电沉积 LEU UO2 靶件生产医用99Mo的工艺研究

⁹⁹Mo是一种重要的医用放射性同位素。采用低浓铀（LEU）靶件生产裂变⁹⁹Mo是发展趋势。本工作进行了电沉积UO₂靶件制备、靶件溶解以及⁹⁹Mo化学分离等工艺研究,确定了电沉积LEU UO₂靶件制备医用裂变⁹⁹Mo的工艺流程。研究表明,于不锈钢管内壁上电沉积UO₂,在pH=7、电流0.5~2 mA/cm²、温度75~90 ℃、镀液中U浓度5 mg/mL条件下,经过约210 h电沉积,不锈钢管内壁上UO₂沉积层质量达到42 mg/cm²;采用6 mol/L HNO₃溶解UO₂镀层。采用α-安息香肟沉淀法实现⁹⁹Mo与大量裂变产物的初步分离,采用阴离子交换法与活性炭色层法联用实现⁹⁹Mo的纯化;纯化后的⁹⁹Mo溶液中,杂质¹³¹I、⁹⁰Sr、⁹⁵Zr、¹⁰³Ru、²³⁸U活度与⁹⁹Mo活度比值分别为4.47×10^-6%、7.40×10^-7%、8.67×10^-7%、2.57×10^-6%、1.69×10^-14%,均小于《欧洲药典》规定值,满足医用要求。本工作建立了电沉积LEU UO₂靶件生产高纯医用裂变⁹⁹Mo的工艺流程,为今后采用LEU技术生产医用裂变⁹⁹Mo,进而实现其自主规模化生产打下了基础。     

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%。     

热中子诱发239Pu裂变的85Krm、87Kr和88Kr产额测量研究

中子诱发²³⁹Pu裂变的⁸⁵Kr^m、⁸⁷Kr和⁸⁸Kr的产额是重要的核参数,目前国外实验数据较少而国内尚未见实验报道。基于西安脉冲堆跑兔系统辐照Pu靶开展了热中子诱发²³⁹Pu裂变的⁸⁵Kr^m、⁸⁷Kr和⁸⁸Kr的产额测量研究。纯化后的钚溶液通过滴定后阴干的方式制靶,靶辐照后结合γ无损分析和气-固分离制源测量等方式测量裂变产物。采用有机玻璃扁平面源等效石英管源、不锈钢大面源等效气体源,并结合蒙特卡罗模拟实现了3类实验样品的γ能峰探测效率曲线的等效法刻度。以相同方式制备的²³⁵U靶开展气-固分离制源实验验证了钚靶中⁸⁵Kr^m、⁸⁷Kr和⁸⁸Kr气体释放率的一致性。根据实测目标产物与⁹⁹Mo的相对产额,以ENDF/BⅧ.0评价数据库中^...     

医用同位素99Mo吸附分离研究进展

医用同位素⁹⁹Mo是一种广泛应用于核医学领域的重要核素。由于常规的高浓缩铀裂变生产⁹⁹Mo的过程中存在安全隐患,人们已经开始寻找其他可靠的⁹⁹Mo生产途径。在分离⁹⁹Mo和^99mTc的方法中柱层析法具有很大优势,其中的关键是层析柱的材料,材料对⁹⁹Mo吸附能力关系到未来新一代⁹⁹Mo-^99mTc发生器的制备。本研究对医用同位素⁹⁹Mo的吸附分离进行综述,介绍⁹⁹Mo生产方式,⁹⁹Mo和^99mTc分离方法 ,以及目前对Mo具有一定吸附效果的吸附材料,为未来利用低比活度⁹⁹Mo吸附制备⁹⁹Mo-^99mTc发生器提供参考。     

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。     

DRAGON用于反应堆中微子实验的基准验证

反应堆中微子实验中需计算输出²³⁵U、²³⁸U、²³⁹Pu、²⁴¹Pu在不同燃耗下的裂变份额,而通常的组件计算程序不输出这些结果。为适应反应堆中微子实验的需求,本文用Takahama-3基准对DRAGON用于压水堆燃耗进行基准验证,给出了反应堆中微子实验中关心的4种核素质量密度实验值与计算值的平均偏差,并利用计算出的裂变份额以及每次裂变释放的能量等,给出了NT3G24组件的反应堆中微子能谱,从而验证了DRAGON应用于反应堆中微子实验的可行性。     

基于核试验监测数据的模拟γ能谱分析国际能力验证

本文介绍了全面禁止核试验条约（CTBT）筹委会临时技术秘书处（PTS）组织的2003年度国际放射性核素实验室能力验证过程。PTS在真实核试验监测数据基础上,通过模拟产生了2003年度能力验证活动的参考γ能谱,能谱中添加了24种裂变产物、5种活化产物和5种天然放射性核素。北京放射性核素实验室分析出了其中的27种核素,核素活度及其活度浓度分析结果与参考值在不确定度范围内一致。利用⁹⁵Zr和⁹⁵Nb活度比计算了核事件的零时,与参考值仅相差0.26 d。根据参考谱中裂变产物和活化产物信息,指出裂变产物应主要由²³⁸U和²³⁹Pu裂变产生,参考谱应源自真实核试验的监测数据。     

高浓铀靶裂变法生产钼-99发展的综述

随着放射性药物化学和核医学的快速发展,放射性诊断核素^99mTc在临床的应用越来越广泛,从而使得当前全球对其母体核素⁹⁹Mo的需求量不断增加。目前高浓铀靶裂变法生产⁹⁹Mo仍是最广泛使用的方法。本文系统介绍了国内外采用高浓铀靶裂变法生产⁹⁹Mo的发展历史和生产工艺。     

0.57、1.0和1.5 MeV中子诱发235U裂变的99Mo产额绝对测量

用HPGe γ能谱法绝对测量了0.57、1.0和1.5 MeV中子诱发²³⁵U裂变产物⁹⁹Mo的产额,使用双裂变室测量了样品辐照过程中的裂变率,应用MCNP ⅣB模拟了铀样品中的中子能谱,并讨论了非主中子的各种来源对产额数据的影响。得到⁹⁹Mo在0.57、1.0和1.5 MeV的产额分别为6.61%、6.62%和6.28%。本工作与美国阿贡实验室的结果有15%以上的相对偏差,主要是由引用的衰变数据不同引起。对阿贡实验室数据进行校正后,本工作与阿贡实验室数据的相对偏差处于实验不确定度范围内。     

Al2O3色层法从低浓铀靶件中分离99Mo条件研究

在裂变⁹⁹Mo的生产工艺中,常用Al₂O₃色层法分离纯化⁹⁹Mo。为建立Al₂O₃色层法从低浓铀(LEU)靶件中分离裂变⁹⁹Mo的工艺,考察吸附时间、温度、酸度、预处理方式等对Al₂O₃吸附Mo效果的影响。研究采用Al₂O₃色层法从不同浓度HNO₃溶液中分离Mo。测定Al₂O₃色层法对Al和主要杂质元素Sr、Ru、Zr、Te、Cs、I的去污系数。研究结果表明,在0.05~0.1 mol/L HNO₃介质中Al₂O₃对Mo有出色的吸附性能,Mo吸附率在99%以上,在NH₄OH溶液中Al₂O₃不吸附Mo。经500 ℃活化3 h预处理得到的Al₂O₃-C具有更大的比表面积,且在HNO₃浓度大于0.1 mol/L时相比于150 ℃活化3 h预处理得到的Al₂O₃-B以及未经高温预处理得到的Al₂O₃-A对于Mo有更好的吸附性能。采用该工艺,通过Al₂O₃色层法从模拟的LEU靶件溶液中提取Mo,Mo回收率大于90%,Al₂O₃色层法对裂变杂质元素Ru、Sr、Zr、Te、Cs等的去除率均大于99.99%,对¹³¹I的去除率大于92%。由此可见,Al₂O₃在HNO₃介质中对Mo的吸附率高,能够有效地去除⁹⁹Mo产品中的杂质核素,适用于从低浓铀靶件中分离裂变⁹⁹Mo。     

239Pu(nth,f)短寿命裂变产物产额测量

采用γ能谱相对测量方法,以97Zr为内标参考核素,完成了239Pu(nth,f)短寿命裂变产物88 Rb、95 Y、101 Mo、101 Tc、138 Csg、142 La等核素的累积产额测量。实验测得88 Rb、95 Y、101 Mo、101 Tc、138 Csg、142 La的累积产额数据分别为(1.32±0.05)%、(4.69±0.22)%、(6.13±0.32)%、(6.10±0.22)%、(6.24±0.24)%、(4.74±0.17)%。对高纯锗探测器中高能端效率刻度、样品封装与辐照、γ谱测量几何条件设计、γ谱测量与数据分析进行了研究。实验测量裂变产物核素分别位于非对称裂变质量分布双驼峰曲线轻峰的左侧、中部和右侧,重峰的中部与右侧。     

Making99Mo in reactors

Conclusions This comparison of two ways of making⁹⁹Mo shows that there is good scope for making it in any area. If there is a thermal reactor having a high flux, one can make⁹⁹Mo from⁹⁸Mo, and in that case, even irradiation in a high flux makes it favorable to use highly enriched⁹⁸Mo and unblocked targets, which raises the specific activity and thus increases the working life in the^99mTc generator.If there is a thermal reactor with low flux or if there is a fast reactor it is best to make⁹⁹Mo from uranium fission products. The target can be highly enriched²³⁵U, which if necessary can be reused, or low-enriched uranium.There are no essential constraints on making⁹⁹Mo, and the production is mainly based on technological tasks.Translated from Atomnaya Énergiya, Vol. 67, No. 2, pp. 104–108, August, 1989.     

Performance of the gas turbine-modular helium reactor fuelled with different types of fertile TRISO particles

Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with ²³⁹Pu, ²³³U or ²³⁵U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of ²³²Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes (²³⁹Pu, ²³³U and ²³⁵U), requires a quantity of thorium 25–30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, ²³³U, ²³⁵U and ²³⁹Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.     

Low enriched uranium foil plate target for the production of fission Molybdenum-99 in Pakistan Research Reactor-1

Low enriched uranium foil (19.99% ²³⁵U) will be used as target material for the production of fission Molybdenum-99 in Pakistan Research Reactor-1 (PARR-1). LEU foil plate target proposed by University of Missouri Research Reactor (MURR) will be irradiated in PARR-1 for the production of 100Ci of Molybdenum-99 at the end of irradiation, which will be sufficient to prepare required ⁹⁹Mo/^99mTc generators at Pakistan Institute of Nuclear Science and Technology, Islamabad (PINSTECH) and its supply in the country. Neutronic and thermal hydraulic analysis for the fission Molybdenum-99 production at PARR-1 has been performed. Power levels in target foil plates and their corresponding irradiation time durations were initially determined by neutronic analysis to have the required neutron fluence. Finally, the thermal hydraulic analysis has been carried out for the proposed design of the target holder using LEU foil plates for fission Molybdenum-99 production at PARR-1. Data shows that LEU foil plate targets can be safely irradiated in PARR-1 for production of desired amount of fission Molybdenum-99.     

从第五不稳定核素系的特性论核燃料的有效利用

以第五不稳定核素系在生长时期的耗裂转化比不断增长的特性以及不同堆型中它的任－衍生核素的饱和含量比值并不相同的特性作为论述的依据，得出了目前从压水堆中取出的废燃料并未获得有效充分利用的论断。认为只须对废燃料经过去除裂变产物的后处理去污流程，即要重新作为动力堆的核燃料使用，避免了使铀钚分离以及＾２８５Ｕ再度浓集的流程。并对如何使用此再制的核燃料提出两种方案，分别适用于压水堆和以天然铀为燃料的坎杜重水堆     

钚及其裂变产物钯 、银 、镉 、锡 、锑 、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组,然后再针对各组目标元素进行分离和纯化,可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证,结果表明,分离流程对6种裂变产物的化学回收率均大于70%,对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³,可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

钚及其裂变产物钯、银、镉、锡、锑、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组，然后再针对各组目标元素进行分离和纯化，可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证，结果表明，分离流程对6种裂变产物的化学回收率均大于70%，对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³，可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

Fission-product yields from neutron-induced fission

Exsting experimental thermal, fast, and 14-MeV neutron-induced fission-product cumulative and independent yieds have been compiled, corrected to common reference values, and listed in tabular form for the following fissile nuclides:Thermal-neutron fission: cumulative yields for ²²⁷Th, ²²⁹Th, ²³³U, ²³⁵U, ²³⁹Pu, ²⁴¹Pu, ²⁴¹Am, ²⁴²Am, ²⁴⁵Cm, ²⁴⁹Cf, ²⁵¹Cf, ²⁵⁴Es, and ²⁵⁵Fm; independent yieds for ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu.Fast-neutron fission: cumulativ yields for ²²⁷Ac, ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³⁵U and ²³⁸U.14-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, ²³³U, ²³⁵U, ²³⁷Np, ²³⁸U, and ²³⁹Pu; independent yields for ²³²Th, ²³³U, ²³⁵U, ²³⁸U, and ²³⁹Pu.11-MeV-neutron fission: cumulative yields for ²³²Th.3-MeV-neutron fission: cumulative yields for ²³¹Pa, ²³²Th, and ²³⁸U.1.1-MeV-neutron fission: cumulative yields for ²³⁷Np.From these experimental values the unknown independent yields are deduced empirically for thermal-neutron fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; the fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu, and ²⁴¹Pu (the chain yields for ²⁴⁰Pu and ²⁴¹Pu used at this energy being predictions); and the 14-MeV-neutron fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U.Finally, by the fitting of the preceding information to condition equations derived from the conservation laws, adjusted sets of chain and independent yields are calculated for thermal fission of ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu; fast fission of ²³²Th, ²³³U, ²³⁵U, ²³⁸U, ²³⁹Pu, and ²⁴¹Pu; and 14-MeV fission of ²³²Th, ²³³U, ²³⁵U, and ²³⁸U. The literature search is probably complete to the end of 1975; some 1976 results are included.This paper replaces and makes obsolete the following UKAEA reports: AERE-R7209, AERE-R7394, AERE-R7680, and AERE-R8152.     

Inherent Protection of Plutonium by Doping Minor Actinide in Thermal Neutron Spectra

The present study focuses on the exploration of the effect of minor actinide (MA) addition into uranium oxide fuels of different enrichment (5% ²³⁵U and 20% ²³⁵U) as ways of increasing fraction of even-mass-number plutonium isotopes. Among plutonium isotopes, ²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu have the characteristics of relatively high decay heat and spontaneous fission neutron rate that can improve proliferation-resistant properties of a plutonium composition. Two doping options were proposed, i.e. doping of all MA elements (Np, Am and Cm) and doping of only Np to observe their effect on plutonium proliferation-resistant properties. Pressurized water reactor geometry has been chosen for fuels irradiation environment where irradiation has been extended beyond critical to explore the subcritical system potential. Results indicate that a large amount of MA doping within subcritical operation highly improves the proliferation-resistant properties of the plutonium with high total plutonium production. Doping of 1% MA or Np into 5% ²³⁵U enriched uranium fuel appears possible for critical operation of the current commercial light water reactor with reasonable improvement in the plutonium proliferation-resistant properties.