Fuel Rod Fabrication Research Articles

Impacts of core parameters on the capability of Cf-252 production in an MSR

Californium-252 (Cf-252) is a powerful neutron source for various applications including neutron radiography, reactor startup, cancer therapy, etc. However, the demand for Cf-252 exceeds its supply at present, and therefore becomes one of the most important restrictions on its applications. Molten salt reactor (MSR) has unique advantages of no fuel rod fabrication, online reprocessing and very deep burnup, which permits flexible core design, substantial target loading into the fuel salt and sufficient depletion time, and is therefore expected to provide an effective solution for Cf-252 production. In this study, four different targets (DU, Pu, Am and Cm) and various mole fractions of the Am target in an MSR are proposed to evaluate the Cf-252 conversion capability. The results indicate that an alluring quality factor of Cf-252 with its mass fraction of over 80% in Cf can be obtained for the selected targets. The highest inventory and conversion rate for Cf-252 with the Cm target can achieve about 700 g and 0.14%, respectively, during the first 10-year operation. The Cf-252 production can be further improved with the increased mole fraction of target but with a deteriorated conversion rate of Cf-252 due to the hardened neutron spectrum and the weakened neutron captures of most isotopes in the reaction chains of Cf-252 formation.

Parametric study on minor actinides transmutation in a graphite‐moderated thorium‐based molten salt reactors

Transmutation of minor actinides (MAs) is an effective solution to reduce the long-term radiotoxicity of nuclear waste. Online refueling and no fuel rod fabrication in thorium-based molten salt reactors (TMSR) are two remarkable advantages of MA transmutation. However, MA solubility limit of the fuel salt and the neutron spectrum in the core are two key parameters that have a strong impact on MA transmutation capability. In this paper, three types of carrier salt with different MA solubility limits and various salt volume fractions (SVFs) are introduced in a TMSR core to evaluate the MA transmutation characteristics. The results indicate that the Flibe salt with the best neutron economy can obtain the highest MA transmutation rate (TR), while the Flinak salt with the largest MA inventory can achieve the highest specific MA transmutation consumption (STC). STC and TR for the case with the Flibe salt and 5% SVF are about 200 kg/GWth.yr and 82%, respectively, which indicates that it can consume the annual MA yields from about 3.37 typical pressurized water reactors (PWRs). Meanwhile, the STC and TR for the case with the Flinak salt and 40% SVF are about 366 kg/GWth.yr and 75%, respectively, which indicates that it can consume the annual MA yields from about 50 typical PWRs. In addition, the total radiotoxicity of MAs after 50-year operation in TMSR can be lowered by as high as about 50%. The significant production of Pu isotopes can be reused as nuclear fuel in fast reactors. In particular, Pu-238, with the mass fraction of more than 60% in Pu isotopes, is a useful material for many nuclear applications. In conclusion, it is feasible to transmute MAs in TMSR and thus achieve the goals of reduction of MA's long-term radioactive hazards.

Validation of the BERKUT fuel rod module against mixed nitride fuel experimental data

The BERKUT module of the EUCLID/V1 integrated computer code has been developed for the performance analysis of fuel rods with nitride or oxide fuels of fast reactors using the quasi two dimensional approach. As far as the uranium-plutonium mixed nitride is considered as one of the basic nuclear fuel materials for the projects of advanced reactor facilities with liquid metal cooled fast reactors developed in Russia, the BERKUT module validation has been carried out against the experimental data observed in the in-pile mixed nitride fuel performance studies. The results of validation obtained taking into account the uncertainties of the fuel rod fabrication data, material properties, and irradiation conditions indicate that the BERKUT module adequately describes the mixed nitride fuel rod behavior in fast reactors.

Comparison of calculated and measured radionuclide inventory of a Zircaloy-4 cladding tube plenum section

ABSTRACTCladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the radionuclide inventory of an irradiated Zircaloy-4 plenum section is analyzed. Quantities of 235/238U, 237Np, 238/239/240/241/242Pu, 241/243Am, 243/244Cm besides 14C, 55Fe, 125Sb, 154Eu, and 134/137Cs were (radio-)chemically determined in digested Zircaloy-4 subsamples. Measured inventories of activation products in the Zr-alloy are in good agreement with calculated values. However, amounts of actinides and fission products exceed the calculated inventory by factor ∼57 (minor actinides and non-volatile fission products) and ∼114 (137Cs). Excess of minor actinides and part of enhanced Cs inventory originate from fuel residues deposited on the inner cladding surface during fuel rod fabrication, whereas vast amount of cesium is volatilized from subjacent fuel pellets and transported to the plenum.

Development and validation of the BERKUT fuel rod module of the EUCLID/V1 integrated computer code

The EUCLID/V1 integrated computer code has been developed for the safety analysis and justification of reactor facilities with fast reactors under normal operating conditions and design basis accidents. The EUCLID/V1 code uses the BERKUT module to describe physical processes occurring in fuel rods with oxide or nitride fuel in fast reactors. The quasi two dimensional approach is employed to simulate the thermomechanical behavior of fuel rods. Verification and validation of the BERKUT module have been carried out using analytical and numerical tests and the experimental data observed in the in-pile fuel performance studies. The results of validation and verification obtained taking into account the uncertainties of the fuel rod fabrication data, material properties and irradiation conditions indicate that the BERKUT module describes adequately the fuel rod behavior in fast reactors.

Development of a method for analysis of thermal performance of VVER fuel

Safe operation and maximum burning of fuel material are two issues that are recently taken into consideration in nuclear fuel rod fabrication industry. Any failure in nuclear fuel and cladding, such as local melting or cracking, may cause release of radioactive fission fragments to the reactor coolant, which is an undesirable issue from the reactor safety point of view. Hence, one of the most important issues in nuclear industry is the preservation of fuel rod integrity during its lifetime. The performance of the irradiated fuel inside the reactor core is affected by various complex phenomena. The main objective of the present paper is to develop valid physical models and an accurate numerical method to study of Water-Water Energetic Reactor (VVER) fuel rod performance. The obtained model can be applied to simulate fuel performance during its lifetime. The correlations used in physical models are chosen in such a way that main parameters, such as gap pressure and fuel centre temperature, can be estimated accurately. The obtained results are in acceptable agreement with available data.

Development of a method for analysis of thermal performance of VVER fuel

Safe operation and maximum burning of fuel material are two issues that are recently taken into consideration in nuclear fuel rod fabrication industry. Any failure in nuclear fuel and cladding, such as local melting or cracking, may cause release of radioactive fission fragments to the reactor coolant, which is an undesirable issue from the reactor safety point of view. Hence, one of the most important issues in nuclear industry is the preservation of fuel rod integrity during its lifetime. The performance of the irradiated fuel inside the reactor core is affected by various complex phenomena. The main objective of the present paper is to develop valid physical models and an accurate numerical method to study of Water-Water Energetic Reactor (VVER) fuel rod performance. The obtained model can be applied to simulate fuel performance during its lifetime. The correlations used in physical models are chosen in such a way that main parameters, such as gap pressure and fuel centre temperature, can be estimated accurately. The obtained results are in acceptable agreement with available data.

Development of End Plug Welding Technique for SFR Fuel Rod Fabrication

In Korea, R&D on a sodium-cooled fast reactor (SFR) was begun in 1997, as one of the national long-term nuclear R&D programs. As one fuel option for a prototype SFR, a metallic fuel, U-Zr alloy fuel, was selected and is currently being developed. For the fabrication of SFR metallic fuel rods, the end plug welding is a crucial process. The sealing of the end plug to the cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions, and parameters were developed for the end plug welding of SFR metallic fuel rods. A gas tungsten arc welding (GTAW) technique was adopted and the welding joint design was developed. In addition, the optimal welding conditions and parameters were established. Based on the establishment of the welding conditions, the GTAW technique was qualified for the end plug welding of SFR metallic fuel rods.

Inert matrix fuel in dispersion type fuel elements

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88MWdkg−1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

Power flattening in the fuel bundle of a CANDU reactor

The strong non-uniformity of the fission power production density in the CANDU fuel bundle could have been mitigated to a great degree. A satisfactory power flattening has been achieved through an appropriately evaluated method by varying the composition of the LWR spent fuel/ThO 2 mixture in a CANDU fuel bundle in radial direction and keeping fuel rod dimensions unchanged. This will help also to greatly simplify fuel rod fabrication and allow a higher degree of quality assurance standardization. Three different bundle fuel charges are investigated: (1) the reference case, uniformly fueled with natural UO 2, (2) a bundle uniformly fueled with LWR spent fuel, and (3) a bundle fueled with variable mixed fuel composition in radial direction leading to a flat power profile (100% LWR spent fuel in the central rod, 80% LWR + 20% ThO 2 in the second row, 60% LWR + 40% ThO 2 in the third row and finally 40% LWR + 60% ThO 2 in the peripheral fourth row). Burn-up grades for these three different bundle types are calculated as ∼7700, ∼27,300, and 10,000 MW.D/MT until reaching a lowest bundle criticality limit of k ∞ = 1.06. The corresponding plant operation periods are 170, 660, and 240 days, respectively.

Strategy of active waste management in the Russian Federation

The paper deals with the main aspects of the strategy for managing currently existing and future nuclear waste in the Russian Federation. There are specific approaches to managing active nuclear waste that was and is being produced from radioactive ore mining and processing and fuel rod fabrication, nuclear power plant operation, spent nuclear fuel reprocessing and defence programme implementation, operation and decommissioning of nuclear powered naval and civil ships and their service bases, nuclear explosions carried out for economic purposes, mining minerals, space programmes and accidents at nuclear units, operation of research nuclear facilities and the use of isotopic products in the economy and medicine. There is a separate review of the issues relevant to spent nuclear fuel management, state of the art for conditioned active waste disposal and promising research trends in the area of active waste management.

Gas chromatography in nuclear fuel rod fabrication

In manufacturing fuel rods, specifications about the gas-content of the sintered pellets and filling gas in the fuel rod have to be met. H 2, A, O 2, N 2, CH 4 and CO are determined in extraction gases of pellets, H 2 being the major component; He, H 2, A, O 2, N 2, CH 4 and CO are determined in the rod filling gas, He being the major component. The principles of gas chromatography are explained, applied to the rest-gas determination in sintered pellets and fuel rod filling gas and some experimental data are shown. Special attention is paid to both extraction techniques which had to be designed for glove box work. A minimum sample volume of 950 μ 1 is required for rest-gas analysis, and a minimum of 6,5 cm 3 is required for rod filling gas. The detection limit for each gas component is better than 0.06 v/o at min, sample volume. The accuracy for the major components is better than 0.5 v/o absolute and for the minor components better than 3% relative.