The demand for ^(238)Pu (nuclear battery heat source) drives the separation of its precursor,^(237)Np,from spent nuclear fuel (SNF).However,the co-existence of multi-valence states (Ⅳ/Ⅴ/Ⅵ) of Np and similar redox b...The demand for ^(238)Pu (nuclear battery heat source) drives the separation of its precursor,^(237)Np,from spent nuclear fuel (SNF).However,the co-existence of multi-valence states (Ⅳ/Ⅴ/Ⅵ) of Np and similar redox behavior with Pu(Ⅳ) hinder the effective separation of Np.N-Butyraldehyde (n-C_(3)H_(7)CHO) selectively reduces Np(Ⅵ) to Np(Ⅴ) without reducing Pu(Ⅳ).Herein,we examined the reduction mechanisms of Np(Ⅵ) and Pu(Ⅳ) by n-C3H7CHO using relativistic density functional theory.Based on the results of the potential energy profiles,the reductions of both Np(Ⅵ) and Pu(Ⅳ) by n-C_(3)H_(7)CHO are thermodynamically feasible,whereas only the former is kinetically achievable.It uncovers that n-C_(3)H_(7)CHO can only reduce Np(Ⅵ) to Np(Ⅴ) owing to kinetically controlled selective reduction.The analyses of spin density and bond distance indicate that the reduction nature for the first Np(Ⅵ)/Pu(Ⅳ) belongs to hydrogen atom transfer,whereas that for the second one involves outer-sphere electron transfer.Localized molecular orbitals (LMOs) analysis discloses the bonding evolution during the reduction process of Np(Ⅵ)/Pu(Ⅳ).This study elucidates the reason behind the kinetically controlled selective reduction of Np(Ⅵ)/Pu(Ⅳ) by nC3H7CHO at the molecular level and offers in-depth perspectives on the isolation of specific metal ions from the view of kinetic control.展开更多
电解法制备四价铀作为核燃料加工中的关键技术,具有产率高且不引入杂质的显著优势,研究以UO_(2)(NO_(3))_(2)-HNO_(3)-N_(2)H_(4)·HNO_(3)体系为基础,通过串联式无隔膜电解装置,系统考察了电流密度与硝酸浓度与肼浓度及温度等关键...电解法制备四价铀作为核燃料加工中的关键技术,具有产率高且不引入杂质的显著优势,研究以UO_(2)(NO_(3))_(2)-HNO_(3)-N_(2)H_(4)·HNO_(3)体系为基础,通过串联式无隔膜电解装置,系统考察了电流密度与硝酸浓度与肼浓度及温度等关键参数对电解效率的影响规律。实验结果表明,电流密度14 m A/cm^(2),硝酸浓度1.6~1.8 mol/L,初始肼浓度0.8 mol/L以及控温50℃以内的工艺条件下,电解产物中四价铀浓度稳定且产率较高,与其他制备方法相比,电解法具有工艺简单与废液少且易于工业化应用的特点,为核燃料后处理工艺提供了技术支持。展开更多
This study proposes a method for^(99)Mo production via electron accelerator irradiation of a natural-uranium-bearing liquid molten salt target,with advantages including low nuclear proliferation risk,online extraction...This study proposes a method for^(99)Mo production via electron accelerator irradiation of a natural-uranium-bearing liquid molten salt target,with advantages including low nuclear proliferation risk,online extraction capability,and low construction costs.The approach primarily produces^(99)Mo through photofission of uranium(~95%),specifically^(238)U(γ,f).Secondary neutrons,originating from photonuclear interactions or fission processes,contribute minimally(~5%)to^(99)Mo production owing to their high energies and low fission cross sections.Key parameter analyses revealed that fluoride salt systems exhibit higher^(99)Mo yield.Their performance stems from high bremsstrahlung energy loss rate and superior photon yield,making them optimal molten salt target materials.To maximize photofission and photoneutron cross sections while minimizing highenergy gamma ray shielding requirements,an electron beam energy range of 40-80 MeV is recommended.To suppress local hot spots and prevent molten salt boiling,flow conditions were introduced to enhance convective heat transfer,effectively reducing the peak temperature.At a flow velocity of 0.5 m/s and under 80 MeV energy conditions,the maximum system temperature is only 808.9 K,which is significantly lower than the boiling point of 1773 K.Under optimized parameters,the maximum annual production capacity of~(99)Mo reaches 4486.49 Ci,sufficient for millions of diagnostic procedures and equivalent to 16.37% of China's projected demand for 2030.This method provides a viable pathway for stable,large-scale^(99)Mo production.展开更多
While nuclear energy represents a low-carbon and high-efficiency energy source that plays a vital role in the global energy mix,the limitations of spent fuel reprocessing technology pose a major challenge to its susta...While nuclear energy represents a low-carbon and high-efficiency energy source that plays a vital role in the global energy mix,the limitations of spent fuel reprocessing technology pose a major challenge to its sustainable development.The PUREX(plutonium uranium redox extraction)process is currently the dominant nuclear fuel reprocessing technology in the world.However,the key extractant in this process is tributyl phosphate(TBP),which degrades under intense radiation,high temperatures,and strong acidity.This leads to the production of dibutyl phosphate,monobutyl phosphate,and other degradation byproducts,which may reduce the extraction efficiency and trigger third-phase formation and equipment corrosion.This paper systematically reviews the degradation mechanisms of TBP and its diluents,the analytical technique suitable for characterizing degradation products,and the impact of degradation products on the post-treatment process.Additionally,optimization strategies employed for suppressing third-phase formation are discussed.This study offers a theoretical foundation and technical insights in optimizing the PUREX process and ensuring the safe operation of the post-treatment process.展开更多
基金supported by the National Natural Science Foundation of China(Nos.22376197,U2441225,22076188).
文摘The demand for ^(238)Pu (nuclear battery heat source) drives the separation of its precursor,^(237)Np,from spent nuclear fuel (SNF).However,the co-existence of multi-valence states (Ⅳ/Ⅴ/Ⅵ) of Np and similar redox behavior with Pu(Ⅳ) hinder the effective separation of Np.N-Butyraldehyde (n-C_(3)H_(7)CHO) selectively reduces Np(Ⅵ) to Np(Ⅴ) without reducing Pu(Ⅳ).Herein,we examined the reduction mechanisms of Np(Ⅵ) and Pu(Ⅳ) by n-C3H7CHO using relativistic density functional theory.Based on the results of the potential energy profiles,the reductions of both Np(Ⅵ) and Pu(Ⅳ) by n-C_(3)H_(7)CHO are thermodynamically feasible,whereas only the former is kinetically achievable.It uncovers that n-C_(3)H_(7)CHO can only reduce Np(Ⅵ) to Np(Ⅴ) owing to kinetically controlled selective reduction.The analyses of spin density and bond distance indicate that the reduction nature for the first Np(Ⅵ)/Pu(Ⅳ) belongs to hydrogen atom transfer,whereas that for the second one involves outer-sphere electron transfer.Localized molecular orbitals (LMOs) analysis discloses the bonding evolution during the reduction process of Np(Ⅵ)/Pu(Ⅳ).This study elucidates the reason behind the kinetically controlled selective reduction of Np(Ⅵ)/Pu(Ⅳ) by nC3H7CHO at the molecular level and offers in-depth perspectives on the isolation of specific metal ions from the view of kinetic control.
文摘电解法制备四价铀作为核燃料加工中的关键技术,具有产率高且不引入杂质的显著优势,研究以UO_(2)(NO_(3))_(2)-HNO_(3)-N_(2)H_(4)·HNO_(3)体系为基础,通过串联式无隔膜电解装置,系统考察了电流密度与硝酸浓度与肼浓度及温度等关键参数对电解效率的影响规律。实验结果表明,电流密度14 m A/cm^(2),硝酸浓度1.6~1.8 mol/L,初始肼浓度0.8 mol/L以及控温50℃以内的工艺条件下,电解产物中四价铀浓度稳定且产率较高,与其他制备方法相比,电解法具有工艺简单与废液少且易于工业化应用的特点,为核燃料后处理工艺提供了技术支持。
基金supported by the National Natural Science Foundation of China(Nos.12435012,12175300,and 12475185)Shanghai Natural Science Foundation(No.24ZR1478500)Nuclear energy development project(HNKF202210(24))。
文摘This study proposes a method for^(99)Mo production via electron accelerator irradiation of a natural-uranium-bearing liquid molten salt target,with advantages including low nuclear proliferation risk,online extraction capability,and low construction costs.The approach primarily produces^(99)Mo through photofission of uranium(~95%),specifically^(238)U(γ,f).Secondary neutrons,originating from photonuclear interactions or fission processes,contribute minimally(~5%)to^(99)Mo production owing to their high energies and low fission cross sections.Key parameter analyses revealed that fluoride salt systems exhibit higher^(99)Mo yield.Their performance stems from high bremsstrahlung energy loss rate and superior photon yield,making them optimal molten salt target materials.To maximize photofission and photoneutron cross sections while minimizing highenergy gamma ray shielding requirements,an electron beam energy range of 40-80 MeV is recommended.To suppress local hot spots and prevent molten salt boiling,flow conditions were introduced to enhance convective heat transfer,effectively reducing the peak temperature.At a flow velocity of 0.5 m/s and under 80 MeV energy conditions,the maximum system temperature is only 808.9 K,which is significantly lower than the boiling point of 1773 K.Under optimized parameters,the maximum annual production capacity of~(99)Mo reaches 4486.49 Ci,sufficient for millions of diagnostic procedures and equivalent to 16.37% of China's projected demand for 2030.This method provides a viable pathway for stable,large-scale^(99)Mo production.
基金supported by the Youth Talent Project of China Nuclear Power Engineering Co.,Ltd.(KY24045).
文摘While nuclear energy represents a low-carbon and high-efficiency energy source that plays a vital role in the global energy mix,the limitations of spent fuel reprocessing technology pose a major challenge to its sustainable development.The PUREX(plutonium uranium redox extraction)process is currently the dominant nuclear fuel reprocessing technology in the world.However,the key extractant in this process is tributyl phosphate(TBP),which degrades under intense radiation,high temperatures,and strong acidity.This leads to the production of dibutyl phosphate,monobutyl phosphate,and other degradation byproducts,which may reduce the extraction efficiency and trigger third-phase formation and equipment corrosion.This paper systematically reviews the degradation mechanisms of TBP and its diluents,the analytical technique suitable for characterizing degradation products,and the impact of degradation products on the post-treatment process.Additionally,optimization strategies employed for suppressing third-phase formation are discussed.This study offers a theoretical foundation and technical insights in optimizing the PUREX process and ensuring the safe operation of the post-treatment process.
