Neutronic evaluation of coating and cladding materials for accident tolerant fuels

In severe accident conditions with loss of active cooling in the core, zirconium alloys, used as fuel cladding materials for current light water reactors (LWR), undergo a rapid oxidation by high temperature steam with consequent hydrogen generation. Novel fuel technologies, named accident tolerant fuels (ATF), seek to improve the endurance of severe accident conditions in LWRs by eliminating or at least mitigating such detrimental steam-cladding interaction. Most ATF concepts are expected to work within the design framework of current and future light water reactors, and for that reason they must match or exceed the performance of conventional fuel in normal conditions. This study analyzed the neutronic performance of ATF when employed in both pressurized and boiling water reactors. Two concepts were evaluated: (1) coating the exterior of zirconium-alloy cladding with thin ceramics to limit the zirconium available for reaction with high-temperature steam; (2) replacing zirconium alloys with alternative materials possessing slower oxidation kinetics and reduced hydrogen production. Findings show that ceramic coatings should remain 10–30 μm thick to limit the neutronic penalty. Alternative cladding materials, with the exception of SiC, enhance neutron loss compared to zirconium-alloys. An extensive parametric analysis concluded that reference performance metrics can be met by employing 300-μm or less thick cladding or increasing fuel enrichment by up to 1.74% depending on material and geometry.     

离子轰击对氮化铀表面氧化行为的影响

采用离子氮化技术制备得到一定厚度的氮化铀层,以不同能量的氩离子轰击考察氮化铀的辐照氧化行为,并与大气中的自然氧化行为进行对比,考察材料在辐照环境下的稳定性。结果表明,氮化铀表面经氩离子轰击后,表面形貌发生了改变;氩离子轰击氧化与大气中的自然氧化行为存在差异,离子轰击增强了氮化铀表面的氧化程度,但其对氧化行为的影响主要在浅表面,大气氧化的氮化铀氧化层更厚;随着氩离子轰击能量的增加,表面氧化物含量及氧化层深度显著增加。总体而言,氩离子辐照对氮化铀层的影响随深度的增加而减弱,并不影响氮化铀的整体稳定性。     

Rock-like oxide fuel behavior under reactivity initiated accident conditions

Pulse irradiation tests of two types of rock-like oxide (ROX) fuel, i.e. yttria stabilized zirconia (YSZ) and YSZ/Spinel composite, were conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under reactivity-initiated accident conditions. The ROX fuels failed with cladding burst at fuel volumetric enthalpies above 10 GJ m⁻³, which was comparable to that of UO₂ fuel. The failure of the ROX fuels, however, occurred with considerable fuel melting and was quite different to that of UO₂ fuel, which was caused by cladding melting and embrittlement due to heavy oxidation. Lower fuel melting temperature of the ROX fuels compared to that of UO₂ contributed to the different fuel failure modes. Certain amount of molten ROX fuel dispersed out at the failure. However, the mechanical energy generation due to the molten fuel/water interaction was negligible for the ROX fuels at peak fuel enthalpies below 12 GJ m⁻³.     

Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective e?ect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.     

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.     

Fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

ABSTRACT

To evaluate the fracture limit of high-burnup advanced fuel cladding tubes under loss-of-coolant accident (LOCA) conditions, laboratory-scale integral thermal shock tests were performed using the following advanced fuel cladding tubes with burnups of 73–85 GWd/t: M-MDA^TM, low-tin ZIRLO^TM, M5®, and Zircaloy-2 (LK3). In total eight integral thermal shock tests were performed for these fuel cladding tube specimens, simulating LOCA conditions including ballooning and rupture, oxidation, hydriding, and quenching. During the tests, the specimens were oxidized to 10%–30% equivalent cladding reacted (ECR) at approximately 1473 K and were quenched under axial restraint load of approximately 520–530 N. The effects of burnup extension and use of the advanced fuel cladding tubes on the ballooning and rupture, oxidation, and hydriding under LOCA conditions were inconsiderable. Further, the high-burnup advanced fuel cladding tube specimens did not fracture in the ECR values equal to or lower than the fracture limits of the unirradiated Zircaloy-4 cladding tube reported in previous studies. Therefore, it can be concluded that the fracture limit of fuel cladding tubes is not significantly reduced by extending the burnup to approximately 85 GWd/t and using the advanced fuel cladding tubes, though it slightly decreases with increasing initial hydrogen concentration.     

我国核燃料循环前端产业的现状和展望

在分析我国核燃料循环产业面临发展机遇的基础上,阐述了天然铀的保障供给、铀转化和铀浓缩、燃料制造领域等核燃料循环产业的现状以及面临的挑战,并对我国核燃料循环前端产业的发展方向进行了分析预测,针对如何应对挑战提出了相应的对策。     

后处理厂1A柱有机相着火事故源项估算

我国对于后处理工业的需求随着核电事业的迅猛发展变得愈发强烈,为了满足后处理工业安全发展必不可少的核应急需求,为核应急工况下后处理厂的核应急响应与决策支持提供依据。针对后处理厂1A柱有机相着火事故这一基准事故,结合实际工艺流程及监测手段,选取了核应急工况下的可获得参数(有机溶剂泄漏质量等)作为输入,在有机相燃烧速率经验公式基础上,结合后处理的工艺特点,引入少量修正建立了后处理厂1A柱有机相着火事故源项估算模型,并使用FORTRAN编程语言开发了相应软件。数值验证结果表明,该估算模型可以满足后处理厂1A柱有机相着火事故的核应急需求。     

Passive measure for preventing recriticality of fuel debris dust for defueling at Fukushima Daiichi nuclear power station

The drilling or cutting of resolidified fuel debris required to defuel the Fukushima Daiichi nuclear power station is certain to generate debris dust. This paper focused on drilling resolidified fuel debris in water and conservatively confirmed by criticality calculations that neutron multiplication effect is higher if debris dust is suspended separately from the debris rather than if it is suspended closely around the debris. No use of vacuuming of debris dust, borated water, and active components was assumed in this study. Also, this paper confirmed that the use of a debris dust guide effectively and passively limited the increase in neutron multiplication by debris dust because the guide distributes dust particles so flatly that sufficient neutron leakage limits neutron multiplication even if the volume fraction of the particles in water reaches the optimum condition. In the actual defueling operation at the nuclear power station, the use of a flatter debris dust guide will be more effective to prevent local recriticality concurrently with the careful control of the mass of debris dust. The physics and ideas in this paper should be applicable to other defueling technologies such as laser cutting as long as debris dust is generated and suspended in water.     

事故容错燃料包壳和芯块材料中子学分析

分别使用SCALE软件包和两种近似方法进行组件和堆芯中子学分析,进而对SiC、先进铁合金和钼等事故容错包壳材料以及U_3Si_2、U~(15)N和U-Mo等芯块材料进行中子经济性评价。结果表明:除了SiC外,金属包壳均有显著的中子惩罚,需通过提高燃料富集度或减少包壳厚度进行补偿;高密度芯块如U_3Si_2通常能够提高中子经济性,但由于过高的~(238)U含量,U~(15)N无明显经济性优势。     

An overview of reactivity initiated accident behavior of rock-like fueled pressurized water reactors

Reactivity initiated accident (RIA) analyses of plutonium rock-like oxide (ROX) fueled PWRs have been carried out with the point kinetics calculations. As a result, the analyses have shown a very severe transient behavior of the ROX fueled PWR, which is unacceptable without any improvement. It was also found that the RIA behavior of ROX fueled PWRs can be improved by increasing the negative fuel temperature coefficient (_f). For this improvement, the additives in the ROX fuel such as UO₂ and ThO₂ were considered, as well as a ROX assembly partial loading UO₂ core. With UO₂ additive, it was successful to have satisfying _f and RIA behavior of ROX fuel core, while the partial loading core must be further improved. Besides the ROX-PWR RIA analytical study, the actual behavior of the ROX fuel pin under RIA condition has been experimentally investigated at the Nuclear Safety Research Reactor (NSRR) of JAERI. Though the ROX fuel pin failure mechanism with fuel melting seems quite different from that of UO₂ pin with cladding melting, the ROX pin failure threshold was found to be roughly the same as that of UO₂ in terms of accumulated energy per unit fuel volume.     

Preparation of High Density Uranium Nitride and Uranium Carbonitride Fuel Pellets

Uranium nitride and uranium carbonitride fuel pellets were prepared for irradiation in the Japan Material Testing Reactor. The pellets are 6.9 mm in diameter and 7 mm long, and are of natural and 5% enriched uranium. Uranium nitride powder was prepared from uranium metal via hydride and higher nitride. Uranium carbide powder was prepared from uranium metal by hydriding and then reacting with propane. The lowest possible reaction temperatures were selected to obtain fine and reactive powders. Uranium nitride and mixed powders of different ratios (UC: UN = 1: 3, 1:1 and 3: 1) were cold pressed without binder. Sintering was carried out in a tungsten crucible in vacuum (10~⁴ mmHg) for 2 hr at 1,900°–2,000°C. The density of the pellets obtained was in the range of 90~95% of the theoretical value with an oxygen content of 1,300~2,100 ppm. No second phase, such as metallic uranium, were observed in the specimens, either by metallography or X-ray diffraction. These pellets of unexpectedly high density without second phase must have been obtained thanks to the good powder characteristics combined with proper sintering conditions. The compositions of uranium carbonitride pellets were found to be slightly nitrogen deficient, compared with the reactants.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

Development of FRACAS-CT module with FRAPCON4.0P01 for simulation of mechanical behaviors for accident-tolerant fuel cladding in a reactor

In order to evaluate the mechanical behaviors of accident-tolerant fuel (ATF) cladding under reactor conditions, a new analytical module, termed FRACAS-CT, was developed. A mechanical model of FRACAS-CT was formulated based on thick-wall theory to consider the multi-layered structure of ATF cladding. Two statuses of a pellet-cladding gap, that is, open and closed, were described by several boundary conditions of pressures, inside radial displacements, and axial strains of the multi-layered ATF cladding. The FRACAS-CT model was verified by comparison with an equivalent finite element (FE) model and was implemented into FRAPCON4.0P1 with consideration of creep and stress relaxation behaviors of the multi-layered ATF cladding. After the implementation, code verification work was performed, and finally, mechanical behaviors and fuel performance of the multi-layered ATF cladding were calculated under normal operating conditions. As a result, the implemented FRACAS-CT can simulate the mechanical response and fuel performance of the multi-layered ATF cladding.     

球形燃料元件中包覆燃料颗粒的化学分析

报道了高温气冷堆球形燃料元件中包覆燃料颗粒的表面铀沾污、自由铀含量及包覆燃料颗粒的装铀量等性能指标的测试方法、范围及测量误差。利用激光荧光法测量并计算了包覆燃料颗粒中的自由铀含量及表面铀 沾污,利用电位滴定法测量了包覆燃料颗粒的装铀量。结果表明,经4层连续包覆的包覆燃料颗粒的质量符合并满足高温气冷堆球形燃料元件对包覆燃料颗粒的设计要求。     

Fabrication of DUPIC fuel pellets using high burn-up spent PWR fuel

Technology for the direct usage of a spent PWR fuel in CANDU reactors (DUPIC) was developed in KAERI to reduce the amount of spent fuel. DUPIC fuel pellets were fabricated using a dry processing method to re-fabricate CANDU fuel from spent PWR fuel without any intentional separation of fissile materials or fission products. The DUPIC fuel element fabrication process satisfied a quality assurance program in accordance with the Canadian standard. For the DUPIC fuels with various fuel burn-ups between 27,300 and 65,000 MWd/tU, the sintered pellet density decreased with increasing fuel burn-ups. Fission gas releases and powder properties of the spent fuel also influenced the DUPIC fuel characteristics. Measurement of cesium content released from green pellets revealed that their sintered density significantly depended on sintering temperature history. It was useful to establish a DUPIC fuel fabrication technology in which a high-burn-up fuel with 65,000 MWd/tU was treated.     

我国乏燃料离堆贮存需求分析

随着我国核电的大力发展,产生了大量的乏燃料。若不能妥善进行处理,会给核电发展带来不利影响。我国后处理技术的发展现状暂时无法有效缓解乏燃料大量累积造成的困境。本文按照我国的核电发展规划,结合现有的乏燃料贮存能力,计算得出了乏燃料的年产生量、累积量,以及离堆贮存需求。建议我国尽快开展压水堆乏燃料离堆贮存设施的研究工作,确保核电的安全发展。     

切尔诺贝利事故释放的~(131)I所致牧草、奶类的污染水平及其变化规律

本文报道了苏联切尔诺贝利核电站事故释放的~(131)I对乌鲁木齐地区牧草、奶类的污染水平及动态变化规律。经估算,~(131)I经奶类所致成人(1.2μSv)和婴儿(120μSv)甲状腺最大待积剂量当量远低于我国放射卫生防护基本标准规定的年剂量限值,因此,对居民健康不会产生影响。     

Effects of Fuel Oxidation and Dissolution on Volatile Fission Product Release under Severe Accident Conditions

The release of volatile fission products from high-burnup UO₂ fuel was examined in a steam atmosphere under severe accident conditions as a part of the VEGA program. The effects of fuel oxidation and dissolution were totally evaluated, by comparing the results with those from previous inert, hydrogen and steam atmosphere tests. It was shown that the oxidation of UO₂ to UO_2+x by steam generally enhances Cs and Kr release. However, the enhancement becomes smaller above the melting temperature of Zircaloy, about 2030 K, likely due to reduction of UO_2+x by molten Zircaloy. The burst release of Cs occurs above about 2300K in the hydrogen atmosphere, while the release rate does not increase so significantly for the examined temperature range (<2800 K) in the steam atmosphere. Analysis of the hydrogen atmosphere test showed that fuel dissolution is apparently connected with the burst release and that a large fraction of Cs is quickly released from the dissolved fuel above 2300 K. It is considered that the fuel dissolution rate in the steam atmosphere is about 1/1000 of that in the hydrogen atmosphere.     

乏燃料后处理厂有机相燃烧事故研究进展

核能的广泛利用伴随着乏燃料的产生和累积,乏燃料后处理技术将乏燃料再循环利用受到重要推崇,但乏燃料后处理设施的安全是发展后处理技术的重要前提,后处理中的有机相着火事故作为后处理的设计基准事故之一,得到了国内外的重要关注。为分析后处理厂在有机相着火事故中,有机相的燃烧行为、放射性气溶胶的扩散和沉积、高效过滤器的性能等,美国、日本等国分别建立了实验设施并进行了有机相燃烧的实验研究。本文综合评述了国内外关于后处理厂有机相着火事故的试验技术方法和研究结果,提出了当前研究存在的问题以及未来有待进一步研究的方向。