加速器驱动钠冷金属燃料快堆次锕系核素嬗变特性研究

分析加速器驱动系统(ADS)钠冷金属燃料快堆重金属燃料不同核素对堆芯有效增殖系数(Keff)的影响,给出了燃料成分的确定方法,详细分析次锕系核素(MA)嬗变特性.运用耦合了MCNP4c3与ORJGEN2的三维燃耗程序COUPLE对堆芯进行稳态与燃耗计算.结果分析表明,调节燃料中239Pu的质量比例并使其在燃耗过程中保持稳定是使Keff达到设计值并在燃耗过程中保持稳定的有效手段.散裂中子引起堆芯内区较外区更硬的中子能谱,有利于提高MA的裂变截面与裂变吸收比.全堆MA嬗变支持比为8.3,具有较好的嬗变效果.由于堆芯内区的高通量,堆芯内外区的嬗变率有明显差异,将MA集中布置于内区有利于减少装料量,改善总体嬗变效果.     

小型模块化氯盐快堆焚烧超铀核素物理特性研究

氯盐快堆具有重金属溶解度高和能谱较硬等特性,是嬗变超铀核素(Transuranic elements,TRU)的理想堆型。本文提出了一种50 MW小型模块化氯盐快堆(small-Modular Chlorine salt Fast Reactor,sm-MCFR)方案,对其焚烧TRU特性进行了初步研究。采用了基于SCALE(Standardized Computer Analyses for Licensing Evaluation)和MODEC(MOlten Salt Reactor Specific DEpletion Code)开发的耦合程序TMCBurnup(TRITON MODEC Coupled Burnup Code),计算并分析了sm-MCFR在TRU+232Th和TRU+DU(Depleted Uranium)两种燃料方案下的临界、燃耗、核素演化和嬗变TRU等物理性能。结果表明：在sm-MCFR运行期间,为维持临界状态,需在线添加TRU,以确保有效增殖系数k_eff>1;满功率运行40 a时,采用TRU+Th燃料方案下堆芯剩余TRU量为657...     

DRAGON&DONJON程序在MSR中堆芯燃耗计算的适用性

DRAGONDONJON组件-堆芯"两步法"程序通过合理简化,理论可适用于任何堆芯与工况。使用蒙特卡罗方法 RMC(Reactor Monte Carlo code)、MCNP(Monte Carlo Neutron Particle transport code)程序验证DRADON程序是否能够承担快/热谱型熔盐堆(Molten Salt Reactor,MSR)焚烧TRU、Th U燃料燃耗计算。选出熔盐增殖堆(Molten Salt Breeder Reactor,MSBR)与熔盐锕系元素再循环和嬗变堆(Molten Salt Advanced Reactor Transmuter,MOSART)堆型进行比较,同时分别利用RMC程序验证DRAGON程序组件燃耗计算的准确性,利用MCNP程序验证DRAGON程序组件均匀化方法以及DONJON程序截面调用和程序全堆扩散的准确性。结果表明,组件燃耗计算中,TRU和Th U燃料满足燃耗计算要求;堆芯临界计算中,快/热谱堆芯计算误差均小于0.001。证明DRADON程序可以胜任快、热谱型MSR焚烧TRU、Th U燃料的物理计算任务。     

加速器驱动的次临界系统的燃耗分析计算和堆芯优化设计

以加速器驱动的次临界系统（ADS）在事故情况下仍处于次临界、k_eff随燃耗时间变化的最大范围不超过1.5%和包壳材料HT9钢可承受的最大辐照损伤的前提下,将堆芯燃料区分为嬗变区和增殖区,并将整个过程保持嬗变区的燃料成分不变。通过对ADS燃料的组成成分、堆芯布置和堆芯功率分布等方面的研究,在Pu的外层富集度与内层富集度之比为1.0～1.5范围内,调整增殖区的燃料成分,并利用MCNP和ORIGEN耦合的COUPLED2程序计算k_eff随燃耗时间的变化。同时,综合考虑功率展平、次锕系核素的嬗变率和燃耗深度等因素,建立1套符合工程实际的次临界系统。     

小型模块化增殖焚烧快堆MA嬗变性能分析

次锕系核素(Minor Actinides,MA)作为长寿命高放射性核废料,在乏燃料放射性中占据主导位置。乏燃料最小化是保证核能可持续发展的重要环节,而嬗变是安全处置乏燃料的有效途径。小型模块化增殖焚烧(Breed and Burn,BB)快堆的中子经济性较好,燃烧寿期长,装料方式灵活多样,可用于增殖产生易裂变核燃料、嬗变长寿命核废料,从而解决核电发展中前端核燃料供给和后端乏燃料处理问题。本文分析对比了U3-MA和U5-MA燃料装载模式的临界、燃耗和安全性能,并系统研究了两种装料模式在BB快堆上嬗变MA的性能。结果表明:两种装料方式均能达到较好的嬗变性能,且MA的添加还能使反应堆寿期更长,堆芯中子经济性更高;此外,从安全性能上来看,添加MA对钍铀燃料循环的缓发中子份额影响较弱,但是对其燃料多普勒系数影响较强,这为后续钍铀、铀钚燃料循环选取合理的MA装载份额提供了参考依据。     

熔盐燃料对加速器驱动的次临界堆中子学性能的影响

加速器驱动的次临界熔盐堆(Accelerator-Driven Subcritical Molten Salt Reactor,ADS-MSR)结合了熔盐堆与ADS的许多优点,在先进核燃料利用方面有独特的优势。为了研究熔盐燃料的使用对ADS系统堆芯的中子学性能的影响,基于MCNP(Monte Carlo N Particle Transport Code)程序,分别计算并分析了熔盐燃料对加速器驱动的次临界堆的外源质子效率、中子能谱以及钍铀转换比等参数的影响。结果表明:相较于氧化物燃料,熔盐燃料的使用将会增加对外源中子和裂变中子的慢化,并且会提高堆芯的入射质子效率。同时,由于熔盐燃料的慢化效应,FLi Be和FLi熔盐燃料燃耗初期的钍铀转换比(CR)分别为1.023和1.062,略低于氧化物燃料的1.068。另一方面,熔盐燃料的在线处理会极大降低燃耗过程中的反应性损失。通过在线燃料处理和在线添料,FLi熔盐和FLi Be熔盐燃料的CR分别在燃耗运行的第1年和第3年超过氧化物燃料,并且能够长期稳定在1.06和1.00左右。     

Am装载对加速器驱动次临界系统瞬态过程的影响分析

加速器驱动次临界系统(ADS)由于外中子源驱动的特点带来大安全裕量,其被广泛设计用于嬗变次锕系(MA)核素。堆芯装载MA核素会减弱燃料的负多普勒效应、增加冷却剂的正温度系数并减小有效缓发中子份额,从而影响系统对扰动的瞬态响应,其中Am同位素的影响尤为重要。本文基于装载金属燃料的ADS嬗变方案,分析了不同Am装载份额条件下系统对于束流过功率和无保护瞬态过功率事故的瞬态过程,发现在深次临界状态下Am装载量对ADS的瞬态过程影响较小,说明ADS具有良好的安全特性,适宜高MA份额燃料的装载。     

双区氯盐快堆的增殖及嬗变性能分析

基于熔盐嬗变堆(Molten Salt Actinide Recycler and Transmuter,简称MOSART)堆芯结构对氯盐快堆(Molten Chloride Salt Fast Reactor,简称MCFR)进行了优化,分析了熔盐成分和后处理方式的影响,使其燃耗性能得到明显的提升,但是相比熔盐快堆(Molten Salt Fast Reactor,简称MSFR)的增殖及嬗变性能仍有一定差距。基于在线连续添料与后处理方式,采用SCALE6.1程序和熔盐堆在线添料和后处理程序(Molten Salt Reactor Reprocessing Sequence,简称MSR-RS)分析了堆芯结构、~(37)Cl富集度对增殖比(Breeding Ratio,简称BR)、核素吸收率、燃耗等方面的影响,提出了双区氯盐快堆的设计,进一步提升了增殖嬗变性能和钍基燃料的利用率,倍增时间缩短到20年左右,超铀核素(Transuranics,简称TRU)嬗变率达到68%左右。     

加速器驱动的次临界系统初步概念设计

基于初始Pu装载对加速器驱动的次临界系统(ADS)嬗变次锕系核素（MA）的影响,提出了6种采用(TRU-10Zr)-Zr*弥散体燃料的ADS概念设计方案。运用MCNP与ORIGEN2程序对ADS嬗变MA堆芯进行稳态与燃耗计算,比较分析MA的嬗变效果、有效增殖因数k_eff、质子束流流强I_p与初始Pu含量的关系。计算结果表明：随着初始Pu含量的增加,MA的嬗变率减小,初始I_p增大;初始Pu含量小于33%,k_eff随时间的变化是先增大后减小,大于33%后一直减小,且随着初始Pu含量的增加,k_eff减小得更加明显。故初始钚含量为33%的方案为最佳,其k_eff的相对变化不超过1%,I_p小于20 mA,MA嬗变率高达28.06%,嬗变支持比为29.23,满足初步设计要求。     

TRU燃料热谱MSR定期换料及嬗变特性数值研究

熔盐反应堆（MSR）燃料制备方便、中子经济性好、燃料管理灵活,具有直接利用轻水堆乏燃料中超铀核素（TRU）的潜力。本文通过优化燃料选取、栅格参数及燃料/石墨体积分数和去除裂变气体和惰性金属等方法,对TRU燃料热谱MSR堆芯寿期、TRU核素积存量、次锕系核素MA嬗变支持比和TRU焚毁率等进行计算分析,证明TRU燃料热谱MSR可实现长周期定期换料,减少在线换料的难度,同时对MA和TRU核素具有一定的嬗变能力,可降低乏燃料放射性毒性。      

加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

运用MCNP与ORIGEN2耦合计算程序COUPLE，对加速器驱动的次临界系统（ADS）钠冷金属燃料快堆堆芯进行稳态与燃耗计算，比较分析次锕系核素（MA）非均匀布置堆芯与均匀布置堆芯在MA嬗变效果与反应性参数方面的差异。计算结果表明，对比均匀布置，非均匀布置具有更高的MA嬗变率与嬗变支持比，在反应性参数方面导致多普勒效应与有效缓发中子分额降低，钠空泡效应增大，在堆芯功率分布与加速器束流功率方面没有明显变化。     

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

Thorium (Th) oxide fuel offers a significant advantage over traditional low-enriched uranium and mixed uranium/plutonium oxide (MOX) fuel irradiated in a Light Water Reactor. The benefits of using thorium include the following: 1) unlike depleted uranium, thorium does not produce plutonium, 2) thorium is a more stable fuel material chemically than LEU and may withstand higher burnups, 3) the materials attractiveness of plutonium in Th/Pu fuel at high burnups is lower than in MOX at currently achievable burnups, and 4) thorium is three to four times more abundant than uranium. This paper quantifies the irradiation of thorium fuel in existing Light Water Reactors in terms of: 1) the percentage of plutonium destroyed, 2) reactivity safety parameters, and 3) material attractiveness of the final uranium and plutonium products. The Monte Carlo codes MCNP/X and the linkage code Monteburns were used for the calculations in this document, which is one of the first applications of full core Monte Carlo burnup calculations. Results of reactivity safety parameters are compared to deterministic solutions that are more traditionally used for full core computations.Thorium is fertile and leads to production of the fissile isotope ²³³U, but it must be mixed with enriched uranium or reactor-/weapons-grade plutonium initially to provide power until enough ²³³U builds in. One proposed fuel type, a thorium-plutonium mixture, is advantageous because it would destroy a significant fraction of existing plutonium while avoiding the creation of new plutonium. ²³³U has a lower delayed neutron fraction than ²³⁵U and acts kinetically similar to ²³⁹Pu built in from ²³⁸U. However, as with MOX fuel, some design changes may be required for our current LWR fleet to burn more than one-third a core of Th/Pu fuel and satisfy reactivity safety limits. The calculations performed in this research show that thorium/plutonium fuel can destroy up to 70% of the original plutonium per pass at 47 GWd/MTU, whereas only about 30% can be destroyed using MOX. Additionally, the materials attractiveness of the final plutonium product of irradiated plutonium/thorium fuel is significantly reduced if high burnups (∼94 GWD/MTU) of the fuel can be attained.     

Plutonium recycling in hydride fueled PWR cores

The classic approach to the recycling of Pu in PWR is to use mixed U-oxide Pu-oxide (MOX) fuel. The mono-recycling of plutonium in PWR transmutes less than 30% of the loaded plutonium, providing only a limited reduction in the long-term radiotoxicity and in the inventory of TRU to be stored in the repository. The primary objective of this study is to assess the feasibility of plutonium recycling in PWR in the form of plutonium hydride, PuH₂, mixed with uranium and zirconium hydride, ZrH_1.6, referred to as PUZH, that is loaded uniformly in each fuel rod. The assessment is performed by comparing the performance of the PUZH fueled core to that of the MOX fueled core. Performance characteristics examined are transmutation effectiveness, proliferation resistance of the discharged fuel and fuel cycle economics. The PUZH loaded core is found superior to the MOX fueled core in terms of the transmutation effectiveness and proliferation resistance. For the reference cycle duration and reference fuel rod diameter and pitch, the percentage of the plutonium loaded that is transmuted in one recycle is 53% for PUZH versus 29% for MOX fuel. That is, the net amount of plutonium transmuted in the first recycle is 55% higher in cores using PUZH than in cores using MOX fuel. Relative to the discharged MOX, the discharged PUZH fuel has smaller fissile plutonium fraction - 45% versus 60%, 15% smaller minor actinides (MA) inventory and more than double spontaneous fission neutron source intensity and decay heat per gram of discharged TRU. Relative to the MOX fuel assembly, the radioactivity of the PUZH fuel assembly is 26% smaller and the decay heat and the neutron yield are only 3% larger. The net effect is that the handling of the discharged PUZH fuel assembly will be comparable in difficulty to that of the discharged MOX assembly while the proliferation resistance of the TRU of the discharged PUZH fuel is enhanced.     

Comparative study of fast critical burner reactors and subcritical accelerator driven systems and the impact on transuranics inventory in a regional fuel cycle

In the frame of Partitioning and Transmutation (P&T) strategies, many solutions have been proposed in order to burn transuranics (TRU) discharged from conventional thermal reactors in fast reactor systems. This is due to the favourable feature of neutron fission to capture cross section ratio in a fast neutron spectrum for most TRU. However the majority of studies performed use the Accelerator Driven Systems (ADS), due to their potential flexibility to utilize various fuel types, loaded with significant amounts of TRU having very different Minor Actinides (MA) over Pu ratios. Recently the potential of low conversion ratio critical fast reactors has been rediscovered, with very attractive burning capabilities. In the present paper the burning performances of two systems are directly compared: a sodium cooled critical fast reactor with a low conversion ratio, and the European lead cooled subcritical ADS-EFIT reactor loaded with fertile-free fuel. Comparison is done for characteristics of both the intrinsic core and the regional fuel cycle within a European double-strata scenario. Results of the simulations, obtained by use of French COSI6 code, show comparable performance and confirm that in a double strata fuel cycle the same goals could be achieved by deploying dedicated fast critical or ADS-EFIT type reactors. However the critical fast burner reactor fleet requires ∼30-40% higher installed power then the ADS-EFIT one. Therefore full comparative assessment and ranking can be done only by a parametric sensitivity study of both the fuel cycle and the electricity generating costs.     

基于单通道模型的ADS堆芯物理热工耦合计算

利用单通道模型,开发了铅铋冷却加速器驱动次临界系统(ADS)堆芯组件温度和密度分布的热工计算程序,并将该程序与MCNP耦合,建立了物理热工耦合计算方法。利用该方法计算了耦合后的功率及热工参数。结果表明,堆芯组件温度及冷却剂流速满足热工安全限值,堆芯径向功率不均匀系数较大,堆芯设计需进一步优化。     

TRU self-recycling in a high temperature modular helium reactor

The Deep Burn Project is developing high burnup fuel based on Ceramically Coated (TRISO) particles, for use in the management of spent fuel Transuranics. This paper evaluates the TRU deep-burn in a High Temperature Reactor (HTR) that recycles its own transuranic production. The DB-HTR is loaded with standard LEU fresh fuel and the self-generated TRUs are recycled into the same core (after reprocessing of the original spent fuel). This mode of operation is called self-recycling (SR-HTR). The final spent fuel of the SR-HTR can be disposed of in a final repository, or recycled again.In this study, a single recycling of the self-generated TRUs is considered. The UO₂ fuel kernel is 12% uranium enrichment and the diameter of the kernel is 500 μm. TRISO packing fraction of UO₂ fuel compact is 26%. In the SR-HTR fuel cycle, it is assumed that the spent UO₂ fuel is reprocessed with conventional technology and the recovered TRUs are fabricated into Deep Burn TRISO fuel. The diameter of 200 μm is used for the TRU fuel kernel. A typical coating thickness is used. The core performance is evaluated for an equilibrium cycle, which is obtained by cycle-wise depletion calculations. From the analysis results, the equilibrium cycle lengths of Case 1 (5-ring fuel block SR-HTR) and Case 2 (4-ring fuel block SR-HTR) are 487 and 450 EFPDs (effective full power days), respectively. And the UO₂ fuel discharge burnups of Case 1 and Case 2 are 10.3% and 10.1%, respectively. Also, the TRU discharge burnups of Case 1 and Case 2 are 64.7% and 63.5%, respectively, which is considered extremely high. The fissile (Pu-239 and Pu-241) content of the self-generated TRU vector is about 52%. The deep-burning of TRU in SR-HTR is partly due to the efficient conversion of Pu-240 to Pu-241, which is boosted by the uranium fuel in SR-HTR. It is also observed that the power distribution is quite flat within the uranium fuel zone. The lower power density in TRU fuel is because the TRU burnup is very high. Also, it is found that transmutation of Pu-239 is near complete in SR-HTR and that of Pu-241 is extremely high in all cases. The decay heat of the SR-HTR core is very similar to the UO₂-only core. However, accumulation of the minor actinides is not avoidable in the SR-HTR core. The extreme high burnup of the Deep Burn fuel greatly reduces the amount of heat producing isotopes that could be problematic in spent fuel repositories (like Pu-238).     

Effect of TRU fuel loading on core performance and plutonium production of FBR

The effect of trans-uranium (TRU) fuel loading on the reactor core performances as well as the actinide and isotopic plutonium compositions in the core and blanket regions has been analyzed based on the large FBR type. Isotopic plutonium composition of TRU fuel is less than that of MOX fuel except for Pu-238 composition which obtains relatively higher composition. A significant increase of plutonium vector composition is shown by Pu-238 for TRU fuel in the core region as well as its increasing value in the blanket region for doping MA case. Excess reactivity can be reduced significantly (5% at beginning of cycle) and an additional breeding gain can be obtained by TRU fuel in comparison with MOX fuel. Doping MA in the blanket regions reduces the criticality for a small reduction value (0.1%) and it gives a reduction value of breeding ratio. Loading MA in the core regions as TRU fuel composition gives relatively bigger effect to increase the void reactivity coefficient mean while it gives less effect for loading MA in the blanket regions. Similar to the void reactivity coefficient profile, loading MA is more effective to the change of Doppler coefficient in the core regions in comparison with loading MA in the blanket regions which gives slightly less negative Doppler coefficient. Obtained Pu-240 vector compositions in the core region are categorized as practically unusable composition for nuclear device based on the Pellaud's criterion. Less than 7% Pu-240 vector compositions in the blanket region are categorized as weapon grade composition for no doping MA case. Obtaining 9% of Pu-238 composition by doping MA 2% in the blanket regions is enough to increase the level of proliferation resistance for denaturing plutonium based on the Kessler's criterion.     

Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-I, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-I were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes.A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core.The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.     

Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.