代中子时间法求解点堆中子动力学方程

由于点堆中子动力学方程是个刚性方程,因此准确、快速、稳定地求解方程是困难的。得益于现代计算机技术的进步,本文直接采用代中子时间计算法求解点堆中子动力学方程,并用C++语言编制了计算程序。经过基准例题和动态-逆动态对比计算,验证了模型、程序计算的准确性和稳定性,而计算时间也是可接受的。     

The determination of reactor kinetic parameters in a two-core reactor

The kinetic parameters, α the coupling coefficient and gt the mean neutron transit time have been determined using a reactor oscillator on the coupled-core of the Queen Mary College research reactor. By using correlation techniques it has proved possible to use detectors small enough to be inserted in the fuel tanks. It is shown that the simplified Baldwin model with one-group diffusion theory is inadequate to describe the kinetic behaviour and the experimentally-determined parameters are dependent upon the positioning of the detectors.     

Measuring neutron noise induced by travelling air bubbles in a research reactor

Travelling air-bubble experiments carried out in a research reactor confirm an earlier proposed model. The sink structure could be found experimentally in the APSD of neutron signals and was used to determine the bubble velocity. The measurements show that neutron detectors measure the velocity of the travelling bubbles, the thermocouples that of the water flow.     

The role of factorial cumulants in reactor neutron noise theory

This article intends to specify the physical meaning and the combinatorial implications of the factorial cumulant of a state variable such as the number of neutrons or the number of neutron counts. Features of the presentation are: (a) the fission process is treated in its entirety without the customary binary emission restriction, (b) the introduction of the factorial cumulants helps in reducing the complexity of the mathematical problems, (c) all the solutions can be obtained analytically.In the present article, the authors deal only with the ergodic hypothesis for the neutron population evolution.     

Identification of reactor internals'' vibration modes of a Korean standard PWR using structural modeling and neutron noise analysis

The vibration characteristics of a Korean standard PWR reactor internals have been estimated through a three-dimensional finite element analyses and verified by using the mode separated power spectral density functions obtained from the ex-core neutron noise signals. Also the natural vibration modes of the fuel assembly have been identified measuring both the ex-core and the in-core neutron noise signals which are close to each other. As a result, the fundamental bending mode frequency of the reactor internal structure is found to be around 8 Hz and the fundamental shell mode frequency 14.5 Hz, respectively. It is also shown that the fundamental bending mode frequency of the fuel assembly is 2.3 Hz and the 2^nd bending mode frequency 5.8 Hz, respectively. These results can be used for the supplements of the Korean standard PWR's CVAP (Comprehensive Vibration Assessment Program) data.     

The role of the eigenvalue separation in reactor dynamics and neutron noise theory

The concept of eigenvalue separation (ES) was introduced in the past for the characterisation of the space-time kinetics of reactor transients, and the stability properties of large loosely coupled cores. However, most of the investigations reported so far concern the determination of the ES itself either from static calculations, or from measurements of the flux tilt or neutron noise cross-correlations. Conclusions on system behaviour were only drawn from the properties of the static eigenfunctions, comparing non-perturbed and perturbed systems, without explicitly solving the time- or frequency-dependent problem. In this paper, we explore the role of the ES on the neutronic response of a critical core to small stochastic perturbations (neutron noise); in particular, the spatial and frequency characteristics of the arising neutron noise as a function of the ES, as well as the spatial structure of the perturbation. It is shown that for systems with small ES and non-uniform perturbations, point kinetics will not dominate even for very low frequencies. The results lend some further insight into the origin and properties of the various types of boiling water reactor instabilities.     

Experimental reactor noise—A review on noise-analytic measurements of thermohydraulic parameters in operating BWRs and their interpretations

This article reviews the results and interpretations of measurements inferred from neutron-noise in BWRs with respect to their applicability for Code-Verification, Monitoring and Anomaly-Detection, which are the final goals of neutron-noise investigations in BWRs.It will be shown that for the time-being, none of these goals have been reached and furthermore, in the near future, rigorous code-verification by means of neutron-noise in operating BWRs (measurements of fluid velocity and voidfraction as well as flow-pattern identification), due to the complexity of the system (the measured values are unknown types of averages of the flows in four 8 × 8 bundles) seems to be rather difficult.     

Optimization study of some neutron imaging parameters in a reactor based neutron radiography facility

Neutron imaging technique can be used as a means of material Non-Destructive testing. One of common neutron sources for neutron radiography is nuclear research reactor. In this work, several neutron imaging parameters such as aperture distance and the radiography plane location from the neutron source as well as the aperture diameter have been computationally optimized to deliver a proposed neutron beam. According to the results, the aperture diameter of 3.5–4 cm which was located at 55–85 cm from the outer layer of reactor core and the position of image plane at 300–400 cm fulfills delivering of the suitable neutron flux and other required conditions. W, Fe and Pb walls with an identified length formed the convergent-divergent collimator and shielded the neutron and gamma out of beam path. Bi and Fluental filters with an optimal dimension were used to efficiently improve the neutron beam profile at a sample position.     

反应堆快中子实验装置辐射场参数测量

利用多箔活化法测量了设计的反应堆快中子实验装置的中子能谱及中子注量,并采用Monte Carlo方法分析了能谱的不确定度.用热释光剂量片法测量了装置的γ剂量.装置各参数测量结果均达到了预期的设计指标.     

Validation of reactor noise linear stability methods by means of advanced stochastic differential equation models

The aim of this paper is to show a validation method of a stability monitor using a BWR model with multiple Wiener noise sources, of additive and multiplicative nature. This model is solved using the modern methods to integrate stochastic differential equation systems, that are based on the stochastic Îto-Taylor expansion, and developed by Kloeden and Platen (1995), Kloeden et al. (1994). The synthetic signals generated with this BWR reduced order model with multiple Wiener processes are then used to obtain what are the optimal ways of filtering the signals for the different methods to estimate the decay ration (DR) and the natural frequency (ω) of the system. Also, for each DR estimation method, we study what is the optimal combination of algorithms to obtain the order and coefficients of the AR model that yields the best prediction of the reactor stability parameters for a broad range of DR values.     

Generalized linear model for identifying the kinetic parameters of a nuclear reactor

Methods of real-time identification of the kinetic parameters of a reactor on the basis of an analysis of the linear couplings in transient neutron processes are examined. The algorithms described can be used in various combinations during reactor operation to identify typical transient processed according to their traces.     

Flow of kinetic parameters in a typical swimming pool type research reactor

Calculations were performed to estimate the variation in kinetic parameters (delayed neutron fraction and prompt neutron generation time) in different core configurations of a typical swimming pool type research reactor. Pakistan research Reactor-1 (PARR-1) was employed for this study. The effect due to burnup of the core was also studied. Calculations were performed with the help of computer codes WIMSD/4 and CITATION. Precursors yield was modified according to the neutron flux averaging only. This is the simple way to calculate the precursor yield for a particular core. The kinetic parameters are different for different core configurations. The β_eff decreases with 1.33 × 10⁻⁶/% burnup whereas prompt neutron generation time increases with 6.42 × 10⁻⁸ s/% burnup. The results were compared with safety analysis report and with published values and were found in good agreement. This study provides the confidence to understand the change in the kinetic parameters of research reactors with core change and also with burnup of the core.     

Positron annihilation study and computational modeling of defect production in neutron-irradiated reactor pressure vessel steels

Positron annihilation spectroscopy (PAS) and a computer simulation were used to investigate a defect production in reactor pressure vessel (RPV) steels irradiated by neutrons. The RPV steels were irradiated at 250 °C in a high-flux advanced neutron application reactor. The PAS results showed that mainly single vacancies were created to a great extent as a result of a neutron irradiation. Formation of vacancies in the irradiated materials was also confirmed by a coincidence Doppler broadening measurement. For estimating the concentration of the point defects in the RPV steels, we applied computer simulation methods, including molecular dynamics (MD) simulation and point defect kinetics model calculation. MD simulations of displacement cascades in pure Fe were performed with a 4.7 keV primary knock-on atom to obtain the parameters related to displacement cascades. Then, we employed the point defect kinetics model to calculate the concentration of the point defects. By combining the positron trapping rate from the PAS measurement and the calculated vacancy concentrations, the trapping coefficient for the vacancies in the RPV steels was determined, which was about 0.97 × 10¹⁵ s⁻¹. The application of two techniques, PAS and computer simulation, provided complementary information on radiation-induced defect production.     

Flow measurements by noise analysis of thermocouple signals from the BOCA experiment at the Studsvik R2 reactor

At the Studsvik research reactor R2, a Boiling Capsule (BOCA) is used for long-term irradiation of BWR and PWR fuel rods. The BOCA experiment consists of a pressurised container that can hold a number of fuel rods in a bundle type configuration. The water flow inside the tube is driven by natural circulation. The coolant flow rate is not normally measured in the BOCA rig. Only thermocouples, measuring the water temperature at pertinent locations, are located inside the pressure tube. To confirm calculated values of the flow rate, transit time determination through the cross-correlation technique has been implemented.

Campaigns of noise measurements have been performed at five different occasions. The measurement campaigns have included 10 thermocouples at 3–4 different power levels. The results for the flow rate range between 0.15 and 0.35 m/s depending on reactor power level. The statistical accuracy of the results has also been evaluated. This paper shows that signal processing of thermocouple signals can be used to obtain rather accurate values of the flow rate in BOCA.