Transient analysis code

A plant-wide transient analysis code is being developed for the analysis of KALIMER s inherent safety and for the assistance in the development of design, where new design features will frequently demand not just new data but new models. Transient and safety analysis code SSC-K is under development based upon the SSC-L code which was developed by BNL for the analysis of loop type... 

Preliminary safety analyses have been performed on the former design (1,100 MWt) with the thermal-hydraulic transient analysis code RETRAN-2 to understand the basic characteristics of SPWR. Examples of the events analyzed are as follows ... 

Validation of the RELAP5-3D input modeling procedure is required against test data or a validated transient analysis code. 

The plant control system has been designed in a similar way to that of BWRs [36-39]. It is shown in Fig. 1.14. The plant transient analysis code SPRAT-DOWN was developed and used in the design work. The node-junction model, shown in Fig. 1.15, contains the RPV, the control rods (CRs), the main feedwater pumps, the turbine control valves, the main feedwater lines, and the main steam lines. The characteristics of the turbine control valves and the changes of the feedwater flow rate according to the core pressure are given in the calculation. 

Fig. 1.15 Node junction model of transient analysis code SPRAT-DOWN...

From the beginning of the conceptual study on supercritical water cooled reactors, several plant transient analysis codes have been developed, modified, and applied to them [1-9]. The general name of these codes is Supercritical Pressure Reactor Accident and Transient analysis code (SPRAT). SPRAT mainly calculates mass and energy conservations, fuel rod heat conduction, and point kinetics. The relation among these calculations is shown in Fig. 4.1. SPRAT can deal with flow, pressure, and reactivity induced transients and accidents at supercritical pressure. The flow chart is shown in Fig. 4.2. 

The system pressurization and line switching are analyzed using a system transient analysis code. Since the startup system and procedures are the same between the Super LWR and Super FR, the Super FR is analyzed as an example. The Super FR has higher power density and smaller heat capacity compared to the Super LWR so that the analysis of the Super FR covers that of the Super LWR from the viewpoint of the criteria of the MCST and the rate of increase in the coolant temperature. 

The plant transient analysis code SPRAT-DOWN, described in Sect. 4.2, is extended for the analyses of abnormal transients and accidents. The calculation model is shown in Fig. 6.12. The models of the equipment, i.e., the AFS, SRV, MSIVs, and turbine bypass valves, are added for safety analyses. A hot chaimel, where the linear heat generation late and maximum cladding surface temperature are the highest in the core, is modeled as well as the average channel in order to calculate the highest values of cladding temperature and pellet enthalpy. The flow chart is shown in Fig. 6.13. 

The plant transient analysis code for the Super FR is called SPRAT-F. It is based on the 1-D node junction model with radial heat transfer and point kinetics models such as SPRAT-DOWN for the Super LWR (see Chaps. 4 and 6). The nodalization is shown in Fig. 7.67. The models used in SPRAT-F are the same as those in SPRAT-DOWN. The turbine control valve regulates the main steam pressure by changing the main steam flow rate as in BWRs and the Super LWR. The relation between its stroke and the steam flow rate is the same as that of the Super LWR (Fig. 4.4). The relation between the core pressure and the feedwater flow rate (with constant pump speed) is also the same as that in the Super LWR (Fig. 4.5). 

The plant transient analysis code SPRAT-F, introduced in Sect. 7.9, is used to calculate the flow distribution and the MCST in the three hot channels with respect to the core power and feedwater flow rate. The nodahzation is shown again in Fig. 7.81 [32]. Since the core power and feedwater flow rate are raised very slowly at the power raising phase, the reactor can be practically treated as in a steady state... 

The first line of any SPICE netlist is the title line. It is used for documentation purposes only. The next few lines usually tell SPICE which analysis will be performed and what the bounds of that analysis will be. For example, we may be requesting a time domain analysis of a circuit (called a transient analysis). The information as to how long the waveform is and what increments and what section of it are of interest is defined in this section of the code. SPICE netlists generally have one function, command, or element per line (Fig. 2.1). Also defined upfront are global constants, subcircuits (models) used repeatedly in the main circuit, and instructions on which nodes are of interest in the final solution, though this structure is not mandatory. 

The Sandia code HECTR (Hydrogen Event Containment Transient Response) is a lumped parameter analysis code for modeling the containment atmosphere under accident conditions involving release, transport, and also combustion of hydrogen [37]. It can handle saturated and supe eated conditions and it covers both short-term transients and long-term convection up to several hundred days. 

In order to predict the T-H-M response of the bentonite, a coupled T-H-M transient analysis was performed with the Finite Element Code FRACON. The governing equations incorporated in the FRACON code were derived from an extension of Biot s (1941) theory of poro-elasticity to include the T-H-M behaviour of the unsaturated FEBEX bentonite. The model formulation(Nguyen, Selvadurai and Armand, 2003) resulted in three governing equations where the primary unknowns are temperature, the displacement vector and the pore fluid pressure, as follows ... 

Sequence of transient analysis using the FRACON code... 

Safety evaluation studies have been conducted for confuming the physical phenomena and integrity of the fuel subassemblies, the core internal structures and the heat transport systems during the normal operation, scram transients and the early stage of postulated accidents. On this account, thermohydraulic experiments related to the decay heat removal by natural circulation have been carried out, and the development and validation of the thermohydraulic safety analysis codes is also in progress. 

TRAC (Transient Reactor Analysis Code)-PWR is an advanced best-estimate systems code designed primarily for analysis of large-break LOCAs (LB-LOCAs) in PWRs, although its versatility allows for the analysis of a wide range of scenarios. This code was developed at Los Alamos National Laboratory under sponsorship of the US Nuclear Regulatory Commission (USNRC). 

Because it is based on well-established LWR technology and implements benchmarked thermal hydraulic safety analysis codes, and because each module is a relatively small reactor, it is possible that all of the information required for design certification may be obtained using a full-scale demonstration of a single module. Of significant interest for design certification would be the impact of neutronic feedback on flow stability, particularly during plant transients. 

The computer code GOBLIN/WATGAS was used for the transient analysis of experiments on level measurement lines with dissolved non-condensables and with boiling effects. The test represents normal cooldown conditions and also depressurization during accidents. 

VIPRE-01 is a finite-volume sub-channel analysis code enable of three-dimensional modelling of reactor cores and other similar geometries in steady-state and transient conditions. VIPRE-01 calculates the detailed steady-state and operational transient core flow distributions, coolant conditions, fuel rod temperature and departure from nucleate boiling ratio (DNBR). 

The turbine trip fault is analysed using LOFTRAN. LOFTRAN computes pertinent plant variables, including the nuclear power transient, the flow coast-down, the primary system pressure transient, and the primary coolant temperature transient. FACTRAN code is then used to calculate the heat flux based on the LOFTRAN analysis results for nuclear power and flow. Finally, VIPRE-01 is used to calculate the DNBR during the transient, using the heat flux from FACTRAN and the flow Ifom LOFTRAN. 

The analysis has been conducted using a transient reactor analysis code (TRAC) coupled with the heat transfer models for steam condensation with the presence of a non-condensable gas in the IC condenser tube. The non-condensable gas, which initially fills the PCV space, will be absorbed into the IC condenser tubes through the broken feedwater pipe. 

Tools for transient thermal-hydraulic and fuel performance analysis. Code development in progress DSNP, MATRA Multi-channel Analyser for steady state and Transients in Rod Arrays... 

There exist powerful simulation tools such as the EMTP [35]. These tools, however, involve a number of complex assumptions and application limits that are not easily understood by the user, and often lead to incorrect results. Quite often, a simulation result is not correct due to the user s misunderstanding of the application limits related to the assumptions of the tools. The best way to avoid this type of incorrect simulation is to develop a custom simulation tool. For this purpose, the FD method of transient simulations is recommended, because the method is entirely based on the theory explained in Section 2.5, and requires only numerical transformation of a frequency response into a time response using the inverse Fourier/Laplace transform [2,6,36, 37, 38, 39, 40, 41-42]. The theory of a distributed parameter circuit, transient analysis in a lumped parameter circuit, and the Fourier/Laplace transform are included in undergraduate course curricula in the electrical engineering department of most universities throughout the world. This section explains how to develop a computer code of the FD transient simulations. 

The safety analysis of the TMSR-SF has also drawn much attention. Three types of transient conditions including ULOF, UOC, and UTOP were examined on the TMSR-SF by an FHR safety analysis code named the FHR Safety Analysis Code (FSAC Xiao et al., 2014). The station blackout anticipated transient without scram (SBO-ATWS) accident was analyzed by the modified RELAP5/MOD 4.0 code with the responses of the passive residual heat removal (PRHR) system (Jiao et al., 2015). 

Xiao, Y., Hu, L., Qiu, S., Zhang, D., Su, G., Tian, W., 2014. Development of a thermal-hydraulic analysis code and transient analysis for a FHTR. In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers pp. V005T016A005. 

The methods used include operating experience and event data, geotectonic historical records, coupled neutronic-thermal-hydraulic transient performance analysis codes. Computer Aided Design-Computer Aided Engineering (CAD-CAE) systems, structural finite element methods, materials stress analysis. Human Reliability Analysis, risk assessments and PRAs. 

The natural recourse for this situation is to use numerical solution methods for both steady-state and transient analyses. The systems-analysis codes that are used for design and safety studies can all handle the coupled NCL case. At the same time, local special- and general-purpose natural-circulation models, methods, and codes are being developed for applications to Gen IV reactor concepts. 

TRAC-B Transient Reactor Analysis Code-Boiling water reactor... 

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