Analysis of temperature noise propagation in a large fuel subassembly of FBR

We have developed a method to calculate the three-dimensional distribution of root-mean-square (RMS) values of temperature noise in the single phase flow in a fast reactor fuel subassembly with a local flow blockage. Employed are the subchannel method in a pin bundle region and the finite difference method in the region downstream of the bundle. We have compared the calculated RMS values of temperature noise with experimental data for a sodium loop test using a wire-spacered 91-pin-bundle fuel sub-assembly with a local blockage. We have investigated the possibility of detection of the blockage by temperature noise by taking into account the influence of structures in the upper part of the subassembly.     

Detection of local boiling in an LMFBR subassembly by temperature fluctuations analysis at the outlet

The aim of the experiments is to detect boiling in a sodium cooled subassembly by measuring fluctuations behind the bundle outlet. The measurements were carried out on an electrically heated 28-rod bundle with a partially blocked section. Fast responding thermocouples were installed downstream of the bundle outlet and downstream of a flow mixing system. Statistical parameters were investigated such as root mean square (RMS) and power spectrum density (PSD). The boiling conditions were generated by reducing the system pressure or flow velocity reduction. The experiments have shown that statistical analysis of temperature fluctuations can produce significant results in the detection of boiling behavior at both the outlet of a subassembly, and behind a flow mixing system.     

Electrochemical noise measurements of stainless steel in high temperature water

Corrosion in a high purity aqueous environment simulating a boiling water reactor (BWR) is addressed in this work. This condition necessitates autoclave experiments under high pressure and temperature.Long-term electrochemical noise measurements were explored as a mean to detect and monitor stress corrosion cracking phenomenon. An experimental set up, designed to insulate the working electrode from external interference, made possible to detect and monitor stress corrosion cracking in slow strain rate tests for sensitized and solution annealed 304 stainless steel at 288 °C. Time-series analysis showed variations in the signature of the current density series due to transgranular stress corrosion cracking (TGSCC) and intergranular stress corrosion cracking (IGSCC).     

Comparison of various models for the simulation of a Core Disruptive Accident in the MARA 10 mock-up

During a Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor melts down partially and that the interaction between hot molten fuel and relatively cold liquid sodium creates a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a Core Disruptive Accident in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents some models available within the EUROPLEXUS code to simulate a Core Disruptive Accident and an analysis of the computed results. In particular, results are compared with experimental measurements and previous numerical simulations carried out with the codes SIRIUS and CASTEM-PLEXUS.     

Measurement of subassembly outlet coolant temperature in the JOYO experimental fast reactor

Thermocouple temperature sensors are installed above the central region of the core in the JOYO experimental fast reactor to monitor the outlet coolant temperature of 115 subassemblies. This paper summarizes the experimental temperature data obtained during initial 50 MWt operation of the reactor. Subassembly outlet coolant temperature distributions that were obtained under various power levels, different main cooling system flowrates, and unequal reactor inlet temperatures from the two cooling loops are described. In addition, coolant temperature and flowrate distributions at the subassembly outlet measured in a zero power experiment are presented.     

The prediction of temperature distribution of a subassembly including intersubassembly heat transfer

A three-dimensional thermal hydraulic code, CLUSTER, has been developed in order to predict the temperature profile in a cluster of seven subassemblies during steady state conditions. The subassemblies under consideration include both pinbundle and reflector types of subassemblies. An implicit finite difference representation of the governing energy equation is employed. Good agreement was obtained between the analytical results and the measured data.     

Thermal transient and the temperature profile in a HELICA mock-up simulated by a new finite element homogenous model

This paper deals with a numerical approach for simulating the thermal and mechanical behaviour of pebble beds used as breeder and neutron multiplier in breeding blanket of nuclear fusion reactor. The model of the pebble beds is based on the results of a theoretical and experimental research activity performed by the Authors on ceramic pebble beds (lithium ortosilicate and lithium metatitanate). The results of this activity permitted to determine the effective thermal conductivity of the beds, versus the temperature and the axial pressure and to implement a homogenous model of pebble bed in a FEM code.This paper illustrates an application of the implemented model, considering pebble beds under several cycles of heating and cooling. The examined geometry corresponds to the HELICA mock-up tested by ENEA in the research centre Brasimone. The experimental tests performed on HELICA have been used as a benchmark problem in order to assess the different approaches for simulating pebble beds. In this paper, the simulations performed with two-dimensional models are illustrated. Moreover the numerical results are compared with the experimental ones. Finally, a discussion on results obtained by other authors involved in the benchmark is reported.     

Results of the temperature noise measurements in subchannels of sodium cooled electrically heated 28 rod bundle behind a blockage

For LMFBR safety studies a 28 rod bundle has been built at Petten (cooperation of GfK and ECN), representing a 60-degrees section of an SNR-300 fuel element having a 70% flat type central blockage. The aims of the temperature noise measurements were to determine the subchannel coolant velocities behind the blockage to study the mixing of coolant in subchannels of different temperature from behind the blockage to the outlet and to study the temperature noise due to boiling in a subchannel. The temperature noise measurements were carried out in parallel to the other measurements (temperature distribution, etc.), using signals of fourteen subchannel thermocouples placed in five measuring planes behind the blockage. The single phase measurements were made with several heat fluxes (5 W/cm² to 120 W/cm²), inlet flows (0.25 to 3 m/s) and inlet temperatures (250°C to 600°C). Two phase flow is initiated and sustained either by a slow and continuous pressure reduction or by stepwise reduction of the main flow. The temperature noise signals were amplified and recorded in analog form. Later the signals were digitized and analysed by digital computers. Part of the signals was also processed by a hardware correlator. The experimental results of the temperature noise measurements will be shown for the different conditions of the loop. Measurements clearly show the following effects:

• - the recirculation flow pattern due to the vortex in the wake behind the blockage;

• - the dependence of r.m.s. value of the noise on the heat flux and the coolant flow;

• - the increase in noise and change in power spectra when going from single phase to the boiling condition.

     

Computation of a core disruptive accident in the MARS mock-up

A hypothetical core disruptive accident in a liquid metal fast breeder reactor (LMFBR) results from the interaction between molten fuel and liquid sodium, which creates a high-pressure bubble of gas in the core. The violent expansion of this bubble loads and deforms the vessel and the internal structures. The MARS experimental test simulates a HCDA in a small-scale mock-up containing all the significant internal components of a fast breeder reactor. The mock-up is filled with water, topped by an argon blanket, and the explosion is generated by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code. The top closure is represented by massive structures and the main internal structures are described by shells. The current numerical results are described and compared with the experimental ones, and previous computations with the CASTEM-PLEXUS code.     

Multi-fluid multi-phase subchannel analysis of subassembly accidents of LMFRs

Analyses of experiments simulating hypothetical subassembly accidents such as a large-scale inlet blockage in a Liquid Metal Cooled Fast Reactor (LMFR) have been performed with computer program KAMUI. With the use of relatively simple but reasonable constitutive models, the code has been applied to the SCARABEE experiments BE+1 and APL1 to validate the analytical capability against the accident conditions under the multi-pin geometry. The results show that the key events such as sodium boiling, clad melting, molten clad relocation, molten clad freezing were adequately simulated taking into account the effect of heat loss to the coolant flow in the outside channel of the test section.     

Deployment analysis and remote accessibility verification for a maintenance task in a PRIDE digital mock-up

This study introduces an advanced modeling and simulation system that can verify and optimize a maintenance procedure at the early design stage of a virtual engineering system. System architecture, which is composed of modules used to analyze deployment of devices that deal with radioactive materials in a digital mock-up, and modules used to simulate accessibility of a remote manipulator for maintenance tasks, are discussed. Based on this architecture, new technology that enables a maintenance analyst to analyze pyroprocessing is proposed. Workspace analysis for remote manipulators, which can optimize a maintenance task at the radiation control area, is introduced. The mathematical background, with respect to the forward and inverse kinematics for a virtual manipulator accessing devices in a virtual environment, is described to establish a remote accessibility. Virtual prototyping is illustrated to carry out an experiment on the deployment analysis of devices and remote accessibility by a haptic device.     

Neutronics analysis and nuclear heating measurement up to the TFC in a mock-up of the ITER inboard shield

A mock-up of ITER including the inboard shield, the vacuum vessel and the coil region, was set up at ENEA Frascati and irradiated with 14 MeV neutrons produced by the Frascati Neutron Generator (FNG). The mock-up dimensions and materials composition are consistent with the current ITER design.The primary objective of the experiment was to validate the MCNP calculations (C) of nuclear heating measured (E) in the region corresponding to the ITER Toroidal Field (TF) coil. An accuracy on C/E ratio ≤±10% was required.The neutronics and shielding properties of the mock-up were also studied throughout the measurement of selected activation reaction rates up to about 1 m depth. Due to the very low activity induced in the foils, the measurements in the deepest experimental positions were performed at the underground low background facility of the Laboratori Nazionali del Gran Sasso, using ultra-low background high purity germanium (HPGe) detectors.The measured reaction rates and nuclear heating were thus compared with the results of the Monte Carlo code MCNP5 coupled with the FENDL-2.1 library.     

Hydraulic simulation of local blockage in a LMFBR fuel subassembly

Experimental studies of local flow blockage in a LMFBR fuel subassembly have been carried out using a simulating model in a water test loop. The studies verified the numerical results of an analytical code. The experiments were conducted in a 61-pin bundle containing a planar blockage without leakage flow. Central and edge blockage were used. The wake flow behind the blockage was visualized with dye or air bubble injection to grasp the flow characteristics. The velocity, static pressure and residence time were measured in the wake flow region.     

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.     

Detection of cooling anomalies in LMFBR subassemblies by means of temperature noise analysis techniques

Experimental and theoretical works, performed in different research centers, indicate that temperature noise analysis is a powerful method to detect, at an early stage, cooling anomalies in LMFBR subassemblies. This technique is under investigation at CNEN. Tests have been performed on electrically heated fuel assemblies mockups and channel exit temperature fluctuations have been studied for both normal and anomalous thermohydraulic conditions. To back up experimental results, a stochastic model has been implemented, reproducing exit temperature noise characteristics.The results are in good agreement with those obtained by other investigators and offer valuable informations on sensor characteristics and position, data collecting procedures and stochastic indicators potentiality.     

北京谱仪附近空间电磁噪声测量和抑制工作

当北京正负电子对撞机（BEPC）提供的e^ /e^-束流通过北京谱仪（BES）时,在对撞区附近会产生一个空间电磁信号。采用单点测量和矩阵测量等方法,对其进行时域和频域测量及分析。为寻找BES附近的干扰噪声源,提供了非常有益的数据。确认了静电分离器的屏蔽改进效果,并且找到了BES附近另一个干扰噪声源-束流管道上安装DCCT（直流电流传感器）位置附近的陶瓷结构的不连续性而泄漏的电磁波。对其采取了屏蔽、接地等一系列电磁兼容措施,极大地净化了BES附近的空间电磁环境,提高了BES对BEPC束流的利用率,改善了BES数据的可靠性。     

On the theoretical-numerical study of the HEXCALIBER mock-up thermo-mechanical behaviour

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facility.The paper presents the main results of the mock-up experimental tests and of their numerical simulations performed adopting the finite element method, which allowed the DIN constitutive model to be assessed and validated.