DOCSLIB.ORG

Modelling of Htrs with Monte Carlo:Sensitivity Due to Different

Total Page：16

File Type：pdf, Size：1020Kb

Download full-text PDF Read full-text
Modelling of Htrs with Monte Carlo:Sensitivity Due to Different 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) Modelling of HTRs with Monte Carlo: Sensitivity due to Different Isotopic Fuel Composition Rita Plukiene1, Danas Ridikas2*, Arturas Plukis1 and Vidmantas Remeikis1 1Physics Institute, Savanoriu pr. 231, LT-2053 Vilnius, Lithuania 2DSM/DAPNIA/SPhN, CEA Saclay, F-91191 Gif-sur-Yvette, France The gas turbine modular helium-cooled reactor (GT-MHR) is a potential candidate for the maximum plutonium destruction in once-through cycle. A particular feature of GT-MHR is that its refractory coated fuel (TRISO particles) is supposed to provide an impermeable barrier to the release of fission products and, at the same time, to resist very deep burn-up rates (more than 90% for 239Pu). In this work we performed detailed Monte Carlo simulations of the GT-MHR operation by loading with different plutonium fuel vectors: plutonium from military applications, plutonium from LWR and RBMK spent nuclear fuel. The comparison of the main GT-MHR performance parameters: keff eigenvalues, the length of the fuel cycle, neutron characteristics and the evolution of fuel composition in particular were obtained. We show that the performance of GT-MHR may be considerably influenced by the plutonium isotopic composition vector used as initial fuel material. This is the first time when incineration possibility of the RBMK-1500 based plutonium isotopic composition was tested using high temperature reactor. Finally, the propagation of statistical errors for the fuel burn-up was examined in detail showing no uncertainty accumulation. KEYWORDS: High Temperature Reactor, Monte Carlo, Pu Incineration 1. Introduction In principle, GT-MHR can be used to burn all types of fuel and offers significant advantages in accomplishing the transmutation of plutonium isotopes and high destruction of 239Pu in particular [1]. The GT-MHR features are: the helium coolant, which remains in a single phase under all conditions; the graphite core, which provides high heat capacity, slow thermal response, and structural stability at high temperatures; the refractory coated particle fuel, which allows high burn-up, retains fission products during operation in high temperatures and retains their integrity in a repository environment for hundreds of thousands years [2]. It utilizes natural erbium as a burn-able poison with the capture cross section having a resonance at a neutron energy such that ensures a strong negative temperature coefficient of reactivity. The lack of interaction of neutrons with coolant (helium gas) makes sure that temperature feedback of fuel and graphite is the only significant contributor to the power coefficient. These features are combined with a closed Brayton energy conversion cycle giving a net efficiency in the range of 45-47 % [2]. In this work we performed detailed simulations of GT-MHR operation by loading it with different Pu fuel vector: plutonium isotopes separated from LWR and RBMK-1500 spent nuclear fuel and plutonium from military applications. The main goal of our study is to compare the main GT-MHR reactor performance parameters as keff - values, the length of the fuel cycle, neutron characteristics, temperature coefficients, and the evolution of fuel composition as a function of different initial fuel isotopic composition. The modelling of GT-MHR and plutonium based fuel cycles within a Monte * Corresponding author, Tel. +33 1 69087847, FAX +33 1 69087584, E-mail: [email protected] Carlo approach was made using double-heterogeneous reactor core geometry including micro-particles [3]. The propagation of Monte Carlo errors was addressed in detail in the last section of this paper. This is the first time when transmutation possibility of the plutonium, originating from the RBMK-1500 spent nuclear fuel (Ignalina nuclear power plant in Lithuania) was tested using GT-MHR. 2. Modeling details The investigation of transmutation feasibility of different isotopic plutonium composition in GT-MHR reactor is based on 3D double-heterogeneous reactor core geometry with micro-particles as described in our recent studies [3, 4]. In brief, the main feature of a double-heterogeneous geometry is an exact active core structure, where fuel and erbium compacts include ceramic - coated fuel and erbium particles mixed uniformly in a graphite matrix. Each coated fuel particle consists of spherical kernel of PuO2-x (200 µm diameter) surrounded by triple protective coatings from pyrocarbon and SiC (so called TRISO coating). The same structure is valid for particles containing burn-able poison - natural erbium in the form of Er2O3. In this work we used the MCNP code and associated data libraries to obtain keff, neutron fluxes and temperature coefficients. As soon as keff is smaller than 1 (within statistical errors) the length of the fuel cycle is determined. In the more realistic description of GT-MHR, control rods would need to be modeled. Here the control rods were not designed since their effect would be very similar in the case of a comparative analysis (relative comparison) if modeled-omitted exactly in the same way. For burn-up calculations we employed the Monteburns code (MCNP+ORIGEN) [5]. ENDF/B-VI data library (as the most often employed with MCNP) was used for the fuel and structure materials, JENDL-3.2 data files (having the biggest number of fission products) were employed for fission products. Neutron spectrum in GT-MHR changes considerably during the fuel burn-up [3]. In this work the neutron fluxes and cross sections were recalculated in every fifty days for the follow-up ORIGEN burn-up calculations. As it was mentioned above, our major interest in this study was to look at the GT-MHR performance parameters by investigating different plutonium fuel isotopic configuration, i.e. plutonium originating from: a) the spent nuclear fuel of LWR (light water reactor), b) the spent nuclear fuel of RBMK - 1500 (graphite - moderated water - cooled reactor), and c) plutonium found in military applications (see Table 1). Table 1. Different GT-MHR fuel and erbium load at the beginning of the fuel cycle.*Note that in RBMK Pu1 case GT-MHR performance parameters where investigated two times – with burnable poison erbium (in Figure 1), and without erbium (in Figure 2). Fuel Military LWR RBMK RBMK RBMK RBMK Pu Pu Pu1* Pu2 Pu3 Pu4 Pu composition, kg 7.0 9.5 238Pu -- 32.4 2.5 6.0 685.5 652.2 239Pu 659.0 661.0 802.1 673.3 429.6 439.0 240Pu 7.8 277.0 361.8 443.1 13.0 14.1 241Pu 4.2 142.0 8.4 12.4 64.9 85.2 242Pu -- 87.6 25.2 65.2 Total plutonium, kg : 701.0 1200.0 1200.0 1200.0 1200.0 1200.0 Er composition, kg : 166Er 132.0 16.8 18.0 -- -- -- 167Er 94.2 11.4 12.2 -- -- -- 170Er 61.4 7.5 8.0 -- -- -- Total erbium, kg : 293.6 35.7 38.2 -- -- -- Since RBMK-1500 SNF (Spent Nuclear Fuel) isotopic composition strongly depends on the initial 235U enrichment and its final burn-up, for the RBMK plutonium several cases were tested: “RBMK Pu1” – 2.0 % 235U initial enrichment, 14 MWd/kg burn-up, “RBMK Pu2” – 2.0 % 235U initial enrichment, 20 MWd/kg burn-up, “RBMK Pu3” – 2.4 % 235U initial enrichment, 22 MWd/kg burn-up, “RBMK Pu4” – 2.6 % 235U initial enrichment, 26 MWd/kg burn-up. RBMK-1500 reactor spent fuel 1.20 composition was calculated using SCALE 4.4a code package (with 1.15 ENDF/B-V nuclear data library) [6]. The corresponding GT-MHR 1.10 plutonium fuel masses for each composition are presented in Table 1. 1.05 eff 1200 kg for civil plutonium cases and k RBMK Pu1 (k =1.156) 700 kg for military plutonium were the 1.00 eff RBMK Pu2 (k =1.087) optimal fuel loads for GT-MHR. The eff RBMK Pu3 (k =1.103) burnable poison (erbium) in natural 0.95 eff RBMK Pu4 (k =1.088) isotopic composition was added to eff 0.90 reactor core to achieve the same keff 0 200 400 600 800 1000 eigenvalue at the beginning of the fuel Time, d cycle for LWR Pu, RBMK Pu1 and Military Pu. A high load of 240Pu in Figure 1. A behaviour of keff for different RBMK other three cases Pu2, Pu3 and Pu4 plutonium cases without burnable poison in GT-MHR. limited the possibility to add burnable poison (see next section). 3. Burn-up results We perform our calculations by simulating the once-through fuel cycle at the constant 600 MWth power. Our primary goal was to check the RBMK plutonium as an initial GT-MHR fuel load. For this purpose GT-MHR performance parameters were analyzed for four RBMK fuel cases without burnable poison. The keff ± 1σ time dependence 1.20 (consequently, its dependence on Military Pu (k =1.132) eff burn-up) for RBMK Pu1, Pu2, Pu3 and LWR Pu (k =1.138) 1.15 eff Pu4 compositions are presented in RBMK Pu1 (k =1.132) eff Figure 1. The most attractive RBMK 1.10 plutonium composition is that of 14 MWd/kg burn-up (up to now there is 1.05 eff about 15% of all SNF), but the use of k other burn-up RBMK SNF plutonium 1.00 in GT-MHR is also possible. In RBMK Pu3 case more than 90% of 239Pu 0.95 transmutation was obtained (see the 0.90 discussion below). 0 200 400 600 800 1000 1200 The GT-MHR with military Time, d plutonium composition was tested in our previous studies [3, 4]. In order to obtain Figure 2. A behaviour of keff for Military, LWR and relative fuel cycle length for different 14 MWd/kg burn-up RBMK plutonium with burnable GT-MHR fuel cases the appropriate poison in GT-MHR.
Load more

Recommended publications
  • 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.
    [Show full text]
  • 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]
  • 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.
    [Show full text]
  • Magnox Corrosion
    Department Of Materials Science & Engineering From Magnox to Chernobyl: A report on clearing-up problematic nuclear wastes Sean T. Barlow BSc (Hons) AMInstP Department of Materials Science & Engineering, The University of Sheffield, Sheffield S1 3JD, UK IOP Nuclear Industry Group, September 28 | Warrington, Cheshire, UK Department Of Materials Science & Engineering About me… 2010-2013: Graduated from University of Salford - BSc Physics (Hons) Department Of Materials Science & Engineering About me… 2013-2018: Started on Nuclear Fission Research Science and Technology (FiRST) at the University of Sheffield Member of the Immobilisation Science Laboratory (ISL) group • Wasteforms cement, glass & ceramic • Characterisation of materials (Trinitite/Chernobylite) • Corrosion science (steels) Department Of Materials Science & Engineering Department Of Materials Science & Engineering Fully funded project based PhD with possibility to go on secondment • 3-4 month taught course at Manchester • 2 mini-projects in Sheffield • 3 years for PhD + 1 year write up • Funding available for conferences and training • Lots of outreach work • Site visits to Sellafield reprocessing facility, Heysham nuclear power station & Atomic Weapons Authority Department Of Materials Science & Engineering Fully funded project based PhD with possibility to go on secondment • 3-4 month taught course at Manchester • 2 mini-projects in Sheffield • 3 years for PhD + 1 year write up • Funding available for conferences and training • Lots of outreach work • Site visits to Sellafield reprocessing facility, Heysham nuclear power station & Atomic Weapons Authority Department Of Materials Science & Engineering Department Of Materials Science & Engineering 2017: Junior Project Manager at DavyMarkham Department Of Materials Science & Engineering PhD projects… 4 main projects 1. Magnox waste immobilisation in glass 2.
    [Show full text]
  • Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States Final Report
    NUREG-1251 Vol. I Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States Final Report Main Report U.S. Nuclear Regulatory Commission p. o AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555 2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica- tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investi- gation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceed- ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula- tions in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
    [Show full text]
  • Fundamentals of Nuclear Reactor Physics
    Fundamentals of Nuclear Reactor Physics E. E. Lewis Professor of Mechanical Engineering McCormick School of Engineering and Applied Science Northwestern University AMSTERDAM • BOSTON • HEIDELBERG • LONDON NEW YORK • OXFORD • PARIS • SAN DIEGO SAN FRANCISCO • SINGAPORE • SYDNEY • TOKYO ELSEVIER Academie Press is an imprint of Elsevier Contents Preface xiii 1 Nuclear Reactions 1 1.1 Introduction 1 1.2 Nuclear Reaction Fundamentals 2 Reaction Equations 3 Notation 5 Energetics 5 1.3 The Curve of Binding Energy 7 1.4 Fusion Reactions 8 1.5 Fission Reactions 9 Energy Release and Dissipation 10 Neutron Multiplication 12 Fission Products 13 1.6 Fissile and Fertile Materials 16 1.7 Radioactive Decay 18 Saturation Activity 20 Decay Chains 21 2 Neutron Interactions 29 2.1 Introduction 29 2.2 Neutron Cross Sections 29 Microscopic and Macroscopic Cross Sections 30 Uncollided Flux 32 Nuclide Densities 33 Enriched Uranium 35 Cross Section Calculation Example 36 Reaction Types 36 2.3 Neutron Energy Range 38 2.4 Cross Section Energy Dependence 40 Compound Nucleus Formation 41 Resonance Cross Sections 42 Threshold Cross Sections 46 Fissionable Materials 47 vii viii Contents 2.5 Neutron Scattering 48 Elastic Scattering 49 Slowing Down Decrement 50 Inelastic Scattering 52 3 Neutron Distributions in Energy 57 3.1 Introduction 57 3.2 Nuclear Fuel Properties 58 3.3 Neutron Moderators 61 3.4 Neutron Energy Spectra 63 Fast Neutrons 65 Neutron Slowing Down 66 Thermal Neutrons 70 Fast and Thermal Reactor Spectra 72 3.5
    [Show full text]
  • System Studies of Fission-Fusion Hybrid Molten Salt Reactors
    University of Tennessee, Knoxville TRACE: Tennessee Research and Creative Exchange Doctoral Dissertations Graduate School 12-2013 SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS Robert D. Woolley University of Tennessee - Knoxville, [email protected] Follow this and additional works at: https://trace.tennessee.edu/utk_graddiss Part of the Nuclear Engineering Commons Recommended Citation Woolley, Robert D., "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS. " PhD diss., University of Tennessee, 2013. https://trace.tennessee.edu/utk_graddiss/2628 This Dissertation is brought to you for free and open access by the Graduate School at TRACE: Tennessee Research and Creative Exchange. It has been accepted for inclusion in Doctoral Dissertations by an authorized administrator of TRACE: Tennessee Research and Creative Exchange. For more information, please contact [email protected]. To the Graduate Council: I am submitting herewith a dissertation written by Robert D. Woolley entitled "SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS." I have examined the final electronic copy of this dissertation for form and content and recommend that it be accepted in partial fulfillment of the equirr ements for the degree of Doctor of Philosophy, with a major in Nuclear Engineering. Laurence F. Miller, Major Professor We have read this dissertation and recommend its acceptance: Ronald E. Pevey, Arthur E. Ruggles, Robert M. Counce Accepted for the Council: Carolyn R. Hodges Vice Provost and Dean of the Graduate School (Original signatures are on file with official studentecor r ds.) SYSTEM STUDIES OF FISSION-FUSION HYBRID MOLTEN SALT REACTORS A Dissertation Presented for the Doctor of Philosophy Degree The University of Tennessee, Knoxville Robert D.
    [Show full text]
  • Risks of Operating Candu 6 Nuclear Power Plants: Gentilly Unit 2 Refurbishment and Its Global Implications
    Risks of Operating Candu 6 Nuclear Power Plants: Gentilly Unit 2 Refurbishment and its Global Implications November 2008 BY GORDON R. THOMPSON Institute for Resource and Security Studies www.greenpeace.ca RISKS OF OPERATING ABSTRACT Operation of any nuclear power plant creates risks. CANDU 6 plants CANDU 6 NUCLEAR POWER PLANTS: pose additional risks arising from their use of natural uranium as fuel and Gentilly Unit 2 Refurbishment and its heavy water as moderator. A CANDU 6 reactor could experience a violent Global Implications power excursion, potentially leading to containment failure and a release of radioactive material to the environment. Spent fuel discharged from a CANDU 6 could be diverted and used to produce plutonium for nuclear BY GORDON R. THOMPSON weapons. Those risks are examined here with special attention to Hydro- Institute for Resource and Security Studies Quebec’s plan for refurbishment and continued operation of the Gentilly 2 plant. That action would lead to continued radiological risk in Quebec Prepared under the sponsorship of and could promote sales of CANDU 6 plants in other countries, thereby Greenpeace Canada contributing to an enhanced risk of nuclear-weapon proliferation. Hydro- Quebec’s plan also faces regulatory risks. Safety issues could increase the © Copyright November 2008 cost of refurbishing Gentilly 2, weakening an already marginal economic case for refurbishment. This report proposes an approach for systematic, GREENPEACE CANADA public assessment of the risks associated with Gentilly 2. 33 Cecil St. Toronto, Ontario ABOUT THE INSTITUTE FOR RESOURCE AND SECURITY STUDIES M5T 1N1 The Institute for Resource and Security Studies (IRSS) is an independent, www.greenpeace.ca nonprofit, Massachusetts corporation, founded in 1984.
    [Show full text]
  • Converting Russian Plutonium- Production Reactors to Civilian Use
    - Science & Global Security, 1994, Volume 5, pp.37-46 @ 1994 OPA (Overseas Publishers Association) Reprints available directly from the publisher Amsterdam B. V. PhorocoPying permitted by license only Published under license by Gordon and Breach Science Publishers SA Printed in the United States of America Converting Russian Plutonium- Production Reactors to Civilian Use Alexander M. Dmitrievo PLUTONIUM-PRODUCTION REACTORS IN RUSSIA The first Soviet reactor designedto produceweapon-grade began operation in Chelyabinsk in June of 1948.This single-purpose,graphite-moderated, single- pass reactor was a water-cooledreactor. Its coolingwater was dumped directly into a lake at roughly 100°C. Several more reactors of this type were subse- quently constructed. In 1958, a reactor of a new type began operation in Tomsk-7. A graphite- moderated production reactor was built to operate in dual-purposemode, pro- ducing both weapons-gradeplutonium and heat and electricity for local resi- dents. The primary reactor circuit was closed, and the reactor plant was equipped with heat exchangers, steam generators and turbines to produce electric power. In 1961, 1964and 1965, four additional dual-purposereactors beganoperation: three in Tomsk-7, and one in Krasnoyarsk-26. Shortly after being commissioned,the Krasnoyarsk-26 reactor was fitted with a system to transmit its low-grade heat (i.e., heated water that has passedthrough steam generators)to meet residential heat demands.(Low-grade heat had already been used for some time to industrial plant buildings at the Tomsk-7 and Krasnoyarsk-26 sites.) Initially, the supply of heat was rather limited, but by 1968a decisionwas made to heat the district of the main city of Tomsk with waste heat from the reactor.
    [Show full text]
  • Rbmk Nuclear Power Plants: Generic Safety Issues
    IAEA-EBP-RBMK-04 XA9744137 RBMK NUCLEAR POWER PLANTS: GENERIC SAFETY ISSUES A PUBLICATION OF THE EXTRABUDGETARY PROGRAMME ON THE SAFETY OF WWER AND RBMK NUCLEAR POWER PLANTS May 1996 INTERNATIONAL ATOMIC ENERGY AGENCY The originating Section of this publication in the IAEA was: Safety Assessment Section International Atomic Energy Agency Wagramerstrasse 5 P.O. Box 100 A-1400 Vienna, Austria RBMK NUCLEAR POWER PLANTS: GENERIC SAFETY ISSUES IAEA, VIENNA, 1996 IAEA-EBP-RBMK-04 ISSN 1025-2754 ©IAEA, 1996 Printed by the IAEA in Austria May 1996 FOREWORD The IAEA initiated in 1990 a programme to assist the countries of central and eastern Europe and the former Soviet Union in evaluating the safety of their first generation WWER-440/230 nuclear power plants. The main objectives of the Programme were: to identify major design and operational safety issues; to establish international consensus on priorities for safety improvements; and to provide assistance in the review of the completeness and adequacy of safety improvement programmes. The scope of the Programme was extended in 1992 to include RBMK, WWER-440/213 and WWER- 1000 plants in operation and under construction. The Programme is complemented by national and regional technical cooperation projects. The Programme is pursued by means of plant specific safety review missions to assess the adequacy of design and operational practices; Assessment of Safety Significant Events Team (ASSET) reviews of operational performance; reviews of plant design, including seismic safety studies; and topical meetings on generic safety issues. Other components are: follow-up safety missions to nuclear plants to check the status of implementation of IAEA recommendations; assessments of safety improvements implemented or proposed; peer reviews of safety studies, and training workshops.
    [Show full text]
  • Co-Operation Among CMEA Member Countries in the Development of Nuclear Energy Its Role in the Implementation of the NPT by A.F
    Co-operation among CMEA Member Countries in the Development of Nuclear Energy Its Role in the Implementation of the NPT by A.F. Panasenkov An important aspect of the multilateral co-operation among the member countries of the Council for Mutual Economic Assistance (CMEA) is the development of nuclear power, a technology which can practically eliminate the threat of fuel and energy shortages and significantly reduce environmental pollution. Nuclear power is gradually becoming an independent specialized branch of power generation. However, the presence of transmission lines and the interconnected fuel cycle, both of which cross national boundaries, make nuclear power an international activity within the framework of CMEA. The nuclear power plants being installed in CMEA member countries use thermal-neutron pressurized-water-cooled water-moderated tank reactors (designated WWER) having power outputs of 440 and 1000 MWe and in the USSR plants with uranium-graphite boiling-water channel-type reactors (designated RBMK) having power outputs or 1000 and 1500 MWe are being built as well. All these reactors use uranium fuel slightly enriched in the isotope uranium-235. Currently, the total power output of operating nuclear power plants in Bulgaria, Czechoslovakia, the German Democratic Republic and the USSR is over 17 000 MWe; the construction of a first nuclear power plant is nearing completion in Hungary, and preparatory design work for the construction of nuclear power plants in Cuba, Poland and Romania is being carried out. However, in thermal-neutron reactors the nuclear fuel is not utilized effectively enough. For this reason, the future development of nuclear power plants in the CMEA member countries is to involve the use of fast-neutron reactors.
    [Show full text]
  • Actinides Input to the Dose in the Irradiated Graphite of RBMK-1500 Reactor
    Nuclear Engineering and Design 300 (2016) 530–535 Contents lists available at ScienceDirect Nuclear Engineering and Design jou rnal homepage: www.elsevier.com/locate/nucengdes Actinides input to the dose in the irradiated graphite of RBMK-1500 reactor a,∗ a a a a a R. Plukiene˙ , A. Plukis , A. Puzas , R. Gvozdaite˙ , V. Barkauskas , G. Duskesasˇ , b b a J.V. Cizdziel , D. Bussan , V. Remeikis a Institute of Physics, Center for Physical Sciences and Technology, Savanoriu˛ pr. 231, LT-02300 Vilnius, Lithuania b University of Mississippi, Department of Chemistry and Biochemistry, 305 Coulter Hall, University, Oxford, MS 38677, USA h i g h l i g h t s • Actinides input to the dose in RBMK-1500 reactor graphite was estimated. • SCALE 6.1 and MCNPX models were used to calculate actinides specific activities. • ORIGEN-ARP was used for gamma power, neutron source and effective dose calculation. • Concentrations of Pu, Am and Cm isotopes in the RBMK graphite sample were measured. • 244 Cm was found to be a critical contributor to effective dose of the personnel. a r t i c l e i n f o a b s t r a c t Article history: The purpose of this work is to indicate the actinides input to the total radiation dose caused by the Received 25 May 2015 irradiated graphite of the RBMK-1500 reactor in comparison to the dose delivered by other nuclides. Received in revised form 27 January 2016 We used computer codes (SCALE 6.1 and MCNPX 2.7) to estimate the dose rate delivered by actinides Accepted 4 February 2016 244 giving special attention to the Cm isotope as a critical contributor to the total activity of actinides in the spent graphite for approximately up to 200 years.
    [Show full text]