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）。     

功率展平的压水堆乏燃料发电包层中子学初步研究

聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现。本工作基于目前国际聚变实验堆（ITER）所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平。计算结果表明：功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%。燃料富集度到运行末期最大可达5.23%。从中子学角度初步论证了该包层的可行性。     

在聚变堆中嬗变~（237）Np的研究

研究了在聚变堆中嬗变长寿命的锕系元素￣（２３７）Ｎｐ，以及转换￣（２３７）Ｎｐ成为可裂变燃料￣（２３９）ｐｕ的物理可行性。探讨了在嬗变包层中￣（２３７）Ｎｐ的浓度、￣（２３９）ｐｕ的中于增殖率、中子壁负载的变化以及嬗变区厚度与￣（２３７）Ｎｐ嬗变率的关系。给出的研究计算结果表明，在１个聚变功率为２００ＭＷ，中子壁负载为１．０ＭＷ／ｍ２的聚变堆包层中，实现年嬗变￣（２３７）Ｎｐ约３．５ｔ，年平均产钚量约２０ｋｇ是可行的。     

加速器驱动洁净核能系统中的核素平衡条件

对加速器驱动洁净核能系统（ＡＤＳ）次临界堆内核素的转换进行了研究。研究结果表明：ＡＤＳ具有充分利用核资源的可能性。次临界热堆中能工作在ψ＝１０＾１５￣１０＾１６ｃｍ＾－２．ｓ＾－１下仍可稳定工作,且平衡时的易裂变核素（＾２３３Ｕ和＾２３９Ｐｕ）数目与初始装料核素的比值远高于热堆的。ＡＤＳ中,外源中子可有效地将可裂变核素转换易裂变核素。为加速达到平衡,初始装料中加入少量＾２３３Ｕ及＾２３９Ｐｕ是一种     

乏燃料后处理溶液中U,Pu的^57Co原激发K—XRF分析

自制一套＾３７Ｃｏ源激发与Ｋ系Ｘ射线荧光（Ｋ－ＸＲＦ）分析系统，用＾３７Ｃｏ的ｋｅＶγ射线激发工艺溶液中Ｕ，Ｐｕ的Ｋ系Ｘ射线荧光，用ＨＰＧｅ探测器－多道微机分析系统进行测量，并以１２２ｋｅＶγ射线康昔顿散射线为内标，建立强度比－浓度校正曲线，快速同地测定了ＰＷＲ乏燃料后处理工艺溶液中Ｕ，Ｐｕ浓度，测定范围为０．５－２００ｇ／Ｌ，精密度为５．０％－１．５％。方法适于ＰＷＲ乏燃料后处理工艺中，Ｕ，Ｐｕ     

聚变-裂变混合堆外中子源效应

聚变-裂变混合堆(FFHR)作为聚变驱动次临界系统(FDS),具有良好的物理性能,能够实现产能、氚增殖、嬗变核废料等功能。采用COUPLE程序研究了水冷混合堆包层的铀水比和中子倍增剂对中子源效率的影响。结果表明:包层能谱越硬,外中子源效率越高;适当加入中子倍增剂Be可使外中子源效率增加。研究结果对进一步改进聚变-裂变混合堆的概念设计具有一定的指导意义。     

乏燃料后处理溶液中U、Pu的~(57)Co源激发K-XRF分析

自制一套５７Ｃｏ源激发Ｋ系Ｘ射线荧光（Ｋ－ＸＲＦ）分析系统。用５７Ｃｏ的１２２ｋｅＶγ射线激发工艺溶液中Ｕ、Ｐｕ的Ｋ系Ｘ射线荧光，用ＨＰＧｅ探测器－多道微机分析系统进行测量，并以１２２ｋｅＶγ射线康普顿散射线为内标，建立强度比－浓度校正曲线，快速无损地测定ＰＷＲ乏燃料后处理工艺溶液中Ｕ、Ｐｕ浓度。测定范围为０．５—２００ｇ／Ｌ，精密度为５．０％—１．５％。方法适于ＰＷＲ乏燃料后处理工艺中Ｕ、Ｐｕ浓度的快速控制分析或在线分析。在同时应用５７Ｃｏ透射源的情况下，精密度达０．５％，方法适于核燃料衡算分析。     

MCORGS及次临界能源堆包层概念研究进展

次临界能源堆是以能源供应为目的的一种聚变裂变混合堆,以聚变驱动,天然铀为裂变燃料,轻水为冷却剂。本文针对该混合堆开发了基于MCNP与ORIGENS的三维中子输运燃耗耦合程序MCORGS,分析了包层三维中子学模型。提出简化干法后处理,设想利用衰变热将乏燃料加热到2 100K,将沸点低于该温度的裂变产物挥发去除。计算了包层各区材料每年发生的原子移位数,建议采用10a左右的换料周期,乏燃料经后处理后可多次复用。第1个寿期内氚增殖比TBR平均约1.15,包层能量放大倍数M平均约12;第2~9个寿期内TBR平均约1.35,M平均约18。利用流体动力学程序完成了包层CAD模型建立、网格划分及稳态传热计算分析,各区材料的最高温度均低于许用温度并有较大裕量。     

托卡马克商用混合堆堆内燃料循环优化设计

简要介绍了聚变－裂变混合堆难内燃料循环研究的方法、程序和程序的改进，提出了适用于托卡马克商用混合堆ＴＣＢ设计的３种堆内燃料循环模型，研究了堆内燃料的装卸模式与增殖燃料２３９Ｐｕ生产量的关系，增殖燃料加浓度的选择，提出了抑制裂变直接加浓核燃料概念，并给出了有关的计算结果。结果表明，在ＴＣＢ设计中，采用抑制裂变直接加浓核燃料模式，可实现年产２３９Ｐｕ燃料２２００ｋｇ且加浓度大于３％。结果还表明，采用分区卸料方式，可有效地减小系统的功率摆动和裂变率，这对商用混合堆设计尤其重要。     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Effect of Transplutonium Doping on Approach to Long-life Core in Uranium-fueled PWR

The present paper advertises doping of transplutonium isotopes as an essential measure to improve proliferation-resistance properties and burnup characteristics of UOX fuel for PWR. Among them ²⁴¹Am might play the decisive role of burnable absorber to reduce the initial reactivity excess while the short-lived nuclides ²⁴²Cm and ²⁴⁴Cm decay into even plutonium isotopes, thus increasing the extent of denaturation for primary fissile ²³⁹Pu in the course of reactor operation. The doping composition corresponds to one discharged from a current PWR. For definiteness, the case identity is ascribed to atomic percentage of ²⁴¹ Am, and then the other transplutonium nuclide contents follow their ratio as in the PWR discharged fuel. The case of 1 at% doping to 20% enriched uranium oxide fuel shows the potential of achieving the burnup value of 100GWd/tHM with about 20% ²³⁸Pu fraction at the end of irradiation. Since so far, americium and curium do not require special proliferation resistance measures, their doping to UOX would assist in introducing nuclear technology in developing countries with simultaneous reduction of accumulated minor actinides stockpiles.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Irradiation analysis of U, Am samples irradiated in experimental fast reactor “Joyo” for protected plutonium production in fast breeder reactor blanket

Americium is a key element to design the FBR based nuclear fuel cycle, because of its long-term high radiological toxicity as well as a resource of even-mass-number plutonium by its transmutation in reactors, which contributes the enhancement of proliferation resistance. The present paper summarizes analysis of the individual Am and U samples irradiation in Joyo to re-evaluate the results of Pu isotopes in the measure of proliferation resistance, and to combine the results for the prediction of DU-Am irradiation especially in the production of Pu isotopes. By the prediction of DU-Am oxide fuel in fast reactor environment with detail fuel irradiation analysis, it was confirmed that neutron moderation and fuel size affects the produced Pu isotope and its vector due to the very high sensitivity of ²³⁸U resonance capture reaction, the larger diameter fuel is more preferable in the case of moderated neutron spectrum environment for denaturing Pu in fast reactor blanket. Finally proliferation resistance of all the Pu produced in U, Am sample irradiation and DU-Am fuel irradiation prediction were evaluated based on decay heat and spontaneous fission neutron rate, and it was confirmed ²⁴¹Am produces un-attractive Pu to abuse from the beginning to the end of irradiation, and more than 2% of ²⁴¹Am doping is required to enhance the proliferation resistance of Pu to MOX grade and Kessler’s proposal in moderated neutron spectrum environment in fast reactor.     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

Determination of americium in plutonium based nuclear fuel materials: An assessment of different methodologies

Determination of americium (Am) is one of the requirements of chemical quality assurance of plutonium (Pu) bearing fuel materials. Though many methods are published for determining Am at picogram to femtogram levels in environmental and biological matrices, yet a few of them are used routinely for Pu based nuclear fuel samples. This paper gives a brief summary of the different analytical methods available and presents results of our experiments on the determination of Am in Pu bearing fuels using gamma spectroscopy. The methods utilizing gamma emissions from ²⁴¹Am and Pu isotopes are fast as they do not involve chemical separation of Pu and Am, do not require an accurate knowledge of the efficiency values of the detector systems and are not dependent on the availability of a radiometric standard for ²⁴¹Am. In addition, for aged Pu samples containing large amounts of ²⁴¹Am, there is no need for dilution and this reduces the volume of analytical radioactive waste solution. Future requirements of reference materials to validate different methodologies for determining Am isotopes are also highlighted.     

Comparisons of HELIOS,ORIGEN2, and Monteburns Calculated 241 Am and 243 Am Concentrations to Measured Values for PWR,BWR, and VVER Spent Fuel

A study was performed at Los Alamos National Laboratory to explore the accuracy of several reactor analysis codes in calculating ²⁴¹ Am and ²⁴³Am concentrations in light water reactor spent fuel. Calculated higher-actinide concentrations were compared to measured values from the literature for three reactor fuels. The fuel samples were taken from the Mihama Unit 3 pressurized water reactor, the Garigliano boiling water reactor, and a VVER-440. The ²⁴¹Am and ²⁴³Am concentrations were calculated using the HELIOS-1.4 lattice-physics code, the ORIGEN2 burnup code, and a linked MCNP/ORIGEN2 code named Monteburns 3.01. Comparisons were made between the calculated and measured values. It was determined that all codes performed consistently well for the Mihama Unit 3 measurements (within ±5% for ²⁴¹Am and ±20% for ²⁴³Am) and the Garigliano measurements (within ±12% for ²⁴¹ Am and ±20% for ²⁴³Am). It was determined that the ORIGEN2 pressurized water reactor libraries are insufficient for the VVER-440 measurements. The HELIOS and MONTEBURNS codes both demonstrated good ability to calculate these isotopes for VVER-440 fuel (±10% for ²⁴¹Am and ±12% for ²⁴³Am). The accuracies of these codes and the associated radiochemical measurements of these higher-actinide isotopes may be insufficient for safeguards and fuel management purposes; thus, development of new methods and modification to existing data libraries may be necessary in order to enable cost-effective safeguarding of these higher-actinide materials.     

钚年龄评估技术

在军控核查技术中应用钚的年龄属性对核弹头认证和“禁产公约”监督具有非常重要的意义。根据^241Pu的不同衰变模式，对母核和子核衰变伴随的γ辐射随时间的变化、各种情况下子核与母核的原子比随时间的变化规律等进行了描述。研究表明，通过某种手段获得^241Pu和^241Am的原子比或测量γ辐射可确定钚的年龄。     

Phase characteristics of a number of U-Pu-Am-Np-Zr metallic alloys for use as fast reactor fuels

Metallic fuel alloys consisting of uranium, plutonium, and zirconium with minor additions of americium and neptunium are under evaluation for potential use to transmute long-lived transuranic actinide isotopes in fast reactors. A series of test designs for the Advanced Fuel Cycle Initiative (AFCI) have been irradiated in the Advanced Test Reactor (ATR), designated as the AFC-1 and AFC-2 designs. Metal fuel compositions in these designs have included varying amounts of U, Pu, Zr, and minor actinides (Am, Np). Investigations into the phase behavior and relationships based on the alloy constituents have been conducted using X-ray diffraction and differential thermal analysis. Results of these investigations, along with proposed relationships between observed behavior and alloy composition, are provided. In general, observed behaviors can be predicted by a ternary U-Pu-Zr phase diagram, with transition temperatures being most dependent on U content. Furthermore, the enthalpy associated with transitions is strongly dependent on the as-cast microstructural characteristics.     

Benchmark of Advanced Burner Test Reactor model using MCNPX 2.6.0 and ERANOS 2.1

Significant research is currently being performed whereby fast reactor cores have been designed to burn transuranic materials reducing the volume and long-term radiotoxicity of spent nuclear fuel. These core and depletion models depend on various computer codes. This research used MCNPX 2.6.0 and ERANOS 2.1 to model a standard 250 MW Advanced Burner Test Reactor (ABTR) core. The intent was to benchmark criticality and burnup results from a stochastic Monte Carlo code and a deterministic depletion code using a standard ABTR model created by Argonne National Laboratory. Because each of these codes solves the transport and burnup problem differently, there is a need to benchmark the core models in order to verify results and identify root causes for significant differences in results between codes. Flux calculations in ERANOS were performed using diffusion theory, Legendre polynomial approximations (using the VARIANT module) and discrete ordinates methods. The k-effective for the higher order transport models remained within 1000 pcm of the MCNPX model. The difference between the total heavy nuclide mass balance in ERANOS using the various flux calculations and the MCNPX depletion model was less than 0.4% out to a burnup of 1095 days (67.45 GWd/MTHM). The percent delta between the codes as a fraction of the fissioned mass was 1.34%. For the isotopes with large concentrations, such as ²³⁸U and ²³⁹Pu, the mass differences were 0.38% and 0.01% respectively. The mass difference for ²⁴¹Am was also small at 0.42%. Notable isotopes in small quantities with larger mass differences were ²⁴²Am, ²⁴²Cm, ²⁴³Cm and ²⁴⁶Cm where differences ranged from 0.1 to 0.2% after 26 days and increased to 11–136% at 1095 days.     

Peak identification of L X-ray spectra of elemental Np and U

The L X-ray photons emitted by transuranic (TRU) elements are expected to be useful for developing nondestructive TRU monitors. Energy spectra of L X-rays emitted by ²⁴¹Am, ²³⁸Pu and ²³⁹Pu sources were measured by a transition edge sensor (TES) microcalorimeter, which allowed precise peak identification with high energy resolution. In the measurements using the TES microcalorimeter, the full width at half-maximum energy resolution was 62.6 eV at 17.222 keV for ²³⁹Pu source, 62.5 eV at 17.222 keV for ²³⁸Pu source and 60.9 eV at 17.751 keV for ²⁴¹Am source. This study demonstrates the separation of ²⁴¹Am and plutonium isotopes by L X-ray spectroscopy using a TES microcalorimeter.