Transient Analysis Code Research Articles

COMPARISON OF THE DECAY HEAT REMOVAL SYSTEMS IN THE KALIMER-600 AND DSFR

A sodium-cooled demonstration fast reactor with the KALIMER-600 as a reference plant is under design by KAERI. The safety grade decay heat removal system (DHRS), which is important to mitigate design basis accidents, was changed in the reactor design. A loss of heat sink and a vessel leak in design basis accidents were simulated using the MARS-LMR system transient analysis code on two plant systems. In the analyses, the DHRS of KALIMER-600 had a weakness due to elevation of the overflow path for the DHRS operation, while it was proved that the DHRS of the demonstration reactor had superior heat transfer characteristics due to the simplified heat transfer mechanism.

Study on Seismic Response Analysis of 3-D Structure

This paper describes the parallel finite element analysis of large scale 3-D structural dynamic problems based on the Domain Decomposition Method (DDM) with preconditioner using Balancing Domain Decomposition (BDD) for massively parallel processors. This domain decomposition has been implemented within a finite element code for linear implicit transient dynamic analysis. Time integration is performed using Newmark-βconstant average acceleration method. The parallel finite element code uses an MPI-based message passing approach to provide portable parallel execution on shared, distributed and distributed shared memory computers. 3-D structural dynamic problems have been conducted on supercomputer to evaluate the performance of the implicit parallel linear finite element code. Results indicate that the proposed methods are highly adaptive for parallel computing and superior in performance.

Validation and Application of the Thermal Hydraulic System Code TRACE for Analysis of BWR Transients

The Karlsruhe Institute of Technology (KIT) is participating on (Code Applications and Maintenance Program) CAMP of the US Nuclear Regulatory Commission (NRC) to validate TRACE code for LWR transient analysis. The application of TRACE for the safety assessment of BWR requires a throughout verification and validation using experimental data from separate effect and integral tests but also using plant data. The validation process is normally focused on safety-relevant phenomena for example, pressure drop, void fraction, heat transfer, and critical power models. The purpose of this paper is to validate selected BWR-relevant TRACE-models using both data of bundle tests such as the (Boiling Water Reactor Full-Size Fine-Mesh Bundle Test) BFBT and plant data recorded during a turbine trip event (TUSA) occurred in a Type-72 German BWR plant. For the validation, TRACE models of the BFBT bundle and of the BWR plant were developed. The performed investigations have shown that the TRACE code is appropriate to describe main BWR-safety-relevant phenomena (pressure drop, void fraction, and critical power) with acceptable accuracy. The comparison of the predicted global BWR plant parameters for the TUSA event with the measured plant data indicates that the code predictions are following the main trends of the measured parameters such as dome pressure and reactor power.

The Analysis of PWR SBO Accident with RELAP5 Based on Linux

RELAP5 is a relatively advanced light water reactor transient hydraulic and thermal analysis code, and it owns the signality of the safe-operating of nuclear reactor system when the safety analysis and operating simulation of the system was done with RELAP5. The RELAP5 operating mode based on Linux operating system was presented in this paper, utilizing Linux operating system's powerful document processing capabilities to deal with the output file of the RELAP5 for the valid data directly, and taking advantage of the system's programmable capabilities to improve the drawing functions of RELAP5. After the operating in Linux system, the precision of the calculating results is guaranteed and the period of the computing is shortened. During the work, for PWR Station Blackout (SBO) accident, the computing with RELAP5 based on Linux and Windows was respectively made. Through the comparison and analysis of the accident response curve of the main parameters such as power of nuclear reactor, average temperature and pressure of primary loop, it shows the operating analysis of nuclear reactor system is safe and reliable with RELAP5 based on Linux.

Development of a thermal–hydraulic safety analysis code RETAC for AP1000

In the present study, a thermal–hydraulic safety analysis code for AP1000 named RETAC (REactor Transient Analysis Code) has been developed using FORTRAN 90 language. A point reactor neutron kinetics model with six groups of delayed neutrons was adopted to describe the core thermal power transient. A distributed parameter model with two-phase drift flux model was used in the U-tube steam generator simulation. In the pressurizer simulation, the RETAC code was equipped with three-region non-equilibrium model and multi-region non-equilibrium model respectively. Similar to current large commercial codes such as RELAP5 and RETRAN series, the four-quadrant analogy curves were adopted for the solution of the transient behaviors of the main coolant pumps. In this paper, a new and reasonable model for the passive residual heat removal heat exchanger (PRHR HX) was proposed based on basic equations of mass, momentum and energy. Gear method and Adams predictor-corrector method were adopted alternately for a better solution to ill-condition differential equations corresponding to detail processes. The PRHR HX inadvertent operation accident and the ADS (automatic depressurization system) inadvertent operation accident were chosen in the transient accident analysis. Furthermore, the simulation results obtained by RETAC were compared with that by Westinghouse-developed LOFTRAN code. The comparison results showed a good agreement and thus proved the accuracy and reliability of the RETAC code. With the adoption of modular programming techniques, the RETAC code can be easily modified and applied to higher power passive safety reactors in the future.

Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available.Currently, at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the DYN3D code is being extended and adopted for the application to block-type High Temperature Gas-Cooled Reactors (HTGRs). In this paper, we give an overview of the latest developments of DYN3D concerning block-type HTGR.The simplified P3 (SP3) transport approximation is implemented into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method previously implemented into DYN3D for square fuel element geometry of LWR is being extended for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes to be able to perform a systematic mesh refinement.One of the main challenges in cross section generation for the HTGR core calculations is the treatment of the so-called “double heterogeneity”. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. The main steps of the RPT method are described. The use of the method for the cross section generation of a simplified HTGR core including its verification is presented.A 3D heat conduction module coupled with a channel-type coolant flow model is implemented to take the temperature reactivity feedback to neutronics physically correctly into account. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Development of TACOS code for loss of flow accident analysis of SCWR with mixed spectrum core

TACOS(Transient Analysis Code Of SCWR) code has been developed to analyze the safety characteristics of SCWR-M for present work. In the code TACOS, the single channel model, point kinetic model, heat transfer model and control system model were used. Four different heat transfer correlations were applied to the steady state calculation of SCWR with mixed spectrum core (SCWR-M). After the comparison of four correlations, Bishop Correlation was selected for transient analysis. Using TACOS code, preliminary analysis of loss of flow accident (LOFA) for SCWR-M has been performed. Partial loss of flow and total loss of flow accident were discussed respectively. For the total loss of flow, the results indicate that the maximum cladding surface temperature (MCST) in the thermal zone was much higher than that in the fast zone. A “special flow mode” similar to natural circulation between coolant channels and moderator channels in the thermal zone has been found. The co-current flow mode design in the thermal zone would cause disadvantage to safety performance because it has a negative effect on the flow rate of coolant channels after the startup of Auxiliary Feedwater System (AFS). The sensitivity analyses of pump coastdown time and AFS flow rate for total loss of flow accident were also carried out.

Stratified flow-induced air-ingress accident assessment of the GAMMA code in HTGRs

An air-ingress accident is a major safety issue pertaining to high-temperature gas-cooled reactors. To see the effect of a stratified flow, which is a multi-dimensional phenomenon that occurs in large broken pipes, we perform 1-D and 2-D air-ingress simulations in the guillotine break of the main coaxial pipe of a 600 MWth GT-MHR with the GAs multicomponent mixture transient analysis (GAMMA) code. We used a 2-D fluid volume to build the coaxial inlet pipe, the lower plenum of the reactor core, and the cavity and simplified the other components as 1-D fluid blocks. After the guillotine break of the main coaxial pipe, the air in the reactor cavity flows into the reactor core in four phases: the blow-down phase, the stratified flow phase, the molecular diffusion phase, and the natural convection phase. In the early stage of a broken pipe, the lower plenum region of the reactor is filled with air within 30 s by a density-driven airflow. In a 1-D simulation, the process of filling the lower plenum with air ingressed from a cavity caused by the diffusion process takes 30 min. However, after 30 s, the flow velocity of air ingressed into the broken pipe decreases and the diffusion phase eventually begins. The natural circulation in this scenario starts after more than 360 h for the 1-D simulation but fails to commence after more than 500 h for the 2-D simulation. The belated natural circulation in the 2-D simulation is mainly attributed to the slower diffusion process in the core region. In turn, the slower diffusion occurs because the temperature of the air in the lower plenum is lower in the 2-D simulation than in the 1-D simulation. The maximum core temperature in the 2-D simulation was by 60 °C lower than that in the 1-D simulation.

Crush Behavior of Corrugated Cores Sandwich Panels

The out-of-plane quasi-static compressive behavior of four types of corrugated cores (V-type, U-type, X-type and Y-type core) has been investigated by experiment and FE simulations. By transient dynamic finite element analysis code MSC.Dytran numerical simulations were performed for calculating crushing forces, deformation mode and energy absorption. The FE simulations predict the crush behavior of cores with reasonable accuracy and provide the whole progressive buckling process and deformation modes. Experiment and simulation indicate that the U-type core, V-type core and X-type core structures show excellent crushing resistance performance and energy absorption characteristic. The crush performance of the Y-type core structures is relatively poor because of bending mode.

Numerical simulation using CFD software of countercurrent gas–liquid flow in a PWR hot leg under reflux condensation

In order to improve the countercurrent flow model of a transient analysis code, countercurrent air–water tests were previously conducted using a 1/15 scale model of the PWR hot leg and numerical simulations of the tests were carried out using the two-fluid model implemented in the CFD software FLUENT 6.3.26. The predicted flow patterns and CCFL characteristics agreed well with the experimental data. However, the validation of the interfacial drag correlation used in the two-fluid model was still insufficient, especially regarding the applicability to actual PWR conditions. In this study, we measured water levels and wave heights in the 1/15 scale setup to understand the characteristics of the interfacial drag, and we considered a relationship between the wave height and the interfacial drag coefficient. Numerical simulations to examine the effects of cell size and interfacial drag correlations on numerical predictions were conducted under PWR plant conditions. Wave heights strongly related with the water level and interfacial drag coefficient, which indicates that the interfacial drag force mainly consists of form drag. The cell size affected the gas velocity at the onset of flooding in the process of increasing gas flow rate. The gas volumetric fluxes at CCFL predicted using fine cells were higher than those using normal cells. On the other hand, the cell size did not have a significant influence on the process of decreasing gas flow rate. The predictions for the PWR condition using a reference set of interfacial drag correlations agreed well with the Upper Plenum Test Facility data of the PWR scale experiment in the region of medium gas volumetric fluxes. The reference interfacial drag correlations employed in this study can be applied to the PWR conditions.

A205 過渡安全最適評価コードのBOPモデル開発とサイクル評価への適用(OS8 軽水炉・新型炉の解析コード)

A205 過渡安全最適評価コードのBOPモデル開発とサイクル評価への適用(OS8 軽水炉・新型炉の解析コード)

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

Development of a thermal-hydraulic analysis software for the Chinese advanced pressurized water reactor

A point reactor neutron kinetics model, a drift-flow U-tube steam generator model, a non-equilibrium three-region pressurizer model and other models were established and a transient analysis code with Visual Fortran 6.5 has been developed to analyze the thermal-hydraulic characteristics of the Chinese advanced pressurized water reactor (AC-600). Visual input, real-time processing and dynamic visualization output were achieved with Microsoft Visual Studio.NET 2003, which greatly facilitate applications in the engineering. The software were applied to analyze the transient thermal-hydraulic characteristics of the loss of feed-water accident, the double loops loss-of-flow accident, the reactivity insertion accident, the sudden increase of feed-water temperature accident and the loss of offsite power accident for the Qinshan nuclear power plant in China. The obtained analysis results are significant to the improvement of design and safety operation of the plant.

Direct implementation of an axial-flow helium gas turbine tool in a system analysis tool for HTGRs

This study concerns the development of dynamic models for a high-temperature gas-cooled reactor (HTGR) through direct implementation of a gas turbine analysis code with a transient analysis code. We have developed a streamline curvature analysis code based on the Newton–Raphson numerical application (SANA) to analyze the off-design performance of helium gas turbines under conditions of normal operation. The SANA code performs a detailed two-dimensional analysis by means of throughflow calculation with allowances for losses in axial-flow multistage compressors and turbines. To evaluate the performance in the steady-state and load transient of HTGRs, we developed GAMMA-T by implementing SANA in the transient system code, GAMMA, which is a multidimensional, multicomponent analysis tool for HTGRs. The reactor, heat exchangers, and connecting pipes were designed with a one-dimensional thermal–hydraulic model that uses the GAMMA code. We assessed GAMMA-T by comparing its results with the steady-state results of the GTHTR300 of JAEA. We concluded that the results are in good agreement, including the results of the vessel cooling bypass flow and the turbine cooling flow.

Elasto-plastic impact response analysis of shear-failure-type RC beams with shear rebars

In this paper, to establish a simple elasto-plastic impact analysis method for shear-failure-type reinforced concrete (RC) beams, falling-weight impact tests and three-dimensional finite element (FE) analyses were conducted. Here, twelve simply supported rectangular RC beams were used, each with dimensions of (width × depth × length) 200 × 400 × 2,400 mm. Shear rebar ratio and impact velocity were taken as variables. Impact load was applied at the mid-span of RC beam by dropping a 400 kg steel weight from pre-determined position. LS-DYNA nonlinear transient finite element analysis code was used for this research. From this study, it is seen that the time histories of impact force, reaction force and mid-span displacement, and crack patterns on the side-surface of RC beam can be predicted accurately by using the proposed FE analysis method.

Analysis of Noncondensable Gas Recirculation Flow in Steam Generator U-Tubes during Reflux Condensation Using RELAP5

Under certain hypothetical accident circumstances in a pressurized water reactor, reflux condensation in steam generator U-tubes may be an effective heat removal mechanism. In a previous study, an empirical correlation of condensation heat transfer coefficients for reflux condensation was derived and incorporated into the transient analysis code RELAP5. In this study, different flow patterns in U-tubes observed in the reflux condensation test 7.2c with nitrogen injection in the Bethsy facility were calculated using a parallelchannel model for steady-state calculations and the RELAP5 code for transient calculations. In the region of low steam flow rates, the pressure in the outlet plenum becomes higher than that in the inlet plenum due to the density head of gases, and nitrogen reverse flow occurs, which was successfully calculated using the RELAP5 code with modified equations for the interfacial and wall friction forces. There are many solutions with different numbers of active U-tubes with steam condensation and the predicted number of active U-tubes agrees well with the measured values based on the assumption that the flow might be most stable in the case with the maximum nitrogen recirculation flow rate.

Improvements to the modelling of two-phase flow and heat transfer in a transient nuclear reactor analysis code

Accurate two-phase flow heat transfer prediction is of great importance in the analysis of reactor safety and TRAC (transient reactor analysis code) is a best estimate (BE) system code developed for such analysis. The work described here forms part of a research project that aims to evaluate and improve the TRAC code behavior by comparing code predictions with a range of “single effect” experiments. It has been shown that the necessary friction factor can be obtained from a voidage correlation and the drift flux parameters and the wall friction coefficient can be further derived from the annular flow model of Owen and Hewitt. These modifications have been combined to give what is regarded as an optimum code for vertical pipe flows. The work has been extended to make further improvements to the heat transfer correlations of the code. The performance of both the original and the improved codes has been tested against experiments on both evaporating and condensing flows. It is found that the improved code, combining the strongest parts of the best available correlations, gives better predictions in every case.

Reflux Condensation Heat Transfer of Steam-Air Mixture under Turbulent Flow Conditions in a Vertical Tube

Under certain hypothetical accident circumstances in a pressurized water reactor, reflux condensation in the steam generator U-tubes may be an effective heat removal mechanism. Non-condensable gas (air) may be present in the reactor coolant system and is well known to inhibit steam from condensing. In a previous study, local condensation heat transfer coefficients for reflux condensation with air were measured, and an empirical correlation intended for a laminar flow of steam-air mixtures was developed, which was a function of the partial pressure ratio, (P steam=P air). In this study, heat transfer coefficients measured under turbulent flows were used to modify the empirical correlation for improved prediction accuracy. The modified empirical correlation considers convection and is proportional to (P steam/P air)0.75 Re g 0.8 where Re g is the steam-air mixture Reynolds number. The correlation was verified against the data of Moon et al. and then incorporated into the transient analysis code RELAP5/SCDAPSIM/MOD3.2. The temperature distributions of the steam-air mixture calculated by the code for six reflux condensation tests are shown to agree well with the measured results.

ICONE15-10138 PRELIMINARY ANALYSIS OF TYPICAL TRANSIENTS IN FUSION DRIVEN SUBCRITICAL SYSTEM (FDS-I)

The potential safety characteristic is expected as one of the advantages of fusion-driven subcritical system (FDS-I) for the transmutation and incineration of nuclear waste compared with the critical reactor. Transients of the FDS-I may occur due to the perturbation of external neutron source, the failure of functional device, and the occurrence of the uncontrolled event. As typical transient scenarios, the following cases were analyzed: unprotected plasma overpower (UPOP), unprotected loss of flow (ULOF), unprotected transient overpower (UTOP). The transient analyses for the FDS-I were performed with a coupled two-dimensional thermal-hydraulics and neutronics transient analysis code NTC2D. The negative feedback of reactivity is the interesting safety feature of FDS-I as temperature increase, due to the fuel form of the circulating particle. The present simulation results showed that the current FDS-I design has a resistance against severe transient scenarios.

Preliminary Safety Analysis on Depressurization Accident without Scram of a Molten Salt Reactor

The Molten Salt Reactor (MSR) concept has recently been considered as one of the candidates for the generation IV nuclear power systems. MSRs have many advantages such as improved safety, proliferation resistance, resource sustainability and waste reduction. But MSR developmental activities have lagged and there are few data available to support detailed analyses. However, the authors believe that additional investigations are merited for future study. From this point of view, the authors have analyzed the depressurization accident of the MSR “Fuji-12” using a newly developed MSR transient analysis code. In Fuji-12, a small amount of helium gas bubbles are circulated in the primary loop for stripping out gaseous fission products. A depressurization of the primary system would cause these bubbles to expand, resulting in a positive reactivity insertion. We have attempted to determine the severity of such an accident. Although the analysis is still preliminary and the assumptions are crude, it can be expected that the depressurization would not result in a severe accident in Fuji-12.