Plutonium utilization in a small modular molten-salt reactor based on a batch fuel reprocessing scheme

A molten salt reactor(MSR)has outstanding features considering the application of thorium fuel,inherent safety,sustainability,and resistance to proliferation.However,fissile material^(233)U is significantly rare at the current stage,thus it is difficult for MSR to achieve a pure thorium-uranium fuel cycle.Therefore,using plutonium or enriched uranium as the initial fuel for MSR is more practical.In this study,we aim to verify the feasibility of a small modular MSR that utilizes plutonium as the starting fuel(SM-MSR-Pu),and highlight its advantages and disadvantages.First,the structural design and fuel management scheme of the SM-MSR-Pu were presented.Second,the neutronic characteristics,such as the graphite-irradiation lifetime,burn-up performance,and coefficient of temperature reactivity were calculated to analyze the physical characteristics of the SM-MSR-Pu.The results indicate that plutonium is a feasible and advantageous starting fuel for a SM-MSR;however,there are certain shortcomings that need to be solved.In a 250 MWth SM-MSR-Pu,approximately 288.64 kg^(233)U of plutonium with a purity of greater than 90% is produced while 978.00 kg is burned every ten years.The temperature reactivity coefficient decreases from -4.0 to -6.5 pcm K^(-1) over the 50-year operating time,which ensures a long-term safe operation.However,the amount of plutonium and accumulation of minor actinides(MAs)would increase as the burn-up time increases,and the annual production and purity of^(233)U will decrease.To achieve an optimal burn-up performance,setting the entire operation time to 30 years is advisable.Regardless,more than 3600 kg of plutonium eventually accumulate in the core.Further research is required to effectively utilize this accumulated plutonium.

First-principles local density approximation (LDA) + U and generalized gradient approximation(GGA) + U studies of plutonium oxides

The electronic structures and properties of PuO2 and Pu2O3 have been studied according to the first principles by using the all-electron projector-augmented-wave （PAW） method. The local density approximation （LDA）＋U and the generalized gradient approximation （GGA）＋U formalisms have been used to account for the strong on-site Coulomb repulsion among the localized Pu 5f electrons. We discuss how the properties of PuO2 and Pu2O3 are affected by choosing the values of U and exchange-correlation potential. Also, the oxidation reaction of Pu2O3, leading to the formation of PuO2, and its dependence on U and exchange-correlation potential have been studied. Our results show that by choosing an appropriate U it is possible to consistently describe structural, electronic, and thermodynamic properties of PuO2 and Pu2O3, which enable the modelling of the redox process involving Pu-based materials.

The abnormal lattice contraction of plutonium hydrides studied by first-principles calculations

Pu can be loaded with H forming complicated continuous solid solutions and compounds,and causing remarkable electronic and structural changes.Full potential linearized augmented plane wave methods combined with Hubbard parameter U and the spin-orbit effects are employed to investigate the electronic and structural properties of stoichiometric and non-stoichiometric face-centered cubic Pu hydrides（PuHx,x=2,2.25,2.5,2.75,3）.The decreasing trend with increasing x of the calculated lattice parameters is in reasonable agreement with the experimental findings.A comparative analysis of the electronic-structure results for a series of PuH x compositions reveals that the lattice contraction results from the associated effects of the enhanced chemical bonding and the size effects involving the interstitial atoms.We find that the size effects are the driving force for the abnormal lattice contraction.

Molecular dynamics simulations of point defects in plutonium grain boundaries

A modified analytic embedded atom method （MAEAM） potential is constructed for fcc 5-Pu. Molecular dynamics （MD） simulations with the potential are performed to investigate the interactions between two symmetrical tilt grain boundaries （GBs） and point defects such as He atom, vacancy and self-interstitial atom （SIA） in Pu. The calculated results show that point defect formation energies are on average lower than those in the lattice but variations from site to site along the GBs are very remarkable. Both substitutional and interstitial He atoms are trapped at GBs. Interstitial He atom is more strongly bound at the GB core than the substitutional He atom. The binding energy of SIA at GB core is higher than those of He atom and vacancy. GB core can bind many He atoms and SIAs due mainly to the fact that it contains many vacancies. Compared with He atom and SIA, the vacancy far from GB core is difficult to diffuse into the core. The GBs can act as sinks and sources of He atoms and SIAs, which may be a reason for the swelling of Pu after a period of self-irradiation because of the higher concentration of vacancy in the bulk.

Evaluating the JEFF 3.1, ENDF/B‑VII.0, JENDL 3.3, and JENDL 4.0 nuclear data libraries for a small 100 MWe molten salt reactor with plutonium fuel

This study evaluated the nuclear data libraries for a small 100 Mega Watt electric(MWe)Molten Salt Reactor with plutonium fuel.The reactor has a power output of 100 MWe,which meets the demand for electricity generation in several regions or provinces outside Java Island.Several nuclear data libraries,such as JEFF 3.1,ENDF/B-VII.0,JENDL 3.3,and JENDL 4.0,were used for a more comprehensive evaluation.LiF–BeF_(2)–ThF_(4)–PuF_(4) was used as the initial fuel composition.The thorium and plutonium concentrations in the fuel salt were varied to obtain the optimum fuel composition,leading to critical conditions.The results showed some neutronic parameters,such as the conversion ratio,neutron spectra,and effective multiplication factors,from three different nuclear data libraries.By changing the plutonium concentration in the initial fuel salt composition,the minimum plutonium loaded for the reactor criticality during 2000 days of operation time was determined to be 0.995,0.91,0.87,and 0.90 mol%for JEFF 3.1,ENDF/B-VII.0,JENDL 3.3,and JENDL 4.0,respectively.The differences in the values of each parameter were due to several factors,such as the cross-section values and number of nuclides in the nuclear data libraries.Several safety parameters were also investigated to ensure the possibility of utilizing PuF_(4) in the reactor.

Man-made plutonium radioisotopes in the salt lakes of the Crimean peninsula

Investigations of ^(239+240)Pu and ^(238)Pu in the surface layer(0–5 cm) of bottom sediment in the Crimean 10 salt lakes from 4 geographical groups were carried out for the first time. The ^(239+240)Pu varied widely between regional geographical groups of lakes as well as within groups too and ranged from 11±4 to 451±43 mBq^(239+240)Pu/kg. The highest levels of ^(239+240)Pu—419±27, 443±24 and 451±43 mBq/kg were observed in the Yevpatoriya(Lake Kyzyl-Yar), the Tarkhankut(Dzharylhach) and the Kerch group(Tobechik), respectively. The lowest values of ^(239+240)Pu were identified in three lakes of the Perekop group and were 20±12, 24±6 and 48±6 mBq/kg. In all lakes ^(238)Pu was an order of magnitude lower than 239+240 Pu and varied from 4.8±2.6 to 30.7±5.5 mBq/kg. The ^(238)Pu activity was decay-corrected to 1986. The characteristic ratio of the ^(238)Pu/239+240 Pu activities in the sediment and percentage of the Chernobyl-derived Pu was calculated. The largest percentages of the Chernobyl-derived Pu were observed in the Evpatoriya group(Lake SasykSivash)—16.2%±8.26%, the Tarkhankut group(Dzharylhach)—8.4%±2.10% and the Kerch group(Aktash)—10.5%±5.56%. The study of the depth distribution of plutonium in the Lake Kyzyl-Yar bottom sediment core(0–25 cm) was fulfilled. It was shown that 239+240 Pu was high enough in all studied layers of bottom sediment, but the highest activity ratio ^(238)Pu/239+240 Pu(0.062±0.020) was found in the deepest layer of 15–20.5 cm and the percentage of Chernobyl-derived Pu was estimated at 6.8%±2.85% in this layer.

Analysis of burnup performance and temperature coefficient for a small modular molten‑salt reactor started with plutonium

In a thorium-based molten salt reactor(TMSR),it is difficult to achieve the pure 232Th–^(233)U fuel cycle without sufficient^(233)U fuel supply.Therefore,the original molten salt reactor was designed to use enriched uranium or plutonium as the starting fuel.By exploiting plutonium as the starting fuel and thorium as the fertile fuel,the high-purity^(233)U produced can be separated from the spent fuel by fluorination volatilization.Therefore,the molten salt reactor started with plutonium can be designed as a^(233)U breeder with the burning plutonium extracted from a pressurized water reactor(PWR).Combining these advantages,the study of the physical properties of plutonium-activated salt reactors is attractive.This study mainly focused on the burnup performance and temperature reactivity coefficient of a small modular molten-salt reactor started with plutonium(SM-MSR-Pu).The neutron spectra,^(233)U production,plutonium incineration,minor actinide(MA)residues,and temperature reactivity coefficients for different fuel salt volume fractions(VF)and hexagon pitch(P)sizes were calculated to analyze the burnup behavior in the SM-SMR-Pu.Based on the comparative analysis results of the burn-up calculation,a lower VF and larger P size are more beneficial for improving the burnup performance.However,from a passive safety perspective,a higher fuel volume fraction and smaller hexagon pitch size are necessary to achieve a deep negative feedback coefficient.Therefore,an excellent burnup performance and a deep negative temperature feedback coefficient are incompatible,and the optimal design range is relatively narrow in the optimized design of an SM-MSR-Pu.In a comprehensive consideration,P=20 cm and VF=20%are considered to be relatively balanced design parameters.Based on the fuel off-line batching scheme,a 250 MWth SM-MSR-Pu can produce approximately 29.83 kg of ^(233)U,incinerate 98.29 kg of plutonium,and accumulate 14.70 kg of MAs per year,and the temperature reactivity coefficient can always be lower than−4.0pcm/K.

Transition toward thorium fuel cycle in a molten salt reactor by using plutonium

The molten salt reactor(MSR), as one of the Generation Ⅳ advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233 U production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning(B&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233 U production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/^(233)U transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period(RP) case and about 1.047 for the 10-day RP case.The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing(RP is 180 days),and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a prebreeding and burning(PB&B) scenario is also analyzed briefly with respect to the net 233 U production and evolution of main nuclides. One can find that it is more efficient to produce 233 U under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.

Hybrid density functional study on lattice vibration,thermodynamic properties,and chemical bonding of plutonium monocarbide

Hybrid density functional theory is employed to systematically investigate the structural,magnetic,vibrational,thermodynamic properties of plutonium monocarbide（Pu C and Pu C0.75）.For comparison,the results obtained by DFT,DFT ＋ U are also given.For Pu C and Pu C0.75,Fock-0.25 hybrid functional gives the best lattice constants and predicts the correct ground states of antiferromagnetic（AFM） structure.The calculated phonon spectra suggest that Pu C and Pu C0.75 are dynamically stable.Values of the Helmholtz free energy ？F,internal energy ？E,entropy S,and constant-volume specific heat Cv of Pu C and Pu C0.75 are given.The results are in good agreement with available experimental or theoretical data.As for the chemical bonding nature,the difference charge densities,the partial densities of states and the Bader charge analysis suggest that the Pu–C bonds of Pu C and Pu C0.75 have a mixture of covalent character and ionic character.The effect of carbon vacancy on the chemical bonding is also discussed in detail.We expect that our study can provide some useful reference for further experimental research on the phonon density of states,thermodynamic properties of the plutonium monocarbide.

Recovery of Uranium and Plutonium from Waste Matrices Using Supercritical Fluid Extraction

Supercritical fluid extraction (SFE) of plutonium in its nitrate form from actual waste, i.e. plutonium bearing cellulose matrix was demonstrated using 0.1 litre capacity extraction vessel. Complete recovery of plutonium was demonstrated using modified supercritical carbon dioxide (Sc-CO2), i.e. Sc-CO2 containing octylphenyl-N, N-diisobutyl-carbamoyl- methylphosphine oxide (φCMPO). Near complete recovery of uranium was demonstrated from simulated waste matri- ces, i.e. uranium bearing teflon, glass and cellulose matrices using preparative scale SFE, i.e. from 1 litre capacity ex- traction vessel. The recovery of uranium was established using Sc-CO2 modified with acetyl acetone.

Comparison of Radionuclide’s Inventories and Activities With Slightly Enriched Uranium and Plutonium Fuel in CANDU Reactors

In these studies the isotopic inventories and corresponding activities of important nuclides for different fuel cycles of a CANDU reactor have been compared. The calculations have been performed using the computer code WIMSD4. The isotopic inventories and activities have been calculated versus the fuel burn-up for the natural UO2 fuel, 1.2% enriched UO2 fuel and for the 0.45% PuO2-UO2 fuel. It is found that 1.2% enriched uranium fuel has the lowest activity as compared to other two fuel cycles and vice versa for the 0.45% PuO2-UO2 fuel.

First application of plutonium in soil erosion research on terraces

The spatial distributions of ^(239+240)Pu and ^(137)Cs in soils from Longji Rice Terraces were investigated to evaluate soil erosion.The activity concentrations of ^(239+240)Pu and ^(137)Cs in the surface soils of the paddy fields were in the range of 0.089–0.734 and1.80–7.88 mBq/g,respectively.The activities of ^(239+240)Pu and ^(137)Cs showed very similar distribution trends,first increasing and then decreasing with increasing elevation.The 240Pu/239Pu atom ratios in the surface soils ranged from 0.162 to 0.232.The activities of ^(239+240)Pu and ^(137)Cs in the soil cores tended to be uniformly distributed within the plowed layer and declined exponentially below this depth.The mean soil erosion rates of Longji Rice Terraces estimated by ^(239+240)Pu and ^(137)Cs tracer methods were 5.44 t/(ha·a)and 5.16 t/(ha·a),respectively,which demonstrated that plutonium can replace ^(137)Cs as an ideal tracer for soil erosion research in the future.Landform features are the main factors affecting the distribution of plutonium and ^(137)Cs as well as soil erosion in the Longji Rice Terraces.

Fukushima Plutonium Effect and Blow-Up Regimes in Neutron-Multiplying Media

It is shown that the capture and fission cross-sections of 238U and 239Pu increase with temperature within 1000 K - 3000 K range, in contrast to those of 235U, that under certain conditions may lead to the so-called blow-up modes, stimulating the anomalous neutron flux and nuclear fuel temperature growth. Some features of the blow-up regimes in neutronmultiplying media are discussed.

Equilibrium composition for the reaction of plutonium hydride with air

There are six independent constituents with 4 chemical elements, i.e. PuH 2.7 (s), PuN(s), Pu 2O 3(s), N 2, O 2 and H 2, therefore , the system described involves of 2 independent reactions ,both ΔG°0. The calculated equilibrium compositions are in agreement with those of the experimental, which indicates that the chemical equilibrium is nearly completely approached. Therefore, it is believed that the reaction rate of plutonium hydride with air is extremely rapid. The present paper has briefly discussed the simultaneous reactions and its thermodynamic coupling effect.

Determination of Plutonium Isotopes in Radioactive Waste Contaminated with Uranium and Thorium

Nucleco is the Sogin Group’s Italian leading company in the sector of radiological services,radioactive waste management,decontamination and reclamation of industrial sites and nuclear power plants.Nucleco is 60%owned by Sogin and 40%by ENEA-National Agency for New Technologies.The waste characterization carried out by Nucleco SpA includes a large variety of radionuclides belonging to the gamma,beta and alpha emitter families.The determination of Uranium and Plutonium isotopes plays a key role in the waste characterization.Nucleco SpA has distinguished itself in characterizations of radioactive waste in complex and nuclear plant matrices.These matrices have large amounts of alpha emitters,in particular isotopes of Uranium,Thorium and Plutonium.A significant presence of U238 results in such a large amount of Th234(and daughters)that interferes with the determination of Pu241.Hence,there is the need of finding a pre-treatment,extraction and subsequent purification technique that would allow Th234 to be detached and thus Pu241 to be determined in complex high-activity matrices.The above elements are extracted using chromatography columns.An isotope of the element to be analyzed,with known activity,is added at the beginning of the process to determine the extraction yield.Before being eluted into the column,the sample undergoes a series of treatments in order to be purified of any interferents.The method developed by Nucleco involves the oxidation of Pu at valence+6 and the subsequent precipitation,in fluorides form,of the elements with valence+4 and+3(i.e.Th234).Pu241 is then measured by LSC(Liquid Scintillation Counting),while the other isotopes are measured by alpha spectrometry after electrodeposition on a metal plate.

长寿命低丰度放射性核素的激光共振电离质谱分析技术研究进展

长寿命低丰度放射性核素分析在核环境监测、核燃料循环和核取证等领域具有重要价值。激光共振电离质谱是一种将激光共振电离技术和质谱技术相结合的质谱分析技术,具有高选择性、高探测灵敏度和高丰度灵敏度等优点,适用于复杂基体条件下超痕量核素分析需求。该文综述了近年来课题组基于自研的磁-电双聚焦激光共振电离质谱仪进行121mSn、126Sn、233U、236U、238Pu等核素丰度分析的相关研究工作,介绍了高效原子化源、高元素选择性以及高同位素选择性实验技术,并对激光共振电离质谱用于超痕量核素分析的发展方向进行了展望。

基于D-D中子源的乏燃料组件钚含量测量装置设计

为了满足国际社会上对于乏燃料组件内特种可裂变材料钚的保障监督需求,以持续提供核材料信息,研究乏燃料组件内钚含量测量的非破坏性分析技术显得尤为重要。本文采用主动法,以压水堆乏燃料组件作为测量对象,D-D中子作为质询中子源,开展了乏燃料组件钚含量测量装置的设计研究。采用MCNPX软件,基于最大化探测器计数率和使各探测器计数率尽量一致的目的,对测量装置中的中子管与探测器组件距离、中子管慢化材料及其厚度、探测器组件与中子管高度差、探测器组件中慢化体厚度等关键参数进行了模拟计算。此研究为中子质询乏燃料组件钚含量测量技术研究及实验验证打下了基础。

钚(Ⅳ)-胶体-不动介质(花岗岩)间的作用机制:实验,表面配位和DLVO作用力

为了准确模拟钚在地下水中的大尺度迁移行为,需要掌握地下水三相体系中钚(Ⅳ)-天然胶体-不动介质(花岗岩)间的作用机制。实验结果表明,钚(Ⅳ)吸附在土壤胶体表面生成的钚(假)胶体有很强的运动能力(钚(假)胶体在装有孔隙介质的柱体中传输后钚的回收率R反映其运动能力),其运动能力R随土壤胶体质量浓度ρ的增大而增大(0≤ρ≤375.4 mg/L,1.3%≤R≤52.5%)。钚(Ⅳ)在土壤胶体表面生成的主要配位种态是≡SOPu(OH)_(3),种态百分数>95%,且种态百分数与胶体浓度正相关。实验和表面配位计算证实,钚(假)胶体增强了钚的运动能力,比水溶态钚污染远场水域的风险更大。DLVO作用势计算结果指出,随着胶体浓度的继续增大,胶体-介质间第二能量最小值(Ф_(min2))的波谷深度加深。该引力相互作用势导致钚(假)胶体在介质表面的吸着沉积效率增大,且与实验观察到的逐渐变弱的钚(假)胶体的运动能力一致(375.4 mg/L≤ρ≤2017.8 mg/L,52.5%≤R≤12.7%)。

地下水成分对钚化学种态分布的影响

钚是核能发展和放射性废物处理处置中的一种重要元素,其在天然水体中的种态分布对于研究其迁移行为至关重要。随着研究的深入,一些新的钚种态不断被发现和确认,原先诸多类似研究工作已经不能准确反映钚在溶液介质中的真实种态占比。结合最新热力学数据,本文通过地球化学计算软件PHREEQC系统研究了钚的种态分布随pH值和共存离子种类及浓度改变的变化规律,并得到了钚在不同水体中的物种形态占比。研究结果表明:在氧化性的地下水中,当水硬度比较低时,六价钚在酸性条件下主要以PuO_(2)^(2+)或PuO_(2)CO_(3)的形态存在,在中碱性条件下主要以PuO_(2)(CO_(3))_(2)^(2−)或PuO_(2)(CO_(3))_(3)^(4−)形态存在;然而,当水硬度比较高,尤其是钙离子浓度比较高时,六价钚虽然在酸性条件下仍主要以PuO_(2)^(2+)或PuO_(2)CO_(3)存在,在中碱性条件下却以CaPuO_(2)(CO_(3))_(3)^(2−)的形式占据主导地位。可见,水的硬度尤其是钙离子浓度是影响钚形态的重要因素之一,因而在钚的地质处置时应予以重点考虑。

钚吸附行为研究进展

Pu是一种非常关键的放射性元素,其分离纯化是乏燃料后处理中的核心技术,同时,研究Pu的分离也对环境保护和放射性污染修复具有重要的意义。目前分离纯化Pu的主要方法有沉淀法、萃取法、吸附法等,其中吸附法是分离纯化Pu的重要方法之一。吸附法具有操作简单,吸附速率快等优点,同时吸附材料具有较大的吸附容量、价格低廉等优点。本文根据吸附材料性质的不同,主要从天然矿物、聚合物树脂、无机氧化物、碳材料以及纳米材料等五个方面介绍了国内外对Pu吸附行为的研究进展,总结了不同吸附材料对Pu的吸附能力及其影响因素,讨论了相关吸附机理,对不同材料的优缺点进行了分析;最后,就Pu吸附分离研究和应用存在的问题和未来发展方向进行了分析和展望,以期对包括Pu在内的放射性元素吸附分离研究和应用提供参考。