DOCSLIB.ORG

Spent Fuel Performance Assessment and Research

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text

Load more

IAEA-TECDOC-1343 Spent fuel performance assessment and research Final report of a Co-ordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR) 1997–2001 March 2003 The originating Section of this publication in the IAEA was: Nuclear Fuel Cycle and Materials Section International Atomic Energy Agency Wagramer Strasse 5 P.O. Box 100 A-1400 Vienna, Austria SPENT FUEL PERFORMANCE ASSESSMENT AND RESEARCH IAEA, VIENNA, 2003 IAEA-TECDOC-1343 ISBN 92–0–102703–6 ISSN 1011–4289 © IAEA, 2003 Printed by the IAEA in Austria March 2003 FOREWORD As of the beginning of 2002, more than 150 000 Mg U spent fuel is currently being stored in various storage facilities. Most of this fuel is under water, but dry storage is becoming a widely used technology with more than 9000 Mg U currently stored in various countries. In the future spent fuel storage capacity needs will continue to increase and some of these fuels will have to be stored for 50 years or more, before their reprocessing or final disposal takes place. As a result, interim storage for long periods of time will continue to be a key technology for all Member States. A number of countries are planning or have already initiated research programmes on spent fuel storage performance, or are developing a high confidence database which evaluates the storage of spent fuel for extremely long periods of time, to determine their consequence on disposal. Spent fuel storage technology (particularly dry storage) is undergoing a rapid evolution, new fuel and material design changes are coming on stream and target burnups are steadily increasing. These improvements will ask for new studies and potential adaptations of the storage technologies currently used. As a consequence, the IAEA initiated in 1997 a Co-ordinated Research Project on Spent Fuel Performance Assessment and Research (SPAR) to collect and exchange spent fuel storage experience of the participating countries, to build a comprehensive international database supporting the licensing of present and future technologies; to carry out research work which will evaluate and justify the storage of spent fuel for extremely long periods of time (more than 50 years); and to assist in defining how the requirements for spent fuel storage and for the whole back end of the fuel cycle are connected. The present publication is based on results obtained in the participating countries. The draft manuscript was prepared and discussed during the last research co-ordination meeting, held in Cordoba, Spain, 1–5 October 2001, and finalized at a consultants meeting in February 2002. The report provides an overview of technical issues related to spent fuel wet and dry storage and summarizes the objectives and major findings of research, carried out within the framework of the CRP. The IAEA gratefully acknowledges the contribution of the CRP participants and the consultants who participated in the drafting and review of the report. The IAEA staff member responsible for this publication was H.P. Dyck of the Division of Nuclear Fuel Cycle and Waste Technology. EDITORIAL NOTE The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal status of such countries or territories, of their authorities and institutions or of the delimitation of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. CONTENTS 1. INTRODUCTION AND OBJECTIVES............................................................................ 1 2. HISTORY OF THE BEFAST AND SPAR CO-ORDINATED RESEARCH PROJECTS .................................................................... 2 3. WET STORAGE ................................................................................................................ 4 3.1. Spent fuel storage experience.................................................................................... 4 3.1.1. General performance...................................................................................... 4 3.1.2. Wet storage experience in the participating countries ................................... 5 3.2. Pool water chemistry................................................................................................. 9 3.2.1. Water quality control ................................................................................... 14 3.2.2. Sources of pool water activity...................................................................... 15 3.3. Inactive content ....................................................................................................... 16 3.3.1. Sources of pool water particulates ............................................................... 17 3.3.2. Water treatment............................................................................................ 17 3.3.3. Water temperature........................................................................................ 18 3.4. Pool components and materials............................................................................... 18 3.4.1. Pool lining.................................................................................................... 18 3.4.2. Storage racks................................................................................................ 19 4. DRY STORAGE .............................................................................................................. 19 4.1. Introduction ............................................................................................................. 19 4.2. Dry storage status and experience........................................................................... 21 4.3. Research activities related to dry storage................................................................ 26 5. SOURCE TERMS ............................................................................................................ 30 5.1. Subcriticality ........................................................................................................... 31 5.2. Thermal calculations ............................................................................................... 32 5.3. Radiation sources .................................................................................................... 32 5.4. Aspects associated with high burnup fuel ............................................................... 33 5.5. Aspects associated with burnup credit .................................................................... 35 6. FUEL INTEGRITY.......................................................................................................... 38 6.1. Definition of fuel integrity ...................................................................................... 38 6.1.1. Loss of fuel cladding integrity ..................................................................... 41 6.1.2. Loss of fuel assembly lifting features .......................................................... 41 6.1.3. Geometrical rearrangement of a fuel assembly ........................................... 41 6.2. Determination of spent fuel integrity ...................................................................... 41 6.2.1. At reactor detection systems ........................................................................ 41 6.2.2. Away from reactor detection systems.......................................................... 44 6.3. Criteria..................................................................................................................... 45 6.3.1. Utilities......................................................................................................... 45 6.3.2. Transportation.............................................................................................. 45 6.3.3. AFR facilities............................................................................................... 45 7. FUEL ASSEMBLY DEGRADATION MECHANISMS IN DRY AND WET STORAGE ........................................................................................... 45 7.1. Degradation mechanisms of fuel assembly components......................................... 45 7.1.1. LWR............................................................................................................. 45 7.1.2. AGR............................................................................................................. 46 7.1.3. Magnox ........................................................................................................ 47 7.1.4. CANDU....................................................................................................... 47 7.1.5. RMBK andWWER-440/1000...................................................................... 49 7.2. Degradation mechanisms affecting the fuel cladding ............................................. 49 7.2.1. LWR............................................................................................................. 49 7.2.2. Wet storage .................................................................................................. 50 7.2.3. Dry storage................................................................................................... 51 7.2.4. AGR............................................................................................................. 57 7.2.5.
Recommended publications
  • Spent Nuclear Fuel Pools in the US

    Spent Nuclear Fuel Pools in the US

    Spent Nuclear Fuel Pools in the U.S.: Reducing the Deadly Risks of Storage front cover WITH SUPPORT FROM: WITH SUPPORT FROM: By Robert Alvarez 1112 16th St. NW, Suite 600, Washington DC 20036 - www.ips-dc.org May 2011 About the Author Robert Alvarez, an Institute for Policy Studies senior scholar, served as a Senior Policy Advisor to the Secre- tary of Energy during the Clinton administration. Institute for Policy Studies (IPS-DC.org) is a community of public scholars and organizers linking peace, justice, and the environment in the U.S. and globally. We work with social movements to promote true democracy and challenge concentrated wealth, corporate influence, and military power. Project On Government Oversight (POGO.org) was founded in 1981 as an independent nonprofit that investigates and exposes corruption and other misconduct in order to achieve a more effective, accountable, open, and ethical federal government. Institute for Policy Studies 1112 16th St. NW, Suite 600 Washington, DC 20036 http://www.ips-dc.org © 2011 Institute for Policy Studies [email protected] For additional copies of this report, see www.ips-dc.org Table of Contents Summary ...............................................................................................................................1 Introduction ..........................................................................................................................4 Figure 1: Explosion Sequence at Reactor No. 3 ........................................................4 Figure 2: Reactor No. 3
  • Of Spent Fuel Pools: Tool Survey Scenarios, Technology Considerations, and Evaluation Criteria

    Of Spent Fuel Pools: Tool Survey Scenarios, Technology Considerations, and Evaluation Criteria

    PNNL-25137 Maintaining Continuity of Knowledge (CoK) of Spent Fuel Pools: Tool Survey Scenarios, Technology Considerations, and Evaluation Criteria January 2016 JM Benz, PNNL JE Tanner, PNNL HA Smartt, SNL MR MacDougall, PNNL PNNL-25137 Maintaining Continuity of Knowledge (CoK) of Spent Fuel Pools: Tool Survey JM Benz, PNNL JE Tanner, PNNL HA Smartt, SNL MR MacDougall, PNNL January 2016 Prepared for the U.S. Department of Energy under Contract DE-AC05-76RL01830 Pacific Northwest National Laboratory Richland, Washington 99352 Contents 1.0 Introduction ................................................................................................................................ 1 2.0 Scenarios ..................................................................................................................................... 1 2.1 Spent Fuel Pools ................................................................................................................. 1 2.2 Spent Fuel Pool Monitoring ............................................................................................... 3 2.3 Bounding Scenarios............................................................................................................ 4 3.0 Technology Considerations/Constraints ..................................................................................... 5 4.0 Evaluation Criteria ...................................................................................................................... 6 5.0 Conclusion .................................................................................................................................
  • The Nuclear Waste Primer September 2016 What Is Nuclear Waste?

    The Nuclear Waste Primer September 2016 What Is Nuclear Waste?

    The Nuclear Waste Primer September 2016 What is Nuclear Waste? Nuclear waste is the catch-all term for anything contaminated with radioactive material. Nuclear waste can be broadly divided into three categories: • Low-level waste (LLW), comprised of protective clothing, medical waste, and other lightly-contaminated items • Transuranic waste (TRU), comprised of long-lived isotopes heavier than uranium • High-level waste (HLW), comprised of spent nuclear fuel and other highly-radioactive materials Low-level waste is relatively short-lived and easy to handle. Currently, four locations for LLW disposal exist in the United States. Two of them, Energy Solutions in Clive, Utah and Waste Control Specialists in Andrews, Texas, accept waste from any U.S. state. Transuranic waste is often a byproduct of nuclear weapons production and contains long-lived radioactive elements heavier than uranium, like plutonium and americium. Currently, the U.S. stores TRU waste at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. High-level waste includes spent nuclear fuel and the most radioactive materials produced by nuclear weapons production. Yucca Mountain is the currently designated high-level waste repository for the United States. 1 | What is Spent Nuclear Fuel? Spent nuclear fuel (SNF), alternatively referred to as used nuclear fuel, is the primary byproduct of nuclear reactors. In commercial power reactors in the U.S., fuel begins as uranium oxide clad in a thin layer of zirconium-aluminum cladding. After several years inside of the reactor, around fi ve percent of the uranium has been converted in some way, ranging from short-lived and highly radioactive fi ssion products to long-lived actinides like plutonium, americium, and neptunium.
  • SAFETY RE-ASSESSMENT of AECL TEST and RESEARCH REACTORS D. J. WINFIELD Chalk River Nuclear Laboratories ATOMIC ENERGY of CANADA

    SAFETY RE-ASSESSMENT of AECL TEST and RESEARCH REACTORS D. J. WINFIELD Chalk River Nuclear Laboratories ATOMIC ENERGY of CANADA

    309 IAEA-SM-310/ 94 SAFETY RE-ASSESSMENT OF AECL TEST AND RESEARCH REACTORS D. J. WINFIELD Chalk River Nuclear Laboratories ATOMIC ENERGY OF CANADA LIMITED 310 IAEA-SM-310/94 SAFETY RE-ASSESSMENT OF AECL TEST AND RESEARCH REACTORS ABSTRACT Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor MAPLE-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration/test reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX and NRU). ZED-2 and PTR are operated by the Reactor Physics Branch, all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault tree analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis and indicates the plan for future analysis.
  • Nuclear Transmutation Strategies for Management of Long-Lived Fission

    Nuclear Transmutation Strategies for Management of Long-Lived Fission

    PRAMANA c Indian Academy of Sciences Vol. 85, No. 3 — journal of September 2015 physics pp. 517–523 Nuclear transmutation strategies for management of long-lived fission products S KAILAS1,2,∗, M HEMALATHA2 and A SAXENA1 1Nuclear Physics Division, Bhabha Atomic Research Centre, Mumbai 400 085, India 2UM–DAE Centre for Excellence in Basic Sciences, Mumbai 400 098, India ∗Corresponding author. E-mail: [email protected] DOI: 10.1007/s12043-015-1063-z; ePublication: 27 August 2015 Abstract. Management of long-lived nuclear waste produced in a reactor is essential for long- term sustenance of nuclear energy programme. A number of strategies are being explored for the effective transmutation of long-lived nuclear waste in general, and long-lived fission products (LLFP), in particular. Some of the options available for the transmutation of LLFP are discussed. Keywords. Nuclear transmutation; long-lived fission products; (n, γ ) cross-section; EMPIRE. PACS Nos 28.41.Kw; 25.40.Fq; 24.60.Dr 1. Introduction It is recognized that for long-term energy security, nuclear energy is an inevitable option [1]. For a sustainable nuclear energy programme, the management of long-lived nuclear waste is very critical. Radioactive nuclei like Pu, minor actinides like Np, Am and Cm and long-lived fission products like 79Se, 93Zr, 99Tc, 107Pd, 126Sn, 129I and 135Cs constitute the main waste burden from a power reactor. In this paper, we shall discuss the management strategies for nuclear waste in general, and long-lived fission products, in particular. 2. Management of nuclear waste The radioactive nuclei which are produced in a power reactor and which remain in the spent fuel of the reactor form a major portion of nuclear waste.
  • Management of Spent Fuel from Nuclear Power Reactors 0ROCEEDINGS�OF�AN�INTERNATIONAL�CONFERENCE 6IENNA �N�*UNE� IAEA SAFETY RELATED PUBLICATIONS

    Management of Spent Fuel from Nuclear Power Reactors 0ROCEEDINGSOFANINTERNATIONALCONFERENCE 6IENNA N*UNE IAEA SAFETY RELATED PUBLICATIONS

    Management of Spent Fuel from Nuclear Power Reactors 0ROCEEDINGSOFANINTERNATIONALCONFERENCE 6IENNA n*UNE IAEA SAFETY RELATED PUBLICATIONS IAEA SAFETY STANDARDS Under the terms of Article III of its Statute, the IAEA is authorized to establish or adopt standards of safety for protection of health and minimization of danger to life and property, and to provide for the application of these standards. The publications by means of which the IAEA establishes standards are issued in the IAEA Safety Standards Series. This series covers nuclear safety, radiation safety, transport safety and waste safety, and also general safety (i.e. all these areas of safety). The publication categories in the series are Safety Fundamentals, Safety Requirements and Safety Guides. Safety standards are coded according to their coverage: nuclear safety (NS), radiation safety (RS), transport safety (TS), waste safety (WS) and general safety (GS). Information on the IAEA’s safety standards programme is available at the IAEA Internet site http://www-ns.iaea.org/standards/ The site provides the texts in English of published and draft safety standards. The texts of safety standards issued in Arabic, Chinese, French, Russian and Spanish, the IAEA Safety Glossary and a status report for safety standards under development are also available. For further information, please contact the IAEA at P.O. Box 100, 1400 Vienna, Austria. All users of IAEA safety standards are invited to inform the IAEA of experience in their use (e.g. as a basis for national regulations, for safety reviews and for training courses) for the purpose of ensuring that they continue to meet users’ needs.
  • Heu Repatriation Project

    Heu Repatriation Project

    HEU REPATRIATION PROJECT RATIONALE In April 2010, the governments of Canada and the United States (U.S.) committed to work cooperatively to repatriate spent highly- enriched uranium (HEU) fuel currently stored at the Chalk River Laboratories in Ontario to the U.S. as part of the Global Threat Reduction Initiative, a broad international effort to consolidate HEU inventories in fewer locations around the world. This initiative PROJECT BACKGROUND promotes non-proliferation This HEU is the result of two decades of nuclear fuel use at the by removing existing weapons Chalk River Laboratories for Canadian Nuclear Laboratories (CNL) grade material from Canada research reactors, the National Research Experimental (NRX) and and transferring it to the National Research Universal (NRU), and for the production of U.S., which has the capability medical isotopes in the NRU, which has benefitted generations of to reprocess it for peaceful Canadians. Returning this material to the U.S. in its existing solid purposes. In March 2012, and liquid forms ensures that this material is stored safely in a Prime Minister Harper secure highly guarded location, or is reprocessed into other forms announced that Canada and that can be used for peaceful purposes. the U.S. were expanding their efforts to return additional Alternative approaches have been carefully considered and inventories of HEU materials, repatriation provides the safest, most secure, and fastest solution including those in liquid form. for the permanent disposition of these materials, thereby eliminating a liability for future generations of Canadians. For more information on this project contact: Email: [email protected] Canadian Nuclear Laboratories 1-866-886-2325 or visit: www.cnl.ca persons who have a legitimate need to PROJECT GOAL know, such as police or emergency response To repatriate highly-enriched uranium forces.
  • Spent Fuel Reprocessing

    Spent Fuel Reprocessing

    Spent Fuel Reprocessing Robert Jubin Oak Ridge National Laboratory Reprocessing of used nuclear fuel is undertaken for several reasons. These include (1) recovery of the valuable fissile constituents (primarily 235U and plutonium) for subsequent reuse in recycle fuel; (2) reduction in the volume of high-level waste (HLW) that must be placed in a geologic repository; and (3) recovery of special isotopes. There are two broad approaches to reprocessing: aqueous and electrochemical. This portion of the course will only address the aqueous methods. Aqueous reprocessing involves the application of mechanical and chemical processing steps to separate, recover, purify, and convert the constituents in the used fuel for subsequent use or disposal. Other major support systems include chemical recycle and waste handling (solid, HLW, low-level liquid waste (LLLW), and gaseous waste). The primary steps are shown in Figure 1. Figure 1. Aqueous Reprocessing Block Diagram. Head-End Processes Mechanical Preparations The head end of a reprocessing plant is mechanically intensive. Fuel assemblies weighing ~0.5 MT must be moved from a storage facility, may undergo some degree of disassembly, and then be sheared or chopped and/or de-clad. The typical head-end process is shown in Figure 2. In the case of light water reactor (LWR) fuel assemblies, the end sections are removed and disposed of as waste. The fuel bundle containing the individual fuel pins can be further disassembled or sheared whole into segments that are suitable for subsequent processing. During shearing, some fraction of the radioactive gases and non- radioactive decay product gases will be released into the off-gas systems, which are designed to recover these and other emissions to meet regulatory release limits.
  • Reactor Physics Calculations in the Nordic Countries

    Reactor Physics Calculations in the Nordic Countries

    ESPOO 2003 VTT SYMPOSIUM 230 The eleventh biennial meeting on reactor physics calculations in the Nordic VTT SYMPOSIUM 230 countries was arranged by VTT Processes in Otaniemi, Espoo and on board Tallink´s m/s Romantika on April 9–10, 2003. General reactor physics, calculational methods, a code system adapted for RBMK reactor analyses, and transmutation of nuclear waste were presented by representatives of universities and programme developers. Computer programmes are the most important tools of reactor physics. At the meeting there were presentations of VTT Processes’ new deterministic 3- dimensional radiation transport code MultiTrans and BWR simulator ARES based upon the AFEN model, and also of new features in internationally wellknown codes like CASMO-4E and POLCA (POLCA-T) together with Reactor physics calculations in the Nordic countries results obtained by these programmes. A code for PWR loading pattern search, called LP-fun, is being developed by Westinghouse and others. On the subject of code validation, measurements on SVEA-96+ fuel bundles in the PROTEUS facility had been analyzed with the PHOENIX4 code, reactor scram experiments in the Loviisa and Mochovce VVER reactors using CASMO-4, MCNP4B and HEXTRAN, results of gamma scanning by the PHOENIX4/POLCA7 combination. Some difficulties in predicting the power distribution in the reactor core with sufficiently good accuracy using any of the available code systems were reported by OKG. Heating of non-fuel regions by gamma radiation and neutrons had been investigated using the HELIOS lattice code. Calculational results for heat deposition from gamma radiation in the moderator tank of the Forsmark-1 reactor were reported by Risø.
  • GAO-15-141, SPENT NUCLEAR FUEL MANAGEMENT: Outreach

    GAO-15-141, SPENT NUCLEAR FUEL MANAGEMENT: Outreach

    United States Government Accountability Office Report to Congressional Requesters October 2014 SPENT NUCLEAR FUEL MANAGEMENT Outreach Needed to Help Gain Public Acceptance for Federal Activities That Address Liability GAO-15-141 D October 2014 SPENT NUCLEAR FUEL MANAGEMENT Outreach Needed to Help Gain Public Acceptance for Federal Activities That Address Liability Highlights of GAO-15-141, a report to congressional requesters Why GAO Did This Study What GAO Found DOE is responsible for disposing of Spent nuclear fuel—the used fuel removed from nuclear power reactors—is commercial spent nuclear fuel. DOE expected to accumulate at an average rate of about 2,200 metric tons per year in entered into contracts with owners and the United States. This spent nuclear fuel is mostly stored wet, submerged in generators of spent nuclear fuel to pools of water. However, since pools have been reaching their capacities, begin disposing of it beginning in 1998, owners and generators of spent nuclear fuel (typically utilities and reactor with plans for disposal in a national operators) have been transferring it to canisters that are placed in casks on repository. DOE, however, was unable concrete pads for dry storage—which is an expensive and time-consuming to meet the 1998 date and, as a result process. When operating reactors’ licenses begin to expire in the 2030s, the rate of lawsuits, the federal government has of spent nuclear fuel accumulation is expected to decrease, but the amount in dry paid out about $3.7 billion for storage storage will increase as the pools are closed and all spent nuclear fuel is costs.
  • Vver and Rbmk Cross Section Libraries for Origen-Arp

    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].
  • Nuclear Fuel Optimization and Problem of Xa9953256 Increasing Burnup and Cost Efficiency of Light Water Reactors in Russia

    Nuclear Fuel Optimization and Problem of Xa9953256 Increasing Burnup and Cost Efficiency of Light Water Reactors in Russia

    NUCLEAR FUEL OPTIMIZATION AND PROBLEM OF XA9953256 INCREASING BURNUP AND COST EFFICIENCY OF LIGHT WATER REACTORS IN RUSSIA M.I. SOLONIN, Yu. K. BIBLASHCVLI, F.G. RESHETNIKOV, F.F. SOKOLOV, A.A. BOCHVAR All-Russia Research Institute of Inorganic Materials, Moscow, Russia Abstract Brief analysis is given of the results of WWER-1000 and WWER-440 fuel assembly (FA) and fuel rod operation to the average burn-up of 44 and 34 MW-day/kg U, respectively. The reliable operation of the fuel is noted which made it possible to outline the paths of further improvements in their fuel cycles. Based on a series of improvements reactors are being switched on a 4-year cycle; a 5-year cycle is contemplated for the WWER- 440. The fuel cycle of the boiling RBMK reactors is subjected to essential alterations. Specifically, the use of erbium oxide as an integral burnable absorber allows a higher enrichment of uranium and bum-up. To extend the burn-up a new zirconium alloy E-635 is assumed to be used as fuel rod claddings and other FA components; the alloy has higher corrosion and irradiation resistance. INTRODUCTION The current technical level of the PWR and BWR fuel production and the perfected system of the reactor management have allowed to provide the successful operation of PWR NPP to the average burn-up of ~45 MW* day/kg U in all countries of developed nuclear power. The differences in the reliability and FA serviceability as well as in the number of leaky fuel rods are rather small and the attention is usually not focused on this issue.