R&D on high temperature components

Experimental studies were carried out to confirm the structural integrity of the key high temperature components, which are necessary to commercialize the HTGR, that is, an intermediate heat exchanger (IHX) and a coaxial hot gas duct. As for the IHX, the following tests were conducted: (1) creep collapse of tube against external pressure, (2) creep fatigue of tube against thermal stress, (3) seismic behavior of tube bundles, (4) thermal hydraulic behavior of tube bundles and (5) in-service inspection technology of tube with full-scale models. As for the coaxial hot gas duct, the thermal insulation performance as well as the structural integrity was investigated with the HENDEL-loop.     

PWR steam generator R&D in France: thermal-hydraulic and vibratory aspects

Concomitant with the launching of the French pressurized water reactor (PWR) nuclear power program, a large research and development (R&D) effort was initiated, devoted to the steam generators (SGs). This program, managed cooperatively by Framatome, the SG designer and manufacturer; Electricité de France (EDF), the French electrical utility; and the Commissariat à l'Energie Atomique (CEA), the French Atomic Energy Commission, primarily responsible for nuclear research; was focused on four main objectives:

1. (1) To obtain a better understanding of the physical phenomena existing in these steam generators and leading to SG performance alterations or operating life reductions.

2. (2) To test and validate improved design solutions for the model 51 Framatome steam generator, which was the first one designed under Westinghouse license.

3. (3) To test and validate new Framatome SG designs.

4. (4) To test and validate new, high-performance design tools.

This vast R&D program covers the following theses:

• - SG thermal-hydraulics,

• - SG tube vibration and wear,

• - SG materials (production, corrosion, etc.),

• - Primary and secondary fluid chemistry,

• - SG technology (manufacturing processes, NDT, etc.),

• - SG in-service inspection, and

• - SG maintenance.

These themes are too numerous to be dealt with in a single article. Consequently, the present article will focus on only the first two themes.     

Fuel R&D at Studsvik I. Introduction and experimental facilities

The different types of fuel research and development programmes at Studsvik are presented. The R2 test reactor and its facilities for irradiation experiments, including INCA, which is a new in-pile facility for waterside corrosion studies, are described. Further, different techniques for non-destructive examinations are described.     

In-service inspections and R&D of PCCVs in Japan

In Japan, the following five plants in operation use prestressed concrete containment vessels (PCCVs): Tsuruga #2 (in operation in 1987), Ohi #3 and #4 (in operation in 1991 and 1993, respectively), and Genkai #3 and #4 (in operation in 1994 and 1997, respectively). These plants have adopted the hemispherical dome type PCCV with two or three buttresses and unbonded prestressing system. For the above five PCCVs, lift-off tests in the in-service inspection have been carried out 1, 3 and 5 years after the beginning of operation. In addition, concrete strain and temperature have been measured periodically. This paper presents the in-service inspection methods and results of the lift-off tests and the measurement of concrete strain. The given results are as follows. Measured strain shows good agreement with calculated creep and shrinkage strain, so tendon tensile force can be predicted by the design method using calculated strain, and it also matches well the results of the lift-off tests.     

Parametric evaluation of large-scale high-temperature electrolysis hydrogen production using different advanced nuclear reactor heat sources

High-temperature electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800-950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an intermediate heat exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies.The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO₂ reactor concepts). Sensitivity studies were then performed with the objective of evaluating the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency of the integrated plant design for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered.     

Main R&D issues for fast reactor structural design standard

For realization of economical and reliable fast reactor (FR) plants, the Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC) are cooperating on the “Feasibility Study on Commercialized FR Cycle Systems”. To certify the design concepts through evaluation of the structural integrity of FR plants, the research and development of the “Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)” is recognized as an essential theme. The FDS focuses on particular failure modes of FRs such as ratchet deformation and creep-fatigue damage due to cyclic thermal loads. For precise evaluation of these modes, the research and development for three main issues is in progress. First, the “Refinement of Failure Criteria” needs to be addressed for particular failure modes of FRs. Secondly, the development of “Guidelines for Inelastic Design Analysis” is conducted to predict elastic plastic and creep deformation under elevated temperature conditions. Lastly, efforts are being made toward preparing “Guidelines for Thermal Load Modeling” for the design of FR components where thermal loads are dominant.     

R&D activities based on fast reactor cycle technologies for transmutation of TRU and LLFP by JNC

Research and development(R&D) activities on partitioning and transmutation of trans-uranium nuclides (TRU) and LLFP and future R&D program in JNC were summarized. Feasibility design studies have been conducting to investigate the characteristics of a fast reactor core with TRU and LLFP transmutation. It was reconfirmed that the fast reactor has a strong potential for transmuting TRU and LLFP, effectively. R&D for establishing partitioning process of TRU apart from the high-level radioactive wastes have been carried out. By several counter-current runs of the TRUEX process using highly active raffinates, a process flow sheet capable of selective partitioning of actinides and fission products was newly developed. JNC has settled a new R&D program concerning partitioning and transmutation of long-lived radioactive waste based on recommendation of check & review for OMEGA program performed by the Ad Hoc Committee under the Atomic Energy Commission of Japan (AEC). The R&D program is composed of the design studies and development of element technologies (fabrication, irradiation) in the “Feasibility Studies” on commercialized fast reactor system and the basic studies with experiments (nuclear data, reactor physics, fuel property, etc.) to establish database and analytical tools for the TRU- and LLFP- containing fuel and core design.     

Synergistic hydrogen production by nuclear-heated steam reforming of fossil fuels

Processes and technologies to produce hydrogen synergistically by the nuclear-heated steam reforming reaction of fossil fuels are reviewed. Formulas of chemical reactions, required heats for reactions, saving of fuel consumption, reduction of carbon dioxide emission, and possible processes are investigated for such fossil fuels as natural gas, petroleum and coal.

In this investigation, examined are the steam reforming processes using the “membrane reformer” and adopting the recirculation of reaction products in a closed loop configuration. The recirculation-type membrane reformer process is considered to be the most advantageous among various synergistic hydrogen production processes. Typical merits of this process are; nuclear heat supply at medium temperature around 550°C, compact plant size and membrane area for hydrogen production, efficient conversion of a feed fossil fuel, appreciable reduction of carbon dioxide emission, high purity hydrogen without any additional process, and ease of separating carbon dioxide for future sequestration requirements.

The synergistic hydrogen production using fossil fuels and nuclear energy can be an effective solution in this century for the world which has to use fossil fuels to some extent, according to various estimates of global energy supply, while reducing carbon dioxide emission.     

Fuel R&D at Studsvik III. Studies of special phenomena in fuel behaviour: lift-off and defect fuel degradation

Two different types of research and development programmes at Studsvik are discussed. Cladding lift-off studies have been pursued, where a high end-of-life overpressure in high burn-up fuel rods may cause outward creep and consequently an increased pellet-cladding gap. Primary defects, mainly caused by debris fretting, may lead to large secondary failures due to internal hydriding. These problems have been studied in the R2 test reactor.     

The effects of deformation on the high-temperature steam oxidation of zircaloy-2

An experimental study was made of the effects of deformation on the oxidation of Zircaloy-2 reactor fuel cladding in flowing steam in the temperature range 700 to 1300°C. The kinetics and mechanism of oxygen penetration and embrittlement were examined using gravimetric and metallographic methods. Tensile deformation during oxidation resulted in cracking of the growing oxide and local oxidation of the metal at the base of the crack. The number of points of local attack was inversely related to the temperature of oxidation. The depth of local penetration of the metal (in excess of the uniform depth of penetration for undeformed cladding) was greatest when the deformation rate was slow. The maximum depth of local attack found in these experiments was equivalent to an increase in total oxygen penetration of about twice that due to uniform diffusion. The possible relevance of these results to the analysis of reactor loss-of-coolant accidents is discussed.     

Solid polymer electrolyte water electrolysis systems for hydrogen production based on our newly developed membranes, Part I: Analysis of voltage–current characteristics

A new solid polymer electrolyte water electrolysis system was constructed using an original proton exchange membrane (PEM). The highly proton-conductive PEM was prepared by the γ-ray-induced post-grafting of styrene into a crosslinked-polytetrafluoroethylene (PTFE) film and subsequent sulfonation. The water vapor to be electrolyzed was controlled at a constant relative humidity and introduced into the cell at different temperatures up to 80 °C. As the cell voltage was increased, the current became higher; the maximum current was 50 mA/cm² at 2.5 V at a temperature of 80 °C, corresponding to a hydrogen production rate of 0.38 mL/min cm² in the normal state (25 °C, 1 atm). The voltage–current characteristics were analyzed with a theoretical model based on Butler–Volmer kinetics for electrodes and transport resistance through the PEM. This analysis revealed that the anode exchange current density and interfacial resistance determined the electrolysis performance.     

Real-time measurement of hydrogen generation level during beryllium dust oxidation by steam depending on the dust arrangement geometry

The paper concentrates on the safety issues in the International Thermonuclear Experimental Reactor (ITER) and describes the experiment on the measurement of hydrogen generation rate in case of Ingress of Coolant Event (ICE)—leak inside the vacuum vessel during interaction between water and beryllium (Be) dust. The ICE situation in ITER was simulated in a facility; the active spectroscopy was used to define the hydrogen content by the dynamics of oxidant concentration at a sampling frequency up to 10 Hz. Hydrogen release in time at temperatures of 500-900 °C is investigated, and different versions of dust arrangement are considered, i.e. on the surface and in a slot between armoring tiles at different initial density. The obtained results are compared with the known experiments.     

Study on efficiency of DCP for nuclear hydrogen production

With many advantages, hydrogen is considered as the fuel of the future. But there is no natural resource of hydrogen and it must be produced by other kinds of energy. As for the primary energy, nuclear energy is a promising alternative. Using heat from nuclear reactor to produce hydrogen is receiving more and more concerns in recent years. This paper mainly emphasizes the study of the direct contact pyrolysis （DCP） of methane using heat from nuclear reactor. A facility was designed to investigate the efficiency of DCP process in certain conditions. The experimental results show that this process produces only hydrogen and carbon. The conversion efficiency increases with temperature and residence time, but decreases as flow rate increases. The highest efficiency of DCP obtained in this exoedment is about 22%.     

Physical characteristics and problems of developing a sodium-cooled fast reactor as a source of high-potential thermal energy for hydrogen production and other high-temperature technologies

Hydrogen as an energy carrier will play a considerable role in the future structure of energy production when nuclear reactors will replace fossil fuels. Investigations performed in recent years have shown that the production of large quantities of energy with high efficiency can be accomplished on the basis of thermochemical cycles using reactors with coolant temperature at the exit from the core of about 900°C. A variant of such a fast reactor with sodium as the coolant is proposed. The main physical characteristics and the main problems which must be solved to build such a reactor are presented. According to its properties, this reactor will meet the modern requirements for nuclear and radiation safety. It can also be used in other promising high-temperature technologies, for example, high-efficiency production of electricity. Deceased. (V. B. Bogush) Translated from Atomnaya énergiya, Vol. 106, No. 3, pp. 129–134, March, 2009.     

Electrochemical calorimetric studies on the electrolysis of water and heavy water (D2O)

Electrochemical calorimetric studies were carried out in H₂O and D₂O using Pd and Pt cathodes with LiOD(H) electrolytes. Two types of calorimetric cell designs were used in this attempt to detect excess enthalpy during the electrolysis of D₂O using Pd cathodes. Control experiments were run side by side using water and/or Pt cathodes. A variety of pretreatments and cathode attachment schemes were employed. No discernible differences were detected between the Pd/D₂O cells and the controls, regardless of conditions. For example, using one type of calorimetric cell and conditions, the mean value for the excess enthalpy was 6.5±4% in D₂O compared to 7.5±7% in H₂O. For the other type cell, the mean result for one series of experiments was 0±4% compared to Pt cathodes.     

Measurement of heat transfer coefficients for steam condensation on a vertical 21.5-mm-O.D. tube in the presence of air

ABSTRACT

Governing the rate of heat transport by condenser tubes in the passive containment cooling system (PCCS), the steam condensation over a vertical cylinder in the presence of air was investigated experimentally. The main objective of this study was to explore if the condensation heat transfer coefficient relies on the tube dimension, which has been a variable missed in most condensation models or has been embraced without experimental demonstration under phase change environments. The mean heat transfer coefficient was measured in the condensation test facility named JERICHO (JNU Experimental Rig for Investigation of Condensation Heat transfer On tube). The outer diameter of the condenser tube used in this study was set to 21.5 mm. The measured heat transfer coefficients were compared to those obtained from the 40-mm-O.D. tube, and a multiplier to correct the variation of the heat transfer coefficient with the tube diameter was proposed for its application to Lee correlation. The proposed correlation was further validated against another set of experimental data obtained from a separate test facility housing the 31.8-mm-O.D. tube.     

聚变高温制氢反应堆锂铅包层结构力学分析

在高温液态锂铅包层结构设计、热工水力学设计和中子学计算基础上,建立包层的三维有限元分析模型,应用商用有限元软件ANSYS对高温液态包层进行热-力结构耦合分析与应力评定。经计算第一壁材料ODS RAFM钢最高温度635℃,最大Von Mises应力379 MPa;包层结构材料RAFM钢最高温度508℃,最大Von Mises应力175 MPa;FCI材料最高温度950℃,最大Von Mises应力218 MPa。初步的分析结果表明结构设计方案是合理、可行的。     

Separation of plutonium from lanthanum by electrolysis in LiCl–KCl onto molten bismuth electrode

This work presents a study on the electroseparation of plutonium from lanthanum using molten bismuth electrodes in LiCl–KCl eutectic at 733 K. The reduction potentials of Pu³⁺ and La³⁺ ions were measured on a Bi thin film electrode using cyclic voltammetry (CV). A difference between the peak potentials for the formation of PuBi₂ and LaBi₂ of approximately 100 mV was found. Separation tests were then carried out using different current densities and salt phase compositions between a plutonium rod anode and an unstirred molten Bi cathode in order to evaluate the efficiency of an electrolytic separation process. At a current density of 12 mA/cm²/wt% (Pu³⁺), only Pu³⁺ ions are reduced into the molten Bi electrode, leaving La³⁺ ions in the salt melt. Similar results were found at two different Pu/La concentration ratios ([Pu]/[La] = 4 and 10). At a current density of 26 mA/cm²/wt% (Pu³⁺), co-reduction of Pu and La was observed as expected by the large negative potential of the Bi cathode during the separation test.     

Reactor core design optimization of the 200 MWt Pb–Bi cooled fast reactor for hydrogen production

In this study reactor core geometrical optimization of 200 MWt Pb–Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 °C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 °C and the maximum coolant outlet temperature less than 700 °C. By taking into account of the hydrogen production as well as corrosion resulting from Pb–Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 °C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.