The evaluation of TRACE/PARCS model for BWR-4 nuclear power plant by startup test transient analyses

Generally, the thermal hydraulic (TH) codes need the results of Neutron Kinetics (NK) codes providing the reactivity properties to calculate neutron flux. Then the TH codes perform the safety analyses obtaining the responses of pressure, temperature, or water level. Two kinds of different codes calculate different physical behaviors sequentially and separately. Simultaneously computing thermal hydraulic and neutron kinetics behaviors can enhance the accuracy of the analysis. Hence, it is crucial to develop the TH-NK coupled model. This study presents the capability of the TH-NK coupled model, developed by TRACE (TRAC/RELAP Advanced Computational Engine) and PARCS (Purdue Advanced Reactor Core Simulator), for the BWR-4 nuclear power plant. The establishment of the TRACE/PARCS model presented the nodal and component modeling methodologies. This model was used to simulate two startup tests of high power level system transients. Principal system responses, calculated by the TRACE/PARCS model, were compared with the measured data in startup tests and the results of the point kinetic calculation of the TRACE (TRACE/PK) to evaluate the model. The evaluation shows that the TRACE/PARCS model can simulate the interaction between thermal hydraulic and neutron kinetics phenomena and predict the transients suitably. Through the comparison, the TRACE/PARCS model can be confident doing the analyses of normal and abnormal operational transients to predict the transient responses.

UNCERTAINTY ANALYSIS TO C5G7-TD BENCHMARK BASED ON THE COST METHOD

A method of Covariance-Oriented Sample Transformation (COST) has been proposed in our previous work to provide the converged uncertainty analysis results with a minimal sample size. The transient calculation of nuclear reactor is a key part of the reactor-physics simulation, so the accuracy and confidence of the neutron kinetics results have attracted much attention. In this paper, the Uncertainty Quantification (UQ) function of the high fidelity neutronics code NECP-X has been developed based on our home-developed uncertainty analysis code UNICORN, building a platform for the UQ of the transient calculation. Furthermore, the well-known space-time heterogeneous neutron kinetics benchmark C5G7 and its uncertainty propagation from the nuclear data to the interested key parameters of the core have been investigated. To address the problem of “the curse of dimensionality” caused by the large number of input parameters, the COST method has been applied to generate multivariate normal-distribution samples in uncertainty analysis. As a result, the law of the assembly/pin normalized power and their uncertainty with respect to time after introducing an instantaneous perturbation has been obtained. From the numerical results, it can be observed that the maximum relative uncertainties for the assembly normalized power can up to be about 1.65% and the value for the pin-wise power distributions can be about 2.71%.

DEVELOPMENT AND VERIFICATION OF T-TRACE/PANTHER COUPLED CODE

Multi-physics coupled simulations have become increasingly important during the last two decades being one of the major field of application in the nuclear technology. The nuclear reactors themselves are complex systems whose responses are driven by interactions between neutron kinetics, thermal-hydraulics, heat transfer, mechanics and chemistry. Probably, in a nuclear system, the most complex and important feedback effect takes place between the core neutron kinetics and thermal-hydraulics. The development of coupled thermal-hydraulic -neutron kinetics codes is a recurrent field of research for the nuclear industry. This contribution, developed in the Consortium for Nuclear Power (CNP) framework, has the objective of develop a dynamic coupling, using TCP/IP based socket communication, between the thermal-hydraulic system code T-TRACE, Tractebel-ENGIE version of the latest US NRC TRACE release, and the multi-group 3-D nodal diffusion and core physics code PANTHER, developed and maintained by EDF Energy (UK). As a first step of the development, a fully temporally explicit coupling scheme has been developed between TRACE and PANTHER based on a boundary conditions exchange at the core level at each temporal iteration. The OECD TMI MSLB benchmark has been selected as verification scenario for testing the ongoing developing T-TRACE/PANTHER coupled code. The developed coupled code is benchmarked code-to-code against TRACE/PARCS and T-RELAP5/PANTHER.

IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE

The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes. An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap. In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).

Detailed Simulation of the Nominal Flow and Temperature Conditions in a Pre-Konvoi PWR Using Coupled CFD and Neutron Kinetics

The aim of the numerical study was the detection of possible vortices in the upper part of the core of a Pre-Konvoi Pressurized Water Reactor (PWR) which could lead to temperature cycling. In addition, the practical application of this Computational Fluid Dynamic (CFD) simulation exists in the full 3D analysis of the coolant flow behavior in the reactor pressure vessel of a nuclear PWR. It also helps to improve the design of future reactor types. Therefore, a CFD simulation of the flow conditions was carried out based on a complex 3D model. The geometry of the model includes the entire Reactor Pressure Vessel (RPV) plus all relevant internals. The core is modelled using the porous body approach, the different pressure losses along and transverse to the main flow direction were considered. The spacer-grid levels were taken into account to the extent that in these areas no cross-flow is possible. The calculation was carried out for nominal operating conditions, i.e., for full load operation. Furthermore, a prototypical End of Cycle (EOC) power distribution was assumed. For this, a power distribution was applied as obtained from a stationary full-core calculation with the 3D neutron kinetics code DYN3D. In order to be able to adequately reproduce flow vortexes, the calculation was performed transiently with suitable Detached Eddy Simulations (DES) turbulence models. The calculation showed fluctuating transverse flow in the upper part of the core, starting at the 8th spacer grid but also revealed that no large dominant vortices exists in this region. It seems that the core acts as a rectifier attenuating large-scale vortices. The analyses included several spacer grid levels in the core and showed that in some areas of the core cross-section an upward increasingly directed transversal flow to the outlet nozzle occurs. In other areas of the core cross-section, on the other hand, there is nearly any cross-flow. However, the following limitations of the model apply: In the model all fuel elements are treated identical and cross flows due to different axial pressure losses for different FA types cannot be displayed. The complex structure of the FAs (eg. flow vanes in spacer grids) could also influence the formation of large-scale vortices. Also, the possible influence of two-phase flows was not considered.

Proportional-Integral Disturbance Observer of Nuclear Reactors

A proportional-integral disturbance observer (PI-DO) for monitoring nuclear reactors is newly proposed, which is driven by the measurements of neutron flux and coolant temperature at reactor inlet as well as their integrations. This PI-DO provides a globally asymptotic estimation with a bounded steady-state error for the reactor key process variables as well as the total disturbances in channels of the neutron kinetics and primary coolant thermal-hydraulics. Moreover, the PI-DO is applied to reconstruct the unmeasurable state variables and total disturbances of a nuclear heating reactor (NHR). Numerical simulation results not only verify the theoretic analysis but also show both the satisfactory performance and the influence of observer parameters.