EAST中性束注入器大厅内辐射剂量计算与测量

利用解析方法对EAST大功率中性束注入器充氢运行时实验大厅内6个关键点的辐射剂量进行了理论计算,并利用光致光剂量计(OSL)对这些位置点进行了辐射剂量测量。理论计算和OSL测量结果表明:理论计算结果与实验测量结果具有一定吻合度。同时还表明:EAST中性束注入器现有的防护装置满足实验运行时辐射防护要求。     

SSRF活动光子挡光器的模拟与实验结果比较

介绍了上海光源插入件光束线前端区高热负载元件活动光子挡光器热缓释设计的数值模拟和实验结果.数值模拟与实验测量结果在10℃的误差范围内一致,且数值模拟结果都高于测量值.测量结果证明了挡光器热缓释设计中采用的数值模拟方法在直接冷却高热负载部件的应用中的可靠性和安全性.     

中子Soller准直器摇动曲线的蒙特卡罗模拟和分析

中子Soller准直器是中子散射谱仪一种常用部件,主要用于限制中子束流的发散度,提高中子谱仪的分辨率。准直器的准直角是描述其性能的一项重要技术指标,通过摇动曲线的实验测量获得。本文基于实验测量需求和方法,利用中子射线追踪模拟蒙特卡罗程序McStas对中子Soller准直器摇动曲线实验测量过程进行了蒙特卡罗模拟计算,并对计算结果进行实验对比和解析分析。分析结果表明,模拟计算、实验数据和理论解析三者符合较好,验证了模拟方法的可行性。通过多方法结合来研究摇动曲线,可提高测量分析结果的可靠性,类似的方法也可扩展应用于其他中子光学部件的测试分析。     

脉冲线测量波荡器磁场的研究

介绍脉冲线测量波荡器磁场一、二次积分分布的原理和一套较高精度的脉冲线测量系统的研制，分析了传感器、实验环境、线的张力和波的反射性等因素对测量精度的影响。该系统对5周期波荡器模型的磁场测量结果与高精度Hall探头测量值作了比较，结果满意。     

γ刻度室的散射测量

本文介绍了铅准直器 γ 刻度室的散射实验研究。用几种方法测量了不同使用距离上的相对散射系数,其结果小于5%。测量了 γ 散射能量,并探讨了准直器的散射,从而了解了辐射场的性质。文中还提出了值得注意的几个问题。     

医院中子照射器热中子束能谱特性研究

医院中子照射器是我国建造的第1座用于医疗目的的微型反应堆,已于2009年12月7日首次达临界,2010年1月22日达到满功率运行。在治疗前,需测量出口处的中子通量密度及能谱等参数,为后续实验提供依据。本文用MCNP建立医院中子照射器模型,得到能谱计算值。选用金箔活化法测量绝对中子通量密度,多箔活化法测量中子能谱,用SAND-Ⅱ程序解谱,并将实验结果与计算结果进行了比较。     

CZT探测器测量核燃料组件铀丰度实验研究

采用CZT探测器、数字谱仪、准直器等组成了1套便携式CZT探测器铀丰度测量装置。该装置可对燃料组件铀丰度进行测定,以便确定相应铀产品丰度符合规定要求。实验研究中,对几类燃料组件丰度进了测量,建立了CZT探测器测量燃料组件铀丰度的方法。现场测量结果表明,铀丰度测量结果相对偏差小于3%,方法简单可靠,装置简便,能满足核材料保障监督和核设施现场测量中的需求。     

~(141)Pr(n,γ)~(142)Pr反应截面的测量

用活化法测量了中子能量在 0.539 MeV,1.090 MeV 和 1.587MeV 的 Pr(n, γ)142Pr 核反应截面值,γ 放射性活度用高纯锗探测 141器测量,实验的测量误差在±(6～7)%范围内。实验结果与其它各家的实验值进行了比较和分析,给出了推荐的激发曲线。     

欧空局单粒子监督器在北京HI-13串列加速器上的单粒子效应校核实验

为检验北京HI-13串列加速器单粒子效应(SEE)实验能力与数据测量的可靠性,利用束流参数校核系统——欧空局单粒子监督器进行了单粒子效应校核实验。实验使用C、F、Cl、Cu 4种离子辐照单粒子监督器,通过改变入射角度获得了有效LET值在1.8~67.4 MeV·cm~2·mg~(-1)之间的单粒子翻转(SEU)截面数据。实验结果与比利时HIF、芬兰RADEF装置上测得的截面数据一致性较好,证实了北京HI-13串列加速器单粒子效应实验束流参数测量的准确性及截面数据测试的可靠性。     

西安脉冲堆中子照相束捕集器参数研究

本文建立了用改进的活化法与蒙特卡罗计算相结合测量中子反照率的方法.实验中用激活片直接测量了西安脉冲反应堆中子照相捕集器表面不同位置处的中子总通量,计算中用蒙特卡罗MCNP程序计算了各激活片上正反方向的中子通量比;测量值经过理论计算值纠偏后,确立了测量位置处入射和反射中子的通量,从而确定了捕集器表面的中子反照率分布这一重要参数,为在中子和光子混合场下开展中子照相束捕集器参数测量和实验应用提供了参考.     

10MW高温气冷堆的燃耗测量研究

10MW高温气冷堆的燃耗测量系统是采用非破坏性高纯锗γ谱仪在线监测来确定燃耗值,利用MCNP4A程序对测量系统的衰减因子进行计算,基于核燃料裂变核素的γ射线能谱分析,以137Cs和134Cs核素活度作为测量对象,并对燃耗测量结果进行讨论.     

燃耗数据库基准检验方法研究

燃耗数据库基准检验方法对于研制高准确度的燃耗数据库至关重要。本文以TAKAHAMA 3压水堆辐照后检验实验中SF95样品的建模为例，研究了建模要素对燃耗计算的影响，确定了燃耗实验建模的方法，开展了燃耗信用制研究感兴趣的锕系和裂变产物核素积存量计算值与实验值的比对。比对结果显示，主锕系核素计算偏差小于2%，大部分次锕系核素偏差小于10%，大部分重要裂变产物核素偏差小于5%。本文还对125Sb积存量随燃耗深度变化规律进行了理论分析，确认了破坏性放化实验测量结果存在缺陷，并进一步获得了125Sb积存量的修正值，使计算偏差从接近170%下降到20%以内。本次研究表明，燃耗数据库基准检验研究不仅需发展适当的燃耗实验建模方法，还需对实验数据进行适当的评价。     

行波焚烧模式的理论分析与初步计算

行波焚烧模式是一种即时原位增殖焚烧模式。本工作从理论上进行初步分析,得出行波焚烧的物理图像。利用欧洲快堆计算程序ERANOS进行二维圆柱模型的方案分析,得到行波焚烧的基本物理特点。结果表明,行波堆在中子学上是可行的,且具有一些很好的物理特性。     

Nondestructive burnup measurements by gamma-ray spectroscopy on spent fuel elements of the RP-10 research reactor

Gamma-ray spectroscopy is an important nondestructive method for the qualification of irradiated nuclear fuels. Regarding research reactors, the main parameter required in the scope of such qualification is the average burnup of spent fuel elements. This work describes the measurement, using nondestructive gamma-ray spectroscopy, of the average burnup attained by Material Testing Reactor (MTR) fuel elements irradiated in the RP-10 research reactor. Measurements were performed at the reactor storage pool area using ¹³⁷Cs as the only burnup monitor, even for spent fuel elements with cooling times much shorter than two years. The experimental apparatus was previously calibrated in efficiency to obtain absolute average burnup values, which were compared against corresponding ones furnished by reactor physics calculations. The mean deviation between both values amounts to 6%.     

HTR-10燃耗自动测量系统

对燃料球进行高效准确的燃耗测量是球床式高温气冷堆实现高利用因子运行的关键环节.10MW高温气冷堆燃耗测量目前未能实现自动运行.结合燃料装卸系统设计原理及燃料循环过程运行特点,对HTR-10原手动燃耗测量提出改进,实现了自动燃耗测量.现场运行结果表明,该方法逻辑准确、可靠性高,能够有效避免人为因素造成的误操作.     

Analysis of Post Irradiation Experiments in PWR Using New Nuclear Data Libraries

For the precise calculation of the burnup of minor actinide isotopes, a code system-SWAT has been developed. This system analyzes burnup problems with neutron spectrum that depends on the type of a reactor and the irradiation history, using latest evaluated nuclear data files JENDL-3 or ENDF/B-Vl. The post irradiation test in TRINO and the recent experiment in typical PWRs in Japan were analyzed with SWAT. These analyses show that the results of U and Pu for high burnup fuels almost agree with experimental results but those for middle burnup fuels do not agree with them. The results for Am and Cm isotopes still have large discrepancy. The average C/E of ²⁴³Am is –0.79, and that of ²⁴⁴Cm is –0.70 for high burnup (–33,000 MWd/tU) samples.

For middle burnup (–25,000 MWd/tU) samples, the C/E for ²⁴⁴Cm is over 2.0. The discrepancy is partially explained by considering the power peaking history of first cycle and second cycle.     

Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions

This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.     

Application of candle burnup strategy for future nuclear energy utilization

The CANDLE burnup strategy is a new reactor burnup concept, where the distributions of fuel nuclide densities, neutron flux, and power density move with the same constant speed along the core axis from bottom to top (or from top to bottom) of the core and without any change in their shapes. Therefore, any burnup control mechanisms are not required, and reactor characteristics do not change along burnup. The reactor is simple and safe. If this burnup scheme is applied to some neutron rich fast reactors, either natural or depleted uranium can be utilized as fresh fuel after second core and the burnup of discharged fuel is about 40%. It means about 40% of natural or depleted uranium can be utilized without either enrichment or reprocessing.

In the ideal nuclear energy utilization system, the radioactive toxicity in the environment should remain or decrease after the utilization. This requirement is very severe and difficult to be satisfied. It may take too much time for its realization. The CANDLE burnup may substitute this period. Though it is a once-through fuel cycle, the discharged fuel burnup is about ten times of the present value for light water reactors. The space necessary for final disposal can be drastically reduced. However, in order to realize such a high burnup of discharged fuels some innovative technologies should be developed. Either new material standing still for such a high burnup or intermediate recladding will be required. Especially new fuel development will take a lot of time. For the time being a small reactor with CANDLE burnup may be a good option for nuclear power generation. Even this kind of reactor requires some innovative technologies and a long period for their developments. For the first stage of CANDLE burnup the prismatic fuel high-temperature gas cooled reactor is preferable. Since the design of this reactor fits to the CANDLE burnup very well, only a little time is required for its research and development.     

基于SCALE的熔盐堆添料与后处理系统分析程序开发及验证

在线添料及在线去除中子毒物是熔盐堆区别于其他固体燃料反应堆的主要特征之一,能够实现较高的燃耗深度和燃料利用率。然而,现有的反应堆物理计算分析软件SCALE不能直接模拟熔盐堆的燃耗计算。因此,本文耦合SCALE中的截面处理模块、临界计算模块以及燃耗计算模块,开发了一套适用于多流体熔盐堆的添料与后处理系统分析程序MSR-RRS,实现熔盐堆的在线添料、裂变产物在线处理或离线批次处理等模拟功能。基于MSR-RRS对现有的单流熔盐增殖堆和双流熔盐快堆的燃耗性能进行了验证。结果表明,MSR-RRS计算结果与基准模型结果符合较好。MSR-RRS适用于多种堆型、多种燃料循环运行模式。     

基于精细燃耗历史及精细燃耗链的球床高温气冷堆燃耗不确定性分析

球床高温气冷堆的燃料管理具有燃料球多次通过堆芯的特点,使得燃料元件经历的燃耗历史十分复杂。球床高温气冷堆堆芯物理设计程序VSOP可以提供燃料元件的精细燃耗历史,但仅包含少量燃耗链和核素种类。而清华大学自主开发的燃耗计算程序NUIT可实现精细燃耗计算,且包含完整燃耗链和核素信息,但不具备精细燃耗历史跟踪功能。本文基于NUIT,结合VSOP提供的球床高温气冷堆精细燃耗历史,开发了球床高温气冷堆堆芯的精细燃耗计算功能,搭建了带有精细燃耗历史模拟和精细燃耗链核素的燃耗分析流程,并实现燃耗不确定性分析功能。在此基础上研究了裂变产额不确定性对球床高温气冷堆燃耗计算不确定性的贡献,并与VSOP的计算结果进行对比。计算分析结果显示,基于NUIT的精细燃耗计算结果和VSOP的燃耗计算结果得到了相互验证,且可以得到更多的核素浓度信息,该计算结果是开展球床高温气冷堆衰变热不确定性研究的基础。