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为评估压水堆核电厂燃料包壳破损时的工作人员辐射风险和燃料包壳破损程度,基于特征物理量建立一回路冷却剂系统中锕系核素质量评估方法。本文基于锕系核素的生成和迁移机理,建立了一回路冷却剂系统中锕系核素的平衡方程组,并选取3种易监测的特征物理量用以评估锕系核素向一回路冷却剂系统的释放量及其分布,并建立了一回路冷却剂系统中锕系核素质量的评估方法。然后分别采用国内在役压水堆核电厂无燃料包壳破损和有燃料包壳破损的实测数据对建立的评估方法进行了验证,验证结果表明:建立的评估方法可在无燃料包壳破损和有燃料包壳破损的情况下对一回路冷却剂系统中锕系核素质量进行评估,评估结果和预期符合。本文研究成果可为压水堆核电厂运行期间一回路冷却剂系统中锕系核素质量及其分布评估提供指导,从而优化后端的工作人员防护措施,降低辐射风险。 相似文献
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新型可燃毒物设计与现代堆芯燃料管理 总被引:3,自引:2,他引:1
阐述了现代压水堆堆芯燃料管理对可燃毒物的新要求,评价了几种新型可燃毒物的设计特点,讨论了以秦山一期为代表的我国压水堆核电厂应用新型可燃毒物改进堆芯燃料管理的途径。 相似文献
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国外一些有核电的国家 ,如美、法、德、日等国 ,都在积极地开展深燃耗燃料元件的研究工作 ,其目标是延长燃料元件使用寿命、提高铀燃料利用率、降低燃料循环成本、降低核电成本。在开展此项研究试验工作中遴选出能经受物理、化学、机械、辐照、裂变气体等多方面作用的燃料元件包壳材料是十分重要的。要求在达到预定的深燃耗目标时 ,元件包壳材料不会发生破损。当前压水堆核电站燃料的燃耗设计值通常是 330 0 0兆瓦日 /吨铀。如果燃料在堆内使用时间由 1 2个月延长到 1 8个月 ,其燃耗提高到 4 5 0 0 0兆瓦日 /吨铀以上 ,那么燃料元件使用的锆… 相似文献
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为研究压水堆完整和破损燃料棒燃料包壳化学相互作用(FCCI)层物相结构组成及影响因素,通过拉曼光谱对燃耗为45 GW·d/tU和41 GW·d/tU的完整和破损燃料棒FCCI层进行了研究分析。结果表明:完整燃料棒形成了周向厚度为14~19μm的FCCI层,主要由两个不同相结构区域组成,分别为靠近包壳界面的单斜和四方相氧化锆混合区域及靠近燃料芯块的四方相区域,在包壳界面附近约7μm范围内,观察到明显的705 cm-1特征峰,该峰的形成源于界面压应力和辐照缺陷的共同作用;破损燃料棒形成了周向厚度为37~61μm的FCCI层,主要由两个不同形貌和相结构区域组成,即靠近包壳界面附近具有多孔、裂纹特征的单斜相氧化锆区域以及靠近燃料芯块的非晶结构区域。对FCCI层相结构的分布及转变影响因素进行了分析讨论,完整燃料棒FCCI层中四方相氧化锆的稳定与界面压应力、中子辐照缺陷和裂变产物作用有关,破损燃料棒FCCI层中单斜相氧化锆的存在则主要来源于应力的弛豫和氧的正常化学计量比。 相似文献
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Takashi Shinozaki Yutaka Udagawa Tomoyuki Sugiyama Masaki Amaya 《Journal of Nuclear Science and Technology》2016,53(9):1426-1434
In order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency. The specimens with an outer surface pre-crack were prepared by using Rolling-After-Grooving (RAG) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain (?tz) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain (?t?) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. From these tests, the data of cladding failure were obtained in the range of strain ratio (?tz/?t?) between about ?0.6 and 0.7: this range of strain ratio covers the range between about 0.0 and 0.7 which is estimated in the case of RIA-simulated test. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state. 相似文献
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风险指引的安全裕度是近十年来核工业界提出的新的安全理念。本文阐述了基于离散动态事件树的风险指引的安全裕度分析方法,给出该方法下核燃料包壳失效概率均值和标准差的数学表达式。针对简化压水堆模型下的全厂断电事故,提出了基于离散动态事件树的风险指引的安全裕度计算流程,计算了两种离散动态事件树分支规则下燃料包壳失效的风险指引的安全裕度及其不确定性。计算结果表明,不同的分支规则、模型参数分布、系统程序最大时间步长对核燃料包壳失效概率均值和标准差均有显著影响。提出了一种改进的可变概率阈值的分支方法,以更好地平衡风险指引的安全裕度分析过程中计算精度与计算资源的匹配问题。 相似文献
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Terada H. Wakayama N. Obata M. Ogawa T. Kobayashi F. Fukuda K. Oyama N. Sakuraba K. Yokouchi I. Yoshida H. Suzuki K. Asami S. 《IEEE transactions on nuclear science》1988,35(3):1041-1045
For pt.I see ibid., vol.34, p.567 (1987). A fuel failure detection (FFD) method based on selective detection of short-life and gaseous fission products, developed for a high-temperature gas-cooled reactor, is described. An improved precipitator was used as a detector for the fission products and the performance of the FFD system was tested using an irradiation rig at the Japan Material Testing Reactor. In the rig, three kinds of samples of coated-particle fuels were irradiated and each sample of the primary helium gas was fed to the FFD system. Failure rates of the three fuel samples called intact, normal, and slightly failed, were estimated at about 10-6, 10-5, and 10 -4, respectively. The FFD system showed a significantly increased response in counting rate for the sample gas with a failure rate of 10-4. The FFD system did not respond to the sample gas with the smaller failure rate of 10-5 even when the background level of long-life fission products in the primary coolant gas increased with fuel temperature and reactor power 相似文献
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《Journal of Nuclear Science and Technology》2013,50(11):991-997
In order to clarify the failure mechanism and determine the failure limit of the High-Temperature Gascooled Reactor (HTGR) fuel under reactivity-initiated accident (RIA) conditions, pulse irradiations were performed with unirradiated coated fuel particles at the Nuclear Safety Research Reactor (NSRR). The energy deposition ranged from 0.578 to 1.869 kJ/gUO2, in the pulse irradiations and the estimated peak temperature at the center of the fuel particle ranged from 1,510 to 3,950 K. Detailed examinations after the pulse irradiations showed that the coated fuel particles failed above 1.40 kJ/gUO2, where the peak fuel temperature reached over the melting point of UO2 fuel. It was concluded that the coated fuel particle was failed by the mechanical interaction between the melted and swelled fuel kernel and the coating layer under RIA conditions. 相似文献
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在压水堆核电站乏燃料元件检验中,完成了4根完整元件棒、4根破损元件棒的γ扫描测量,元件燃耗分布在9600~45000 MW•d/t(U)之间,获得了完整元件轴向相对燃耗分布、破损元件137Cs分布及迁移流失情况。结果显示,破损元件均存在不同程度的Cs迁移流失,破口处存在137Cs计数突变(降低)。破损元件134Cs/137Cs原子比分布与相邻完整元件基本一致,表明134Cs、137Cs流失比例近似相等,可用134Cs/137Cs原子比表征其相对燃耗分布;破口处可通过低挥发性核素154Eu计数水平判断燃料芯块是否缺失。检验结果可为燃料元件破损原因分析及堆内行为分析提供重要依据。 相似文献
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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. 相似文献
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W. Marshall J. Wagner 《Packaging, Transport, Storage and Security of Radioactive Material》2014,25(1):1-7
Commercial used nuclear fuel (UNF) in the USA is expected to remain in storage for periods potentially greater than 40 years. Extended storage (ES) time and irradiation to high burnup values (>45 GWd t?1) may increase the potential for fuel failure during normal and accident conditions involving storage and transportation. Fuel failure, depending on the severity, could result in changes to the geometric configuration of the fuel, which has safety and regulatory implications. The likelihood and extent of fuel reconfiguration and its impact on the safety of the UNF are not well understood. The objective of this work is to assess and quantify the impact of fuel reconfiguration due to fuel failure on the criticality safety of UNF in storage and transportation casks. Criticality analyses are conducted considering representative UNF designs covering a range of enrichments and burnups in multiple cask systems. Prior work developed a set of failed fuel configuration categories, and specific configurations were evaluated to understand trends and quantify the consequences of worst case potential reconfiguration progressions. These results are summarised here and indicate that the potential impacts on subcriticality can be rather significant for certain configurations (e.g. >20% Δkeff). However, for credible fuel failure configurations from ES or transportation following ES, the consequences are judged to be manageable (e.g. <5% Δkeff). The current work expands on the previous efforts by including part length rods in fresh boiling water reactor fuel assemblies and studying the effect of damage in varying numbers of fuel assemblies. 相似文献
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Abderrafi M. Ougouag Jan Leen Kloosterman Wilfred F.G. van Rooijen Hans D. Gougar William K. Terry 《Nuclear Engineering and Design》2006,236(5-6):669-2004
Models and methods are presented for determining practical limits of the packing density of TRISO particles in fuel pebbles for a pebble-bed reactor (PBR). These models are devised for designing and interpreting fuel testing experiments. Two processes for particle failure are accounted for: failure of touching particles at the pressing stage in the pebble manufacturing process and failure due to inner pressure buildup during irradiation. The second process gains importance with increasing fuel temperature, which limits the particle packing density and the corresponding fuel enrichment. Suggestions for improvements to the models are presented. 相似文献
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《Annals of Nuclear Energy》2005,32(5):479-492
We have developed a method for detecting and diagnosing a disk wear failure and a foreign object failure among the various failure modes of check valves. The method is based on the acoustic emission sensors which can detect the sound wave of the leakage flow and the estimation of the power spectral densities with an auto-regressive model. For validating the method, we implemented a hydraulic test loop with an artificially failed check valve. We have found that the frequency spectrums from the acoustic signals are strongly dependent on the failure modes of the check valve and that they are nearly independent of the failure size and operating pressure through an estimation of the power spectral density with an auto-regressive signal processing model. In addition, the root mean square values of the acoustic signal and the amplitudes of the power spectral density as well as the loop pressure have a strong dependency on the failure size in each failure mode of the check valve. We developed a diagnosis algorithm by using neural network models in order to identify the type and size of the failure in the check valve. The diagnosis algorithm consists of a hierarchical model composed of three back-propagation neural networks. The results of our research and the experiments show that the diagnosis algorithm is proven to be a good solution for identifying the failures of the check valves without any disassembling work. 相似文献