Analysis of Strategies for Improving Uranium Utilization in Pressurized Water Reactors PDF Download

Author: Joseph A. Sefcik
Publisher:
ISBN:
Category : Nuclear fuel elements
Languages : en
Pages : 241

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Book Description
Systematic procedures have been devised and applied to evaluate core design and fuel management strategies for improving uranium utilization in Pressurized Water Reactors operated on a once-through fuel cycle. A principal objective has been the evaluation of suggested improvements on a self-consistent basis, allowing for concurrent changes in dependent variables such as core leakage and batch power histories, which might otherwise obscure the sometimes subtle effects of interest. Two levels of evaluation have been devised: a simple but accurate analytic model based on the observed linear variations in assembly reactivity as a function of burnup; and a numerical approach, embodied in a computer program, which relaxes this assumption and combines it with empirical prescriptions for assembly (or batch) power as a function of reactivity, and core leakage as a function of peripheral assembly power. State-of-the-art physics methods, such as PDQ-7, were used to verify and supplement these techniques. These methods have been applied to evaluate several suggested improvements: (1) axial blankets of low-enriched or depleted uranium, and of beryllium metal, (2) radial natural uranium blankets, (3) low-leakage radial fuel management, (4) high burnup fuels, (5) optimized H/U atom ratio, (6) annular fuel, and (7) mechanical spectral shift (i.e. variable fuel-to-moderator ratio) concepts such as those involving pin pulling and bundle reconstitution. The potential savings in uranium requirements compared to current practice were found to be as follows: (1) O0-3%, (2) negative, (3) 2-3%; possibly 5%, (4) "15%, (5) 0-2.5%, (6) no inherent advantage, (7) 10%. Total savings should not be assumed to be additive; and thermal/hydraulic or mechanical design restrictions may preclude full realization of some of the potential improvements.

Author: Massachusetts Institute of Technology. Energy Laboratory
Publisher:
ISBN:
Category : Pressurized water reactors
Languages : en
Pages : 73

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This is the final summary progress report on a research program carried out within the MIT Energy Laboratory/Nuclear Engineering Department under the US Department of Energy's program to increase the effectiveness of uranium utilization in light water reactors on the once-through fuel cycle. Two major themes, methodology and applications, characterize the research. A simple built accurate set of algorithms, designated as "the linear reactivity method" were developed to permit self-consistent evaluations of a broad spectrum of changes in core design and fuel management tactics. More than a dozen suggested improvements were then evaluated, focusing primarily on retrofittable modifications and pressurized water reactors. In common with the findings of many other investigators, high burnup and routine end-of-cycle coastdown were identified as preferred options.

Author:
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ISBN:
Category :
Languages : en
Pages : 5

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This document is an interim report under ACDA BOA AC9NX707, Task Order 80-03, which covers the evaluation of certain potential improvements in pressurized water reactor designs intended to enhance uranium fuel utilization. The objective of these evaluations is to seek advanced, non-retrofittable improvements that could possibly be commercialized by the end of the century, and, on the basis of a preliminary evaluation, to select compatible improvements for incorporation into a composite advanced pressurized water reactor concept. The principal areas of investigation include reduced parasitic absorption of neutrons (Task 1), reduced neutron leakage (Task 2), and alternative fuel design concepts (Task 3). To the extent possible, the advanced concept developed in an earlier study (Retrofittable Modifications to Pressurized Water Reactors for Improved Resource Utilization, SSA-128, October 1980) is used as a basis in developing the advanced composite concept. The reference design considered typical of present PWR commercial practice is the system described in RESAR-414, Reference Safety Analysis Report, Westinghouse Nuclear Energy Systems, October 1976.

Author: Yigal Ronen
Publisher: CRC Press
ISBN: 9780849360817
Category : Technology & Engineering
Languages : en
Pages : 282

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Book Description
The purpose of this book is to describe concepts related to advanced water reactors, with particular focus on Advanced Pressurized Water Reactors. It discusses the severe disadvantages which water reactors have with respect to uranium utilization. It also reveals new concepts in which the conversion ratio and the uranium utilization is improved. This interesting work includes information on various others ways used in addition to the increase in the conversion ratio. This is an informative, useful book for all nuclear scientists and engineers, and anyone who is interested in high converting water reactors.

Author: R. M. Fleischman
Publisher:
ISBN:
Category :
Languages : en
Pages :

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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 5

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This report summarizes work performed for the U.S. Arms Control and Disarmament Agency under BOA AC9NX707 (Task Order 80-02), as part of the Agency's continuing program on improved fuel utilization in light water reactors. The objective of the study was to investigate improvements in fuel management and design of water reactors (PWRs) that could potentially increase the utilization of natural uranium resources in a once-through fuel cycle (i.e., without using spent fuel reprocessing and recycle). For the present study, potential improvements were limited to retrofittable concepts, i.e., those which could be modifications to the reactor system or balance of plant. The potential improvements considered were not necessarily restricted to those which might be economical under current uranium ore prices or to those which might be acceptable to the nuclear industry at the present time. A six-month fuel cycle, for example, although technically possible, would be neither economical nor accept able to the industry at the present time. Although all potential improvements are not necessarily compatible with each other, the target objective was to seek a composite system of compatible improvements that, if possible, could increase uranium resource utilization by 30% or more. Economic factors, risks involved in the introduction, and potential licensing concerns are also addressed in the report.

Author: Charles P. Kliewer
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 320

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Book Description
A significant amount of interest has been aroused recently concerning the advancement of the current pressurized water reactor core designs with a special emphasis towards improving the conversion characteristics of these reactors. Most reports have been divided into two camps; those that deal with the neutronic aspects and those that deal with the thersohydraulic concerns. Seldom do these two areas get combined for purposes of evaluating a new design. In this effort, the author takes a pragmatic approach to this area in so far as looking into ways of incorporating this advanced technology into current operational power plants. In so doing the Trojan Nuclear Power Plant was selected to serve as the reference design plant. This was done since it is a Westinghouse designed reactor, as are a large portion of the PWR's in the United States, and because it has one of the largest thermal power ratings in the nation as well. Both neutronic and thermohydraulic aspects are examined as well as an alternative fuel concept. In order to carry out the analysis computer codes COBRA-IV and LEOPARD were employed on a CYBER 170/710 mainframe computer. COBRA-IV was used to obtain results relating to the associated pressure loss and core temperature characteristics while LEOPARD was used for the neutronic aspects. A parametric study was initiated using fuel enrichment and pitch as the variables that would be systematically changed. As an additional factor to assure cross compatibility, the fuel rod diameter was held to a constant value throughout this analysis. Results of this research strongly indicate that current operational power plants can be effectively altered to become converters or low level breeders with only configurational changes in the core itself and no major equipment changes. Hence the author concludes that this concept is both feasible and readily attainable with the current level of technology. An additional benefit that would be realized under the adoption of this design would be the marked improvement in the utilization of uranium ore which ultimately becomes fuel. This would directly result in the extension of the power generation capability associated with nuclear power well into the next century.

Author: Guillaume Giudicelli
Publisher:
ISBN:
Category :
Languages : en
Pages : 124

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This work aims at investigating the potential benefits of nitride fuel use in pressurized water reactors. The AP1000 is chosen as the reference power plant. Both oxide and nitride fuel are considered and compared using a steady state thermal hydraulics and mechanics parametric optimization study to achieve a maximal core power. A subsequent neutronics study determined the achievable energy extracted per fuel mass (burnup) and sets the core power that allows for an 18-months fuel cycle length. The impact of the change in the core operating temperature on the steam cycle efficiency is considered in order to provide a final evaluation of the electric power uprate. The steady state limits considered are pressure drop, minimum departure from nucleate boiling ratio, fretting and sliding wear and fuel average and centerline temperatures. These limits were set by the reference design's performance. Two strategies were used to raise the core power while remaining within specified limits: increasing the core mass flow rate and decreasing the core inlet temperature. These two strategies were implemented in a simplified MATLAB tool using correlations and a MATLAB-VIPRE (subchannel simulation tool) interface to better model cross-flows. Designs with smaller pins but with similar pitch-todiameter ratios compared to the reference design were found to be optimal with regards to these performances for both strategies. Fretting wear was found to be the limiting constraint for these designs for the first strategy, and additional spacer grids can be introduced to reduce fretting wear and to allow a further power increase. MDNBR was found to be the limiting constraint for these designs in the second strategy. The fuel temperature was not limiting for these designs and both oxide and nitride fuel can be utilized with the same uprates. Both tools provided similar results: smaller fuel pins with similar pitch over diameter ratios allow for better performances than the nominal design in the aforementioned criteria. The most promising strategy proved to be decreasing the core inlet temperature. With this strategy, the possible uprate is 16%, or 550 MWth, in both tools. Such an uprate requires an additional steam generator, and when lowering the core inlet temperature the efficiency of the steam cycle is lowered by 1% as we also need to lower the steam generator saturation pressure. This will require a larger high-pressure turbine. The optimized nitride-fueled design was compared with the oxide-fueled nominal core in terms of neutronics performances. I showed that the new design can reach an 18 month cycle length, at an uprated power, with a 4.3% enrichment and a 60 assembly feed using uranium nitride, compared with a 4.6% enrichment and a 68 assembly feed for the nominal design at the nominal power. With a higher enrichment and a higher feed, a two-year cycle length can be reached even with the uprate. The moderator temperature coefficient, the shutdown margin and the power coefficient of both designs satisfied licensing requirements. A 5% increase in fuel cycle costs was noted with the nitride optimized core, minor compared to the revenue of a 150 MWe uprate. Transient performances, and more extensive fuel performance studies are left for future studies.

Author:
Publisher:
ISBN:
Category : Power resources
Languages : en
Pages : 404

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Author: Christian Diego Di Sanzo
Publisher:
ISBN:
Category :
Languages : en
Pages : 155

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Breed and Burn reactors (B & B), also referred to Traveling Wave Reactors, are fast spectrum reactors that can be fed indefinitely with depleted uranium only, once criticality is achieved without the need for fuel reprocessing. Radiation damage to the fuel cladding limits the fuel utilization of B & B reactors to ~ 18-20% FIMA (Fissions of Initial Metal Atoms) - the minimum burnup required for sustaining the B & B mode of operation. The fuel discharged from this type of cores contain ~ 10% fissile plutonium. Such a high plutonium content poses environmental and proliferation concerns, but makes it possible to utilize the fuel for further energy production. The objectives of the research reported in this dissertation are to analyze the fuel cycle of B & B reactors and study new strategies to extend the fuel utilization beyond ~ 18-20% FIMA. First, the B & B reactor physics is examined while recycling the fuel every 20% FIMA via a limited separation processing, using either the melt refining or AIROX dry processes. It was found that the maximum attainable burnup varies from 54% to 58% FIMA - depending on the recycling process and on the fraction of neutrons lost via leakage and reactivity control. In Chapter 3 the discharge fuel characteristics of B & B reactors operating at 20% FIMA and 55% FIMA is analyzed and compared. It is found that the 20% FIMA reactor discharges a fuel with about ~ 80% fissile plutonium over total plutonium content. Subsequently a new strategy of minimal reconditioning, called double cladding is proposed to extend the fuel utilization in specifically designed second-tier reactors. It is found that with this strategy it is possible to increase fuel utilization to 30% in a sodium fast reactor and up to 40% when a subcritical B & B core is driven by an accelerator-driven spallation neutron source. Lastly, a fuel cycle using Pressurized Water Reactors (PWR) to reduce the plutonium content of discharged B & B reactors is analyzed. It was found that it is possible to burn the B & B discharged fuel up to an additional 105.6 GWd/MTIHM and 66 GWd/MTIHM, for melt refining and AIROX, respectively.