次锕系核素在铅冷快堆中的嬗变性能

乏燃料中大部分次锕系（minor actinides, MA）核素半衰期较长,对环境具有长期放射性危害。分离 嬗变技术将次锕系核素从高放废液中分离出来,并通过反应堆嬗变为短寿命或稳定核素,从而消除其放射性危害。为研究次锕系核素与燃料均匀混合、制成嬗变棒和做燃料芯块镀层装载方式下在铅冷快堆中的嬗变特性,采用MCNP和SCALE程序进行模拟计算。结果表明,三种方式下²³⁷Np、²⁴¹Am、²⁴³Am和混合次锕系核素使有效增殖因数k_eff降低,而²⁴⁴Cm和²⁴⁵Cm使k_eff升高,且²⁴⁵Cm可使k_eff大幅度增加。不同质量的混合次锕系核素装载后,三种方式下堆芯k_eff都随装载量的增加而降低,降低幅度由小到大分别为嬗变棒、均匀混合和镀层。不同次锕系核素装载量以均匀混合方式在堆芯经过550 d辐照后,²³⁷Np、²⁴¹Am和²⁴³Am嬗变率均为正值,其中²⁴¹Am嬗变率最大,而²⁴⁴Cm和²⁴⁵Cm嬗变率均为负值,²⁴⁵Cm增加明显,总的次锕系核素嬗变率为14%,可为次锕系核素在铅冷快堆中嬗变性能评价提供参考。     

Cyclic mode of neptunium, americium and curium transmutation in heavy-water reactor

Characteristics of process of transmutation of neptunium, americium and curium from spent nuclear fuel in heavy-water reactor during first 10 lifetimes and at transition to equilibrium mode are calculated. During transmutation, dangerous nuclides, first of all, ²⁴⁴Cm and ²³⁸Pu are accumulated. They cause an increase of radiotoxicity. At first 10 cycles of transmutation, the radiotoxicity is increased by 8.7 times in comparison with radiotoxicity of initial load of transmuted actinides. Heavy-water reactor with thermal power of 1000 MW can transmute neptunium, americium and curium extracted from 3.7 VVER-1000 type reactors. It means, that the required power of transmutation reactor makes about 8% of thermal power of VVER-1000 type reactors.     

Highly enriched isotopes of uranium and transuranium elements for scientific investigation

The paper describes the production of highly enriched isotopes of uranium, plutonium, americium and curium by means of electromagnetic separation for scientific and applied research in physics, chemistry, geology and other fields. The equipment and radiochemical methods used allows to provide the isotopic pure samples in quantities sufficient to set up nuclear physics experiments, to produce reference materials and standard sources for calibration of radiometrical and mass spectrometrical equipment and for use in radionuclear metrology. For a series of nuclei unique characteristics of isotopic enrichment and radiochemical and chemical purity were achieved: ²³³U: 99.97%; ²³⁵U: 99.97%; ²³⁶U: 98.0%; ²³⁸U: 99.997%; ²³⁸Pu: 99.6%; ²³⁹Pu: 99.9977%; ²⁴⁰Pu: 99.9–100%; ²⁴¹Pu: 96.998%; ²⁴²Pu: 97.8–99.96%; ²⁴⁴Pu: 96.7%; ²⁴¹Am: 99.6%; ^242mAm: 85.6%; ²⁴³Am: 99.2–99.94%; ²⁴³Cm: 99.99%; ²⁴⁵Cm: 99.998%; ²⁴⁶Cm: 99.8%; ²⁴⁷Cm: 90%; ²⁴⁸Cm: 97%. Methods of radiochemical and chemical separation, product certification, fabrication of special sources or targets and layers of highly enriched isotopes on various substrates are presented.     

Improvement of plutonium proliferation resistance by doping minor actinides

This paper focuses on improving the proliferation resistance of plutonium resulting from uranium-based fuel irradiation. Intrinsic properties of plutonium isotopes with even mass numbers (²³⁸Pu, ²⁴⁰Pu and ²⁴²Pu) — in terms of their intense decay heat and high spontaneous fission neutron rates — were used as a measure to improve the proliferation resistance of plutonium itself. The present study explores MA addition effect into LEU (5%²³⁵U) and HEU (20%²³⁵U) with regard to plutonium proliferation resistance characteristics. Consideration goes beyond critical condition to examine the potential of subcritical system in enhancing the plutonium proliferation properties. Results show that even the doping level of 1% of Np, TrPu or all MA elements into low enriched uranium improves the proliferation-resistant properties of plutonium. A potential for further improvement is achieved by higher doping of minor actinides into high enriched uranium irradiated in a subcritical mode.     

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。     

Performance of natural uranium- and thorium-fueled fast breeder reactors (FBRs) for U fissile production

The performance of natural uranium and thorium-fueled fast breeder reactors (FBRs) for producing ²³³U fissile material, which does not exist in nature, is investigated. It is recognized that excess neutrons from FBRs with good neutron economic characteristics can be efficiently used for producing ²³³U. Two distinct metallic fuel pins, one with natural uranium and another with natural thorium, are loaded into a large sodium-cooled FBR. ²³³U and the associated-U isotopes are extracted from the thorium fuel pins. The FBR itself is self-sustained by plutonium produced in the uranium fuel pins. Under the equilibrium state, both uranium and thorium spent fuels are periodically discharged with a certain discharge rate and then separated. All discharged fission products are removed and all discharged actinides are returned to the FBRs except the discharged uranium utilized for fresh fuel of the other thorium-cycled reactors. ²³³U-production rate of the FBRs as a function of both the uranium–thorium fuel pins fraction in the core and the discharge fuel burnup is estimated. The result shows that larger fraction of uranium pins is better for the FBR criticality while larger fraction of thorium fuel pins and lower fuel burnup give higher ²³³U production rate.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU⁻¹. Characteristic test of this device was done with a ²⁵²Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU⁻¹ at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.     

乏燃料废包壳残留物质核素分析计算

为分析计算乏燃料废包壳残留物质的核素含量，以M310型核电机组及燃料组件为分析对象，建立了乏燃料废包壳残留物质核素含量分层计算模型，用SCALE程序计算分析了²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比等主要参数随燃耗及冷却时间的变化。计算结果表明，²⁴⁴Cm含量、总Pu含量及²⁴⁴Cm/Pu比随燃耗及冷却时间的变化均可用三阶多项式拟合。本文工作为废包壳残留物质非破坏性测量方法研究提供了数据支持。     

可燃毒物提高小型压水堆堆芯寿期研究

针对长寿期堆芯的应用需求,开展了提高小型压水堆堆芯寿期研究。以棒状燃料为对象,对不同栅格尺寸和不同可燃毒物的选取进行计算,得出小型压水堆堆芯寿期相关影响因素。通过对不同尺寸的燃料栅格进行输运 燃耗计算,得到燃耗最佳栅格尺寸。以燃耗最佳栅格尺寸建立组件,并选择转换性能好的锕系核素²⁴⁰PuO₂作为可燃毒物,利用²⁴⁰Pu吸收中子转换成易裂变核素²⁴¹Pu的特性,对堆芯实现反应性控制和寿期延长。本研究通过对燃料栅格尺寸和可燃毒物的合理选择,提高了燃料利用率,达到延长堆芯寿期的目的。     

高钚含量样品中α能谱法分析~(237)Np的可行性研究

通过对244 Cm的α实验谱进行拟合得到单能峰的峰形参数,采用随机抽样技术表征谱计数的统计涨落,建立了一种模拟半导体α能谱的方法。利用该方法模拟238Pu和243Am的α能谱,与实验谱基本吻合,证明了方法的可靠性。在此基础上,研究了239Pu对237Np的α能峰的影响,结果表明,当239Pu与237Np的活度比A(239Pu)/A(237Np)≤10时,通过解谱得到的A(239Pu)/A(237Np)与设定值的相对偏差≤2.0%。对于A(239Pu)/A(237Np)约为3 000的样品,如果对钚的去污系数达到300以上,则可由α能谱法测量样品中的237 Np。     

Build-Up and Decay of Actinide Nuclides in Fuel Cycle of Nuclear Reactors

In order to assess the feasibility of utilizing plutonium in thermal reactors, build-up and decay of actinide nuclides have been studied for BWR, PWR, HWR, HTGR and LMFBR, which are uranium-oxide fueled or mixed-oxide fueled, and which produce electric power of 1,000MW. The following items were examined;

1. quantities of actinide nuclides build-up in the reactor

2. build-up and decay of activities of actinides in the spent fuel

3. build-up and decay of activities of actinides after reprocessing, and

4. variation of isotopie composition of plutonium with high burn-up.

It is concluded from the calculated results that precautions should be taken against high activities of resultant actinides if plutonium is utilized as a fissile material for thermal reactors. To make reprocessing and high-level waste management easy and practical, it is recommended that a thermal reactor should be fueled with uranium, the plutonium produced in a thermal reactor should be used in a fast reactor, and plutonium produced in the blanket of a fast reactor is more appropriate as fast reactor fuel than that from a thermal reactor.     

Multi-recycling in the ENHS and the equilibrium core

This work investigates the effect of initial fuel composition, power density and number of recycles on the pitch-to-diameter (P/D) ratio and TRans-Uranium isotopes (TRU) loading required for attaining one of the most important design goals of the Encapsulated Nuclear Heat Source (ENHS) – nearly zero burnup reactivity swing over the 20 years of core life. It is found that the required P/D ratio is sensitive to, primarily, the initial concentration of the short-lived isotope ²⁴¹Pu in the fuel loaded into the first core and to the core power density. The longer is the cooling time of the TRU from LWR spent fuel the smaller becomes the relative ²⁴¹Pu concentration and the smaller becomes the fraction of ²⁴¹Pu lost via radioactive decay and, hence, the smaller needs be the conversion ratio required for nearly zero burnup reactivity swing and the larger can be the P/D ratio. Likewise, the higher is the ENHS power density, the smaller becomes the fraction of ²⁴¹Pu lost via radioactive decay and the larger becomes the P/D required for the first core. The optimal P/D ratio tends to increase with the number of times the fuel is recycled from one ENHS core to the next one. The optimal P/D ratio for the equilibrium composition core is in between 1.53 and 1.59. For a given discharge burnup it tends to somewhat increase with the equilibrium core power density. However, if structural materials will be developed to enable a 20 years core life at elevated power densities, the higher the power density the smaller is the required equilibrium P/D ratio.     

铀、钚等在某岩土介质上的吸附及其动力学

选择甘肃嘉峪关地区细砂、强风化花岗岩、粉质粘土三种具有代表性岩土介质,通过静态批式实验获得U、239 Pu、241 Am和244 Cm在三种岩土介质上的吸附动力学过程,采用准二级动力学模型及动边界模型对实验结果进行了拟合,结果表明:U、239 Pu、241 Am和244 Cm在三种岩土介质上的吸附均符合准二级动力学模型...     

福岛核事故向环境释放的Pu研究进展

福岛核事故向环境释放的放射性核素中包含了锕系元素Pu,其中以极毒组的239Pu、240Pu和高毒组的241Pu为主。本文总结并分析了针对福岛核事故向环境释放的Pu的相关研究。据估计,福岛核事故向环境中排放的239+240Pu总量约为109 Bq,是切尔诺贝利核事故排放量的万分之一。此次事故排放的Pu同位素原子比(240Pu/239Pu和241Pu/239Pu)及活度比(A(238Pu)/A(239+240Pu))明显异于全球沉降值,可作为事故中Pu溯源的判定依据。事故所排放的Pu全部来源于核电站1~3号反应堆堆芯而非乏燃料池。现有研究报道的数据表明,在福岛核电站周围30km范围内的陆地环境中存在来自核事故排放的Pu污染,污染相对严重的"热点"区域和该地区与核电站的相对位置没有明显关联,主要是受地形和降水的影响。而对于人们关心的海洋环境,来自福岛核事故的Pu污染非常小。核事故向海洋中排放的Pu相对于核事故前海洋环境中的Pu污染水平可忽略不计。     

托卡马克实验混合堆 FEB 嬗变 MA 可行性研究

研究了在聚变实验混合堆ＦＦＢ设计中，嬗变长寿命放射性少锕系（ＭＡ，ＭｉｎｏｒＡｃ－ｔｉｎｉｄｅｓ）核废物的可行性。应用改进的一维中子输运和燃耗计算程序ＢＩＳＯＮ３．０，完成了嬗变中子学与核素贫化计算。研究了核废物的嬗变率与辐照时间、包层厚度和废物装载量的关系，并对系统有关参数的选择进行了优化设计。结果表明，该设计（ＭＡ＋Ｐｕ）可年嬗变处置来自５５座相同功率的ＰＷＲ卸出的ＭＡ核废物，同时输出热功率５．４ＧＷ（ｔｈ）。     

Feasibility of a critical molten salt reactor for waste transmutation

This study assesses the feasibility of designing a Molten Salt Reactor (MSR) using the salt mixture of LiF (15 mol%), NaF (58 mol%) and BeF₂ (27 mol%) to be critical when fuelled with TRU from LWR spent fuel without exceeding the actinides solubility limit and while extracting fission products at realistic rates. The first part of the study investigated the graphite-to-MS volume ratio on the neutron balance, transmutation characteristics and graphite lifetime. It is found that a core without graphite moderator is the preferred design option; it offers the best neutron balance, most compact design and alleviated graphite lifetime problem. The second part of the study investigated sensitivity of the epithermal spectrum core to the feed composition, power density, fission products residence time and actinides loss fraction. It is found that the transmutation effectiveness improves with increasing power density and that the shorter the LWR spent fuel cooling time is, the better becomes the MSR neutron balance. The optimal MSR design offers a remarkably high transmutation capability – fissioning of as high as 99.8% of the TRU fed. The transmutation capability of the MSR is also rated in terms of final waste radiotoxicity, decay heat, spontaneous fission neutrons emission, fissile and ²³⁷Np inventory.     

Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea

Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.     

Thermal analysis of a storage cask for 24 spent PWR fuel assemblies

Abstract

The purpose of this paper is to perform a thermal analysis of a spent fuel storage cask in order to predict the maximum concrete and fuel cladding temperatures. Thermal analyses have been carried out for a storage cask under normal, off-normal and accident conditions. The environmental temperature is assumed to be 27°C under the normal condition. The off-normal condition has an environmental temperature of 40°C. An additional off-normal condition is considered as a partial blockage of the air inlet ducts. Four of the eight inlet ducts are assumed to be completely blocked. The accident condition is defined as a 100% blockage of air inlet ducts. The storage cask is designed to store 24 PWR spent fuel assemblies with a burn-up of 55,000 MWD/MTU and a cooling time of 7 years. The decay heat load from the 24 PWR assemblies is 25.2 kW. Thermal analyses of the ventilation system have been carried out for the determination of the optimum duct size and shape. The finite-volume computational fluid dynamics code FLUENT was used for the thermal analysis. From the results of the analysis, the maximum temperatures of the fuel rod and concrete overpack were lower than the allowable values under the normal, off-normal and accident conditions.     

Expansion of the capabilities of the GA-4 legal weight truck spent fuel shipping cask

Abstract

General Atomics has developed the model GA-4 legal weight truck spent fuel cask, a high-capacity cask for the transport of four pressurised water reactor (PWR) spent fuel assemblies, and obtained a certificate of compliance (CoC, No. 9226) in 1998 from the US Nuclear Regulatory Commission (NRC). The currently authorised contents for this CoC, however, are much more limiting than the actual capability of the GA-4 cask to transport spent PWR fuel assemblies. The purpose of this paper is to show how the authorised contents can be significantly expanded by additional analyses without any changes to the physical design of the package. Using burn-up credit as outlined in US NRC Interim Staff Guidance 8, Revision 2, the authorised contents can be significantly expanded by increasing the maximum enrichment as the burn-up increases. Use of burn-up credit eliminates most of the criticality imposed limits on authorised package contents, but shielding still limits the use of the cask for higher burn-up, short-cooled fuel. By reducing the number of assemblies transported (downloading) to two and using shielding inserts, even high-burn-up fuel with reasonable cooling times can be transported.