COCOSYS: Status of development and validation of the German containment code system

For the simulation of severe accident propagation in containments of nuclear power plants it is necessary to assess the efficiency of severe accident measures under conditions as realistic as possible. Therefore the German containment code system COCOSYS is under development and validation at GRS. The main objective is to provide a code system on the basis of mostly mechanistic models for the comprehensive simulation of all relevant processes and plant states during severe accidents in the containment of light water reactors covering the design basis accidents, too. COCOSYS is being used for the identification of possible deficits in plant safety, qualification of the safety reserves of the entire system, assessment of damage-limiting or mitigating accident management measures, support of integral codes in PSA level 2 studies and safety evaluation of new plants.COCOSYS is composed of three main modules, which are separate executable files. These modules are covering thermal hydraulics including hydrogen combustion, fission products mainly aerosols and iodine behaviour, and corium behaviour with molten corium concrete interaction. The communication between these modules is realized via PVM (parallel virtual machine).COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently subject of the just started OECD-THAI project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project.For example COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. At present COCOSYS is in use for the licensing process of the new Finnish EPR plant on the industrial side.Improvements and model extensions like pyrolysis processes, direct containment heating and the combined use with CFD models are just ongoing.     

Modeling and code development for concrete structures: An overview

The present status of efforts to model concrete behavior under projected extreme loadings suitable for finite element formulation is presented. The difficulties in modeling introduced by problems associated with quality control of the concrete mix and dependency of test results on a number of variables are pointed out. A variety of modeling schemes which attempt to take into account such peculiarities of concrete behavior are discussed, and advantages and shortcomings of each scheme are mentioned. Future analytical and experimental research needs are indicated, especially with regard to reinforced concrete element modeling. The present paper constitutes a state-of-the-art report on computer modeling of plain and reinforced concrete behavior and contains appropriate recommendations from the authors.     

A study of the stress intensity factors for single or multiple cracks in thick-walled cylinders

The variation of stress intensity factors of a single semi-elliptical crack and multiple semi-elliptical cracks which are radial symmetric or unsymmetric array in an internal pressurized thick-walled cylinder is studied by use of the “frozen-stress” photo-elastic method. The method of determining mixed-mode stress intensity factors K_I and K_II is given. By means of experimental results and the relative results of other authors, the approximate expression for evaluating stress intensity factors of straight border, semi-circular, semi-elliptical internal surface cracks in thick-walled cylinders are presented.     

Development of a thermal–hydraulic system code for simulators based on RELAP5 code

A thermal–hydraulic system code for simulators, RELAPSIM, was developed at NSSE based on RELAP5. The development procedure consists of three major parts. Firstly, time control function was added into the code to meet real-time calculation needs. Secondly, controlled dynamic data communication was improved, so that thermal–hydraulic parameters can be easily modified for further applications. Finally, functions controlling the computation procedure were embedded to achieve a full capability to simulate multiple operations, such as start-up, shutting down or freeze. This paper describes the main features of the new code. The results of code assessment and code application are presented and discussed.     

Survey of experiments and code development for the passive residual heat removal system of PWR in China

Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.     

The development of a stress analysis code for nuclear graphite components in gas-cooled reactors

Most of the UK nuclear power reactors are gas-cooled and graphite moderated. As well as acting as a moderator the graphite also acts as a structural component providing channels for the coolant gas and control rods. For this reason the structural integrity assessments of nuclear graphite components is an essential element of reactor design. In order to perform graphite component stress analysis, the definition of the constitutive equation relating stress and strain for irradiated graphite is required. Apart from the usual elastic and thermal strains, irradiated graphite components are subject to additional strains due to fast neutron irradiation and radiolytic oxidation. In this paper a material model for nuclear graphite is presented along with an example of a stress analysis of a nuclear graphite moderator brick subject to both fast neutron irradiation and radiolytic oxidation.     

FAST: An advanced code system for fast reactor transient analysis

One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems.     

The development and validation of a natural circulation analysis code for marine reactors

A code PNCMC (Program for Natural Circulation under Motion Conditions) has been developed for natural circulation simulation of marine reactors. The code is based on one-dimensional two-fluid model in noninertial frame of reference. The body force term in the momentum equation is considered as a time dependent function, which consists of gravity and inertial force induced by three-dimensional ship motion. Staggered mesh, finite volume method, semi-implicit first order upwind scheme and Successive Over Relaxation (SOR) method are used to discretize and solve two-phase mass, momentum and energy equations. Single-phase natural circulation experiments under rolling condition performed in Institute of nuclear and new energy technology of Tsinghua University and two-phase natural circulation experiments under rolling condition performed by Tan and colleagues are used to validate PNCMC. The validation results indicate that PNCMC is capable to investigate the single-phase and two-phase natural circulation under rolling motion.     

GOAT: a simulation code for high-intensity beams

A simulation code, GOAT, is developed to simulate single-bunch intensity-dependent effects and their interplay in the proton ring of the Electron-Ion Collider in China(EicC) project. GOAT is a scalable and portable macroparticle tracking code written in Python and coded by object-oriented programming technology. It allows for transverse and longitudinal tracking,including impedance, space charge effect, electron cloud effect, and beam-beam interaction. In this paper, physical models and numerica...     

New code for digital imaging system for track measurements

A code for the numerical analysis of images from a CCD camera, including correction of acquisition system defects, has been written and applied to the visualization of α-particle tracks in the solid state nuclear track detectors CR39 and LR115. A standard mask having different diameter holes permitted the calibration of the imaging system. Comparison of manual scanning results with those given by the automatic system are presented for detectors exposed to α-particles of ²⁴¹Am source. As application, the system was used to measure specific activities of ²³⁸U in the Moroccan phosphates and their phosphogypsums samples by contact autoradiography.     

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.     

The development and validation of a thermal hydraulic code in rolling motion

An advanced thermal hydraulic code is established on the basis of RELAP5/MOD3.3 code for the investigation of the thermal hydraulic behavior of nuclear power systems. The RELAP5 code is modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The experimental data are used to validate the theoretical models and calculation results. It is shown that the advanced flow and heat transfer models could correctly predict the frictional resistance and heat transfer coefficients in rolling motion. The thermal hydraulic code is used to simulate the operation of a natural circulation system in rolling motion. The calculation results are in good agreement with experimental data. The relative discrepancies between calculation results and experimental data are less than 5%.     

Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1

A multi-dimensional thermal-hydraulic system code MARS has been developed by consolidating and restructuring the RELAP5/MOD3.2.1.2 and COBRA-TF codes. The two codes were adopted to take advantage of the very general, versatile features of RELAP5 and the realistic three-dimensional hydrodynamic module of COBRA-TF. In the course of code development, major features of each code were consolidated into a single code first. The resulting source programs were rewritten in standard fortran 90, and then were restructured using modular data structures based on “derived type variables” and a new “dynamic memory allocation” scheme. In addition, the Windows graphics features were implemented for user friendliness. This paper presents the developmental activities up to mars version 1.3.1 including the code consolidation, the code restructuring and modernization, and the results of the developmental assessment.     

The development of a transient version of the HOTROD computer code

The full length fuel rod steady state performance code HOTROD has been developed to predict fuel performance during a transient. This paper explains the theory used to calculate the transient fuel temperatures using the Crank-Nicolson method. Transient HOTROD predictions of SGHWR and PWR axial clad temperatures during a loss-of-coolant accident are given and are compared with those predicted by other codes. The code is being further developed to model Zircaloy clad ballooning and the creep equations coupled to the heat transfer equation derived in this paper.     

An improved numerical scheme with the fully-implicit two-fluid model for a fast-running system code

A new computational method is implemented in the FISA-2 (Fully-Implicit Safety Analysis-2) code to simulate the thermal-hydraulic response to hypothetical accidents in nuclear power plants. The basic field equations of FISA-2 consist of the mixture continuity equation, void propagation equation, two phasic momentum equations, and two phasic energy equations. The fully-implicit scheme is used to eliminate a time step limitation and the computation time per time step is minimized as much as possible by reducing the matrix size to be solved. The phasic energy equations written in the nonconservation form are solved after they are set up to be decoupled from other field equations. The void propagation equation is solved to obtain the void fraction. Spatial acceleration terms in the phasic momentum equations are manipulated with the phasic continuity equations so that pseudo-phasic mass flux may be expressed in terms of pressure only. Putting the pseudo-phasic mass flux into the mixture continuity equation, we obtain linear equations with pressure variables only as unknowns. By solving the linear equations, pressures at all the nodes are obtained and in turn other variables are obtained by back-substitution. The above procedure is performed until the convergence criterion is satisfied. Reasonable accuracy and no stability limitation with fast-running are confirmed by comparing results from FISA-2 with experimental data and results from other codes.     

CATHARE-3: A new system code for thermal-hydraulics in the context of the NEPTUNE project

After a thorough analysis of the industrial needs and of the limitations of current simulation tools, EDF and CEA (Commissariat à l’Energie Atomique) launched the NEPTUNE Project in 2001 (see Guelfi et al., 2007) with the support of AREVA-NP and IRSN. The NEPTUNE activities include software development, research in physical modeling and numerical methods, development of advanced instrumentation techniques and new experimental programs. Four different simulation scales were addressed including DNS (Direct Numerical Simulation), CFD in open medium (Computational Fluid Dynamics), component (subchannel-type analysis) and system (reactor modeling) scales.In 2006 CEA, EDF, AREVA-NP and IRSN defined the strategy for the system scale of NEPTUNE and the CATHARE-3 development was launched. The main objectives are:

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advanced physical modeling of two-phases flows, mainly by using multi-field and turbulence models,

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improved 3D modeling by the use of fine and non conforming structured meshes,

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generalized coupling capabilities with other thermal-hydraulic scales and with other disciplines (core physics, structural mechanics, …),

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extension of the applicability to new Gen IV reactors (Sodium Cooled Fast Breeder Reactors, Gas Cooled Reactors, Supercritical Light Water Reactors),

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a true object-oriented code architecture.

At the same time CATHARE-3 is in continuity with the CATHARE-2 code which is the current industrial version of CATHARE and internationally used for nuclear power plant safety analysis, in simulators and in coupled simulation tools. The road map of these two codes will allow a smooth transition from CATHARE-2 to CATHARE-3 for all users.This paper gives an overview of the choices made for the development of CATHARE-3 including new physical models, validation strategy and experimental programs, numerical improvements, enhanced coupling capability and software architecture evolution. The current status of the project as well as the overall schedule will be presented.     

Specific purpose Monte Carlo code development for nuclear well logging applications

The authors have been developing specific purpose Monte Carlo simulation programs for the suite of nuclear well logging devices that are in present use. To date codes called McPNL and McDNL have been developed and tested for the pulsed neutron lifetime and dual-spaced or compensated neutron porosity logs, respectively. Another code called McLDL is presently under development for the gamma-ray litho-density logs. These codes are discussed and results are presented as to their construction, advantages, and uses as compared to the general purpose codes MCNP and McBEND. Computational benchmark problems are specified for all three logs for future quantitative comparisons.     

Development of a real-time radon monitoring system for simultaneousmeasurements in multiple sites

A real-time radon monitoring system that can simultaneously measure radon concentrations in multiple sites was developed and tested. The system consists of maximum of four radon detectors, optical fiber cables and a data acquisition personal computer. The radon detector uses a plastic scintillation counter that collects radon daughters in the chamber electrostatically. The applied voltage on the photocathode for the photomultiplier tube (PMT) acts as an electrode for radon daughters. The thickness of the plastic scintillator was thin, 50 μm, so as to minimize the background counts due to the environmental gamma rays or beta particles. The energy discriminated signals from the radon detectors are fed to the data acquisition personal computer via optical fiber cables. The system made it possible to measure the radon concentrations in multiple sites simultaneously     

Development of core fuel management code system for WWER-type reactors

1 Introduction Over the past decades, although many in-core fuel management code systems for PWRs with square fuel assemblies have been developed, there are only a few codes for the cores with hexagonal assemblies (such as Russian pressurized water type WWER reac- tors). The Tianwan Nuclear Power Station in Jiangsu Province, China, is imported from Russia, which adopts the WWER-1000 reactor, and will be put into operation; therefore, the research of core fuel man- agement for WWER-typ…