Fractional Scaling Analysis for IRIS pressurizer reduced scale experiments

About twenty organizations joined in a consortium led by Westinghouse to develop an integral, modular and medium size pressurized water reactor (PWR), known as international reactor innovative and secure (IRIS), which is characterized by having most of its components inside the pressure vessel, eliminating or minimizing the probability of severe accidents.     

A statistically-engineered approach for assessing ageing effects on thermal-hydraulic elements for CANDU reactors

The purpose of this study was to develop a valid method to assess ageing effects on thermal-hydraulic elements for CANDU reactors. This method consisted of six steps. The ageing elements used in this method allowed for the immediate consideration of the code input without adjusting preexisting NPP simulation codes, and it also comprehensively considered the change in NPP’s thermal-hydraulic elements due to ageing effects. Each ageing element was selected from among the many thermal-hydraulic factors in which each element would have the greatest effect on the thermal-hydraulic conditions determined by the analysis of the ageing effect on the reactor. In this process, sensitivity analysis for each ageing element was done to understand the effects of each ageing element on the thermal-hydraulic conditions and peak cladding temperature. In addition, a degradation model capable of anticipating the values of the ageing elements over time was developed based on statistical interpretation methods and measured data, and the results’ conservativeness was guaranteed by conservatively selecting optimized combinations of ageing elements and their effects. The inherent uncertainty found in the complex nature of ageing for thermal-hydraulic elements can be reduced by being very conservative. Thus, the concept of the safety margin was introduced to propose a criterion for the assessment of ageing effects on thermal-hydraulic elements in NPPs. In addition, a preliminary analysis of Wolsong Unit 2 has been done. The results show that the 3rd highest value of PCT during LBLOCA was higher than that of the baseline with a value of 21.4 K. Thus, the ageing effect which is not taken into consideration in existing accident analysis evaluation methods was evaluated in this study. Moreover, it was found that meaningful differences may occur from the consideration of the safety analysis of NPP accidents. Accordingly, this method could synthetically assess the ageing effects on thermal-hydraulic elements in CANDU reactors, and this is expected to make considerable contributions to secure reliable safety margins for NPPs.     

Piping system response induced by thermal-hydraulic transient: A review of recent analysis methods and experiments

Complete thermal-hydraulic and structural dynamic response analysis of piping systems subjected to a thermal hydrodynamic transient, such as a safety/relief valve (S/RV) opening, is a complex multi-step process. The four links of this analysis chain are thermal-hydraulic analysis, mechanical loads calculation, structural dynamics analysis, and transient thermal stress analysis. This paper presents summaries of the individual analysis steps, guidelines for the performance of these analyses, and a review of recent experimental versus analytical prediction comparison studies. Finally, research needs are discussed.     

The IRBIS integrated approach to designing pipe structures

Keldysh Applied Mathematics Institute, Russian Academy of Sciences. Electrochemical Construction Research Institute. Translated from Atomnaya Énergiya, Vol. 72, No. 5, pp. 510–517, May, 1992.     

Some challenges to the thermal-hydraulic codes

The thermal-hydraulic codes were developed with the data and correlations obtained from separate effect tests. As such. There are some system-related phenomena which cannot be depicted properly by the codes. In this paper we discuss the difficulties encountered by code modeling for the following systems: feedback loop, multichannel system, multidimensional flow and multiloop circulation. The discussion shows that codes can only give probable answers; the difficulties encountered are due to maldistribution of heat and flow, primary-secondary interaction, feedback effect, instrumentation-control interaction and other unknown factors.     

A thermal-hydraulic code (THYD) for the miniature neutron source reactor thermal-hydraulic transients

The transient thermal-hydraulic problem of MNSR is represented by ten differential equations solved numerically using Runge–Kutta method.Computational results are then compared with experimental measurements. Fuel grids and cooling coil models are incorporated in the model too. Radiating energy from the clad is taken into account in the energy balance in the reactor. The pool is divided into three sections in the model. The effect of the cooling coil of the pool upper section on reactor thermal-hydraulic parameters is discussed. The only input parameter of the reactor is the power temporal distribution. Good agreement between calculated and measured data was obtained.     

Analytical benchmarks for verification of thermal-hydraulic codes based on sub-channel approach

Over the last year (2007), preliminary tests have been performed on the Moroccan TRIGA MARK II research reactor to show that, under all operating conditions, the coolant parameters fall within the ranges allowing the safe working conditions of the reactor core. In parallel, a sub-channel thermal-hydraulic code, named SACATRI (Sub-channel Analysis Code for Application to TRIGA reactors), was developed to satisfy the needs of numerical simulation tools, able to predict the coolant flow parameters. The thermal-hydraulic model of SACATRI code is based on four partial differential equations that describe the conservation of mass, energy, axial and transversal momentum. However, to achieve the full task of any numerical code, verification is a highly recommended activity for assessing the accuracy of computational simulations. This paper presents a new procedure which can be used during code and solution verification activities of thermal-hydraulic tools based on sub-channel approach. The technique of verification proposed is based mainly on the combination of the method of manufactured solution and the order of accuracy test. The verification of SACATRI code allowed the elaboration of exact analytical benchmarks that can be used to assess the mathematical correctness of the numerical solution to the elaborated model.     

Transferability of results of small scale experiments to real structures

Numerical evaluations in combination with experiments on the basis of the J-integral methods are a necessary step in the chain of transferability from small specimens to real structures.For three cases, single-edge notched specimens of different thicknesses, flat plates under tension containing two through-cracks and side-grooved compact specimens of various steels, both finite element calculations including crack growth and experiments using the partial unloading technique were performed.The results show a good agreement of the experimental and numerical J-values and confirm the experimental procedure to evaluate J from the work done on the specimen.Moreover, for the single-edge notched specimen the strong influence of the angular stiffness of the loading system on the specimen behaviour is demonstrated.     

A new computational approach to nuclear aerosol problems

The nonlinear kinetic aerosol equation, describing the time evolution of an aerosol distribution within a well-stirred container, is formulated in a mathematically “conservative” form. A numerical method is then developed for which conservation of mass is automatically satisfied. This procedure simplifies the derivation of conservative numerical schemes by reducing the number of approximations that must be made. Comparisons between an exact solution of the kinetic aerosol equation and numerical approximations show the following: numerical solutions based on the conservative form of the kinetic equation are more accurate and are obtained more efficiently than numerical solutions based on the standard “nonconservative” form of the kinetic equation.     

A new approach to energy-dependent neutron transport theory

A new approach to energy-dependent neutron transport theory is described which treats the asymptotic solution of the Boltzmann equation exactly and approximates the spatial transient by the separable kernel model.The method is illustrated by an application to the Milne and pulsed neutron problems, and exact expressions are obtained for the extrapolated endpoint and the emergent angular distribution.It is pointed out that the method can be extended to include criticality problems and three-dimensional geometries.     

A new approach to automatic radiation spectrum analysis

The application of adaptive methods to the solution of the automatic radioisotope identification problem using the energy spectrum is described. The identification is carried out by means of neural networks, which allow the use of relatively reduced computational structures, while keeping high pattern recognition capability. In this context, it has been found that one of these simple structures, once adequately trained, is quite suitable to identify a given isotope present in a mixture of elements as well as the relative proportions of each identified substance. Preliminary results are presented, and are deemed good enough to consider these adaptive structures as powerful and simple tools in the automatic spectrum analysis     

Introducing and validating a new method for coupling neutronic and thermal-hydraulic calculations

In this study, a new and innovative method is introduced for analyzing neutronic and thermal-hydraulic calculation. For this aim, VVR-S research reactor was selected, and the calculation procedure was performed for it. WIMS, CITATION and COBRA-EN codes were used for investigation. Calculation model consists of two sub-models: neutronic and thermo-hydraulic. The neutronic model uses WIMS and CITATION codes for neutronic simulation of the reactor core and calculating flux and power distribution over it. WIMS code simulates the fuel assemblies and CITATION models the core. The thermal-hydraulic model uses COBRA-EN code for performing the relative calculation. In this study, FORTRAN 90 program is used for linking two sub-models and performing the calculation. The proposed procedure is performed for VVR-S analysis and finally, the obtained results are compared with the experimental results that show good agreement with it.     

A new approach to plasma profile control in ITER

Advanced tokamak operation in ITER, such as the steady-state and hybrid modes, requires an active real-time feedback control of plasma profiles to achieve the advanced regimes for sustained operation. In this work, we have explored a potentially robust control technique that simplifies the active real-time control of electron temperature and safety factor profiles in ITER. As a new and simple approach, static responses of the plasma profiles to power changes of auxiliary heating and current drive are modelled and updated in real-time, differing from the techniques which use a dynamic model deduced from identification experiments, or even a simplified explicit model. To allow real-time update of the plasma profile response model, the underlying physics is simplified with several assumptions. The electron temperature profile response is modelled by simplifying the electron heat transport equation. The safety factor profile response is modelled by directly relating it to the changes of source current density profiles. The required actuator power changes are calculated using the singular value decomposition technique, taking the saturation of the actuator powers into account. The potential of this control technique has been tested by applying it to simulations of the ITER hybrid mode operation using CRONOS. In these simulations, the electron temperature and safety factor profiles were well controlled either independently or simultaneously.     

Numerical investigation on thermal-hydraulic performance of new printed circuit heat exchanger model

Three-dimensional numerical analysis was performed to investigate heat transfer and pressure drop characteristics of supercritical CO₂ flow in new Printed Circuit Heat Exchanger (PCHE) model using commercial CFD code, Fluent 6.3. First, numerical analysis for conventional zigzag channel PCHE model was performed and compared with previous experimental data. Maximum deviation of in-outlet temperature difference and pressure drop from experimental data is about 10%. A new PCHE model has been designed to optimize thermal-hydraulic performance of PCHE. The new PCHE model has several airfoil shape fins (NACA 0020 model), which are designed to streamlined shape. Simulation results showed that in the airfoil shape fin PCHE, total heat transfer rate per unit volume was almost same with zigzag channel PCHE and the pressure drop was reduced to one-twentieth of that in zigzag channel PCHE. In airfoil shape fin PCHE model, the enhancement of heat transfer area and the uniform flow configuration contributed to obtain the same heat transfer performance with zigzag channel PCHE model. And the reduction of pressure drop in airfoil shape fin PCHE model was caused by suppressing generation of separated flow owing to streamlined shape of airfoil fins.     

Reduced scale PWR passive safety system designing by genetic algorithms

This paper presents the concept of “Design by Genetic Algorithms (DbyGA)”, applied to a new reduced scale system problem. The design problem of a passive thermal-hydraulic safety system, considering dimensional and operational constraints, has been solved. Taking into account the passive safety characteristics of the last nuclear reactor generation, a PWR core under natural circulation is used in order to demonstrate the methodology applicability. The results revealed that some solutions (reduced scale system DbyGA) are capable of reproducing, both accurately and simultaneously, much of the physical phenomena that occur in real scale and operating conditions. However, some aspects, revealed by studies of cases, pointed important possibilities to DbyGA methodological performance improvement.     

聚变制氢堆高温液态包层热工水力学新概念研究

在深入分析聚变堆包层设计要求和目前技术发展水平的基础上,根据热化学工艺制氢需要高温热的要求,提出了一个基于技术相对成熟的低活化铁素体/马氏体钢作为主要结构材料、高压氦气与液态LiPb合金作为冷却剂、具有创新性“多层流道插件”结构方案以获得高温热能的包层热工水力学概念,建立了热工水力学模型,在利用有限元数值模拟程序进行模拟计算的基础上分析了这种新概念包层的可行性。     

A momentum integral network method for thermal-hydraulic systems analysis

The momentum integral network method employed in the MINET code is described in detail. In using this method, pressure is calculated dynamically in only a few key locations, and long runs of piping, etc., are each covered by one integral (over multiple nodes) momentum equation. Advantages of this approach for large and complex thermalhydraulic systems such as balance of plant, are described. Validation studies are summarized and cited. A comparison is made between this approach and the local pressure representation. A nodalization study is reported, in which the computational time is shown to increase linearly (or less) with the number of nodes, in contrast to a higher order increase expected from calculations based on a local pressure representation.     

Transient experiments from the thermal-hydraulic ADS lead bismuth loop (TALL) and comparative TRAC/AAA analysis

TALL is a medium-size experimental facility constructed at KTH, to study the steady-state and transient thermal-hydraulics performance of LBE-cooled reactors, with the primary purpose of supporting the European transmutation demonstration (ETD) using LBE-cooled accelerator-driven systems (ADS). This paper presents the results of transient experiments performed on the TALL test facility, whose aim is to provide a data base for validation of computer codes which may be used for the analysis of the safety of those systems. This paper also presents the results of the post-test calculations, carried out at PSI, using the TRAC/AAA code. The transient experiments performed include the loss of heat sink, the loss of pump, the loss of both primary and secondary flows, overpower, overcooling, heater trip, and the operational transients of start-up and shut-down. The experimental results show the excellent natural circulation performance of a LBE-cooled system which should contribute to a good safety performance. The TRAC/AAA calculations provide results which agree well with the experimental data.     

A new approach to radioactive waste self-burial using high penetrating radiation

The method of self-burial of radioactive waste in geological formations using direct heating of rocks by radiation is proposed in this paper. In the currently known studies, thermal conductivity is considered as a main heat transfer mechanism. Application of high penetrating gamma radiation for direct melting of surrounding rocks will reduce the energy absorption inside the sinking device and will lower maximum temperature and temperature gradients in the elements of the device. In this paper, conditions of realization of the direct heating by radiation mechanism are presented and requirements to heat-generating radionuclides have been derived. Assessments of the spatial distribution of energy release in the surrounding rocks for the point and plane sources with the radionuclide ⁶⁰Co have been performed. Based on these data, the temperature distributions in the surrounding rocks and the expression for determining the descent velocity as a function of ⁶⁰Co surface activity in the sinking device have been obtained. Estimations of energy absorption fraction inside the spherical heat-generating elements filled with ⁶⁰Co and surface activity of ⁶⁰Co, necessary to achieve velocity of about 1 km per year, have been made. The results are given for granite and salt rocks.