The properties of WWER-1000 type materials obtained on the basis of a surveillance program

The surveillance test results of the reactor pressure vessels (RPV) of three Russian WWER-1000 units designated unit-1, -2 and -3 are given and the embrittlement rates compared to those predicted by the Russian Regulatory Guide. Dependence of the radiation behavior of WWER-1000 type RPV steels on metallurgical variables and the damage dose is considered. The trend curves for the steels under investigation are proposed.     

Using the MCU computer program to analyze the results of critical experiments with HTGR fuel pellets on the Astra testing stand

The Astra testing stand and the critical assemblies simulating on this stand the special physical features of HTGRs are described. The experimental data, obtained using the Astra stand, on the critical parameters of assemblies with HTGR fuel pellets and an analysis of the data using the MCU computer program, which employs the Monte Carlo method, are presented taking account of the requirements of the international program on critical benchmark experiments. It is shown that the experimental data are in good agreement with the computational results.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 243–252, October, 2004.     

Design and experimental results of a 28 GHz,400 kW gyrotron for electron cyclotron resonance heating

A high-power 28 GHz gyrotron has been successfully developed at the Institute of Applied Electronics, China Academy of Engineering Physics. This gyrotron was designed for electron cyclotron resonance heating(ECRH) in the spherical tokamak XL-50. A diode magnetron injection gun was designed to produce the required gyrating electron beam. The gyrotron operates in the TE8,3mode in a cylindrical open cavity. An internal quasi-optical mode converter was designed to convert the operating mode into a f...     

Experimental research on the gravity-driven boron injection system for a 200 MW nuclear heating reactor

The gravity-driven boron injection system (GDBIS), designed by the Institute of Nuclear Energy Technology (INET) of the Tsinghua University, PR China, is a new type of passive system to be applied in the 200 MW nuclear heating reactor (NHR-200), also designed by INET. The function of this system is to shut down the reactor in an emergency, in case control rods do not operate properly. A borate water tank is located 10 m above the top of the pressure vessel. When the pressure of the reactor and the boron tank balances, the borate water will be driven by gravity to flow into the reactor, and thus shut down the reactor. The thermal hydraulic performances of the system for cold (room temperature nitrogen) and hot (mixture of hot steam and nitrogen) operating conditions, especially the response time of pressure and water injection, have been researched under different initial conditions. Firstly, several factors, e.g. orifice on steam lines, and the volume ratio of the gas–steam spaces of the reactor and the boron tank, have effects on the pressure and water injection response time and other thermal hydraulic performance of the system. Secondly, the steam and liquid communication modes, namely the acting time and sequence of the action of valves connecting steam and liquid lines, have great influences on the performance of the system. Thirdly, the limited pressure balance time (about 1.0 s) can be achieved under the cold condition. This investigation shows that GDBIS can be properly used in the 200 MW nuclear heating reactor.     

Verification test results of a cutting technique for the ITER blanket cooling pipes

For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.     

Comparison of computational results obtained with different programs for radiation transfer

Main Science Center of the Russian Federation — Institute of High-Energy Physics. Translated from Atomnaya énergiya, Vol. 79, No. 4, pp. 269–279, October, 1995.     

Nontraditional concepts of nuclear power plants with inherent safety (new nuclear technology for the next stage of large-scale production of nuclear power)

This paper is a continuation of the discussion, initiated by A. A. Abagyan, G. I. Biryukov, S. V. Bryunin, et al. in their paper Status and problems of the development of nuclear power in the USSR (Atomnaya Énergiya,69, No. 2, 67–79, February, 1991), of the prospects for further development of nuclear power. It is the hope of the editorial staff that discussions of this important subject will continue.     

Verification of a TRACE EPR™ model on the basis of a scaling calculation of an SBLOCA ROSA test

In cooperation with the Finnish Radiation and Nuclear Safety Authority (STUK), a project has been launched at the Paul Scherrer Institute (PSI) aimed at performing safety evaluations of the Olkiluoto-3 nuclear power plant (NPP), the first EPR™, a generation III pressurizer water reactor (PWR); with particular emphasis on small-and large-break loss-of-coolant-accidents (SB/LB-LOCAs) and main steam-line breaks.As a first step of this work, the best estimate system code TRACE has been used to develop a model of Olkiluoto-3. In order to test the nodalization, a scaling calculation from the rig of safety assessment (ROSA) test facility has been performed. The ROSA large scale test facility (LSTF) was built to simulate Westinghouse design pressurized water reactors (PWR) with a four-loop configuration. Even though there are differences between the EPR™ and the Westinghouse designs, the number of similarities is large enough to carry out scaling calculations on SBLOCA and LOCA cases from the ROSA facility; as a matter of fact, the main differences are located in the secondary side. Test 6-1 of the ROSA 1 programme, an SBLOCA with the break situated in the upper head of the reactor pressure vessel (RPV), was of special interest since a very good agreement with the experiment was obtained with a TRACE input deck. In order to perform such scaling calculation, the set-points of the secondary relief and safety valves in the EPR™ nodalization had to be changed to those used in the ROSA facility, the break size and the core power had to be scaled by a factor of 60 (according to the core power and core volume) and the pumps coast down had to be adapted to the ones of the test. The calculation showed very similar results as the experiment and the ROSA-TRACE calculation. The only significant difference observed was a faster primary depressurization after the break flow turned to single-vapor flow. This difference could be explained on the basis of geometrical differences between the EPR™ and ROSA/Westinghouse RPV's designs.     

Cross-correlation detection of nonstationary signals due to a leak in the first loop of a nuclear power plant

A leak-detection system employing high-temperature microphones has been developed for RBMK and ATR (Japan) reactors. Further improvement of the system focused on using cross-correlation analysis of the spectral components of the signal to detect a small leak at an early stage of development. Frequency-time cross analysis was used to analyze nonstationary signals due to a leak, and the wavelet transformation was used to single out the envelopes corresponding to different spectral components of the signal. Since envelope processes are less affected by distortions than are wave processes, they give a higher-degree of correlation and can be used to detect leaks with lower signal/noise ratios. Many simulation tests performed at the Fugen nuclear power plant with an ATR reactor have shown that the proposed methods can be used to detect and find the location of a small leak. __________ Translated from Atomnaya énergiya, Vol. 102, No. 3, pp. 183–188, March, 2007.     

An exploration of the design for a PCCV head slab with very large penetrations using a dynamic relaxation computer program

A prestressed concrete containment vessel for a Liquid Metal Fast Breeder Reactor has been developed. The current design for the head slab for this vessel differs geometrically from previous head slabs for other reactors. This head slab has a single very large central penetration which is surrounded by an annulus in which there are a number of large and small penetrations. The total area taken by penetrations is in excess of 50% of the total pressurised plan area.This paper describes a computer analysis carried out using a Dynamic Relaxation code to assess the structural behaviour of the head slab. The effects of variations in the geometry and boundary conditions have been examined. The mathematical models considered were based on one of a series of small scale physical models intended to validate the design of the containment structure.The dynamic relaxation technique adopted to analyse the models allows for non-linear behaviour of materials with the provision for the application of step by step increments of both pressure and prestress forces. Concrete is modelled by assuming that it has a limited tensile stress capacity and once this is exceeded it acts as an orthotropic material. In addition, a yield condition can be specified to allow for triaxial stress states and the possible plastic flow. The method predicts crack propagation, hinge positions if any, tendon forces and overall structural deformations.The analysis has demonstrated that for a very slender end slab design, the presence of large unlined penetrations reduces the strength of the slab. It has also been demonstrated that the use of an upstand ring beam appears to give a greater reserve of strength than flat slabs even in situations in which failure of concrete rather than the prestressing system is the cause of structural collapse.Comparison between the experimental model results and those of the theoretical analyses indicate general agreement in crack patterns and failure modes. However, there are some discrepancies in actual failure pressures. These differences may however be attributed to some uncertainties that exist in the material parameter values and assumptions in the computer code.     

Considerations for in-service inspection on pressure vessel of 200MW nuclear heating reactor

The 200MW nuclear heating reactor adopts an integrated arrangement for the primary circuit. It is designed to be operated at lower temperature and lower pressure as compared to large reactors. A steel containment serves a barricade for the reactor pressure vessel. The pressure vessel has some safety characteristics, such as low stress level, low induced integral neutron flux, and high toughness etc. Among them, the most important is its LBB behavior. Based on the safety analysis for the pressure, the requirements and procedures of in-service inspection are layed-out accordingly.     

Experimental simulation study on start-up of the 5 MW nuclear heating reactor

An experimental simulation study on the start-up of a low temperature, natural circulation nuclear heating reactor (5 MW developed by the Institute of Nuclear Energy of Tsinghua University, Beijing) is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instability, namely geysering, flashing instability and low steam quality density wave instability on the start-up are described. The mechanism of flashing instability, which has never been well studied in this field, is especially interpreted. Based on the study of these instabilities, it is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: increasing of initial pressure by means of a noncondensable gas (N₂), start-up of the reactor at this pressurized condition (single-phase regime operation), and transition to a lower pressure, boiling operation. Three transition methods are discussed. As a result of these studies, the method of transition with low heat flux and low inlet subcooling is proposed. A stable start-up process of the 5MW reactor is achieved by careful selection of the thermohydraulic parameters.     

Recent results on the application of electron cyclotron heating to tokamaks

Many experiments using Electron Cyclotron Heating (ECH) of plasmas in tokamaks have been reported over the past several years. At a power level of 4 MW, ECH has achieved electron temperatures as high as 10 keV in the T-10 tokamak, and the H-mode has been attained in divertor discharges in DIII-D and JFT-2M. Regarding global energy confinement in either L-mode or Hmode, ECH appears to be quite similar in efficiency to neutral injection, but in addition to bulk heating it has been useful for many purposes, including study of local electron heat diffusivity through pulse-modulated heating; suppression of sawteeth, Edge Localized Modes, and other MHD activity; suppression of disruptions; preionization and startup; and current drive. In this paper, progress in these areas which has been reported since the IAEA meeting in 1986 will be summarized.Work supported by U.S. Department of Energy Contract No. DEAC03-89ER51114.     

The structural design of superconducting magnets for the large coil program

Fusion reactor designs based on magnetic confinement will require the use of superconducting magnets to make them economically viable. For a tokamak fusion reactor; large magnetic field coils are required to produce a toroidal magnetic confinement volume. Although superconductors have been used for approximately 20 years, several requirements for their application in fusion reactors are beyond demonstrated technology in existing magnets. The Large Coil Program (LCP) is a research, development, and demonstration effort specifically for the advancement of the technologies involved in the production of large superconducting magnets. This paper presents a review of the status of the structural designs, analysis methods, and verification tests being performed by the participating LCP design teams in the US, Switzerland, Japan, and the Federal Republic of Germany. The significant structural mechanics concerns that are being investigated with the LCP are presented.     

An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.