Validation of the CFX4 CFD code for containment thermal-hydraulics

In order to determine the risk associated with the presence of hydrogen in a nuclear power plant containment during a hypothetical severe accident, thermal hydraulic codes are used. Amongst other codes, NRG uses the commercial computational fluid dynamics code CFX4 for this purpose. Models to describe condensation have been implemented by user coding. This paper describes these models. In addition, an overview is given of validation activities with the CFX4 model. Experimental results from the following sources have been used: the Kuhn condensation model; the PHEBUS test facility; the PANDA test facility; and the TOSQAN, MISTRA, and THAI test facilities within the OECD International Standard Problem 47. The CFX4 model predictions are fairly good. Deviations originate primarily from the applied wall treatment. Several recommendations for further development are therefore proposed.     

COBRA-3M: A modified version of COBRA for analyzing thermal-hydraulics in small pin bundles

COBRA-3M, a modified version of COBRA computer code, is most suitable for the analysis of thermal-hydraulics in small pin bundles commonly used in in-reactor or out-of-reactor experiments. It includes detailed thermal models for the fuel pins and duct walls. It can handle nonuniform power distribution across the bundle and/or within a fuel pin. Temperature dependence of material properties and fuel-cladding gap conductance can be treated. Heat generation in the duct walls and the effect of heat loss to the surroundings can also be simulated. COBRA-3M has been used extensively in the design and analysis of TREAT and SLSF experiments.     

LMFBR system analysis: Impact of heat transport system on core thermal-hydraulics

In discussing LMFBR thermal-hydraulic analysis, this paper focuses on the heat transport system and its impact on the predicted core behavior, particularly during off-normal or protected accident transients. Following a brief background of related work in the area of system simulation for both loop and pool-type LMFBR designs, modeling considerations for individual components such as reactor core, piping, pumps, heat exchangers and check valves, together with the overall integrated approach to system simulation, are discussed. The need for, and current approaches to, modeling pool stratification are also examined. The role of buoyancy forces in the system is clarified, with particular emphasis on its increasing influence during flow decay. Sample results are presented to illustrate the influence of system modeling details, and selection of component parameters and operational mode, on predicted core thermal-hydraulic response during protected loss-of-flow transients. From a systematic study of the effect of pump inertia for a flow coastdown to natural circulation event in a loop-type design, it is found that certain combinations of primary and secondary pump inertias can lead to core flow reversal for a sustained period, and eventual boiling in the hot fuel channel. This effect, based on its impact on core flow, is even more pronounced in pool-type designs.     

ATWS analysis for PWR using the coupled code system DYN3D/ATHLET

The ATWS transient “Loss of main feed water supply” in a generic four-loop PWR at the nominal power of 3750 MW was analyzed using the coupled code system DYN3D/ATHLET. A variation of the MOX-fuel-assembly portion in the core has an effect on the reactivity coefficients of the fuel temperature and the moderator density. These two parameters mainly influence the behaviour of the coolant pressure, which is safety-relevant. It has been demonstrated that the pressure maximum decreases with an increasing portion of MOX. For all core loadings considered, both primary-circuit mechanical integrity and sufficient core cooling are guaranteed.     

IGSHIELD: A new interactive point kernel gamma ray shielding code

IGSHIELD is an interactive gamma ray shielding code (stand alone) developed in Visual Basic Version 6.0 for windows operating system. The computational methodology is based on the point kernel technique employed in QADCGPIC (and hence QADCGGP) code. In IGSHIELD, all the features of QADCGPIC code written in FORTRAN and its user interface GUI2QAD-3D written in visual basic have been incorporated and several additional improvements have been made. The new improvements made in the code are (i) handling of multiple source shapes with accompanying energy distributions, (ii) incorporation of hexagonal cylinder as a shield body in the combinatorial geometry routine, (iii) provision to declare hexagonal cylinder and truncated right circular cone as source bodies, (iv) provision to declare point, line and plane as source bodies and (v) arbitrary orientation of sources such as cylinder, hexagonal cylinder and truncated cone. The newly introduced source bodies will be of immense use for calculations involving hexagonal shaped cylindrical sub-assemblies of fast reactors and cone shaped radioactive plume release from stack. The source array option will be of use for calculations involving stacked radioactive drums in vast storage areas. The code is validated by comparing the results with (i) simple source-shield configurations that can be solved using analytical equations and MCNP code, (ii) a point source in an infinite shield medium configuration that is solved using simple analytical equation and (iii) a standard problem in which an irradiated fuel is surrounded by various shielding materials and the problem is solved using MCNP. Several sample problems encountered in nuclear industry and other useful information required for shield designers and analysts are the other highlights of the code. The input for any kind of situations can be prepared easily using the forms that flash with appropriate prompts. Since entry of data is the important part of the input, its validity is checked immediately and the user is cautioned against incorrect data. The input module accepts only the correct data. If the data happens to be an object, the code instantaneously display the 3D view of all the objects built so far. Thus, the Interactive Gamma Ray Shielding Code will cater the needs of a shield designer as well as serve as an excellent educational tool.     

A coupled analysis of system thermal-hydraulics and three-dimensional reactor kinetics for a 12-finger control element assembly drop event in a PWR plant

A coupled system thermal-hydraulics (T-H) and three-dimensional reactor kinetics code, MARS/MASTER, was developed to attain more accurate predictions for nuclear system transients that involve strong interactions between neutronic and T-H phenomena. In this paper, a 12-finger control element assembly (CEA) drop event in a two-loop pressurized water reactor (PWR) plant under a full power condition was analyzed, where the 12-finger CEA that is nearest to the hot leg of Loop 2 is assumed to incidentally drop. This instantaneously results in an asymmetric radial power distribution and, in turn, asymmetric loop behavior, which may lead to a reactor trip due to a low departure from nucleate boiling (DNB) ratio at the intact side of the core or an excessive difference between the cold leg coolant temperatures. This event clearly requires a coupled calculation of system T-H and three-dimensional reactor kinetics to realistically investigate the thermal-hydraulic behavior of the reactor core. A simple theoretical modeling is also devised to evaluate the cold leg temperature difference under a quasi-steady state.     

FMSR: A code system for in-core fuel management calculation of aqueous homogeneous solution reactor

Aqueous homogeneous solution reactor is a promising concept for the production of medical isotopes. But some characteristics of aqueous solution reactors, such as no traditional assembly in the core, the gas bubbles’ generation in fuel solution, isotopes distillation, unstructured geometry, strong anisotropic scattering, etc., make the fuel management calculation very complicated. This study establishes a suitable calculation model for aqueous homogeneous solution reactors and developed an in-core fuel management calculation code FMSR (Fuel Management for Solution Reactors) based on the 3D transport solver DNTR. Numerical results indicate that FMSR can be used for the fuel management calculation of homogeneous aqueous solution reactor as a trial.     

Simulation of a main steam line break accident using a coupled “system thermal-hydraulics,three-dimensional reactor kinetics,and hot channel analysis” code

For a realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients.     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

LEIS for the prediction of turbulent multifluid flows applied to thermal-hydraulics applications

The paper centres on the use of the so-defined LEIS approach (Large-Eddy & Interface Simulation) for turbulent multifluid flows present in thermal-hydraulics applications. Interfacial flows involving deformable, sheared fronts separating immiscible fluids are shown to be within reach of this new approach, featuring direct resolution of turbulence and sheared interface deformations within the interface tracking (ITM) framework, such as level sets and VOF. In this technique supergrid turbulence and interfacial scales are directly solved whereas the sub-grid (SGS) parts are modelled, at least the turbulence part of it. First results are shown (feasibility), and difficulties and open issues are discussed. The connection between these two particular scales will also be discussed, and potential modelling routes evoked, including combining two-fluid and ITM, local grid refinement, or combing particle tracking and ITM for sub-grid inclusions smaller than the grid size.     

A new code for predicting the thermo-mechanical and irradiation behavior of metallic fuels in sodium fast reactors

An engineering code to predict the irradiation behavior of U–Zr and U–Pu–Zr metallic alloy fuel pins and UO₂–PuO₂ mixed oxide fuel pins in sodium-cooled fast reactors was developed. The code was named Fuel Engineering and Structural analysis Tool (FEAST). FEAST has several modules working in coupled form with an explicit numerical algorithm. These modules describe fission gas release and fuel swelling, fuel chemistry and restructuring, temperature distribution, fuel–clad chemical interaction, and fuel and clad mechanical analysis including transient creep-fracture for the clad. Given the fuel pin geometry, composition and irradiation history, FEAST can analyze fuel and clad thermo-mechanical behavior at both steady-state and design-basis (non-disruptive) transient scenarios.FEAST was written in FORTRAN-90 and has a simple input file similar to that of the LWR fuel code FRAPCON. The metal–fuel version is called FEAST-METAL, and is described in this paper. The oxide–fuel version, FEAST-OXIDE is described in a companion paper. With respect to the old Argonne National Laboratory code LIFE-METAL and other same-generation codes, FEAST-METAL emphasizes more mechanistic, less empirical models, whenever available. Specifically, fission gas release and swelling are modeled with the GRSIS algorithm, which is based on detailed tracking of fission gas bubbles within the metal fuel. Migration of the fuel constituents is modeled by means of thermo-transport theory. Fuel–clad chemical interaction models based on precipitation kinetics were developed for steady-state operation and transients. Finally, a transient intergranular creep-fracture model for the clad, which tracks the nucleation and growth of the cavities at the grain boundaries, was developed for and implemented in the code. Reducing the empiricism in the constitutive models should make it more acceptable to extrapolate FEAST-METAL to new fuel compositions and higher burnup, as envisioned in advanced sodium reactors.FEAST-METAL was benchmarked against the open-literature EBR-II database for steady state and furnace tests (transients). The results show that the code is able to predict important phenomena such as clad strain, fission gas release, clad wastage, clad failure time, axial fuel slug deformation and fuel constituent redistribution, satisfactorily.     

CATHENA: A thermalhydraulic code for CANDU analysis

This paper describes the Canadian algorithm for thermal hydraulic network analysis (CATHENA) transient, thermalhydraulics code developed for the analysis of postulated upset conditions in CANDU®¹ reactors. The core of a CANDU reactor consists of a large number of horizontal pressure tubes containing fuel bundles. As a result of the unique design of the CANDU reactor, the CATHENA thermalhydraulic code has been developed with a number of unique modelling capabilities. The code uses a one-dimensional, two-fluid, nonequilibrium representation of two-phase flow. Some of the unique features of the CATHENA code are the one-step semi-implicit numerical method used and the solid heat transfer modelling capability that allows horizontal fuel bundles to be represented in detail. The code has been used in the design and analysis of CANDU-3, CANDU-6 and CANDU-9 reactors. The code has also been used for the design and analysis of the multiple applied lattice experimental (MAPLE) class of reactors and for the analysis of thermalhydraulic experimental programs conducted by Atomic Energy of Canada Limited (AECL).     

BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis

Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements.     

A new cross-section generation model in the FAST code system and its application to gas- and sodium-cooled Generation IV fast reactors

The cross-section generation scheme employed in the 3D spatial kinetics PARCS code included in the FAST code system being used and developed at the Paul Scherrer Institute (PSI) is currently based on region-wise macroscopic cross-sections for reference conditions and their first-order derivatives with respect to the state variables. Since for some transients, feedback effects may likely not in this way be precisely approximated in their interrelations, this standard method was recently complemented for (hex, z) geometry by a more rigorous cross-section representation scheme. The main idea behind the new approach within the FAST code system is that of preparing sets of microscopic cross-sections for a studied design. The full library consisting of such isotopic tables is generated based on using the ECCO cell code of the code system ERANOS developed by the French Atomic Energy Commission (CEA) for a suitable range of temperatures and background cross-sections (σ₀) on a common grid. In the paper, this σ₀-model is described. In addition, within a detailed verification study needed due to its complexity, it is extensively compared to the standard method for both steady-state and transient conditions. Thereby use is made of current Generation IV fast-spectrum concepts. Reactivity and power evolution indicate overall good agreement of the two methods. Such a good consistence in various transient situations for systems characterized by different neutron spectra gives a large degree of confidence in the correct implementation and suitability of the microscopic cross-section methodology. Therefore, besides for the safety analysis of advanced fast-spectrum core concepts in general, it is foreseen, within the FAST code system, to use the σ₀-model for the assessment of uncertainty propagations in transient calculations in conjunction with the available ERANOS isotopic covariance matrices. The new development makes it also in principle possible to use the PARCS code as a reactor kinetics solver for severe accident analysis, allowing through the use of microscopic cross-sections to account for relocation of the core material during the accident.     

MARGARET: A comprehensive code for the description of fission gas behavior

The relevant phenomena concerning stable-fission gas behavior in nuclear fuels are combined in a single model: MARGARET. This same tool can be used for base irradiations up to high burnup, ramp tests and annealing tests. The representation of intragranular or intergranular bubbles and fabrication pores is highly mechanistic. The partition of fission gas between these cavities and dissolved in the solid permits determination of swelling of the fuel. The released gas is obtained by difference between the created and retained gas in the fuel. The model has been validated against base irradiations, ramps and annealing tests of UO₂ fuel. The article presents the complete equations of the model in the base irradiation condition (Part I), followed by a detailed analysis of the behavior of a fuel irradiated up to 61 GWd/tU, extensively examined after irradiation (Part II). Part III presents the specific additional terms used for the calculation of transient and annealing conditions.     

今日的尤卡山计划和多重屏障系统的新概念

主要论述了美国尤卡山项目的最新进展和比利时工程屏障研究的新进展。在尤卡山项目新进展方面介绍了尤卡山项目修改的近期计划,以及两个环境评价补充报告和两位民主党参议员参加2008年美国总统竞选时对尤卡山项目的态度。还介绍了比利时为加强高放废物地质处置的安全性而开发的多重屏障的新概念。     

PWS: an efficient code system for solving space-independent nuclear reactor dynamics

The reactor kinetics equations are reduced to a differential equation in matrix form convenient for explicit power series solution involving no approximations beyond the usual space-independent assumption. The coefficients of the series have been obtained from a straightforward recurrence relation. Numerical evaluation is performed by PWS (power series solution) code, written in Visual FORTRAN for a personal computer. The results are applied to the step reactivity insertion, ramp input, zigzag input, and oscillatory reactivity changes. When the reactivity is given, including the case in which the feedback reactivity is a function of neutron density, the developed method can provide a straightforward procedure for computing reactor dynamics problems. The solution of this method was compared to some other analytical and numerical solutions of the point reactor kinetics equations; the results proved that the approach is both efficient and accurate to several significant figures.