High-Temperture Structural Analysis on a Small-Scale PHE Prototype Considering Mechanical Properties in the Heat-Affected Zone and in the Weld Zone

PHE (Process Heat Exchanger) is a key component in transferring the high temperature heat generated from a VHTR (Very High Temperature Reactor) to the chemical reaction for massive production of hydrogen. A performance test on a small-scale PHE prototype made of Hastelloy-X is currently undergoing in a small-scale gas loop at the Korea Atomic Energy Research Institute. Previous researches on the high-temperature structural analysis of the small-scale PHE prototype had been performed using parent material properties over the whole region. In this study, high-temperature elastic structural analysis considering mechanical properties in the weld zone was performed and the analysis result was compared with previous researches.

Development of Long Rotating Blade Applied in the New Generation Nuclear Steam Turbine

In order to meet the high economic requirement of the 3rd generation Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) applied in currently developing nuclear power plants, a series of half-speed extra-long last stage rotating blades with 26 ∼ 30 m2 nominal exhaust annular area is proposed, which covers a blade-height range from 1600 mm to 1900 mm. It is well known that developing an extra long blade is a tough job involving some special coordinated sub-process. This paper is dedicated to describe the progress of creating a long rotating blade for a large scaled steam turbine involved in the 3rd generation nuclear power project. At first the strategy of how to determine the appropriate height for the last-stage-rotating-blade for the steam turbine is provided. Then the quasi-3D flow field quick design method for the last three stages in LP casing is discussed as well as the airfoil optimization method. Furthermore a sophisticated blade structure design and analyzing system for a long blade is introduced to obtain the detail dimension of the blade focusing on the good reliability during the service period. Thus, except for CAD and experiment process, the whole pre-design phase of the extra-long turbine blade is presented which is regarded as an assurance of the operation efficiency and reliability.

Design and Control of Bypass Condensers

In waste to energy plants and certain genre of cogeneration plants, it is mandatory to condense the steam from the boiler or HRSG in a separate bypass condenser when the steam turbine is out of service. The steam from the boiler or HRSG is attemperated in a pressure reducing desuperheating valve and then condensed in a bypass condenser. To avoid flashing of condensate in downstream piping it is customary to subcool the condensate in the bypass condenser. Circulating water from the steam surface condenser is used to condense the steam in the bypass condenser. Some of the challenges involved in the design of the bypass condenser are: • High shellside design pressure and temperature • Condensate subcooling • Large circulating water (tubeside) flow rate • Relatively low circulating water (tubeside) inlet temperature • Large Log Mean Temperature Difference (LMTD) • Large shell diameters • Small tube lengths The diverse requirements complicate the mechanical and thermal design of the bypass condenser. This paper highlights the complexities in the design and performance of the bypass condenser. Similarities with the design and operation of steam surface condensers and feedwater heater are reviewed. The uniqueness of the bypass condenser’s design and operation are discussed and appropriate solutions to ensure proper performance are suggested.

Thermochemical Hydrogen Production IS process

The Japan Atomic Energy Agency (JAEA) is conducting research and development on nuclear hydrogen production using High Temperature Gas-cooled Reactor and thermochemical water-splitting Iodine-Sulfur (IS) process aiming to develop large-scale hydrogen production technology for “hydrogen energy system”. In this paper, the present status of R&D on IS process at JAEA is presented which focuses on examining integrity of such components as chemical reactors, separators, etc. Based on previous screening of materials of construction mainly from the viewpoint of corrosion resistance in the harsh process conditions of IS process, it was planned to fabricate the IS components and examine their integrity in the process environments. At present, among the components of IS process plant consisting of three chemical reaction sections, i.e., the Bunsen reaction section, the sulfuric acid decomposition section and the hydrogen iodide decomposition section, key components in the Bunsen reaction section was fabricated.

Characteristics Analysis of SCWR During Partial Loss of Reactor Coolant Flow Transient

Partial loss of reactor coolant flow is one of the most important transients for safety analysis of supercritical water-cooled reactor (SCWR). Taking the super LWR concept provided by Japan as research object, transient analysis of partial loss of coolant flow rate is given by coupled neutronics and thermal hydraulics calculation method. The results show that, when 5% partial loss of coolant flow is happening, maximum cladding temperature would increase firstly with the decreasing of fuel channel inlet coolant flow. Then followed with the neutronic feedback and control operation, maximum cladding temperature decreases and finally return to normal. When 50% partial loss of coolant flow is happening, a scram signal will be given to ensure system safety, but the maximum cladding temperature still shows a significant increase early. On this basis, sensitivity analysis is performed considering the influence of core power and main coolant flow. It is found that maximum peaking value increases significantly following the initial flow rate decreasing, but shows a very little increase caused by core power increasing.

Dynamic Response Analysis of Control Rod Drive Mechanism Under the Stepping Impact Force for Nuclear Power Plants

The control rod drive mechanism (CRDM) is an important safety-related component in the nuclear power plant (NPP). When CRDM steps upward or downward, the pressure-containing housing of CRDM is shocked axially by an impact force from the engagement of the magnetic pole and the armature. To ensure the structural integrity of the primary coolant loop and the functionality of CRDM, dynamic response of CRDM under the impact force should be studied. In this manuscript, the commercial finite element software ANSYS is chosen to analyze the nonlinear impact problem. A nonlinear model is setup in ANSYS, including main CRDM parts such as the control rod, poles and armatures, as well as nonlinear gaps. The transient analysis method is adopted to calculate CRDM dynamic response when it steps upward. The impact loads and displacements at typical CRDM locations are successfully obtained, which are essential for design and stress analysis of CRDM.

Semi-Analytical Study of Aerosol Washout by Rain

The accidents of Chernobyl and Fukushima have shown the necessity to better understand all the mechanisms implied in the scavenging of aerosol particles released to the atmosphere during a nuclear accident. Among all the phenomena involved in the deposition of aerosol particles, we focus here on the aerosol particles scavenging by the raindrops below the clouds, also called washout (as opposed to the rainout, which concerns scavenging inside the clouds). The strategy of IRSN to enhance the knowledge and the modelling of any mechanism involved in the washout of aerosol particles by rain spans from environmental studies, to analytical ones. The semi-analytical approach chosen here is halfway between these two modes of reasoning. A companion paper is also submitted to the conference to present the microphysical approach chosen at IRSN. In order to perform this study, aerosol particles were dispersed in the TOSQAN chamber, which is a large cylindrical enclosure (4.8 m height with 1.5 m internal diameter). The aerosol particles once dispersed, synthetic rains of different kinds (from stratiform to convective rains) can be activated. Finally, the instantaneous spectral scavenging coefficients are determined from the spectral decrease of aerosol particles concentration in the chamber as a function of time. In order to be able to produce synthetic rains representative of any tropospheric events, a special generator has been designed; it is based on a vibro-rotative disk. This generator is able to produce monodispersed rains at the top of the TOSQAN chamber with rainfall rates from 7 to 15 mm/h and drops diameters from 0.5 to 2.5 mm injected at velocities close to their terminal one. During these tests, the spectral aerosol concentration is measured in line with the help of a Welas granulometer. This instrument is based on white light scattering. The results of these experiments highlight the influence of “meteorological” conditions inside the chamber on the washout of the chamber atmosphere, especially when the relative humidity is reaching saturation.

Hydrogen Detonation in a Flat Semi-Confined Layer

Hydrogen accumulation at the top of the containment or reactor building may occur due to an interaction of molten corium and water followed by a severe accident of a nuclear reactor (TMI, Chernobyl, Fukushima Dai-ichi). Hydrogen accumulates usually in a containment of nuclear reactor as a stratified semi-confined layer of hydrogen-air mixture. Detonation of such mixture may lead to significant damage of the containment structure. A series of large scale experiments on hydrogen combustion and detonation in a semi-confined layer of uniform and non-uniform hydrogen-air mixtures in presence of obstructions or without them was performed at the Karlsruhe Institute of Technology (KIT). Critical conditions for deflagration-to-detonation transition and then for steady state detonation propagation were experimentally evaluated in a flat semi-confined layer. The experiments were performed in a horizontal semi-confined layer with dimensions of 9×3×0.6 m with/without obstacles opened from below. The hydrogen concentration in the mixtures with air was varied in the range of 0–34 vol.% without or with a gradient of 0–1.1 mol. %H2/cm. Effects of hydrogen concentration gradient, thickness of the layer, geometry of the obstructions, average and maximum hydrogen concentration on critical conditions for detonation onset and then propagation were investigated with respect to the safety analysis. Blast wave strength and mechanical response of the safety volume was experimentally measured as well.

Reactor Kinetics in a Loss-of-Forced-Cooling (LOFC) Test of HTGRs

The High-Temperature Engineering Test Reactor (HTTR) is the first High-Temperature Gas-cooled Reactor (HTGR) with a thermal power of 30 MW and a maximum reactor outlet coolant temperature of 950 °C; it was built at the Oarai Research and Development Center of JAEA. At present, test studies are being conducted using the HTTR to improve HTGR technologies in collaboration with domestic industries that also contribute to foreign projects for the acceleration of HTGR development worldwide. To improve HTGR technologies, advanced analysis techniques are currently under development using data obtained with the HTTR, which include reactor kinetics, thermal hydraulics, safety evaluation, and fuel performance evaluation data (including the behavior of fission products). In this study, a three gas circulator trip test and a vessel cooling system (VCS) stop test were performed as a loss of forced cooling (LOFC) test to demonstrate the inherent safety features of HTGR. The VCS stop test involved stopping the VCS located outside the reactor pressure vessel to remove the residual heat of the reactor core as soon as the three gas circulators are tripped. All three gas circulators were tripped at 9, 24 and 30 MW. The primary coolant flow rate was reduced from the rated 45 t/h to 0 t/h. Control rods (CRs) were not inserted into the core and the reactor power control system was not operational. In fact, the three gas circulator tripping test at 9 MW has already been performed in a previous study. However, the results cannot be disclosed to the public because of a confidentiality agreement. Therefore, we cannot refer to the difference between the analytical and test results. We determined that the reactor power immediately decreases to the decay heat level owing to the negative reactivity feedback effect of the core, although the reactor shutdown system was not operational. Moreover, the temperature distribution in the core changes slowly because of the high heat capacity due to the large amount of core graphite. Core dynamics analysis of the LOFC test for the HTTR was performed. The relationship among the reactivities (namely, Doppler, moderator temperature, and xenon reactivities) affecting recriticality time and reactor peak power level as well as total reactivity was addressed. Furthermore, the analytical results for a reactor transient of hundred hours are presented. Based on the results, emergency operating procedures can be developed for the case of a loss of coolant accident in HTGR when the CRs are not inserted into the core and the reactor power control system is not operational. The analytical results will be used in the design and construction of the Kazakhstan High-Temperature Reactor and the realization of commercial Very High-Temperature Reactor systems.

Simulation of a Reinforced Concrete Slab Punching Impact

Reinforced concrete wall subjected to an impact by a hard steel missile with a mass of 47 kg and an impact velocity of 135 m/s was one case study in the IRIS 2010 benchmark exercise in OECD/NEA/CSNI/IAGE framework. The wall had dimensions of 2m × 2m × 0.25m and it was simply supported. The perforation of the missile was expected. Fortum Power and Heat Ltd. participated in the benchmark. In this paper, we present our modeling and blind prediction of the benchmark case. The test results of the benchmark were released after the predictions were made. Based on the result comparison, we concluded that our model gave conservative results.