Securing the safe performance of graphite reactor cores

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The safety margin prediction of 0. Cite this paper: A. Oludare, M.

Kundrecensioner

Agu, A. Umar, S. Adedayo, O. Omolara, L. Article Outline 1. Introduction 2. Graphite-Moderated Reactors 3. Shrinkage of Material 4. Linear Expansion 5. Effects on Strain 6. Area Expansion 7. Volumetric Expansion 8. Accident Analysis 9. Risk Analysis Future Reactors Methodology The Research Objectives The Research Motivation Results and Analyses Introduction Researches have shown that cooling problem of graphite moderated reactor especially during an accident can contribute to the acceleration of the reactor melt-down, since the design dimension of graphite height and diameter play significant role in the safety of these types of reactor[1].

This design dimension in terms of height and diameter of the reactor graphite core need to be considered in other to minimize high increase in the fuel temperature during operation and accident in other to avoid reactor melt-down and to keep the reactor safe. The large size of graphite core will provide low power density reactor that will minimize heat conservation in the reactor core and also disallow core melting. The small size of graphite core could contribute to the causes of decay heat within reactor core of nuclear power plant either during operation or accident.

The decay heat in the core assemblies could degenerate to hydrogen built-up that can make reactor to fail, as identified in some reactor accidents[2]. Recently, data for the dimensional change of AGR graphite have been successfully fitted to irradiation temperature, ignoring any effect of fast neutron flux level Eason e tal, In this work comparism of different test on graphite moderated reactor design GMRD models parameter specification of height and diameter was carried out before conclusion. This paper purpose was to test fuel and materials. The reactor design concept is intended to allow for the high temperature effect on dimensional variation in the stability of commercial graphite moderated power reactors.

It is hope that this conceptual design would provide a good, novel approach and method for multi-objective decision-making in the development of the nuclear industry when applied.

Graphite-Moderated Reactors A graphite reactor is a nuclear reactor that uses carbon as a neutron moderator, which allows un-enriched uranium to be used as nuclear fuel. Nuclear graphite is any grade of graphite , usually electro-graphite , specificallymanufactured for use as a means of production moderator or reflector within nuclear reactors.

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NEA Publications: ISSUE BRIEF No. 5 - ADVANCED WATER REACTOR TECHNOLOGY

Graphite is an important material for the construction of both historical and modern nuclear reactors as it is one of the purest materials manufactured at industrial scale and it retains its properties including strength even at high temperatures. Vrain Generating Station -is a natural gas powered electricity generating facility. Shrinkage of Material There are three types of shrinkage: i Shrinkage of the liquid, ii Solidification shrinkage and iii Patternmaker's shrinkage. The shrinkage of the liquid is rarely a problem because more material is flowing into the mold behind it.

Solidification shrinkage occurs because metals are less dense as a liquid than a solid, so during solidification the metal density dramatically increases. Patternmaker's shrinkage refers to the shrinkage that occurs when the material is cooled from the solidification temperature to room temperature, which occurs due to thermal contraction. An equation of state can be used to predict the values of the thermal expansion at all the required temperatures and pressures , along with many other state functions.

In the major equations of state, if for a given amount of substance contained in a system, the temperature, volume, and pressure are not independent quantities; they are connected by a relationship of the general form: 1 In the following equations the variables are defined as follows. Any consistent set of units may be used, although SI units are preferred.

Linear Expansion To a first approximation, the change in length measurements of an object "linear dimension" as opposed to, e. It is the fractional change in length per degree of temperature change. Assuming negligible effect of pressure, we may write : 4 where L is a particular length measurement and is the rate of change of that linear dimension per unit change in temperature.

If it does, the equation must be integrated. Effects on Strain For solid materials with a significant length, like rods or cables, an estimate of the amount of thermal expansion can be described by the material strain, given by and defined as: 6 where Linitial is the length before the change of temperature and Lfinal is the length after the change of temperature. Area Expansion The area thermal expansion coefficient relates the change in a material's area dimensions to a change in temperature. It is the fractional change in area per degree of temperature change.

Ignoring pressure, we may write: 10 where A is some area of interest on the object, and is the rate of change of that area per unit change in temperature. Volumetric Expansion For a solid, we can ignore the effects of pressure on the material, and the volumetric thermal expansion coefficient can be written: 12 where V is the volume of the material, and is the rate of change of that volume with temperature.

This means that the volume of a material changes by some fixed fractional amount. For example, a steel block with a volume of 1 cubic meter might expand to 1. This is an expansion of 0.


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If we had a block of steel with a volume of 2 cubic meters, then under the same conditions, it would expand to 2. The volumetric expansion coefficient would be 0. If we already know the expansion coefficient, then we can calculate the change in volume 13 where is the fractional change in volume e. The above example assumes that the expansion coefficient did not change as the temperature changed. This is not always true, but for small changes in temperature, it is a good approximation. If the volumetric expansion coefficient does change appreciably with temperature, then the above equation will have to be integrated: 14 where To is the starting temperature and is the volumetric expansion coefficient as a function of temperature T.

Management of Ageing in Graphite Reactor Cores RSC Special Publications

The temperature of the fuel varies as a function of the distance from the center to the rim. At distance d from the center the temperature T d is described by the equation where is the thermal conductivity. When thermal motion causes a particle to move towards the observer, the emitted radiation will be shifted to a higher frequency. Likewise, when the emitter moves away, the frequency will be lowered. Since there is a distribution of speeds both toward and away from the observer in any volume element of the radiating body, the net effect will be to broaden the observed line.

Nuclear Engineering Securing the Safe Performance of Graphite Reactor Cores

Figure 9 shows the tree structure of the OPT process [ 39 ]. The application of OPT is expected to contribute from early stages to the built-in safety approach proclaimed as part of Generation IV safety assessment philosophy. Its final objective is to ensure that Defense in Depth satisfies the desirable attributes of being exhaustive, balanced, progressive, tolerant and forgiving [ 5 ].

During the pre-conceptual and conceptual design phases OPT will serve as an important guidance to research efforts in order to achieve the mentioned Defense in Depth attributes by means of robust, reliable and simple design solutions. Best-estimated deterministic computer codes are preferred, incorporating sensitivity analyses to cover the existing uncertainties, depending on the design stage. One important challenge is the upgrade, development and validation of deterministic computer codes and their necessary input data to perform convincing safety assessments of some innovative reactor concepts.

RSWG supports the idea of applying PSA from the earliest practical stages of the design process, and continuing its application iteratively throughout the evolution of the design concept until its maturity, in the stages of final design, licensing and operation. The PSA scope should comprise both internal and external events.

PSA is recognized as a fundamental tool to prioritize properly design and operational issues which are more significant to safety, contributing in this way to a proper balance between costs and safety effectiveness of Generation IV Nuclear Energy Systems. PSA advantages for a systematic understanding and evaluation of risk uncertainties is also remarked. It is expected that PSA will contribute to understand differences in the level of safety of diverse technical proposals and select the designs that better fulfill the selection safety criteria for a given Generation IV reactor concept.

To achieve its leading role, PSA must be validated as part of a rigorous quality assurance program, including a peer review conducted by independent experts. ISAM is a relatively new methodology which is still under development and adjustment. It is recognized that the methodology will need to be modified or updated based on the lessons and findings derived from the Fukushima accident [ 5 , 44 ].

RSWG demonstration that ISAM can be used to evaluate and document the safety of Generation IV nuclear systems, supported by an extended use of the methodology in practical applications, is only beginning but has excellent perspectives of becoming a future reality. It will certainly depend on international efforts and progress in the materialization of Generation IV reactor concepts.

Its creation followed the recommendations from the economics crosscut group of the Generation IV roadmap project that a standardized cost estimating protocol be developed to provide decision makers with a credible basis to assess, compare, and eventually select future nuclear energy systems, taking into account a robust evaluation of their economic viability.

The methodology developed by the EMWG is based upon the economic goals of Generation IV nuclear energy systems, as adopted by GIF: to have a life cycle cost advantage over other energy sources, to have a level of financial risk comparable to other energy projects i. This section briefly describes an economic model for Generation IV nuclear energy systems [ 6 ] and the accompanying software [ 45 ] in which the guidelines and models were implemented. These tools will integrate cost information prepared by Generation IV system development teams during the development and demonstration of their concept, thus assuring a standard format and comparability among concepts.

This methodology will allow the Generation IV International Forum GIF Experts Group to give an overview to policy makers and system development teams on the status of available economic estimates for each system and the relative status of the different systems with respect to Generation IV economic goals. Figure 10 displays the structure of the integrated nuclear energy economic model. The following discussion is based on this figure. Cost estimates prepared by system design teams should report the overall direct and indirect costs for reactor system design and construction base construction cost and an estimate of the reactor annual operation and maintenance costs.

The intent is that these costs be developed using the GIF COA described in [ 6 ], prepared by the methods outlined therein. The decision maker, however, needs more than just the overall costs in each life cycle category. Of particular interest are the cost per kilowatt of installed capacity and the cost of electricity generation cost per kilowatt-hour from such systems, including the contribution of capital and non-fuel operations.

This total cost is amortized over the plant economic life so that the capital contribution to the levelized unit of energy cost LUEC can be calculated. Fuel cycle materials and services are purchased separately by the utility or the fuel subcontractor. For fuel cycles commercially deployed, there are mature industries worldwide that can provide these materials and services. Markets are competitive, and prices are driven by supply and demand. The fuel cycle model requires as inputs the amount of fuel needed for the initial core and subsequent equilibrium cores, along with the fissile enrichment of the uranium or plutonium, and, for uranium, the transaction tails assay assumed by the enrichment service provider.

The EMWG model uses algorithms similar to those described in [ 48 ] to estimate the overall cost for each step and ultimately the unit cost contribution of fuel to electricity cost. Background material on the economic aspects of fuel cycle choices including information on nuclear materials and fuel cycle service unit costs for conventional reactor types that use commercially available fuels can be found in NEA reports [ 48 , 49 ].

These documents include cost data on fuel reprocessing and high-level waste disposal for closed fuel cycles and spent fuel disposal for the once-through option. Innovative fuel cycles or fuel cycle steps for which no industrial scale or commercial facilities currently exist, especially for fuel fabrication, reprocessing, and waste disposal are also addressed. For example, the Very-High-Temperature Reactor system will require high-temperature particle fuel and the SFR Sodium-Cooled Fast Reactor system might require innovative pyrometallurgical and pyrochemical facilities for fuel fabrication, reprocessing, and re-fabrication.

For such systems, price data for fuel cycle services generally are not readily available. The design team must supply data on the design and construction costs for the facilities, along with an estimate of their annual production rates and operation costs. Algorithms discussed in [ 6 ] can produce rough approximations of the unit costs.