对先进核能系统中U-Pu循环的研究

利用新开发的数据库和程序对加速器驱动的次临界反应堆中的Ｕ－Ｐｕ循环进行了详细的理论研究．通过对反应堆功率、临界系数、γ和中子通量以及裂变核的演化等计算,得到了令人满意的中子学结果.The U-Pu cycle in accelerator driven subcritical reactor is studied by means of new data library and code. The satisfactory neutronics results are obtained by calculating the reactor power, critical value, gamma and neutron flux and evolution of fissioning nuclear density. The detailed analysis is also presented.     

启明星Ⅱ次临界非稳态热中子注量率测量

中子注量率及分布是反应堆的重要参数,本工作通过核数守恒在非稳态情况下的推导和求解,从理论上论证了次临界反应堆非稳态情况下中子注量率测量的可行性。将活化法和固体径迹法有机结合,利用固体径迹探测器标定活化片的测量数据,测量了启明星Ⅱ零功率装置的He-3管实验孔道内及反应堆外壁的中子注量率的分布,并与模拟计算结果进行了比较,利用MCNPX程序得到的模拟计算结果与实验结果的趋势一致,证明了该测量方法可以测量低通量的中子注量率,可实现反应堆不同时刻、不同位置的中子注量率测量,为CiADS技术的研发提供了实验数据与技术支撑。Neutron flux measurements were carried out at VENUS-Ⅱ lead-based zero power reactor by neutron activation method combined with solid-state nuclear track detectors (SSNTD). This experimental method was proposed based on the principle of nuclear number conservation when a foil was irradiated in an unsteady-state neutron field. By this method, thermal neutron flux distributions inside the He-3 duct were measured when VENUS-Ⅱ was operated under unsteady-state. The neutron flux distributions were also calculated with MCNPX code and were consistent with the experimental data. In addition, the neutron fluxes in the outer layer of VENUSⅡ were measured under steady-state. These results would benefit the further study of experimental methods for neutron flux measurement and provide important support for the design of CiADS.     

时空中子倍增公式解法研究

通过将空间概念引入中子倍增公式, 建立了时空中子倍增公式, 针对时空中子倍增公式计算时间效率过低的缺点, 对时空倍增公式的计算方法进行了研究, 新解法在保证精确性的基础上, 又提高了计算效率。新的计算方法可以应用于研究反应堆反应性发生动态变化的情况, 例如反应堆启堆或停堆以及变功率运行的过程, 对具有很强机动性要求的小型反应堆的安全分析有一定的理论意义和实用价值。 The space time neutron multiplication formula was established by introducing the concept of space time to the neutron multiplication formula. Because of the rather long computing time by using the formula, the solution of space time neutron multiplication formula is optimized which is efficient and accurate. The new method can be used to study the reactor with dynamic change in reactivity. It can be also used for studying the process of startup, outage and power drawn operation, which is of great importance to the safety analysis of the mini nuclear reactor.     

中子注量率波动修正因子及其应用

在中子核反应研究中,尤其是在利用活化法进行中子核反应截面测量研究时,需要准确测量样品辐照的中子注量。监督反应标准截面法简便可行,在一些核反应截面测量研究中经常用来定量样品辐照的中子注量。在利用监督片核反应剩余核的放射性活度计算平均中子注量率时,中子注量率波动修正因子是一个很重要的参数。对中子注量率波动修正因子进行了详细阐述,通过理论推导给出了中子注量率波动修正因子的定义,从实际应用出发讨论了中子注量率波动修正因子的使用条件和监督反应的选择原则。Incident neutron flux has to be measured accurately in the neutron reaction study especially in the neutron reaction cross-section measurement with activation method. Average neutron flux in the irradiated sample is usually determined by the monitor reaction with reference cross-section values. However, the average incident neutron flux, based on the radioactivity of the residual nuclei produced in the monitor reaction, is dependent upon the neutron flux fluctuation. In the procedure of the average neutron flux calculation, the correction factor for the neutron flux fluctuation plays a key role. In this paper, definition of the neutron flux fluctuation correction factor is inferred heoretically. The selection principles of the monitor reaction and the utilization of the correction factor have been discussed.     

用于中子测量的俘获探测器研究

为了在弱中子场和有限小空间内测量绝对裂变率, 制作了俘获探测器, 研究了俘获探测器的性能。 介绍了用于中子测量的俘获探测器和铅屏蔽室, 以及该探测器系统在特定条件下测量裂变反应率的结果, 并与裂变室测量结果进行了比较。 探讨了铅屏蔽室大小对测量结果的影响。 To detect the absolute neutron flux in a weak neutron field and restricted space, the fission fragment trapping detector was fabricated and the properties of the detector were studied. In this paper, the detector and shielding chamber used in neutron detection were described and the experimental measurements of the fission rate in specific condition were performed with the detection system and the result has been compared with that obtained by fission chamber. The influence of the shielding chamber on the measured results was analyzed.     

体积填充率变化对球床钨颗粒散裂靶中子学特性影响的研究

应用MCNPX程序,构建质子束轰击球床钨颗粒散裂靶的物理模型,模拟散裂靶泄露中子产额、能谱、通量轴向分布以及散裂靶沉积能分布。针对不同钨颗粒直径和体积填充率,研究了不同钨颗粒直径下体积填充率变化对球床散裂靶中子学特性的影响。模拟结果表明,钨颗粒体积填充率增加,散裂靶的最大沉积能密度减小。在1～20 mm的范围内,钨颗粒的直径越小,散裂靶泄漏中子产额越大,散裂靶泄漏中子产额随钨颗粒体积填充率变化的波动越小,有利于维持CIADS系统反应堆功率稳定。The physical model of the high-energy proton bombarding the tungsten pebble bed spallation target is simulated by the MCNPX program. The effect of the filling rate on the neutronic characteristics with different particle diameters is studied, by calculating the leakage neutron yield, leakage neutron spectrum axial neutron flux distribution and the energy deposition of the target. The result shows that when the diameter increases from 1 to 20 mm,the maximum deposited energy density decreases in the target, but the leakage neutron yield increases. When the filling rate reaches 74%, leakage neutron yields are almost the same value with different particle diameters. When the target is piled up with 1 mm tungsten particles, neutron leakage yield changes smaller with the variation of the filling rate than the other diameter particles which is beneficial to maintain the reactor power stability in ADS.     

基于D-T紧凑型中子源的快中子照相准直屏蔽体设计及中子束特性模拟研究

中子照相是一种重要的无损检测技术,它能用于火工产品、毒品和核燃料元件等的检测。基于紧凑型D-T中子发生器,完成了一个用于快中子照相的准直屏蔽体系统（BSA）的物理设计。根据D-T中子源的能谱和角分布建立了中子源模型,采用MCNP4C蒙特卡罗程序,模拟了准直屏蔽体系统中中子和γ射线的输运,准直中子束相对于单位源中子的中子注量可以达到9.30×10-6 cm-2,准直中子束中主要是能量大于10 MeV的快中子;在设置的样品平面直径14 cm的照射视野范围,准直束中子注量的不均匀度为4.30%,准直束中中子注量与γ注量的比值为17.20,中子通量和中子注量比值J/Φ为0.992,说明准直中子束有好的平行性;准直屏蔽体外的泄露中子注量率与准直束中子注量率相比降低了2个量级。所设计的准直屏蔽体能满足快中子照相的要求。Neutron radiography is an important nondestructive testing technique. It can be used to detect the explosive devices, drug and the nuclear fuel element, etc. A beam-shaping-assembly (BSA) based on a compact D-T neutron generator is designed for fast neutron radiography in this paper. D-T neutron source model is constructed based on the neutron energy spectrum and angular distribution data. The transportation of neutron and γ-ray in the BSA is simulated using MCNP4C code. The neutron fluence of the collimated neutron beam with respect to the neutron source of the unit source is 9.30×10-6 cm-2. The collimated neutron beams is mainly fast neutrons with energies greater than 10 MeV. In the irradiation field range with a diameter of 14 cm, the neutron fluence uniformity of the collimated beam is 4.3%, the ratio of the neutron fluence to the gamma fluence in the collimated beam is 17.20, and the neutron flux and the neutron fluence ratio (J/Φ) is 0.992 which indicates that the collimated neutron beam has good parallelism. The leakage neutron fluence in outside of BSA is two orders of magnitude lower than that of the collimated neutron beam. The designed BSA can meet the need of fast neutron radiography.     

Real-time surface figure monitoring of optical elements in continuous polishing

A real-time monitoring system is set up based on a computer, dynamic interferometer, beam expanding system,and a beam reflecting system. The stability and repeatability of the monitoring system is verified. A workpiece and a glass monitoring plate are placed in the same ring. The surface figure of the workpiece, monitored by the monitoring plate, synchronizes with the surface of the glass monitoring plate in terms of peak–valley and power.The influence of the reflection and transmission surface are discussed in theory and a numeral deviation in online and offline testing data is quantitatively analyzed. The new method provides a quick and easy real-time method to characterize changes to the optical surface during polishing.     

Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D~+ beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the threedimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S (α,β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors.     

中子源有源法核查技术研究

用有源 (主动 )方法研究了贫化铀组合系统的中子诱发裂变缓发中子探测技术 .在不同屏蔽和组合等条件下测量和比较了贫化铀系统的缓发裂变中子分布 ,进一步研究了实验系统的可核查性 .探讨了区分核与非核系统的方法. The technique for detecting the delayed neutrons from neutron induced fission in uranium systems was studied by using an active method with 3He proportional counting tube array and a 14 MeV D T neutron source. Under the conditions of different shielding and combination, the distributions of delayed fission neutrons from depleted uranium systems were measured and the reliability of the systems was studied. The method to distinguish a nuclear system from a non nuclear one was discussed.     

Monitoring method for neutron flux for a spallation target in an accelerator driven sub-critical system

In this paper, we study a monitoring method for neutron flux for the spallation target used in an accelerator driven sub-critical(ADS) system, where a spallation target located vertically at the centre of a sub-critical core is bombarded vertically by high-energy protons from an accelerator. First, by considering the characteristics in the spatial variation of neutron flux from the spallation target, we propose a multi-point measurement technique,i.e. the spallation neutron flux should be measured at multiple vertical locations. To explain why the flux should be measured at multiple locations, we have studied neutron production from a tungsten target bombarded by a 250 Me V-proton beam with Geant 4-based Monte Carlo simulations. The simulation results indicate that the neutron flux at the central location is up to three orders of magnitude higher than the flux at lower locations. Secondly, we have developed an effective technique in order to measure the spallation neutron flux with a fission chamber(FC), by establishing the relation between the fission rate measured by FC and the spallation neutron flux. Since this relation is linear for a FC, a constant calibration factor is used to derive the neutron flux from the measured fission rate. This calibration factor can be extracted from the energy spectra of spallation neutrons. Finally, we have evaluated the proposed calibration method for a FC in the environment of an ADS system. The results indicate that the proposed method functions very well.     

基于脉冲中子源法的CiADS反应性测量蒙卡模拟分析

脉冲中子源法(PNS)是加速器驱动次临界系统反应性测量的一种重要技术.利用蒙卡软件建立CiADS次临界反应堆模型,模拟注入脉冲质子束的过程,获得的中子通量的时间演化谱.采用Python语言编程实现脉冲叠加过程,给出稳定缓发中子本底,实现脉冲中子源法的次临界反应堆的反应性测量模拟,给出脉冲周期注入下堆芯不同位置处的中子变...     

加速器驱动次临界系统散裂靶参数优化计算

采用国际开源程序包Geant4,构建高能质子束轰击加速器驱动次临界系统(ADS)散裂靶的物理模型,模拟计算质子轰击液态金属铅、铅-铋合金和汞靶的泄漏中子谱分布,以及计算不同能量质子对应的铅靶泄漏中子产额和轴向积分分布,获得1 Ge V质子对应的铅圆柱靶优化参数,考虑入射质子的利用率和整个堆芯的体积质量,优化靶半径范围为16～24 cm,靶高为100 cm,相关研究结果可为(ADS)散裂靶的物理和工程设计提供理论依据。     

Calibration of ITER Instant Power Neutron Monitors: Recommended Scenario of Experiments at the Reactor

Instant power is a key parameter of the ITER. Its monitoring with an accuracy of a few percent is an urgent and challenging aspect of neutron diagnostics. In a series of works published in Problems of Atomic Science and Technology, Series: Thermonuclear Fusion under a common title, the step-by-step neutronics analysis was given to substantiate a calibration technique for the DT and DD modes of the ITER. A Gauss quadrature scheme, optimal for processing “expensive” experiments, is used for numerical integration of ²³⁵U and ²³⁸U detector responses to the point sources of 14-MeV neutrons. This approach allows controlling the integration accuracy in relation to the number of coordinate mesh points and thus minimizing the number of irradiations at the given uncertainty of the full monitor response. In the previous works, responses of the divertor and blanket monitors to the isotropic point sources of DT and DD neutrons in the plasma profile and to the models of real sources were calculated within the ITER model using the MCNP code. The neutronics analyses have allowed formulating the basic principles of calibration that are optimal for having the maximum accuracy at the minimum duration of in situ experiments at the reactor. In this work, scenarios of the preliminary and basic experimental ITER runs are suggested on the basis of those principles. It is proposed to calibrate the monitors only with DT neutrons and use correction factors to the DT mode calibration for the DD mode. It is reasonable to perform full calibration only with ²³⁵U chambers and calibrate ²³⁸U chambers by responses of the ²³⁵U chambers during reactor operation (cross-calibration). The divertor monitor can be calibrated using both direct measurement of responses at the Gauss positions of a point source and simplified techniques based on the concepts of equivalent ring sources and inverse response distributions, which will considerably reduce the amount of measurements. It is shown that the monitor based on the average responses of the horizontal and vertical neutron chambers remains spatially stable as the source moves and can be used in addition to the staff monitor at neutron fluxes in the detectors four orders of magnitude lower than on the first wall, where staff detectors are located. Owing to low background, detectors of neutron chambers do not need calibration in the reactor because it is actually determination of the absolute detector efficiency for 14-MeV neutrons, which is a routine out-of-reactor procedure.     

Measurement of the neutron spectrum of a Pu-C source with a liquid scintillator

The neutron response function for a BC501A liquid scintillator (LS) has been measured using a series of monoenergetic neutrons produced by the p-T reaction. The proton energies were chosen such as to produce neutrons in the energy range of 1 to 20 MeV. The principles of the technique of unfolding a neutron energy spectrum by using the measured neutron response function and the measured Pulse Height (PH) spectrum is briefly described. The PH spectrum of neutrons from the Pu-C source, which will be used for the calibration of the reactor antineutrino detectors for the Daya Bay neutrino experiment, was measured and analyzed to get the neutron energy spectrum. Simultaneously the neutron energy spectrum of an Am-Be source was measured and compared with other measurements as a check of the result for the Pu-C source. Finally, an error analysis and a discussion of the results are given.     

Utilization of the BARC critical facility for ADS related experiments

The paper discusses the basic design of the critical facility, whose main purpose is the physics validation of AHWR. Apart from moderator level control, the facility will have shutdown systems based on shutoff rods and multiple ranges of neutron detection systems. In addition, it will have a flux mapping system based on 25 fission chambers, distributed in the core. We are planning to use this reactor for experiments with a suitable source to simulate an ADS system. Any desired sub-criticality can be achieved by adjusting the moderator level. Apart from perfecting our experimental techniques, in simple configurations, we intend to study the one-way coupled core in this facility. Preliminary calculations, employing a Monte Carlo code TRIPOLI, are presented.      

Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.     

ADS瞬态分析中的概率论-确定论耦合方法

针对加速器驱动次临界系统(ADS)瞬态问题,采用预估校正改进准静态方法(PCQS)处理时空中子动力学方程中的时间自变量,采用蒙特卡罗方法处理相应的空间-角度-能量自变量,重点解决了低次临界度下模拟计算不稳定的问题,验证了TWGIL-Seed-Blanket动力学基准问题和小型模拟ADS问题,得到瞬态过程的功率变化结果,与基于其他方法的程序比较,经初步验证取得了较好结果,证明了该耦合方法可行。     

Prospects for transmutation of nuclear waste and associated proton-accelerator technology

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, a major concern or objection is the long-term burden that is constituted by the radiotoxic waste from the spent fuel. The concept of Partitioning & Transmutation, a scientific and technological answer, is therefore of high interest. Its deployment may use dedicated “Transmuter” or “Burner” reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a solution of high interest. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at an introduction into the field in order to focus, in its later part, on the development of the required accelerator technology.     

用于洁净核能源的ADS-EA方案

加速器驱动系统 (ADS)是一种新型的洁净能源装置 ,它采用加速器提供的强流高能量质子束驱动次临界的核反应堆 ,既有安全可靠和产生核废料少的优点 ,还可以处理传统反应堆留下的核废料 .能量放大器是一个基于回旋加速器组合的 ADS方案 ,由三级回旋加速器组成的加速器系统可以产生流强为12 m A和能量为1 .2 Ge V的质子束 (束流功率1 4.4MW) ,用以驱动1500 MW的核反应堆. Accelerator Driving System (ADS) is a new device for cleaning energy. A high intensity, high power proton beam provided by accelerators is used to drive a sub critical nuclear reactor. It is safe, reliable and can produce less nuclear waste, and also can be used to treat the nuclear waste from the classical reactor. An Energy Amplifier (EA), which is composed of three cyclotrons, is one type of ADS. It will be used to produce 14.4 MW proton beam (12 mA, 1.2 GeV) and to operate a 1 500 MW nuclear reactor.