Study on dynamic characteristics of fission products in 2 MW molten salt reactor

In this study,a numerical flow model of the fission products(FPs)in the primary loop system of a molten salt reactor(MSR)was established and solved using Mathematica 7.0.The simulation results were compared with those of the ORIGEN-S program in the static burnup mode,and the deviation was found to be less than 10%,which indicates that the results are in good agreement.Furthermore,the FPs distribution in the primary loop system under normal operating conditions of the 2 MW MSR was quantitatively analyzed.In addition,the distribution phenomenon of the FPs under different flow rate conditions was studied.At the end of life,the FPs activity in the core region(including active region,and upper and lower plenum regions)accounted for 77.3%,and that in the hot leg #1,main pump,hot leg #2,heat exchanger,and cold leg region accounted for 1.2%,16.15%,0.99%,2.5%,and 1.9%,respectively,of the total FPs in the primary loop under normal operating conditions.The proportion of FPs in the core decreased with the increase in flow rate in the range of 2.24-22,400 cm^3 s^-1.The established analytical method and conclusions of this study can provide an important basis for radiation safety design of the primary loop,radioactive source management design,thermal-hydraulic safety analysis,and radiochemical analysis of FPs of 2 MW MSRs.

Numerical analysis on element creation by nuclear transmutation of fission products

A burnup calculation was performed to analyze the Apr`es ORIENT process, which aims to create highlyvaluable elements from fission products separated from spent nuclear fuels. The basic idea is to use nuclear transmutation induced by a neutron capture reaction followed by a β-decay, thus changing the atomic number Z of a target element in fission products by 1 unit. LWR(PWR) and FBR(MONJU) were considered as the transmutation devices. High rates of creation were obtained in some cases of platinum group metals(44Ru by FBR,46 Pd by LWR) and rare earth(64Gd by LWR,66 Dy by FBR). Therefore, systems based on LWR and FBR have their own advantages depending on target elements. Furthermore, it was found that creation rates of even Z(= Z + 1) elements from odd Z ones were higher than the opposite cases. This creation rate of an element was interpreted in terms of "average 1-group neutron capture cross section of the corresponding target element σc Z defined in this work. General trends of the creation rate of an even(odd) Z element from the corresponding odd(even) Z one were found to be proportional to the 0.78th(0.63th) power of σc Z, however with noticeable dispersion. The difference in the powers in the above analysis was explained by the difference in the number of stable isotopes caused by the even-odd effect of Z.

Study on High-yielding Seed Production Technologies of Japonica Hybrid Rice

[Objective] This study aimed to investigate the high-yielding seed production technologies of japonica hybrid rice. [Method] Key technologies affecting the seed production of japonica hybrid rice were investigated based on manual transplanting experiment, mechanical transplanting experiment, paclobutrazol chemical control experiment, leaf-clipping experiment, gibberellic acid treatment experiment and experiment of different row ratios of male and female parents. [Result] The duration from seeding to heading was significantly shortened with sowing date postponing. The sowing date of mechanical-transplanting parents should be 3 -4 d earlier than that of manual-transplanting parents. Spraying paclobutrazol had significantly control effect on plant height of rice seedlings, and the spraying dose should be determined based on the sensitivities of different parents to paclobutrazol, with an appropriate range of 600-900 g/hm2 . Leaf clipping had certain improving effect on the seed-setting rate. The appropriate spraying dose of gibberellin for japonica hybrid rice was 75-90 g/hm2 . The yield would be relatively high when row ratio of male and female parent was 6∶2. [Conclusion] This study laid the foundation for improving the yield of produced japonica hybrid rice seeds.

First-principle studies of radioactive fission productions Cs/Sr/Ag/I adsorption on chrome-molybdenum steel in Chinese 200 MW HTR-PM

Chrome-molybdenum steel(2·1/4Cr1Mo) is one of the main products of steam generation.The adsorption behaviors of radioactive fission products on2·1/4Cr1Mo surface are critical in the analysis of HTR-PM.Here,the adsorption behavior of cesium,strontium,silver and iodine on 2·1/4Cr1Mo was investigated with first-principle calculations that the Ag and I atoms prefer to be adsorbed at the square hollow site of the face-centered cubic iron cell with a binding energy of about 1 and 3 eV,respectively.In contrast,Cs and Sr atoms are not adsorbed on the surface of the 2·1/4Cr1Mo.These results are again confirmed via analysis of charge density differences and the densities of state.Furthermore,the adsorption rates of these fission products show that only I and Ag have significant adsorption on the metal substrate.These adsorption results explain the amount of adsorbed radionuclides for an evaluation of nuclear safety in HTR-PM.These micro-pictures of the interaction between fission products and materials are a new and useful way to analyze the source term.

Verification of neutron-induced fission product yields evaluated by a tensor decompsition model in transport-burnup simulations

Neutron-induced fission is an important research object in basic science.Moreover,its product yield data are an indispensable nuclear data basis in nuclear engineering and technology.The fission yield tensor decomposition(FYTD)model has been developed and used to evaluate the independent fission product yield.In general,fission yield data are verified by the direct comparison of experimental and evaluated data.However,such direct comparison cannot reflect the impact of the evaluated data on application scenarios,such as reactor transport-burnup simulation.Therefore,this study applies the evaluated fission yield data in transport-burnup simulation to verify their accuracy and possibility of application.Herein,the evaluated yield data of235U and239Pu are applied in the transport-burnup simulation of a pressurized water reactor(PWR)and sodium-cooled fast reactor(SFR)for verification.During the reactor operation stage,the errors in pin-cell reactivity caused by the evaluated fission yield do not exceed 500 and 200 pcm for the PWR and SFR,respectively.The errors in decay heat and135Xe and149Sm concentrations during the short-term shutdown of the PWR are all less than 1%;the errors in decay heat and activity of the spent fuel of the PWR and SFR during the temporary storage stage are all less than 2%.For the PWR,the errors in important nuclide concentrations in spent fuel,such as90Sr,137Cs,85Kr,and99Tc,are all less than 6%,and a larger error of 37%is observed on129I.For the SFR,the concentration errors of ten important nuclides in spent fuel are all less than 16%.A comparison of various aspects reveals that the transport-burnup simulation results using the FYTD model evaluation have little difference compared with the reference results using ENDF/B-Ⅷ.0 data.This proves that the evaluation of the FYTD model may have application value in reactor physical analysis.

Yield of long-lived fission product transmutation using proton-, deuteron-, and alpha particle-induced spallation

The transmutation of long-lived fission products through spallation induced by light nuclides was investi-gated for the purpose of determining the feasibility of this approach for long-lived fission products,in both economic and environmental terms.The cross-section data were obtained from the TALYS Evaluated Nuclear Data Library(TENDL).A thick target model was used to study the consumption of the target isotopes in the transmutation process.The transmutation yield was calculated using the highest beam intensity available with the China initiative accelerator-driven system.It was found that the light nuclide-induced spallation reaction can significantly reduce the radio toxicity of the investigated long-lived fission products.Using the transmutation target made of elemental LLFP and the proton beam with an intensity of 5 mA,the consumption of 90 Sr,93 Zr,107 Pd,or 137 Cs can reach approximately 500 g per year.

Morphological Effects of Natural Products on Schizosaccharomyces pombe Measured by Imaging Flow Cytometry

Gaining a full understanding of the mechanisms of action of natural products as therapeutic agents includes observing the effects of natural products on cellular morphology,because abnormal cellular morphology is an important aspect of cellular transformations that occur as part of disease states.In this study a set of natural products was examined in search of small molecules that influence the cylindrical morphology of fission yeast Schizosaccharomyces pombe.Imaging flow cytometry of large populations of S.pombe exposed to natural products captured cell images and revealed changes in mean length and aspect ratio of cells.Several natural products were found to alter S.pombe’s morphology relative to control,in terms of elongating cells,shrinking them,or making them more round.These results may facilitate future investigations into methods by which cells establish and maintain specific shapes.

Structural integrity evaluation of irradiated LEU targets for the production of molybdenum‑99 using thermo‑mechanical behavior simulation coupled with pressure of fission gas release calculation

Irradiated low-enriched uranium as target plates is used to produce,via neutron radiation and from the molybdenum-99 fission product,technetium-99m,which is a radio-element widely used for diagnosis in the field of nuclear medicine.The behavior of this type of target must be known to prevent eventual failures during radiation.The present study aims to assess,via prediction,the thermal–mechanical behavior,physical integrity,and geometric stability of targets under neutron radiation in a nuclear reactor.For this purpose,a numerical simulation using a three-dimensional finite element analysis model was performed to determine the thermal expansion and stress distribution in the target cladding.The neutronic calculation results,target material properties,and cooling parameters of the KAERI research group were used as inputs in our developed model.Thermally induced stress and deflection on the target were calculated using Ansys-Fluent codes,and the temperature profiles,as inputs of this calculation,were obtained from a CFD thermal–hydraulic model.The stress generated,induced by the pressure of fission gas release at the interface of the cladding target,was also estimated using the Redlich–Kwong equation of state.The results obtained using the bonded and unbonded target models considering the effect of the radiation heat combined with a fission gas release rate of approximately 3%show that the predicted thermal stress and deflection values satisfy the structural performance requirement and safety design.It can be presumed that the integrity of the target cladding is maintained under these conditions.

ELECTRON MICROSCOPIC AUTORADIOGRAPHIC STUDY ON SUBCELLULAR LOCALIZATION OF FISSIONPRODUCT ^（147）Pm INTISSUE CELLS

The early risk of internal contaminated accumulation of 147Pm is in blood cells and endothelial cells, especially in red blood cells. Then 147Pm is selectively deposited in ultrastructure of liver cells, such as in nucleus, nucleolus, rough endoplasmic reticulum, mitochondria and microbodies. Dense tracks also appear in mitochondria and lysosome of pedal cells within renal corpuscle, and so does in nucleus as well as in mitochondria and microbodies of epicyte of kidney near-convoluted tubule. With the prolongation of observing time, 147Pm is selectively and steadily deposited in subcellular level of organic component for bone. Substantial amount of 147Pm is taken up into the nuclear fraction of osteoclasts and osteoblasts. Particularly, in organelles 147Pm is mainly accumulated in rough endoplasmic reticulum and in mitochondria.Autoradiographic tracks especially localize in combined point between Golgi complex and transitive vesicle of rough endoplasmic reticulum. In addition, numerous 147Pm deposited in collagenous fibre within interstitial of bone cells is hardly excreted.

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.

Analysis of Fission Fragments Contributors on Total Decay Heat of Thermal Neutron-Induced Fission of U-235

Calculation of the decay heat from the decay/buildup of radionuclides generated after nuclear fission is one of the highest priorities in the nuclear industry. These calculations become more important if they are made together with the analysis of the most important isotopes affecting the decay heat. They are useful in designing the necessary nuclear safety for spent fuels, and their importance cannot be overlooked in the designs of transporting fuel storage containers as well as in the management of the radioactive waste generated. In this paper, by using MATLAB, the decay heat after the thermal fission of a U-235 nucleus was numerically calculated by solving linear differential equations for all the buildups/decays of the fission products. Also, the most contribution of radioactive isotopes to the decay heat was analyzed by using Microsoft Excel. The most influential isotopes were deduced in two ways;either by calculating the most influential isotopes at specific times, or by determining the largest influences in a cumulative manner. All required nuclear data such as decay constants their branching ratios, independent fission yield, and average α-, β-, and γ-energies released per disintegration of any nuclide, have been extracted from the latest version of the Evaluated Nuclear Data Files (ENDF) database ENDF/B-VIII.0. The two different methods used showed a difference in the contributing isotopes, which is logical for the difference in the method of calculation. The first method is suitable for instantaneous data while the second method is more suitable when there is a need to know the cumulative calculations. In sum, we can say that both methods complement each other, and neither of them can be dispensed with in the accurate calculations related to transportation and storage of spent fuel.

MOX燃料与包壳化学相互作用研究进展

简要介绍了MOX燃料芯块微观组织特点和主要裂变产物行为及其对化学相互作用层的影响,归纳总结了国内外对化学相互作用层微观结构的研究进展,分析了现有研究的不足和仍待解决的问题,以期对我国未来MOX燃料的研究和应用提供部分参考。

FLiNaK熔盐中CsF的定向凝固分离研究

熔盐反应堆采用氟化物熔盐作为冷却剂和燃料载体,运行后的燃料成分为铀、钍、裂变产物和载体盐,对裂变产物进行分离并回收有效组分复用,可以提高反应堆的运行经济性,并且使放射性废物最小化。定向凝固技术是利用多元混合物的相平衡特性和凝固过程元素迁移机制实现物质分离,具有工艺操作简单、无副产物产生等优点,有望用于燃料盐中裂变产物分离。在自制的冷棒式定向凝固实验装置上,研究了FLiNaK熔盐体系内典型裂变产物CsF在不同工艺条件下定向凝固后的含量分布。研究结果表明:通过控制冷却凝固速度,得到的凝固盐中Cs元素的含量在径向上呈现出梯度分布,由内向外依次递减,外侧凝固盐中Cs含量相较液相中最高降低约90%,表明燃料盐中裂变产物定向凝固分离具有一定的可行性。

钍基熔盐反应堆内化学研究进展和展望

熔盐反应堆是第四代核能系统中唯一使用液态燃料的反应堆,在钍基熔盐反应堆研制和运行中有许多直接与化学相关的关键问题,堪比“化学堆”。于是,因熔盐反应堆研发和运行的需要诞生了放射化学在裂变能利用中的一门新分支学科——熔盐反应堆化学。本实验室利用加速器驱动的中子源和γ能谱分析技术开展了钍基熔盐反应堆化学研究。本文介绍了钍铀转换中间核素~(233)Pa和裂变产物~(131)I及~(95)Nb在熔盐反应堆模拟燃料盐中分布和行为的研究进展。基于对美国橡树岭国家实验室(ORNL)的熔盐反应堆实验装置运行中的燃料盐、锕系元素和裂变产物等相关若干问题分析,提出了在钍基熔盐反应堆框架内熔盐反应堆内化学方面应该进一步开展的研究内容,包括钍基熔盐反应堆运行的化学检测和诊断、影响熔盐氧化还原电势的因素、熔盐氧化还原电势检测的新技术等。熔盐反应堆化学研究的进一步深入将拓展熔盐反应堆化学实践和理论,使钍基熔盐反应堆化学水平提升到新高度,为未来钍基熔盐反应堆高效安全运行提供科学技术方面的支撑和保障。

压水堆破口事故下裂变产物源项释放及衰变热分析

反应堆严重事故时堆芯发生熔化现象,导致部分放射性源项不再留存于燃料中,将会通过各种途径释入环境,对周围造成严重的放射性污染。为了研究不同模型下裂变产物在压力容器内外释放量及其衰变热分布,分析喷淋系统对控制源项释放及热量的影响。基于典型的百万千瓦级压水堆核电站模型,利用一体化安全分析程序MAAP建模,分析计算CORSOR-M、CORSOR-O和ORNL-BOOTH三种源项释放模型反应堆一回路热管段破口叠加高、低压安注失效的事故序列和后果。结果表明:裂变产物源项主要在压力容器内释放,释放量远多于压力容器外的释放量。CORSOR-O模型下压力容器最晚融穿,安全壳失效最早;ORNL-BOOTH中压力容器虽最先融穿,但安全壳失效远晚于其他两种模型。源项释放差异导致不同模型衰变热现象不同,主要热源皆为挥发性裂变产物。开启喷淋不仅可以使悬浮碘化物充分控制在安全壳内,还能有效带走源项产生的衰变热,降低安全壳压力,保证安全壳完整性。

压水堆核电厂燃料棒破损诊断分析研究

燃料棒是核电厂包容放射性物质的第一道屏障。燃料棒破损会导致冷却剂裂变产物活度升高,严重时机组须在数小时内后撤到停堆。通过取样监测的冷却剂放射化学数据可以一定程度上反映堆芯内装载燃料棒的破损情况。本研究介绍了压水堆核电厂功率运行期间冷却剂内裂变产物的来源,分析了裂变产物通过反冲和扩散方式的产生机理,通过求解迁移方程得到稳态情况下裂变产物活度的解析解。基于最小二乘法对反冲释放和扩散释放的裂变产物释放产生比进行解谱,建立了诊断压水堆燃料棒破损时间、破口程度、锕系核素泄漏、燃耗和燃料批次的定量分析模型。采用某百万千瓦压水堆运行中发生二次氢化的燃料循环的冷却剂裂变产物监测数据进行了验证,理论模型的分析结果也与机组停堆后啜漏检查和热室检查结果相符。

高燃耗UO_(2)陶瓷燃料芯块微观结构演化行为研究进展

UO_(2)陶瓷燃料芯块作为轻水堆等商业堆的核心部件,在运行时的变化会影响反应堆运行安全。延长燃料元件使用燃耗从而达到更高经济效益是核电厂未来发展方向,然而随着燃料燃耗的增加,燃料芯块的微观结构如晶粒尺寸、孔隙等会发生变化,裂变产物的产生与迁移量增加,芯块包壳相互作用的发生概率升高导致反应堆运行安全系数下降,因此高燃耗下陶瓷燃料芯块微观结构演化行为和裂变产物迁移的研究受到重视。基于学者们所做的研究工作,总结了陶瓷UO_(2)芯块裂变产物研究相关的几个内容:边缘高燃耗结构的特征与形成机理、裂变产物的分布及影响,微观结构演化与裂变产物迁移的联系,讨论了高燃耗下芯块微观结构演化和其对裂变产物迁移的影响。为提高燃料使用燃耗,结合研究现状,对高燃耗下燃料芯块的进一步研究提出展望。

First-principles study on the diffusion behavior of Cs and I in Cr coating

Cs and I can migrate through fuel-cladding interfaces and accelerate the cladding corrosion process induced by the fuel-cladding chemical interaction.Cr coating has emerged as an important candidate for mitigating this chemical interaction.In this study,first-principles calculations were employed to investigate the diffusion behavior of Cs and I in the Cr bulk and grain boundaries to reveal the microscopic interaction mitigation mechanisms at the fuel-cladding interface.The interaction between these two fission products and the Cr coating were studied systematically,and the Cs and I temperature-dependent diffusion coefficients in Cr were obtained using Bocquet’s oversized solute-atom model and Le Claire’s nine-frequency model,respectively.The results showed that the Cs and I migration barriers were significantly lower than that of Cr,and the Cs and I diffusion coefficients were more than three orders of magnitude larger than the Cr self-diffusion coefficient within the temperature range of Generation-IV fast reactors(below 1000 K),demonstrating the strong penetration ability of Cs and I.Furthermore,Cs and I are more likely to diffuse along the grain boundary because of the generally low migration barriers,indicating that the grain boundary serves as a fast diffusion channel for Cs and I.

Nuclear decay database in fission product mass region

Accurate and reliable nuclear decay databases are essential for fundamental and applied nuclear research studies.However,decay data are not usually as accurate as expected and need improvement.Hence,a new Chinese nuclear decay database in the fission product mass region(A=66−172)based on several major national evaluated data libraries has been developed under joint efforts in the CNDC working group.A total of 2358 nuclides have been included in this decay database.Two main data formats,namely ENSDF and ENDF,have been adopted.For the total meanβandγenergies,available data from total absorption gamma ray spectroscopy measurements have been adopted.For some nuclides without experimental measurements,theoretically calculated values have been added.