新型熔盐快堆的物理优化设计及99Mo产量分析

为提高新型熔盐快堆的堆芯中子经济与安全性能,并利用²³⁵U的裂变反应进行⁹⁹Mo同位素生产,应用SCALE6.1程序进行了堆芯几何参数优化,基于优化后的堆芯对⁹⁹Mo同位素的生产进行相关分析。结果表明：适当增加燃料元件半径、减小燃料栅元半径可提高有效增殖因子,同时降低冷却剂温度系数;当燃料元件容器壁厚为0.1 cm、燃料元件半径为3.5 cm、栅元半径为5 cm、活性区半径和反射层厚度分别为63 cm和100 cm时,堆芯运行寿期满足32个月,此时总反应性温度系数为-1.615×10^-5K^-1,保证了堆芯的固有安全性;选最外层燃料元件作为⁹⁹Mo生产的燃料靶件可提高⁹⁹Mo的产量,当燃料靶件提取周期为7 d时,⁹⁹Mo出堆年产量达到6.25×10¹⁶Bq,比活度为2.77×10¹⁵Bq·g^-1。     

252Cf、238Pu与237Np三种核素的生产供应

²⁵²Cf、²³⁸Pu、²³⁷Np三种核素是用于深空探测和核能发电等领域的重要放射性同位素,国内主要依靠进口,来源有限。了解这三种核素的生产供应情况,对国内开展相关研究工作有重要参考价值。本文分别介绍了²⁵²Cf、²³⁸Pu、²³⁷Np的特性和用途,并概述了其生产供应情况。目前²⁵²Cf主要在美国橡树岭国家实验室（ORNL）和俄罗斯原子反应堆研究所（RIAR）的高中子通量反应堆辐照生产。冷战结束之后,²³⁸Pu两大生产国——美国和俄罗斯的生产能力逐渐丧失,随着深空探测任务对同位素电池的需求,近些年美俄两国正在陆续恢复生产。²³⁷Np作为²³⁸Pu生产的原材料,主要存在于裂变产物或高放废物中,通过后处理流程分离提取。为保障国内反应堆的稳定运行和深空探测任务的开展,建议尽快实现上述三种战略核素的自主供应能力。     

237Np/235U裂变截面比和俘获裂变截面比的测量与分析

在中国实验快堆（CEFR）中直接测量²³⁷Np的裂变截面和俘获截面较困难且误差很大,根据CEFR采用UO₂燃料的特点,可通过测量²³⁷Np/²³⁵U裂变截面比和俘获裂变截面比以获取²³⁷Np的相关数据。本文通过分析截面比的测量结果得到²³⁷Np的较重要数据,为后期在CEFR上进行²³⁷Np嬗变研究和²³⁸Pu生产提供了数据支持。     

~(238)Pu的辐照生产及制备

介绍237Np靶件堆内辐照生产238Pu的途径及其相应物理过程。从237Np靶件几何设计、辐照靶件中237Np装量、238Pu产量及236Pu相对含量随时间的变化、238Pu辐照生产实验测量结果等方面分析了238Pu辐照生产技术,简单介绍靶件辐照后的化学分离提纯、238Pu粉末和热源制作等工艺过程。     

TBP-TEVA萃取色层分离-α能谱法测定辐照镎靶溶解液中痕量236Pu的可行性

²³⁶Pu的含量控制是钚热源的一项重要参数，通过α能谱准确测量镎靶溶解液中痕量²³⁶Pu，建立镎靶辐照靶件溶解液中钚的分离方法。根据杂质组成特点采用TBP-TEVA萃取色层双柱分离，用氨基磺酸亚铁以及亚硝酸钠对钚进行调价，对靶件溶解液中的Al、Fe、U、Th和Np等进行分离，去污系数均大于10⁴，钚的回收率为90.7%。研究大量²³⁸Pu对α能谱测定²³⁶Pu的干扰，结果表明，大量²³⁸Pu会造成仪器本底升高，²³⁸Pu能谱峰分辨率降低；在7 500 Bq ²³⁸Pu干扰下，测量4.3 h 时，²³⁶Pu的最小可检测活度为1.20×10^-2 Bq（当量质量为6.11×10^-16 g）。计算结果表明，镎靶溶解液样品中钚的同位素比值n(²³⁶Pu)/n(²³⁸Pu) ≥4.63×10^-8时，取合适的样品量使得电沉积源中²³⁸Pu 活度在 450~7 500 Bq范围内，均可测量其中的痕量²³⁶Pu，同时可准确测定同位素比值n(²³⁶Pu)/n(²³⁸Pu)。     

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%。     

238Pu溶液α吸收剂量率的测定

为了了解α粒子对无盐试剂的辐解效应,需要确定α辐解源的辐照剂量率。采用硫酸亚铁剂量计(Fricke)测定了²³⁸Pu溶液的α吸收剂量率,求得单位浓度²³⁸Pu的剂量率为40.0~43.1 Gy•L/（g•min）;按照²³⁸Pu α粒子的能量计算,质量浓度为1 g/L的²³⁸Pu溶液每分钟吸收的能量为32.58 J/min, 即32.58 Gy•L/(g•min),计算值和理论值偏差为23%~33%,此与α辐解测定难度和实验环境有关。     

加速器质谱法测量含铀微粒中铀同位素比的方法研究

准确测定含铀微粒同位素比在核保障中有重要的应用价值。本文采用将含铀微粒溶解并加入高纯Fe粉烘干的方法制样,采用中国原子能科学研究院的HI-13串列加速器质谱测量靶样中的同位素比。通过对CRM铀系列同位素标准样品的分析表明,该方法可测定高于10^-5的²³⁶U/²³⁸U同位素比;对于²³⁵U/²³⁸U同位素比在10^-4～10^-1范围内的含铀微粒,²³⁵U/²³⁸U同位素比相对扩展不确定度均小于10%。     

水样中Pu、Np、Am和Cm的联合快速分析方法

环境监测、辐射防护、核取证和核应急等领域对环境和生物样品中²³⁸Pu、²³⁹Pu、²⁴⁰Pu、²⁴¹Pu、²³⁷Np、²⁴¹Am、²⁴³Cm和²⁴⁴Cm测定的需求日渐增大。本研究提出一个自上而下串联TEVA树脂、UTEVA树脂和DGA树脂的联合、快速、可靠、可批量操作的分析方法，该方法首先通过水合氧化钛（HTO）共沉淀将待测核素从样品基质中分离，其后使用串联层析柱中的TEVA树脂柱分离纯化Pu与Np，DGA层析柱分离纯化Am与Cm。对于α放射性核素，通过CeF₃微沉淀法制备薄层α测量源，使用高分辨率α谱仪分别测量^239+240Pu、²³⁸Pu、²³⁷Np、²⁴¹Am与^243+244Cm；对于β放射性核素²⁴¹Pu，使用液体闪烁计数器测量。²³⁶Pu和²³⁴Am示踪表明该流程的化学回收率大于80%，加标实验结果表明期望值与测量值相吻合，证明了该方法的高可信度及稳定性。α谱仪测量48 h，最小可探测活度²⁴¹Am为0.40 mBq，^243+244Cm为0.33 mBq，²³⁸Pu为0.72 mBq，^239+240Pu为0.44 mBq，²³⁷Np为0.72 mBq。液闪计数器测量1 800 s，²⁴¹Pu的最小可探测活度为0.17 Bq。使用12孔真空盒同时制备12个样品，可加快制样时间，批次制样时间小于3 h，极大地降低了样品的使用量、制备时间和分析成本。     

缓发中子计数法定量235U、239Pu初步研究

利用缓发中子计数法对²³⁵U-²³⁹Pu混合物中²³⁵U和²³⁹Pu含量的快速测定进行了初步研究。在中国原子能科学研究院30 kW微型反应堆(简称微堆)垂直孔道辐照²³⁵U、²³⁹Pu以及²³⁵U-²³⁹Pu混合物样品30 s,冷却2 s,用缓发中子探测器测量100 s,得出²³⁵U和²³⁹Pu的探测限分别为0.14和0.18 μg;探测器效率为0.015 0±0.001 0;当²³⁵U和²³⁹Pu质量比m(²³⁵U)/m(²³⁹Pu)=1.2时,²³⁵U、²³⁹Pu含量计算值与标称值的相对偏差分别为0.8%和6.9%。     

Recovery of minor actinides from spent molten salt waste and decontamination of molten salt waste

Recovery of minor actinides from spent molten salt is one of the important issues. Decontamination of spent molten salt waste is also the problem to be solved for establishment of pyrochemical reprocessing. The decontamination method of spent molten salt waste with recovery of minor actinides has been proposed. Our proposed process is based on the hydrometallurgical process. This process consists of the following processes. First, the spent molten salt waste is dissolved in aqueous solution. Next, the minor actinides are recovered by chromatographic techniques using the pyridine resin in the methanolic hydrochloric acid solution. In the last process, the spent molten salt waste is decontaminated by the cation-exchange chromatography. In the present paper, the adsorption behavior of minor actinides, rare earth elements, alkaline earth elements, and alkali metal elements on pyridine resin is reported. The demonstration experiment of the recovery of the minor actinides from simulant spent molten salt waste is also reported.     

Non-proliferation attributes of molten salt reactors

Molten salt reactors (MSR) have many non-proliferation attributes. They can operate on the thorium-uranium fuel cycle which protects the fissile material by the daughter products of the inseparable U-232. MSRs can completely fission all plutonium and HEU, and as desired, ‘convert’ them to U-233. This also results in high, and efficient resource utilization, while diminishing the plutonium stock. On line processing, when applied, could free the waste from all fissile material. The fuel in the reactor stays protected by the intense radiation of the fission products. Fuel can also be protected in the reactor as well as outside the reactor by denaturing with natural uranium. A wide variety of MSRs are available, from ‘once through’ minimum processing reactors to ones with fuel processing which can breed fuel for converters. MSRs are extremely safe and simple reactors with good economic potential.     

Structural investigation of thorium in molten lithium-calcium fluoride mixtures for salt treatment process in molten salt reactor

X-ray absorption fine structure (XAFS) measurements on thorium fluoride in molten lithium-calcium fluoride mixtures and molecular dynamics (MD) simulation of zirconium and yttrium fluoride in molten lithium-calcium fluoride mixtures have been carried out. In the molten state, coordination number of thorium (N_i) and inter ionic distances between thorium and fluorine in the first neighbor (r_i) are nearly constant in all mixtures. However the fluctuation factors (Debye-Waller factor (σ²) and C₃ cumulant) increase until xCaF₂ = 0.17 and decrease by addition of excess CaF₂. It means that the local structure around Th⁴⁺ is disordered until xCaF₂ = 0.17 and stabilized over xCaF₂ = 0.17. The variation of fluctuation factors is related to the number density of F⁻ in ThF₄ mixtures and the stability of local structure around Th⁴⁺ increases with decreasing the number density of F⁻ in ThF₄ mixtures. This tendency is common to those in the ZrF₄ and YF₃ mixtures. However, in the case of YF₃ mixtures, the local structure around Y³⁺ becomes disordered until xCaF₂ = 0.40 and it becomes stabilized by addition of excess CaF₂. The difference between ThF₄ mixtures and YF₃ mixtures is related to the difference of Coulumbic interaction between Th⁴⁺-F⁻ and Y³⁺-F⁻. Therefore, the variation of local structure around cation is related to not only number density of F⁻ in molten salts but also the Coulumbic interaction between cation and anion.     

火星熔盐堆堆芯概念设计

火星探测近来成为空间研究的一个主流趋势。建立火星基地是人类研究和开发火星的必然选择。与太阳能储能系统相比,核反应堆作为火星基地的能源系统,在系统质量、操作灵活性和环境适应性等方面具有显著优势。给出了火星熔盐堆(Mars Molten Salt Reactor,M~2SR-1)的堆芯设计方案,并建立堆芯计算模型,以MCNP(Monte Carlo N Particle Transport Code)和ORIGEN为计算工具,从物理、安全、热工等方面对M~2SR-1进行了计算分析。分析结果表明:M~2SR-1在满功率运行下可满足8 a的寿期要求,在不同假设掉落环境下,有效增殖因数均小于0.98,满足临界安全要求。本研究可以为星球表面熔盐堆设计提供参考。     

Characterization of molten 2LiF–BeF_2 salt impregnated into graphite matrix of fuel elements for thorium molten salt reactor

The impregnation behavior of molten 2LiF–BeF_2(FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor(TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques.It was found that the entrance pore diameter of the graphite matrix is less than 1.0 μm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%,corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures(0.8 MPa and1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.     

Coupled dynamics in the physics of molten salt reactors

The present paper is devoted to the analysis of the coupled thermo-fluid and neutronic dynamics of fast fluid-fuel multiplying nuclear systems. A completely coupled model is needed since in some fast reactors designs, the velocity pattern could be very complicated and strongly affected by the neutron dynamics via the heat source from fission reactions. Furthermore, the neutron dynamics is strongly affected by the thermohydrodynamics via the motion of precursors and by feedback effects. The methods typical of solid fuel reactors of previous generations are not sufficient to handle these more highly coupled concepts. In the preset paper, we consider the coupling between neutronics and thermohydrodynamics with simple but realistic hypotheses assumed to model the evolution of all the variables involved in the calculation. The numerical scheme used represents the current state of the art in the solution of non-linear systems: the Newton–Krylov algorithm. Several calculations are presented to demonstrate the ability of the methods described here to study the behavior of molten salt reactors in both steady state and transient situations.     

熔盐快堆U-Pu燃料循环增殖性能分析

熔盐快堆增殖是当前国际上关注的热点,本文基于堆芯结构双流体方案,利用氟化或氯化熔盐中铀钚重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。对铀钚燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+PuF_4+UF_4、NaF+PuF_4+UF_4和NaCl+PuCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数。分析了增殖比BR(Breeding Ratio)受反应堆裂变区、增殖区和中子反射层的尺寸影响,熔盐中~6Li和~(35)Cl同位素丰度对BR的影响,以及BR随运行时间动态变化。计算结果表明:氯盐方案(BR=1.46)与两种氟盐方案(BR≈1.06)相比较,具有更大的增殖能力优势。结合熔盐相图、BR随重金属摩尔浓度变化和BR最大值随熔盐平均工作温度变化曲线,可以在熔盐快堆设计中快速确定熔盐的工作温度、重金属摩尔浓度和反应堆增殖比。     

熔盐快堆Th-U燃料循环增殖性能分析

熔盐快堆是当前国际上关注的热点之一,本文基于堆芯结构双流体方案,即裂变熔盐燃料和增殖熔盐介质各自独立冷却循环,利用氟化或氯化熔盐中钍铀重金属盐高温下的高溶解度特性,获得熔盐快堆的高增殖。通过比较钍铀燃料循环熔盐快堆的三种可行性熔盐燃料方案(LiF+ThF_4+UF_4、NaF+ThF_4+UF_4和NaCl+ThCl_3+UCl_3),采用基于反应堆安全分析和设计的综合性模拟程序SCALE(Standardized Computer Analyses for Licensing Evaluation),计算了中子能谱、反应性温度系数,分析了增殖比BR(breeding ratio)受反应堆裂变区、增殖区和ZrC中子反射层的尺寸影响、熔盐中~6Li和~(35)Cl同位素丰度的影响,以及熔盐密度误差对BR计算值的准确性影响、易裂变核素随反应堆运行时间演化等。在钍铀燃料循环熔盐快堆中,通过优化处理得到三种熔盐燃料方案的增殖比BR约为1.2。     

FLiNaK熔盐中CsF的蒸发与分离

Cs的同位素是核裂变的主要产物之一,在熔盐反应堆液态燃料盐中以Cs F的化学形态存在,定期从燃料盐中除去或减少其含量将有助于提高反应堆的中子经济性。本文用FLi Na K熔盐模拟熔盐堆载体盐FLi Be体系,研究了Cs F在不同蒸发条件下的蒸发行为,并尝试进行了减压蒸馏和金属Li还原蒸发技术分离Cs F的实验研究。研究表明,在5 Pa蒸馏压力下,Cs F的蒸发量随温度呈线性上升趋势,780oC时Cs F的含量由1%降到0.14%,分离率达86%,但此时载体盐的蒸发量达9.5%;在常压、700oC条件下,熔盐中Cs F的蒸发比例随还原剂Li的添加量而提高,当添加的金属Li的摩尔浓度与Cs F为120:1时,Cs F分离率达91%。研究结果为了解Cs F在氟盐体系中的蒸发行为和建立可行的分离方法提供基础实验依据。     

DYN1D-MSR dynamics code for molten salt reactors

This paper reports about the DYN1D-MSR code development and dynamics studies of the molten salt reactors (MSR) – one of the ‘Generation IV International Forum’ concepts. In this forum the graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered.The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution.In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip).