Transport of MOX fuel: continuous challenge

Abstract

For 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the spent nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. This paper focuses on the casks used to transport the fresh and used mix oxide (MOX) fuel. To transport the fresh MOX boiling water reactor and pressurised water reactors fuel, TN International has developed two designs of casks: the MX 6 and the MX 8. These casks are and have been used to transport MOX fuel for French, German, Swiss and in a near future Japanese nuclear power plants. A complete set of baskets have been developed to optimise the loading in terms of integrated dose and also of course capacity. Mixed oxide used fuel has now its dedicated cask: the TN 112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in September 2008 in the Electricité de France nuclear power plant of Saint-Laurent-des-Eaux. By its continuous involvement in the nuclear transportation field, TN International has been able to face the many challenges linked to the radioactive materials transportation especially talking of MOX fuel. TN International will also have to face the increasing demand linked to the nuclear renaissance.     

Packaging Design and Qualification: The Experience of the CDTN/CNEN

Abstract

Since 1982 the CDTN, the Nuclear Technology Development Centre, has been designing, testing and qualifying packaging for radioactive materials. These packagings are used for the transport of radioisotopes and disposal of spent sealed sources, wastes generated in the nuclear fuel cycle and the wastes produced in the radiological accident that occurred in the city of Goiânia. For radioactive tracers and medical/industrial radioisotopes, the packagings used are cardboard and wood boxes, while the spent sealed sources are preferably conditioned in metal drums containing lead shielding and a gas absorber material. To condition and transport the wastes from the various nuclear cycle activities, metal drums and boxes are used in Brazil. For the higher active wastes from the nuclear power plant Angra I, a metallic drum in a concrete overpack is used. The wastes generated in the accident were first conditioned in the readily available packaging. Later on, more appropriate packaging was designed by the CDTN staff. CDTN has carried out a programme since 1983 to evaluate the durability of commercial drums used for waste conditioning.     

TRANSPORT OF MOX FUEL

Abstract

The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations.     

Another option for separated plutonium management: Storage MOX

Currently there is no plan for the management of civilian plutonium that does not create a stockpile of separated plutonium. As a result, a number of nations with nuclear technology hold a large stockpile (about 240 tonnes) of separated plutonium. This paper suggests a timely, cost-effective solution for managing this material: storage MOX. A storage MOX plan would use existing MOX fuel fabrication facilities to make a simple MOX waste form suitable for long-term storage. Alternative waste forms to MOX are also possible, such as zirconia and pyrochlore, which provide more reliable durability and radiation damage control over thousands to hundreds of thousands of years.     

Oxygen chemical diffusion in hypo-stoichiometric MOX

Kinetics of the oxygen-to-metal ratio change in (U_0.8Pu_0.2)O_2−x and (U_0.7Pu_0.3)O_2−x was evaluated in the temperature range of 1523-1623 K using a thermo-gravimetric technique. The oxygen chemical diffusion coefficients were decided as a function of temperature from the kinetics of the reduction process under a hypo-stoichiometric composition. The diffusion coefficient of (U_0.7Pu_0.3)O_2−x was smaller than that of (U_0.8Pu_0.2)O_2−x. No strong dependence was observed for the diffusion coefficient on the O/M variation of samples.     

VALMOX: validation of nuclear data for high burn-up MOX fuels

VALMOX, an acronym for validation of nuclear data for high burn-up MOX fuels, is one of the projects of the cluster evolutionary fuel concepts: high burn-up and MOX fuels (EVOL). It covers 30 months, from October 2001 to March 2004.It considers the evaluation of the actinide inventory of MOX fuel at high burn-up (typically 60 GWd/t) in light water reactors, with special attention to the helium production. Calculated values for the spent fuel isotopic masses are compared to the measured ones, with sensitivity analyses made in support. The JEF 2.2 nuclear data file is taken as a basis for calculation. The resulting recommendations on nuclear data should be employed in the preparation and testing of the next JEFF3 file.So far, the major effort was placed on the evaluation of MOX fuel irradiations in pressurised water reactors, and first results will be presented and compared.     

Effective thermal conductivity of MOX raw powder

The radial temperature distribution of plutonium and uranium mixed oxide powder loaded into a cylindrical vessel was measured in air and argon gas, and the effective thermal conductivity was calculated from the measured temperature distribution and the decay heat. The effective thermal conductivities were small values of 0.061-0.13 W m^-1 K^-1 at about 318 K, and changed significantly with O/M, bulk density and atmospheric gas. The results in this work were analyzed by the model of Hamilton and Crosser and a new model for the effective thermal conductivity of the powder was derived as functions of powder properties and thermal conductivity of atmospheric gas.     

CeO2代替PuO2模拟制造MOX燃料芯块的可行性

MOX燃料在国外已成功应用于轻水堆和快中子增殖堆中.Pu元素的放射性和毒性给MOX燃料芯块制造技术带来很大不便.本文通过对U-Ce-O系统、U-Pu-O系统的热力学性质、相图等方面的比较和对制造工艺的模拟方式进行的初步探讨表明,在MOX燃料的生产工艺试验中,U-Ce-O系统可用来模拟U-Pu-O系统.     

MOX燃料混料过程的优化

本文采用House holder变换法对MOX燃料混料过程中Pu同位素均一化问题进行优化计算,并用轨迹求解法对球磨中的转速问题进行了初步探讨。     

Mixed-oxide (MOX) fuel performance benchmarks

Within the framework of the OECD/NEA Expert Group on Reactor-based Plutonium disposition (TFRPD), fuel modeling code benchmarks for MOX fuel were initiated. This paper summarizes the calculation results provided by the contributors for the first two fuel performance benchmark problems. A limited sensitivity study of the effect of the rod power uncertainty on code predictions of fuel centerline temperature and fuel pin pressure also was performed and is included in the paper.     

MOX燃料堆芯热工特性及设计限值研究

使用MOX燃料的快堆核电站以其线功率高、燃耗高、堆芯出口温度高等特点,对堆芯热工设计提出了新的问题.本文在对MOX燃料热工性能分析的基础上,给出了主要的热工设计限值,并以电功率870 MW电站为参考,初步分析了其堆芯热工特性和设计裕量.结果表明对于MOX燃料,较高的堆芯热工参数合理可行,且具有足够的裕量.     

Low Current Efficiency in MOX Deposition Tests

A series of MOX deposition tests has been performed since 2001 at RIAR to clarify its complex phenomena and to improve its poor current efficiency. In the 2001 tests, the cathode current efficiency was between 60 and 100% but the Pu fraction in the MOX was between 5 and 20%. In 2002 tests, the fraction was raised to more than 30% by modifying the test conditions but the current efficiency fell to between 20 and 60%. A new method was proposed to simulate the parasitic current due to the electrode reactions of UO₂ ²+/UO₂+, Pu⁴+/Pu³+ and Fe³+/Fe²+ at the cathode. It was found that the parasitic current due to the UO₂ ²+/UO₂+ reaction significantly lowers the current efficiency especially when the cathode potential is kept near the equilibrium value during the electrolysis to increase the Pu fraction in the MOX deposit.     

MOX Co-deposition Tests at RIAR for SF Reprocessing Optimization

The Oxide Electrowinning method has been studied as one of the candidate dry reprocessing concepts of the future fuel cycle systems. On the MOX co-deposition process, the main process of that method, some fundamental experiments have been performed to confirm its feasibility. In the experiments, several parameters were set to study the suitable electrolysis condition to obtain desired granule of MOX. The concentrations of uranium, plutonium, fission products(FP) simulators, and corrosion products(CP) simulators were adopted as the parameters. The blowing gas composition (O₂, Cl₂, Ar) during the electrolysis was also set as the variable condition. Through these experiments, it was clarified that the partial pressure of chlorine gas during electrolysis was important to obtain MOX granule with high Pu concentration (about 30%) without generating bottom precipitation in melt. Finally, adequacy of the process control method for MOX co-electrolysis was confirmed through the test using spent fast reactor(FR) fuel.     

ADAGIO technique: From UO2 fuels to MOX fuels

The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning ‘Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas’). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of ⁸⁵Kr release kinetics which was developed for UO₂ was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO₂ matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.     

Coupled neutronics-fluids modelling of criticality within a MOX powder system

This paper is concerned with the reactivity during a transient of a layered MOX, and moderating lubricant (zinc stearate) mixing system which is used as part of fuel processing. The reactivity effects with increases in temperature and mixing (moderation) are investigated here. The transient and static simulations reveal that large increases in reactivity and temperature are possible. The nuclear criticality modelling of powders is performed using the Finite Element Transient Criticality (FETCH) code. This models criticality transients in spatial and temporal detail using a continuum multi-fluid model of the gas and solid-powder phases. The neutronics model in FETCH solves the neutron transport in full phase space. In this study a number of hypothetical criticality scenarios with the same generic mixing device and MOX powder are investigated in order to gain an understanding of the dynamics of powder criticality. An additive zinc stearate is also introduced. The zinc stearate, can act as a moderator and thus, on mixing, a ramp reactivity can be introduced. The mixing device is introduced into the simulations to help investigate this ramp reactivity.     

Towards validation of criticality calculations for systems with MOX powders

A series of experiments referred to as BFS/MOX was conducted in the BFS-1 experimental facility at IPPE, Russia. The program was designed to provide a basis for validation of criticality calculations for MOX fuel manufacturing processes and particularly those with low-moderated MOX fissile material. An extensive experimental program was performed, including criticality and reactor-type parameter measurements. The experiments were evaluated, peer reviewed, and analyzed with various codes and cross section data. The criticality validation study was performed employing a sensitivity/uncertainty technique based on the generalized linear least squares method. This paper briefly describes the experimental program, shows different tools’ performance when calculating criticality for the BFS/MOX configurations, and focuses upon the validation study and results for generic applications with weapons-grade plutonium.     

混合氧化物燃料应用前景及核燃料循环战略

本文叙述了不同国家在不同核电发展条件下的核燃料循环战略初步考虑,并着重叙述了在快堆商业化进程推迟后所发展的混合氧化物(MOX)燃料的设计、制造技术及定量经济分析的基本情况,并建议我国近期应主要采用中间贮存并配以建造后处理中间试验工厂,为今后的商业后处理厂做好技术准备,同时还要跟踪 MOX 燃料技术。     

模拟MOX燃料粉末混合均匀性研究

进行了模拟MOX (UO2-10%CeO2)燃料粉末的球磨混合和烧结实验,讨论了行星式高能球磨工艺参数对模拟MOX燃料粉末的混合均匀性(或变异系数CV)和模拟MOX烧结芯块中Ce分布均匀性的影响,以及可能存在的粉末球磨混合机理.采用优化的粉末球磨混合工艺参数,可使模拟MOX粉末的混合均匀度达到98%以上,主要的混合机理是扩散.电子探针(EPMA)分析证明,烧结芯块中Ce元素也达到了微观均匀分布.     

Feasibility Study on the Analysis of MOX Particle Behavior

A plenty of plutonium is dealt in Plutonium Fuel Fabrication Facility and the facility is required to confine plutonium within a limited space such as glove box (GB) because plutonium is a-emitter and causes an internal exposure. The MOX particles entrainment occurs and some of them are transiting to the outlet of GB without deposition to floor and wall. The entraining rate and the transiting rate are reported as Airborne Release Fraction (ARF) and Respirable Fraction (RF) in the literatures. However, no activities of model development and analytical approach have been found for ARF and RF. Thus, a feasibility study is done in this paper on the behavior of MOX particles in GB such as entraining and transiting. A modeling code has been developed by improving AQUA-SF code and the RF values for abnormal occurrences, such as free-fall spill, outflow and fire, have been analyzed and compared with those reported. This paper also shows the analytical results of the improved code together with the simulated experimental results. It is found that the calculated values are almost corresponded to those reported and that the improved code can estimate MOX particle behavior in GB well.