LWR Power Plants Research Articles

Performance of a transmutation advanced device for sustainable energy application

Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas-cooled pebble bed accelerator driven system, TADSEA (Transmutation Advanced Device for Sustainable Energy Application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, cooled by helium which enables high temperatures (in the order of 1200 K), to generate hydrogen from water either by high temperature electrolysis or by thermochemical cycles. For designing this device several configurations were studied, including several reflectors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW of thermal power. In this paper new studies performed on deep burn in-core fuel management strategy for LWR waste are presented. The fuel cycle on TADSEA device has been analyzed based on both: driven and transmutation fuel that had been proposed by the General Atomic design of a gas turbine-modular helium reactor. The transmutation results of the three fuel management strategies, using driven, transmutation and standard LWR spent fuel were compared, and several parameters describing the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain, are also presented.

Level 2 analysis to estimate LERF risk from a thermally induced rupture of a steam generator tube defect

This paper outlines the Level 2 portion of a methodology for determining the incremental induced steam generator tube rupture large early release fraction caused by an actual through-wall defect. This defect was responsible for the minor steam generator tube leak that occurred in September 2002 at the Comanche Peak Steam Electric Station Unit 1. In order to quantify the performance of the defect over the operating cycle, a range of defect lengths were input to the PROBFAIL computer code [Kenton, M., 2001. PROBFAIL: A Computer Code for Evaluating the Likelihood of Steam Generator Tube Rupture in Severe Nuclear Power Plant Accidents, CREARE TM-2138], using appropriate boundary conditions derived from MAAP4 [Henry, R., et al., May 1994. MAAP4—Modular Accident Analysis Program for LWR Power Plants, Computer Code Manual, EPRI Research Project 3131-02] runs. From the analysis of the calculated times of burst for each assumed defect length, the minimum through-wall defect length necessary for tube burst to occur prior to hot leg or surge line creep rupture was calculated. The probability that the defect would actually have this length was then estimated by determining the fraction of the cycle for which the defect would be at least that long. The methodology development and implementation relied on MAAP4 runs, which are discussed extensively in connection with their role in: (1) guiding the construction of the accident progression event tree, (2) generating relevant information for probability assignments in the various underlying fault trees and (3) obtaining boundary conditions of pressure and temperature for use in PROBFAIL. The overall increment in LERF due to the existence of the defect was calculated to be approximately 4E−08.

F211 高低温水合流部の高サイクル熱疲労設計評価基準策定 : 全体概要と考え方

This paper reports current efforts to establish a JSME code for the evaluation of high-cycle thermal fatigue for thermal striping in mixing tees with hot and cold water. The proposed design evaluation method has four steps to screen the design parameters one-by-one from the viewpoint of the severity of the thermal load predicted from design conditions. Five charts were produced by thermal striping tests for broad operating conditions of actual LWR power plants and a thermal stress model considering the effects of boundary layer attenuation for fluid temperature fluctuation, in order to realize the design evaluation method.

Systematic methodology for diagnosis of water hammer in LWR power plants

The paper gives the dimensions of the knowledge base that is necessary to carry out a diagnosis of water hammer susceptibility/root cause analyses for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) nuclear power plant systems. After introducing some fundamentals, water hammer phenomena are described. Situations where each phenomenon is encountered are given and analytical models capable of simulating the phenomena are referenced. Water hammer events in operating plants and their inclusion in the knowledge base is discussed. The diagnostic methodology is presented through an application on a system in a typical light water reactor plant. The methodology presented serves as a possible foundation for the creation of an expert water hammer diagnosis system.

The potential net energy gain from DT fusion power plants

The determination of the potential net energy gain from deuterium—tritium-fueled fusion power plants on the basis of tokamak reactors of present-day design is treated in four main sections. In the first the principle of net energy balancing is discussed in detail and the total net energy gain is deduced as a magnitude characterizing the exploitation of an energy source. Then the net energy balance for fusion power plants is determined by means of a method combining the advantages of the energetic input-output method with those of process chain analysis. Thirdly, these results are compared with LWR, HTR, FBR, and coal-fired power plants. Finally, a comparison of the results with those of other authors is made. As a result, the potential total net energy gain from fusion power plants considerably exceeds that from LWR, HTR, and coal-fired power plants and is in the same range as that of FBR power plants.

The potential net energy gain from DT fusion power plants

The determination of the potential net energy gain from deuterium—tritium-fueled fusion power plants on the basis of tokamak reactors of present-day design is treated in four main sections. In the first the principle of net energy balancing is discussed in detail and the total net energy gain is deduced as a magnitude characterizing the exploitation of an energy source. Then the net energy balance for fusion power plants is determined by means of a method combining the advantages of the energetic input-output method with those of process chain analysis. Thirdly, these results are compared with LWR, HTR, FBR, and coal-fired power plants. Finally, a comparison of the results with those of other authors is made. As a result, the potential total net energy gain from fusion power plants considerably exceeds that from LWR, HTR, and coal-fired power plants and is in the same range as that of FBR power plants.

照射後試験燃料輸送用ＮＨ‐２５型容器による燃料輸送の経験

Since the Japan Atomic Energy Research Institute and Nippon Nuclear Fuel Development Co. hot laboratories are located far off from the port which can handle spent fuel shipping casks, it is necessary to use a trailer-mounted cask which can be transported by public roads, bridges and intersections for the transportation of spent fuel specimens to these hot laboratories.Model NH-25 shipping cask was designed, manufactured and qualification tested to meet Japanese regulations and was officially registered as a BM type cask. The NH-25 cask accomodates two BWR fuel assemblies, one PWR assembly or one ATR fuel assembly using interchangeable inner containers. The cask weight is 29.2t. The cask has three concentric stainless steel shells. Gamma shielding is lead cast between the inner shell and the intermediate shell. Neutron shielding consists of ethylene-glycol-aqueous solution layer formed between the intermediate shell and the outer shell.The NH-25 cask now has been in operation for 2.5 yr. It was used for the transportation of spent fuel assemblies from six LWR power plants to the port on shipping cask carrier "Hinouramaru" on the sea, as well as from the port to the hot laboratory on a trailer. The capability of safe handling and transporting of spent fuel assemblies has been well demonstrated.

Cost analysis for LWR power plants

A statistical analysis of light water reactor power plant capital costs that uses a data base that is larger and of higher quality than that used in a previous study. The data span six years, and include virtually all U.S. LWR power plants presently in commercial operation. During this period average capital costs increased at the rate of about $140/k We (1978 dollars) per year, and average construction time increased at the rate of about 4 months per year. Significant economies of scale, either in construction time or capital cost, were not detected. Other findings were that plants built in the Northeast continued to show higher average costs than those in the rest of the country, the experience of the architect-engineer is a factor in reducing costs, and the costs of plants with cooling towers could not be distinguished from those without.