压水堆内钍-铀增殖循环研究——堆芯设计

在全UOX（铀氧化物）堆芯平衡循环的基础上,研究了UOX/PuThOX（钚钍混合氧化物）混合堆芯和UOX/U3ThOX（工业级²³³U-钍混合氧化物）混合堆芯的燃料管理方案设计,实现了钍 铀增殖循环。U3ThOX燃料组件在堆内停留6个燃料循环,平均循环长度较参考的全UOX堆芯增加5 EFPD;U3ThOX燃料组件卸料后冷却1年时易裂变核素存量较装料时增加了7%。为比较分析,设计了UOX/MOX（钚铀混合氧化物）混合堆芯的燃料管理方案。核特性分析结果表明：1）装载PuThOX燃料对堆芯核特性产生的影响与装载MOX燃料类似,硼微分价值和控制棒价值减小、临界硼浓度变大、慢化剂温度系数更负、停堆裕量减小、多普勒亏损更大;2） UOX/U3ThOX混合堆芯和参考的全UOX堆芯具备相似的核特性。     

基于粒子群优化算法的压水堆换料优化初步研究

基于大亚湾核电站压水堆堆芯燃料组件布置建立优化的数学模型,采用离散的多目标粒子群优化算法和基于有限元法的堆芯计算软件"Donjon",及组件计算软件"Dragon",将粒子群程序与堆芯物理计算软件结合,编制完整的堆芯燃料管理优化程序。应用此优化程序对大亚湾核电站初装料进行优化的结果表明:粒子群优化算法得到的优化方案与参考方案相比,keff增加11.9%,且功率峰因子低于1.4,满足安全限值。     

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。     

超临界水冷堆MOX燃料组件控制棒特性研究

超临界水冷堆(SCWR)因具有较高的热效率和较强的经济竞争性等优势引起许多国家和地区的广泛关注。MOX燃料即普通燃料UO_2与PuO_2的混合陶瓷燃料替换UO_2会给SCWR堆芯安全带来一定的不确定性。因而MOX燃料组件的反应性控制与普通燃料有较大差异。论文采用MCNP5软件对SCWR采用传统核燃料与MOX燃料组件的控制棒控制性能进行了分析和对比,结果表明:MOX燃料组件中子能谱硬化,控制棒中硼(B)的丰度越大,控制棒直径越大,其控制效果越理想。控制棒对径向功率峰抑制效果明显,而对轴向功率分布影响较小。计算结果对压水堆新型MOX燃料组件控制棒设计有一定参考意义。     

超临界水冷堆MOX燃料特性分析

针对超临界水冷堆组件,采用不同Pu含量的MOX燃料进行组件计算,得到不同燃料条件下的燃耗深度、功率分布因子、慢化剂温度反应性系数等结果,并对比分析在超临界水冷堆中应用MOX燃料与应用UO2燃料对组件性能的影响,以及不同Pu含量MOX燃料间的性能区别。分析结果表明,在超临界水冷堆设计中,应用MOX燃料与应用UO2燃料有相似的功率分布,应用MOX燃料可以增加燃耗深度,并有良好的慢化剂温度反应性系数。经过合理设计的MOX燃料可较好应用于超临界水冷堆中,且产生更好的性能。     

含MOX燃料组件堆芯顶端注量率峰问题研究

通过详细的建模分析,阐述了顶端注量率峰产生的现象及原因,发现钚铀氧化物混合(MOX)组件顶端热中子注量率峰普遍存在于UO2燃料与MOX燃料的混合堆芯中,针对这一问题,提出了改变MOX组件或UO2组件的顶端组成,在组件顶端加入不同的材料,达到减少中子注量率峰的目的。计算表明,通过在MOX燃料棒顶部加入12～18 cm的Gd2O3毒物的效果最好;采用MOX燃料中心开孔的方式,也能解决这一问题。     

灵活燃料循环革新型水冷反应堆（FLWR）的概念设计和循环特性

提出了一种基于轻水反应堆（LWR）技术丰富经验、应用灵活的燃料循环的革新型水冷反应堆（FLWR）概念设计。该设计的目的是通过两个阶段的钚多次循环,实现有效和灵活地利用铀和钚的资源。在第一阶段中,FLWR堆芯是实现高转换型堆芯的概念设计,基本上平稳地保持现有轻水堆和来自轻水堆铀-钚混合氧化物（MOX-LWR）燃料技术的技术连续性,从技术的观点看没有重大的差异;第二阶段的堆芯将是一种慢化剂-减少型水冷反应堆（RMWR）堆芯的概念设计,达到大于1．0的高转换率。钚（Pu）的多次循环,对于长期持续的能量供应是有利的。FLWR是一种沸水堆型（BWR）反应堆,其堆芯设计特点为：堆芯呈短粗状,装载以三角形的栅格排列的燃料棒组成的六角形燃料组件,装有高富集度的混合金属氧化物（MOX）燃料和Y形控制棒。堆芯在两个阶段中使用一致的和相同尺寸的燃料组件,因此在反应堆运行寿期内,在同一个反应堆系统中,前一个反应堆堆芯概念设计可以过渡到后一个堆芯概念设计,这样就可以灵活地响应天然铀资源未来情况的预期变化,或建立金属氧化物乏燃料的经济的后处理技术。 完成了堆芯设计的详细研究,结合其他有关的研究,迄今为止所获得的结果已经表明所提出的这种反应堆概念设计是可行并具有发展前景的。     

MOX燃料组件装入现役M310堆芯对堆芯核设计的影响研究

国际上的MOX燃料技术目前已较为成熟,且已有在压水堆中运行的工程经验。本文对MOX燃料组件的中子学性能进行了分析,对其在我国现役M310堆芯应用的可行性进行了研究,得到了M310堆芯由全部使用UO2燃料组件向使用30%的MOX燃料组件过渡的堆芯燃料管理方案,并对使用MOX燃料组件的堆芯的部分中子学参数进行了初步分析。结果表明:使用30%的MOX燃料组件的堆芯可达到与全UO2堆芯相当的循环长度;堆芯反应性控制能力可满足要求;慢化剂温度系数、Doppler温度系数、Doppler功率系数、氙和钐的动态特性均趋向使堆芯运行更加安全和稳定。本文的研究结果可为MOX燃料在M310堆芯中应用的进一步研究提供参考。     

西德准备将其全部钚返回轻水堆

【美国《核燃料》1987年第12卷第13期第4页报道】西德准备将瓦克斯多夫后处理厂回收的钚全部返回轻水堆。从1987至1995年,使用西德钚制造的1164个混合氧化物燃料组件(MOX)将装入11座反应堆,其中包括3座沸水堆。这就需要400吨重金属,即今后9年平均每年45吨重金属。MOX组件中钚含量主要取决于堆芯设计,像格拉芬菜因费尔德这样的压水堆,其MOX组件含重金属300公斤,其中大部分是钚,16个组     

基于混合自适应遗传退火算法的AHTR堆芯优化研究

板状先进高温堆（AHTR）的预设计采用均一富集度的燃料组件，导致功率峰因子（PPF）过大，总PPF高达2.09，一定程度制约了反应堆的安全性与经济性。文章采用富集度分区法对其进行改进优化，为了加快堆芯燃料最优化布置的搜索速度，设计了一种自适应的混合智能算法，该算法整个优化过程均基于一个用MATLAB语言编辑的程序自动完成，优化后的径向功率峰因子降低至1.122，相比原设计降低25.02%。温度场模拟结果表明，优化方案温度分布更均匀，峰值温度从1030 K降低至1010 K，有效地提高了堆芯的安全裕量。      

Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated.Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin.The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core.Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities.The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B₄C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution. The temperature reactivity coefficients of the TOX core were found to be always negative. The TOX core has a slightly reduced, as compared to UOX core, but still sufficient shutdown margin.In the TOX core β_eff is smaller by about a factor of two in comparison to the UOX core and even lower than that of the MOX core. The combination of small β_eff and reduced control materials worth may potentially deteriorate the performance under RIA conditions and requires an additional examination. The behavior of the considered cores during the most limiting RIAs, such as rod ejection, main steam line break, and boron dilution, is further investigated and reported in Part II of the paper.     

Approach to equilibrium fuelling scheme of 500 MWe PFBR based on 3-D core burnup modeling

Approach to equilibrium fuelling scheme of 500 MWe prototype fast breeder reactor (PFBR) has been predicted using detailed 3-D core burnup modeling. Equilibrium is reached after two cycles of 180 effective full power days (efpd) each. One-third core is refueled every time in a repeatable scatter load scheme after every 3 cycles. Considering the constraints of linear heat rating (LHR) on fuel and blanket pins it is found that the nominal core achieves full power only in mid-cycle. A novel interpolation scheme is used to find the peak LHR in any axial section of a fuel/blanket sub-assembly. Breeding ratio is adequate for self-sufficient Pu generation in a closed fuel cycle with Pu from axial blankets and two rings of radial blanket sub-assemblies.     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

Effect of Pu-rich agglomerates in MOX fuel on reactivity analysis of light water reactor MOX core physics experiments

The effect of Pu-rich agglomerates in U-Pu mixed oxide (MOX) on the reactivity analysis of light water reactor MOX core physics experiments was studied with the continuous-energy Monte Carlo calculation code MVP II. First, the following three different models were compared in the analysis of a representative unit cell of a MOX core tested at the KRITZ reactor: a Lattice model where Pu-rich agglomerates were assumed to exist in a fixed pitch, a statistical geometry (STG) model of MVP II, and a Random model where the random distribution of Pu-rich agglomerates was directly modeled. Since the three models gave comparable results, the STG model was used in parametric calculations to systematically understand the reactivity effect depending on the characteristics of Pu-rich agglomerates. In addition, the selected unit cells composing the MOX cores and one representing MOX core tested at the EOLE criticality facility were analyzed with the measured characteristics of Pu-rich agglomerates in MOX fuel. Consequently, the reactivity differences between the calculations assuming the homogeneous Pu distributions and those considering Pu-rich agglomerates were less than 0.0005 Δk/k/k', indicating that the effect of Pu-rich agglomerates was small on the reactivity analysis of the MOX cores tested in the EOLE facility.     

Concept of innovative water reactor for flexible fuel cycle (FLWR)

In order to ensure sustainable energy supply in the future based on the matured light water reactor (LWR) and coming mixed oxide (MOX)-LWR technologies, a concept of innovative water reactor for flexible fuel cycle (FLWR) has been investigated in Japan Atomic Energy Research Agency (JAEA). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming MOX-LWR technologies without significant technical gaps. The second part represents the reduced-moderation water reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-developed LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the future fuel cycle circumstances during the reactor operation period around 60 years.At present, reprocessed plutonium from the LWR spent fuel is to be utilized in MOX-LWR. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or MOX-LWR. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the MOX-LWR utilization. When the fuel cycle for plutonium multiple recycling with MOX fuel reprocessing is realized, the fuel assembly will be replaced with another type of the tight-lattice one for RMWR with different rod diameter, rod gap width and so forth even in the same reactor system, being flexibly corresponding to the fuel cycle circumstances.Investigation on the core for both the parts of the FLWR concepts has been performed, including the core conceptual design, the core characteristics under Pu multiple recycling, the thermal hydraulic investigation in the tight-lattice core, and so forth. Up to the present, promising results have been obtained.     

Symbiosis system for transmutation on nitride fuel FBR and ADS

A conceptual scheme for mass flow of transmuting Plutonium (Pu), minor actinides (MA) and long-lived fission products (LLFP) is studied. In this feature, the existing light-water reactors (LWRs) cycle will be main stream for nuclear electric generation during a long-term period more than 50 years, and Pu will be utilized in mixed oxide fuel (MOX)-LWRs. In future, when Pu recycling system will be achived by introducing high-conversion LWRs (HCLWRs) and/or fast breeder reactors (FBRs), the accelerator driven transmutation system (ADS) transmutes Pu, MA and Iodine from Purex or Dry reprocessing. This is due to reduce burden for transmuting the excess or remained Pu, MA and LLFP by commercial reactor plants in Pu-recycling system. For this purpose, we introduce a concept of symbiosis system for transmutation based on nitride fuel FBR and ADS. The core design for lead-bismuth (Pb-Bi) cooled FBRs and ADS, Pb-Bi technologies, 15N enrichment and 14C toxicity are studied. And the mass flows for MA and Iodine are discussed based on an estimated scenario for nuclear electric plants introduction in future.     

Feasibility study of large MOX fueled FBR core and applicability of various coolants and fuels aimed at the Self-Consistent Nuclear Energy System

The potential of a large MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long -lived fission products (LLFPs), Se-79, Tc-99, Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and part of a lower axial blanket region without any significant impact on the reactor's nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and a sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, re-criticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. The fundamental applicability of various coolants and fuels is evaluated based on neutron balance toward the final goal of the ideal SCNES. The results show that gas coolant has a potential for increasing the transmutation efficiency of LLFPs. And an improved SCNES with several conventional FBRs and a FP transmutation reactor is also studied.