Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

To perform nuclear reactor simulations in a more realistic manner,the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions.For simplicity,efficiency,and robustness,the matrixfree Newton/Krylov(MFNK)method was applied to the steady-state coupling calculation.In addition,the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK.For the transient coupling simulation,the operator splitting method with a staggered time mesh was utilized to balance the computational cost and accuracy.Finally,VERA Problem 6 with power and boron perturbation and the NEACRP transient benchmark were simulated for analysis.The numerical results show that the MFNK method can outperform Picard iteration in terms of both efficiency and robustness for a wide range of problems.Furthermore,the reasonable agreement between the simulation results and the reference results for the NEACRP transient benchmark verifies the capability of predicting the behavior of the nuclear reactor.

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.

Anti-noise performance of the pulse coupled neural network applied in discrimination of neutron and gamma-ray

In this study,the anti-noise performance of a pulse-coupled neural network(PCNN)was investigated in the neutron and gamma-ray(n-γ)discrimination field.The experiments were conducted in two groups.In the first group,radiation pulse signals were pre-processed using a Fourier filter to reduce the original noise in the signals,whereas in the second group,the original noise was left untouched to simulate an extremely high-noise scenario.For each part,artificial Gaussian noise with different intensity levels was added to the signals prior to the discrimination process.In the aforementioned conditions,the performance of the PCNN was evaluated and compared with five other commonly used methods of n-γdiscrimination:(1)zero crossing,(2)charge comparison,(3)vector projection,(4)falling edge percentage slope,and(5)frequency gradient analysis.The experimental results showed that the PCNN method significantly outperforms other methods with outstanding FoM-value at all noise levels.Furthermore,the fluctuations in FoM-value of PCNN were significantly better than those obtained via other methods at most noise levels and only slightly worse than those obtained via the charge comparison and zerocrossing methods under extreme noise conditions.Additionally,the changing patterns and fluctuations of the FoMvalue were evaluated under different noise conditions.Hence,based on the results,the parameter selection strategy of the PCNN was presented.In conclusion,the PCNN method is suitable for use in high-noise application scenarios for n-γdiscrimination because of its stability and remarkable discrimination performance.It does not rely on strict parameter settings and can realize satisfactory performance over a wide parameter range.

Core and blanket thermal-hydraulic analysis of a molten salt fast reactor based on coupling of OpenMC and OpenFOAM

In the core of a molten salt fast reactor(MSFR),heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture that acts as both fuel and coolant.Fission energy is released from the fuel salt and transferred to the second loop by fuel salt circulation.Therefore,the MSFR is characterized by strong interaction between the neutronics and the thermal hydraulics.Moreover,recirculation flow occurs,and nuclear heat is accumulated near the fertile blanket,which significantly affects both the flow and the temperature fields in the core.In this work,to further optimize the conceptual geometric design of the MSFR,three geometries of the core and fertile blanket are proposed,and the thermal-hydraulic characteristics,including the three-dimensional flow and temperature fields of the fuel and fertile salts,are simulated and analyzed using a coupling scheme between the open source codes OpenMC and OpenFOAM.The numerical results indicate that a flatter core temperature distribution can be obtained and the hot spot and flow stagnation zones that appear in the upper and lower parts of the core center near the reflector can be eliminated by curving both the top and bottom walls of the core.Moreover,eight cooling loops with a total flow rate of0.0555 m3 s-1 ensur an acceptable temperature distribusure an acceptable temperature distribution in the fertile blanket.

A set of new nucleon coupling constants and the proto neutron star matter

Using a new set of nucleon coupling constants CZll the properties of a proto neutron star are examined within the framework of the relativistic mean-field theory for the baryon octet system. It is found that the relative number density of A,≡ , and ≡0 for CZll are all smaller than those for GL97 and for both CZ11 and GL97, ∑-∑0 and ∑＋ do not appear. It is also found that the pressure and the maximum mass for CZll are all smaller than those for GL97. The maximum mass for CZ11 decreases by approximately 9 percent compared with that for GL97.

Effects of Tensor Couplings on Nucleonic Direct URCA Processes in Neutron Star Matter

The relativistic neutrino emissivity of the nucleonic direct URCA processes in neutron star matter is investigated within the relativistic Hartree-Fock approximation. We particularly study the influences of the tensor couplings of vector mesons ω and ρ on the nucleonic direct URCA processes. It is found that the inclusion of the tensor couplings of vector mesons w and p can slightly increase the maximum mass of neutron stars. In addition, the results indicate that the tensor couplings of vector mesons ω and ρ lead to obvious enhancement of the total neutrino emissivity for the nucleonic direct URCA processes, which must accelerate the cooling rate of the non- superfluid neutron star matter. However, when considering only the tensor coupling of vector meson ρ, the neutrino emissivity for the nucleonic direct URCA processes slightly declines at low densities and significantly increases at high densities. That is, the tensor coupling of vector meson ρ leads to the slow cooling rate of a low-mass neutron star and rapid cooling rate of a massive neutron star.

Phase Transitions in Neutron Stars Within Modified Quark-Meson Coupling Model

K^- condensation and quark deconfinement phase transitions in neutron stars are investigated. We use the modified quark-meson coupling model for hadronic phase and the MIT bag model for quark phase. With the equation of state （EOS） solved self-consistently, we discuss the properties of neutron stars. We find that the EOS of pure hadron matter with condensed K- phase should be ruled out by the redshift for star EX00748-676, while EOS containing unpaired quark matter phase with B1/4 being about 180 MeV could be consistent with both this observation and the best measured mass of star PSR 1913 ＋ 16. But if the recent inferred massive star among Terzan 5 with M 〉 1.68M is confirmed, all the present EOSes with condensed phase and deconfined phase would be ruled out.

小角中子散射原位热力耦合加载装置

热力耦合近工况条件下材料微观结构的原位实验研究,对于深入理解材料服役性能演化机制十分重要,可给出样品微观上的纳米结构尺度分布。为充分发挥小角中子散射统计性好、取样体积大可开展原位实验等优势,本文基于中国先进研究堆小角中子散射谱仪,设计并研制了一台高温和拉力同时加载的原位实验装置,并实现了高温高压下原位测量材料的纳米尺度形貌变化。实验测试结果表明,装置最大载荷可达20 kN,最高温度800℃,控温精度优于±1℃。利用该装置对镍基单晶高温合金样品进行了原位小角中子散射测试,发现温度拉力条件下样品内部纳米结构的明显变化,表明基于该装置可开展热力耦合加载下的原位小角中子散射实验。该装置及其相应实验方法,可用于核电不锈钢等多种高温结构材料的原位加载实验研究,提供微观结构演化数据。

反射中子对金属快中子脉冲堆特性参数的影响研究

快中子脉冲堆对墙壁反射中子比较敏感,反射中子会改变快中子脉冲堆波形,当反射中子较多时可能会对脉冲堆的运行安全造成不利影响。本文建立了考虑墙壁反射中子效应的点堆动力学方法、蒙特卡罗中子学计算方法和ANSYS热力学计算方法三者耦合的“核-热-力”耦合方法,并对含有墙壁反射中子效应的快中子脉冲堆Godiva-Ⅰ瞬态过程进行分析。结果表明:反射中子使脉冲后沿提高,使冲坪时的反应性变低,使堆芯位移、应力有所提高。

板状燃料富集度对燃料性能的影响研究

核燃料的富集度对燃料性能有着重要的影响。为探究燃料富集度对U3Si2-Al板状燃料性能的影响,本研究基于BEEs-Plates/OpenMC/ZEBRA耦合程序,针对15%、20%和25%富集度的组件开展了中子物理-热工水力-燃料性能耦合计算,并分析了组件平均燃耗达125.71 GWd·tU^(-1)时的中子物理参数和燃料性能参数。计算结果表明,燃料组件富集度越高,其体积应变、蠕变应变和塑性应变越大。对中子物理参数的分析结果表明,25%富集度的组件功率密度比15%富集度的组件高出约18%,且富集度越高的组件最大燃耗也越大。对燃料性能参数的分析结果表明,25%富集度的组件塑性应变比15%富集度的组件最大高约40%,说明富集度越高的组件越容易发生失效。

激光惯性约束聚变堆包层能量沉积特性

参考国内外聚变堆技术,建立了一种200 MW激光惯性约束聚变堆包层概念设计,包层采用超临界二氧化碳和锂铅双冷结构。研究构建了瞬态和稳态耦合模型计算包层温度分布及变化。靶丸内爆反应使用MULTI-IFE进行计算,核热耦合部分基于蒙特卡罗程序OpenMC和自编程换热模型对包层模型结构、冷却和产氚进行计算。研究结果表明,核热耦合模型能够完成对包层的初步计算分析,周期性的瞬态载荷会引起第一壁面温度的振荡,但包层内部的温度最终会收敛到稳态计算结果。堆腔尺寸对于降低温度以及震荡效果明显,但仍需氙展平辐射功率峰。包层产氚和能量导出同时受到堆腔尺寸和增殖区的影响,在200 MW工况下,3 m半径和0.25 m增殖区尺寸计算结果最能满足需求。

百千瓦级空间核反应堆屏蔽优化研究

屏蔽体尺寸和重量对空间核反应堆和核动力航天器性能有着重要影响,因而屏蔽设计优化是空间核动力系统设计的关键。本文以JIMO项目反应堆为对象,在铍-碳化硼-钨-氢化锂分层组合屏蔽方案的基础上,考虑到辐照剂量的径向分布,采用蒙特卡罗方法计算了负载处辐照剂量和氢化锂中子剂量,分析了屏蔽设计原理,并提出了分步优化方法以实现屏蔽优化。根据结果分析,调整了铍和碳化硼的厚度比例、钨半径及布置位置,获得了优化的屏蔽方案,在满足屏蔽要求的基础上质量减少了98.41 kg。提出的屏蔽方案及设计流程可为空间核电源屏蔽设计优化提供参考。

摇摆条件下棒束通道自然循环核热耦合特性实验研究

为研究海洋条件下核热耦合对棒束通道自然循环热工参数波动特性的影响,在机械摇摆台上设计并搭建了5×5棒束通道自然循环实验回路。基于单群点堆中子动力学模型,通过实时采集热工参数并计算实时核功率,实现了运动-热工-物理的耦合。对棒束通道单相自然循环核热耦合的瞬态特性进行了实验研究,研究结果显示:在静态工况下,核热耦合使得系统功率发生小幅度波动,当温度反馈系数低于-5×10^(-4)℃^(-1)时,燃料温度反馈对功率的影响比冷却剂温度反馈更大。核热耦合对热工系统的时均参数无显著影响,增大燃料温度反馈系数会降低系统的稳定性。在摇摆运动工况下,摇摆幅度越小或摇摆周期越短,核热耦合引入的功率波动幅度就越小。此外,在摇摆运动启动的瞬态过程中,核热耦合使得系统重新建立稳定自然循环的时间大幅延长。

Preliminary safety analysis for heavy water-moderated molten salt reactor

The heavy water-moderated molten salt reactor(HWMSR)is a newly proposed reactor concept,in which heavy water is adopted as the moderator and molten salt dissolved with fissile and fertile elements is used as the fuel.Issues arising from graphite in traditional molten salt reactors,including the positive temperature coefficient and management of highly radio-active spent graphite waste,can be addressed using the HWMSR.Until now,research on the HWMSR has been centered on the core design and nuclear fuel cycle to explore the viability of the HWMSR and its advantages in fuel utilization.However,the core safety of the HWMSR has not been extensively studied.Therefore,we evaluate typical accidents in a small modular HWMSR,including fuel salt inlet temperature overcooling and overheating accidents,fuel salt inlet flow rate decrease,heavy water inlet temperature overcooling accidents,and heavy water inlet mass flow rate decrease accidents,based on a neutronics and thermal-hydraulics coupled code.The results demonstrated that the core maintained safety during the investigated accidents.

KRUSTY-HEU热管空间堆体积优化及热力耦合特性研究

空间热管堆具有小型化、长寿期、环境适应性强等优点,在航天领域中有广阔的应用前景。本文针对由洛斯阿拉莫斯国家实验室提出的高富集度热管空间堆高浓缩铀方案KRUSTY-HEU,在保持反应性不变的前提下,通过添加慢化剂材料的方法进行体积优化,减小控制棒的体积并降低反射层厚度,用以降低航天器的发射成本和发射载荷。在慢化剂布置方面,本文提出了芯块内层添加慢化剂、芯块内部中间层添加慢化剂和慢化剂与燃料弥散3种方案,并对反应性和安全性进行对比。结果表明,芯块内层添加慢化剂方案优于其他两种方案,最低的反射层厚度为8.69 cm,较之前降低了3.11 cm,体积降低了约30%。针对优化后的方案进行了中子学和热力学分析,结果表明温度对反应性影响较小且移动反射层能够很好地控制反应性,同时燃耗引起的反应性损失较小。在正常运行时,堆芯温度和热位移与之前相差不大,热应力最大处为慢化剂和芯块的交界面,超过了堆芯材料屈服极限。通过在慢化剂和芯块间添加间隙,在材料屈服极限内有效地降低了热应力。单管失效分析表明堆芯温度、位移和热应力发生变化,但未影响其安全性。综上所述,本文提出的芯块内层添加慢化剂方案可有效降低KRUSTY-HEU的体积,热力耦合特性表明优化后的堆芯仍具有较高的安全性和稳定性。

Flow field effect of delayed neutron precursors in liquid-fueled molten salt reactors

In molten salt reactors(MSRs),the liquid fuel salt circulates through the primary loop and a part of the delayed neutron precursors(DNPs)decays outside the reactor core.To model and analyze the flow field effect of DNPs in channel-type liquid-fueled MSRs,a three-dimensional space-time dynamics code,named ThorCORE3D,that couples neutronics,core thermalhydraulics,and a molten salt loop system was developed and validated with the Molten Salt Reactor Experiment(MSRE)benchmarks.The effects of external loop recirculation time,fuel flow rate,and core flow field distribution on the delayed neutron fraction loss of MSRE at steadystate were modeled and simulated using the ThorCORE3D code.Then,the flow field effect of the DNPs on the system responses of the MSRE in the reactivity insertion transient under different initial conditions was analyzed systematically for the channel-type liquid-fueled MSRs.The results indicate that the flow field condition has a significant effect on the steady-state delayed neutron fractions and will further affect the transient power and temperature responses of the reactor system.The analysis results for the effect of the DNP flow field can provide important references for the design optimization and safety analysis of liquid-fueled MSRs.

Influence of Interactions on Populations for Hyperons in Neutron Stars

The numerical results of the populations for the baryon octet in neutron star matter have been presented by solving a set transcendental equations in the framework of the relativistic mean field approximation. The influence of the hyperon interactions on hyperon populations in neutron star matter is discussed. The results manifest that when the ratio of the hyperon-to-nucleon couplings increases, all hyperons appear towards low baryon density direction.

Impact of photoneutrons on reactivity measurements for TMSR-SF1

The solid-fueled thorium molten salt reactor(TMSR-SF1) is a 10 MW_(th) test reactor design to be deployed in 5-10 years by the TMSR group.Its design combines coated particle fuel and molten FLiBe coolant for great intrinsic safety features and economic advantages.Due to a large amount of beryllium in the coolant salt,photoneutrons are produced by(y,n) reaction,hence the increasing fraction of effective delayed neutrons in the core by the photoneutrons originating from the long-lived fission products.Some of the delayed photoneutron groups are of long lifetime,so a direct effect is resulted in the transient process and reactivity measurement.To study the impact of photoneutrons for TMSR-SF1,the effective photoneutron fraction is estimated using k-ratio method and performed by the Monte Carlo code(MCNP5) with ENDF/B-Ⅶ cross sections.Based on the coupled neutronphoton point kinetics equations,influence of the photoneutrons is analyzed.The results show that the impact of photoneutrons is not negligible in reactivity measurement.Without considering photoneutrons in on-line reactivity measurement based on inverse point kinetics can result in overestimation of the positive reactivity and underestimation of the negative reactivity.The photoneutrons also lead to more waiting time for the doubling time measurement.Since the photoneutron precursors take extremely long time to achieve equilibrium,a "steady" power operation may not directly imply a "real" criticality.

From the Dark Neutron to the Neutron Decay Anomaly and Lithium Cosmologic Problem

In the context of the geometric model of particles (PGM), we show two different forms of the structure of the quark positions making up the neutron: first, an ordinary form, while the second is a “dark” form (difficult to detect). By the “dark” form we attempt of explaining the anomaly of the neutron lifetime (τ) in its decay observed in two different experiments as that in “bottle” and “in beam” and expressed by discrepancy between the two lifetimes (τbottle ≠ τbeam). Using the structure equation of the dark neutron, we calculate its mass. In this framework, two problems can be resolved: the asymmetry between matter and antimatter and the abundance into universe of Lithium 7Li than the 6Li.

基于非结构网格输运模型的核-热耦合程序的开发与验证

反应堆内存在着中子物理、流动传热等多种物理场的紧密耦合和相互反馈。为了能准确地模拟反应堆内的真实情况,本研究针对先进复杂反应堆开发了非结构网格核-热耦合程序MORPHY。中子物理求解采用三角形变分节块法方法结合刚性限制法求解时空中子输运方程;热工水力求解基于一维的并联通道模型和圆柱导热模型。采用TWIGL基准题验证了中子动力学的准确性,堆芯相对功率与参考结果的偏差小于0.5%。与Dodds基准题结果对比,验证了程序对于非结构网格的描述能力。基于NEACRP压水堆基准题对程序的核热耦合计算能力进行验证,并分析对比了不同耦合方法、角度离散阶数对结果的影响。结果表明:MORPHY程序计算值与TWIGL、Dodds以及NEACRP基准题参考值吻合良好,能够用于堆芯稳态和瞬态核热耦合分析模拟。