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.     

Time-integrated radiotoxicity index for study of transmutation of long-lived nuclides

Radiotoxicity of actinides is investigated based on the Annual Limit on Intake (ALI). A new radiotoxicity index, which takes the radiotoxicity of a parent and its all daughters into account, is defined. Three time-integrated radiotoxicity indices are also introduced for three time intervals of 0 to ∞ (ultimate), 0 to 1 × 10⁴ years (short-term) and 1 × 10⁴ to 1 × 10⁸ years (long-term). For 29 actinides, these indices are calculated theoretically and numerically. The ultimate indices of the actinides are almost the same. The long-term indices, however, show a wide range of values. Neptunium-237 and Am-241 have the largest long-term indices. On the contrary, Pu-240, Pu-242 and Pu-244, Cm-244 and Cm-246 have very small values which are 11/50 to 1/100 of those of Np-237 and Am-241 If these indices are used in the study of transmutation, one can examine and compare easily radiotoxicity and its time dependence for any material.     

Fusion-driven transmutation of selected long-lived fission products

The long-term radiological burden associated with nuclear power production is usually attributed to long-lived fission products (LLFP). Their lifetime and large equilibrium mass and hence radioactivity accumulated in the course of fission energy generation make their storage a rather formidable task to solve. Therefore the idea of artificial incineration of LLFP through their transmutation has been quite naturally incorporated into the concept of self-consistent nuclear energy system (SCNES) based primarily on fast breeder reactor technologies. However it is now acknowledged that neutron environment of fission facilities including fast breeder reactors does not seem most appropriate for LLFP transmutation. The issue has been then extensively developed within the framework of multi-component self-consistent nuclear energy system (MC-SCNES). Neutrons of specific quality required for LLFP transmutation are proposed there to be of non-fission origin. Given neutron excess available and neutron quality, a fusion neutron source (FNS) is appearing as the candidate No. 1 to consider for LLFP transmutation. Research on LLFP transmutation by means of FNS has very long history and has received an additional boost during the decade passed. In the present study, potential of thermal flux blanket of FNS is exemplified by transmutation of ⁹³Zr and ¹²⁶Sn, the most difficult LLFP to transmute. It is shown that transmutation of ⁹³Zr is effective even with a rather modest neutron loading of 1 MWt·m⁻², typical for ITER project. Transmutation of ¹²⁶Sn, however, requires neutron loading of as high as 3 MWt·m⁻² for DD fusion and is quite unattractive for DT fusion. In the latter case, transmutation through the threshold (n,2n) reaction may be preferable.     

Transmutation of long-lived fission products with use of available excess neutrons in BWR core

Feasibility of transmutation of long-lived Fps in BWR is studied from the view point of neuron balance. Pointing out the equilibrium mass amount of LLFP such as ¹²⁹I, ⁹⁹Tc, and ¹²⁶Sn in BWR core, the neutron balance analysis shows that the available neutrons produced by the current BWR core modification is satisfactory to the required amount of neutrons for LLFP transmutation. Core performance of the LLFP-recycled BWR is then analyzed and the result shows satisfactory results for the present core design criteria.

The results confirm the potential of the transmutation of LLFP in BWR and that the eventual radioactive hazard of the radwaste coming form the LLFP recycled BWR can be reduced well below the hazard level of the burned uranium.     

Recycle transmutation of MA and LLFP using BWR for sustaining geologic disposal

The actinides and fission products produced in nuclear fuels constitute an important part of the HLW. Therefore, methods for reducing the radiotoxicity of the MA and LLFP in HLW are presently under investigation. The purposes of this study are to evaluate the effectiveness of MA transmutation by taking advantage of neutron spectrum hardening due to void fraction along BWR axial direction; to understand the effectiveness of LLFP transmutation in BWR considering the large capture cross section of FP in thermal region; and to evaluate the macroscopic characteristics of longer residential period of LLFP target in the high burnup BWR core. Conceptual B/T BWR supposed in this study was reactor which the performance comparable to the current BWR. In MA transmutation case, the calculation was focused on varying the void fraction of 0 to 40% along the axial direction, which were directly associated to the lower and upper region of the BWR core. The performance of B/T BWR was evaluated in which four components of MA (²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm) with fixed fraction were blended with UO₂ in B/T fuel. While, for LLFP transmutation, the B/T BWR was assumed to have two homogeneous regions: {1} the region for UO₂ driver fuel (99% of fuel weight), and {2} the region for LLFP (⁹⁹Tc and ¹²⁹I) target capsules (1% of fuel weight), in which metallic Tc rods and iodine in the form of CeI₃ was contained in cylindrical target capsules. The evaluation functions are {1} fission-to-transmutation ratio, [F/T ratio]_MA, and {2} transmutation fraction, Tf_LLFP. Results show that the hardening neutron spectrum due to increase of void fraction in B/T BWR would result a higher [F/T ratio] of MA transmutation performance. Np and Am would be effectively loaded in the upper region of the core, while Cm could be loaded in any region of the core. At the EOC of equal or more than 50 GWd/Mg(HM), technetium has a higher transmutation fraction compared to iodine. To obtain higher LLFP transmutation fraction, the residential time in the LLFP targets in the core, should be kept for long time, for instance about 10 to 30 years. For that purpose, it was proposed that the number of B/T BWR system for LLFP treatment corresponds to the residential time of the LLFP target, i.e. 10 to 30 units.     

基于启明星1#的ADS嬗变研究

以ADS次临界试验平台启明星1#为研究对象,计算分析了MA和LLFP共9种核素的嬗变反应率。通过实验测量了LLFP中¹³⁷Cs的实际嬗变反应率,发现该装置对¹³⁷Cs嬗变速度是其本身自然衰变的10倍。实验结果表明启明星1#具有一定的嬗变能力,经分析确认实验测量结果和理论计算结果吻合。     

Accelerator mass spectrometry with time-of-flight measurement

A time-of-flight detection system for the measurement of microscopic concentrations of long-lived radioisotopes has been set up at the Munich MP tandem accelerator. The technique allows an unambiguous mass determination for heavy ions up to the actinides. As a first test of the system, the concentration of ¹²⁹I in iodine has been measured for several samples of mineral and biological origin. The smallest ¹²⁹I/¹²⁷I ratios measured were a few times 10⁻¹³. The high sensitivity of the method was achieved by using a 90° injector with a good mass resolution, a special technique of stabilizing the terminal voltage with an auxiliary beam, and a time-of-flight system with a mass resolution of 800.     

三步分离法制备高丰度的142La

为精确测量¹⁴²La的衰变数据,需制备出高活度、高丰度、无载体的¹⁴²La样品。为去除或降低混合裂片中放射性镧的其他同位素,依据母子体衰变关系和同一质量链上独立产额的差异,提出了“三步分离”的分离模式。以一次BaCl₂•2H₂O沉淀、两次HDEHP萃取色层分离为手段建立了快速分离¹⁴²La的化学流程,并研制了批式快速化学分离装置。流程耗时约20 min,化学回收率约80%,对其他核素的去污因数大于10³,产品中¹⁴²La与¹⁴¹La活度比大于3.5。     

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。     

放射性核素从废物体中向外释放行为初步研究

以去离子水为初始释放环境,考察了1年释放周期内,两类典型低、中水平放射性废物固化体在被水完全浸泡情景下,3种关键核素⁶⁰Co、¹³⁷Cs、⁹⁰Sr的向外释放行为。使用平均有效扩散系数作为表征参数。多个样品的测试结果表明：3种核素向外释放表现出规律性的趋势,即¹³⁷Cs最快,⁶⁰Co最慢, ⁹⁰Sr相比居中。对于不同的核素,起始活度对释放行为的影响是不同的。借助SEM探讨了废物体微观结构对核素释放的影响。     

Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter

This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZ_MA, and LLFP transmutation zone, TZ_FP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m². The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.     

Radiotoxicity hazard of U-free PuO2+ZrO2 and PuO2+ThO2 spent fuels in LWR

The radiotoxicity hazard of U-free Rock-like oxide: ROX (PuO₂+ZrO₂) and Thorium oxide: TOX (PuO₂+ThO₂) LWR spent fuels is investigated and radiotoxicity hazard of MOX spent fuel is considered as a reference case. The long-term ingestion radiotoxicity hazard of ROX spent fuel is one third and nearly one fourth of that of TOX and MOX spent fuels, respectively. This is because the discharged Pu and long lived Np in ROX fuel is less than that of TOX and MOX fuels. In TOX fuel, discharged Pu and MA are lower than that of MOX fuel but the long-term radiotoxicity hazard of spent fuel is nearly the same as MOX spent fuel. At the cooling 10⁵ years, the radiotoxicity hazard of TOX spent fuel is approximately ten and three times higher than that of ROX and MOX spent fuels, respectively due to higher toxic contribution of ²²⁹Th in TOX spent fuel.     

Development of ceramic waste forms for actinide-rich waste— Radiation stability of perovskite and phase and chemical stabilities of ZR-and AL-based ceramics

Ceramics are considered as most promising materials for conditioning of long-lived radionuclides because of their outstanding durability for long term. The Japan Atomic Energy Research Institute (JAERI) has developed ceramic waste forms, e.g. Synroc and zirconia-based ceramics, for the actinide-rich wastes arising from partitioning and transmutation processes. In the present study, -decay damage effects on the density and leaching behavior of perovskite (one of three main minerals forming Synroc) were investigated by an accelerated experiment using the actinide doping technique. A decrease in density of Cm-doped perovskite reached 1.3 % at a dose of 9 × 10¹⁷ -decays·g⁻¹. The leach rates (MCC-1 leach test inpH 2 solution at 90°C for 2 months) of perovskite specimens with accumulated doses of 1.6 × 10¹⁷, 4.0 × 10¹⁷ and 8.3 × 10¹⁷ -decays·g⁻¹ were 1.7, 2.3 and 3.0 μ·m⁻²·day⁻¹, respectively. Application of zirconia- and alumina-based ceramics for incorporating actinides was also investigated by the experiments using non-radioactive elements (Ce and Nd) with an emphasis on crystallographic phase stability and chemical durability. The yttria-stabilized zirconia was stable crystallographically in the wide ranges of Ce and/ or Nd content and had excellent chemical durability.     

废放射源近地表处置活度限值推导

以我国低中放固体废物北龙处置场和西北处置场为参考场址,针对钻探景象和钻探后景象,考虑了几个重要放射性核素,推导了废放射源在两个场址中处置的活度限值。除¹³⁷Cs的关键景象为钻探景象外,其他所选核素的关键景象均为钻探后景象。且单个废物包中的废放射源活度限值不受处置场的限制,推荐了适合于近地表处置的单个废物包内废放射源的接收限值,⁹⁰Sr、¹³⁷Cs、²³⁸Pu、^63Ni、²⁴¹Am、²²⁶Ra和²³⁹Pu分别为2.0×10¹⁰、6.0×10⁹、1.0×10⁶、1.0×10¹¹、1.0×10⁷、1.0×10⁶和8.0×10⁶Bq。     

Method for precision measurement of the masses of short-lived nuclides with application of stable isobars

The distinction of the suggested scheme of the experiment is the application of the isobaric doublets method in determining the masses of short-lived nuclides by means of a high-resolution mass spectrometer. This method enables us to considerably decrease the systematic errors of the measurement results in comparison with the previously applied isotopic reference method.

The work was carried out on an experimental complex, the main component of which was a prism mass spectrometer (PMS) with resolution R = 3 × 10⁴, installed on-line with a mass separator and a synchrocyclotron of the LNPI of the Academy of Sciences of Russia. On this installation we measured the masses of the isotopes ^91–97Rb. Samples containing stable reference isotopes ^91,92,94,96Zr, ⁹³Nb or ^95,97Mo were introduced into the ionizer of the PMS for calibration of the mass scale, that was ensured by their ionization simultaneously with radioactive isotopes of rubidium coming from the mass separator. The values of the masses of the stable isotopes used as reference in the suggested method are known with a very high precision (ΔM/M 3×10⁻⁸), therefore they made a negligible contribution to our experimental errors.     

在短寿命放射性医疗废物中发现长寿命杂质核素

为了确定短寿命放射性医疗废物能否清洁解控及短期的衰变贮存是否有效,本研究使用γ谱仪分析已放置了10个半衰期以上的低水平含99Tcm的放射性废物中残留的放射性核素,通过核素的全能峰定性,通过感兴趣区的净峰面积定量.经过了衰变贮存,虽然所有核素的活度浓度均低于相应的清洁解控水平,但在99Tcm废物中检出了长寿命核素137Cs、155Eu、123Tem、154Eu,其中137Cs、155Eu分析为99Tcm的母体99Mo生产过程中产生的杂质核素进入99Tcm所致,123Tem、154Eu可能为99Mo靶中杂质核素衰变而成.半衰期最长的长寿命核素137Cs的半衰期为30a,短期的衰变贮存并不能使这些长寿命核素活度显著减少,可见控制放射性药物的核纯度具有重要意义,可防止后期产生的医疗废物处置复杂化.     

Evaluation on transmutation performance of minor actinides with high-flux BWR

The performance of high-flux BWR (HFBWR) for burning and/or transmutation (B/T) treatment of minor actinides (MA) and long-lived fission products (LLFP) was discussed herein for estimating an advanced waste disposal with partitioning and transmutation (P&T). The concept of high-flux B/T reactor was based on a current 33 GWt-BWR, to transmute the mass of long-lived transuranium (TRU) to short-lived fission products (SLFP). The nuclide selected for B/T treatment was MA (Np-237, Am-241, and Am-243) included in the discharged fuel of LWR. The performance of B/T treatment of MA was evaluated by a new function, i.e. [F/T ratio], defined by the ratio of the fission rate to the transmutation rate in the core, at an arbitrary burn-up, due to all MA nuclides. According to the results, HFBWR could burn and/or transmute MA nuclides with higher fission rate than BWR, but the fission rate did not increase proportionally to the flux increment, due to the higher rate of neutron adsorption. The higher B/T fraction of MA would result in the higher B/T capacity, and will reduce the units of HFBWR needed for the treatment of a constant mass of MA. In addition, HFBWR had a merit of higher mass transmutation compared to the reference BWR, under the same mass loading of MA.     

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.     

Influence of void fraction change on plutonium and minor actinides recycling in BWR with equilibrium burnup

A study on the influence of void fraction change on plutonium and minor actinides recycling in standard boiling water reactor (BWR) with equilibrium burnup model has been conducted. We considered the equilibrium burnup model since it is a simple time independent burnup method that can handle all possible produced nuclides in any nuclear system.

The uranium enrichment for the criticality of the reactor diminishes significantly for the plutonium and minor actinides recycling case compared to that of the once-through cycle of BWR case. This parameter decreases much lower with the increasing of the void fraction. A similar propensity was also shown in the required natural uranium per annum. The annual required natural uranium was calculated by assuming that the uranium concentration in the tail of the enrichment plant is 0.25 w%. The amount of loaded fuel reduces slightly with the increment of the void fraction for plutonium and minor recycling in BWR.     

A concept of nitride fuel actinide recycle system based on pyrochemical reprocessing

A fuel cycle system coupled with nitride fuel fast reactors and a pyrochemical reprocessing has been investigated in order to establish an actinide transmutation recycle system with long-lived radioactive nuclides. Core performance of the nitride fuel fast reactor can provide design flexibility of excellent safety characteristics and a new concept of core composed with Na- and He- bonded fuel assemblies is proposed. The effect of ¹⁵N enrichment on nuclear characteristics and the evaluation of toxicity of ¹⁴C generated from ¹⁴N are appeared, and futhermore, excellent performance for the minor actinide (MA) transmutation is shown.

The study of the pyrochemical process shows that the actinides are reasonably separated from fission products in the nitride spent fuels, and that the high level wastes are of nearly actinide-free form.