主产裂变99Mo的研究堆初步热工安全设计

针对主产裂变⁹⁹Mo的研究堆的需求,设计了堆芯总体方案。根据中子物理计算功率分布和该反应堆的热工水力设计限值,应用COTH程序确定了反应堆的热工水力设计参数。通过结果分析可知,该堆芯设计方案可满足设计要求。     

101Moγ射线发射几率测定

¹⁰¹Mo是裂变燃耗诊断的重要核素之一，它的半衰期短，准确测定¹⁰¹Mo核衰变数据的难度大。本工作采用强度平衡法和全程记录法测量¹⁰¹Mo和¹⁰¹Tc的γ射线发射几率P_γ。¹⁰¹Mo P_γ的强度平衡法测量结果比全程记录法的测定结果高5%，而¹⁰¹TcP_γ的两种方法测定结果在不确定度范围内一致。有关¹⁰¹Mo的γ射线发射几率测量方法及其测定值有待进行进一步研究和测定。      

99Tcm-BnAO-乙二醇独甲醚的合成及其生物分布

⁹⁹Tc^m-BnAO(HL91)是一种非硝基咪唑类乏氧显像剂,为了提高其乏氧显像性能,合成了BnAO-乙二醇独甲醚,并对其进行了⁹⁹Tc^m标记。结果表明,⁹⁹Tc^m-BnAO及⁹⁹Tc^m-BnAO-乙二醇独甲醚均为电中性,⁹⁹Tc^m-BnAO-乙二醇独甲醚脂溶性有所提高。其荷S180瘤小鼠体内生物分布结果表明,随着时间的延长,⁹⁹Tc^m-BnAO及⁹⁹Tc^m-BnAO-乙二醇独甲醚在肿瘤组织的相对摄取率逐渐增加。但⁹⁹Tc^m-BnAO-乙二醇独甲醚的肿瘤与血和肿瘤与肌肉的放射性摄取比（T/NT）较⁹⁹Tc^m-BnAO略有降低。     

99Tcm-DTPA-DG标记物的制备

为了在合成二乙三胺五乙酸-脱氧葡萄糖（DTPA-DG）基础上，制备⁹⁹Tc^m-DTPA-DG标记物，以SnCl₂·2H₂O为还原剂，还原⁹⁹Tc^mO^-₄并与DTPA-DG形成⁹⁹Tc^m-DTPA-DG标记物。进行了DTPA-DG用量、SnCl₂·2H₂O用量、反应介质pH值、反应温度等对标记率的影响实验。结果表明，25mg DTPA-DG ，500μg SnCl₂·2H₂O，pH=6，加入Na⁹⁹Tc^mO⁴淋洗液0.5~4mL，在25℃以上放置30min或沸水浴反应10min时，用9g/L NaCl和丙酮作展开剂纸层析法鉴定标记物,放射化学纯度>99%。标记物在室温放置6h，放射化学纯度仍达98.6%。⁹⁹Tc^m-DTPA-DG标记率高，标记物的体外稳定性好，操作简便，便于临床应用。     

前哨淋巴结显像剂99Tcm-rituximab体外特性和安全限度的实验研究

用⁹⁹Tc^m标记rituximab（美罗华）评价特异性前哨淋巴结（SLN）示踪剂⁹⁹Tc^m-rituximab体外特性及其体内应用的安全性。采用聚丙烯酰胺凝胶电泳（SDS PAGE）检测了⁹⁹Tc^m-rituximab的分子完整性;采用间接酶联免疫荧光分析（ELISA）及流式细胞术检测了⁹⁹Tc^m-rituximab的免疫活性;最后依据中华药典的要求行⁹⁹Tc^m-rituximab安全限度及在正常小鼠体内分布实验。结果显示：⁹⁹Tc^m-rituximab标记率为90%～95%;标记化合物分子完整,免疫活性保留完全,标记化合物无菌无热源。受试小鼠以60 mg/kg剂量注射⁹⁹Tc^m-rituximab（相当于人体用量的500倍）,1周内未见小鼠死亡。小鼠体内分布结果显示：⁹⁹Tc^m-rituximab入血后主要通过肾脏排泄,放射性摄取值由1 h的(14.01±0.61)%ID/g减少为24 h的(3.51±0.48)%ID/g;肝脏亦可见摄取。各器官未见有明显的放射性滞留。因此,⁹⁹Tc^m-rituximab结构及性能稳定,无急性毒性,使用安全,可应用于临床。     

Re/99Tcm(CO)3+黄酮复合物的合成及其与Aβ斑块的结合特性

为研制新型⁹⁹Tc^m(CO)₃⁺标记的黄酮类Aβ显像剂,设计合成了同型Re/⁹⁹Tc^m(CO)₃⁺黄酮类衍生物,用荧光法研究了Re(CO)₃⁺黄酮类衍生物在体外与Aβ斑块的结合特性,并初步观察了其在昆明小鼠体内的生物分布。结果表明,Re(CO)₃⁺黄酮类衍生物的亲和常数(K_d=5.43 nmol/L)比放射性碘标记的黄酮类衍生物高。正常小鼠的动物分布结果表明,2 min内脑初始摄取较高（0.46±0.23 %ID/g）,且清除较快（120 min时为0.13±0.04 %ID/g)。以上结果表明,⁹⁹Tc^m(CO)₃⁺黄酮类衍生物有进一步研究的价值。     

99Tcm(CO)3-CNRGD的制备及生物学评价

合成了含异腈基团的多肽偶联物（CNRGD）,并用［⁹⁹Tc^m(CO)₃(H₂O)₃］⁺标记,得到具有与整合素α_vβ₃受体多结合位点的⁹⁹Tc^m(CO)₃-CNRGD,并对其进行了体内外生物学评价。结果表明,在优化的标记条件下,⁹⁹Tc^m(CO)₃-CNRGD的标记率达到77%,纯化后,标记物放射化学纯度大于96%。体外稳定性实验显示其具有很高的稳定性;脂水分配系数显示其具有较好的脂溶性。正常小鼠体内分布显示,⁹⁹Tc^m(CO)₃-CNRGD在血液中清除较快,主要通过肝肾代谢。荷MCF-7人乳腺癌裸鼠体内分布显示,注射1、4h后,标记物在肿瘤部位的摄取值达(2.38±0.37)%ID/g和(1.57±0.21)%ID/g,瘤/血比分别达0.71±0.09、1.15±0.15,表明该标记物在肿瘤细胞中有一定的摄取和较长的滞留时间。     

99Tcm标记NGR及其在荷人HePG2肝癌裸鼠体内的生物分布及SPECT显像

用⁹⁹Tc^m标记了含有天冬酰胺-甘氨酸-精氨酸（Asn-Gly-Arg）序列的血管靶向性短肽NGR,评价了标记物⁹⁹Tc^m-NGR的放化性质以及在荷HePG2肝癌模型裸鼠体内的生物分布和SPECT显像。标记结果显示,⁹⁹Tc^m-NGR的标记率>90%,放化纯度>95%。荷瘤裸鼠体内生物分布结果显示,⁹⁹Tc^m-NGR在肾脏和肝脏的摄取率较高,注射后1 h肿瘤摄取达（2.52±0.62）%ID/g,最高达（7.26±2.71）%ID/g,12 h仍然达（3.93±1.93）%ID/g,但在竞争性抑制组中摄取率为（1.29±0.85）%ID/g。荷瘤裸鼠的SPECT显像结果显示,除肿瘤外,其他组织器官的放射性摄取随时间延长逐渐降低,肿瘤与肌肉组织的放射性攝取比（T/NT）4 h时最高,可达3.25。注射后1 h肿瘤可见,12 h时最为清晰。以上结果提示,⁹⁹Tc^m-NGR易于制备,具有良好的靶向性, 在肿瘤的诊疗中具有良好的研究前景。     

糖基化生长抑素99Tcm标记及生物学评价

以天然生长抑素(Somatostatin，SMS)、葡聚糖-10(Dextran¹⁰，Dx¹⁰)及巯基乙胺（Cysteamine）为原料，合成糖基化生长抑素配体化合物SMS-Dx¹⁰-Cysteamine；以¹²⁵I-奥曲肽（¹²⁵I-Tyr₃-Octreotide）为放射性配基，进行受体竞争结合实验，测定SMS-Dx¹⁰-Cysteamine的IC₅₀值；利用葡庚糖转换络合进行SMS-Dx¹⁰-Cysteamine 的⁹⁹Tc^m标记，重点探讨⁹⁹Tc^m标记条件及标记物体外稳定性；并用⁹⁹Tc^m-Cysteamine-Dx¹⁰-SMS进行正常SD大鼠体内分布、血浆清除及肿瘤模型动物显像实验。结果表明：配体化合物SMS-Dx¹⁰-Cysteamine保持了对生长抑素2型受体高亲和力，其IC₅₀值与SMS相近；在最佳标记条件：0.3 g/L SnCl₂，5 g/L SMS-Dx¹⁰-Cysteamine，标记介质pH=6.0，室温下反应30 min，⁹⁹Tc^m-Cysteamine-Dx¹⁰-SMS标记率约为85%，经分离纯化后，其放射化学纯度大于99%，标记物体外稳定；⁹⁹Tc^m-Cysteamine-Dx¹⁰-SMS在正常大鼠体内血浆半衰期为2.38 h，主要浓聚于肝、脾脏并经肾排泄；荷胰腺癌裸鼠显像表明，注射后6 h，肿瘤组织具有明显的放射性摄取，⁹⁹Tc^m-Cysteamine-Dx¹⁰-SMS有望成为一种生长抑素受体阳性肿瘤显像剂。     

特异性前哨淋巴结显像剂99Tcm-rituximab药盒的制备及生物评价

采用2-巯基乙醇修饰Rituximab（利妥昔单克隆抗体）,制得其冻干药盒。所制得的冻干药盒性质稳定,可在-20 ℃冷冻保存3个月以上。利用⁹⁹Tc^m-葡庚糖酸钠（GH）交换法,标记冻干药盒得到⁹⁹Tc^m-rituximab,标记率和放化纯度均大于90%。生物分布结果显示：SD大鼠经前脚掌皮下注射⁹⁹Tc^m-rituximab后,前哨淋巴结的摄取值在1~4 h内逐渐增加,在4 h时达到（2.14±0.46）%ID,前哨淋巴结与注射点的放射性摄取比在4 h时达到最大（14.80%±2.11%）,且在18 h时,这一比值基本保持不变。SPECT显像结果表明,在30 min~18 h内,大鼠的前哨淋巴结清晰可见,无次级淋巴结显影。本研究提供了一种特异性前哨淋巴结显像剂的药盒化制备方法,该方法操作简便,性能稳定,便于在临床推广应用。     

New method of producing <Superscript>99</Superscript>Mo in molten-salt fluoride fuel

The status and the problems of world ⁹⁹Mo production are presented. A comparative analysis is made of reactor methods of ⁹⁹Mo production. It is noted that the currently used technologies and research reactors are not satisfying the growing demand in medicine for this isotope. It is underscored that the role of alternative production technologies has grown. In the development of new ⁹⁹Mo production technologies, the experimental results obtained on the basis of research program conducted on the MSRE reactor with molten-salt fluoride fuel have been analyzed. The analysis revealed a special behavior of certain fission products including ⁹⁹Mo: they leave the melt spontaneously and enter the gas phase. The authors hypothesize that highly volatile fluorides of the indicated products are formed in the melt; this explains the effect indicated. The effect is used as a basis to propose a new reactor method of producing fissionproduced ⁹⁹Mo. Concrete examples of a way to implement the new method of producing fission-produced ⁹⁹Mo using molten-salt fluoride nuclear fuel are presented.     

应用溶液堆生产裂变99Mo

裂变^99Mo的供应对保障核医学的应用和发展具有非常重要的作用。应用溶液堆生产裂变^99Mo具有运行安全、没有靶件制备、溶解工艺和生产工艺相对较简单、消耗铀燃料较少的优点。因此应用溶液堆生产裂变^99Mo具有很好的应用前景。     

中国先进研究堆(CARR)的设计特点和创新技术

中国先进研究堆(CARR)是一座多用途、高性能指标的研究堆。本文就其先进技术和设计理念的采用、系统设备的布局、运行模式的确定和应用等设计特点进行了归纳。这些设计特点,体现了CARR设计的先进性、安全性、经济性和创新性,是我国研究堆设计水平的具体体现。CARR的建成和投入运行,将成为中国原子能科学研究院的发展提供重要的科研平台。     

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.     

Obtaining molybdenum-99 in the IRT-T research reactor using resonance neutrons

The yield of ⁹⁹Mo from the ⁹⁸Mo(n,γ)⁹⁹Mo reaction significantly depends of the energy spectrum of the neutron flux. It is well known that the cross-section for this reaction is about 130 mb, whereas the resonance integral of the reaction is 6.9 b. The aim of this work was to investigate the conditions that let to increase ⁹⁹Mo yield from the targets with natural and enriched isotope composition under irradiation by resonance neutrons at the IRT-T research reactor.The calculations of integrated cross-sections of all Mo isotopes in the region of the ⁹⁸Mo resonances showed that screening in the target with natural isotope composition by other isotopes is relatively small. So the ⁹⁸Mo in the natural mixture can be activated by resonance neutrons approximately in the same manner as pure ⁹⁸Mo.Experimental measurements of the ⁹⁸Mo(n,γ) effective cross-section using the MoO₃ sample with natural and enriched composition in the reactor channels with the beryllium moderator with the thickness of 20 up to 90 mm showed that the effective cross-sections in these channels reach the value of 700 mb. The contribution of the epithermal neutrons into the ⁹⁸Mo activity was 68% for the enriched targets and 78% for natural molybdenum, respectively.At that channel it is possible to produce ⁹⁹Mo with specific activity up to 3.4 Cu/g with samples of natural isotope composition and up to 15 Cu/g with enriched samples on the base of reactors with neutron flux of (1.7 × 10¹⁴ n/(cm² s)). Such ⁹⁹Mo specific activity is enough not only to realize extraction technologies production of ^99mTc, but to manufacture sorption generators of ^99mTc without wastes.     

Improving SFR economics through innovations from thermal design and analysis aspects

Achieving economic competitiveness as compared to LWRs and other Generation IV (Gen-IV) reactors is one of the major requirements to attract large-scale investment in commercial sodium cooled fast reactor (SFR) power plants. Advances in R&D for advanced SFR fuel and structural materials provide key long-term opportunities to improve SFR economics. In addition, other new opportunities are emerging to further improve SFR economics. This paper provides an overview on potential ideas from the perspective of thermal hydraulics to improve SFR economics. These include: (1) a new hybrid loop-pool reactor design to further optimize economics, safety, and reliability of SFRs with more flexibility, (2) a multiple-reheat and intercooling helium Brayton cycle to improve plant thermal efficiency and to reduce safety related overnight and operation costs, and (3) modern multi-physics thermal analysis methods to reduce analysis uncertainties and associated requirements for over-conservatism in reactor design. This paper reviews advances in all three areas and their potential beneficial impacts on SFR economics.     

Thermodynamic considerations and prediction of the primary coolant activity of Tc

A physical model has been developed to describe the coolant activity behaviour of ⁹⁹Tc, during constant and reactor shutdown operations. This analysis accounts for the fission production of technetium and molybdenum, in which their chemical form and volatility is determined by a thermodynamic treatment using Gibbs-energy minimization. The release kinetics are calculated according to the rate-controlling step of diffusional transport in the fuel matrix and vaporization from the fuel-grain surface. Based on several in-reactor tests with defective fuel elements, and as supported by the thermodynamic analysis, the model accounts for the washout of molybdenum from the defective fuel on reactor shutdown. The model also considers the recoil release of both ⁹⁹Mo and ⁹⁹Tc from uranium contamination, as well as a corrosion source due to activation of ⁹⁸Mo. The model has provided an estimate of the activity ratio ⁹⁹Tc/¹³⁷Cs in the ion-exchange columns of the Darlington Nuclear Generating Station, i.e., 6 × 10⁻⁶ (following ∼200 days of steady reactor operation) and 4 × 10⁻⁶ (with reactor shutdown). These results are consistent with that measured by the Battelle Pacific Northwest Laboratories with a mixed-bed resin-sampling device installed in a number of Pressurized Water Reactor and Boiling Water Reactor plants.     

ARIES-AT magnet systems

This report presents a conceptual design of the magnet systems for an advanced tokamak fusion reactor (ARIES-AT). The main focus of the paper is to anticipate and extrapolate the current state-of-the-art in high temperature superconductors and coil design, and apply them to an advanced commercial fusion reactor concept. The current design point is described and supported with a preliminary structural analysis and a discussion of the merits, performance, and economics of high temperature vs. low temperature superconductors in an advanced fusion reactor design.     

Mo在Al2O3上的吸附、解吸性能及其与U、Sr、Cs、I等杂质元素的分离

目前国际⁹⁹Mo面临供应危机,急需新技术和新反应堆。医用同位素生产堆是以²³⁵UO₂(NO₃)₂溶液为燃料的专用反应堆,生产成本低、三废少、经济效益高。本工作利用Al₂O₃为分离材料,从模拟的医用同位素生产堆（MIPR）燃料溶液中分离和纯化Mo。结果表明,经两次分离,Mo的总回收率大于 60%, U、Sr、Cs、I等杂质可以被除去,采用Al₂O₃从MIPR燃料溶液中提取⁹⁹Mo工艺可行。