Application of reactor noise analysis in the candu reactors of Ontario hydro

Reactor noise analysis techniques are being applied in Ontario Hydro's CANDU nuclear generating stations to monitor the dynamic characteristics of critical plant components and processes. A comprehensive analysis of stationary signal fluctuations (noise) of the standard instrumentation of Pickering-B, Bruce-B and Darlington units has been carried out in the past two years. In these measurements the feasibility of applying noise analysis techniques to actual operating data has been demonstrated. The results indicated that the detection and characterization of instrument and process failures, and validation of process signals and instrument functionality can be based on the existence of certain statistical signatures derived from the measured reactor noise signals.     

Propagation noise calculations in VVER-type reactor core

Neutron noise induced by propagating disturbances in VVER-type reactor core is addressed in this paper. The spatial discretization of the governing equations is based on the box-scheme finite difference method for triangular-z geometry. Using the derived equations, a 3-D 2-group neutron noise simulator (called TRIDYN-3) is developed for hexagonal-structured reactor core, by which the discrete form of both the forward and adjoint reactor dynamic transfer functions (in the frequency domain) can be calculated. In addition, both types of noise sources, namely point-like and traveling perturbations, can be modeled by TRIDYN-3. The results are then benchmarked in different cases. Considering the noise source as propagating perturbations of the macroscopic absorption cross sections, the induced neutron noise is calculated throughout the reactor core. For the first time, adjoint approach is applied and examined for modeling moving noise sources. Moreover, the space- and frequency-dependence of the propagation noise are investigated in this paper.     

Two-group theory of neutron noise in Molten Salt Reactors

In a previous paper, the kinetics, dynamics and the neutron noise in a Molten Salt Reactor (MSR) was investigated in a simple homogeneous reactor model in one-group diffusion theory. In this paper those investigations are extended to two-group theory in the same reactor model. In addition, unlike in the previous paper where in the quantitative work data from a thermal LWR were used, here material data of a conceptual MSR with a thermal spectrum and thorium fuel are used, along with data from both fast and thermal LWRs. Among other things, the relative weight and the range of the local component is investigated. It is found that the strong neutronic coupling in an MSR, which was pointed out in the preceding paper, diminishes the role of the local component as compared to light water reactors. Some further new features of the noise in MSR, not directly related to the two-group approach, are also found.     

单光子分辨时间谱仪信噪比的研究

研究了单光子分辨时间谱仪的信噪比和影响信噪比的因素,探讨了降低本底干扰、提高探测效和信噪比的方法。结果表明,信噪比与探测器对源张的立体角和探测效率成正比,与电子学噪声成反比,合理选择辐射源的活度对减小偶然符合计数是重要的,加符合门和单光子鉴别器为提高测量精度和探测效率提供了一种有效途径。     

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.     

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.     

Study of the MTC estimation by noise analysis in 2-D heterogeneous systems

The effect of a heterogeneous distribution of the temperature noise on the MTC estimation by noise analysis is investigated. This investigation relies on 2-group diffusion theory, and all the calculations are performed in a 2-D realistic heterogeneous core. It is shown, similarly to the 1-D case, that the main reason of the MTC underestimation by noise analysis compared to its design-predicted value lies with the fact that the temperature noise might not be homogeneous in the core, and therefore using the local temperature noise in the MTC noise estimation gives erroneous results. A new MTC estimator, which was previously proposed for 1-D 1-group homogeneous cases and which is able to take this heterogeneity into account, was extended to 2-D 2-group heterogeneous cases. It was proven that this new estimator is always able to give a correct MTC estimation with an accuracy of 3%. This small discrepancy comes from the fact that the reactor does not behave in a point-kinetic way, contrary to the assumptions used in the noise estimators. This discrepancy is however quite small.     

On the diffusion coefficient calculation in two-step light water reactor core analysis

This paper presents consistent and rigorous accuracy assessments of various methods for calculating the diffusion coefficients in a two-step reactor core analysis of light water reactors (LWRs). The diffusion coefficients are significantly affected by the transport correction and critical spectrum calculations. There are various methods for the transport corrections (inflow/outflow/hybrid corrections) and critical spectrum calculations (B₁/P₁/CASMO-4E methods) so that it is necessary to decide the best combination to achieve a high accuracy in the transport/diffusion two-step analysis. Numerical tests are performed step-by-step to search for the best combination of the methods by comparing each other the transport one-step results, transport/diffusion two-step results, and Monte Carlo results. Numerical test results with a large and a small LWR core show that the combination of inflow transport correction and CASMO-4E critical spectrum calculation is most accurate than the other combinations in terms of eigenvalues and assembly power distributions.     

Reactor noise measurements in the safety and regulating systems of CANDU stations

Reactor noise measurements of safety and regulating system intrumentation are performed in the CANDU nuclear power stations of Ontario Power Generation (OPG) and Bruce Power. Station signals included in the noise measurements are in-core flux detectors (ICFD), ion chambers (I/C), flow transmitters, pressure transmitters, and resistance temperature detectors (RTD). Their frequency dependent noise signatures are regularly measured during steady-state operation, and are used for parameter estimation and anomaly detection.

The specific applications include the following areas:

Flux noise measurements to detect and characterize (a) anomalies of in-core flux detectors, ion chambers and their electronics, (b) mechanical vibration of fuel channels and in-core detector tubes induced by coolant/moderator flow.

Pressure and flow noise measurements to estimate the in-situ response times of flow/pressure transmitters and their sensing lines installed in the reactor's coolant loops.

Temperature noise measurements to estimate the in-situ response times of thermal-well or strap-on type RTDs installed in the reactor's coolant and moderator loops.

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.     

反应堆脉冲中子和γ射线辐照SiGe HBT器件的退火因子及典型直流参数测量分析

本工作测量了反应堆脉冲中子、γ辐照SiGe HBT典型直流电参数和退火因子.在反应堆1×1013 cm2的脉冲中子注量和257 Gy(Si)γ总剂量辐照后,SiGe HBT静态共射极直流增益减小了20%.辐照后基极电流、结漏电流增大,集电极电流、击穿电压减小.初步分析了SiGe HBT瞬态中子、γ辐射损伤机理.     

Analysis of an MTC noise measurement performed in ringhals-2 using gamma-thermometers and in-core neutron detectors

A noise measurement in the Swedish Ringhals-2 PWR was performed in January 2002 by using twelve gamma-thermometers and two in-core neutron detectors, all located on the same axial level in the reactor. The gamma-thermometers are very versatile tools since they allow estimating the core-averaged moderator temperature noise throughout the core. This core-averaged temperature noise was then used to estimate the MTC by noise analysis, via a new MTC noise estimator. It was shown that whatever the location of the neutron detector might be, the MTC is always correctly estimated by this new MTC noise estimator, without any calibration to a known value of the MTC prior to the noise measurement. For the purpose of comparisons, the MTC was also estimated by using a single gamma-thermomemeter and a single core-exit thermocouple, together with an in-core neutron detector. In such cases, the MTC was systematically underestimated, with a stronger bias for the core-exit thermocouple than for the gamma-thermometer. This shows that the main reason of the MTC underestimation by noise analysis in all the experimental work until now was due to the radially non-homogeneous temperature noise throughout the core. The resulting deviation from point-kinetics of the reactor response has a negligible effect.     

铀部件质量属性测量中的信噪比初步分析

利用蒙特卡罗程序对252Cf快电离室诱发不同质量金属铀部件的中子输运过程进行了模拟,获得屏蔽与直穿两种实验布局条件下252Cf快电离室与探测器之间的中子时间关联符合计数,初步评估测量过程中的信噪比,作为实验前端布局的重要依据。结果表明,两种布局下的信噪比均与铀部件质量成正比,直穿布局下的信噪比较高。     

The study of aeroball system for measuring 3D neutron flux distribution in reactor core

Aeroball system is attractive in several aspects because it can easily transport the map of neutron flux distribution to be measured from incore to outside of a reactor vessel.However,before the aeroball system is put to practical use in the heating reactor.there are four topics that have to be further studied.They are the stability of the activated positions,enhancement of signal/noise(S/N)ratio,distributed control and data-acquisition system and on-lin nbeutron flux distribution reconstruction.Besides describing the rasons for them,this paper gives out the theory,concept and solution about the first two topics and it is helptul to give the possibility to enhance the reactor-power.     

Safety analysis of the IAEA reference research reactor during loss of flow accident using the code MERSAT

Using the thermal hydraulic code MERSAT detailed model including primary and secondary loop was developed for the IAEA's reference research reactor MTR 10 MW. The developed model enables the simulation of expected neutronic and thermal hydraulic phenomena during normal operation, reactivity and loss of flow accidents.Two different loss of flow accident (LOFA) have been simulated using slow and fast decrease time of core mass flow. In both cases the expected flow reversal from downward forced to upward natural circulation has been successfully simulated. The results indicate that in both accidents the limit of onset of subcooled boiling was not arrived and consequently no exceed of design limits in term of thermal hydraulic instability or DNB is observed. Finally, the simulation results show good agreement with previous international benchmark analyses accomplished with other qualified channel and thermal hydraulic system codes.     

A new navigation strategy for underwater robot in reactor pool combined propeller speed detection and dynamics analysis with sonar data correction

The navigation reliability of remote-operated vehicle (ROV) in nuclear power plant is very important. It is necessary to carry out redundant navigation design using components with stronger anti-radiation performance compared with traditional navigation devices, to ensure relatively high precision navigation in the case of failure of traditional navigation equipment. The paper proposes a new navigation algorithm for ROV working in the reactor pool combined propellers’ rotating speed detection with hydrodynamics analysis and sonar data correction without using other sensors, just as a redundant navigation strategy. The designed ROV can work at any water depth thanks to the high-precision depth control algorithm. At first, the hydrodynamics analysis of ROV is completed. After that, the dead reckoning (DR) is implemented combined with rotating speed detection of high-performance propellers with dynamics analysis of ROV. Then, the error of DR is analyzed and compensated through repeated field experiments. At last, the sonar data are added to correct the deduced position of ROV when it stops. Field experiment verifies that the precision of the proposed algorithm is high enough to be used as a redundant navigation strategy for ROV working in the reactor pool of the nuclear power plant.     

Design and application of MTR fuel assemblies in new proposed inverted mode

MTR fuel is one of the most used fuel types in research reactors. Loading high positive excess reactivity enables a research reactor to be in service for an extended period of time without refueling. This in turn ends up to asymmetric burn up distribution in axial direction. Minor modifications in present standard fuel assembly design allow use of each fuel assembly in dual directions along vertical axis. Thus, a smoother axial burn up distribution can be achieved. Detailed calculations showed a real improvement in neutron economy and extra positive reactivity is gained.     

Safety analysis of beyond design basis accidents in RBMK-1500 reactors

At present the design basis accidents for RBMK-1500 are rather thoroughly investigated. The performed analyses helped to develop and implement a number of safety modifications. Further plant safety enhancement requires developing emergency procedures that would enable beyond design basis accidents management by preventing core damage or mitigating consequences of severe accidents.