本文選題：乏燃料貯存 + 中子吸收性能�。� 參考：《太原理工大學(xué)》2015年碩士論文

【摘要】：隨著我國核電產(chǎn)業(yè)的快速發(fā)展，核電站乏燃料的持續(xù)安全貯存成為核電技術(shù)自主化過程中亟待解決的關(guān)鍵性問題。乏燃料貯存用輻射屏蔽材料需同時滿足較高的結(jié)構(gòu)性能、腐蝕性能和屏蔽性能要求。本文基于MonteCarlo方法，以6061鋁合金為載體，通過添加碳化硼作為中子吸收組元、鉛作為γ輻射屏蔽組元，為設(shè)計(jì)結(jié)構(gòu)/功能一體化的新型鋁基乏燃料貯存用輻射屏蔽材料提供理論依據(jù)。 本文采用MCNP5.0和Super MC/MCAM軟件，研究了碳化硼顆粒增強(qiáng)鋁基復(fù)合材料的組分配比、厚度、增強(qiáng)相尺寸與中子透射系數(shù)的關(guān)系，分析了鋁基碳化硼材料與中子的相互作用機(jī)制；研究了鉛-碳化硼鋁基復(fù)合材料的組分配比、射線能量與中子透射系數(shù)、次級γ射線和衰變γ射線吸收率的關(guān)系；研究了貯存乏燃料的硼酸溶液硼濃度和乏燃料組件密集化程度對輻射屏蔽性能造成的影響；分析了在乏燃料中235U衰變條件下，鋁基輻射屏蔽材料的屏蔽效果。 結(jié)果表明：由于中子吸收核素10B的存在，B4C含量對中子吸收性能影響很大、使次級γ射線吸收率微弱降低；由于其密度較小使材料整體密度降低，會導(dǎo)致一次γ射線的吸收率略有下降；隨著微米級B4C顆粒存在使中子透射系數(shù)提高11%以上；B4C顆粒形狀改變引起的中子在材料內(nèi)穿行距離的變化也會對中子吸收性能造成影響。 由于γ射線吸收核素Pb具有較大的原子序數(shù)、較高γ射線的質(zhì)量吸收系數(shù)和高密度，其含量提高會使鋁基屏蔽材料的中子吸收性能有所提高、對次級γ射線與的吸收能力均顯著增強(qiáng)。2cm厚、含鉛量為25%的鋁基復(fù)合材料的一次γ射線屏蔽能力與0.5cm鉛板相當(dāng)。 入射粒子的能量直接決定著材料的輻射屏蔽效果，這是由于材料的宏觀吸收截面通常隨射線能量增大而減小。當(dāng)入射中子的能量低于200eV時，中子吸收性能與次級γ射線屏蔽性能隨能量減小而顯著提高；當(dāng)一次γ射線的能量低于660KeV時，對其的屏蔽性能隨能量減小而顯著提高。 由于乏燃料貯存介質(zhì)硼酸溶液對入射高能中子先“慢化”后“熱化”的作用，，鋁基屏蔽材料的中子吸收性能隨硼原子濃度提高而降低、隨組件柵距增大而減小；次級γ射線屏蔽性能隨硼原子濃度提高而降低、隨組件柵距增大而減�。毁A存介質(zhì)變化對材料吸收乏燃料衰變產(chǎn)生的γ射線的吸收影響不大。在乏燃料組件密集化貯存（組件柵距23cm）、屏蔽層厚度為0.7cm的條件下，乏燃料235U衰變中子經(jīng)硼濃度2500ppm的硼酸溶液環(huán)境充分慢化、熱化后，B4C wt.%為30%的B4C/Al復(fù)合材料作為屏蔽層時，可以吸收98.04%的中子、45.44%的次級γ射線和20.21%的一次γ射線；B4Cwt.%為30%、Pb wt.%為25%的Pb-B4C/Al復(fù)合材料作為屏蔽層時，可以吸收98.82%的中子、61.05%的次級γ射線和47.08%的一次γ射線。
[Abstract]:With the rapid development of nuclear power industry in China, the continuous and safe storage of spent fuel in nuclear power plant has become a key problem to be solved urgently in the process of nuclear power technology autonomy. The radiation shielding material for spent fuel storage needs to meet high structural performance, corrosion performance and shielding performance. Based on the MonteCarlo method, 6061 aluminum alloy is used in this paper. As a carrier, by adding boron carbide as a neutron absorption component and lead as a shielding element of gamma radiation, it provides a theoretical basis for the design of a new type of aluminum based spent fuel storage shielding materials for the design of structural / functional integration.
In this paper, MCNP5.0 and Super MC/MCAM software were used to study the relationship between the group distribution ratio, thickness, the thickness, the phase size of the reinforced aluminum matrix composite and the neutron transmission coefficient, and the interaction mechanism of the aluminum based boron carbide material with the neutron, and the group distribution ratio, the ray energy and the neutron energy of the lead carbon boric aluminum matrix composites were studied. The relationship between the transmission coefficient, the secondary gamma ray and the absorption rate of decay gamma ray, the effect of boron concentration in the boric acid solution and the intensity of the spent fuel assembly on the radiation shielding performance of the spent fuel were studied, and the shielding effect of the aluminum based radiation shielding material under the 235U decay condition in the spent fuel was analyzed.
The results show that because of the existence of the neutron absorption nuclide 10B, the B4C content has a great influence on the absorption properties of the neutron, which makes the secondary gamma ray absorptivity weakly reduced. Because of its smaller density, the overall density of the material decreases, which will lead to a slight decrease in the absorption rate of the primary gamma ray, and the neutron transmission coefficient increases by 11% with the existence of the micron grade B4C particles. On the other hand, the variation of the neutron penetration distance caused by the change of B4C particle shape will also affect the neutron absorption performance.
Because of the large atomic number and high mass absorption coefficient and high density of gamma ray absorption nuclide Pb, the neutron absorption properties of the aluminum based shielding material will be improved and the absorption capacity of the secondary gamma ray and the absorptive capacity of the secondary gamma ray increases significantly, and the primary gamma ray shielding energy of the aluminum matrix composites with 25% lead content is 25%. The force is equivalent to the 0.5cm lead plate.
The energy of the incident particle is directly determined by the radiation shielding effect of the material, because the macroscopic absorption cross section of the material decreases with the increase of the ray energy. When the energy of the incident neutron is lower than 200eV, the absorption and secondary gamma ray shielding performance of the neutron is significantly increased with the decrease of energy; when the energy of the primary gamma ray is lower than 660KeV The shielding performance of the device increases with the decrease of energy.
The neutron absorption performance of the aluminum based shielding material decreases with the increase of boron concentration and decreases with the increase of the component spacing, and the secondary gamma ray shielding performance decreases with the increase of boron concentration, and decreases with the increase of the component spacing. The change of storage medium has little effect on the absorption of gamma ray produced by the absorption of spent fuel decay. Under the condition of the dense storage of the spent fuel assembly (component distance 23cm) and the thickness of the shielding layer of 0.7cm, the spent fuel 235U decay neutrons are fully slowed by the boric acid solution environment of boron concentration 2500ppm, and the B4C wt.% is 30% of the B4C/Al composite after the heating. As a shielding layer, 98.04% neutrons, 45.44% secondary gamma rays and 20.21% gamma rays can be absorbed; when B4Cwt.% is 30% and Pb wt.% is 25% Pb-B4C/Al composite material as shielding layer, it can absorb 98.82% neutrons, 61.05% secondary gamma rays and 47.08% gamma rays.
【學(xué)位授予單位】：太原理工大學(xué)
【學(xué)位級別】：碩士
【學(xué)位授予年份】：2015
【分類號】：TB333

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