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.

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.

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.

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.

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.

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.

Prediction of the cross-sections of isotopes produced in deuteron-induced spallation of long-lived fission products

The spallation cross-section data for the long-lived fission products (LLFPs) are scarce but required for the design of accelerator driven systems. In this paper, the isospin dependent quantum molecular dynamics model and the statistical code GEMINI are applied to simulate deuteron-induced spallation in the energy region of GeV/nucleon. By comparing the calculations with the experimental data, the applicability of the model is verified. The model is then applied to simulate the spallation of 90Sr, 93Zr, 107Pd, and 137Cs induced by deuterons at 200, 500 and 1000 MeV/nucleon. The cross-sections of isotopes, the cross-sections of long-lived nuclei, and the reaction energy are presented. Using the above observables, the feasibility of LLFP transmutation by spallation is discussed.

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

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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。

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.

Uncertainty and sensitivity analysis of in-vessel phenomena under severe accident mitigation strategy based on ISAA-SAUP program

The phenomenology involved in severe accidents in nuclear reactors is highly complex.Currently,integrated analysis programs used for severe accident analysis heavily rely on custom empirical parameters,which introduce considerable uncertainty.Therefore,in recent years,the field of severe accidents has shifted its focus toward applying uncertainty analysis methods to quantify uncertainty in safety assessment programs,known as“best estimate plus uncertainty(BEPU).”This approach aids in enhancing our comprehension of these programs and their further development and improvement.This study concentrates on a third-generation pressurized water reactor equipped with advanced active and passive mitigation strategies.Through an Integrated Severe Accident Analysis Program(ISAA),numerical modeling and uncertainty analysis were conducted on severe accidents resulting from large break loss of coolant accidents.Seventeen uncertainty parameters of the ISAA program were meticulously screened.Using Wilks'formula,the developed uncertainty program code,SAUP,was employed to carry out Latin hypercube sampling,while ISAA was employed to execute batch calculations.Statistical analysis was then conducted on two figures of merit,namely hydrogen generation and the release of fission products within the pressure vessel.Uncertainty calculations revealed that hydrogen production and the fraction of fission product released exhibited a normal distribution,ranging from 182.784 to 330.664 kg and from 15.6 to 84.3%,respectively.The ratio of hydrogen production to reactor thermal power fell within the range of 0.0578–0.105.A sensitivity analysis was performed for uncertain input parameters,revealing significant correlations between the failure temperature of the cladding oxide layer,maximum melt flow rate,size of the particulate debris,and porosity of the debris with both hydrogen generation and the release of fission products.

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。

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.

Waste Transmutation and Nuclear Energy Generation Using a Tokamak Fusion-Fission Hybrid Reactor

A tokamak fusion-fission hybrid reatcor is proposed as one of candidates for disposal ofthe long-lived actinides and fission product wastes and supply of future energy.To assess thefeasibility of transmutation of long-lived radiowastes using fusion-fission hybrid reactors,afusion core design is presented and several possible conceptual blankets are studied,for,re-spectively,actinides transmutation and fission product transmutation.The results show thatactinides and fission products may be effectively transmuted using the presented hybrid reac-tors.

三代非能动核电厂乏燃料贮运系统衰变热计算及关键因素研究

本文以三代非能动核电厂国和一号乏燃料组件为研究对象,基于ANS5.1—2005和ORIGEN-S软件的衰变热计算开展对比研究,分析了不同冷却时间下裂变产物、锕系元素衰变热变化规律。结果显示,对于冷却时间在5年以上的乏燃料,锕系元素占总衰变热的贡献接近20%甚至更高,锕系元素的主要贡献来自于Cm-244、Pu-238和Am-241。ANS 5.1-2005对锕系元素仅考虑了U-239和Np-239,对于冷却时间较长的乏燃料贮运系统,相对ANS 5.1—2005,ORIGEN-S衰变热计算结果更加保守。建议在三代非能动核电厂乏燃料的贮运系统衰变热计算中采用基于ORIGEN-S等类似功能的专用程序进行计算。

镎与裂片元素分离工艺研究

本文设计并验证了镎与裂片元素分离的工艺。所设计的镎提取纯化工艺分为两个工艺段,镎提取与纯化工艺段(2NA)和镎反萃工艺段(2NB)。2NA萃取槽工艺:实现镎钚的共同萃取,镎浓缩3倍,同时与裂片元素进行分离;设计工艺条件下,在2NA工艺段,Np收率大于99.9%,2NA段对Tc的净化为4.0×10^(4),2NA段对Ru的净化为2.1×10^(3),2NA段对Zr的净化为4.0×10^(4);2NB反萃取槽:将镎反萃进入水相。在2NB工艺段,Np反萃率大于99.9%,Np与裂片没有净化效果。验证了提取工艺对高浓度镎提取的可行性。结合之前工作,基本可以确定设计的流程可以实现用萃取法从中放废液中纯化和浓缩镎的目的。

快堆氧化物燃料微观组织和裂变产物研究进展

氧化物燃料作为快堆燃料元件的核心部件,在堆内服役条件下会发生一系列结构变化和裂变产物的扩散、迁移和释放,一直是国内外关注和研究的重点。针对快堆氧化物燃料微观组织和裂变产物研究状和进展进行了综述,总结了快堆氧化物燃料现有研究的不足、需要解决的问题以及技术发展动态,提出了未来需要开展研究的一些思路,以期为我国快堆氧化物燃料辐照性能研究提供部分参考,研究成果可以为快堆氧物燃料的优化设计、快堆高燃耗安全运行提供理论依据技术支撑。

快堆严重事故钠燃烧过程裂变产物释放模拟试验研究

钠燃烧过程产生的裂变产物及钠气溶胶迁移是快堆严重事故重要的源项之一。本研究对钠燃烧过程裂变产物随钠蒸汽和钠气溶胶迁移的行为进行分析,针对钠蒸发作用下裂变产物释放、钠燃烧作用下裂变产物释放以及气相空间气溶胶迁移行为分别提出了物理模型,并在确定计算方法的基础上通过CFD软件建模进行了仿真计算,最后通过开展小规模钠燃烧试验,获取了真实钠燃烧过程裂变产物沉降数据,对计算模型进行了修正和补充。试验数据与仿真计算结果表明,气溶胶迁移模型能够较好地表征裂变产物及钠气溶胶迁移行为,钠燃烧作用下裂变产物的释放系数为10-3时计算结果与试验结果较吻合。

用于制备^(99)Mo的低浓铀靶件研究进展

^(99m)Tc是目前核医学临床应用最广泛的放射性核素,^(99m)Tc核素一般从其母体核素^(99)Mo衰变得到。目前^(99)Mo主要由^(235)U靶件辐照、提取制备。出于防止核扩散的考虑,铀靶正在由高浓缩铀(HEU)向低浓缩铀(LEU)转化,本研究对各种LEU铀靶的特点进行详细阐述,对靶件的优缺点进行总结,并对进一步研究工作提出建议。