Boiling-sodium tests with the BOR-60 reactor

Conclusions These first experiments on the BOR-60 reactor have shown that in principle it is possible to detect the boiling of sodium from the acoustic and neutron fluctuations; useful information has been obtained on the character of the signals and the scope for using various processing methods. However, further measurements and calculations are needed before we can design a reliable real-time system for monitoring for sodium boiling in the core of a fast reactor.USSR, GDR. Translated from Atomnaya Énergiya, Vol. 45, No. 5, pp. 338–342, November, 1978.     

The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapor bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapor bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low (<50°C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations, flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary.     

Progress for on-line acoustic emission monitoring of cracks in reactor systems

This paper reviews accomplishments and planned tasks for the NRC-sponsored research program concerned with “Acoustic Emission/Flaw Relationships for Inservice Monitoring of Nuclear Reactor Pressure Boundaries”. The objective of the acoustic emission (AE) monitoring program is to develop and validate the use of AE methods for continuous surveillance of reactor pressure boundaries to detect flaw growth. Topics discussed include testing AE monitoring on reactors, refinement of an AE signal identification relationship, study of slow crack growth rate effects on AE generation, and activity to produce an ASTM standard for AE monitoring and to gain ASME code acceptance of AE monitoring.     

Results of acoustic and temperature noise measurements on sodium cooled electrically heated 7 rod bundle

Assembly cooling deficiency in a LMFBR is one of the most important safety problems for reactor design and operation.

Studies on early detection and diagnosis of local accident by means of noise analysis techniques have been initiated at CNEN. Acoustic and temperature noise measurements have been carried out on a 7 rod bundle during slow power transients up to boiling conditions. The test section, simulating the italian PEC reactor fuel element, was mounted on ENA-2 sodium loop located at the CSN Casaccia.

Acoustic noise spectral analysis up to 32 kHz shows the appearance, in presence of boiling, of power increase at certain frequencies. Power spectra and rms values are updated and recorded every 0.3 sec and show large variations going from single phase to boiling.

Temperature noise spectral analysis shows that the power, between 1 and 50 Hz, increases, in presence of boiling, by a factor bigger than 30. It has been tested the sensitivity of other indicators of the temperature fluctuations, like skewness and flatness, to reveal boiling.     

Potential U.S. contributions to in-reactor experiments for fast reactor surveillance systems

It is maintained that special features of FFTF make it an ideal system to test sodium boiling detection techniques by acoustic/neutronic methods and to test the response of acoustic/neutronic sensors to vibrations. It is shown that accumulated research results indicate that such tests in FFTF are feasible, predictable, promising and safe.     

Acoustic detection of boiling in LMFBRs: An estimate of sensitivity derived from experiments during the commissioning of PFR

This paper describes a programme of experiments undertaken during the commissioning of the 600-MW PFR at Dounreay. The objective of the programme was to obtain data on reactor background acoustic noise levels and to estimate the sensitivity of the installed acoustic boiling detection system.

The boiling-noise detection system installed in PFR consists of seven waveguides. The waveguides are solid metal rods which extend from the reactor top to just above the core top level. Five are arranged on a 0.75m radius circle with two at the edge of the core. The sensing elements are standard accelerometers.

For the commissioning experiment three rigs were installed in the core locations below three of the waveguides. Each rig was over 12m long and carried three sodium-proof microphones at levels corresponding to core top, mid-core and core bottom. Each rig also carried an electrically-powered vapour generator which could inject a stream of sodium vapour into the surrounding 250°C sodium, so producing a boiling type noise. In addition to these nine microphones in the core, microphones were also installed at the inlet and outlet of each of the three pumps. To handle the data for this experiment a system of amplifiers and three fourteen-track tape recorders was used. To check amplifier gains a calibration signal was automatically injected at the end of each record. On line one-third octave and narrow-band analysis was also available, but the bulk of the analysis was done from the tape records.

The commissioning programme was in two parts: a measurement of background noise and an investigation of the detection sensitivity at the waveguides of the noise from the vapour generator (this implicitly included the transmission loss). In the background measurement programme the noise level at each microphone was recorded at pump speeds from 200 to 960 rev/min, full speed, in approximately 10% steps. To determine the sensitivity, the signal at each transducer was recorded with each vapour generator operating in turn. The power input to the vapour generators was 1100W, of which some 800W required to balance the heat losses. The power available for boiling consequently was only 300W; nevertheless the signal was detectable at pump speeds up to 750 rev/min on waveguides radially 0.75m from the source.

Since the noise output of the vapour generator was measured by a microphone close to the source, this result could be used, with data obtained on background noise and measurements of sodium boiling noise obtained on force convection rig experiments, to deduce that a highly sub-cooled boiling source involving 40kW of thermal energy would be detectable in PFR with a margin of 10dB. The conclusion rests on the assumption that there is no major adverse change in the transmission losses at the time detection is required, such as might be caused by a large quantity of free gas in the sodium.     

Sabena: Subassembly boiling evolution numerical analysis

This paper describes the computer code SABENA that has been used in subassembly sodium boiling evolution numerical analysis as a contribution to fast breeder reactor safety analysis. SABENA is a two-fluid model subchannel code system to calculate coolant boiling and two-phase flow in a rod bundle together with external loop characteristics which affects the overall boiling behavior in the bundle section. With the use of relatively simple but reasonable constitutive models, the SABENA code has been applied to and validated against many multi-pin sodium boiling problems. The results have shown excellent agreement with the experiments. The numerical methods and models employed in the code have proven to be robust and efficient in light of the extreme severity of the conditions characterizing low-pressure sodium boiling.     

钠沸腾噪声探测研究

文章叙述了钠沸腾噪声探测研究进展,建立了离线和在线均可进行的高频和低频信号采集和处理系统,引进、开展、改进和编制了信号处理、故障诊断、事故报警和自回归模型分析等软件包。应用这些硬软件对水和钠沸腾噪声进行了探测和分析。结果表明,沸腾噪声信号的自功率谱密度(APSD)的幅值明显大于沸腾时的值,用自回归模型判别因子分析,可实现钠沸腾在线实时诊断和监护。     

Steam leak detection in advance reactors via acoustics method

Prediction of LOCA (loss of coolant activity) plays very important role in safety of nuclear reactor. Coolant is responsible for heat transfer from fuel bundles. Loss of coolant is an accidental situation which requires immediate shut down of reactor. Fall in system pressure during LOCA is the trip parameter used for initiating automatic reactor shut down. However, in primary heat transport system operating in two phase regimes, detection of small break LOCA is not simple. Due to very slow leak rates, time for the fall of pressure is significantly slow. From reactor safety point of view, it is extremely important to find reliable and effective alternative for detecting slow pressure drop in case of small break LOCA. One such technique is the acoustic signal caused by LOCA in small breaks. In boiling water reactors whose primary heat transport is to be driven by natural circulation, small break LOCA detection is important. For prompt action on post small break LOCA, steam leak detection system is developed to detect any leak inside the reactor vault. The detection technique is reliable and plays a very important role in ensuring safety of the reactor. Methodology developed for steam leak detection is discussed in present paper. The methods to locate the leak is also developed and discussed in present paper which is based on analysis of the signal.     

The acoustic environment in large HTGRS

Well-known techniques for estimating acoustic vibration of structures have been applied to a General Atomic high-temperature gas-cooled reactor (HTGR) design. It is shown that one must evaluate internal loss factors for both fluid and structure modes, as well as radiation loss factors, to avoid large errors in estimated structural response. At any frequency above 1350 rad/s there are generally at least 20 acoustic modes contributing to acoustic pressure, so statistical energy analysis may be employed. But because the gas circuit consists mainly of high-aspect-ratio cavities, reverberant fields are nowhere isotropic below 7500 rad/s, and in some regions are not isotropic below 60 000 rad/s. In comparison with isotropic reverberant fields, these anisotropic fields enhance the radiation efficiencies of some structural modes at low frequencies, but have surprisingly little effect at most frequencies. The efficiency of a dipole sound source depends upon its orientation.     

Development of Sensors for On-Line Monitoring of Nonmetallic Impurities in Liquid Sodium

Sensors for on-line monitoring of hydrogen and carbon in sodium and hydrogen in argon cover gas circuits over sodium have been developed. The performance of these sensors in fast breeder test reactor (FBTR) and large sodium facilities is evaluated. A sensor for monitoring oxygen in sodium is under development. The in-sodium electrochemical hydrogen sensors are found to detect about 10 ppb increase in hydrogen concentration over a background of 50 ppb. The cover gas hydrogen monitoring sensor system is found to sense hydrogen down to 2 vppm in argon over sodium systems. Electrochemical carbon sensors are capable of detecting down to 1 ppm of carbon in sodium.     

Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are 2-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.     

Model development for analysis of the Korea advanced liquid metal reactor

The SSC-K code is under development for analysis of the Korea Advanced LIquid MEtal Reactor (KALIMER) design adopting a pool-type reactor in Korea. The SSC-L code which was originally developed at Brookhaven National Laboratory for analysis of a loop-type liquid metal reactor, is its precursory code. The main reason for the development is that SSC-L cannot be applied directly to the KALIMER design because its application is limited to only a loop-type reactor. The SSC-K code represents the core with multiple coolant channels incorporated with a point kinetics model for calculation of the reactivity feedback. It can provide detailed one-dimensional thermal-hydraulic simulations not only for the primary and secondary sodium coolant circuits, but also the steam/water circuit of the balance-of-plant. This paper presents an overview of the recent developments on the physical models for SSC-K. Those developments are concerned with the two-dimensional hot pool model for analysis of the thermal stratification phenomena in the hot pool, the model for the passive decay heat removal system, the sodium boiling model in the core, and other physical models necessary for the KALIMER analysis. It also demonstrates the analysis results for the unprotected accidents like unprotected transient over power, unprotected loss of flow, and unprotected loss of heat sink postulated in the preliminary KALIMER design. The major focus of these analyses is made on confirmation of the inherent safety characteristics for the reactivity feedback in the core.     

Some Experimental Work Performed on the BOR-60 Reactor

The experimental work performed on a BOR-60 reactor over a period of 30 years of reactor operation is briefly reviewed. The results of investigations of the neutron-physical, thermohydraulic, and dynamical characteristics and the safety parameters of the reactor are presented. The investigations performed and the analysis of transient and emergency regimes made it possible to improve the standard shutdown and cooldown systems in order to soften the temperature conditions on the reactor components.The result of a series of experiments on the safety of fast sodium reactors, among which the introduction of gas into the core, sodium boiling, blocking of the flow in the experimental fuel assembly with destruction of fuel elements, interloop leakage in the steam generators, and so on, are discussed. A complex of diagnostics systems has been developed and tested on the basis of the safety investigations.Analysis of the radiation parameters and characteristics of the reactor made it possible to develop methods and means for monitoring and improving the radiation conditions and the safety of the reactor.Experimental irradiation of various initial materials, using threshold and other reactions, enabled the serial production of radionuclides for medical purposes.     

Sodium boiling experiments in a 7-pin bundle: Investigations into some safety aspects of the KNK II driver zone

A series of sodium boiling experiments, in which the thermohydraulic characteristics of KNK II driver subassemblies were simulated, has been carried out for the purpose of studying the effects of two types of cooling disturbances: rapid flow interruption and those flow reductions which develop rather gradually. Information about the spatial and temporal development of boiling, voidage and dryout was obtained. Furthermore the feasibility of individual subassembly temperature monitoring has been investigated as well as that of two integral boiling detection methods based on acoustic noise and reactivity measurements.     

Experimental study on the application of acoustic emission during the hydraulic pressurization of a reactor pressure vessel

The investigations were aimed at demonstrating the state of the art of acoustic emission testing (AET) of reactor pressure vessels. The object under investigation was the large reactor pressure vessel of the MPA in Stuttgart, a boiling-water reactor pressure vessel, which was provided with a multitude of flaws in weld seams and in the base material. Six hydrostatic tests approximately up to the working pressure of a boiling-water reactor (71 bar) were carried out. In addition to the global multichannel locating technique, also local monitoring techniques were applied. Global location permitted a large number of different indications to be detected simultaneously. Not all of the known flaws did, however, show the expected number of AE events. On the other hand, it was possible to detect flaws previously unknown to the AE staff in some weld seams; these indications were confirmed by nondestructive testing. It was demonstrated that the locating accuracy of local monitoring using signal analysis was improved by a factor of 20 to 30 compared to global monitoring.     

Detection of acoustic emission events in the presence of parasitic noise

The optimum threshold for the detection of acoustic emission bursts, which are masked by parasitic noise is calculated. The theoretical results are applied to the acoustic noise generated by boiling sodium.     

Experimental research on heat transfer to liquid sodium and its incipient boiling wall superheat in an annulus

Liquid sodium is mainly used as a cooling fluid in the liquid metal fast breeder reactor （LMFBR）, whose heat transfer, whether convective heat transfer or boiling heat transfer, is different from that of water. So it is important for both normal and accidental operations of LMFBR to perform experimental research on heat transfer to liquid sodium and its boiling heat transfer. This study deals with heat transfer with high temperature （300-700℃） and low Pe number （20-70） and heat transfer with low temperature （250-270℃） and high Pe number （125-860）, and its incipient boiling wall superheat in an annulus. Research on heat transfer involves theoretical research and experiments on heat transfer to liquid sodium. It also focuses on the theoretical analysis and experimental research on its incipient boiling wall superheat at positive pressure in an annulus. Semiempirical correlations were obtained and they were well coincident with the experimental data.     

防城港核电站堆内中子通量测量系统指套管碰磨分析

防城港核电站1号机组主泵惰走试验期间,在对核电站松脱部件和振动监测系统13路加速度通道进行背景噪声例行检查时发现,通过松脱部件和振动监测系统的声音监听设备监测到,安装于反应堆压力容器底部堆内中子通量测量系统导向管上通道有"哒哒哒"的异常信号。为找出异常信号源,利用松脱部件监测系统声监测功能对压力容器底部监测到的异常信号进行分析,该信号不是由松脱部件产生的信号。通过听音棒的辅助监听,最后综合分析得出该信号是由堆内中子通量测量系统指套管在管道路径上碰磨引起。该事件的分析与解决,不仅解决了工程建设需要,对核安全局批准下一步工作开展提供了支持依据,而且对通过松脱部件监测系统来开展由于流致振动引起的中子通量测量系统指套管异常振动诊断有重大的实用价值。     

Experience with identification of loose parts by acoustic monitoring of primary system

Substantial progress has been achieved in the identification of loose parts which had been detected by acoustic monitoring of reactor primary system. Several years of practical experience and the use of the offline digital analysis system MEDEA proved that acoustic monitoring is very successful for detecting component failures at an early stage. ISTec is involved in loose parts monitoring in several nuclear power plants in Germany. Advanced powerful tools for classification and evaluation of burst signals have been realised.

Loose parts monitoring systems, which are installed in all German nuclear power plants (NPPs), indicated specific impact conditions at lower plenum of two BWR's. Flow tests were carried out with various coolant flow rates of internal axial pumps and use of model nuts in one case. More than 2000 different bursts have been analysed to provide information in detail about impact occurrences, their spectral characteristics and impact sequences. Burst shape parameters could be determined and signal amplitudes have been trended. Determination of the sound origin — fixed origin in one case, flow-induced moving origin in the other case - and mass estimation of the loose parts could be performed by application of advanced burst analysis methods. Characteristics of the impact signals are presented in the paper.