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.

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.

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.

Development of New Simulation Software System to Evaluate Radiation Dose Rates in a Wide Radioactively Contaminated Area

The Japanese government plans to conduct decontamination tasks in radioactively contaminated areas. For such a situation, we developed a system that evaluates radiation dose rates in a wide radioactively contaminated area by utilizing our radiation dose evaluation technology. This system can not only generate present maps of radiation dose rate in the air based on the dose rate measured at the surface of the contaminated areas, but can also quickly calculate the reduction effect of dose rate due to decontamination tasks by entering decontamination factors. The system can then formulate decontamination plans and make it possible to plan measures to reduce radiation exposure for workers and local residents. Radioactive nuclides that contribute to gamma-ray dose rate are mainly Cs-134 and Cs-137 in soil, on trees, buildings, and elsewhere. Shapes of such radiation sources are assumed to be 10m square or 100m square. If it is unsuitable that the radiation sources assume to squares, the radiation sources can assume to point. The relation between distance from the surface or point source and the radiation dose rate is calculated using MCNP5 code (A General Monte Carlo N-Particle Transport Code - Version 5), and approximated using four-parameter empirical formula proposed by Harima et al. In addition, the system can consider shielding such as soil, concrete, and iron. When setting such shielding, the skyshine dose rate is taken into account in dose rate calculation.

Underground Siting of Nuclear Power Plants: Insights From Fukushima

Siting of nuclear power plants in an underground nuclear park has been proposed by the authors in many previous publications, first focusing on how the present 1200 to 1600 MW-electric light water reactors could be sited underground, then including reprocessing and fuel manufacturing facilities, as well as high level permanent waste storage. Recently the focus has been on siting multiple small modular reactor systems. The recent incident at the Fukushima Daiichi site has prompted the authors to consider what the effects of a natural disaster such as the Japan earthquake and subsequent tsunami would have had if these reactors had been located underground. This paper addresses how the reactors might have remained operable — assuming the designs we previously proposed — and what lessons from the Fukushima incident can be learned for underground nuclear power plant designs.

Stress Test Evaluation of Refueling and Storage Pools for Loviisa NPP

The paper will describe the following analysis of Loviisa plant’s spent fuel pools. As a consequence of stress tests for the existing NPP’s in Finland after the experiences gathered from the Tohoku -Taheiyou-Oki event in Japan in March of this year Fortum Power and Heat Oy (Fortum) has initiated the following analyses of the Loviisa power plant’s refueling pools and spent fuel intermediate storage pools for the combined cooling loss and the earthquake loads. The following loads will be analyzed: 1) The spent fuel pool and refueling pools water temperature is 100 degrees Celsius. The heat load duration is undetermined; 2) The earthquake ground motion applied simultaneously with the thermal load is defined as follows: (ground motion response spectrum is defined in Guide YVL 2.6), the maximum horizontal acceleration is assumed to be 0.1g, 0.2g, 0.3g and 0.4g, respectively; 3) Own weight; 4) The pool of water, hydrostatic pressure load plus the sloshing load because of earthquake motion. Te analysis is aimed to demonstrate the structural integrity and leak-tightness of the pools under the effect of above loads. The analysis is nonlinear taking into account the cracking of the concrete. As a result of the analysis the maximum strains will be determined in the pool stainless steel liner as well in the pool concrete walls.

Validation and Benchmarking of Comprehensive Vibration Assessment Program for Prototype Reactor Internals

Reactor internals important to nuclear power plant safety shall be designed to accommodate steady-state and transient vibratory loads throughout the service life of the reactor. Operating experience has revealed failures of reactor internals in both pressurized water reactors (PWRs) and boiling water reactors (BWRs) due to flow-induced vibrations (FIVs). U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) that the NRC staff considers acceptable for use in verifying the structural integrity of reactor internals for FIV prior to commercial operation. A CVAP supports the NRC reviews of applications for new nuclear reactor construction permits or operating licenses under 10 CFR Part 50, as well as design certifications and combined licenses that do not reference a standard design under 10 CFR Part 52. The overall CVAP should be implemented in conjunction with preoperational and initial startup testing. For prototype reactor internals, the comprehensive program should consist of a vibration and fatigue analysis, a vibration measurement program, an inspection program, and a correlation of their results. Validation and benchmarking processes should be integrated into the CVAP throughout each individual program. Based on the authors’ experiences in Advanced Boiling Water Reactor and AP1000® CVAPs and based on detailed reviews of the U.S. Evolutionary Power Reactor and the U.S. Advanced Pressurized Water Reactor CVAPs, this article summarizes the essential CVAP validation and benchmarking processes with proper consideration of bias errors and random uncertainties. This article provides guidance to a successful CVAP that satisfies the NRC requirements and ensures the reliability of the evaluation of potential adverse flow effects on nuclear power plant components.