Statistical analysis of cross sections—A new tool for data validation

Several neutron activation cross section libraries have been produced recently with data above the energy of 20 MeV. These are needed because of new applications, such as the International Fusion Material Irradiation Facility (IFMIF) and Accelerator Driven Systems (ADS). A standard validation procedure for such libraries is a comparison with differential and integral experimental data. However, the number of differential data above 20 MeV is relatively small and there are very few integral experiments with neutron spectra dominated by high-energy (>20 MeV) neutrons. Furthermore the size of these libraries (about 775 targets and >60,000 reactions) favours global test procedures. It is shown that the method of ‘Statistical analysis of cross sections’ (SACS) can be used for this purpose. This is based on the assumptions that cross section excitation curves for all targets have a similar (bell) shape shifted by differences in Q-value, and that their properties, such as the maximum of the excitation curve (σ_max), the incident neutron energy at which σ_max occurs (E_max) and the width of the excitation curve at half its maximum (Δ_1/2) have a pronounced smooth trend as a function of A or s. It will be demonstrated that this method of analysis is a very effective and novel way to validate reaction cross section data, both for their shape and their magnitude. This will be illustrated using the recent European Activation Library EAF-2005, produced within the EFDA Fusion Technology Programme. The majority of the cross section data in the library are based on calculations with the nuclear reaction code TALYS.     

A new procedure to generate resonance integral table with an explicit resonance interference for transport lattice codes

Resonance interference could not be considered explicitly in the conventional resonance treatment employing subgroup and direct resonance integral methods when using coarse energy group structure. This problem comes from the lack of information for the resonance shapes of resonant nuclides in the resonance interference formulas. As energy group boundaries get coarser, inaccuracy in estimating self-shielded cross sections with resonance interference gets bigger. A new method has been proposed to conserve the self-shielded cross sections for each group through an explicit consideration of resonance interference effect, which results in a good accuracy in predicting the multiplication factor. This method can be applicable to various mixing combinations of constituent resonant nuclides with resonance interference and can cover wide dilution range. The MERIT code has been used to generate resonance integral tables and to estimate resonance interference effects. And the 2-D transport lattice code KARMA has been used to perform sample calculations to see the effectiveness of the newly developed method. Sample calculations have been performed for single pins with various temperatures, ²³⁵U enrichments and dilution levels with the 47 and 190 energy group structures. The computational results show that this method is able to estimate self-shielded cross sections in each coarse energy group accurately for various temperatures and various geometry and composition configurations.     

A benchmark exercise on the use of CFD codes for containment issues using best practice guidelines: A computational challenge

In the framework of the 5th EU-FWP project ECORA the capabilities of CFD software packages for simulating flows in the containment of nuclear reactors was evaluated. Four codes were assessed using two basic tests in the PANDA facility addressing the transport of gases in a multi-compartment geometry. The assessment included a first attempt to use Best Practice Guidelines (BPGs) for the analysis of long, large-scale, transient problems. Due to the large computational overhead of the analysis, the BPGs could not fully be applied. It was thus concluded that the application of the BPGs to full containment analysis is out of reach with the currently available computer power. On the other hand, CFD codes used with a sufficiently detailed mesh seem to be capable to give reliable answers on issues relevant for containment simulation using standard two-equation turbulence models. Development on turbulence models is constantly ongoing. If it turns out that advanced (and more computationally intensive) turbulence models may not be needed, the use of the BPGs for ‘certified’ simulations could become feasible within a relatively short time.     

A global approach to the physics validation of simulation codes for future nuclear systems

This paper presents a global approach to the validation of the parameters that enter into the neutronics simulation tools for advanced fast reactors with the objective to reduce the uncertainties associated to crucial design parameters. This global approach makes use of sensitivity/uncertainty methods; statistical data adjustments; integral experiment selection, analysis and “representativity” quantification with respect to a reference system; scientifically based cross-section covariance data and appropriate methods for their use in multigroup calculations. This global approach has been applied to the uncertainty reduction on the criticality of the Advanced Burner Reactor (both metal and oxide core versions) presently investigated in the frame of the GNEP initiative. The results obtained are very encouraging and allow to indicate some possible improvements of the ENDF/B-VII data file.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

Bats: A new tool for AMS data reduction

A data evaluation program was developed at ETH Zurich to meet the requirements of the new compact AMS systems MICADAS and TANDY in addition to the large EN-Tandem accelerator. The program, called “BATS”, is designed to automatically calculate standard and blank corrected results for measured samples. After almost one year of routine operation with the MICADAS C-14 system BATS has proven to be an easy to use data reduction tool that requires minimal user input. Here we present the fundamental principle and the algorithms used in BATS for standard-sized radiocarbon measurements.     

The cova programme for the validation of computer codes for fast reactor containment studies

The UKAEA and the Joint Research Centre, Euratom, Ispra are engaged in a collaborative experimental programme carrying out a series of small scale, well instrumented tests aimed at providing high quality data of the stresses, strains and loads occurring when a well characterized source is released within a fluid in a containment vessel. In the UK the data are being used to validate the computer codes ASTARTE and SEURBNUK which are used in studies of the response of the fast reactor primary containment system in the event of a hypothetical reactor excursion. This paper describes the UK experimental programme and the development of the low density explosive which is used as the energy source; the rationale of the code development programme is presented together with a report on the progress that has been made in validating the two codes against the experimental data.     

Verification and validation of the HTGR systems CFD code Flownex

Regulatory requirements prescribe extensive verification and validation (V&V) of computer codes that are used in the design and analysis of accident conditions in nuclear plants. Flownex is a dynamic systems CFD code used as the primary thermal-fluid simulation code by the Pebble Bed Modular Reactor Company (PBMR).Stringent quality assurance processes have been implemented to ensure that the code conforms to the set standards. These processes include the comparison of Flownex with analytical results as well as with experimental data.The results of this process are summarized in this paper. Analytical solutions are used to verify Flownex's element models so as to ensure that the basic theory is correctly implemented in the computer code. As part of the analytical V&V effort various well-defined problems are solved using numerical methods implemented in independent computer codes.Comparison with experimental and plant data is a very important feature of the V&V program to validate that the chosen theory is fit for purpose. For this, validation data from the pebble bed micro model (PBMM) is used. In addition to the PBMM experimental data Flownex is compared to a number of small thermal-fluid experiments in which certain specific component phenomena is validated. These experiments were developed in collaboration with North-West University (previously Potchefstroom University).     

A new web-based tool for data visualization in MDSplus

The Java tool jScope has been widely used for years to display acquired waveform in MDSplus. The choice of the Java programming language for its implementation has been successful for several reasons among which the fact that Java supports a multiplatform environment and it is well suited for graphics and the management of network communication. jScope can be used both as a local and remote application. In the latter case, data are acquired via TCP/IP communication using the mdsip protocol. Exporting data in this way however introduces several security problems due to the necessity of opening firewall holes for the user ports. For this reason, and also due to the fact that JavaScript is becoming a widely used language for web applications, a new tool written in JavaScript and called WebScope has been developed for the visualization of MDSplus data in web browsers.Data communication is now achieved via http protocol using Asynchronous JavaScript and XML (AJAX) technology. At the server side, data access is carried out by a Python module that interacts with the web server via Web Server Gateway Interface (WSGI).When a data item, described by an MDSplus expression, is requested by the web browser for visualization, it is returned as a binary message and then handled by callback JavaScript functions activated by the web browser.Scalable Vector Graphics (SVG) technology is used to handle graphics within the web browser and to carry out the same interactive data visualization provided by jScope. In addition to mouse events, touch events are supported to provide interactivity also on touch screens. In this way, waveforms can be displayed and manipulated on tablet and smartphone devices.     

Analyzing the alternative shutdown cooling behaviors for Chinshan Nuclear Power Plant using CFD simulation

The Chinshan Nuclear Power Plant (CSNPP) is a GE-designed BWR4 plant, having two identical units with rated core thermal power of 1804 MWt each unit. Several alternative shutdown cooling methods driven by natural or mixed convection has been proposed by the plant for studying the core cooling capability when the Residual Heat Removal (RHR) systems are not available or the refueling tasks, such as the In Vessel Visual Inspection (IVVI) work etc., is being proceeded. One of the examples is to connect a pipe from the outlet of the new spent fuel heat exchanger to the reactor cavity. The design of the alternatives shall ensure that the maximum core fluid temperature is limited below the boiling temperature of water. In this study, a Computational Fluid Dynamics (CFD) model is developed to analyze the natural convection phenomena during the shutdown period. Through a series of assumption, modeling and meshing processes, a calculation domain with approximate four million meshes including the RPV, reactor cavity and spent fuel pool, have been solved in this study. The analysis results showed that the passive alternative shutdown cooling system could provide sufficient heat removal capability to maintain the maximum core fluid temperature below boiling temperature. The results also indicated that the alternative shutdown cooling system could be served as an appropriate solution for CSNPP when the RHR is inoperable.     

基于CFD方法的反应堆流量分配结构的优化设计

采用商用软件ANSYS CFX12.1对反应堆压力容器内流体区域进行计算流体动力学(CFD)分析,获取了相关结构的压降、堆芯入口流量分布和交混特性等参数,研究球形和格架两种不同流量分配结构形式对堆芯入口水力特性的影响,通过结构参数的优化,确定最终流量分配结构设计方案。     

A new magnetic scanning system for proton microprobe

Design of a new beam scanning system with ferrite-cored magnetic coils is described. Results of testing of the system are presented. Deflecting magnetic field produced by the scanning system shows high degree of uniformity. The fringing fields are decreased sharply along the magnet axis.     

The TONUS CFD code for hydrogen risk analysis: Physical models, numerical schemes and validation matrix

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code, called the TONUS code, which incorporates both lumped-parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present the governing equations, numerical strategy and schemes used for the CFD approach.Several benchmark exercises based on experimental results obtained on large-scale facilities, such as MISTRA, TOSQAN and RUT, are presented. They have been used as verification and assessment procedures for modelling and numerical approaches of the code. Specific emphasis is given to the sensitivity analysis of the computed results with respect to numerical and physical parameters. The powerful Design-Of-Experiments technique for sensitivity analysis is successfully applied to the ISP47 MISTRA test case.The TONUS CFD code is presently used to support the hydrogen risk assessment for the European Pressurized Reactor (EPR) plant and to investigate the impact of the two-room concept on hydrogen distribution in the EPR containment.     

3D Micro-PIXE at atmospheric pressure: A new tool for the investigation of art and archaeological objects

The paper describes a novel experiment characterized by the development of a confocal geometry in an external Micro-PIXE set-up. The position of X-ray optics in front of the X-ray detector and its proper alignment with respect to the proton micro-beam focus provided the possibility of carrying out 3D Micro-PIXE analysis. As a first application, depth intensity profiles of the major elements that compose the patina layer of a quaternary bronze alloy were measured. A simulation approach of the 3D Micro-PIXE data deduced elemental concentration profiles in rather good agreement with corresponding results obtained by electron probe micro-analysis from a cross-sectioned patina sample. With its non-destructive and depth-resolving properties, as well as its feasibility in atmospheric pressure, 3D Micro-PIXE seems especially suited for investigations in the field of cultural heritage.     

Nuclear track formation and skewness of particle trajectories in the target: A new perspective

It is discussed why track detectors such as CR-39 and CN-85, which are only composed of light elements, are highly sensitive and efficient compared with detectors such as glass and mica, which are partly or fully composed of heavier elements. Weak scattering of incident particles by light target atoms does not significantly deviate incident particles from their straight trajectories while the target atoms recoil considerably, damaging the detector. Heavier atoms scatter incident particles through wide angles, significantly deviating them from their straight paths while the target atoms recoil weakly, producing less damage. Simulations are presented to demonstrate these concepts about nuclear track formation.