压水堆核燃料循环情景模式初步研究

根据我国核电发展现状和中长期发展规划及中长期（2030、2050）发展战略研究,假设2050年前我国压水堆核电发展规模,基于压水堆乏燃料后处理,回收的钚做成MOX燃料放入压水堆中使用,MOX燃料只使用1次的循环模式,进行核能发展情景研究。基于压水堆可装载30%比例MOX燃料的已有研究结果,考虑我国主要的两种压水堆堆型M310和AP1000,进行压水堆核燃料循环分析。利用核能发展情景动态分析程序DESAE-2,给出了不同情景模式下天然铀需求量、乏燃料累计量等。结果表明：至2050年,B₁和B₂模式较A模式分别节省天然铀4.1万t和2.9万t。     

2020年前我国核燃料循环情景初步研究

根据我国核电现状和中短期发展规划,对2020年前我国核电规模提出了三种预测方案,并根据各种方案对压水堆电站的核燃料循环情景进行了计算。重点研究了压水堆核电所需的铀资源、分离功,卸出的乏燃料以及乏燃料中Pu和次要锕系元素(MA)的产生量。     

核燃料循环技术经济和战略综述

本文简要综述了核燃料循环前端和尾端各环节包括采矿、水法治金、铀转化、铀富集、燃料元件制造、乏燃料后处理（处置）、废物处理和处置等工艺的技术现状和发展方向；分析了核燃料循环各环节的成本构成和在核电成本中的份额，目前核燃料循环成本约占核电成本27%，随着燃料元件性能的改善和燃耗的提高，核燃料循环成本在核电中的份额将有大幅度下降；核燃料循环战略其最大分歧点是关于乏燃料的管理，即对乏燃料元件是采取后处理回收轴钚再循环还是采取直接处置，出于国情，各国采取了不同的战略，由于核安全和环境保护是全球性问题，这两种路线之争还将继续。     

2050年前我国压水堆核燃料循环模式研究

根据中国工程物理研究院的《中国能源中长期(2030、2050)发展战略研究》报告,以我国的两种主流堆型—大亚湾M310和三代AP1000为研究对象,由我国核燃料循环现状和未来发展目标,假定了2050年前我国压水堆核燃料循环的几种可能模式,并利用DESAE-2(核能情景分析软件)计算了假定模式下的铀钚资源需求与高放废物的提取量。计算结果可为我国核能发展策略提供数据参考。计算结果的比较与分析表明,扩大装有MOX燃料的在运营压水堆规模能更有效地节省铀资源,而决定装有MOX燃料的在运营压水堆规模大小的关键在于所具有压水堆乏燃料后处理能力的大小。     

核燃料循环“一次通过”情景分析研究

在我国大力发展核电和建立先进闭式燃料循环体系的前提下,为了核燃料循环各环节相互匹配,有机协调地发展核电,需对核燃料循环各环节的规模、布局等开展模拟分析研究。为了综合比较闭式燃料循环在铀资源利用率、环境友好型等方面的优势,还需对核燃料循环的"一次通过"方式进行分析研究。根据我国核电目前和中长期发展规划,假设2050年中国核电装机容量分别达到100GW、200GW和300GW,核电站全部采用压水堆,核燃料循环采用"一次通过"方式,利用简化模型,计算分析了近100年内累计需要的天然铀、每年需要的天然铀、每年需要的分离功、累计产生的乏燃料和需要离堆暂存的乏燃料量等关键数据。     

烧掉锕系元素——未来的展望

【《欧洲核综览》1998年 5— 6月刊第 4 0页报道】　在世界范围的核电生产中 ,轻水堆 (L WR)是一个主要堆型。它们每年产生1万吨乏燃料。这种燃料仍然含有四分之一的原始2 3 5U以及另外产生的 75吨钚 ,三分之二以上是易裂变材料。通过后处理 ,从裂变产物中分离铀和钚 ,其工业规模大约只有 15 0 0吨 /年。大多数乏燃料存放在中间贮藏设施中 ,等待转运到最终地下贮存库。全世界乏燃料所含的钚达到 80 0吨。从民用核燃料中已经分离出约 2 0 0吨的钚 (并且已循环使用 ) ;另有 2 30吨武器级钚还在核武器上。实际上所有乏燃料都是可以回收的 ,但…     

日本确立快堆乏燃料预处理技术

【日本《原子能视野》2003年2月刊报道】 核燃料循环开发机构与三菱重工业公司、三菱材料公司合作,确立了用于快堆乏燃料干法后处理的预处理技术。预处理技术是将乏燃料棒粉碎,然后从中以高纯度回收氧化铀和氧化钚,最后再将剩余乏燃料送往后处理主体的一系列技术。 核燃料循环开发机构这次开发了机械式粉碎机、磁分离技术及高频感应加热方式。采用这些分离技术对燃料棒进行粉碎处理,并将金属棒里面的氧化物近乎100%地分离。干法处理技术是快堆时代不可缺少的技术,所以正被当作下一代后处理技术加以研究,而这次开发的这一系列技术将使干法后…     

日本高放废物的处置现状及各国的情况

【日本《原子能视野》2002年12月刊报道】 目前,核电占日本发电总量的约1/3。由于资源匮乏,所以日本把“核燃料循环”作为其核政策的基本方针。虽然日本自2002年8月以来发生了一连串的核事件,国民对核电的信赖瞬间化为乌有,但从确保能源长期供应的角度出发,日本不会改变其核燃料循环的方针。 核电厂日常运行产生的低放废物将被装入屏蔽容器罐、用水泥固化,然后运往青森县六所村的低放废物埋藏设施,埋藏于浅地层的混凝土处置场内。日本目前采用的乏燃料后处理方式是,先将乏燃料溶于硝酸溶液,然后再分离和提取出其中的铀和钚,同时产生高放废…     

跨国研究团队公布核燃料循环对废物处置影响的研究结果

【美国《核新闻》2009年1月刊报道】美国电力研究院（EPRI）与法国电力公司（EDF）最近合作开展了一项关于核燃料循环对废物处置影响的研究。在这项研究中,假设的核燃料循环情景是先进燃烧堆（ABR）已投入商业应用并对乏燃料进行后处理。这项研究的结论是使用先进核燃料循环能够降低核废物的处置需求,但目前还需要开展大量的研发工作,以确保能够安全且经济地将商业乏燃料后处理和循环技术投入使用。     

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。     

HTGR fuel and fuel cycle

This paper describes some of the basic charactritics of the HTGR fuel with emphasis on the 1160 MW(e) plant now being offered commercially by Gulf General Atomic and some of the aspects of the fuel cycle which are unique to the HTGR. The fuel cycle is based on highly enriched (93%) uranium for the initial and the make-up fissile material; thorium for the fertile material, with the bred ²³³U being recycled at the earliest opportunity. The fuel elements consist only of ceramic materials with the thorium/uranium carbides or oxides in the form of coated particles.     

Nuclear fuel management via fuel quality factor averaging

The numerical procedure of prime number averaging is applied to the fuel quality factor distribution of once- and twice-burned fuel in order to evolve a fuel management scheme. The resulting fuel shuffling arrangement produces a near optimal flat power profile both under beginning-of-life and end-of-life conditions. The procedure is easily applied requiring only the solution of linear algebraic equations.     

Plutonium fuel and fuel elements for power reactors

Conclusions The use of plutonium in the fuel cycle during complex utilization of thermal and fast reactors in nuclear energetics permits solving the problem of ensuring nuclear fuel for a long period. Oxide uranium-plutonium fuel facilitates the development of technology of fast reactors and so far it is considered as the basic type of fuel. At the same time, oxide fuel cannot ensure the required rate of plutonium accumulation, in view of which the investigations of more efficient fuel and constructional materials become a pressing problem. The use of uranium-plutonium oxide fuel in thermal reactors requires improvements in the construction of fuel elements and organization of large-scale completely automatic production.Translated from Atomnaya Énergiya, Vol. 43, No. 5, pp. 412–417, November, 1977. Editors' Remarks. For the completeness of the discussion of the problem it is, of course, necessary to consider the possibility of using plutonium in fast and thermal reactors as done by the authors. However, it should be kept in mind that by its nuclear-physical parameters plutonium as a nuclear fuel is more suitable for use in fast reactors than in thermal reactors. The use of plutonium in thermal reactors can reduce the demands of natural uranium for the development of nuclear power in all by 10–15%, whereas its use in fast reactors reduces the demand for uranium by a factor of 10.All this indicates the feasibility of using plutonium only in fast reactors even if its accumulation is required over a certain period.     

Investigating fuel element melt in simulators with fuel compositions

Nizhegorod Polytechnical Institute. Translated from Atomnaya Énergiya, Vol. 70, No. 2, pp. 108–110, February, 1991.     

Nitration by a simulated fuel technique for nitride fuel re-fabrication

Nitration reaction of a spent nuclear oxide fuel through a carbothermic reduction and the change in thermal conductivity of the resultant nitride specimens were investigated by a simulated fuel technique for use in nitride fuel re-fabrication from spent oxide fuel. The simulated spent oxide fuel was formed by compacting and sintering a powder mixture of UO₂ and stable oxide fission product impurities. It was pulverized by a 3-cycle successive oxidation-reduction treatment and converted into nitride pellet specimens through the carbothermic reduction. The rate of the nitration reaction of the simulated spent oxide fuel was decreased due to the fission product impurities when compared with pure uranium dioxide. The amount of Ba and Sr in the simulated spent oxide fuel was considerably reduced after the nitride fuel re-fabrication. The thermal conductivity of the nitride pellet specimen in the range 295-373 K was lower than that of the pure uranium nitride but higher than the simulated spent oxide fuel containing fission product impurities.     

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

This paper provides an overview of high-temperature phenomena in nuclear fuel elements and bundles, with particular relevance to the CANDU fuel design. The paper describes heat generation, fuel thermal response, and thermophysical properties of the fuel and sheath that can affect the thermal and mechanical response of the fuel element. Sources of chemical heat that can arise during accident conditions in the fuel element are also detailed. Specific phenomena associated with fuel restructuring, fuel sheath deformation, fuel-to-sheath heat transfer, fuel sheath failure criteria, oxidation, hydriding and embrittlement of the Zircaloy sheath, gap transport processes in failed elements, fuel/sheath interaction and fuel dissolution by molten cladding are detailed as important phenomena that can impact reactor safety analysis. Fuel behaviour during a power pulse and fuel bundle behaviour that occurs during a severe reactor accident are further considered. The review also points out areas of further research that are needed for a more complete understanding.     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.