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熔盐堆作为第四代先进核能系统,具有中子经济性好、固有安全性高、在线换料、燃烧钍燃料等优点。本文针对熔盐快堆和熔盐热堆,采用MCNP5计算了熔盐堆中常用的9种燃料盐寿期初的临界性能和转换系数,并用中子平衡方法分析了影响转换系数的因素。从寿期初的计算数据分析,由于233U具有较高的平均裂变中子数及较小的中子俘获截面,有利于提高反应堆增殖系数和燃料利用率。另外,熔盐中的23Na相对于7 Li中子俘获截面更大,导致含23Na燃料盐增殖系数相对较低,但对热堆的影响较小;而在快堆中,熔盐中采用Na元素相比采用Li元素更有利于中子能谱硬化,更适合快堆的增殖。 相似文献
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压水堆内钍-铀增殖循环研究——堆芯设计 总被引:1,自引:1,他引:0
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。 相似文献
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熔盐快堆是当前国际上关注的热点之一,本文基于堆芯结构双流体方案,即裂变熔盐燃料和增殖熔盐介质各自独立冷却循环,利用氟化或氯化熔盐中钍铀重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。通过比较钍铀燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+ThF_4+UF_4、NaF+ThF_4+UF_4和NaCl+ThCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数,分析了增殖比BR(breeding ratio)受反应堆裂变区、增殖区和ZrC中子反射层的尺寸影响、熔盐中~6Li和~(35)Cl同位素丰度的影响,以及熔盐密度误差对BR计算值的准确性影响、易裂变核素随反应堆运行时间演化等。在钍铀燃料循环熔盐快堆中,通过优化处理得到三种熔盐燃料方案的增殖比BR约为1.2。 相似文献
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气冷快堆兼具高温气冷堆的经济性和快堆的可持续性等优点,在四代堆型中具有独特的技术优势。为了适应气冷快堆高温、高中子通量的堆芯环境,本文基于耐事故燃料模型,提出了一种块状气冷快堆燃料组件设计方案,并对该组件中铀钚混合燃料中的钚含量、冷却孔道的直径及数量、栅距比、包壳及组件盒厚度等物理参数对中子学特性的影响规律开展了敏感性分析研究。分析结果表明:在研究的6个参数中,钚含量和栅距比对组件的中子学特性影响最大,冷却孔道数量主要影响组件内的功率分布,其余参数对组件中子学特性几乎无影响。最后针对块状燃料组件低冷却剂份额的特点,利用单通道模型进行组件内的温度分布计算,给出了热工限值对组件参数的要求。 相似文献
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武器级钚(WGPu)与反应堆级钚(RGPu)可以分别从废旧拆除的核武器中以及轻水堆乏燃料中获得,二者均可以作为钍基燃料的驱动燃料。为对上述2种驱动燃料特性进行研究,利用DRAGONV4程序以及JEFF3.11-295群截面库进行反应堆物理计算。采用修正4因子公式对WGPu与RGPu驱动条件下的SB 6+3型组件初始无限增殖系数进行分析。同时,为确定WGPu与RGPu增殖性能最优的空间分离尺度和钍含量,进一步对比了不同空间分离尺度的SB型组件和MOX组件寿期末的233U质量。结果表明,钍含量相同时,WGPu具有较高的热中子裂变系数,导致其初始无限增殖系数和燃耗深度均大于RGPu,并且不随钍含量的大小而改变。RGPu作为驱动燃料的SB5+4-70%Th组件具有最优增殖性能。WGPu作为驱动燃料时,MOX型组件233U质量大于SB型组件,并在70%钍含量时达到最大值。 相似文献
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The use of thorium in pressurized water reactor fuel assemblies is investigated in this paper. The novelty of the reported work is to study a fuel design primarily intended to control the excess of reactivity at beginning of life, and flatten the intra-assembly power distribution rather than converting fertile Th-232 into fissile U-233. The fuel assembly is a traditional 17 × 17 pressurized water reactor fuel design. The majority of the fuel pins contain a mixture of uranium and thorium oxides, while a few fuel pins contain a mixture between uranium and gadolinium oxides. The calculation were performed by two-dimensional transport calculations with the Studsvik Scandpower CASMO-4E code in order to determine the main neutronic properties of the new fuel design, compared with the traditional uranium-based fuel assembly containing gadolinium used as reference. The majority of the neutronic properties of the uranium-thorium-based fuel assembly were similar to the reference fuel assembly. The Doppler and the moderator temperature coefficients of reactivity were found to be appreciably more negative in the uranium-thorium-based design, but still within acceptable limits. One advantage of this new uranium-thorium-based design is a reduction of the pin peak power at beginning of life, because of smaller amount of gadolinium being used. This is important from an operational and safety viewpoint, since the margin to departure from nucleate boiling becomes larger. Consequently, this new type of thorium-based fuel assembly shows advantageous properties for use in power-uprated cores. 相似文献
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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。 相似文献
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Klara Insulander Björk Valentin Fhager Christophe Demazière 《Progress in Nuclear Energy》2011,53(6):618-625
With the aim of investigating the technical feasibility of fuelling a conventional BWR (Boiling Water Reactor) with thorium-based fuel, computer simulations were carried out in a 2D infinite lattice model using CASMO-5. Four different fissile components were each homogenously combined with thorium to form mixed oxide pellets: Uranium enriched to 20% U-235 (LEU), plutonium recovered from spent LWR fuel (RGPu), pure U-233 and a mixture of RGPu and uranium recovered from spent thorium-based fuel. Based on these fuel types, four BWR nuclear fuel assembly designs were formed, using a conventional assembly geometry (GE14-N). The fissile content was chosen to give a total energy release equivalent to that of a UOX fuel bundle reaching a discharge burnup of about 55 MWd/kgHM. The radial distribution of fissile material was optimized to achieve low bundle internal radial power peaking. Reactor physical parameters were computed, and the results were compared to those of reference LEU and MOX bundle designs. It was concluded that a viable thorium-based BWR nuclear fuel assembly design, based on any of the fissile components, can be achieved. Neutronic parameters that are essential for reactor safety, like reactivity coefficients and control rod worths, are in most cases similar to those of LEU and MOX fuel. This is also true for the decay heat produced in irradiated fuel. However when Th is mixed with U-233, the void coefficient (calculated in 2D) can be positive under some conditions. It was concluded that it is very difficult to make savings of natural uranium by mixing LEU (20% U-235) homogenously with thorium and that mixing RGPu with thorium leads to more efficient consumption of Pu compared to MOX fuel. 相似文献
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The possibility of utilizing thorium as a fuel in a pressurized water reactor(PWR)has been proven from the neutronic perspective in our previously published work without assessing the thermal hydraulic(TH)and solid structure performances.Therefore,the TH and solid structure performances must be studied to confirm these results and ensure the possibility of using a thorium-based fuel as an excellent accident-tolerant fuel.The TH and solid structure performances of thorium-based fuels were investigated and compared with those of U02.The radial and axial power peaking factors(PPFs)for U02,(232Th,235U)02,and(232Th,233U)02 were examined with a PWR assembly to determine the total PPF of each one.Both Gd203 and Er203 were tested as burnable absorbers(BAs)to manage the excess reactivity at the beginning of the fuel cycle(BOC)and reduce the total PPF.Er203 resulted in a more significant reduction to the total PPF and,therefore,a greater reduction to the temperature distribution compared to Gd203.Given these results,we analyzed the effects of adding Er203 to thorium-based fuels on their TH and solid structure performances. 相似文献
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《核技术(英文版)》2016,(4):207-213
Fertile fuel, such as thorium or depleted uranium, can be bred into fissile fuel and burnt in a breed-andburn(BB) reactor. Modeling a full core with fertile fuel can assess the performance of a BB reactor with exact quantitative estimates, but costs too much computation time. For simplicity, performing the recently developed neutron balance method with a zero-dimensional(0-D)model can also provide a reasonable result. Based on the0-D model, the feasibility of the BB mode for thorium fuel in a fast reactor cooled by sodium was investigated by considering the(n, 2n) and(n, 3n) reaction rates of fuel and coolant in this work, and compared with that of depleted uranium fuel. Afterward, the performance of the same thorium-based fuel core, but cooled by helium, lead-bismuth, and FLi Be, respectively, is discussed. It is found that the(n, 2n) and(n, 3n) reactions should not be neglected for the neutron balance calculation for thorium-based fuel to sustain the BB mode of operation. 相似文献
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氢化锆慢化熔盐堆钍铀转换性能初步分析 总被引:3,自引:0,他引:3
中子能谱对钍基燃料在熔盐堆中的利用效率及温度反馈系数等安全问题有较大影响,所以对熔盐堆新型慢化剂的研究具有重要意义。本工作基于SCALE6计算程序,对不同几何栅元结构的氢化锆栅元组件在熔盐堆的物理性能进行了研究,分别计算了中子能谱、钍铀转换比、~(233)U浓度、总温度反馈系数以及燃耗等中子物理参量。结果表明,减小六边形栅元对边距或者增加熔盐占栅元体积比可以增加钍铀转换比和改善温度反应性系数;当加入的氢化锆慢化剂体积份额为0.1时就可以将熔盐堆~(233)U初始浓度降低到2.5×10~(-2)以内;氢化锆慢化熔盐堆在超热谱条件下,其~(233)U初装载量和超铀核素产量较小,同时堆芯较为紧凑。 相似文献
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钍资源及其利用 总被引:4,自引:1,他引:4
钍是一种赋存在自然界中的天然放射性元素,在地壳中比铀更丰富,其丰度约为铀的3~4倍。广泛分布在各种不同的地质环境中。世界各国现已查明可经济回收的钍资源量达数百万吨。钍可广泛应用于光学、无线电、航空、航天、冶金、化工、材料等领域,更重要的是它可用作核燃料。随着核电发展对铀需求的不断增加,钍基燃料循环的研发工作业已引起广泛关注,通过大量的研究证实,钍在核能方面的应用具有广阔的前景,未来可有效地补充铀资源的不足。结合钍的物理、化学性质,以及近年世界各国对钍基燃料循环的研发成果,简要介绍世界钍资源的分布、钍资源量、钍资源的地质类型和产出地质背景,以及钍在核能中的应用潜力。 相似文献