Verification of NETFLOW code using plant data of sodium cooled reactor and facility

The plant simulation code NETFLOW on PC applicable to the liquid-metal cooled reactors has been developed on the basis of the models developed for single-phase and two-phase light water flow systems. The functions of this code have been verified by individual tests for light water flow systems and a sodium flow system. In order to apply this code to a sodium cooled fast reactor, several extra functions were verified using the plant data obtained using 50 MW steam generators and the Monju fast breeder reactor. Finally, the turbine trip transient of the Monju was simulated and the result was compared with the measured plant data. Good agreements were obtained in these verifications. As a result of the present study, the code can be applied as an education tool for students.     

Development of a steady state analysis code for a molten salt reactor

The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.     

Methods for calculating coupled reactor systems

Theoretical approaches to analyzing neutron kinetics in coupled reactor systems are reviewed. A systematic exposition of methods based on the use of integral parameters and an integral equation decribing neutron transport with the energy and angular variables eliminated is given. Special attention is given to a reactor-subcritical assembly as a coupled system.Translated from Atomnaya Énergiya, Vol. 97, No. 6, pp. 403–414, December, 2004.This revised version was published online in April 2005 with a corrected cover date.     

Criticality qualification of a new Monte Carlo code for reactor core analysis

In order to accurately simulate Accelerator Driven Systems (ADS), the utilization of at least two computational tools is necessary (the thermal–hydraulic problem is not considered in the frame of this work), namely: (a) A High Energy Physics (HEP) code system dealing with the “Accelerator part” of the installation, i.e. the computation of the spectrum, intensity and spatial distribution of the neutrons source created by (p, n) reactions of a proton beam on a target and (b) a neutronics code system, handling the “Reactor part” of the installation, i.e. criticality calculations, neutron transport, fuel burn-up and fission products evolution. In the present work, a single computational tool, aiming to analyze an ADS in its integrity and also able to perform core analysis for a conventional fission reactor, is proposed. The code is based on the well qualified HEP code GEANT (version 3), transformed to perform criticality calculations. The performance of the code is tested against two qualified neutronics code systems, the diffusion/transport SCALE-CITATION code system and the Monte Carlo TRIPOLI code, in the case of a research reactor core analysis. A satisfactory agreement was exhibited by the three codes.     

Vectorization and parallelization of a production reactor assembly code

In order to efficiently use new features of supercomputers, production codes, usually written 10 – 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantional speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090.     

Vectorization and parallelization of a production reactor assembly code

In order to efficiently use new features of supercomputers, production codes, usually written 10 – 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantional speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090.     

Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled RELAP5-3D system code and a CFD code

The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.     

Comparison of two models for a high temperature reactor coupled to a gas turbine

This paper presents two independent dynamic models of a nuclear gas turbine power plant. Both the high temperature nuclear reactor (HTR) and its energy conversion system (ECS) based on a direct Brayton cycle have been modelled. One model utilises RELAP5 for the ECS, the other Aspen Custom Modeler (ACM). The reactor model used in both models is a point kinetic model derived from a detailed reactor model. The ECS model is described and compared componentwise, with an emphasis on the turbomachinery. The total plant models are compared with each other by calculating two representative transients: one load rejection transient and one transient with the system at part load.     

BACO (Barra Combustible), a computer code for simulating a reactor fuel rod performance

The basic philosophy and mathematical structure of the fuel performance simulation code BACO is described. This code is based on a central finite-difference quasi-bidimensional approximation. Within that approximation, the thermoelastic-plastic behaviour of a in-service fuel rod is calculated by a set of equations which are linearized and solved for each time step by a sparse matrix inversion subroutine. The numerical method is shown to be stable and to converge rapidly to physically sound results for the stresses and strains. Changes in the fuel shape due to cracking and restructuring are included in the calculation within a self-consistent mathematical frame. Code convergence and accuracy are discussed by comparing some predictions against thermoelastic and plastic analytic solutions. An example of the code predictions for the rod state during a reactor shutdown is presented and discussed.     

Study on corrosion of welded steel for LBE-cooled fast reactors

Lead-bismuth eutectic (LBE) is a candidate coolant for fast reactors. Its physical, chemical and neutronic properties make it a prospect coolant for the reactors. However, corrosion of structure is the main problem of utilizing LBE as a coolant. Compatibility of welded structure with LBE at high temperature should be considered clearly for feasibility of lead-bismuth-cooled fast reactors. This study was preformed to investigate the mechanical properties and corrosion characteristics of the welded ferritic-martensitic (FM) steel, HCM12A, in LBE at 650 °C for 500 h. The welding methods were TIG welding (137 mm/min; 480 W), YAG laser welding (780 mm/min; 287 W) and electron beam welding (1000 mm/min; 60 kW). The oxygen concentration of test environment was maintained at 7 x 10⁻⁷ wt% by injecting Ar-H2-steam gas mixture. Vickers hardness test and SEM/EDX analysis were conducted on the cross section of welded HCM12A. It was found that oxide layer was larger in the weld zones than base metal (BM). However, outer layer was detached on some areas.     

Qualification of the CFD code Trio_U for full scale reactor applications

The article presents a procedure to qualify the Trio_U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).¹ The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio_U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.     

Efficient code system of core characteristics for small cylindrical thermal reactor

A new code system for the overall neutronic calculation of a thermal reactor by a simple and effective way is presented. The code covers microscopic library compilation, macroscopic constant generation, cell calculations by multi-group treatment for neutron transport equation and core calculations over three zones for fuel and one zone for moderator. The Dancoff correction factor required in the interpolation of the self-shielding factors of resonance nuclides is automatically calculated by the installed collision probability routines. The burn-up calculation and Garrison and Ross model of fission product have been included. Also the effect of control rod on the reactivity of the reactor with special treatment for the control rod based on the homogenization technique has been included. Making a comparison with SRAC95 code system has checked the adopted code.     

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

The Molten Salt Reactor(MSR) is one of the six advanced reactor nuclear energy systems for further research and development selected by Generation IV International Forum(GIF),which is distinguished by its core in which the fuel is dissolved in molten fluoride salt.Because fuel flow in the primary loop,the depletion of MSR is different from that of solid-fuel reactors.In this paper,an MCNP5 and ORIGEN2 Coupled Burnup(MOCBurn) code for MSR is developed under the MATLAB platform.Some new methods and novel arrangements are used to make it suitable for fuel flow in the MSR.To consider the fuel convection and diffusion in the primary loop of MSR,fuel mixing calculation is carried out after each burnup time step.Modeling function for geometry with repeat structures is implicated for reactor analysis with complex structures.Calculation for a high-burnup reactor pin cell benchmark is performed using the MOCBurn code.Results of depletion study show that the MOCBurn code is suitable for the traditional solid-fuel reactors.A preliminary study of the fuel mixture effect in MSR is also carried out.     

Development of a thermal-hydraulic analysis code TAC-DS for spent fuel dry-storage system for high-temperature pebble bed reactor

We report the development of a thermal-hydraulic analysis code (called TAC-DS: Thermal-hydraulic Analysis Code for Dry-storage System). The spent fuel dry-storage system of High-Temperature Reactor Pebble-bed Modules in China is simulated using the TAC-DS to confirm the design basis and to analyze the transient behavior following an accident involving blower failure. The TAC-DS includes mathematical models for the air-coolant system, heat conduction within spent fuel canisters, and thermal radiation between heat structures. The time-dependent hydrodynamic model of the TAC-DS is formulated using one-dimensional mass, momentum and energy equations, and solved using semi-implicit finite-difference scheme. The complicated heat transfer models of heat structure are incorporated into the hydrodynamic system implicitly with enclosure correlations. Code is written in Fortran 90. A validation calculation is performed by solving a simplified model. Thermal performance of the buffer storage region in the system under forced ventilation scenario is studied with TAC-DS to validate the design requirement, as well as to provide the initial condition for the transient analysis. Blower failure accident is studied to assess the performance of the safety features during the transient accident. Since the code is modular, TAC-DS can be easily modified and applied to other spent fuel dry-storage system in the future.     

Simulation of a main steam line break accident using a coupled “system thermal-hydraulics,three-dimensional reactor kinetics,and hot channel analysis” code

For a realistic analysis of thermal-hydraulics (T-H) transients in light water reactors, KAERI has developed the best-estimate T-H system code, MARS. The code has been improved from the consolidated version of the RELAP5/MOD3 and COBRA-TF codes. Then, the MARS code was coupled with a three-dimensional (3-D) reactor kinetics code, MASTER. This coupled calculation feature, in conjunction with the existing hot channel analysis capabilities of the MARS and MASTER codes, allows for more realistic simulations of nuclear system transients.     

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.