DOCSLIB.ORG

Rbmk Fuel Assemblies: Current Status and Perspectives

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Rbmk Fuel Assemblies: Current Status and Perspectives RBMK FUEL ASSEMBLIES: CURRENT STATUS AND PERSPECTIVES A.I. KUPALOV-YAROPOLK, V.A. NIKOLAEV, YU.M. CHERKASHOV Research and Development Institute of Power Engineering, Moscow A.K. PANYUSHKIN XA9846755 Machine Construction Company, Ehlectrostal, Moscow region A.M. FEDOSOV Russian Scientific Center Kurchatov Institute (RSC KI), Moscow Russian Federation Abstract The safety enhancement measures implemented since 1986 have led to substantial burnup reduction in the RBMK fuel assemblies and consequently to economical losses. With the purpose to compensate the losses, computer analysis and experiments were performed during the last decade. The works were aimed at the RBMK fuel charge perfection to reduce void reactivity effect and to increase fuel burnup. The paper presents principle results of the studies which are currently under implementation or are supposed to be implemented in the nearest future. Approach to fuel assembly design modification As one result of the safety enhancement measures implemented after 1986, in particular, by installing a large number of additional absorbers to the core, the fuel burnup of the RBMK fuel assemblies (FAs) was reduced by 30-35 % compared to the design value. To compensate the losses, computer analysis and experimental works aimed at modifying the RBMK fuel charge were performed during the last decade with the purpose to reduce void reactivity effect and increase the fuel burnup. The activities on fuel assemblies modification were performed in two lines: • with change in geometrical dimensions of fuel assemblies involved; • without changing basic overall dimensions. One RBMK FA for both the 1000 MW reactors and RBMK-1500 contains 18 fuel element of 13.6 mm diameter, all are located over two circumferences: the first radius for 6 fuel elements and the second one - for 12 fuel elements (Fig. 1). The first line, aimed at alternating the FA design, envisaged development of two options, i.e. with fuel elements of ether diminished or enlarged diameters. The fuel elements designs 10.2 mm and 14.8 in diameter are worth noting. In the first case, the FA incorporated fuel elements of three radii, in the second case, it had two, the same as in the standard case. The reactor process channel was also envisaged to get larger in diameter where in a fuel assembly of 80 to 86 mm diameter was installed. The analysis showed that the change for 10.2 mm fuel elements practically did not change the void reactivity effect. Thus, it makes up is ~0.9 P for the 2 % enriched fuel in the core with 80 additional absorbers, whereas for the 10.2 mm fuel elements it is ~1.0 (3. The fuel elements with diameters enlarged from 13.6 up to 14.8 mm allows additional absorbers to be discharged from the core and reduce the cost by ~15 %. In this case the fuel assembly uranium charge gets higher by -20 %. However, bearing in mind, that actual change for channels with large 143 I i t 144 diameter may be realised only when the core is updated as well as the RBMK units service life design terms, this line did not get progression. Therefore, the investigations were reoriented during last several years such that application of burnable absorbers should allow the geometry of the fuel assembly and fuel channel to remain unchanged. Development and implementation of burnable absorber fuel A potentiality for application of different burnable absorbers: B, Dy, Gd, Hf, Er was subjected to analysis. Also different options of their location in fuel assemblies (in structure components, partially in fuel elements, etc.) were considered. The analysis performed showed that erbium may get to be better absorber for the RBMK reactors. The options considered for Er location in FAs allowed to conclude that the highest effect may be attained when Er is added directly to the fuel of all FA fuel elements. The calculations demonstrated that 0.4 % of Er added to the 2.4 % enrichment fuel enables to practically have the same voidage effect in the core with no additional absorbers (AA) as in the reactor with 80 AAs (Table 1). At the same time, application of additional absorbers permits to substantially reduce (-10 %) the FA maximum power which realises at the initial stage of the FA operation (Fig. 2). This margin can be used for increasing the fuel enrichment. The calculations illustrated that Er application in RBMK allows to pass to 2.4 % enrichment without exceeding the FA maximum power limits. The higher enrichment and discharge of additional absorbers from the core lead to a fuel burnup higher by 30 %. The fuel burnup for the RBMK-1000 reactors get higher by 18 % as the enrichment is changed for 2.6%. The presence of erbium in fuel results in a substantial (by 10-15 %) reduction in the fresh fuel reactivity as compared to standard which allows FA reload that are performed in RBMKs on power by more "mild" control. This does not require significant movement of control rods to suppress local peaks in power density. TABLE 1. BASIC CALCULATED CHARACTERISTICS OF RBMK REACTOR CORES WITH NEW FUEL Characteristics RBMK-1000 RBMK-1500 Enrichment, % 2.4 2.6 3.0 2.0 2.4 Erbium content by mass, % without Er 0.41 0.76 without Er 0.41 Number of AA in core 80 0 0 54 0 Discharged fuel burnup, MW day/kgU 20.9 26.6 28.8 15.1 21.0 Average-over-service life voidage, P +0.6 +0.5 -0.39 +0.7 +0.3 Annual economic effect due to reduction in fuel - 4.0 - - 9.0 component per 1 power unit, mln. USD Possible date of FA serial production commencement 1997 1999 - 1997 - 145 FC ultimate power 4,0 3,5 - 3,0 2,5 - - 2,0 10 15 20 25 Fuel burnup, MWd/kg U 2.4% without erbium; 2.4%+0.4% of erbium; standard FC 2% Fig. 2. Maximum FA power versus burnup. 146 On completing preliminary computer studies in 1993, the activities on introduction of uranium- erbium fuel were developed along the lines as follows: • computer studies of reactor neutronics as experimental erbium FA (EFA) lots are installed into core, optimisation of RBMK units on changing for uranium-erbium fuel; • computer studies of the core thermal and engineering reliability, analysis of emergency situations; • fuel properties studies, fuel element (FE) design justification; • optimisation of fuel pellet technology, techniques for their quality control, preparation of production for manufacturing experimental lots; • computer safety analysis of cooling ponds with spent fuel discharged therein. Basic results of the above activity lines are: Full-scale core calculations confirmed the said neutronics characteristics of the core with uranium- erbium fuel. The calculated analysis which applies the most conservative approaches showed that emergency situations in a core with uranium-erbium fuel runs in a more "mild" manner. Similarity in properties of uranium-erbium and standard fuels permitted to apply the FE design practically without any changes. The pellets with a central hole are recommended for application to reduce temperature levels. The technological flow-chart for manufacturing uranium-erbium pellets is slightly different from the standard one. The calculations performed are quite natural to require to be confirmed by in-pile experiments. To this end, in 1995 the first lot (150 pieces) of the RBMK-type fuel assemblies with uranium-erbium (UEFA) was manufactured. It was charged into the Ignalina Nuclear Power Plant (INPP) Unit 2 from June through January, 1996. The computer prognosis predicted reduction in void reactivity effect resulting from the experimental lot charge of 0.4 p. The EFAs charge started at INPP on June 26, 1995. Following the test program on charging every 25 EFAs, the core basic neutronic characteristics were to be measured, first of all the void reactivity effect. The results obtained are presented in Table 2. The value of the void reactivity effect reduced from 0.8 P down to 0.5 p. The maximum power of one fuel channel with EFAs did not overcome the maximum power of a FA with the standard fuel. Bearing in mind positive results of the first stage, it was decided that the tests will be expanded and the EFA (500 pieces) batch will be charged together with simultaneous discharge of additional absorbers. At INPP-2 more than 100 additional EFAs are so far charged and more than 260 EFAs are presently under operation. Experimental results are also consistent with computer predictions. By the end of 1996 about 25 EFAs will attain the bumup of 10.0 MW d/kg U. The first batch of the RBMK-1000 EFAs with fuel enrichment of 2.6 % are presently installed at Leningrad NPP. The documents for charging the EFAs lots at the Chernobyl and Kursk NPPs are under agreement. It is scheduled that RBMK units will be transferred to uranium-erbium fuel by 2000. 147 l 00 TABLE 2. REACTOR PERFORMANCE IN THE COURSE OF LOADING EXPERIMENTAL LOT OF ERBIUM FUEL ASSEMBLIES Date 17.03.95 08.08.95 15.08.95 08.12.95 29.01.96 25.06.95 30.06.96 16.07.96 23.07.96 Number of EFAs 0 50 74 100 150 195 195 236 240 Thermal power, MW 4060 4000 3420 3900 3700 3400 3370 3960 3470 FA average power production, 858 846 834 851 842 848 859 858 868 MW day/FA Operative reactivity 53.9 54.2 56.9 53.7 54.6 55.0 55.9 54.5 56.4 margin, MC rods Number of A A 53 53 53 53 53 53 49 49 45 0.83 0.7 0.6 0.7 0.5 0.6 0.6 0.6 0.7 Improvement of RBMK fuel reliability and burnup increase For the purpose of attaining higher fuel burnup and FAs reliability simultaneous activities are in progress along the following lines: • further increase in enrichment of fuel of RBMK-1000 EFAs; • introduction of zirconium grids; • design change of FE fastening in FA for improving fuel operating conditions, among those for operation according to scheduled variable loads; • repair of defected FAs and their repeated use.
Load more

Recommended publications
  • Vver and Rbmk Cross Section Libraries for Origen-Arp
    VVER AND RBMK CROSS SECTION LIBRARIES FOR ORIGEN-ARP Germina Ilas, Brian D. Murphy, and Ian C. Gauld, Oak Ridge National Laboratory, USA Introduction An accurate treatment of neutron transport and depletion in modern fuel assemblies characterized by heterogeneous, complex designs, such as the VVER or RBMK assembly configurations, requires the use of advanced computational tools capable of simulating multi-dimensional geometries. The depletion module TRITON [1], which is part of the SCALE code system [2] that was developed and is maintained at the Oak Ridge National Laboratory (ORNL), allows the depletion simulation of two- or three-dimensional assembly configurations and the generation of burnup-dependent cross section libraries. These libraries can be saved for subsequent use with the ORIGEN-ARP module in SCALE. This later module is a faster alternative to TRITON for fuel depletion, decay, and source term analyses at an accuracy level comparable to that of a direct TRITON simulation. This paper summarizes the methodology used to generate cross section libraries for VVER and RBMK assembly configurations that can be employed in ORIGEN-ARP depletion and decay simulations. It briefly describes the computational tools and provides details of the steps involved. Results of validation studies for some of the libraries, which were performed using isotopic assay measurement data for spent fuel, are provided and discussed. Cross section libraries for ORIGEN-ARP Methodology The TRITON capability to perform depletion simulations for two-dimensional (2-D) configurations was implemented by coupling of the 2-D transport code NEWT with the point depletion and decay code ORIGEN-S. NEWT solves the transport equation on a 2-D arbitrary geometry grid by using an SN approach, with a treatment of the spatial variable that is based on an extended step characteristic method [3].
    [Show full text]
  • The Effect of Burnup and Separation Efficiency on Uranium Utilization and Radiotoxicity
    INL/CON-10-20154 PREPRINT The Effect of Burnup and Separation Efficiency on Uranium Utilization and Radiotoxicity Eleventh Information Exchange Meeting on Partitioning and Transmutation Samuel Bays Steven Piet November 2010 This is a preprint of a paper intended for publication in a journal or proceedings. Since changes may be made before publication, this preprint should not be cited or reproduced without permission of the author. This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party’s use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the United States Government or the sponsoring agency. THE EFFECT OF BURNUP AND SEPARATION EFFICIENCY ON URANIUM UTILIZATION AND RADIOTOXICITY Samuel Bays, Steven Piet Idaho National Laboratory, United States Abstract This paper addresses two fundamental cradle-to-grave issues of fuel cycle sustainability. The two primary issues of interest are efficient use of the natural uranium resource (cradle), and management of nuclear waste radiotoxicity (grave). Both uranium utilization and radiotoxicity are directly influenced by the burnup achieved during irradiation (transmutation related) and where applicable the separation efficiency (partitioning related). Burnup influences the in-growth of transuranics by breeding them into the fuel cycle.
    [Show full text]
  • Nuclear Fuel Cycles Technology Assessment
    Clean Power Quadrennial Technology Review 2015 Chapter 4: Advancing Clean Electric Power Technologies Technology Assessments Advanced Plant Technologies Biopower Carbon Dioxide Capture and Storage Clean Power Value-Added Options Carbon Dioxide Capture for Natural Gas and Industrial Applications Carbon Dioxide Capture Technologies Carbon Dioxide Storage Technologies Crosscutting Technologies in Carbon Dioxide Capture and Storage Fast-spectrum Reactors Geothermal Power High Temperature Reactors Hybrid Nuclear-Renewable Energy Systems Hydropower Light Water Reactors Marine and Hydrokinetic Power Nuclear Fuel Cycles Solar Power Stationary Fuel Cells Supercritical Carbon Dioxide Brayton Cycle U.S. DEPARTMENT OF Wind Power ENERGY Clean Power Quadrennial Technology Review 2015 Nuclear Fuel Cycles Chapter 4: Technology Assessments Introduction and Background The Nuclear Fuel Cycle (NFC) is defined as the total set of operations required to produce fission energy and manage the associated nuclear materials. It can have different attributes, including the extension of natural resources, or the minimization of waste disposal requirements. The NFC, as depicted in Figure 4.O.1, is comprised of a set of operations that include the extraction of uranium (U) resources from the earth (and possibly from seawater), uranium enrichment and fuel fabrication, use of the fuel in reactors, interim storage of used nuclear fuel, the optional recycle of the used fuel, and the final disposition of used fuel and waste forms from the recycling processes. Thorium (Th) fuel cycles have been proposed also, but have not been commercially implemented). Figure 4.O.1 Schematic of the uranium based Nuclear Fuel Cycle 1 Quadrennial Technology Review 2015 Clean Power TA 4.O: Nuclear Fuel Cycles The nuclear fuel cycle is often grouped into three classical components (front-end, reactor, and back-end): Front End: The focus of the front end of the nuclear fuel cycle is to deliver fabricated fuel to the reactor.
    [Show full text]
  • The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation
    GNA bNLI-I MOL.19980810.0383 PNNL-11 929 Nationail Labolratory Opeate Batelb fornimthe"• The Burnup Dependence of Light Water Reactor Spent Fuel Oxidation B. D. Hanson July 1998 z"D z co Prepared for the U.S. Department of Energy under Contract DE-AC06-76RLO 1830 ' DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTEELLE for the UNITED STATES DEPARTMENT OF ENERGY under ContractDE-ACO6-76RLO 1830 Printed in the United States of America Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; prices available from (615) 576-8401. Available to the public from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal Rd., Springfield, VA 22161 This document was printed on recycled paper.
    [Show full text]
  • Nuclear Reactors and Plutonium Production
    Nuclear Reactors and Plutonium Production ISIS Course-Week 4 October 23, 2014 Nuclear Reactors • A nuclear reactor is an apparatus or machine in which fissile material can be made to undergo a controlled, self-sustaining nuclear reaction with the consequent release of energy. The Basic Parts of a Reactor http://en.wikipedia.org/wiki/Magnox • Fuel region • Heat removal equipment • Control system • Refueling capability Purposes of Reactors • Nuclear reactors are used for civilian purposes to – generate electricity – produce heat via steam – produce isotopes for medical, industrial, or research purposes – conduct research – propel surface ships, such as ice-breakers • Nuclear reactors are also used for military purposes to – produce plutonium for nuclear weapons – produce tritium for nuclear weapons – propel military submarines and surface ships – power aircraft and rockets (abandoned) Classification of Reactors • Power reactors – typically civilian, almost all now generate electricity. Some produced steam for district heating. Typically make large quantities of reactor-grade plutonium. • Research reactors – typically civilian aimed at making isotopes or conducting research. – In non-proliferation context, ostensibly civilian research reactors have also been used to make plutonium for nuclear weapons, e.g. Cirus and Dhruva reactors in India. The misuse of the Taiwan research reactor was planned. Is the Arak reactor in Iran in this category, ostensibly a civilian research reactor but in reality it would also have produced plutonium for nuclear weapons? • Propulsion reactors – either civilian or military • Production reactors – almost always military and dedicated to plutonium and tritium production for nuclear weapons, which means that the goal is the production of weapon- grade plutonium. Reactors • Reactor types vary greatly, some are far simpler to build than others • Production reactors and research reactors are the simplest to build.
    [Show full text]
  • Fuel Burnup and Enrichment Extension Preparation Strategy
    1 APPENDIX A: FUEL BURNUP AND ENRICHMENT EXTENSION 2 PREPARATION STRATEGY 3 4 Based on stakeholder interactions, the NRC staff is aware of industry’s plans to request higher 5 fuel burnup limits along with the deployment of near-term ATF concepts. Additionally, the staff 6 expects that the extension of fuel burnup limits, and the economic drive to achieve those 7 burnups, will result in requests to increase fuel enrichments to greater than the current standard 8 of 5 weight percent U235. Therefore, the staff is proactively assessing the current knowledge 9 and experimental database associated with extending both burnup and enrichment for light 10 water reactor (LWR) fuels. This plan focuses on the strategy to prepare the NRC for review of 11 future licensing actions in which industry requests to go beyond current licensed limits with 12 burnups up to ~75 GWd/MTU rod-average and enrichments up to ~8 weight percent U235. Staff 13 will continue to engage with industry and the fuel vendors on these topics and adjust this 14 strategy as industry plans for higher burnup and increased enrichments evolve. 15 16 Overview of Preparatory Activities 17 18 As with other ATF activities related to advanced cladding and fuel materials, the staff has 19 grouped its burnup and enrichment preparatory activities into four tasks. The highlights of each 20 task are briefly described below; subsequent sections within this appendix describe these tasks 21 in greater detail. 22 23 Task 1: Regulatory Framework: In-Reactor Performance 24 • Participate in coordinated PIRT exercises on in-reactor performance of fuels with 25 increased enrichment under a wide array of conditions, performance -based metrics, and 26 analytical criteria to ensure acceptable performance.
    [Show full text]
  • Burnup Determinaton of Nuclear Fuel
    MASS SPECTROSCOPY Original Papers Vol17.No4,December1969 Burnup Determinaton of Nuclear Fuel KIYOSHI I NOUE,*KAORU TANIGUCHI,*TOSHIFUMI MURATA,** HIDEHIKO MITSUI*AND AKIRA DOI* (Received August16,1969) A general description of burnup determining experiments is presented for making vivid a part of mass spectrometry,in comparison of results obtained from radiochemical and nondestructive analyses. Specimen of nuclear fuel was grains of uranium dioxide irradiated in a nuclear reactor for about200 days,and its burnup was found to be8,000MWD/T with sufficient consistency in the experimental results. 1.Introduction method is included in the first approach, Extensive studies on nuclear fuel and the second approach includes radio burnup analysis have been made in the chemical and nondestructive methods. past,since burnup is one of fundamental Both of the radiochemical and nondes quantities required in a fuel research structive methods essentially measure program.Burnup is also an important radioactivities of fission products formed quantity for design and operation of in a nuclear fuel.The former has an ac power reactors from a standpoint of ceptable precision with low cost and safety as well as operability. simple procedure,but the development A wide variety of methods have of a good method for the later remains been developed to determine burnup of an engineering target. irradiated fuel.Principal procedures This paper presents a general de used so far have been based on meas scription on the burnup determining ex ments of: periments for a uranium dioxide fuel (1)Changes in isotopic composition of specimen.Comparing the mass spectro fissioning elements in a nuclear metry method with the other methods, fuel brought about during irradia the contribution it can make to the tion.
    [Show full text]
  • Burnup Credit Criticality Safety Analysis Using STARBUCS
    ACTIVITIES OF THE OECD/NEA EXPERT GROUP ON ASSAY DATA FOR SPENT NUCLEAR FUEL I. C. Gauld Oak Ridge National Laboratory P.O. Box 2008 Oak Ridge, TN 37831-6370 USA [email protected] Y. Rugama OECD Nuclear Energy Agency 12, Bd des Iles – 92130 Issy-les-Moulineaux, France Management of spent nuclear fuel is a key issue for many NEA member countries. In nuclear criticality safety, the decision of many countries to advance burnup credit as part of their licensing strategy has heightened recent interest in experimental data needed to validate computer codes used in burnup credit calculations. This paper discusses recent activities of an Expert Group on assay data, formed under the OECD/NEA/NSC/WPNCS (Working Party on Nuclear Criticality Safety) to help coordinate isotopic assay data activities and facilitate international collaboration between NEA member countries developing or implementing burnup credit methodologies. Recent activities of the Expert Group are described, focusing on the planned expansion of the Spent Fuel Isotopic Composition Database (SFCOMPO), and preparation of a state-of-the-art report on assay data that includes sections on recommended radiochemical analysis methods, techniques, and lessons learned from previous experiments. I. INTRODUCTION Management of spent nuclear fuel is a key issue for many Nuclear Energy Agency (NEA) member countries. As interim storage facilities in many countries reach their design capacities, the need to optimize spent fuel storage is becoming an increasingly important issue to managing fuel cycle costs while reducing associated risks. In nuclear criticality safety, the decision by many countries to advance burnup credit as part of their licensing strategy has heightened recent interest in measurement data that are needed to validate computer code calculations for a burnup credit methodology.
    [Show full text]
  • Reactor Fuel Isotopics and Code Validation for Nuclear Applications
    ORNL/TM-2014/464 Reactor Fuel Isotopics and Code Validation for Nuclear Applications Matthew W. Francis Charles F. Weber Marco T. Pigni Approved for public release; Ian C. Gauld distribution is unlimited. September 2014 DOCUMENT AVAILABILITY Reports produced after January 1, 1996, are generally available free via US Department of Energy (DOE) SciTech Connect. Website http://www.osti.gov/scitech/ Reports produced before January 1, 1996, may be purchased by members of the public from the following source: National Technical Information Service 5285 Port Royal Road Springfield, VA 22161 Telephone 703-605-6000 (1-800-553-6847) TDD 703-487-4639 Fax 703-605-6900 E-mail [email protected] Website http://www.ntis.gov/help/ordermethods.aspx Reports are available to DOE employees, DOE contractors, Energy Technology Data Exchange representatives, and International Nuclear Information System representatives from the following source: Office of Scientific and Technical Information PO Box 62 Oak Ridge, TN 37831 Telephone 865-576-8401 Fax 865-576-5728 E-mail [email protected] Website http://www.osti.gov/contact.html This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof.
    [Show full text]
  • Performance Assessment of Spent Fuel Assemblies Removed from Shutdown Rbmk-1000 Power Units for Reburning in Operating Power Units
    PERFORMANCE ASSESSMENT OF SPENT FUEL ASSEMBLIES REMOVED FROM SHUTDOWN RBMK-1000 POWER UNITS FOR REBURNING IN OPERATING POWER UNITS B. Kanashov, I. Kuzmin, A. Kostyuchenko, S. Perepelkin, V. Chesanov (Sosny R&D Company) INTRODUCTION The scheduled operating life-time of RBMK-1000 reactor units, that is 30 years, has already expired. Comprehensive analysis of the reactor conditions allowed to extend the operating life-time of RBMK units up to 45 years [1]. The additional time to operate each reactor will depend on its peculiarities, however, we can predict the NPP will shut down its reactors in the order of their commissioning. Table 1. RBMK units decommissioning schedule subject to their extended operating life-time NPP Unit 1 Unit 2 Unit 3 Unit 4 Leningrad 2018 2020 2024 2026 Kursk 2020 2023 2028 2030 Smolensk 2027 2030 2035 – A peculiarity of the RBMK reactors is non-stop core refueling. On the average, one burnt-up fuel assembly (FA) is replaced for a fresh one daily. Sometimes the irradiated assemblies that have not reached the maximum design burnup are reinserted into the core, which already contains the assemblies of different burnups. FA distribution in terms of burnup depends on the duration of the previous period. If the reactor was loaded with the same type fuel with approximately the same speed, the FA burnup distribution in the core does not change. In this case, the refueling is steady-state. Over the period of RBMK reactors operation, the enrichment of fresh fuel has experienced a staged growth. Nowadays, RBMK-1000 units are practically converted to uranium-erbium fuel enriched to 2.8 % in U-235 with erbium content 0.6 %.
    [Show full text]
  • Advances by the Integral Fast Reactor Program De91 011174
    I A // / A / ANL/CP--71374 ADVANCES BY THE INTEGRAL FAST REACTOR PROGRAM DE91 011174 M. J. Lineberry, D. R. Pedersen, of the EBR-11 reactor reactor with a uranium alloy metallic L. C. Waiters, and J. E. Cahalan fuel. ll Argonne National Laboratory f 9700 S. Cass Avenue The IFR safety approach for the Integral Fast Reactor (IFR) Argonne, Illinois 60439 concept capitalizes on the characteristics of metallic fuels and of pool-type liquid metal reactors to provide enhanced ABSTRACT safety margins. The fundamental safety approach guiding The advances by the Integral Fast Reactor Program at the IFR program is: Argonne National Laboratory are the subject of this paper. ; 1. A simple, economic, high quality fuel system must The Integral Fast Reactor (IFR) is an advanced liquid- ', be developed which will allow normal operation of metal-cooled reactor concept being developed at Argonne j reprocessed fuel with minimal fuel failures. National Laboratory. The advances stressed in the paper : 2. The metallic fuel system must be tolerant of fuel include fuel irradiation performance, improved passive , failures and local faults. safety, ^nd the development of a prototype fuel cycle ' 3. The reactor design must be such that for any system facility. failure, including those in the balance of plant, no active systems have to operate to maintain the INTRODUCTION reactor in a safe state. The Integral Fast Reactor (IFR) (Ref. 1) is an advanced 4. Even though the reactor can achieve passive shut- liquid-metal-cooled reactor concept being developed at down for any system failure, the reactor should be Axgonne National Laboratory.
    [Show full text]
  • HTR-N Plutonium Cell Burnup Benchmark: Definition, Results & Intercomparison
    PHYSOR 2004 -The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments Chicago, Illinois, April 25-29, 2004, on CD-ROM, American Nuclear Society, Lagrange Park, IL. (2004) HTR-N Plutonium Cell Burnup Benchmark: Definition, Results & Intercomparison J.C. Kuijper 1, N. Cerullo5, F. Damian2, J.L. Kloosterman4, G. Lomonaco5, J. Oppe1, X. Raepsaet2, H.J. Ruetten3 1NRG, Petten, the Netherlands 2CEA, Saclay, France 3FZJ, Juelich, Germany 4IRI, Delft University of Technology, Delft, the Netherlands 5Universita di Pisa, Pisa, Italy In order to obtain a further validation of several reactor physics code systems to be used for the analysis of High Temperature gas-cooled Reactors (HTR) for plutonium burning applications, a HTR plutonium cell burnup calculational benchmark exercise was defined. For simplicity, the benchmark configuration only consists of a single spherical, 6 cm diameter HTR (“pebble”) fuel element containing coated (PuO2) fuel particles. The requested calculations concerned the tracking of the multiplication factor, nuclide densities in the fuel particles and other relevant parameters during the irradiation of the fuel element at constant power up to the unusually high burnup of 800 MWd/kgHM. Generally a good agreement is found between the results of 3 out of 4 participants, representing 4 out of 5 code systems. The remaining differences in results between the 3 participants can be largely attributed to differences in the modeling of the reaction paths, which are amplified by the unusually high flux levels typical to this particular benchmark. KEYWORDS: HIGH TEMPERATURE GAS-COOLED REACTOR, PLUTONIUM BURNING, HTR-N, EUROPEAN UNION FIFTH FRAMEWORK PROGRAM 1. Introduction In the “HTR-N” and “HTR-N1” projects of the European Fifth Framework Program [1-3] investigations are being made concerning the feasibility of burning plutonium in “High Temperature Gas-Cooled Reactors (HTR)”.
    [Show full text]