共查询到19条相似文献,搜索用时 953 毫秒
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应用图论原理,建立了一种基于数据库的多分支递次衰变计算方法,编制了计算机程序,用于计算复杂衰变链中各子体放射性强度随时间的变化。设计的递归算法可对衰变路径进行自动搜索和存储,并对路径重复部分进行适当标记,实现了衰变链的程序化计算。 相似文献
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利用放射性随时间衰变的规律,编制了239Pu裂变产物不同时刻γ放射性强度计算程序。该程序在衰变路径的处理上采用图的深度优先遍历计算方法,可以有效地计算停止辐照后不同时刻的裂变产物能量强度谱,计算任意时刻的放射性总强度,实现了裂变产物衰变链计算的程序化。 相似文献
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《核科学与工程》2017,(6)
主控室是对核电厂正常运行和事故状态实施控制的场所,为了保护控制室内操作人员,法规要求对控制室进行可居留性分析。在一般计算模型中,为了简化模型,没有考虑衰变子核对于人员剂量的贡献。本文中,针对通用主控室模型在考虑了衰变链对人员剂量的影响的基础上,建立了核素平衡方程,并开发了主控室可居留性剂量评价程序CROSS进行计算。基于RG1.183规定的冷却剂丧失事故(LOCA)源项,使用CROSS程序分别在非放射性通风系统新风模式和应急可居留系统通风模式下对主控室可居留性进行了分析。计算中对比了是否考虑衰变子核对个人剂量的影响。计算结果表明,考虑衰变链后,对于非放射性通风系统新风模式:个人有效剂量增到了1.28倍,其中主要增加为内照射有效剂量,增到了1.64倍;对于应急可居留系统通风模式:个人有效剂量增到了1.27倍,其中主要增加为内照射有效剂量,增到了1.30倍。在核电厂事故工况下计算主控室人员剂量时,需要考虑衰变链对个人有效剂量的影响。 相似文献
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应用混合堆放射性计算程序FDKR和衰变链数据库AF-DCDLIB,计算了托卡马克实 验混合堆FEB(Fusion Experimental Beeder)概念设计中活化产物、裂变产物和锕系元素的放射性、衰变余热和潜在生物危害因子BHP值。计算的结果表明,对于FEB设计来说,在150MW聚变功率下运行一年,停堆时刻的总放射性、余热和BHP值分别为5.74×10~(20)Bq 相似文献
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《核科学与工程》2017,(6)
PWR-GALE是美国核管会编制并使用的压水堆核电厂气液态流出物源项计算程序,现有的配套核数据库已有长达四十年之久未进行更新,无法确定是否能够满足先进压水堆的计算和审评需求,需要通过基于最新版本的核评价数据库制作新的配套数据库对其进行适用性的评价。因此,本文基于核评价数据库ENDF/B-VII.0,提取衰变子库中相关信息,根据直接裂变产额、衰变信息以及保留的裂变产物核素得到更新的沿衰变链归并的产额数据,通过中子学-燃耗耦合计算获得了更新的中子微观反应截面数据;并与现有的配套数据库进行了对比分析;然后,通过计算一系列面向不同机型的算例进行了整体的对比验证与分析。结果表明:现有的PWR-GALE配套核数据可以满足先进压水堆的计算和评审需求。 相似文献
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根据合成239Pa实验的特定条件,利用递次衰变关系首次确定了丰中子新核素239Pa的半衰期。计算结果检验并支持了实验中合成了239Pa的结论 相似文献
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反应堆停堆后的余热导出是反应堆的重要安全功能之一,停堆初期余热由裂变功率和衰变热构成,停堆后期余热主要取决于衰变热。本文开发了应用于钠冷快堆系统分析程序FR-Sdaso的衰变热计算模型,该模型可考虑裂变功率和功率历史的影响。通过与ANSI/ANS-5.1-2005标准和SAS4A/SASYS-1程序对比进行了模型验证。FR-Sdaso程序的计算结果与ANSI/ANS-5.1-2005标准的最大相对偏差约为0.1%,与SAS4A/SASYS-1的最大相对偏差在10-8量级,初步证明了所开发模型的正确性。最后,基于中国实验快堆的设计数据,分析了紧急停堆过程中裂变功率对衰变热的影响,结果表明,忽略裂变功率的影响导致衰变热的最大相对偏差约-7%,出现在停堆初期。因此,计算停堆初期衰变热时应考虑裂变功率的影响。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(10):727-734
Decay heat in the blanket and shield of the Fusion Experimental Reactor (FER) was calculated using a newly developed radioactivation calculation code system THIDA-2. The decay heat at various time periods after shutdown were calculated. The decay heat level in the FER blanket was found to be at least one order of magnitude lower than in fission reactors at all time periods after shutdown. The necessity of following the transport of decay γ-rays in obtaining the detailed distribution of decay heat in the blanket was demonstrated. The validity of the γ-ray kerma factors used in the evaluation was also shown. 相似文献
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In this study, a decay heat analysis was performed for prism type VHTR cores by combining Monte Carlo depletion calculation with McCARD code and the decay cooling calculation with ORIGEN-2 code. In the Monte Carlo depletion approach, the McCARD multi-cycle core depletion calculation was performed up to an equilibrium cycle, involving a great details of core geometry and material inventory. ORIGEN-2 performs only the decay cooling calculation with the full scope of ORIGEN-2 nuclides inventory provided by the McCARD depletion calculation. The accuracy of the decay heat analysis procedure developed in the previous work by using HELIOS and ORIGEN-2 codes was also verified. The HELIOS/ORIGEN-2 procedure showed a good accuracy for a short period of cooling time. However, a relatively large discrepancy between the two was observed for a long period of cooling time. As expected, the decay heat of a TRU fueled DB-MHR core was much higher than that of uranium fueled PMR200 core due to the fuel composition difference, which means that more attention for effective removal of the decay heat should be paid in designing the TRU fueled deep burn cores to ensure the safety of the deep burn core during the conduction cooling events. 相似文献
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In this work, we performed an evaluation of decay heat power of advanced, fast spectrum, lead and molten salt-cooled reactors, with flexible conversion ratio. The decay heat power was calculated using the BGCore computer code, which explicitly tracks over 1700 isotopes in the fuel throughout its burnup and subsequent decay. In the first stage, the capability of the BGCore code to accurately predict the decay heat power was verified by performing a benchmark calculation for a typical UO2 fuel in a Pressurized Water Reactor environment against the (ANSI/ANS-5.1-2005, “Decay Heat Power in Light Water Reactors,” American National Standard) standard. Very good agreement (within 5%) between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power for fast reactors with different coolants and conversion ratios, for which no standard procedure is currently available. Notable differences were observed for the decay power of the advanced reactor as compared with the conventional UO2 LWR. The importance of the observed differences was demonstrated by performing a simulation of a Station Blackout transient with the RELAP5 computer code for a lead-cooled fast reactor. The simulation was performed twice: using the code-default ANS-79 decay heat curve and using the curve calculated specifically for the studied core by BGCore code. The differences in the decay heat power resulted in failure to meet maximum cladding temperature limit criteria by ∼100 °C in the latter case, while in the transient simulation with the ANS-79 decay heat curve, all safety limits were satisfied. The results of this study show that the design of new reactor safety systems must be based on decay power curves specific to each individual case in order to assure the desired performance of these systems. 相似文献
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温度是影响熔岩玻璃体溶解速度的关键因素,为此,本文计算了核试验后10~300 000d内熔岩玻璃体中核素衰变热功率,评估了核素衰变热功率对熔岩玻璃体的温度和溶解速度的影响程度。采用了国际原子能机构给出的100kt TNT当量地下核试验产生的、半衰期大于1a的放射性核素含量,利用其中裂变产物核素137 Cs的含量推算累积裂变产额大于0.1%、半衰期为1d~1a的短寿命裂变产物核素的含量。分析了各核素的放射性衰变特点,采用ENDF/BⅦ库中核素衰变辐射的平均α能量、平均电子能量和平均电磁辐射能量计算各核素在熔岩玻璃体内因衰变而沉积的能量。计算结果表明:核素衰变热功率呈分段幂函数衰减;在10~2 000d、2 000~60 000d和60 000d之后的时段内,衰变热功率分别主要源于短寿命裂变产物核素、长寿命裂变产物核素和锕系元素。核素衰变热功率对熔岩玻璃体的温度和溶解速度的影响不大,1 000d后影响非常小。 相似文献
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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。 相似文献