球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

快堆闭式燃料循环对提高铀资源利用率的分析研究

快堆结合闭式燃料循环提高铀资源利用率需对乏燃料进行回收和再循环。对工业钚在大型MOX(混合铀钚)燃料钠冷增殖快堆中多次循环的特性进行了计算分析,结果表明,钚成分经多次循环后可达平衡,其中易裂变核维持在约74%的较高比例。从成分品质看,工业钚在增殖快堆中的循环次数不受限制。构建模型并分析了快堆闭式燃料循环对于铀资源利用率的提高。快堆闭式循环策略下,回收铀、钚多次循环后可大幅度提高铀资源利用率。提高燃料燃耗和乏燃料后处理回收率能显著提升铀利用率;但在最初的几次循环中后处理回收率的影响较小,循环次数增加后,将会对利用率有明显提升。较低的燃料燃耗和回收率情况下,将存在较低的无限次循环铀利用率上限。     

球床式高温气冷堆示范工程球形燃料元件的研制

为满足球床式高温气冷堆（HTR-PM）示范工程对球形燃料元件大批量生产、单球高铀含量和低破损率的要求,必须对10 MW高温气冷堆时期的球形燃料元件生产工艺进行改进和优化。通过对基体石墨粉、穿衣、压制、车削和热处理等关键设备及相关工艺进行重新设计和优化,建立了规模化的球形燃料元件生产工艺。采用该工艺生产的球形燃料元件,冷态性能如压碎强度、热导、磨损和腐蚀等均满足HTR-PM的技术指标,特别是球形燃料元件的平均自由铀含量与HTR-PM球形燃料元件的自由铀含量指标（6.0×10^-5）相差近1个数量级。采用优化后的规模化生产工艺,成功地研制出符合HTR-PM技术要求的球形燃料元件。     

重水堆核燃料循环

重水堆(HWR)一个最重要的特点就是中子经济性好，高的中子经济性使得重水堆可以使用天然铀。重水堆除可以用天然铀之外，还可用低富集度铀、轻水堆乏燃料回收的铀、MOX燃料和钍燃料等。这使重水堆的燃料循环具有更大的灵活性。     

球床模块式高温气冷堆失冷事故特性研究

利用高温气冷堆专用系统分析软件THERMIX程序,对球床模块式高温气冷堆(HTR-PM)失冷失压和失冷不失压事故的动态特性进行了研究,分析了堆芯功率、燃料最高温度及堆舱水冷壁余热载出功率等关键参数的变化过程,并对影响余热排出功率和燃料最高温度的不确定性进行了评价.研究结果表明,在失冷事故下,堆芯余热可通过热传导、辐射和自然对流等非能动方式传至最终热阱大气,燃料元件和压力容器等重要部件的最高温度均在设计限值内.这为HTR-PM保持模块式高温气冷堆固有安全性不变的同时实现单堆250 MW的功率方案奠定了基础,也为后续高温气冷堆电站示范工程进一步的深入设计研究提供了依据.     

球床式高温气冷堆的余热不确定性分析

反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。     

高温气冷堆乏燃料分析实验室设计与设备研究

针对我国高温气冷堆乏燃料研究设施的空白,研究设计了专门用于高温气冷堆球形燃料元件辐照后性能研究的乏燃料分析实验室和专用工艺设备。基于球形燃料元件与包覆燃料颗粒的特殊结构,所设计的乏燃料分析实验室包括5间热室、6个手套箱和辅助设施,研究设计了专用的工艺实验设备,能够对辐照后的高温气冷堆燃料元件和包覆燃料颗粒进行宏观检查、燃耗测量、元件解体、模拟事故条件加热、辐照微球γ测量分析破损率,通过金相显微镜和扫描电镜进行微观结构分析,开展燃料元件的辐照失效机理研究。     

250 MW球床模块式高温气冷堆进水事故研究

基于250 MW球床模块式高温气冷堆(HTR-PM)的初步设计,以高温气冷堆专用系统分析软件TINTE程序为主要工具,对蒸汽发生器1根传热管双端断裂设计基准的进水事故进行了分析,研究了反应堆温度和压力的变化特性、球床石墨的腐蚀率以及安全阀开启所造成的可燃气体排放等.此外,还分析了风机挡板关闭失效情况下堆内温度分布差异所造成的自然循环对事故后果的影响.计算结果表明:在蒸汽发生器1根传热管双端断裂、最大进水量600 kg情况下,事故后燃料元件的最高温度远低于设计限值,化学反应所引起的石墨腐蚀不会造成反应堆结构强度的破坏和燃料元件的意外破损,释放到反应堆舱室的可燃气体含量也不存在爆炸危险.     

高温气冷堆不确定性分析的新进展

球床高温气冷堆由于采用流动球床堆芯和燃料多次通过的运行方式,不能直接套用轻水堆中一般采用的“系统分解,逐级传递”的分析思路,其不确定性的传播和分析具有特殊性。清华大学核能与新能源技术研究院基于高温气冷堆的设计分析经验,开展了高温堆的不确定性研究,并取得了一些进展。目前高温气冷堆已建立起完整的不确定性分析计算框架。在此框架内,基于VSOP程序,开发能反映球床高温气冷堆实际运行特点的不确定性分析程序VSOP-UAM,实现了核数据不确定性隐式效应和显式效应的完整分析。然后使用SCALE/TSUNAMI-3D和VSOP-UAM程序,建立燃料球、堆芯单元、初装堆芯和平衡堆芯的分析模型,量化了核数据的不确定性对各种模型关键参数的影响。此外,还量化了球流混流效应、燃料富集度、燃料孔隙率这些球床堆芯参数的不确定性对堆芯有效增殖因数k_eff和功率分布的影响。从计算结果可看出,高温气冷堆的不确定性分析显示出了有别于传统轻水堆的结果。     

球床式高温气冷堆球流混流的影响分析

研究球床式高温气冷堆球流存在的混流对堆芯关键参数的影响。开发了能模拟球流混流过程与效果的MFVSOP程序。选择球床模块式高温气冷堆核电站示范工程(HTR-PM)平衡堆芯为研究对象,对比分析不同的混流程度对堆芯功率峰值、功率密度等参数的影响及其不确定性。分析发现,混流对球床式高温气冷堆关键参数的不确定性影响不大,多次通过的燃料循环方式可降低不确定性。     

Radiotoxicity hazard of U-free PuO2+ZrO2 and PuO2+ThO2 spent fuels in LWR

The radiotoxicity hazard of U-free Rock-like oxide: ROX (PuO₂+ZrO₂) and Thorium oxide: TOX (PuO₂+ThO₂) LWR spent fuels is investigated and radiotoxicity hazard of MOX spent fuel is considered as a reference case. The long-term ingestion radiotoxicity hazard of ROX spent fuel is one third and nearly one fourth of that of TOX and MOX spent fuels, respectively. This is because the discharged Pu and long lived Np in ROX fuel is less than that of TOX and MOX fuels. In TOX fuel, discharged Pu and MA are lower than that of MOX fuel but the long-term radiotoxicity hazard of spent fuel is nearly the same as MOX spent fuel. At the cooling 10⁵ years, the radiotoxicity hazard of TOX spent fuel is approximately ten and three times higher than that of ROX and MOX spent fuels, respectively due to higher toxic contribution of ²²⁹Th in TOX spent fuel.     

HTGR reactor physics and fuel cycle studies

The high-temperature gas-cooled reactor (HTGR) appears as a good candidate for the next generation of nuclear power plants. In the “HTR-N” project of the European Union Fifth Framework Program, analyses have been performed on a number of conceptual HTGR designs, derived from reference pebble-bed and hexagonal block-type HTGR types. It is shown that several HTGR concepts are quite promising as systems for the incineration of plutonium and possibly minor actinides.These studies were mainly concerned with the investigation and intercomparison of the plutonium and actinide burning capabilities of a number of HTGR concepts and associated fuel cycles, with emphasis on the use of civil plutonium from spent LWR uranium fuel (first generation Pu) and from spent LWR MOX fuel (second generation Pu). Besides, the “HTR-N” project also included activities concerning the validation of computational tools and the qualification of models. Indeed, it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel.These validation and qualification activities have been centred round the two HTGR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3D diffusion theory codes yield acceptable results. Special attention, however, has to be given to the modelling of neutron streaming effects. For the HTR-10 the analyses focused on first criticality, temperature coefficients and control rod worth. Also in these studies a good correspondence between calculation and experiment is observed for the 3D diffusion theory codes.     

基于MOX燃料组件的177混合堆芯装料方案数值研究

通过计算华龙一号（HPR1000）压水堆平均卸料燃耗得到乏燃料中钚（Pu）同位素的含量,以此成分比例来设计铀钚混合氧化物（MOX）燃料。采用离散型燃料组件设计,通过不同Pu含量的MOX燃料棒离散型布置来降低与UO₂燃料组件间的功率梯度。采用程序MCNP和COSLATC模拟堆芯功率分布和热中子注量率分布,采用分区分层的低泄漏装料方案,降低不同燃料组件间的功率梯度,展平堆芯的功率分布。在不考虑可燃毒物的前提下,利用3种Pu含量的MOX组件将混合堆芯的功率峰因子控制在1.77左右,明显优于原堆芯的功率峰因子,为国产三代压水堆引入MOX燃料提供了具有参考价值的装料方案。      

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

Radiotoxicity hazard of inert matrix fuels after burning minor actinides in light water reactors

Ingestion radiotoxicity hazard index of inert matrix spent fuels are investigated after burning minor actinide (MA) isotopes in LWRs and compared with the hazard index of MOX and MA burning MOX (MOX+MA) spent fuels. As U-free fuels, ROX: (PuO₂+ZrO₂) and TOX: (PuO₂+ThO₂), are considered, in which MA's are added as oxides. The radiotoxicity hazard index of ROX+MA spent fuel is less than that of TOX+MA and MOX+MA spent fuels due to the lower density of actinides in spent fuel. Some of cooling years the toxic yield of ROX+MA spent fuel is even less than that of MOX spent fuel, if the initial loaded MA in ROX is about 0.5 at %.     

Core characteristics of fast reactor cycle with simple dry pyrochemical processing

Fast reactor core concept and core nuclear characteristics are studied for the application of the simple dry pyrochemical processing for fast reactor mixed oxide spent fuels, that is, the Compound Process Fuel Cycle, large FR core with half of loaded fuels are recycled by the simple dry pyrochemical processing. Results of the core nuclear analyses show that it is possible to recycle FR spent fuel once and to have 1.01 of breeding ratio without radial blanket region. The comparison is made among three kinds of recycle fuels, LWR UO₂ spent fuel, LWR MOX spent fuel, and FR spent fuel. The recycle fuels reach an equilibrium state after recycles regardless of their starting heavy metal compositions, and the recycled FR fuel has the lowest radio-activity and the same level of heat generation among the recycle fuels. Therefore, the compound process fuel cycle has flexibility to recycle both LWR spent fuel and FR spent fuel. The concept has a possibility of enhancement of nuclear non-proliferation and process simplification of fuel cycle.     

压水堆内钍-铀增殖循环研究——乏燃料特性分析

利用ORIGENS程序对压水堆钍基乏燃料的特性进行分析,揭示了钍基乏燃料在放射性毒性、衰变热、γ射线等方面的特性,相关结果可为钍基乏燃料的贮存、后处理和地质处置提供必要的参考。研究的乏燃料是压水堆内钍-铀增殖循环堆芯设计方案中的4种,包括UOX（铀氧化物）、MOX（钚铀混合氧化物）、PuThOX（钚钍混合氧化物）和U3ThOX（工业级²³³U-钍混合氧化物）。研究结果表明：1）由于超铀核素的含量极低,在卸料后1 000年内,U3ThOX的放射性毒性显著低于超铀核素含量高的乏燃料;2）由于²³²U衰变链中²⁰⁸Tl的贡献,钍基乏燃料中2.6 MeV能量附近的γ射线强度明显高于铀基乏燃料,而这一能量附近的γ射线强度在卸料后约10年达到局部峰值,所以,钍基乏燃料的后处理最好避开此时间。     

超临界水冷快堆增殖因子敏感性初步研究

从长远观点来看,超临界水冷快堆（SCFWR）的增殖性能是一个重要问题,由于超临界水堆中冷却剂密度仅相当于当前沸水堆（BWR）的1/3,加之稠密性栅格布置,SCFWR具有增殖的潜力。为了探究SCFWR的增殖性问题,利用基于多群三维细网有限差分中子扩散方程的堆芯核计算方法,设计不同的算例,分别计算了堆芯冷却剂流型、不锈钢和ZrH_1.7的利用、堆型布置、棒径大小、MOX燃料中PuO₂的份额、堆芯燃耗深度及堆芯尺寸等因素对SCFWR增殖性能的影响。计算结果表明,增大堆芯转换比的途径有：采用对流式流型、加入ZrH_1.7层、采用合适的堆芯布置、增加棒径、提高MOX燃料中PuO₂的份额及增大堆芯尺寸而减少中子泄漏等。从而为提高SCFWR的转换比提供了可参考的依据路线。     

铀钚萃取洗涤-共反萃工艺的离心萃取器台架实验研究

由于快堆MOX乏燃料放射性强,需要缩短停留时间以降低溶剂辐解,本工作以离心萃取器为萃取设备,在短停留时间下进行了快堆MOX乏燃料后处理铀钚萃取洗涤-共反萃工艺研究。研究结果显示,该工艺在单级停留时间约20s时具有良好的铀钚收率,萃取洗涤过程中铀和钚收率均大于99.99%,共反萃过程中铀和钚收率分别为99.99%和99.94%;同时能有效防止第三相的形成,避免钚的聚合沉淀。