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.     

Thermal hydraulic analysis of SMART for heat removal transients by a secondary system

SMART is an integral type reactor of 330 MW, which enhances its safety by adopting inherent safety design features. Thermal hydraulic characteristics of transients in heat removal by a secondary system for the SMART have been carried out by means of the TASS/SMR and MATRA codes. The primary, secondary, and passive residual heat removal systems RHRS of the SMART were modeled properly. Then, a set of transients for the whole system was investigated. The results of the analyses using the conservative initial and boundary conditions showed that the safety features of the SMART design carried out their functions well and there was a strong moderator temperature coefficient due to the soluble boron free reactor affected by the transient behavior. The natural circulation was well established in the primary and passive residual heat removal systems during the transients and was enough to ensure a stable plant shutdown condition after a reactor trip.     

A model for analysis of fuel behaviour in transients

For many applications, analysis of fuel element behaviour must take non-linear thermal, and elasto-plastic effects into account. This is particularly true if the fuel undergoes large deformations and rapid temperature transients. To meet this need a multi-dimensional fuel model based on finite element stress and thermal analysis has been developed. The model is solved for the transient temperature distribution by a step-by-step time incremental procedure. The temperature is then introduced into the elasto-plastic analysis as a thermal load and stresses and deformations are calculated. A model for treatment of creep and a special element for the gap between fuel pellet and cladding is incorporated together with semi-empirical procedures for calculating fission gas release, fuel pellet to cladding heat transfer coefficients, etc.The fuel model has been compared with both analytical solutions and in-reactor experimental results. The observed and predicted results are in good agreement.     

A fuzzy modeling approach to the identification of transients in nuclear components

In this paper, a fuzzy-logic modeling approach is adopted for the identification of the causes of transients in a component of a nuclear power plant. The if–then rules, representing the underlying processes are inferred from the available input–output signal data. The method is applied to the early classification of the causes of transients in a steam generator of a pressurized water reactor (PWR). Based on the measured signals, the forcing function responsible for the transient is readily classified and its amplitude is estimated. The case of two concurrent causes of transients is also considered.     

Multivariate analysis with covariance matrix applied to separative power modeling of a gas centrifuge

In this work, the least-squares methodology with covariance matrix is applied to determine a data curve fitting to obtain a performance function for the separative power δU of an ultracentrifuge as a function of variables that are experimentally controlled. The experimental data refer to 460 experiments on the ultracentrifugation process for uranium isotope separation. The experimental uncertainties related to these independent variables are considered in the calculation of the experimental separative power values, determining an experimental data input covariance matrix. The process variables, which significantly influence the δU values, are chosen to give information on the ultracentrifuge behaviour when submitted to several levels of feed flow rate F, cut θ and pressure in the product line, P_p. After the model goodness-of-fit validation, a residual analysis is carried out to verify the assumed basis concerning its randomness and independence and mainly the existence of residual heteroscedasticity with any explained regression model variable.     

Analysis of low load transients for the Vandellòs-II NPP: Application to operation and control support

The Technical University of Catalonia (UPC) has been jointly working with the Asociación Nuclear Ascó-Vandellòs (ANAV) for a number of years in order to establish, qualify and use best estimate (BE) models for the reactors under the control of ANAV. ANAV is the consortium that is responsible for operation of the Ascó and Vandellòs-II reactors. The reactors are Westinghouse-design three-loop PWRs with an approximate electrical power of 1000 MW. The existing integral plant models for each plant are currently used for many different purposes among which are support of plant operation and control. Quite a number of studies have been done in order to improve both safety and plant competitiveness. Most of these dynamic analyses were carried out in relation to transients starting at nominal full power or at least, very close to full power. This paper develops a specific use of the Vandellòs-II plant model for operation and control support at low power involving new ranges of system actuation parameters. It also examines scenarios that are somewhat different from those typically analysed. The study starts showing the results of an assessment case, which is a start-up test and provides some additional qualification, and subsequently attempts to establish calculations to support both an improvement in feed water controls and to set up operating recommendations for low-load manual operation of feed water turbo-pump. Both results hopefully, will produce an outcome, which leads to an improvement in safety and reduces reactor trip probability.     

Tugsim-10, a computer code for transient analysis of closed gas turbine cycles and specific applications

A code has been developed to compute the kinetic and thermohydraulic transients in a nuclear gas turbine power plant for both operational and accidental conditions. A depressurization accident analysis demonstrates the performance of the code and shows how the core may be cooled by coolant circulation through the turbomachinery, using the afterheat to drive the gas turbine. Emergency shut-down calculations for a 30 MW(e) fossil fuel plant are compared with measurements.     

Project of the GT-MHR high-temperature helium reactor with gas turbine

In February 1995, MINATOM of Russia and General Atomics (USA) signed the Agreement for the development and design of the GT-MHR facility with a modular helium reactor and a gas turbine intended to be constructed in Russia. This Agreement was subsequently expanded by the participation of Framatom and Fuji Electric. The GT-MHR facility is designed for burning weapons-grade plutonium and utilization of the heat produced in the direct gas-turbine cycle with electricity production efficiency of about 50%. In future such facilities with uranium fuel will be proposed for use as commercial NPPs. A GT-MHR prototype and fuel production facility are intended to be constructed at the Siberian Chemical Combine in Seversk (Tomsk-7). In accordance with the Agreement, a conceptual design of the GT-MHR should be developed in September 1997. As a part of the conceptual design, a reactor module with a power conversion system is being designed and plutonium fuel is being developed.     

Engineering analysis of mass flow rate for turbine system control and design

The mass flow rate is determined in the steam turbine system by the area formed between the stem disk and the seat of the control valve. For precise control the steam mass flow rate should be known given the stem lift. However, since the thermal hydraulic characteristics of steam coming from the generator or boiler are changed going through each device, it is hard to accurately predict the steam mass flow rate. Thus, to precisely determine the steam mass flow rate, a methodology and theory are developed in designing the turbine system manufactured for the nuclear and fossil power plants. From the steam generator or boiler to the first bunch of turbine blades, the steam passes by a stop valve, a control valve and the first nozzle, each of which is connected with piping. The corresponding steam mass flow rate can ultimately be computed if the thermal and hydraulic conditions are defined at the stop valve, control valve and pipes. The steam properties at the inlet of each device are changed at its outlet due to geometry. The Compressed Adiabatic Massflow Analysis (CAMA) computer code is written to predict the steam mass flow rate through valves. The Valve Engineered Layout Operation (VELO) test device is built to experimentally study the flow characteristics of steam flowing inside the control valve with the CAMA input data. The Widows’ Creek type control valve was selected as reference. CAMA is expected to be commercially utilized to accurately design and operate the turbine system for fossil as well as nuclear power plants.     

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.     

Research and developments of guidelines for thermal load modeling

In order to realize reliable and economical plants, Japanese commercialized fast reactors adopt innovative component design such as simplification of a reactor vessel, integration of intermediate heat exchangers with primary pumps, shortening of pipes and reduction of loop numbers. Thermal load, one of the principal loads in fast reactor components, becomes critical in those components. For realistic evaluation and mitigation of thermal loads “Interim Guidelines for Thermal Load Modeling” were developed. These guidelines are referred from “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)”.Current guidelines deal with system thermal transient load and thermal striping load. As for system thermal transient load, two kinds of modeling method were provided. Concerning thermal striping load, thermal fatigue evaluation method was developed considering attenuation effects of thermal stress related to frequency of temperature fluctuation.Interpretation documents of the guidelines and exemplars of application problems were also provided to support designers.     

Parametric studies on different gas turbine cycles for a high temperature gas-cooled reactor

The high temperature gas-cooled reactor (HTGR) coupled with turbine cycle is considered as one of the leading candidates for future nuclear power plants. In this paper, the various types of HTGR gas turbine cycles are concluded as three typical cycles of direct cycle, closed indirect cycle and open indirect cycle. Furthermore they are theoretically converted to three Brayton cycles of helium, nitrogen and air. Those three types of Brayton cycles are thermodynamically analyzed and optimized. The results show that the variety of gas affects the cycle pressure ratio more significantly than other cycle parameters, however, the optimized cycle efficiencies of the three Brayton cycles are almost the same. In addition, the turbomachines which are required for the three optimized Brayton cycles are aerodynamically analyzed and compared and their fundamental characteristics are obtained. Helium turbocompressor has lower stage pressure ratio and more stage number than those for nitrogen and air machines, while helium and nitrogen turbocompressors have shorter blade length than that for air machine.     

Development of a 2D conduction code for a finned fuel element analysis

The fuel element of KMRR (Korea Multi-purpose Research Reactor) has 8 longitudinal, rectangular fins to enhance the heat transfer performance. The existence of these fins makes it difficult to analyze the heat transfer phenomena within the fuel element using the conventional one-dimensional heat conduction model. As the uncertainty in the computation of the maximum sheath temperature significantly affects the core thermal margin, a computer code, called, TEMP2D, which is based on a two-dimensional heat conduction model has been developed to deal with the finned element and validated. This computer code TEMP2D has a fully implicit numerical scheme and can solve both the steady state and transient problems such as the changes in coolant thermal-hydraulic conditions and fuel pin power. The code accuracy, which proved to be an excellent one, was verified by comparing its results with those from two widely accepted computer codes, MARC and ADINA. The result of this code calculation has been used to compute the KMRR core thermal margin and to develop a correlation for the equivalent 1D heated diameter which can reproduce the maximum cladding temperature (or heat flux) at various steady states when used in the 1D heat conduction model.     

Two-phase modeling of a gas phase fluidized bed reactor for the fluorination of uranium tetrafluoride

A bubbling fluidized bed reactor for the fluorination of uranium tetrafluoride by fluorine gas was simulated employing two-phase models, with the bubble phase assumed to be in plug flow, and the emulsion phase in plug flow (P-P model) and in perfectly mixed flow (P-M model). The model calculations were compared with actual data in term of fluorine conversion. The comparison showed that the P-M model predicted the data satisfactorily. The P-M model was then applied to understand the influence of operating parameters on the reactor performance. A sensitivity study of various operating parameters showed the extent each parameter can influence the rates of fluorine conversion and uranium hexafluoride production.     

A simple approach to the prediction of waterhammer transients in a pipe line with entrapped air

The pressure histories within entrapped air bubbles in a pipe line during a waterhammer transient are treated theoretically. A convenient integral method is introduced, which takes full account of air/water interface movement and liquid compressibility. The significance of the method is that it provides a simple equation set for approximating, with good accuracy and with a small degree of conservatism, the solution to a problem that otherwise involves coupled partial differential equations on time dependent domains with non-linear boundary conditions. The accuracy of the method is defined by its comparison with available numerical-solution-predictions and measurements of the pressure within an entrapped-air-bubble at a dead end in a pipe. The method is shown to be a computationally simple and efficient way of assessing the impact of liquid compressibility on pressure rise when multiple water columns and air pockets are present in a pipe line.     

The effect of certain cycle parameters on the efficiency of a nuclear gas turbine unit

The effect is analyzed of the increase of the effective pressure ratio, the regeneration factors, the initial gas pressure, temperature of the fuel element cladding, hydraulic resistance of the gas circuit on the internal efficiency of a nuclear gas turbine unit, taking into account the characteristics of the active zone of the reactor. The results are given of the effect on the efficiency of a nuclear gas turbine unit (NGTU) of the intermediate heating and cooling of the gas. A possible circuit for a NTGU is discussed, with one intermediate heating and three-stage cooling of the gas;Translated from Atomnaya Énergiya, Vol. 20, No. 5, pp. 412–415, May, 1966.     

LEIS: A reliable tool for surface composition analysis?

Different single and polycrystalline surfaces of Cu and Ag have been investigated by time-of-flight low-energy ion scattering using ⁴He⁺ ions. The fraction of ions that survived single scattering from the outermost surface layers, P⁺, was measured in different neutralization regimes. At low energies, a distinct difference in P⁺ was observed for non-equivalent Cu crystal surfaces for projectiles backscattered in a single collision. The polycrystalline surface was found to exhibit similar neutralization behaviour as the (1 1 1) single crystal surface. At higher energies, P⁺ shows a strong dependence on the angular orientation of the single crystal. The impact of these findings on quantitative surface composition analysis by LEIS is discussed.     

Study of a 105 GHz short-pulse dummy load for the electron cyclotron resonance heating system with the quasi-optical method

The accurate power measurement is important for an ECRH system in tokamak. The dummy load is designed and developed for the measurement of the millimeter wave power. This work analyzes the dummy load based on the quasi-optical method and the ray tracing method. The reflectivity and thermal deposition of the dummy load have been considered to ensure the safety of the entire system. High-power tests have been carried out at a 105 GHz/500 kW ECRH system. The results of the tests indicate that the designed dummy load is stable and valid.