Near-complete transuranic waste incineration in a thorium fuelled pressurised water reactor

The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Incineration, and virtual elimination, of waste stockpiles is possible in a thorium (Th) fuelled critical or subcritical fast reactor. Fuel cycles producing a net decrease in TRUs are possible in conventional pressurised water reactors (PWRs). However, minor actinides (MAs) have a detrimental effect on reactivity and stability, ultimately limiting the quality and quantity of waste that can be incinerated. In this paper, we propose using a thorium-retained-actinides fuel cycle in PWRs, where the reactor is fuelled with a mixture of thorium and TRU waste, and after discharge all actinides are reprocessed and returned to the reactor. To investigate the feasibility and performance of this fuel cycle an assembly-level analysis for a one-batch reloading strategy was completed over 125 years of operation using WIMS 9. This one-batch analysis was performed for simplicity, but allowed an indicative assessment of the performance of a four-batch fuel management strategy. The build-up of 233U in the reactor allowed continued reactive and stable operation, until all significant actinide populations had reached pseudo-equilibrium in the reactor. It was therefore possible to achieve near-complete transuranic waste incineration, even for fuels with significant MA content. The average incineration rate was initially around 330 kg per GW th year and tended towards 250 kg per GW th year over several decades: a performance comparable to that achieved in a fast reactor. Using multiple batch fuel management, competitive or improved end-of-cycle burn-up appears achievable. The void coefficient (VC), moderator temperature coefficient (MTC) and Doppler coefficient remained negative. The quantity of soluble boron required for a fixed fuel cycle length was comparable to that for enriched uranium fuel, and acceptable amounts can be added without causing a positive VC or MTC. This analysis is limited by the consideration of a single fuel assembly, and it will be necessary to perform a full core coupled neutronic-thermal-hydraulic analysis to determine if the design in its current form is feasible. In particular, the potential for positive VCs if the core is highly or locally voided is a cause for concern. However, these results provide a compelling case for further work on concept feasibility and fuel management, which is in progress. © 2011 Elsevier Ltd. All rights reserved.

The performance of closed reactor grade plutonium-thorium fuel cycles in reduced-moderation pressurised water reactors

The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Previous work has indicated that TRU waste can be virtually eliminated in a pressurised water reactor (PWR) fuelled with a mixture of thorium and TRU waste, when all actinides are returned to the reactor after reprocessing. However, the optimal configuration for a fuel assembly operating this fuel cycle is likely to differ from the current configuration. In this paper, the differences in performance obtained in a reduced-moderation PWR operating this fuel cycle were investigated using WIMS. The chosen configuration allowed an increase of at least 20% in attainable burn-up for a given TRU enrichment. This will be especially important if the practical limit on TRU enrichment is low. The moderator reactivity coefficients limit the enrichment possible in the reactor, and this limit is particularly severe if a negative void coefficient is required for a fully voided core. Several strategies have been identified to mitigate this. Specifically, the control system should be designed to avoid a detrimental effect on moderator reactivity coefficients. The economic viability of this concept is likely to be dependent on the achievable thermal-hydraulic operating conditions. © 2012 Elsevier Ltd. All rights reserved.

Safety implications of reactivity variations in fast thorium ADSRs

Nuclear power generation offers a reliable, low-impact and large-scale alternative to fossil fuels. However, concerns exist over the safety and sustainability of this method of power production, and it remains unpopular with some governments and pressure groups throughout the world. Fast thorium fuelled accelerator-driven sub-critical reactors (ADSRs) offer a possible route to providing further re-assurance regarding these concerns on account of their properties of enhanced safety through sub-critical operation combined with reduced actinide waste production from the thorium fuel source. The appropriate sub-critical margin at which these reactors should operate is the subject of continued debate. Commercial interests favour a small sub-critical margin in order to minimise the size of the accelerator needed for a given power output, whilst enhanced safety would be better satisfied through larger sub-critical margins to further minimise the possibility of a criticality excursion. Against this background, this paper examines some of the issues affecting reactor safety inherent within thorium fuel sources resulting from the essential Th90232→Th90233→Pa91233→U92233 breeding chain. Differences in the decay half-lives and fission and capture cross-sections of 233Pa and 233U can result in significant changes in the reactivity of the fuel following changes in the reactor power. Reactor operation is represented using a homogeneous lumped fast reactor model that can simulate the evolution of actinides and reactivity variations to first-order accuracy. The reactivity of the fuel is shown to increase significantly following a loss of power to the accelerator. Where the sub-critical operating margins are small this can result in a criticality excursion unless some form of additional intervention is made, for example through the insertion of control rods. © 2012 Elsevier Ltd. All rights reserved.

Comparison among MCNP-based depletion codes applied to burnup calculations of pebble-bed HTR lattices

The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX. © 2009 Elsevier B.V. All rights reserved.

Effect of95 241Am(n, γ) reaction branching ratio on fuel cycle and reactor design characteristics

The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

Flexible conversion ratio lead cooled reactor design

This paper presents the neutronic design of a lead cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core a wide range of transuranic actinides management strategies: from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients.

The effect of Ln(III)/An(III) separation on TRU multi-recycling in CONFU assembly

This work examines the basic feasibility of the net-zero-balance TRU multi-recycling concept in which trivalent lanthanide fission products (Ln(III) ) are not separated from trivalent actinides (An(III)). The TRU together with Eu and Gd isotopes are recycled in a standard PWR using Combined Non-Fertile and UO2 (CONFU) assembly design. The assembly assumes a heterogeneous structure where about 20% of U02 fuel pins on the assembly periphery are replaced with Inert Matrix Fuel (IMF) pins hosting TRU, Gd, and Eu generated in the previous cycles. The 2-D neutronic analysis show potential feasibility of Ln / An recycling in PWR using CONFU assembly. Recycling of Ln reduces the fuel cycle length by about 30 effective full power days (EFPD) and TRU destruction efficiency by about 5%. Power peaking factors and reactivity feedback coefficients are close to those of CONFU assembly without Ln recycling.

A combined nonfertile and UO2 PWR fuel assembly for actinide waste minimization

A new combined Non Fertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept, ∼20% of the UO2 fuel pins are replaced with fertile free fuel hosting the transuranic elements (TRUs) generated in the previous cycle. This leads to a fuel cycle sustainable with respect to net TRU generation, and the amount and radiotoxicity of the nuclear waste can be significantly reduced in comparison with the conventional once-through UO2 fuel cycle. It is shown that under the constraints of acceptable power peaking limits, the CONFU assembly exhibits negative reactivity feedback coefficients comparable in values to those of the reference UO2 fuel. Feasibility of the PWR core operation and control with complete TRU recycle has been shown based on full-core three-dimensional neutronic simulation. However, gradual buildup of small amounts of Cm and Cf challenges fuel reprocessing and fabrication due to the high spontaneous fission rates of these nuclides and heat generation by some Pu, Am, and Cm isotopes. Feasibility of the processing steps becomes more attainable if the time between discharge and reprocessing is 20 yr or longer.

Liquid salt cooled flexible conversion ratio fast reactor: Neutronic design

This paper presents the neutronic design of a liquid salt cooled fast reactor with flexible conversion ratio. The main objective of the design is to accommodate interchangeably within the same reactor core alternative transuranic actinides management strategies ranging from pure burning to self-sustainable breeding. Two, the most limiting, core design options with unity and zero conversion ratios are described. Ternary, NaCl-KCl-MgCl2 salt was chosen as a coolant after a rigorous screening process, due to a combination of favourable neutronic and heat transport properties. Large positive coolant temperature reactivity coefficient was identified as the most significant design challenge. A wide range of strategies aiming at the reduction of the coolant temperature coefficient to assure self-controllability of the core in the most limiting unprotected accidents were explored. However, none of the strategies resulted in sufficient reduction of the coolant temperature coefficient without significantly compromising the core performance characteristics such as power density or cycle length. Therefore, reactivity control devices known as lithium thermal expansion modules were employed instead. This allowed achieving all the design goals for both zero and unity conversion ratio cores. The neutronic feasibility of both designs was demonstrated through calculation of reactivity control and fuel loading requirements, fluence limits, power peaking factors, and reactivity feedback coefficients. © 2009 Elsevier B.V. All rights reserved.

The closed thorium-transuranic fuel cycle in reduced-moderation PWRs and BWRs

Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios, which is desirable to maximize the TRU burning rate. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage TRU burning cycle, where the first stage is Th-Pu MOX in a conventional PWR feeding a second stage continuous burn in RMPWR or RBWR, is technically reasonable, although it is more suitable for the RBWR implementation. In this case, the fuel cycle performance is relatively insensitive to the discharge burn-up of the first stage. © 2013 Elsevier Ltd. All rights reserved.

First-principles LDA plus U and GGA plus U study of neptunium dioxide

We have performed a systematic first-principles investigation to calculate the electronic structures, mechanical properties, and phonon-dispersion curves of NpO2. The local-density approximation+U and the generalized gradient approximation+U formalisms have been used to account for the strong on-site Coulomb repulsion among the localized Np 5f electrons. By choosing the Hubbard U parameter around 4 eV, the orbital occupancy characters of Np 5f and O 2p are in good agreement with recent experiments [A. Seibert, T. Gouder, and F. Huber, J. Nucl. Mater. 389, 470 (2009)]. Comparing to our previous study of ThO2, we note that stronger covalency exists in NpO2 due to the more localization behavior of 5f electrons of Np in line with the localization-delocalization trend exhibited by the actinides series.

钇和镧系元素萃取热力学及双水相体系分离氨基酸的研究

本文研究了一种中性萃取剂支链三烷基氧化膦（Cyanex 925）和一种羧酸类萃取剂仲辛基苯氧基取代乙酸（CA-12）萃取稀土元素和钇的热力学性能。探讨了利用CA -12为萃取剂将钇与其他稀土分离的可行性，遵循基础－应用的原则，完成了从龙南离子型矿的浸出液中提取钇的分馏串级模拟实验。在此基础上还进一步研究了“绿色溶剂”离子液为溶剂，CA-12萃取稀土和钇的热力学机理。我们还考察了双水相中氨基酸的分离，为利用双水相体系萃取稀土元素奠定了一定的基础。具体的研究内容如下： 1.系统的研究了Cyanex 925在硝酸体系中萃取稀土和Y的规律，由斜率分析方法确定了反应机理，发现了明显的四分组效应，并确定了Y在萃取中所处的位置。同时发现Cyanex 925有可能用于轻、重稀土分组，易反萃。 2.CA-12对Y萃取具有高的选择性，研究了Y与其他稀土分离的可能性。进行了CA-12从混合稀土中提取Y的工艺模拟实验，并获得纯度为99.5％Y2O3，该工艺高效简便，具有好的应用前景。 3.系统的研究了离子液作为溶剂，CA-12从硝酸介质中萃取稀土和Y的规律。考察了不同水相酸度、水相中相关各种离子及萃取剂浓度变化对CA-12萃取稀土和钇的影响，从而推导了萃取反应方程式及机理。并发现在同样萃取剂浓度和水相条件下，CA-12-离子液体系中萃取稀土和钇的能力低于CA-12-庚烷体系中。 4.研究了赖氨酸、蛋氨酸、苯丙氨酸和半胱氨酸在聚乙二醇（PEG）－磷酸盐双水相体系（ATPSs）中分配行为，分别考察了PEG分子量、水相pH、氨基酸侧链结构等对分配比的影响，得出氨基酸在双水相中的分配行为取决于双水相体系的性质和氨基酸的支链结构与带电情况。

稀土对红细胞及带3蛋白胞质片段结构和生理功能的影响

稀土在农业和医学中的广泛应用使人们日益关心稀土的安全性问题。本论文以红细胞为研究对象，围绕稀土对红细胞形态的影响，稀土能否进入红细胞以及稀土对红细胞带3蛋白胞质片段结构和功能的影响三方面开展研究工作，主要研究成果如下：首先，在宏观上通过超高倍光学显微镜观察了稀土阳离子及其络阴离子对正常人红细胞形态的影响。结果表明当红细胞与硝酸悯作用后其体积发生膨胀，表面出现棘状凸起，细胞间发生聚集。而以往被人们认为毒性较小的柠檬酸悯则使红细胞呈现囊泡状凸起，随着与稀土作用时间的延长多数囊泡状结构可脱落。EDTA的加入可使部分在低浓度（10~(-7)M）稀土离子条件下变形的红细胞恢复原状，说明细胞形态变化主要是由环境中稀土的存在引起的。其次，根据稀土离子跨膜研究中存在的一些问题，在总结前人工作的基础上，建立一种方法测定了体外温育和静脉注射条件下红细胞中稀土含量。进一步证实文献报道含有络合剂的洗涤缓冲液能够将进入细胞内部的稀土带出，影响测定结果的准确性。本论文中应用不含络合剂的洗涤缓冲液洗涤与稀土温育后的红细胞，在10mMTris-HCl印H7.0）低渗缓冲液中溶胀，ICP-MS测定结果显示无论是稀土阳离子还是稀土络阴离子均可以跨膜，且稀土络阴离子跨膜速度较快。耳静脉注射稀土后仅在兔血浆及红细胞膜上检测到稀土，说明在短期静脉注射条件下存在于血浆中的稀土不能进入兔红细胞。最后，应用基因工程技术克隆、表达、纯化了带3蛋白胞质片段及其融合蛋白。计算机结构模拟、荧光光谱以及生物活性的测定证实重组带3蛋白胞质片段与天然蛋白具有相似的空间结构和生物活性。稀土离子对醛缩酶、醛缩酶与带3蛋白胞质片段相互作用的影响表现为：0-10μMLa~(3+) 对醛缩酶活性有促进作用，当体系中L~(3+)浓度达到6μM时，带3蛋白胞质片段基本完全失去对醛缩酶活性的抑制作用。蛋白质内源荧光和同步荧光光谱研究结果表明，La~(3+)对带3蛋白胞质片段与醛缩酶结构均具有一定影响。本论文实验结果表明，低浓度稀土可导致调节细胞内糖酵解速率的带3蛋白胞质片段失去活性，使糖酵解速率无序增强。由于红细胞主要碳源来自于血糖，糖酵解速率的加快很可能会引起生物体血糖浓度的降低。

茚基含氮、氧侧链分子内配位稀土有机化合的合成和性质

稀土有机碳σ-键配合物和氢化物不仅可以催化许多有机反应,而且还可以催化极性单体与非极性单体的聚合.作为稀土有机碳σ-键配合物与氢化物的前体,双配稀土氯化物(C<,9>H<,7>)<,2>LnCl一直是稀土有机化学中研究的热点.1.合成了一系列双配(四氢糠基茚基)稀土氯化物(C<,4>H<,7>OCH<,2>C<,9>H<,6>)<,2>LnCl(Ln=La,Pr,Nd,Sm,Gd,Dy,Y,Ho,Er,Yb,Lu).除了Pr以外,所有化合物的晶体结构都被X-射线衍射表征.2.合成并用X-衍射表征了3-(2-吡啶甲基)茚基锂(C<,5>H<,4>NCH<,2>C<,9>H<,6>)Li(THF)<,2>的晶体结构.3.合成了双配[3-(2-吡啶甲基)茚基]稀土氯化物(C<,5>H<,4>NCH<,2>C<,9>H<,6>)<,2>LnCl(Ln=Sm,Nd),并得到了配合物Nd的晶体结构.4.用二碘化钐(镱)与3-(2-吡啶甲基)茚基锂反应合成了二价双配[3-(2-吡啶甲基)茚基]稀土配合物(C<,5>H<,4>NCH<,2>C<,9>H<,6>)<,2>Ln(Ⅱ)(THF)(Ln=Sm,Yb).5.在用无水氯化稀土YbCl<,3>与3-(2-吡啶甲基)茚基锂反应合成双配[3-(2-吡啶甲基)茚基]稀土氯化物时,由于发生了还原反应,得到了二价双配[3-(2-吡啶甲基)茚基]镱化物(C<,5>H<,4>NCH<,2>C<,9>H<,6>)<,2>Yb(Ⅱ)(THF).6.二价双配[3-(2-吡啶甲基)茚基]稀土配合物(C<,5>H<,4>NCH<,2>C<,9>H<,6>)<,2>Ln(Ⅱ)(THF)(Ln=Sm,Yb)对已内酯具有很好的催化聚合活性.聚合反应可控,并具有活性聚合的特征.

Competition in rotation-alignment between high-j neutrons and protons in transfermium nuclei

The study of rotation-alignment of quasiparticles probes sensitively the properties of high-j intruder orbits. The distribution of very-high-j orbits, which are consequences of the fundamental spin-orbit interaction, links with the important question of single-particle levels in superheavy nuclei. With the deformed single-particle states generated by the standard Nilsson potential, we perform Projected Shell Model calculations for transfermium nuclei where detailed spectroscopy experiments are currently possible. Specifically, we study the systematical behavior of rotation-alignment and associated band-crossing phenomenon in Cf, Fm, and No isotopes. Neutrons and protons from the high-j orbits are found to compete strongly in rotation-alignment, which gives rise to testable effects. Observation of these effects will provide direct information on the single-particle states in the heaviest nuclear mass region.