Radiation damage and simulated fission product effects on the properties of inert matrix materials

Candidate inert matrix materials for actinide transmutation (MgAl₂O₄, CeO₂) or immobilization (ZrSiO₄) containing ²⁴¹Am were characterized. The currently most considered material, ZrO₂, was produced, with La₂O₃ as stand-in for Am, and with and without simulated fission products to investigate burnup effects. The oxygen potential was measured using an EMF cell. The accumulation of radiation damage due to Am decay was investigated by periodically measuring lattice parameters and hardness. Sequential leaching tests in deionized water, aimed at correlating the leaching behaviour of Am and of the matrix with radiation damage, showed significant release of Am and of some matrix components.     

Dissolution and analysis of erbium doped cermet fuels

Cermets are suggested as new kind of nuclear fuel to reduce global costs. They need high enriched fuel and thus use of burnable poison. Special pellets were developed and irradiated to test such concepts. Some pellets consist of a cermet fuel. With an improved fuel thermal conductivity (by using metal matrix), lower temperatures than standard fuel are obtained. Some pellets were made of cermet and erbium in small quantity. Studies on erbium were launched to determine the influence of this neutron poison. Standard dissolutions (HNO₃, HF) on cermet (Mo-UO₂) and on erbium doped cermet show a large amount of insoluble matter. Tests have been carried out in order to establish a procedure for a complete dissolution of active pellets. Consequently, an optimal process was defined. Irradiated pellets from experimental reactor SILOE will be dissolved. Analytical chemistry studies were undertaken. Thermal Ionization Mass Spectrometry (TIMS) and Glow Discharge Mass Spectrometry (GDMS) have been applied. The U and Er isotopic composition has been determined on different samples.     

Dissolution behavior of (U,Zr)O2-based simulated fuel debris in nitric acid

To explore the possibility of dissolving fuel debris into nitric acid as a potential pre-treatment for waste treatment in which the U and Pu are removed from the inventory, dissolution tests of U_1?xZr_xO₂ and (U,Pu)_1?xZr_xO₂ were carried out in 6 M HNO₃ at 353 K. At the end of the dissolution test (after 4 h), the ratio of dissolved uranium decreased with an increase in the Zr contents, x. While the dissolution of U-rich samples was congruent, a preferential leaching of U was observed with Zr-rich samples. Taking into account these different dissolution phenomena, the dissolution rate analysis was carried out using surface-area model to calculate the instantaneous dissolution rate (IDR). The IDR decreased from 10^?5 down to 10^?10 mol cm^?2 min^?1 as x increased from 0 to 0.95. From these findings, dissolution with HNO₃ is expected to be only applicable in U-rich part of fuel debris (x < 0.3) if the dissolution in 6 M HNO₃ at 353 K is assumed. Application of complexing acids, such as mixture of HNO₃ and HF, should be considered to increase the dissolution rate of the Zr-rich part.     

先进核燃料循环体系研究进展

概述了先进核燃料循环体系的概念 ,论述了目前后处理与分离 嬗变领域的研究进展和技术发展趋势     

Studies on LLFP transmutation in a pressurized water reactor

A systematic study on the long-lived fission product (LLFP) transmutation in a pressurized water reactor (PWR) is performed, aiming at an optimal transmutation strategy for present nuclear energy development. The LLFPs selected in the analysis include ⁹⁹Tc and ¹²⁹I discharged from light water reactors (LWRs). The isotope ¹²⁷I is also considered to avoid the difficulties in isotopes separation. To minimize the negative impacts of LLFPs on the core performance and safety parameters, metallic technetium or MgI₂ target pins mixed with ZrH₂ are designed and investigated. Through the numerical analysis on equilibrium cycles, the transmuted amounts of ⁹⁹Tc and ¹²⁹I equal to the yields from 1.94 and 4.22 PWRs with a power of 1000 MWe, respectively. Numerical results indicate that both ⁹⁹Tc and ¹²⁹I can be transmuted conveniently in present PWRs in the form of target pins.     

Heterogeneous fuels for minor actinides transmutation: Fuel performance codes predictions in the EFIT case study

Plutonium recycling in new-generation fast reactors coupled with minor actinides (MA) transmutation in dedicated nuclear systems could achieve a decrease of nuclear waste long-term radiotoxicity by two orders of magnitude in comparison with current once-through strategy.     

加速器驱动次临界系统（ADS）与核能可持续发展

描述了ADS系统的主要技术特点和在我国核能可持续发展战略中的作用及地位;介绍了国内外ADS研究的状态和发展趋势;提出了ADS研发必须解决的关键技术问题及解决这些问题的时间表;分析了ADS研发与国内核能相关发展计划的关系;并就我国开展ADS的研发提出了一些建议。     

Elements of comparison between different inert matrix fuels as regards plutonium utilisation and safety coefficients

The current international trend is to focus towards the utilization of plutonium. The use of composite fuels in inert matrix (U-free) is a potentially efficient solution to this problem. This document deals with the cermet fuels, selected for their excellent behaviour under irradiation and their high thermal conductivity. The emphasis is placed on the study of kinetic coefficients. Comparisons are performed with other solutions that use other composite fuels, especially the Solid Solutions and ROXs. As core control requires a heterogeneous assembly, an assembly whose characteristics are compared to the APA reference is proposed.     

加速器驱动先进核能系统的乏燃料循环再生研究北大核心CSCD

乏燃料后处理是核燃料循环的关键环节,制约核电的可持续发展。借助于加速器驱动先进核能系统(ADANES)提供的高通量、硬能谱的外源中子,其乏燃料后处理只需除去乏燃料中的挥发性裂变产物和影响次锕系元素嬗变的中子毒物,长寿命的次锕系元素Np、Am、Cm可与二氧化铀一起转化为新的燃料元件在加速器驱动燃烧器中燃烧、嬗变、增殖和产能。基于此,本课题组提出了加速器驱动的乏燃料后处理及再生制备的技术路线,包括高温氧化粉化与挥发、选择性溶解分离和燃料再生制备。本文主要介绍了近几年本课题组在这三方面所取得的一些成就,希望能为加速器驱动先进核能系统的乏燃料后处理提供基础数据。     

Status of the French Research Program on Partitioning and Transmutation

The production of nuclear energy in France has been associated, since its inception, with the optimization of radioactive waste management, including the partitioning and the recycling of recoverable energetic materials. From the start, the Commissariat à l'Energie Atomique (CEA) has devoted considerable effort to the management of the back end of the cycle in order to extract the re-usable materials, uranium and plutonium, and to condition the resulting waste with the vitrification process.

At the end of the December 1991 Waste Management Act, it is considered that partitioning techniques, which have been validated on real solutions with aqueous process, have been brought to a point where there is reasonable assurance that industrial deployment can be successful. For transmutation, CEA has conducted programs proving at lab scale the feasibility of the elimination of minor actinides and scenario studies have also allowed assessing the efficiency of transmutation in terms of the quantitative reduction of the final waste inventory depending on the reactor fleet of Pressurized Water Reactor-Fast Neutron Reactor-Accelerator Driven Systems (PWR-FR-ADS).

For the future, the new Waste Management Act passed by the French Parliament on June 28, 2006, demands that Partitioning and Transmutation research continues in strong connection to GEN IV systems and ADS development, allowing to assess the industrial perspectives of such systems in 2012 and to put into operation a transmutation demo facility in 2020.     

Initial dissolution rate of a Japanese simulated high-level waste glass P0798 as a function of pH and temperature measured by using micro-channel flow-through test method

Aqueous dissolution tests were performed for a Japanese type of simulated high-level waste (HLW) glass P0798 by using a newly developed test method of micro-channel flow-through (MCFT) method, and the initial dissolution rate of glass matrix, r ₀, was measured as a function of solution pH (3–11) and temperature (25–90°C) precisely and consistently for systematic evaluation of the dissolution kinetics. The MCFT method using a micro-channel reactor with a coupon shaped glass specimen has the following features to provide precise and consistent data on the glass dissolution rate: (1) any controlled constant solution condition can be provided over the test duration; (2) the glass surface area actually reacting with solution can be determined accurately; and (3) direct and totally quantitative analyses of the reacted glass surface can be performed for confirming consistency of the test results. The present test results indicated that the r ₀ shows a “V-shaped” pH dependence with a minimum at around pH 6 at 25°C, but it changes to a “U-shaped” one with a flat bottom at neutral pH at elevated temperatures of up to 90°C. The present results also indicated that the r ₀ increases with temperature according to an Arrhenius law at any pH, and the apparent activation energy evaluated from Arrhenius relation increases with pH from 54 kJ/mol at pH 3 to 76 kJ/mol at pH 10, which suggests that the dissolution mechanism changes depending on pH.     

Recent activities in the field of nuclear materials and nuclear fuels

On the research and development of nuclear materials and fuels, many of outstanding papers have been presented in scientific journals, including the Journal of Nuclear Science and Technology. Some topics have been covered in this summary from the latest activities of nuclear materials and fuels.     

Dynamics aspects of plutonium burning in an inert matrix

Burnup calculations have been performed on a mini fuel assembly containing 21 fuel rods and four water holes at the corners. The fuel rod positions were filled with 4% enriched UO₂ fuel and with either reactor grade or weapons grade plutonium mixed in an inert matrix. The ratio between the UO₂ and the IMF rods was varied to investigate the influence of the UO₂ fuel on the dynamics of the assembly. From a simple reactor model with one delayed neutron group and first-order fuel and temperature feedback mechanisms, the linear transfer function from reactivity to reactor power was calculated that was subsequently used in a root-locus analysis. From this, it is concluded that only 20% of the fuel rods need to be made of UO₂ to have a fuel that is linearly stable up to 1000 days of irradiation.     

美国先进核燃料循环计划概述及对我国的启发

美国已重启核能计划,其中先进的核燃料循环计划是核能计划的核心。若这一计划得以顺利实施,将可以消除人们曾担忧的核能开发中的三大问题,核扩散、高放废物的处置和铀资源的可持续性问题。美国这一计划对我国的核能发展应有一些启发。     

Utilization Factor of Nuclear Material in LMFBR

The effects of γ-irradiation on a simulated nuclear waste glass were studied by electron spin resonance spectroscopy (ESR), and were compared with the results on silica glass and Pyrex glass. Three kinds of glasses were γ-irradiated up to the dose of 1.22 MGy and the ESR spectra were obtained. The intensity of ESR spectra were obtained as a function of irradiation dose and annealing temperature.

The spectrum of the waste glass was characteristic of two typical peaks, Peak 1 was the strong resonance at g=4.3 showing the existence of four coordinated Fe³⁺ and Peak 2 was the weak and broad resonance at g= 2.0 showing the existence of six coordinated Fe³⁺. The ESR spectra of the waste glass before and after γ-irradiation were almost overlapped and a little difference only in the intensity was observed. While in silaca glass and Pyrex glass, the peaks from E'γ center and boron-oxygen hole center (BOHC) were observed to arise after irradiation. The absolute intensity of. Peaks 1 and 2 described above changed in complicated way depending on the dose. The result suggests oxidation or reduction of iron takes place in the waste glass depending on the dose. The isochronal annealing of irradiated glasses shows most of γ-ray- induced damages in the waste glass are restored even at room temperature, although most of the damages in silica glass and Pyrex glass are disappeared at the temperature from 550 to 600 K. The results show that the waste glass with a few weight percent of iron is resistent to radiation than other commercial glasses.     

Neutronics assessment of the use of thorium fuels in current pressurized water reactors

The use of thorium fuel in current PWRs in a once-through fuel cycle is an attractive option due to potential advantages such as high conversion ratio and low minor actinide generation. The current neutronics assessments indicate that the thorium fuel cycle could supplement the current uranium–plutonium fuel cycle to improve operational performance and spent fuel consideration in current PWRs without core and subassembly modifications. Neutronics safety parameters in the PWR cores with the thorium fuels are within the range of current PWRs.The PWR cores with thorium fuels have significantly higher conversion ratios which could enable efficient fuel utilization. Further, it is shown that the use of thorium as a fertile material can reduce minor actinide generation and the radio-toxicity of spent fuels. In considerations related to proliferation resistance, the results of the current analyses show no significant difference between the studied thorium fuels and the standard oxide fuel for the assumed characteristics and burnup levels.     

A study on the burn-up characteristics of inert matrix fuels

Proper disposal of minor actinides (MA), long-lived fission products (LLFPs), and transuranium element (TRU) plays a key role in the sustainable development of fission nuclear power. Adoption of inert matrix fuels (IMFs) can effectively reduce the amount of ²³⁷Np and Np element in the spent fuel of present-day commercial power reactors. In order to study the burn-up characteristics of IMFs caused by the unique composition, burn-up calculations and MA accumulation of two typical IMFs, PuO₂ + ZrO₂ + MgO and PuO₂ + ThO₂, are performed in this paper. Results indicate that k_inf at beginning of life (BOL) and reactivity drop with burn-up for PuO₂ + ZrO₂ + MgO are much larger than those of PuO₂ + ThO₂ IMF. The yields of ²³⁷Np and Np element in IMFs are two orders smaller than those of UO₂ and mixed oxide (MOX) fuels. For the same PuO₂ volume fraction and a certain burn-up, the masses of ²³⁷Np, Np element, and ²⁴¹Am for PuO₂ + ZrO₂ + MgO are smaller than those of PuO₂ + ThO₂; however, the mass of total MA is larger. IMF has high destruction efficiencies of TRU and plutonium (Pu). The results and conclusion provide basic data for the ongoing IMF design and application study.     

星座型裂变燃料核反应堆的物理构想

从分析钍在持续中子辐照过程中各代子体含量的演变出发,着重研究有多代子体均达到各自的饱和值时的情况和所具有的特性,提出星座型裂变物质核反应堆的物理构想,并就此堆的特性和应用前景作了简单阐述和讨论。     

First-principles study on dissolution and diffusion properties of hydrogen in molybdenum

Employing a first-principles method, we have investigated dissolution and diffusion properties of hydrogen (H) in molybdenum (Mo), one of potential candidates for plasma facing materials in a nuclear fusion Tokamak. We show that single H atom is energetically favorable sitting at the tetrahedral interstitial site (TIS) instead of octahedral interstitial site and diagonal interstitial site. This can be confirmed by the electron localization function result. Bader charge analysis suggests that the bonding between H and surrounding Mo is mainly ionic mixed with slight covalent component. Double H atoms tend to be paired up at the two neighboring TIS’s along the 〈1 1 0〉 direction with the distance of ∼0.221 nm and the binding energy of 0.03 eV. This suggests a weak attractive interaction between H atoms, with the implication that self-trapping of H and thus formation of the H₂ molecules are quite difficult in an intrinsic Mo environment. We demonstrate that the diffusion barrier of H that jumps between the TIS’s is 0.16 eV, and the dissolved concentration of H in the intrinsic Mo is 2.6 × 10⁻⁸ at a typical temperature of 600 K. The diffusion coefficients of H, D, and T are different due to the different masses, which are calculated to be 1.3 × 10⁻⁷ m²/s, 9.2 × 10⁻⁸ m²/s, and 7.5 × 10⁻⁸ m²/s at 600 K.     

Effect of clayey groundwater on the dissolution rate of the simulated nuclear waste glass SON68

The influence of clayey groundwater on the dissolution rate of SON68 glass was investigated under rate drop conditions. Leaching in contact with groundwater resulted in larger amounts of altered glass than obtained with initially pure water. Clayey groundwater delays the rate drop and subsequently results in dissolution rates higher than in pure water. This effect is due to the presence of magnesium, which precipitates in secondary phases with silicon while the other elements found in clayey water have no effect on the glass kinetics. Modeling the test results showed that the precipitation of secondary magnesium phases sustains the dissolution of the passivating layer of the glass and lowers the leachate pH, thereby increasing the diffusivity of this layer. The simulations also showed that the precipitation of magnesium phases is not limited by their precipitation kinetics, but is controlled by their solubility and the flow of silicon from glass dissolution. When all the magnesium in solution has precipitated, the pH slowly returns to the values usually measured during leaching of glass in initially pure water. Then the dissolution rate reflects the values measured in pure water. This study demonstrates that the reactivity of magnesium phases in the geological environment and the transport of magnesium in solution could have a significant impact on the long-term behavior of the glass.