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。     

燃耗信任制临界计算中保守性因素研究

在运用燃耗信任制技术进行乏燃料储存、运输等环节的临界安全分析时,临界计算所采用的条件是否具有足够的包络性十分关键。本文借助于OECD/NEA发布的若干燃耗信任制临界安全基准题,使用SCALE5.1软件中的STARBUCS模块进行分析,对信任核素选取、乏燃料冷却时间以及端末效应等因素对乏燃料系统临界安全性的影响进行了研究,得出了各参数保守性的有关结论。     

PWR核电厂乏燃料贮存临界计算重要核素的选取

信用核素选取是基于燃耗信用制乏燃料贮存临界安全分析的关键一步。通过对不同富集度、燃耗深度及停堆冷却时间下典型PWR燃料组件分析,以核素中子吸收份额大小排序为依据,筛选出对总的中子吸收起主要贡献的核素。结果显示,47个核素即可包络停堆后0~20a内影响乏燃料贮存系统反应性的所有核素中的99%。通过核素敏感性因子分析证明依据中子吸收份额排序选取重要核素的方法是合理的,与基准算例的结果对比证明所筛选出的核素能足够代表影响系统反应性的所有重要核素。     

基于CASMO5的燃耗历史对乏燃料反应性的影响计算

基于乏燃料贮存领域常用的锕系加裂变产物(APU-2)级燃耗信任制,应用二维组件燃耗计算程序CASMO5,计算了燃耗过程中功率密度和运行历史对乏燃料k∞的影响。结果表明:燃耗计算中,选择堆芯额定功率对应的平均功率密度,同时k∞附加0.002 3的包络裕度,运行历史选择循环内及循环间无停堆额定功率运行,同时k∞附加0.004 5的包络裕度,可满足燃耗信任制中包络性原则。     

燃耗信任制下燃耗计算对临界计算的偏差及不确定度的研究

为量化燃耗信任制中燃耗计算传递给临界计算的不确定度,本文基于参数统计法对燃耗计算的核素偏差及偏差不确定度展开分析,并以蒙特卡罗（MC）抽样方法计算的kinf不确定度为基准,比较不同抽样方法对临界计算不确定度的影响。结果表明,核素偏差与偏差不确定度是随样品燃耗变化的分段函数。对于临界计算,拉丁超立方抽样（LHS）方法与MC抽样方法的kinf不确定度计算结果吻合较好,且LHS方法可考虑参数间的相关性,计算结果更真实,可进一步提升电厂的经济性。     

田湾核电站乏燃料水池采用燃耗信任制的计算研究

以田湾核电站(TNPS)2×5排列的贮存格架构成的乏燃料水池为例,研究采用燃耗信任制技术的密集贮存和临界安全问题。采用MONK9A程序计算分析不同富集度、不同燃耗的乏燃料装载情况下系统的k_eff. 根据系统k_eff随不同初始富集度燃料的燃耗变化情况给出了水池的参考装载曲线。采用燃耗信任制技术的密集贮存方案能提高贮存能力31%。     

田湾核电站乏燃料干式贮存临界安全计算研究

以CASTOR 1000/19干式贮存容器装载田湾核电站六角形乏燃料组件为例,研究六角形乏燃料干式贮存的临界安全问题。基于新燃料假设,应用MONK9A程序对贮存容器满装载乏燃料进行不同工况下k_eff的计算。计算结果表明：正常工况下,k_eff远小于临界安全限值,是临界安全的;事故工况下,当²³⁵U富集度大于3.15%时,系统存在临界安全风险,须减少乏燃料装载量来确保临界安全。考虑燃耗信任制后,采用相同的模型计算得出贮存容器满装载的参考装载曲线,按此曲线要求装载能确保所有工况下的系统临界安全。采用燃耗信任制技术提高了贮存容器的利用率。该研究可为田湾核电站采用乏燃料干式贮存方案提供依据。     

乏燃料贮存燃耗信任制计算的敏感元素分析

燃耗信任制的应用分析需要计算乏燃料系统的反应性,为了保证临界安全需要保证分析结果的包络性,为此需要考虑到各种不同燃耗因素对于分析结果的影响。采用SCALE程序包中的STARBUCS模块,通过对OECD/NEA发布的Phase-IA及Phase-IIA基准题的验证分析,研究了不同信任等级、堆芯运行参数等因子对乏燃料贮存系统临界安全的敏感性分析,并得出有益的结论。     

“CANDLE” burnup regime after LWR regime

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment.

If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth.

The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium.     

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%.     

Active reduction of the end effect by local installation of neutron absorbers

In the burnup credit analyses of interim or long-term spent fuel (SF) storage facilities and transport casks, when the average burnup value is greater than approximately 30 GWd/t, the neutron multiplication factor becomes greater if we consider the axial burnup distribution of the spent fuel assembly rather than assuming an average burnup. This phenomenon is called the “end effect” and it is one of the main technical issues in burnup credit research. The end effect is characterized by an increase of the neutron flux around the end regions of the spent fuel assemblies in the criticality calculation. However, such increase of the neutron flux has not been observed in experiments using actual spent fuel assemblies.     

Thinking outside the box: options for transport of high burnup spent fuel

Abstract

The current uncertainty surrounding the licensing and eventual opening of a long term geologic repository for the nation’s civilian and defense spent nuclear fuel and high level radioactive waste has shifted the window for the length of time spent fuel could be stored to periods of time significantly longer than the current licensing period of 40 years for dry storage. An alternative approach may be needed to the licensing of high burnup fuel for storage and transportation based on the assumption that spent fuel cladding may not always remain intact. The approach would permit spent fuel to be retrieved on a canister basis and could lessen the need for repackaging of spent fuel. This approach is being presented as a possible engineering solution to address the uncertainties and lack of data availability for cladding properties for high burnup fuel and extended storage time frames. The proposed approach does not involve relaxing current safety standards for criticality safety, containment, or permissible external dose rates.     

国产压水堆核电站40 GW•d/tU乏燃料棒金相检验

为评价国产燃料棒在较高燃耗水平下的辐照性能,在中国原子能科学研究院燃料与材料检验设施（303热室）对燃耗为40 GW•d/tU的国产压水堆核电站乏燃料棒进行了金相检验。检验内容包括芯块宏观与微观组织、包壳水侧腐蚀与氢化物分布、芯块-包壳相互作用状况等。金相检验结果表明：40 GW•d/tU燃耗下,芯块未发生明显的轮廓变化,气孔率为3.3%～5.8%,晶粒组织为等轴晶,平均晶粒尺寸为7.2 μm;Zr合金最大水侧氧化膜厚度为23 μm,氢化物分布和含量正常,最大氢含量约为150 μg/g,同时不同部位的包壳氢含量与水侧氧化膜厚度基本呈线性关系,水侧腐蚀处于正常水平;包壳内壁有局部轻微腐蚀,包壳与芯块之间存在间隙,未发生包壳与芯块相互作用情况。     

蒙特卡罗方法在CZT探测器测量乏燃料组件燃耗中的应用

利用碲锌镉（CZT）探测器组成的γ谱探测系统是一种测量乏燃料组件燃耗的较有效的方法。本文利用蒙特卡罗方法,借助于Geant4软件包计算在两种测量方式、多个准直高度条件下组件中¹³⁷Cs和¹³⁴Cs的全能峰探测效率,对测量结果的正确评价分析具有一定意义。另外,采用偏倚抽样方法确定源粒子发射方向,极大提高了CZT探测器全能峰探测效率。     

乏燃料燃耗与中子发射强度的关系

对辐照过的低富集氧化铀燃料的燃耗与中子发射强度间的关系进行分析和研究。计算了不同初始富集度、不同燃耗、不同冷却时间的乏燃料的中子发射强度,经分析,证实了燃耗与中子发射强度间存在的幂函数关系,并对影响幂函数关系的各种因素进行了研究。发现幂函数关系中的系数受初始富集度、冷却时间、燃耗范围的影响;如果冷却时间大于2a,这个关系不受辐照历史的影响;如果冷却时间小于2a,这个关系受辐照历史的影响。     

通过测量模拟样品中236U与238U原子个数比估算燃耗的探索研究

自然界中²³⁶U与²³⁸U原子个数比约10^-14,不同反应堆类型及核燃料辐照情况辐照后的核材料中²³⁶U与²³⁸U原子个数比不同,一般为天然²³⁶U与²³⁸U原子个数比的10⁷～10¹¹倍。通过测量环境样品中的²³⁶U与²³⁸U原子个数比可探知取样点附近进行过的辐照活动、环境污染的来源及对应核燃料的燃耗。本研究使用配制的模拟样品,建立了多接收电感耦合等离子质谱（MC-ICP-MS）技术测定²³⁶U与²³⁸U原子个数比的方法以及估算核燃料燃耗的工作方案,并与其他燃耗计算方法比较,燃耗的相对偏差约10%。     

先进非能动压水堆核电站燃料组件轴向燃耗分布研究

燃料组件的轴向燃耗分布以及末端效应是燃耗信任制技术应用中的难点。基于先进非能动压水堆核电站的运行模式及组件设计特点,结合可能的燃料管理策略,统计轴向两端使用低富集度抑制区的乏燃料组件,生成轴向燃耗包络线。以 AP1000堆型的乏燃料贮存单元为例,通过分析统计,证明生成的轴向燃耗包络线用于临界安全分析是保守的。在此基础上,详细研究燃料组件顶部的低富集度抑制区对末端效应的贡献,并对燃料组件进行设计改进,减小至消除末端效应,为简化乏燃料组件相关的临界安全分析提供了一个方法。相关研究工作及成果,是先进非能动压水堆核电站乏燃料组件相关的设施设备的临界安全设计的基础,可为其他堆型的相应研究提供参考和借鉴。     

先进非能动压水堆核电站燃料组件轴向燃耗分布研究

燃料组件的轴向燃耗分布以及末端效应是燃耗信任制技术应用中的难点。基于先进非能动压水堆核电站的运行模式及组件设计特点,结合可能的燃料管理策略,统计轴向两端使用低富集度抑制区的乏燃料组件,生成轴向燃耗包络线。以AP1000堆型的乏燃料贮存单元为例,通过分析统计,证明生成的轴向燃耗包络线用于临界安全分析是保守的。在此基础上,详细研究燃料组件顶部的低富集度抑制区对末端效应的贡献,并对燃料组件进行设计改进,减小至消除末端效应,为简化乏燃料组件相关的临界安全分析提供了一个方法。相关研究工作及成果,是先进非能动压水堆核电站乏燃料组件相关的设施设备的临界安全设计的基础,可为其他堆型的相应研究提供参考和借鉴。     

燃料元件破坏性燃耗测量过程的质量控制

燃耗是核燃料元件最重要的性能指标之一,其准确测量对新型燃料元件研制和换料周期确定等具有重要意义。破坏性燃耗测量属于强放射性下的精细化学分析,需建立系统的方法对测量过程进行质量控制,确保测量数据准确可靠。本文从方法适用性分析、数据预估、质谱干扰分析、样品污染分析、多种方法验证等方面介绍了破坏性燃耗测量过程质量控制的方法。剖析了重同位素法、¹⁴⁸Nd监测体法、¹⁴⁵Nd+¹⁴⁶Nd监测体法、¹³⁷Cs监测体法等的优缺点和适用范围。介绍了由裂变产额比值预估钕同位素丰度比、由铀同位素丰度比预估燃耗值的方法。分析了质谱测量时由Ce和Sm同位素造成的同量异位素干扰及其检测、排除手段。针对天然本底污染和样品间的交叉污染,分别论述了两种污染源的判断和修正技巧。还探讨了对燃耗值、稀释剂浓度、同位素丰度比等关键数据进行对比验证的方法。     

Estimation of Spent Fuel Compositions from Light Water Reactors

The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future.

In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.

On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U0₂ and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.