Validation of coupled codes using VVER plant measurements

A data set of five transients at different VVER type nuclear power plants was collected in order to validate neutron kinetics/thermal hydraulics codes. Two of these transients ‘drop of control rod at nominal power at Bohunice-3’ of VVER-440 type and ‘coast-down of 1 from 3 working MCPs at Kozloduy-6’ of VVER-1000 type, were then utilised for code validation. Eight institutes contributed to the validation with 10 calculations using 5 different combinations of coupled codes. The thermal hydraulic codes were ATHLET, SMABRE and RELAP5 and the neutron kinetic codes DYN3D, HEXTRAN, KIKO3D and BIPR8. The general behaviour of both the transients was quite well calculated with all the codes. Even an elementary modelling of coolant mixing in reactor pressure vessel under asymmetric transients improved correspondence to the measurements. Some differences between the calculations seem to indicate that fuel modelling and treatment of VVER-440 control rods need further consideration. The simultaneous validation interacted with the data collection effort and thus improved its quality. The complexity of data collection systems and sometimes conflicting data, however, called for compromises and interpretation guides that also taught the analysts balanced plant modelling.     

Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code

Incorporation of full three-dimensional models of the reactor core into system thermal–hydraulic transient codes allows better estimation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations to verify and validate the capability of the so-called coupled codes technique. For these purposes appropriate Light Water Reactor (LWR) transient benchmarks based upon programmed transients performed in Nuclear Power Plants (NPP) were recently developed on a higher ‘best-estimate’ level. The reference problem considered in the current framework is a Main Coolant Pump (MCP) switching-on transient in a VVER1000 NPP. This event is characterized by a positive reactivity addition as consequence of the increase of the core flow. In the current study the coupled RELAP5/PARCS code is used to reproduce the considered test. Results of calculation were assessed against experimental data and also through the code-to-code comparison.     

Sampling based uncertainty analysis of station blackout in PSB VVER integral test facility

Best estimate accident analysis with uncertainty evaluation is being encouraged in the present licensing scenarios of nuclear power plants. This paper deals with uncertainty and sensitivity analysis for station blackout in PSB VVER integral test facility under the framework of coordinated research project of IAEA. Nodalization was developed using best estimate system code RELAP5/MOD3.2 and its steady state and transient level qualifications are achieved. Sampling based approaches are used to carry out uncertainty and sensitivity/importance analysis. The objective of the analysis is to get confidence for uncertainty methodology by comparing with the experimental results and extend its applicability to NPPs. Uncertainty analysis is carried out by selecting nine important input parameters with specified ranges and its uniform distributions. A design matrix of 45 × 9 is generated for variations of input parameters with the Latin Hypercube Sampling and 45 code runs were taken. Linear regression was also carried out to quantify the effect of each individual input parameter on output parameters in terms of standard rank regression coefficients. Uncertainty band in output parameters is defined between 95th and 5th percentile value. It is observed that most of the experimental values and code calculated reference values are lying within the uncertainty band. For most of the parameters, width of uncertainty band increases with transient progression time.     

Nuclear data sensitivity/uncertainty analysis for XT-ADS

The XT-ADS, an accelerator-driven system for an experimental demonstration, has been investigated in the framework of IP EUROTRANS FP6 project. In this study, the sensitivity and uncertainty analyses were performed to comprehend the reliability of the XT-ADS neutronic design.     

Benchmark calculations of pressurizer model for Maanshan nuclear power plant using TRACE code

The pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) nuclear power plants. An accurate modeling of the pressurizer is needed to determine the pressure response of the primary coolant system, and thus to successfully simulate overall PWR nuclear power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: (1) turbine trip test from 100% power (Test PAT-50); (2) large-load reduction at 100% power (Test PAT-49); (3) net-load trip at 100% power (Test PAT-51); and (4) net-load trip at 50% power (Test PAT-21). The simulation results show that the predictions of the pressure response are in reasonable agreement with the power plant's start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.     

NURESIM - A European simulation platform for nuclear reactor safety: Multi-scale and multi-physics calculations, sensitivity and uncertainty analysis

Starting in 2005 with the NURESIM Integrated Project (FP6), a European Reference Simulation Platform for Nuclear Reactors called NURESIM is being developed. This development follows a roadmap which is consistent with the SRA (Strategic Research Agenda) of the European SNETP (Sustainable Nuclear Energy Technology Platform). After delivery of two successive versions during the course of the NURESIM project, the numerical simulation platform is presently being developed in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries.NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results.The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between nonconforming meshes).More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more user-friendly.The platform also provides the informatics environment for testing and comparing different codes. For this purpose, it is essential to permit connection of the codes in a standardized way. The standards are being progressively built, concurrently with the process of developing the platform.The NURESIM platform and the individual models, solvers and codes are being validated through challenging applications corresponding to nuclear reactor situations, and including reference calculations, experiments and plant data. Quantitative deterministic and statistical sensitivity and uncertainty analyses tools are also developed and provided through the platform.A Users’ Group of European and non-European countries, including vendors, utilities, TSOs, and additional research organizations (beyond the current partners) has also been established in order to enhance the role of the simulation platform in meeting the needs of the nuclear industry, as applied to current and future nuclear reactors.This presentation summarizes the achievements and ongoing developments of the simulation platform in core physics, thermal-hydraulics, multi-physics, uncertainties and code integration.     

Dimensionally adaptive dynamic switching and adjoint sensitivity analysis: new features of the RELAP5/PANBOX/COBRA code system for reactor safety transients

This paper highlights two novel features that have been implemented into the coupled RELAP5/PANBOX2/COBRA3 (R/P/C) code system. On the one hand, the R/P/C code system has been extended to include a dimensionally adaptive algorithm that uses the underlying physical phenomena to switch dynamically between three-dimensional (3D), one-dimensional (1D), and point-kinetics models, thereby reducing computational times up to a factor of five while preserving the accuracy, within user-defined error criteria, of the 3D reference calculation. On the other hand, the R/P/C system has also been extended to include the Adjoint Sensitivity Analysis Procedure (ASAP) for the RELAP5/MOD3.2 two-fluid model with non-condensables, thus enabling the efficient calculation of local sensitivities of RELAP5 results to various parameters in the RELAP5 code.     

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.     

Exact nuclear data uncertainty propagation for fusion neutronics calculations

Recently, we have presented an exact method (now called “Total Monte Carlo”) to propagate uncertainties of fundamental nuclear physics experiments, models and parameters to different types of criticality-safety benchmarks. We now show that such exact uncertainty calculations are directly relevant to the optimal and safe design of fusion reactors by applying this methodology to a series of fusion shielding benchmarks, namely those connected to the Oktavian, Fusion Neutronics Source and LLNL Pulsed Sphere experiments. Uncertainties on neutron and gamma leakage fluxes for 13 shielding benchmarks are obtained, in the mass range from ^natMg to ^natW. Uncertainties for cross-sections, angular distributions, single- and double-differential emission spectra, and gamma-ray production cross-sections are considered in this uncertainty propagation scheme.     

Post-test calculation and uncertainty analysis of the experiment QUENCH-07 with the system code ATHLET-CD

In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam–argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B₄C oxidation do not affect significantly the total calculated hydrogen release rates.     

Development and validation of the code COUPLE3.0 for the coupled analysis of neutron transport and burnup in ADS

The analysis of the fuel depletion behavior is critical for maintaining the safety of accelerator-driven subcritical systems(ADSs). The code COUPLE2.0 coupling 3-D neutron transport and point burnup calculation was developed by Tsinghua University. A Monte Carlo method is used for the neutron transport analysis, and the burnup calculation is based on a deterministic method. The code can be used for the analysis of targets coupled with a reactor in ADSs. In response to additional ADS analysis requirements at the Institute of Modern Physics at the Chinese Academy of Sciences, the COUPLE3.0 version was developed to include the new functions of(1) a module for the calculation of proton irradiation for the analysis of cumulative behavior using the residual radionuclide operating history,(2) a fixed-flux radiation module for hazard assessment and analysis of the burnable poison, and(3) a module for multi-kernel parallel calculation, which improves the radionuclide replacement for the burnup analysis to balance the precision level and computational efficiency of the program. This paper introduces thevalidation of the COUPLE3.0 code using a fast reactor benchmark and ADS benchmark calculations. Moreover,the proton irradiation module was verified by a comparison with the analytic method of calculating the210 Po accumulation results. The results demonstrate that COUPLE3.0 is suitable for the analysis of neutron transport and the burnup of nuclides for ADSs.     

On the implementation of disturbance analysis systems for detection and diagnosis of safety-related transients in a nuclear power plant

Considerable interest has developed in the last few years in the possible applications of a computerized disturbance analysis system (DAS) to nuclear power plant fault detection and diagnosis. The present paper, after introducing some general aspects of the DAS methodology, focuses specifically on the problems associated with its application to safety related malfunctions and transients. An approach called ‘significant event analysis’ is proposed for possible application when considerable diagnostic depth is deemed necessary from the DAS. A fairly detailed application of this approach to a particular class of disturbances potentially associated with power supply systems is then described. Finally, consideration is given to the relation between the specific types of failures actually recorded in electrical systems and the potential benefits from implementation of a DAS in those systems.     

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.     

Uncertainty and sensitivity analyses of the Kozloduy pump trip test using coupled thermal–hydraulic 3D kinetics code

The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulation and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper, a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data. Deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect. Further investigations through the use of two “antagonist” uncertainty GRS and the CIAU methods, were considered in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and further enhancements for both methods are needed.     

Control strategies for transients of hydrogen production plant in VHTR cogeneration systems

Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design.In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine-Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant.It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.