Effects of α-radiation on a direct disposal system for spent nuclear fuel – (2) review of research into safety assessments of direct disposal of spent nuclear fuel in Europe and North America

The Japanese geological disposal programme has started researching disposal of spent nuclear fuel (SF) in deep geological strata (hereafter “direct disposal of SF”) as an alternative management option other than reprocessing followed by vitrification and deep geological disposal of high-level radioactive waste (HLW). In the case of direct disposal of SF, the radioactivity of the waste is higher and the potential effects of the radiation are greater. Specific examples of the possible effects of radiation include: increased amounts of canister corrosion; generation of oxidizing chemical species in conjunction with radiation degradation of groundwater and accompanying oxidation of reducing groundwater; and increase in the dissolution rate and the solubility of SF. Therefore, the influences of radiation, which are not expected to be significant in the case of geological disposal of vitrified waste, must be considered in safety assessments for direct disposal of SF. Focusing especially on the effects of α-radiation in safety assessment, this study has reviewed safety assessments in countries other than Japan that are planning direct disposal of SF. The review has identified issues relevant to safety assessment for the direct disposal of SF in Japan.     

Utilization of spent PWR fuel-advanced nuclear fuel cycle of PWR／CANDU synergism

High neutron economy, on line refueling and channel design result in the unsurpassed fuel cycle flexi-bility and variety for CANDU reactors. According to the Chinese national conditions that China has both PWR and CANDU reactors and the closed cycle policy of reprocessing the spent PWR fuel is adopted, one of the advanced nu-clear fuel cycles of PWR/CANDU synergism using the reprocessed uranium of spent PWR fuel in CANDU reactor is proposed, which will save the uranium resource (-22.5%), increase the energy output (-41%), decrease the quantity of spent fuels to be disposed (-2/3) and lower the cost of nuclear poower, Because of the inherent flexibility of nuclearfuel cycle in CANDU reactor, and the low radiation level of recycled uranium(RU), which is acceptable for CANDU reactor fuel fabrication, the transition from the natural uranium to the RU can be completed without major modifica-tion of the reactor core structure and operation mode.It can be implemented in Qinshan Phase Ⅲ CANDU reactors with little or no requirement of big investment in new design. It can be expected that the reuse of recycled uranium of spent PWR fuel in CANDU reactor is a feasible and desirable strategy in China.     

The action of α-radiation on aqueous acid solutions

We investigated the action of polonium -radiation on a 0.8 N aqueous solution of sulfuric acid.It was established that over a wide range of dissolved polonium concentrations (0.1–12 mC/ml), the initial radiation-chemical hydrogen peroxide yield remains the same and equals 1.20 molecules per 100 ev of absorbed energy and that the hydrogen peroxide concentration tends toward a limit. The equilibrium limit of hydrogen peroxide concentration in 0.8 N sulfuric acid is 5–8·10¹⁸ molecules/ml.     

Influence of α-clustering nuclear structure on the rotating collision system

In recent years, the collective motion properties of global rotation of the symmetric colliding system in relativistic energies have been investigated. In addition, the initial geometrical shape effects on the collective flows have been explored using a hydrodynamical model, a transport model, etc. In this work, we study the asymmetric ~(12)C+~(197)Au collision at 200 GeV/c and the effect of the exotic nuclear structure on the global rotation using a multi-phase transport model. The global angular momentum and averaged angular speed were calculated and discussed for the collision system at different evolution stages.     

Challenges in spent nuclear fuel final disposal： conceptual design models

The disposal of spent nuclear fuel is a long-standing issue in nuclear technology. Mainly, UO2 and metallic U are used as a fuel in nuclear reactors. Spent nuclear fuel contains fission products and transuranium elements, which would remain radioactive for 104 to 108 years. In this brief communication, essential concepts and engineering elements related to high-level nuclear waste disposal are described. Conceptual design models are described and discussed considering the long-time scale activity of spent nuclear fuel or high level waste. Notions of physical and chemical barriers to contain nuclear waste are highlightened. Concerns regarding integrity, self-irradiation induced decomposition and thermal effects of decay heat on the spent nuclear fuel are also discussed. The question of retrievability of spent nuclear fuel after disposal is considered.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Industrial processing of the spent fuel of npp operating with water-moderated VVÉR-440 reactors

Mayak Production Complex. Sverdlovsk Scientific-Research Institute of Chemical Machinery. Research-cum-production Complex VNIPIET. Ministry of Atomic Energy, Russian Federation. A. A. Bochvar All-Union Scientific-Research Institute of Inorganic Materials. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 432–436, May, 1992.     

Three-dimensional FE analysis of the thermal–mechanical behaviors in the nuclear fuel rods

In order to implement numerical simulation of the thermal–mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal–mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal–mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm³, the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm² K to 0.01 W/mm² K, while increases up to 54.7% when h decreases from 0.01 W/mm² K to 0.005 W/mm² K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal–mechanical behaviors in the fuel rod; when the gap size varies from 0.03 mm to 0.1 mm, the highest temperature in the pellet increases by 19.7%, and the maximum first principal stress at the outer pellet surface decreases by 17.4%; it is critical to optimize the gap size in order to reduce the pellet–cladding mechanical interaction and avoid their contact at early stage. This study lays a foundation for the further research on the irradiation-induced mechanical behaviors in the fuel rods.     

Can redox sensitive radionuclides be immobilized on the surface of spent nuclear fuel? - A model study on the reduction of Se(IV)aq on Pd-doped UO2 under H2 atmosphere

Spent nuclear fuel contains noble metal particles composed of fission products (Pd, Mo, Ru, Tc, Rh and Te, often referred to as ε-particles). Studies have shown that these particles play a major role in catalyzing oxidative dissolution as well as H₂ reduction of the oxidized UO₂ fuel matrix, depending on the conditions. Thus it is possible that these particles also could have a major impact on the state of other redox sensitive radionuclides (such as the long lived fission product ⁷⁹Se) present in spent nuclear fuel. In this study, Pd-doped UO₂ pellets are used to simulate noble metal particles inclusions in spent nuclear fuel and the effect on dissolved selenium in the form of selenite (250 μM selenite) in simulated ground water solution (10 mM NaCl, 10 mM NaHCO₃) at 1 and 10 bar hydrogen pressure. The selenite was found to be reduced to elemental Se, forming colloidal particles. At hydrogen pressures of 10 bar, the rate of selenite reduction was found to be linearly correlated to the fraction of Pd in the UO₂ pellets. No selenium was detected on the surface of the pellets. For the lowest Pd loading (0.1% Pd) the selenite reduction does not appear to proceed to completion indicating that the surface becomes less active.     

Monitoring ofγ-radiation on flat landscapes in the rapid-search mode for above-background spots

Institute of General and Nuclear Physics, Russian Scientific Center "Kurchatovskii institut." Translated from Atomnaya énergiya, Vol. 76, No. 2, pp. 153–155, February, 1994.     

Quantification and local distribution of hydrogen within Zircaloy-4 PWR nuclear fuel cladding tubes at the nuclear microprobe of the Pierre Süe Laboratory from μ-ERDA

Hydrogen content and its distribution in in-core materials of nuclear plants are known to have a strong influence on their behaviour, especially on their mechanical properties but also on their corrosion resistance. This point has to be largely investigated in the case of the nuclear fuel cladding (Zr based alloys) of pressurized water reactors (PWR).Two situations have been considered here, with regards to the hydrogen content and its spatial distribution within the thickness of the tubes:

(1)

Irradiated fuel cladding tubes after a nominal period under working conditions in a PWR core.

(2)

Non-irradiated fuel cladding previously exposed to conditions representative of an hypothetical “loss of coolant accident” scenario (LOCA).

As far as micrometric distributions of H were required, μ-ERDA has been performed at the nuclear microprobe of the Pierre Süe Laboratory. This facility is fitted with two beam lines. In the first one, used for non-active sample analysis, the μ-ERDA configuration has been improved to reduce the limits of detection and the reliability of the results. The second one offers the unique feature of being dedicated to radioactive samples. We will present the nuclear microprobe and emphasize on the μ-ERDA configuration of the two beam lines. We will illustrate the performance of the setup by describing the results obtained for Zircaloy-4 cladding both on non-irradiated and irradiated samples.     

The role of Fe on the solubility of Nb in α-Zr

The solid solubility of Nb in α-Zr is an important parameter that has a potential impact on the corrosion properties of Zr-Nb alloys at reactor operating temperatures, i.e. below the monotectoid temperature. Work on dilute Zr-Nb alloys has shown that Fe is a common impurity that confounds the assessment of the solid solubility limit for Nb in Zr. This is because Fe has a very low solubility limit and it forms precipitates with both Nb and Zr. To assess the effect of Fe on the phases formed in the binary Zr-Nb alloy system, alloys containing 0.1-0.7 wt% Nb and <11 to 470 wt ppm Fe were heat-treated at temperatures between 575 °C and 600 °C and examined by transmission electron microscopy. Results indicate that, even at a concentration ? 24 ppm, Fe readily combines with Nb to form precipitates in the alloys with Nb contents in the range of 0.20 to 0.29 wt%. However, β-Nb particles were not observed for these same alloys and were only seen when the Nb content was ? 0.49 wt%. Because β-Nb particles were not found in the 0.29 wt% Nb alloy and the precipitation was estimated to have a negligible effect on the amount of Nb remaining in solution (reduced by <0.001 wt%), it is proposed that the solubility limit of Nb in a true binary Zr-Nb alloy would be between 0.29 and 0.49 wt%.     

Assembly?oriented reliability analysis method for the top?connection structure of a nuclear fuel assembly

The nuclear fuel assembly is the core component of a nuclear reactor. In a pressurized water reactor fuel assembly, the topconnection structure connects the top nozzle to the guide thimble. Its performance reliability is essential for the stability of the nuclear fuel assembly. In this study, an assembly-oriented reliability analysis method for top-connection structures is presented by establishing an assembly-oriented top-connection structure parameter modeling method and a nonlinear contact ga...     

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.