both advanced fission reactor concepts and fusion energy systems demand materials that can survive extremely harsh operating environments having persistent high temperature and high neutron flux conditions. Silicon carbide fiber/silicon carbide matrix (SiC–SiC) composites have shown promise for these applications, which include fuel cladding and reactor structural components. However, the composite fabrication process is time consuming and the fabrication of complicated geometries can be difficult.In this work, SiC–SiC and carbon fiber–SiC composite samples were fabricated using chemical vapor infiltration (CVI), and the mechanical and thermal properties of samples with a range of densities and total infiltration times were characterized and compared. Both sample density and the reinforcing fiber material were found to have a very significant influence on the composite mechanical and thermal material properties. In particular, internal porosity is found to have a significant effect on the mechanical response, as can be observed in the crack propagation in low density samples. In order to better understand the densification of the composites, a computer model is being developed to simulate the diffusion of reactants through the fiber preform, and SiC deposition on the fiber surfaces. Preliminary modeling has been correlated with experimental results and shows promising results.     

A comprehensive study on neutronics of a lead–bismuth eutectic cooled accelerator-driven sub-critical system for long-lived fission product transmutation

The aim of this study is to investigate the high-level waste (HLW) transmutation and fissile breeding potentials of a lead–bismuth eutectic (LBE) cooled accelerator-driven system (ADS) for the various configurations (the target radius, R_T = 10–50 cm and the radial thickness of the sub-critical core, δ_SC = 50–80 cm) and for the various fuel compositions (the fuel volume fraction, VF_F = 10%, 12%, 15% and 20% and the fissile fraction, FF = 10–24%) under sub-critical condition. The long-lived fission products (LLFPs: ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs nuclides) and the uranium mono carbide (UC) ceramic fuel are considered as the HLW and the fissile fuel, respectively. The neutronic calculations have been performed per the incident proton (1000 MeV) with the high-energy Monte Carlo code MCNPX in coupled neutron and proton mode using the LA150 library. The numerical results bring out that the case of R_T = 30 cm, δ_SC = 80 cm, VF_F = 10% and FF = 23% is the optimum configuration and fuel composition, from the energy gain point of view, and has a high neutronic performance for an effective LLFP transmutation and fissile breeding.     

Coupled neutronics/thermal–hydraulics analysis of CANDU–SCWR fuel channel

At supercritical pressure condition, the thermal–hydraulics behavior of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal–physical properties across the pseudo-critical line. A coupling analysis of neutronics and thermal–hydraulics has become important for SCWR, because of the strong link between the water density and the neutron spectrum and subsequently the power distribution. The neutronics code Monte Carlo N-Particle code (MCNP) and the subchannel code Advanced Thermal–Hydraulics Analysis Subchannel (ATHAS) are used in a coupled way to better understand the design characteristics of a pressure tube type SCWR fuel channel. The results show that: the developed coupled code system can be used to analyze pressure tube type SCWR fuel bundles; improved radial fuel enrichment profile will optimize the coolant and cladding temperature distribution to meet the design criteria; smaller pressure tube pitch will result in more flatten axial power distribution and more uniform radial power distribution.     

On monitoring the tritium breeder in a lead–lithium cooled ceramic breeder (LLCB) module of the ITER

The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead–lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D–T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method.Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.     

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.     

Thermodynamic analysis of uranium–gadolinium fuel stability at high temperatures

Using the most accurate measurements of the liquidus temperature in the UO₂–Gd₂O₃ system up to 30 mol.% of Gd₂O₃, thermodynamic models of the melt and cubic solution GdO_1.5 in UO₂ are constructed. The equilibrium phase diagram of the system UO₂–GdO_1.5 in the interval 1900–3200 K is calculated in the entire composition range and the metastable diagram is calculated assuming that no cubic solid solutions are formed. The upper and lower boundaries of the melting onset temperature (solidus) of uraniumgadolinium fuel are presented. The phase composition of the pellets made from such fuel and, ultimately, the technology determine the melting onset temperature uniquely.     

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.     

Analysis of criticality in shipment and storage of fuel at a nuclear power plant with a VVÉR reactor

The substantiation of nuclear safety during shipment and storage of fresh and spent fuel at nuclear power plants with VVéR reactors is examined in the light of the more stringent nuclear safety rules. Possible technical measures for satisfying the safety criterion are examined, for example, the concept of subcritical fresh fuel. An example of the estimation of the probability of the formation of a critical mass as result of fuel assemblies falling randomly out of a container is presented. Certain characteristic features of the calculation of the neutron-physical characteristics of fuel in a cooling pond are presented, for example, the nonconservative nature of a separate analysis in the infinite approximation. 4 figures, 5 references. OKB “Gidropress”. Translated from Atomnaya éneriya, Vol. 87, No. 1, pp. 11–16, July, 1999.