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Sample records for vver-440 fuel cycles

  1. VVER fuel cycle development at Slovakia

    International Nuclear Information System (INIS)

    Darilek, P.; Chrapiak, V.; Majerik, J.

    1995-01-01

    Four VVER-440 units are now under exploitation at Bohunice-site in Slovakia. Fuel cycle development of Unit No.3 and No.4 (type 213) is discussed and compared with equilibrium cycles in this paper. (author)

  2. Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440. Preliminary assessment of operating efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinskiy, Alexey [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2017-09-15

    Since the introduction of VVERs-440, their fuel assemblies are subject to ongoing improvements. Until now, the basic structural parameters of fuel, such as rod diameter of 9.1 mm, have never changed. This paper focuses on computational estimates of basic neutronic parameters of the fuel cycle that involves assemblies consisting of fuel rods with diameter reduced to 8.9 mm.

  3. The FARC fuel archive of VVER

    International Nuclear Information System (INIS)

    Zizin, M.N.; Parfenova, N.A.; Proselkov, V.N.; Shishkov, L.K.

    1998-01-01

    The principles of organisation are explained and the structure of the FARC fuel archive for VVER reactors is described. The objective of the archive is accumulation of fuel data, data storage and obtaining the fuel using characteristics. The working version of fuel archive on 01.07.98 is realised, in which the data tables for fuel assemblies for 169 VVER-440 cycles and 35 VVER-1000 cycles are stored. There are two different versions of fuel archive - for VVER-440 (FARC) and for VVER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. The algorithms and codes for checking integrity, reasonableness and reliability of fuel archive data are developed. (author)

  4. Improving the VVER-440 fuel design and technology

    International Nuclear Information System (INIS)

    Aksenov, P.; Bondar, Y.; Kolosovsky, Y.; Kochergin, V.; Luzan, Y.; Malakhov, A.; Krapivtsev, V.; Bauman, N.; Shumeev, A.; Filippov, V.

    2009-01-01

    Operational performance of VVER-440 fuel has long been demonstrating good reliability of the fuel. However, assembly failures occur, and fuel suppliers should always take measures to maintain its reliability. For several years, OAO MSZ has been fabricating working assemblies with detachable shrouds and removable fuel rods. The next step is the supply of demountable assemblies to allow inspection or repair of fuel rods after removal of the shroud. With the help of corresponding program the Russian organizations have carried out research and development work to advance and study operational features of demountable VVER-440 CFAs. The main engineering solutions are consistent with the working assemblies. The pilot demountable CFAs are running in the Kola-4 core. The obtained results can be used when deciding on the demountable CFAs delivery issues. The experiment-calculated research results of coolant mixing in the present design VVER-440 have been analysed. It has been found out that coolant mixing in the WA head is incomplete and that is why leading to conservatism when determining the reactor operational limits. The proposed WA head design includes an upgraded bumper grid with additional planes intensifying coolant mixing in the head. The bumper grid drawing and a pilot model is available. The thermohydraulics and rigidity features of the proposed design have been studied by experiment-calculated methods

  5. Possibility of implementation of 6-year fuel cycle at NPP with VVER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Heraltova, L., E-mail: lenka.heraltova@fjfi.cvut.cz [UJV Rez a.s., Hlavni 130, 250 68 Husinec-Rez (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, Brehova 7, 115 19 Praha 1 (Czech Republic)

    2015-12-15

    Highlights: • Possibility of extension of fuel cycle. • Increase of enrichment above 5% {sup 235}U. • Core properties calculated by diffusion code ANDREA. • Back end fuel cycle characteristic. - Abstract: This paper discusses possibility of an extension of a fuel cycle at a VVER-440 reactor for up to 6 years. The prolongation of a fuel cycle was realized by optimization of a fuel design and increasing of a fuel enrichment. The modified design of the fuel assembly covers change of pellet geometry, decreasing of parasitic absorption in construction materials, improved moderation of fuel pins and also increase of enrichment. Fuel assemblies with enrichment up to 7% {sup 235}U are considered for prolonged fuel batches. Three different batch lengths were considered for evaluation of core properties – 12, 18 and 24 months, and two types of burnable absorbers were included – Gd{sub 2}O{sub 3} and Er{sub 2}O{sub 3}. Comparison of proposed fuel assemblies was realized by length of a batch, average burnup, maximal power of fuel assembly or fuel pin, control fuel assembly worth, reactivity coefficients, and effective delayed neutrons fraction. Comparison of characteristics of a burned fuel discharged from a reactor core is discussed in the last part of the paper.

  6. Nuclear fuel for VVER reactors. Actual state and trends

    International Nuclear Information System (INIS)

    Molchanov, V.

    2011-01-01

    The main tasks concerning development of FA design, development and modernization of structural materials, improvement of technology of structural materials manufacturing and FA fabrication and development of methods and codes are discussed in this paper. The main features and expected benefit of implementation of second generation and third generation fuel assembly for VVER-440 Nuclear Fuel are given. A brief review of VVER-440 and VVER-1000 Nuclear Fuel development before 1997 since 2010 is shown. A summary of VVER-440 and VVER-1000 Nuclear Fuel Today, including details about TVSA-PLUS, TVSA-ALFA, TVSA-12 and NPP-2006 Phase 2 tasks (2010-2012) is presented. In conclusion, as a result of large scope of R and D performed by leading enterprises of nuclear industry modern nuclear fuel for VVER reactors is developed, implemented and successfully operated. Fuel performance (burnup, lifetime, fuel cycles, operating reliability, etc.) meets the level of world's producers of nuclear fuel for commercial reactors

  7. Feasibility of VVER-440 type SFAT

    International Nuclear Information System (INIS)

    Kaartinen, J.; Tarvainen, M.

    1995-05-01

    Spent fuel attribute tester, SFAT, has been constructed and tested for gross defect verification of VVER-440 type spent fuel assemblies. Based on earlier optimisation studies, the VVER-440 SFAT is kept hanging from the mast of the fuel handling machine moved by the operator. The device tested includes a standard 2' x 2' NaI(T1) detector connected to a commercial MCA. The results achieved with normal VVER-440 spent fuel assemblies at the Loviisa npp in Finland in November 1994 show that the method is feasible. The design of the so-called fuel follower assemblies, however, prevents SFAT verification, at least with moderate measurement times. Verification of the presence of the assemblies based on the detection of the fission product 137 Cs (662 keV) is possible even in 10-30 seconds. Measurement times of the order of 1-2 minutes make it possible to draw also semi-quantitative conclusions of the burnup and cooling time of the operator declared data (consistency check). (orig.) (7 refs., 11 figs., 3 tabs.)

  8. Impact of burnable absorber Gd on nuclide composition for VVER-440 fuel (Gd-2)

    International Nuclear Information System (INIS)

    Zajac, R.; Chrapciak, V.

    2010-01-01

    The latest version of Russian fuel VVER-440 includes burnable absorber in 6 pins. In this article is impact of burnable absorber on nuclide composition and criticality analyzed. In part 1 was analyzed whole burnup interval 0-50 MWd/kgU. In present part 2 are detailed analysis only for first cycle (burnup 0-10 MWd/kgU). (Authors)

  9. Results of VVER fuel rods tests in the MIR.M1 reactor under power cycling conditions

    International Nuclear Information System (INIS)

    Burukin, A.; Izhutov, A.; Ovchinnikov, V.; Kalygin, V.; Markov, D.; Pimenov, Y.; Novikov, V.; Medvedev, A.; Nesterov, B.

    2011-01-01

    The paper presents the main results of the 50 ... 60 MWd/kgU burnup VVER fuel rods tests performed in the MIR.M1 reactor loop facilities under power cycling. The non-destructive PIE results are presented as well. A series of experiments was performed, including overall measurement of fuel rod parameters test, in one of which 300 cycles were done. Irradiation under power cycling conditions and PIE of high-burnup VVER fuel rods showed the following: 1) all fuel rods claddings preserved their integrity under irradiation at linear heat rate (LHR) higher than the NPP operating one; 2) experimental data were obtained on the axial and radial cladding strain and fission gas release (FGR) from 50 ... 60 MWd/kgU burnup VVER-440 and VVER-1000 fuel rods as well as on the kinetics of the change in these parameters and fuel temperature under the power cycling; 3) non-destructive PIE results are in a satisfactory correlation with the data obtained by means of in-pile measurement gages during irradiation. (authors)

  10. The further development of WWER-440 fuel design performance

    International Nuclear Information System (INIS)

    Lushin, V.; Vasilchenko, I.; Ananjev, J.; Abashina, G.

    2011-01-01

    The most distinguished stages in VVER-440 fuel development of the latest ten years are: designing of second generation FA complex; and designing of sheathless working fuel assembly of the third generation (RK-3) which are presented in this report. Designing of fuel assemblies of the second generation and RK-3 is characterized by the tendency to power increase of VVER-440 operating units with V-213-type reactor, that, in turn, has given a stimulus to further design enhancement of fuel assemblies specified. The further development of the second generation fuel assembly design and the change-over to the third generation working assemblies will allow for fuel utilization to be considerably increased under the conditions of application the more long-term fuel cycles for VVER-440 reactors and operation of the Units at the increased power

  11. Safety of VVER-440 reactors

    CERN Document Server

    Slugen, Vladimir

    2011-01-01

    Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

  12. Experimental study of hydrodynamically induced vibrational processes in VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Solonin, V.I.; Perevezentsev, V.V.; Rekshnya, N.F.; Krapivtsev, V.G.

    2000-01-01

    Investigations are described of hydrodynamically induced vibrations in a single fuel assembly of a VVER-440 reactor, performed on a full-scale model installed in a closed loop filled with distilled water; the model fuel elements contained simulators of fuel pellets. Data on hydrodynamic loads were obtained by measuring pressure oscillations along the height of the fuel assembly case. Results of the measurements are presented in graphs and are discussed in some detail. (A.K.)

  13. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

  14. Influence of Bypass on Thermo-Hydraulics of VVER 440 Fuel Assembly

    Directory of Open Access Journals (Sweden)

    Jakubec Jakub

    2017-04-01

    Full Text Available The paper deals with CFD modelling and simulation of coolant flow within the nuclear reactor VVER 440 fuel assembly. The influence of coolant flow in bypass on the temperature distribution at the outlet of the fuel assembly and pressure drop was investigated. Only steady-state analyses were performed. Boundary conditions are based on operating conditions. ANSYS CFX is chosen as the main CFD software tool, where all analyses are performed.

  15. Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

    In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

  16. Hydrodynamics around a spacer of a VVER-440 fuel rod bundle

    International Nuclear Information System (INIS)

    Mayer, G.; Hazi, G.; Kavran, P.

    2004-01-01

    Recently, an intensive research has been started in our institute, focusing on the hydrodynamics of fuel rod bundles. Numerical computations have been planed to be carried out in a three level bottom-up hierarchy, using direct numerical simulation, large eddy simulation and Reynolds averaged Navier-Stokes approach. Here, we give a description of the numerical method applied for direct numerical and large eddy simulation. We present some preliminary results obtained by the simulation of the flow around a spacer of a VVER-440 fuel rod bundle. (author)

  17. VVER fuel. Results of post irradiation examination

    International Nuclear Information System (INIS)

    Smirnov, V.P.; Markov, D.V.; Smirnov, A.V.; Polenok, V.S.; Perepelkin, S.O.; Ivashchenko, A.A.

    2005-01-01

    The present paper presents the main results of post-irradiation examination of more than 40 different fuel assemblies (FA) operated in the cores of VVER-1000 and VVER-440-type power reactors in a wide range of fuel burnup. The condition of fuel assembly components from the viewpoint of deformation, corrosion resistance and mechanical properties is described here. A serviceability of the FA design as a whole and interaction between individual FA components under vibration condition and mechanical load received primary emphasis. The reasons of FA damage fuel element failure in a wide range of fuel burnup are also analyzed. A possibility and ways of fuel burnup increase have been proved experimentally for the case of high-level serviceability maintenance of fuel elements to provide for advanced fuel cycles. (author)

  18. Diversification of the VVER fuel market in Eastern Europe and Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Kirst, Michael [Westinghouse EMEA, Brussels (Belgium); Benjaminsson, Ulf; Oenneby, Carina [Westinghouse Electric Sweden AB, Vaesteraes (Sweden)

    2015-03-15

    There are a total of 33 VVER active reactors in the EU and Ukraine, accounting for the largest percentage of the total electricity supply in the countries operating these. The responsible governments and utilities operating these units want too see an increased diversification of the nuclear fuel supply. Westinghouse is the only nuclear fuel producer outside Russia, which has taken the major steps to develop, qualify and manufacture VVER fuel designs - both for VVER-440 and VVER-1000 reactors. The company has delivered reloads of VVER-440 fuel to Loviisa 2 in Finland, VVER-1000 fuel for both the initial core and follow-on regions to Temelin 1-2 in the Czech Republic and more recently reloads of VVER-1000 fuel to South Ukraine 2-3. Technical challenges in form of mechanical interference with the resident fuel have been encountered in Ukraine, but innovative solutions have been developed and successfully implemented and today Ukraine has, for the first time in its history, a viable VVER-1000 fuel design alternative, representing a tremendous lever in energy security for the country.

  19. Multiple recycle of REMIX fuel at VVER-1000 operation in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P. N.; Bobrov, E. A., E-mail: evgeniybobrov89@rambler.ru; Chibinyaev, A. V.; Teplov, P. S.; Dudnikov, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The basic features of loading the VVER-1000 core with a new variant of REMIX fuel (REgenerated MIXture of U–Pu oxides) are considered during its multiple recycle in a closed nuclear fuel cycle. The fuel composition is produced on the basis of the uranium–plutonium regenerate extracted at processing the spent nuclear fuel (SNF) from a VVER-1000, depleted uranium, and the fissionable material: {sup 235}U as a part of highly enriched uranium (HEU) from warheads superfluous for defense purposes or {sup 233}U accumulated in thorium blankets of fusion (electronuclear) neutron sources or fast reactors. Production of such a fuel assumes no use of natural uranium in addition. When converting a part of the VVER-1000 reactors to the closed fuel cycle based on the REMIX technology, the consumption of natural uranium decreases considerably, and there is no substantial degradation of the isotopic composition of plutonium or change in the reactor-safety characteristics at the passage from recycle to recycle.

  20. Sensitivity study applied to the CB4 VVER-440 benchmark on burnup credit

    International Nuclear Information System (INIS)

    Markova, Ludmila

    2003-01-01

    A brief overview of four completed portions (CB1, CB2, CB3, CB3+, CB4) of the international VVER-440 benchmark focused on burnup credit and a sensitivity study as one of the final views of the benchmark results are presented in the paper. Finally, the influence of real and conservative VVER-440 fuel assembly models taken for the isotopics calculation by SCALE sas2 on the system k eff is shown in the paper. (author)

  1. Verification of the enrichment of fresh VVER-440 fuel assemblies at NPP Paks

    Energy Technology Data Exchange (ETDEWEB)

    Almasia, I.; Hlavathya, Z.; Nguyena, C. T. [Institute of Isotopes, Hungarian Academy of Sciences, Budapest, (Hungary); others, and

    2012-06-15

    A Non Destructive Analysis (NDA) method was developed for the verification of {sup 235}U enrichment of both homogeneous and profiled VVER-440 reactor fresh fuel assemblies by means of gamma spectrometry. A total of ca. 30 assemblies were tested, five of which were homogeneous, with {sup 235}U enrichment in the range 1,6% to 3,6%, while the others were profiled with pins of 3,3% to 4,4% enrichment. Two types of gamma detectors were used for the test measurements: 2 coaxial HPGe detectors and a miniature CdZnTe (CZT) detector fitting into the central tube of the assemblies. It was therefore possible to obtain information from both the inside and the outside of the assemblies. It was shown that it is possible to distinguish between different types of assemblies within a reasonable measurement time (about 1000 sec). For the HPGe measurements the assemblies had to be lifted out from their storage rack, while for the CZT detector measurements the assemblies could be left at their storage position, as it was shown that the neighbouring assemblies do not affect measurement inside the assemblies' central tube. The measured values were compared to Monte Carlo simulations carried out using the MCNP code, and a recommendation for the optimal approach to verify the {sup 235}U enrichment of fresh VVER-440 reactor fuel assemblies is suggested.

  2. Response of Soviet VVER-440 accident localization systems to overpressurization

    International Nuclear Information System (INIS)

    Kulak, R.F.; Fiala, C.; Sienicki, J.J.

    1989-01-01

    The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. Descriptions of the VVER accident localization structures may be found in the report DOE NE-0084. The objective of this paper is to evaluate the structural integrity of the VVER-440 ALS at the Soviet design pressure, and to determine their response to pressure loadings beyond the design value. Complex, three-dimensional, nonlinear, finite element models were developed to represent the major structural components of the localization systems of the VVER-440 models V230 and V213. The interior boundary of the localization system was incrementally pressurized in the calculations until the prediction of gross failure. 6 refs., 9 figs

  3. Containment leak-tightness enhancement at VVER 440 NPPs

    International Nuclear Information System (INIS)

    Prandorfy, M.

    2001-01-01

    The hermetic compartments of VVER 440 NPPs fulfil the function of the containment used at NPPs all over the word. The purpose of the containment is to protect the NPP personal against radioactive impact as well as to prevent radioactive leakage to the environment during a lost of coolant accident. Leak-tightness enhancement in NPPs with VVER 440/213 and VVER 440/230 reactors is an important safety issue. New procedures, measures and methods were adopted at NPPs in Mochovce, J. Bohunice, Dukovany and Paks for leak identification and sealing works performed by VUEZ Levice. (authors)

  4. Interactive nuclear plant analyzer for the VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer (NPA) has been developed for a VVER-440 model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator. Results of the interactive calculation can be through the user-defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperatures of other metal structures. In addition, changes in the status of various components and system can be initiated and/or displayed both numerically and graphically on the mask

  5. APROS multifunctional simulator applications for VVER-440

    International Nuclear Information System (INIS)

    Porkholm, K.; Kantee, H.; Tiihonen, O.

    2000-01-01

    Fortum Engineering Ltd and the Technical Research Centre of Finland have developed APROS simulation software since 1986. APROS is a multifunctional simulator, which is used for process and automation design, safety analysis and training simulator applications. APROS has unique features and models developed especially for VVER-440 reactors. At first the paper gives a short overview of APROS multifunctional simulator. The rest of the paper deals with different kind of applications of APROS in VVER-440 reactors' improvement and operation development. (author)

  6. The choice of the fuel assembly for VVER-1000 in a closed fuel cycle based on REMIX-technology

    International Nuclear Information System (INIS)

    Bobrov, E.; Alekseev, P.; Chibinyaev, A.; Teplov, P.; Dudnikov, A.

    2016-01-01

    REMIX (Regenerated Mixture) fuel is produced directly from a non-separated mix of recycled uranium and plutonium from reprocessed used fuel and the fabrication technology of such fuel is called REMIX-technology. This paper shows basic features of different fuel assembly (FA) application for VVER-1000 in a closed fuel cycle based on REMIX-technology. This investigation shows how the change in the water-fuel ratio in the VVER FA affects the fuel characteristics produced by REMIX technology during multiple recycling. It is shown that for for the traditional REMIX-fuel it does not make sense to change anything in the design of VVER FA, because there are no advantages in the fuel feed consumption. The natural uranium economy by the fifth cycle reached about 29%. In the case of the REMIX fuel based on uranium-plutonium from SNF MOX fuel, it would be appropriate to use fuel assemblies with a water-fuel ratio of 1.5

  7. Interactive nuclear plant analyzer for VVER-440 reactor

    International Nuclear Information System (INIS)

    Shier, W.; Horak, W.; Kennett, R.

    1992-05-01

    This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator

  8. VVER-440 fuel cycles possibilities using modified FA design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.; Razym, V.; Dostal, M.; Jenik, J.; Krupar, P.

    2009-01-01

    A nearly equilibrium five-year cycle has been achieved at Dukovany NPP over the last years. This means that working fuel assemblies (WFA) with an average enrichment of 4.25 w% (control assemblies (CA) with an average enrichment of 3.82 w%) are normally loaded and reloaded for five years. Operation at uprated thermal power (105% of the original one, increase from 1375 MW t to 1444 MW t ) is being prepared by use of WFA with an average enrichment of 4.38 w% (CA with an average enrichment of 4.25 w%). With the aim of fuel cycle economy improvement, the fuel residence time in the core has to be prolonged up to six years with one cycle duration time up to 18 months and preserving loadings with very low leakage. In order to achieve this goal, at least neutron-physical characteristics of FA must be improved and such changes should be evaluated from other viewpoints. Some particular changes have already been analyzed earlier. Designs of new fuel assemblies with higher (and in the central part of a FA the highest possible, i.e. 4.95 w%) enrichment with preserving low pin power non-uniformity are described in the presented paper. An FA with an average enrichment of 4.66 w% (lower than originally evaluated) containing six fuel pins with 3.35 w% Gd 2 O 3 content was selected in the end. Fuel pins have bigger pellet diameter, bigger pin pitch and thinner FA shroud. A newly designed FA was evaluated from the viewpoint of physics (pin power non-uniformity, criticality of fuel at transport and storage and determination of basic quantities for spent fuel storage purposes by ORIGEN code), thermo-hydraulics (comparison of subchannel output temperatures and the departure from nucleate boiling ratio - DNBR) and mechanical properties. The purpose of this study was to simulate an FA subject to the loads during its six- year lifetime whereas normal working conditions were taken into account. There are presented two models with different shroud thickness undergoing these analyses. Both

  9. Determination of mixing factors for VVER-440 fuel assembly head

    Energy Technology Data Exchange (ETDEWEB)

    Tóth, S., E-mail: toth@reak.bme.hu [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary); Aszódi, A. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Műegyetem rkp. 9, H-1111 Budapest (Hungary)

    2013-11-15

    CFD models have been developed for the heads of the old, the present and the new type VVER-440 fuel assemblies using the experience of a former validation process. With these models temperature distributions are investigated in the heads of some typical assemblies and the in-core thermocouple signals are calculated. The analyses show that the coolant mixing is intensive but not-perfect in the assembly heads. The difference between the thermocouple signal and the cross-sectional average temperature at the measurement level depends on the assembly type. Using the results of these CFD calculations the weight factors of the rod bundle regions for the in-core thermocouple have been determined. With these factors the thermocouple signals are estimated and the results are statistically tested using the registered data of the Hungarian nuclear power plant. This test shows that the deviations between the measured and the calculated temperatures can be significantly decreased and consequently monitoring uncertainties can be reduced with using the weight factors.

  10. ASTEC applications to VVER-440/V213 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: ivstt@nextra.sk; Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir

    2014-06-01

    Since the beginning of ASTEC development by IRSN and GRS the code was widely applied to VVER reactors. In this paper, at first specific features of VVER-440/V213 reactor design that are important from the modelling point of view are briefly described. Then the validation of ASTEC code with focus on its applicability to VVER reactors is briefly summarised and the results obtained with the ASTEC V2.0-rev1 version for the ISP-33 PACTEL natural circulation experiment are presented. In the next section the application of ASTEC V2.0-rev1 code in upgrade of VVER-440/V213 NPPs to cope with consequences of severe accidents is described. This upgrade includes adoption of in-vessel retention via external reactor vessel cooling and installation of large capacity passive autocatalytic recombiners. Results of analysis with focus on corium localisation and stabilisation inside reactor vessel, hydrogen control in confinement and prevention of long-term confinement pressurisation are presented.

  11. Fuel Cycle of VVER-1000: technical and economic aspects

    International Nuclear Information System (INIS)

    Kosourov, E.; Pavlov, V.; Pavlovichev, A.

    2009-01-01

    The paper contains estimations of dependences of technical and economic characteristics of VVER-1000 fuel cycle on number of charged FAs and their enrichment. In the study following restrictions were used: minimum quantity of loaded fresh FAs is equal 36 FAs, a maximum one - 78 (79) FAs and fuel enrichment is limited by value 4,95 %. The following technical and economic characteristics are discussed: cycle length, average burnup of spent fuel, specific consumption of natural uranium, specific quantity of separative work, annual production of thermal energy, fuel component of electrical energy cost, electricity generation cost. Results of estimations are presented as dependences of researched characteristics on cycle length, quantity of loaded FAs and their enrichments. The presented information allows to show tendencies and ranges of technical and economic characteristics at change of fuel cycle parameters. This information can be useful for definition of the fuel cycle parameters which satisfy the requirements of power system and exploiting organizations. (authors)

  12. Comparison of PWR-IMF and FR fuel cycles

    International Nuclear Information System (INIS)

    Darilek, Petr; Zajac, Radoslav; Breza, Juraj; Necas, Vladimir

    2007-01-01

    The paper gives a comparison of PWR (Russia origin VVER-440) cycle with improved micro-heterogeneous inert matrix fuel assemblies and FR cycle. Micro-heterogeneous combined assembly contains transmutation pins with Pu and MAs from burned uranium reprocessing and standard uranium pins. Cycle analyses were performed by HELIOS spectral code and SCALE code system. Comparison is based on fuel cycle indicators, used in the project RED-IMPACT - part of EU FP6. Advantages of both closed cycles are pointed out. (authors)

  13. Advanced fuel cycles and burnup increase of WWER-440 fuel

    International Nuclear Information System (INIS)

    Proselkov, V.; Saprykin, V.; Scheglov, A.

    2003-01-01

    Analyses of operational experience of 4.4% enriched fuel in the 5-year fuel cycle at Kola NPP Unit 3 and fuel assemblies with Uranium-Gadolinium fuel at Kola NPP Unit 4 are made. The operability of WWER-440 fuel under high burnup is studied. The obtained results indicate that the fuel rods of WWER-440 assemblies intended for operation within six years of the reviewed fuel cycle totally preserve their operability. Performed analyses have demonstrated the possibility of the fuel rod operability during the fuel cycle. 12 assemblies were loaded into the reactor unit of Kola 3 in 2001. The predicted burnup in six assemblies was 59.2 MWd/kgU. Calculated values of the burnup after operation for working fuel assemblies were ∼57 MWd/kgU, for fuel rods - up to ∼61 MWd/kgU. Data on the coolant activity, specific activity of the benchmark iodine radionuclides of the reactor primary circuit, control of the integrity of fuel rods of the assemblies that were operated for six years indicate that not a single assembly has reached the criterion for the early discharge

  14. New code for VVER-440 loading pattern design

    International Nuclear Information System (INIS)

    Bajgl, J.; Lehmann, M.

    1999-01-01

    This paper describes the main attributes of a new computer program OPTIMAL used for loading pattern design in Dukovany NPP (4 reactors VVER-440). We have been developed this program in Nuclear Research Institute Rez since 1994 on the base of special contract between Dukovany NPP and Nuclear Research Institute Rez. General information about the optimisation methodology is given in the first part. The organisation of the optimisation process is described in part 2. Construction of the optimisation functional is shown in part 3. Procedures used during one-cycle optimisation are described in part 4. (Authors)

  15. Modernizing the VVER-440/230

    International Nuclear Information System (INIS)

    Mink, F.J.

    1991-01-01

    The modernization of the VVER-440/230s is not fundamentally different from backfit projects on older pressurized water reactors which Westinghouse has completed elsewhere. However, carrying out such programmes only makes sense if the plants are expected to continue operation for their projected life or beyond. This clearly requires some licensing and political stability; both are essential if investors in the upgrading project are to be found. (author)

  16. The in-core fuel management code system for VVER reactors

    International Nuclear Information System (INIS)

    Cada, R.; Krysl, V.; Mikolas, P.; Sustek, J.; Svarny, J.

    2004-01-01

    The structure and methodology of a fuel management system for NPP VVER 1000 (NPP Temelin) and VVER 440 (NPP Dukovany) is described. It is under development in SKODA JS a.s. and is followed by practical applications. The general objectives of the system are maximization of end of cycle reactivity, the minimization of fresh fuel inventory for the minimization of fed enrichment and minimization of burnable poisons (BPs) inventory. They are also safety related constraints in witch minimization of power peaking plays a dominant role. General structure of the system consists in preparation of input data for macrocode calculation, algorithms (codes) for optimization of fuel loading, calculation of fuel enrichment and BPs assignment. At present core loading can be calculated (optimized) by Tabu search algorithm (code ATHENA), genetic algorithm (code Gen1) and hybrid algorithm - simplex procedure with application of Tabu search algorithm on binary shuffling (code OPAL B ). Enrichment search is realized by the application of simplex algorithm (OPAL B code) and BPs assignment by module BPASS and simplex algorithm in OPAL B code. Calculations of the real core loadings are presented and a comparison of different optimization methods is provided. (author)

  17. Investigation of station blackout scenario in VVER440/v230 with RELAP5 computer code

    Energy Technology Data Exchange (ETDEWEB)

    Gencheva, Rositsa Veselinova, E-mail: roseh@mail.bg; Stefanova, Antoaneta Emilova, E-mail: antoanet@inrne.bas.bg; Groudev, Pavlin Petkov, E-mail: pavlinpg@inrne.bas.bg

    2015-12-15

    Highlights: • We have modeled SBO in VVER440. • RELAP5/MOD3 computer code has been used. • Base case calculation has been done. • Fail case calculation has been done. • Operator and alternative operator actions have been investigated. - Abstract: During the development of symptom-based emergency operating procedures (SB-EOPs) for VVER440/v230 units at Kozloduy Nuclear Power Plant (NPP) a number of analyses have been performed using the RELAP5/MOD3 (Carlson et al., 1990). Some of them investigate the response of VVER440/v230 during the station blackout (SBO). The main purpose of the analyses presented in this paper is to identify the behavior of important VVER440 parameters in case of total station blackout. The RELAP5/MOD3 has been used to simulate the SBO in VVER440 NPP model (Fletcher and Schultz, 1995). This model was developed at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences (INRNE-BAS), Sofia, for analyses of operational occurrences, abnormal events and design based scenarios. The model provides a significant analytical capability for specialists working in the field of NPP safety.

  18. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  19. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2009-01-01

    The SCORPIO-VVER system includes following features: 1) Validation of plant measurements and identification of sensor failures. 2) Optimum combination of measurements and calculations to obtain precise values of important parameters. 3) On-line 3D power distribution calculation with pin power reconstruction. 4) Limit checking and thermal margin calculation allowing for surveillance of VVER core limits such as DNBR, Sub-cooling margin, FdH and FQ peeking factors. 5) Integrated modules for monitoring fuel performance and coolant activity for identification of fuel failures. 6) Predictive capabilities and strategy planning, offering the possibility to check the consequences of operational manoeuvres in advance, prediction of critical parameters, etc. 7) Convenient monitoring of approach to criticality during reactor start-up. 8) Automated transition between cycles (fuel reload). The SCORPIO-VVER core monitoring system with its flexible and modular framework successfully responses to the plant operating needs and advances in nuclear fuel cycle strategies and fuel design. Modular framework allows for easy modifications of the system and implementation of new methods in physical modules. Even if the system is installed only on VVER-440 reactors, it could be adapted for VVER-1000 needs

  20. Regeneration and localization of radioactive waste in the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Egorov, N.N.; Kudryavtsev, E.G.; Nikipelov, B.V.; Polyakov, A.S.; Zakharkin, B.S.; Mamaev, L.A.

    1993-01-01

    Normal functioning of the nuclear-power industry is only possible with a closed fuel cycle, including regeneration of the spent fuel from atomic power plants, the production and recycling of the secondary fuel, and localization of the radioactive waste. Despite the diversity of contemporary attitudes toward the structure of the nuclear fuel cycle around the world, the closure of the fuel cycle has been fundamental to the atomic-power industry in the USSR since the very beginning, and has taken on even greater significance in Russia today. From the beginning, the idea of a closed fuel cycle has been based essentially on one fundamental criterion: the concept of expanded productivity on the basis of fuel regeneration, i.e., the economic factor. Important as economic factors are, safety issues have taken on great significance in recent years: not only power-station reactors but all the ancillary stages of the fuel cycle must meet fundamentally new reliability, safety, and environmental hazards. The RT-1 plant is a versatile operation, regenerating spent fuel from VVER-440, BN-350, and BN-600 reactors, nuclear icebreakers and submarines, research reactors, and other power units. The plant can reprocess 400 ton/year of basic VVER-440 fuel. World-class modern processes have been introduced at the plant, meeting the necessary quality standards: zonal planning, remote operation to eliminate direct contact of the staff with radioactive material, extensive monitoring and control systems, multistage gas-purification systems, and new waste-treatment methods

  1. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  2. Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

    Science.gov (United States)

    Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

    2018-02-01

    The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

  3. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  4. Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

    International Nuclear Information System (INIS)

    Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

    2007-01-01

    The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

  5. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    Felix, E.; Dessars, N.

    2003-01-01

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  6. Outlook on to fuel cycle perspectives at WWER-440

    International Nuclear Information System (INIS)

    Stech, S.; Bajgl, J.

    2005-01-01

    Current internal fuel cycle in NPP Dukovany 4x440 MWe is shortly characterized with new types of fuel assemblies and advanced fuel cycles which have been introduced in the last years. The modernization activities accomplished until now might be extrapolated to the further period in fuel design - mechanic, thermal-hydraulic and neutronic respectively - with additional increase in fuel enrichments and burnups on the way to the 6-year cycle. Reaktor power up rating together with Unit thermal efficiency improvements could bring an increase in the electric output to the value nearly 500 MWe. The reasons are given for long-term cooperation with Fuel Supplier and Plant Designer in the area of fuel cycle as well as in Unit Design Basis. All innovations mentioned in the article including future fuel and fuel cycle changes might be a quite realistic perspective at the end of the first decade of the new century (Authors)

  7. Standard and hydrazine water chemistry in primary circuit of VVER 440 units

    International Nuclear Information System (INIS)

    Burclova, J.

    1992-01-01

    Standard ammonia-potassium-boron water chemistry of 8 units with VVER 440 in CSFR is discussed as well as the corrosion product activity in the coolant during steady state and shut-down period and surface activity, dose rate build-up and occupational radiation exposure. Available data on hydrazine application (USSR, Hungary) indicate the possibility of the radiation field decreasing. Nevertheless the detailed analysis of 55 cycles of operation under standard water chemistry in Czechoslovakia allows to expect the comparable results for both water chemistries. (author)

  8. Main results on pilot operation during 5 years of the 3rd generation fuel in VVER-440 reactors of Kola NPP

    International Nuclear Information System (INIS)

    Saprykin, V.; Sumarokov, M.; Gagarinskiy, A.; Sumarokova, A.; Adeev, V.

    2015-01-01

    In the report the results of comparison of main neutron-physical data of exploitation of nuclear fuel are presented for the average enrichment (on U - 235) of 4.87 for the 2nd and 3rd (12 piece) generations with the results of calculations by the complex of the programs KASKAD for 5 fuel loadings of Kola NPP Unit 4 with the reactor VVER- 440. The basic feature of fuel of the 3rd generation as compared with the 2nd is a presence of ribs of inflexibility at corners instead of cover of the fuel assembly and also the increased amount of uranium. The arrangement of fuel rods with different enrichment in fuel assemblies of the 2nd and 3rd generations is chosen identical for the convenient comparison of neutronic and thermohydraulic characteristics of the fuel of different generations. The fuel of 3rd generation was situated in the core symmetrically to the fuel of 2nd one

  9. Information about AER WG a on improvement, extension and validation of parametrized few-group libraries for VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Mikolas, P.

    2009-01-01

    Joint AER Working Group A on 'Improvement, extension and validation of parameterized few-group libraries for VVER-440 and VVER-1000' and AER Working group B on 'Core design' eighteenth meeting was hosted by Skoda JS a.s. in Plzen (Czech Republic) during the period of 4 to 6 May 2009. There were present altogether 16 participants from 6 member organizations and 13 presentations were read. Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation. Presentations were devoted to some aspects of few group libraries preparations and to the benchmark dealing with VVER-440 follower modeling in calculations. Gy. Hegyi gave some new information about NURESIM-NURISP EU project (ZR-6), R. Zajac spoke about the development of data libraries for codes BIPR-7 and PERMAK, P. Darilek compared FA's with Gd during burning process and Yu. Bilodid described further development of plutonium-based burnup history modeling in DYN3D burnup calculations. G. Hordosy presented results of control rod follower induced local power peaking computational benchmark and J. Svarny described Monte Carlo VVER-440 control rod follower benchmark computations. Future activities are also shortly described in the end of the paper. (author)

  10. Contemporary and prospective fuel cycles for WWER-440 based on new assemblies with higher uranium capacity and higher average fuel enrichment

    International Nuclear Information System (INIS)

    Gagarinskiy, A.A.; Saprykin, V.V.

    2009-01-01

    RRC 'Kurchatov Institute' has performed an extensive cycle of calculations intended to validate the opportunities of improving different fuel cycles for WWER-440 reactors. Works were performed to upgrade and improve WWER-440 fuel cycles on the basis of second-generation fuel assemblies allowing core thermal power to be uprated to 107 108 % of its nominal value (1375 MW), while maintaining the same fuel operation lifetime. Currently intensive work is underway to develop fuel cycles based on second-generation assemblies with higher fuel capacity and average fuel enrichment per assembly increased up to 4.87 % of U-235. Fuel capacity of second-generation assemblies was increased by means of eliminated central apertures of fuel pellets, and pellet diameter extended due to reduced fuel cladding thickness. This paper intends to summarize the results of works performed in the field of WWER-440 fuel cycle modernization, and to present yet unemployed opportunities and prospects of further improvement of WWER-440 neutronic and operating parameters by means of additional optimization of fuel assembly designs and fuel element arrangements applied. (Authors)

  11. Informational system to assist decision making at spent nuclear fuel transportation from VVER-440, VVER-1000 and RBMK-1000 nuclear power plants

    International Nuclear Information System (INIS)

    Kuryndin, A.V.; Kirkin, A.M.; Stroganov, A.A.

    2012-01-01

    The developed informational system provides an automated estimations of nuclear and radiation safety parameters during spent nuclear fuel transportation from WWER-440 and WWER-1000 and RBMK-1000 nuclear power plants to the nuclear fuel cycle facilities, and allows us to determine the optimum cask loading from the dose rates distribution outside of protection point of view [ru

  12. Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuopanportti, Jaakko [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

    2013-09-15

    An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

  13. Automatic loading pattern optimization tool for Loviisa VVER-440 reactors

    International Nuclear Information System (INIS)

    Kuopanportti, Jaakko

    2013-01-01

    An automatic loading pattern optimization tool called ALPOT has been developed for Loviisa VVER-440 reactors. The ALPOT code utilizes combination of three different optimization methods. The first method is the imitation of the equilibrium pattern that is the optimized pattern in case the cycle length and the operation conditions are constant and the same shuffling pattern is repeated from cycle to cycle. In practice, the algorithm imitates assemblies' operation year distribution of the equilibrium pattern stochastically. The function of the imitation algorithm is to provide initial patterns quickly for the next optimization phase, which is performed either with the stochastic guided binary search algorithm or the deterministic burnup kernel method depending on the choice of the user. The former is a modified version of the standard binary search. The standard version goes through all possible swaps of the assemblies and chooses the best swap at each iteration round. The guided version chooses one assembly, tries to swap it with every other possible assembly and performs the best swap at each iteration round. The search is guided so that the algorithm chooses the assemblies at or near the most restrictive fuel assembly first. The kernel method creates burnup kernel functions to estimate burnup variations that are required to achieve desired changes in the power distribution of the reactor. The idea of the kernel method is first determine the optimal burnup distribution that minimizes the maximum relative assembly power using the created kernel functions and a common solver routine. Then, the burnups of the available fuel assemblies are matched with the obtained burnup distribution. (orig.)

  14. Advanced fuel cycles options for LWRs and IMF benchmark definition

    International Nuclear Information System (INIS)

    Breza, J.; Darilek, P.; Necas, V.

    2008-01-01

    In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

  15. Operation of Nuclear Fuel Based on Reprocessed Uranium for VVER-type Reactors in Competitive Nuclear Fuel Cycles

    Energy Technology Data Exchange (ETDEWEB)

    Troyanov, V.; Molchanov, V.; Tuzov, A. [TVEL Corporation, 49 Kashirskoe shosse, Moscow 115409 (Russian Federation); Semchenkov, Yu.; Lizorkin, M. [RRC ' Kurchatov Institute' (Russian Federation); Vasilchenko, I.; Lushin, V. [OKB ' Gidropress' (Russian Federation)

    2009-06-15

    Current nuclear fuel cycle of Russian nuclear power involves reprocessed low-enriched uranium in nuclear fuel production for some NPP units with VVER-type LWR. This paper discusses design and performance characteristics of commercial nuclear fuel based on natural and reprocessed uranium. It presents the review of results of commercial operation of nuclear fuel based on reprocessed uranium on Russian NPPs-unit No.2 of Kola NPP and unit No.2 of Kalinin NPP. The results of calculation and experimental validation of safe fuel operation including necessary isotope composition conformed to regulation requirements and results of pilot fuel operation are also considered. Meeting the customer requirements the possibility of high burn-up achieving was demonstrated. In addition the paper compares the characteristics of nuclear fuel cycles with maximum length based on reprocessed and natural uranium considering relevant 5% enrichment limitation and necessity of {sup 236}U compensation. The expedience of uranium-235 enrichment increasing over 5% is discussed with the aim to implement longer fuel cycles. (authors)

  16. Investigations of SPND noise signals in VVER-440 reactors

    International Nuclear Information System (INIS)

    Kiss, S.; Lipcsei, S.; Hazi, G.

    2001-01-01

    This paper describes and characterises SPND noise measurements of an operating VVER-440 nuclear reactor. Characteristics of the signal can be radically influenced by the geometrical properties of the detector and the cable and by the measuring arrangement. Structure of phase spectra showing propagating perturbations measured on uncompensated SPN detectors is studied through models.(author)

  17. In core-fuel management in approach to equilibrium of WWER-440 reactor; Prelazni rezim iskoriscenja goriva u nuklearnom reaktoru tipa VVER-440

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, N [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    For the need of in core fuel management and prediction of fuel cycle costs as well as operating of a nuclear power plant behaviour of main physical parameters and refueling scheme during approach to equilibrium operation are indispensable. An estimation of a refueling scheme during forst six years of exploitation for a commercially proven PWR reactor of WWER-440 type is shown in this paper. (author)

  18. Analysis of the VVER-440 reactor steam generator secondary side with the RELAP5/MOD3 code

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1993-01-01

    Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the steam generator secondary side. The model has been used to calculate steady state flow conditions in the secondary side of a VVER-440 reactor steam generator. (orig.) (1 ref., 9 figs., 2 tabs.)

  19. Contributions of Modranska potrubni a.s. to the safety improvement of piping systems and valves of NPS type VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Slach, J.

    2004-01-01

    The following activities are described: (i) Installation of pipe whip restraints on piping for high pressure and temperature steam and feed piping; (ii) Installation of air receivers for quick-acting valves with air actuator on VVER 440 units at the Jaslovske Bohunice V2 NPP; (iii) Replacement of the technical water distribution system material in the reactor hall of the Temelin VVER 1000 units; Installation of measuring nozzles on main steam piping DN 600 at the Temelin VVER 1000 units. (P.A.)

  20. Neutron dosimetry in EDF experimental surveillance programme for VVER-440 nuclear power plants

    International Nuclear Information System (INIS)

    Brumovsky, M.; Erben, O.; Novosad, P.; Zerola, L.; Hogel, J.; Trollat, C.

    2001-01-01

    Fourteen chains containing experimental surveillance material specimens of the VVER 440/213 nuclear power reactor pressure vessels were irradiated in the surveillance channels of the Nuclear Power Plant Dukovany in the Czech Republic. The irradiation periods were one, two or three cycles. The chains contained different number and types of containers, the omitted ones were replaced by chain elements. All of the containers were instrumented with wire neutron fluence detectors, some of the containers in the chain had spectrometric sets of neutron fluence monitors. For the absolute fluence values evaluation it was taken into account time history of the reactor power and local changes of the neutron flux along the reactor core height, correction factors due to the orientation of monitors with respect to the reactor core centre. Unfolding programs SAND-II or BASA-CF were used. The relative axial fluence distribution was obtained from the O-wire measurements. Neutron fluence values above 0.5 MeV energy and above 1.0 MeV energy in the container axis on the axial positions of the sample centres and fluence values in the geometric centre of the samples was calculated making use the exponential attenuation model of the incident neutron beam. Received fast neutron fluence values can be used as reference values to all VVER-440 type 213 nuclear power plant reactors. (author)

  1. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-01-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

  2. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  3. Experimental investigation of power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    Mikus, J.

    2004-01-01

    This paper presents some results of the axial power (fission density) distribution measurements carried out on the light-water, zero-power reactor LR-0 in a WWER-440 type core in vicinity of the WWER-440 control rod model at zero boron concentration in moderator, modelling the conditions at the end of the WWER-440 fuel cycle. Further information concerns the control rod model description, specification of the LR-0 core, fuel assemblies and measurement conditions. The aim of performed experiment is enlargement of the available power peaking database to enable the validation of the calculation codes by means of the measured data that correspond to the end of WWER-440 fuel cycle. (author)

  4. Assessment of computer codes for VVER-440/213-type nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Szabados, L.; Ezsol, Gy.; Perneczky [Atomic Energy Research Institute, Budapest (Hungary)

    1995-09-01

    Nuclear power plant of VVER-440/213 designed by the former USSR have a number of special features. As a consequence of these features the transient behaviour of such a reactor system should be different from the PWR system behaviour. To study the transient behaviour of the Hungarian Paks Nuclear Power Plant of VVER-440/213-type both analytical and experimental activities have been performed. The experimental basis of the research in the PMK-2 integral-type test facility , which is a scaled down model of the plant. Experiments performed on this facility have been used to assess thermal-hydraulic system codes. Four tests were selected for {open_quotes}Standard Problem Exercises{close_quotes} of the International Atomic Energy Agency. Results of the 4th Exercise, of high international interest, are presented in the paper, focusing on the essential findings of the assessment of computer codes.

  5. Experience of developments and implementation of advanced fuel cycles of WWER-440 reactors

    International Nuclear Information System (INIS)

    Gagarinski, A.A.; Lizorkin, M.P.; Novikov, A.N.; Proselkov, V.N.; Saprykin, V.V.

    2000-01-01

    The paper presents the experience of development and implementation of advanced four- and five-year fuel cycles in the WWER-440 reactors, the results of experimental operation of the new fuel design and the main neutronic characteristics of the core. (Authors)

  6. Main results of post-irradiation examinations of new-generation fuel assemblies VVER-1000

    International Nuclear Information System (INIS)

    Zvir, E.; Markov, D.; Polenok, V.; Zhitelev, V.; Kobylyansky, G.

    2009-01-01

    To increase the competitiveness of Russian nuclear fuel at the foreign market and to improve its technical and economic performance in order to provide a necessary level of safety, it is necessary to solve certain important tasks: Increase of fuel burn-up; Extension of operational lifetime of fuel assemblies and operational reliability of nuclear fuel; Introduction of cost-beneficial and flexible fuel cycles. Alternative fuel assemblies TVSA VVER-1000 and TVS-2 are used as a basis to optimize the nuclear fuel and develop advanced fuel cycles for nuclear power plants with VVER-1000 reactor types. Four fuel assemblies TVSA operated during 1 and up to 6 reactor cycles, reference fuel assembly TVS-2 operated during three reactor cycles and achieved an average fuel burnup of 48MW·day/kgU as well as failed fuel assembly TVS-2 operated during one cycle were examined at RIAR in recent years. The main objectives of these examinations were to obtain experimental data in support of operational integrity of products or to find out reasons of their failure. The performed post-irradiation examinations confirmed the operational integrity of alternative fuel assemblies TVSA including their geometrical stability up to the average fuel burnup of 55 MW·day/kgU over the fuel assembly (FA) (up to the maximal fuel burnup of ∼73 MW·day/kgU in fuel rods) and of TVS-2 up to the average fuel burnup of 48 MW·day/kgU over the fuel assembly. The changes introduced in the design of VVER-1000 fuel assembly during the development of alternative fuel assembly TVSA and TVS-2 did not make any negative effect on fuel rods. It was proved that causes of fuel rod failure were not related to design features of fuel assemblies. The design features and operating conditions of fuel assemblies under examinations are briefly described. Post-irradiation examinations proved the geometrical stability of fuel assemblies TVSA and TVS-2 under operation up to the fuel burnup of ∼50 MW day/kgU, as for the

  7. Power peak in vicinity of WWER-440 control rod at end of fuel cycle

    International Nuclear Information System (INIS)

    Mikus, J.

    2003-01-01

    This paper presents some results of the axial power distribution measurements carried out in a WWER-440 type core on the light-water, zero-power reactor LR-O in the vicinity of the WWER-440 control rod model at zero boron concentration in moderator. Further presented information concern the description of the control rod model, LR-0 core arrangement, specification of the fuel assemblies and measurement conditions. The aim of performed experiment is enlargement of the available 'power peaking database' to enable the calculation codes validation also by means of data that correspond to the end of WWER-440 fuel cycle (Authors)

  8. Development of TVSA VVER-1000 fuel

    International Nuclear Information System (INIS)

    Samoilov, O.; Kaydalov, V.; Romanov, A.; Falkov, A.; Morozkin, O.; Sholin, E.

    2013-01-01

    The TVSA fuel assemblies with a rigid angle-piece skeleton operate at 21 VVER-1000 units of Kalinin NPP, and Ukrainian, and Czech and Bulgarian NPPs. The total of more than 6,000 TVSA fuel assemblies have been fabricated. High lifetime performance has been achieved, namely, the maximum FA burnup is 65 MW∙day/kgU; maximum fuel rod burnup is 72 MW∙day/kgU; the lifetime is 50,000 EFPH. The TVSA fuel assembly is being improved to enhance its technical and economic performance and competitiveness of the Russian fuel for the VVER-1000 reactor: 1) Reliability and safety are being enhanced; repairability is being ensured. 2) High burnup levels in fuel are being ensured. 3) The uranium content in FAs is being increased. 4) The operational life is being extended. 5) Thermal-technical characteristics of FAs are being improved. The basic TVSA fuel assembly design evolved into the TVSA-PLUS with the fuel column elongated by 150 mm. The TVSA-PLUS fuel assembly has been in operation since 2010 at Kalinin NPP power units; an eighteen-month cycle is implemented at the uprated power of 104%. The TVSA-12PLUS fuel assembly has been developed with an elongated fuel column, optimized spacer grid positions (the spacer grid pitch is 340 mm) and with ensuring higher rigidity for the skeleton. It is provided for that fuel rods with the elevated uranium content and mixing intensifier grids will be used. The TVSA-T is developed for VVER-1000 reactor cores at the Temelin NPP. The TVSA-T is characterized by a load-carrying skeleton formed with angle-pieces and combined spacer grids that incorporate mixer grids. The TVSA-T design won the international tender to supply fuel to the Temelin NPP in the Czech Republic, and currently Temelin NPP Unit 1 and 2 are operating with the cores fully loaded with TVSA-Ts

  9. Test facility of the VVER-440 condensation-type pressure suppression system

    International Nuclear Information System (INIS)

    Wolff, H.; Arndt, S.

    2004-01-01

    Since the early nineties, GRS has supported regulatory authorities in Central and Eastern Europe in performing safety assessments of nuclear power plants. Especially studies of the condensation-type pressure suppression system of VVER-440/V-213-type plants have been important in this respect. Major steps in demonstrating complete functioning of the condensation-type pressure suppression system under accident conditions by experiments run in the Russian large scale test facility, BC V-213, have been completed in the past two years within the framework of various international experimental programs. The test results were used to validate specifically for power plants with VVER-400/V-213 reactors the COCOSYS GRS computer code, which is used in the safety assessments. The results of recalculations of the C02 EREC test, which simulates a break of a main steam pipe, demonstrate the present state of validation of COCOSYS for VVER condensation-type pressure suppression systems. (orig.) [de

  10. Qualification of UT methods and systems used for in-service inspections of VVER 440 vessels

    International Nuclear Information System (INIS)

    Skala, Z.; Vit, J.

    2003-01-01

    SKODA JS has been performing automated in-service inspections VVER reactor pressure vessels for more than twenty years. All of these inspections were performed by ultrasonic pulse echo method, combined from 1996 with eddy current testing. The Time of Flight Diffraction Method (TOFD) is one of modern methods of ultrasonic testing. The accuracy of sizing the through wall extent of a flaw by TOFD is much better than the accuracy achievable by the pulse echo method. A series of laboratory tests were performed by SKODA JS and confirmed the suitability of TOFD method for VVER reactor parts testing. The Czech Atomic law demands the qualification of systems and methods used for the in-service inspections of nuclear reactors. The qualification is done in accordance with ENIQ methodology and consists of preparation of the Technical Justification and practical tests made under the surveillance of Qualification Body. SKODA JS intends to qualify systems and methods used for the automated ultrasonic testing of VVER 440 and VVER 1000 reactor components from the inner as well as from the outer surface. The accuracy of the flaw through wall extent sizing by TOFD was confirmed by the qualification of methods and systems used for the testing of VVER 440 vessel circumferential weld and so the TOFD method shall be used routinely by SKODA JS for the inspection of vessel circumferential welds root area and for sizing of flaws exceeding the acceptance level. (author)

  11. Phase and structural transformations in VVER-440 RPV base metal after long-term operation and recovery annealing

    Science.gov (United States)

    Kuleshova, E. A.; Gurovich, B. A.; Maltsev, D. A.; Frolov, A. S.; Bukina, Z. V.; Fedotova, S. V.; Saltykov, M. A.; Krikun, E. V.; Erak, D. Yu; Zhurko, D. A.; Safonov, D. V.; Zhuchkov, G. M.

    2018-04-01

    This study was carried out to evaluate the possibility of 1st generation VVER-440 reactors lifetime extension by recovery re-annealing with the respect to base metal (BM). Comprehensive studies of the structure and properties of BM templates (samples cut from the inner surface of the shells in beltline region) of operating VVER-440 reactor (after primary standard recovery annealing 475 °C/150 h and subsequent long-term re-irradiation within reactor pressure vessel (RPV)) were conducted. These templates were also subjected to laboratory re-annealing 475 °C/150 h. TEM, SEM and APT studies of BM after laboratory re-annealing revealed significant recovery of radiation-induced hardening elements (Cu-rich precipitates and dislocation loops). Simultaneously a process of strong phosphorus accumulation at grain boundaries occurs since annealing temperature corresponds to the maximum reversible temper brittleness development. The latter is not observed for VVER-440 weld metal (WM). Comparative assessment of the properties return level for the beltline BM templates after recovery re-annealing 475 °C/150 h showed that it does not reach the one typical for beltline WM after the same annealing.

  12. On detection of the possible use of VVERs for unreported production of plutonium. Final report for the period July 1988 - December 1989

    International Nuclear Information System (INIS)

    Simov, R.; Nelov, N.; Stoyanova, I.; Kovachev, N.; Yonchev, P.

    1989-01-01

    The study includes an analysis of the feasibility of unreported production of plutonium-239 in VVER-440 reactors. It is shown that for VVER-440 reactors 36 natural uranium oxide fuel assemblies in the peripheral region of the core need to be loaded to produce 8 kg of extra plutonium in one cycle. Substituting the peripheral fuel assemblies with natural uranium oxide fuel assemblies, the changes in the power peaking are negligible and do not affect reactor safety. Unreported production outside the core is not practical due to physical and mechanical constraints, low flux level, etc. The feasibility of unreported removal of irradiated material in spent fuel cask has been also assessed. After about a month cooling time, still within the refueling period, the irradiated natural uranium fuel assemblies could be removed off-site without significant health hazard to the workers. To improve the effectiveness of the safeguards objectives, additional inspection activities are suggested. 10 figs

  13. ANDREA 2.2 and 2.3. Advances in modelling of VVER cores

    Energy Technology Data Exchange (ETDEWEB)

    Havluj, Frantisek; Hejzlar, Jonatan; Vocka, Radim; Vysoudil, Jiri [UJV Rez, Husinec-Rez (Czech Republic)

    2017-09-15

    In 2016 a new version of code ANDREA for core design and reload safety analysis of VVER reactors has been released. The new code version includes several major improvements. The first of them is a seamless incorporation of short time kinetics calculations (without temperature feedback) into the code. This new feature accompanied by the possibility of excore detector signal predictions enables precise interpretation of dynamic measurements of control assembly weight during the reactor startup. Second important enhancement resides in new flexible format of cross section libraries and in new fuel temperature model based on results of TRANSURANUS fuel performance code. The new code version has been thoroughly tested and validated for both VVER440 and VVER-1000 reactors. Furthermore for the new version 2.3 which is to be released shortly we have implemented the possibility of fluent control assemblies' motion and of non-equidistant axial nodalization schemes in VVER-440 calculations.

  14. The status of the Bubbler Condenser Containment System for the Reactors of the VVER-440/213 Type

    International Nuclear Information System (INIS)

    Karwat, H.; Rosinger, H.E.

    1998-01-01

    VVER-440/213 Pressurized Water Reactors have a pressure-suppression containment structure called a 'Bubbler Condenser' tower which can reduce the design pressure of the entire containment following a design basis accident (DBA), such as a loss-of-coolant accident (LOCA). The bubbler condenser pressure suppression system provides reduction of the LOCA containment pressure by the condensation of released steam in a water pool. World-wide there are 14 nuclear power plants of the VVER-440/213 type in Eastern Europe and Russia. One of the safety concerns for the VVER-440/213 reactors relates to the ability of the bubbler condenser containment system to function satisfactorily and to maintain its integrity following certain postulated accidents and thus limit the release of radioactive material to the environment. The complicated geometry of the bubbler condenser unit, and the dependence on several moving devices and interlocks are the main doubts expressed by different specialists with regard to the design. General description of the bubbler condenser containment system, the physical processes, concerns and design assessment of the bubbler condenser containment system, presentation of the OECD's Unified Bubbler Condenser Research Project (UBCRP) and the European Commission PHARE/TACIS project. Recent utility investigations are also discussed

  15. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    International Nuclear Information System (INIS)

    Kostov, M.K.; Prato, C.A.; Stevenson, J.D.

    1993-01-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made

  16. Floor response spectra for seismic qualification of Kozloduy VVER 440-230 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kostov, M.K. [Bulgarian Academy of Sciences, Sofia (BG). Central Lab. for Seismic Mechanics and Earthquake Engineering; Ma, D.C. [Argonne National Lab., IL (United States); Prato, C.A. [Univ. of Cordoba (AR); Stevenson, J.D. [Stevenson and Associates, Cleveland, OH (US)

    1993-08-01

    In this paper the floor response spectra generation methodology for Kozloduy NPP, Unit 1-2 of VVER 440-230 is presented. The 2D coupled soil-structure interaction models are used combined with a simplified correction of the final results for accounting of torsional effects. Both time history and direct approach for in-structure spectra generation are used and discussion of results is made.

  17. Accident loads for a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Techy, Z. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Lajtha, G. [Institute for Electric Power Research (VEIKI), Budapest (Hungary); Taubner, R. [Institute for Electric Power Research (VEIKI), Budapest (Hungary)

    1995-08-01

    Specific features of the VVER-440/213 containment are the subdivided rectangular building and the localization system including the bubbler trays and air traps. Accident loads are calculated for a large break loss of coolant accident (LBLOCA). The maximum pressure and temperature loads are calculated with different codes during the blowdown phase of the LBLOCA. The uncertainty margins of the maximum pressure are given in this case. Sensitivity studies are performed for different leakage rates and hydraulic data of the containment. The effects of the active and passive spray systems on the depressurization are presented in this paper. The maximum pressure loads are also examined in case of degraded conditions of the localization system. (orig.).

  18. VVER-specific features regarding core degradation - Status Report

    International Nuclear Information System (INIS)

    Hozer, Z.; Trambauer, K.; Duspiva, J.

    1999-01-01

    and metal masses of VVER reactors results for some accident sequences in later core degradation. The unique construction of VVER-440 control assemblies plays a special role during accident progression, having several fuel assemblies below the core and creating a heterogeneous core structure with absorber assemblies. Some events (e.g. B 4 C melting in the VVER-1000) are more similar to processes in BWRs. The early phase of core degradation seems to be similar in VVERs and PWRs, however the role of boron steel absorber assembly melting can change the sequence for the VVER-440. Accident progression during the late phase of core degradation can be influenced by the interaction of molten material with lower plenum structures in VVER-440. The mechanism of vessel failure can show some differences due to the VVER bottom heads being elliptical and without penetrations, compared with those of PWRs which are hemispherical and where penetrations are present. The severe accident phenomena were compared with the help of categories used for PWRs and BWRs. Most of the phenomena were found not to be VVER-specific, which means that the phenomena takes place in a similar way in the reactor types compared. Very few phenomena were found not relevant to VVER (e.g. AIC control rod related ones). When the phenomena was given a VVER-specific character the role of design features was discussed. In some cases there was experimental evidence, however in most of the cases the effect of VVER design could only be estimated. Some phenomena are not known in detail even for Western LWRs, so the comparison can show only some likelihood of differences. The review of related experiments showed, that the VVER experimental database is not as extensive as that for PWRs and BWRs. A number of separate-effect and integral tests indicated that the behaviour of materials used in VVERs is generally similar to that of PWRs. Some specific areas are not covered by experiments at all and their investigation should be

  19. Fuel designs for VVER reactors

    International Nuclear Information System (INIS)

    Simonov, K.V.; Carbon, P.; Silberstein, A.

    1995-01-01

    That progresses in efficiency and safety through progresses in technology and better prediction with fully benchmarked upgraded computer codes is a common goal for on the one hand the original designer of the VVER reactors and their respective fuels and on the other hand for EVF a western company resulting from a combined force with highly diversified and complementary talents in reactor and fuel design and manufacturing. It can be expected that this new challenge and dialogue between the two Russian and European industrial ventures will be mutually beneficial and yield innovative and high quality products and as a consequence strong return will be produced for the best interest of utilities operating VVER reactors. (orig./HP)

  20. Analysis of some fuel characteristics deviations and their influence over WWER-440 fuel cycle design

    International Nuclear Information System (INIS)

    Stoyanova, I.; Kamenov, K.

    2001-01-01

    The aim of this study is to estimate the influence of some deviations in WWER-440 fuel assemblies (FA) characteristics upon fuel core design. A large number of different fresh fuel assemblies with enrichment of 3.5 t % are examined related to the enrichment, mass of initial metal Uranium and assembly shroud thickness. Infinite multiplication factor (Kinf) in fuel assembly has been calculated by HELIOS spectral code for basic assembly and for different FA with deviation of a single parameter. The effects from single parameter deviation (enrichment) and from two parameter deviations (enrichment and wall thickness) on the neutron-physics characteristics of the core are estimated for different fuel assemblies. Relatively week burnup dependence on Kinf is observed as result of deviation in the enrichment of the fuel and in the wall thickness of the assembly. An assessment of a FA single and two parameter deviations effects on design fuel cycle duration and relative power peaking factor is also considers in the paper. As a final conclusion can be settled that the maximum relative shortness of fuel cycle can be observed in the case of two FA parameters deviations

  1. Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

    International Nuclear Information System (INIS)

    Gurin, Andrey V.; Alekseev, P.N.

    2017-01-01

    This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

  2. Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system

    Energy Technology Data Exchange (ETDEWEB)

    Gurin, Andrey V. [National Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation); Alekseev, P.N.

    2017-09-15

    This paper presents a study of scenarios of transition to a closed fuel cycle in the system of nuclear power, built basing on resource availability requirements at the stage of full life-cycle reactors. Conventionally, there are three main scenarios for the development of nuclear energy: with VVER reactors operating in an open fuel cycle; with VVER reactors operating in a closed fuel cycle; and co-operating VVER and BN, operating in a closed fuel cycle. For the considered scenarios, a quantitative estimation of change in time of material balances were performed, including spent fuel balance, balance of plutonium, reprocessed and depleted uranium, radioactive waste, and the analysis of the fuel component of the cost of electricity.

  3. Economical Feedback of Increasing Fuel Enrichment on Electricity Cost for VVER-1000

    Directory of Open Access Journals (Sweden)

    Mohammed Saad Dwiddar

    2015-08-01

    Full Text Available A methodology of evaluating the economics of the front-end nuclear fuel cycle with a price change sensitivity analysis for a VVER-1000 reactor core as a case study is presented. The effect of increasing the fuel enrichment and its corresponding reactor cycle length on the energy cost is investigated. The enrichment component was found to represent the highly expenses dynamic component affecting the economics of the front-end fuel cycle. Nevertheless, the increase of the fuel enrichment will increase the reactor cycle length, which will have a positive feedback on the electricity generation cost (cent/KWh. A long reactor operation time with a cheaper energy cost set the nuclear energy as a competitive alternative when compared with other energy sources.

  4. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels; Microstructure et fragilisation des aciers de cuve des reacteurs nucleaires VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Hennion, A

    1999-03-15

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  5. Fuel for new Russian reactor VVER-1200

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, Ivan Nikitovich [GRPress, 21, Ordzhonikidze Street, 142103 Podolsk, Moscow region (Russian Federation)

    2009-06-15

    A great program is accepted in Russia on increasing the nuclear power capacities. The basis of the program is commissioning of VVER-1200 Units of AES-2006 design. This is largely an evolutionary project of VVER-1000 reactor plant. It is referred also to reactor core. The plant electric power is increased due to increase in the reactor thermal power and forcing the main parameters and the efficiency increase. With this, reactor pressure increases from 15,7 to 16,2 MPa. The reactor inlet temperature increases from 290 deg. C to 298 deg. C, and outlet temperature from 319 deg. C to 329 deg. C. In a set of the design for four Units (2 Units at Novovoronezh NPP and 2 Units at Leningrad NPP) two base fuel cycles are developed: 5 year and 3 year. To provide such fuel cycles the fuel loading is increased by 8 tons, as compared to VVER-1000 base design, due to fuel column increase by 200 mm and change of fuel pellet sizes. In the mentioned fuel cycles the average burnup in the unloaded batch will be {approx}57 MW.day/kg U and 52 MW.day/kg U (maximum burnup over FAs is 64,5 MW.day/kg U and 60,3 MW.day/kg U), respectively. Specific consumption of natural uranium will be reduced by 5% as compared to that reached at VVER-1000 reactor. In spite of increase in Unit power the limiting permissible fuel rod linear heat rate is decreased from 448 W/cm to 420 W/cm. Refueling pattern is used with small neutron escape. The safety criteria are used that were established for VVER-1000, except for those that did not comply with EUR. For instance, the number of leaky fuel rods under accident is limited. The more stringent requirements are stated on efficiency margin of CPS rods for reactor shutdown that is ensured by the increased number of CPS rods. The well-proved design of fuel assembly TVS-2 and its close modification TVS-2M, operated at Balakovo NPP and Rostov NPP, is laid down in the basis of the core design. The load-carrying component of this structure is a rigid skeleton formed by

  6. Experimental studies of resistance fretting-wear of fuel rods for VVER-1000 and TVS-KVADRAT fuel assemblies

    International Nuclear Information System (INIS)

    Makarov, V.; Afanasiev, A.; Egorov, Yu.; Matvienko, I.

    2015-01-01

    The paper covers the results of the studies performed to justify the wear resistance of fuel rods in contact with the spacer grids of TVS VVER-1000 fuel assembly and TVS-KVADRAT square fuel assembly of Russian design for PWR-900 reactor. The presented results of three testing stages comprise: Testing of mockup fuel rods of VVER TVS fuel assembly for fretting wear under the conditions of the water chemistry of VVER reactor; Testing models of different design embodiments of the fuel rods for VVER TVS fuel assembly for fretting wear in still cold water; Testing mockup fuel rods of TVS-KVADRAT square fuel assembly for PWR reactor for frettingwear under the conditions of PWR water chemistry. The effect of structural and operational factors was determined (amplitudes, fuel rod vibration frequencies, values of cladding-to-spacer grid cell gap for the depth of fuel rod cladding wear etc.), an assessment was made of the threshold values of fuel rod vibration parameters, which, if not exceeded, provide the absence of the fuel rod cladding fretting wear in the fuel rod-to spacer grid contact area. Key words: fretting wear, fuel rod, spacer grid, VVER, PWR (author)

  7. Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

    International Nuclear Information System (INIS)

    Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

    2008-01-01

    Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

  8. VVER-440 and VVER-1000 reactor dosimetry benchmark - BUGLE-96 versus ALPAN VII.0

    International Nuclear Information System (INIS)

    Duo, J. I.

    2011-01-01

    Document available in abstract form only, full text of document follows: Analytical results of the vodo-vodyanoi energetichesky reactor-(VVER-) 440 and VVER-1000 reactor dosimetry benchmarks developed from engineering mockups at the Nuclear Research Inst. Rez LR-0 reactor are discussed. These benchmarks provide accurate determination of radiation field parameters in the vicinity and over the thickness of the reactor pressure vessel. Measurements are compared to calculated results with two sets of tools: TORT discrete ordinates code and BUGLE-96 cross-section library versus the newly Westinghouse-developed RAPTOR-M3G and ALPAN VII.0. The parallel code RAPTOR-M3G enables detailed neutron distributions in energy and space in reduced computational time. ALPAN VII.0 cross-section library is based on ENDF/B-VII.0 and is designed for reactor dosimetry applications. It uses a unique broad group structure to enhance resolution in thermal-neutron-energy range compared to other analogous libraries. The comparison of fast neutron (E > 0.5 MeV) results shows good agreement (within 10%) between BUGLE-96 and ALPAN VII.O libraries. Furthermore, the results compare well with analogous results of participants of the REDOS program (2005). Finally, the analytical results for fast neutrons agree within 15% with the measurements, for most locations in all three mockups. In general, however, the analytical results underestimate the attenuation through the reactor pressure vessel thickness compared to the measurements. (authors)

  9. Assessment of In-vessel corium retention for VVER-440/V213

    International Nuclear Information System (INIS)

    Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

    2011-01-01

    In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

  10. Fuel leak testing performance at NPP Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Slugen, V.; Krnac, S.; Smiesko, I.

    1995-01-01

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR's. 1 tab., 2 figs., 3 refs

  11. Fuel leak testing performance at NPP Jaslovske Bohunice

    Energy Technology Data Exchange (ETDEWEB)

    Slugen, V; Krnac, S [Slovak Technical Univ., Bratislava (Slovakia); Smiesko, I [Nuclear Powr Plant EBO, Jaslovske Bohuce (Slovakia)

    1996-12-31

    The NPP Bohunice VVER-440 fuel leak testing experience are relatively extensive in comparison with other VVER-440 users. As the first Europe NPP was adapted Siemens (KWU) in core-sipping equipment to VVER-440 units and since this time were have done these tests also for NPP Paks (Hungary) and NPP Dukovany (Czech Republic). The occurrence of leaking fuel assemblies in NPP is in the last 5 years relatively stabilised and low. A significant difference can be observed between type V-230 (31 leaks) and type V-213 (1 leak). None of of the indicated leaking fuel assemblies has been investigated in the hot cell. Therefore cannot be confirm the effective causes of leak occurrence. Nevertheless, the fuel failure rate and the performance of leak testing in NPP Bohunice are comparable to the world standard at PWR`s. 1 tab., 2 figs., 3 refs.

  12. Present status and recent improvements of water chemistry at Russian VVER plants

    International Nuclear Information System (INIS)

    Mamet, V.; Yurmanov, V.

    2001-01-01

    Water chemistry is an important contributor to reliable plant operation, safety barrier integrity, plant component lifetime, radiation safety, environmental impact. Primary and secondary water chemistry guidelines of Russian VVER plants have been modified to meet the new safety standards. At present 14 VVER units of different generation are in operation at 5 Russian NPPs. There are eight 4-loop pressurised water reactors VVER-1000 (1000 MWe) and six 6-loop pressurised water reactors VVER-440 (440 MWe). Generally, water chemistry at East European VVER plants (about 40 VVER-440 and VVER-1000 units in Ukraine, Bulgaria, Slovakia, Czech Republic, Hungary, Finland and Armenia) is similar to water chemistry at Russian VVER plants. Due to similar design and structural materials some water chemistry improvements were introduced at East European plants after they has been successfully implemented at Russian plants and vice versa. Some water chemistry improvements will be implemented at modern VVER plants under construction in Ukraine, Slovakia, Czech Republic, Iran, China, India. (R.P.)

  13. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  14. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Gy. Ézsöl

    2012-01-01

    Full Text Available The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440, the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed.

  15. PMK-2, the First Integral Thermal-Hydraulics Tests for the Safety Evaluation of VVER-440/213 Nuclear Power Plants

    International Nuclear Information System (INIS)

    Ezsol, G.; Perneczky, L.; Szabados, L.; Toth, I.

    2012-01-01

    The PMK-2 facility is a full-pressure thermal-hydraulic model of the primary and partly the secondary circuit of the VVER-type units of Paks NPP. The facility was the first integral-type facility for VVERs. The PMK-2 was followed later by the PACTEL (for VVER-440), the ISB, and PSB for VVER-1000. Since the startup of the facility in 1985, 55 experiments have been performed primarily in international frameworks with the participation of experts from 29 European and overseas countries forming a scientific school to better understand VVER system behaviour and reach a high level of modelling of accident sequences. The ATHLET, CATHARE, and RELAP5 codes have been validated including both qualitative and quantitative assessments. The former was almost exclusively applied to the early phase of validation by integral experiments, while the quantitative assessments have been performed by the Fast Fourier Transform Based Method. Paper gives comprehensive information on the design features of PMK-2 facility with a special respect to the representativeness of phenomena, the experiments performed, and the results of the validation of ATHLET, CATHARE, and RELAP5 codes. Safety significance of the PMK-2 projects is also discussed

  16. Physical model of the nuclear fuel cycle simulation code SITON

    International Nuclear Information System (INIS)

    Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

    2017-01-01

    Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

  17. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1995-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  18. Application of the thermal-hydraulic codes in VVER-440 steam generators modelling

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, P; Vranca, L; Vaclav, E [Nuclear Power Plant Research Inst. VUJE (Slovakia)

    1996-12-31

    Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.

  19. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  20. Vver-1000 Mox core computational benchmark

    International Nuclear Information System (INIS)

    2006-01-01

    The NEA Nuclear Science Committee has established an Expert Group that deals with the status and trends of reactor physics, fuel performance and fuel cycle issues related to disposing of weapons-grade plutonium in mixed-oxide fuel. The objectives of the group are to provide NEA member countries with up-to-date information on, and to develop consensus regarding, core and fuel cycle issues associated with burning weapons-grade plutonium in thermal water reactors (PWR, BWR, VVER-1000, CANDU) and fast reactors (BN-600). These issues concern core physics, fuel performance and reliability, and the capability and flexibility of thermal water reactors and fast reactors to dispose of weapons-grade plutonium in standard fuel cycles. The activities of the NEA Expert Group on Reactor-based Plutonium Disposition are carried out in close co-operation (jointly, in most cases) with the NEA Working Party on Scientific Issues in Reactor Systems (WPRS). A prominent part of these activities include benchmark studies. At the time of preparation of this report, the following benchmarks were completed or in progress: VENUS-2 MOX Core Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); VVER-1000 LEU and MOX Benchmark (completed); KRITZ-2 Benchmarks: carried out jointly with the WPRS (formerly the WPPR) (completed); Hollow and Solid MOX Fuel Behaviour Benchmark (completed); PRIMO MOX Fuel Performance Benchmark (ongoing); VENUS-2 MOX-fuelled Reactor Dosimetry Calculation (ongoing); VVER-1000 In-core Self-powered Neutron Detector Calculational Benchmark (started); MOX Fuel Rod Behaviour in Fast Power Pulse Conditions (started); Benchmark on the VENUS Plutonium Recycling Experiments Configuration 7 (started). This report describes the detailed results of the benchmark investigating the physics of a whole VVER-1000 reactor core using two-thirds low-enriched uranium (LEU) and one-third MOX fuel. It contributes to the computer code certification process and to the

  1. Sequence of decommissioning of the main equipment in a central type VVER 440 V-230

    International Nuclear Information System (INIS)

    Andres, E.; Garcia Ruiz, R.

    2014-01-01

    IBERDROLA Ingenieria y Construccion S.A.U., leader of consortium with Empresarios Agrupados and INDRA, has developed the Basic Engineering for the decommissioning of contaminated systems and building of a VVER 440 V-230 Nuclear Power Plant, establishing the sequence and methodology for the main equipment fragmentation. For that, it has been designed dry and wet cutting zones to be set up in the area where steam generators, main cooling pumps and pressurizer are located; these components will be dismantled previously. (Author)

  2. Special features of embrittlement of welded joints in shells of VVER-type reactors

    International Nuclear Information System (INIS)

    Kasatkin, O.G.

    1999-01-01

    At present, the atomic power engineering of Russia and Ukraine is based on water-water energy reactors of the VVER-440 and VVER-1000 type, with the electric power of 440 and 1000 MW, respectively. The majority of the VVER-440 reactors are installed in Russia, and VVER-1000 reactors operate in Ukraine. The reactors' shell (RS) is produced from cylindrical shells and a dished end welded together by circular joints under a flux. The RS of the VVER-440 reactor is produced from 15Kh3MFA steel, and the VVER-1000 reactors are produced from 15Kh2NMFA steel. The shell of the VVER-1000 reactor has an internal austenite coating. The condition of the RS metal is determined mainly by the critical brittleness temperature T b at which the impact toughness of specimens with a sharp notch reaches 60 J/cm 2 . The energy reactors, working in western countries, are characterised by a service life of 40 years and discussion is being carried out to extend this lifetime to 60 years. The design service life of the domestic reactors varies from 30 (RS VVER-440) to 40 (RS VVER-1000) years. According to investigations, the service life of the shells of these reactors is restricted by the properties of welded joints which are characterised by higher susceptibility to embrittlement than that of the parent metal, especially due to a higher content in the weld of phosphorus (RS VVER-440) or nickel (RS VVER-1000). Therefore, some experts believe that the actual service life of the RS is shorter than the design life. The accurate evaluation of the service life of welded joints in the RS is very important for the safety of service and also in the economic aspects, because the unjustified decrease of the permissible service life and premature shutdown of units of the nuclear power station result in huge losses

  3. SCORPIO-VVER core monitoring and surveillance system with advanced capabilities

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2010-01-01

    In this work authors present 12 years of operation experience of core monitoring and surveillance system with advanced capabilities on nuclear power plants on 6 unit of VVER-440 type of reactors at two different NPPs. The original version of the SCORPIO (Surveillance of reactor CORe by PIcture On-line display) system was developed for the western type of PWR reactors. The first version of the SCORPIO-VVER Core Monitoring System for Dukovany NPP (VVER-440 type of reactor, Czech Republic) was developed in 1998. For SCORPIO-VVER implementation at Bohunice NPP in Slovakia (2001) the system was enhanced with startup module KRITEX.

  4. Improving nuclear safety of VVER-440 units

    International Nuclear Information System (INIS)

    Nochev, T.; Sabinov, S.

    2001-01-01

    In this paper authors deals with improvement of nuclear safety of WWER-440 units in Kozloduy NPP. Main directions for improving nuclear safety of WWER-440 units were: - to expand number of the design accident; - to increase reliability of equipment important for the safety; - to decrease the probability of initiating events; - improvements the integrity of the primary circuit (application LBB concept, qualification of the pressure safety valves to avoid pressurized thermal shock); - improvement of the fire protection; - improvement of the operation including upgrading and improvement of operational documents, implementation of new system for training the operators and etc.; - reassessment of the seismic response of the plant. Main actions were made at NPP Kozloduy to increase nuclear safety of VVER-440 units. 1. Modernization of Emergency High Pressure Safety Injection System. The modernization includes dividing of independent channels with reservation of active elements. Pumps were exchanged with more effective and reliable ones. HPSIS was increased reliability in general through decrease number of active elements and exchanged with passive. 2. For the purpose of avoiding fast cooling at the primary circuit and obtaining thermal shock of reactor vessel, Main Safety Insulation Valves are installed at NPP Kozloduy. 3. Modernization of Emergency power supplies AC. Oil breakers VMP-10 are exchanged with gas FS-4. 4. Generator breakers are installed to decrease probability of loss power supply and blackout. They provide reliable power supply to the system important for the safety in case of failure on generator. 5. I and C system has been qualified and optimized. 6. Reassessments of Limiting Conditions of Operation and new scram signals have been introduced. 7. An operators-oriented Informational System has been developed. It includes ensuring and updating of equipment data, new informational support of operator and etc. 8. A new auxiliary independent system for

  5. Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

    Energy Technology Data Exchange (ETDEWEB)

    Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

    2017-04-01

    Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

  6. Calculation study of the WWER-440 fuel performance for extended burnup

    International Nuclear Information System (INIS)

    Kujal, J.; Pazdera, F.; Barta, O.

    1984-01-01

    The results of preliminary calculational study of extended burnup cycling schemes impact on WWER-440 fuel performance are presented. Two high burnup schemes were proposed with three and four cycles, resp. Comparison was made with three cycle reference case. The thermal mechanical analysis was performed with PIN and RELA codes. The values of rod internal pressure, fuel centerline temperatures and fuel-cladding gap are expressed as function of power history. (author)

  7. Thermal-hydraulic studies on the safety of VVER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Tuunanen, J.

    1994-01-01

    The thesis includes several thermal-hydraulic analyses related to the Loviisa VVER-440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transients and development of a calculational model for calculation of boric concentrations in the reactor. In the first part of thesis, in the case of simulation of boric acid solution behaviour during long-term cooling period of LOCAs, experiments were performed in scaled-down test facilities. The experimental data together with the results of RELAP5/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. In the second part, in the case of simulation of horizontal generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments as well as earlier REWETT-III natural circulation tests, were analyzed with RELAP5/MOD3 Version 5m5 code. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAP5/MOD1-Eur, RELAP5/MOD3 and CATHARE codes. (56 refs., 9 figs.)

  8. Analysis of transients for NPP with VVER-440 using the code SiTAP

    International Nuclear Information System (INIS)

    Kalinenko, V.

    1994-06-01

    The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)

  9. NNSA / IAEA VVER reactor safety workshops. May 2002 - April 2003. Executive summary

    International Nuclear Information System (INIS)

    Evans, M.; Petri, M. C.

    2003-01-01

    Over the past year, the U.S. National Nuclear Security Administration (NNSA) has sponsored four workshops to compare the probabilistic risk assessments (PRAs) of Soviet-designed VVER power plants. The ''International Workshop on Safety of First-Generation VVER-440 Nuclear Power Plants'' was held on May 20-25, 2002, in Piestany, Slovakia. A short follow-on workshop was held in Bratislava, Slovakia, on November 5-6, 2002, to complete the work begun in May. Piestany was the location also for the ''International Workshop on Safety of Second-Generation VVER-440 Nuclear Power Plants'' (September 9-14, 2002) and the ''International Workshop on Safety of VVER-1000 Nuclear Power Plants'' (April 7-12, 2003). The four workshops were held in cooperation with the International Atomic Energy Agency (IAEA), the Nuclear Regulatory Authority of Slovakia (UJD), the Center for Nuclear Safety in Central and Eastern Europe (CENS), and Argonne National Laboratory (ANL). The objectives of the workshops were to identify the impact of the improvements on the core damage frequency; the contribution to the PRA results of different assumptions about events that can occur at the plants; and to understand, identify, and prioritize potential improvements in hardware and plant operation of VVER nuclear power plants. These objectives were achieved based on insights gained from recent PRAs completed by the plants and their technical support organizations. Nine first-generation VVER-440 plants (nominally of the VVER-440/230 design) are currently operating in Armenia, Bulgaria, Russia, and Slovakia. Sixteen VVER-440/213 plants are currently operating in the Czech Republic, Hungary, Russia, Slovakia, and Ukraine. Twenty-three VVER-1000 plants are currently operating in Bulgaria, the Czech Republic, Russia, and Ukraine. Eleven addition plants are in the advanced stages of construction in various parts of the world. The workshops reviewed the current configuration and safety status of each plant

  10. Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes

    Energy Technology Data Exchange (ETDEWEB)

    Aleshin, Sergey S.; Bikeev, Artem S.; Bolshagin, Sergey N.; Kalugin, Mikhail A.; Kosourov, Evgeniy K.; Pavlovichev, Aleksandr M.; Pryanichnikov, Aleksandr V.; Sukhino-Khomenko, Evgenia A.; Shcherenko, Anna I.; Shcherenko, Anastasia I.; Shkarovskiy, Denis A. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    Two types of calculations were made to compare BIPR-7A and MCU results for 3D full-scale models. First EPS (emergency protection system) efficiency and in-core power distributions were analyzed for an equilibrium fuel load of VVER-1000 assuming its operation within an 18-month cycle. Computations were performed without feedbacks and with fuel burnup distributed over the core. After 3D infinite lattices of full-scale VVER-1000 fuel assemblies (A's) with uranium fuel 4.4% enrichment and uranium-erbium fuel 4.4% enrichment and Er{sub 2}O{sub 3} 1 % wt were considered. Computations were performed with feedbacks and fuel burnup at the constant power level. For different time moments effective multiplication factor and power distribution were obtained. EPS efficiency and reactivity effects at chosen time moments were analyzed.

  11. The integration of fast reactor to the fuel cycle in Slovakia

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Necas, V.

    2009-01-01

    A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

  12. WWER-440 fuel cycles possibilities using improved fuel assemblies design

    International Nuclear Information System (INIS)

    Mikolas, P.; Svarny, J.

    2008-01-01

    Practically five years cycle has been achieved in the last years at NPP Dukovany. There are two principal means how it could be achieved. First, it is necessary to use fuel assemblies with higher fuel enrichment and second, to use fuel loading with very low leakage. Both these conditions are fulfilled at NPP Dukovany at this time. It is known, that the fuel cycle economy can be improved by increasing the fuel residence time in the core up to six years. There are at least two ways how this goal could be achieved. The simplest way is to increase enrichment in fuel. There exists a limit, which is 5.0 w % of 235 U. Taking into account some uncertainty, the calculation maximum is 4.95 w % of 235 U. The second way is to change fuel assembly design. There are several possibilities, which seem to be suitable from the neutron - physical point of view. The first one is higher mass content of uranium in a fuel assembly. The next possibility is to enlarge pin pitch. The last possibility is to 'omit' FA shroud. This is practically unrealistic; anyway, some other structural parts must be introduced. The basic neutron physical characteristics of these cycles for up-rated power are presented showing that the possibilities of fuel assemblies with this improved design in enlargement of fuel cycles are very promising. In the end, on the basis of neutron physical characteristics and necessary economical input parameters, a preliminary evaluation of economic contribution of proposals of advanced fuel assemblies on fuel cycle economy is presented (Authors)

  13. Core designs of modern VVER projects

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Kushmanov, S.; Vjalitsyn, V.; Vasilchenko, R.

    2015-01-01

    The presented operational experience of TVS - 2M (pilot-commercial operation started in 2006 at Balakovo NPP -1) enables to use it as reference for new projects because of similarity in designs and operational conditions. In the paper main parameters of fuel cycles, stability to impact of damaging factors, pilot operation of MG, new alloys, ADF and NTMC, upgrade of FA - 2M for the further power uprating, profiling of Gd-fuel rods for 18-month Fuel Cycle (FC) and perfection of absorber element design are the discussed issues. At the end author concluded that: 1) Core designs of new projects AES-2006 and VVER-TOI are based on extensive successful operational experience of the close prototype of TVS - 2M. 2) All improvements both of technical and economic parameters of fuel are subjected to representative examination by pilot operation at the power units with VVER-1000 being close prototypes of new designs

  14. VVER-440 training simulators upgrades - Experience of CORYS T.E.S.S

    International Nuclear Information System (INIS)

    Bartak, J.; Fallon, B.

    2006-01-01

    The paper presents recent projects of upgrading screen operated simulators of VVER-440 nuclear power plants to full scale replica simulators, implemented by CORYS TESS. Control room replica full scope simulators were built for the Bohunice NPP in Slovakia and the Novovoronezh NPP in Russia. The scope of simulation was extended to reflect the current status of the units, which have undergone significant modernization programs over the last few years. The paper describes the software and hardware adaptations and evolutions of the existing simulators, the implementation in the simulator of modern supervision systems as well as of systems and equipment designed in the seventies and still used on the reference units. The training benefits of parallel use of control room replica and screen-operated simulators in the training process are discussed. (author)

  15. Surveillance of WWER-440 fuel performance

    International Nuclear Information System (INIS)

    Simko, J.; Urban, P.

    1999-01-01

    In this lecture next problems of surveillance of WWER-440 fuel performance are presented: surveillance of WWER-440 fuel performance at Mochovce NPP; basic data of WWER-440 reactor; in-core reactor measuring system 'SVRK'; basic level of SVRK; information output of basic level of SVRK; surveillance of fuel performance; table of permissible operation conditions of the reactor; limitation of the unit 1 power at the beginning of the operation; cyclic changes of power; future perspectives

  16. Experimental support of the bleed and feed accident management measures for VVER-440/213 type reactors

    International Nuclear Information System (INIS)

    Szabados, L.

    2002-01-01

    In the original design of the VVER-440/213 type nuclear power plants event oriented emergency operating procedures (EOP) were implemented. In the last years, however, new symptom based procedures of Westinghouse-type have been developed and partly implemented for the plants in Central Europe including the Paks Nuclear Power Plant. Paper gives a short review of the experiments performed in the PMK-2 facility to study the effectiveness of the bleed and feed strategies and to get experimental data bases for the validation of thermohydraulic system codes like RELAP5, ATHLET and CATHARE.(author)

  17. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  18. Current state of spent fuel management in the Russian Federation

    International Nuclear Information System (INIS)

    Makarchuk, T.F.; Spichev, V.V.; Tikhonov, N.S.; Simanovsky, V.M.; Tokarenko, A.I.; Bespalov, V.N.

    1998-01-01

    Twenty nine power units of nine nuclear power plants of total installed capacity 22 GW(e) are now in operation in the Russian Federation. They produce approximately 12% of electric power in the country. The annual spent fuel arising is about 790 tU. The spent fuel from VVER-440 and BN-600 is reprocessed at the RT-1 plant near Chelyabinsk. The VVER-1000 spent fuel is planned to be reprocessed at the reprocessing plant RT-2 which is under construction near Krasnoyarsk. The RBMK-1000 spent fuel is not reprocessed because of its low fissile content. It is meant to be stored in intermediate storage facilities at the NPP sites and in a centralized storage facility during a period not less than 50 years and then to be disposed of in geological formations. State of the art of spent fuel reprocessing, storage and transportation is considered in the paper. Problems of nuclear fuel cycle back-end in Russia are taken into account. (author)

  19. Fire protection upgrading of four Russian 440/230 VVER units

    International Nuclear Information System (INIS)

    Corsini, G.; Yelfimov, S.

    1995-01-01

    The main goal of TACIS 3.6 a project funded by the Commission of the European Communities (CEC), was the front-end engineering for upgrading the Fire Protection System (FPS) of the safety-related equipment of Novovoronezh, Units 3 and 4, and Kola, Units 1 and 2, VVER 440/230 nuclear power plants. As a first step, all the safety-related equipment had to be identified, evaluation criteria had to be established and the existing FPS reviewed against the criteria. In the second step, the selection of the upgrading measures, depending on feasibility and cost estimate, has been accomplished, room by room. The third step, carried out on schedule and completed end July 95, has been essentially the preparation of the Technical Specifications for procurement of the needed equipment including remaining detail engineering. The Russian sub-contactor Atom Energo Project (AEP), who have been the designers of these older NPP s, have done the work with the Italian Ansaldo as the consultants of their Russian colleagues. Practical aspects of the engineering work are discussed and examples of improvements selected for retrofitting described. (author)

  20. Effect of uncompensated SPN detector cables on neutron noise signals measured in VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, S. E-mail: kisss@sunserv.kfki.hu; Lipcsei, S. E-mail: lipcsei@sunserv.kfki.hu; Hazi, G. E-mail: gah@sunserv.kfki.hu

    2003-03-01

    The Self Powered Neutron Detector (SPND) noise measurements of an operating VVER-440 nuclear reactor are described and characterised. Signal characteristics may be radically influenced by the geometrical properties of the detector and the cable, and by the measuring arrangement. Simulator is used as a means of studying the structure of those phase spectra that show propagating perturbations measured on uncompensated SPN detectors. The paper presents measurements with detectors of very different sizes (i.e. 20 cm length SPNDs and the 200 cm length compensation cables), where the ratios of the global and local component differ significantly for the different detector sizes. This phenomenon is used up for signal compensation.

  1. Experience in modernization of safety I and C in VVER 440 nuclear power plants Bohunice V1 and Paks

    International Nuclear Information System (INIS)

    Martin, M.

    2000-01-01

    For nuclear power plants which have been in operation for more than 15 years, backfitting or even complete replacement of the instrumentation and control (I and C) equipment becomes an increasingly attractive option, motivated not only by the objective to reduce the cost of I and C system maintenance and repair but also to prolong or at least to safeguard the plant life-time: optimized spare-part management through use of standard equipment; reduction of number and variety of different items of equipment by implementing functions stepwise in application software; adding new functionality in the application software which was not possible in the old technology due to lack of space; safeguarding of long-term After-Sales-Service. Some years ago Bohunice V1 NPP, Slovak Republic and Paks NPP, Hungary intended to replace parts of their Safety I and C, mainly the Reactor Trip System, the Reactor Limitation System and the Neutron Flux Excore Instrumentation and Monitoring Systems. After a Basic Engineering Phase in Bohunice V1 and a Feasibility Study in Paks both plants decided to use the Siemens' Digital Safety I and C System TELEPERM XS to modernize their plants. Both Bohunice, Unit 2 and Paks, Unit 1 have been back on line for over six months with the new Digital Safety I and C. At the present time Bohunice, Unit 1 and within the next few months Paks, Unit 2 will be replaced. Trouble-free start-ups and no major problems under operation in the first two plants were based on: thorough understanding of the VVER 440 technology; comprehensive planning together with the plant operators and authorities; the possibility to adapt TELEPERM XS to every plant type; the execution of extensive pre-operational tests. Regarding these modernization measures Siemens, as well as the above Operators, have gained considerable experience in the field of I and C Modernization in VVER 440 NPPs. Important aspects of this experience are: how to transfer the VVER technology to TELEPERM XS; how to

  2. The experimental definition of the acoustic standing wave series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

    International Nuclear Information System (INIS)

    Bulavin, V.V.; Pavelko, V.I.

    1995-01-01

    On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops. (author)

  3. Greifswald and Rheinsberg: East European VVERs with a new mission

    International Nuclear Information System (INIS)

    Sterner, H.; Leushacke, D.; Rittscher, D.

    1995-01-01

    Plans for the decommissioning of the VVER reactors at Greifswald and Rheinsberg in the former German Democratic Republic are described. The decision to decommission the eight VVER-440s (two of which are still under construction) at Greifswald and the Russian prototype PWR, VVER-2, at Rheinsberg, was taken because of a lack of public acceptance and financial considerations. Three main phases are scheduled for the project. The first, the post-operation phase, has already commenced at both sites. It involves: the operation of all systems needed safely to store fuel elements, to remove them and to condition operational waste; dismantling of mainly inactive, systems not needed for fuel handling; and system decontamination. The second phase comprises the dismantling of contaminated systems, remote dismantling and conditioning of the dismantled material. Finally, in the site restoration phase, following the dismantling of remaining systems and building decontamination and demolition, the site will be adapted for other uses. Three projects for new site use at Greifswald being pursued. As a first step, an international Technology Centre is to be created to collect and evaluate information on decommissioning experience. The centre will require international financial support and co-ordination. New facilities are already being constructed to deal with decontamination of dismantled materials and the interim storage of radioactive waste and reactor fuel. (UK)

  4. TAREG 2.01/00 Project, ''Validation of neutron embrittlement for VVER 1000 and 440/213 RPVs, with emphasis on integrity assessment''

    International Nuclear Information System (INIS)

    Ahlstrand, R.; Margolin, B.; Kostylev, V.; Yurchenko, E.; Akbashev, I.; Piminov, V.; Nikolaev, Y.; Koshkin, V.; Kharshenko, V.; Chyrko, L.; Bukhanov, V.; Comsa, O.

    2012-01-01

    The irradiation embrittlement and integrity of the VVER reactors has been an important issue in many EC supported TACIS and PHARE projects since 1990. In the EC annual program 2000 two TACIS projects (TAREG 2.01/00 and 2.01/03) were approved on the issue in order to improve the neutron irradiation embrittlement databases, elaborate new trend curves for the embrittlement and to assess the integrity of the RPVs (Reactor Pressure Vessel) by analysing PTS transients (Pressurized Thermal Shock) for some selected Russian and Ukrainian VVER 1000 and 440/213 NPPs. In this paper the TAREG 2.01/00 project is briefly described with some details from the twin project 2.01/03, which served as a materials testing project, providing inputs for the 1st project. As a result of the project new trend curves for neutron irradiation embrittlement were elaborated, based on upgraded and more reliable surveillance results databases. The PTS study shows that the integrity of the selected VVER RPVs can be ensured to the end of RPV design life. (author)

  5. Non-Proliferative, Thorium-Based, Core and Fuel Cycle for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Todosow, M.; Raitses, G.; Galperin, A.

    2009-01-01

    Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the

  6. Technology of repair of selected equipment in the power plant type VVER 440

    International Nuclear Information System (INIS)

    Barborka, J.; Magula, V.

    1998-01-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored

  7. Technology of repair of selected equipment in the power plant type VVER 440

    Energy Technology Data Exchange (ETDEWEB)

    Barborka, J.; Magula, V. [Welding Research Inst. (WRI), Bratislava (Slovakia)

    1998-11-01

    This article is divided in two parts: The first part is studying the effect of individual parameters by the usual and pulsed welding of 15CH2MFA steel. It can be concluded that by use of mechanized or automatic TIG process in PC position with addition of a cold wire with high nickel content the desired quality of repair welded joints of a pressure vessel of VVER 440 reactor can be achieved. Based on the results of the second laboratory study of the renovation technology applied for the rotary surfaces of pressure-tight cover and spindle of the main closing armature type DN 500 it can be concluded, that the developed technology for surfacing the sealing surfaces by TIG process with addition of a high-nickel cold wire the functional capability of the mentioned parts can be fully restored.

  8. Application of the coupled code system KIKO3D/ATHLET to the boron dilution transients in VVER-440 Type NPP

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Trosztel, I.

    2003-01-01

    The transient caused by a perturbation of boron concentration and coolant temperature at the inlet of a Russian developed reactor (VVER-440) is reanalysed as a part of the modernisation (introduction of a new type profiled fuel) and power upgrading (up to 108 %) project. This task is one of the basis cases to be investigated in the safety analysis of the pressurized water reactor (PWR) and the criteria of Anticipated Operational Occurrences (AOO) have to be fulfilled for it. First detailed planning calculations were performed with the thermal hydraulic system code ATHLET and neutron physical code system KARATE-440 to find out the appropriate initial parameter set taking into account the active safety system of the NPP. Finally the most reactive case is analysed by the KIKO3D/ATHLET coupled system code. Whereas the investigation is done for safety analysis, conservative assumptions are imposed on reactivity characteristics. Moreover at the core inlet no-mixing is supposed from the unaffected loops. The presented calculations show, how the coupled code system with a detailed description of plant functions and core behaviour can help to understand better the local phenomena in the study of a potential risk of dilution accident, as it offers the possibility to evaluate the plant safety in a more realistic and versatile manner. (author)

  9. Fuel reliability experience in Finland

    International Nuclear Information System (INIS)

    Kekkonen, L.

    2015-01-01

    Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

  10. FP 6 EU - COVERS. Coordination action - VVER safety research

    International Nuclear Information System (INIS)

    Vasa, I.

    2008-01-01

    In this work research program of the European Union FP 6 - COVERS coordinated by the NRI Rez is presented. COVERS is designed to improve professional and communication environment in the specific area covering all aspects of safe and reliable operation of nuclear power plants with VVER-440 and VVER-1000 reactors. Project Consortium is composed of 26 research and development, engineering and technical support organisations of European VVER-operating and other EU and non-EU countries.

  11. Modelling of thermal mechanical behaviour of high burn-Up VVER fuel at power transients with special emphasis on the impact of fission gas induced swelling of fuel pellets

    International Nuclear Information System (INIS)

    Novikov, V.; Medvedev, A.; Khvostov, G.; Bogatyr, S.; Kuzetsov, V.; Korystin, L.

    2005-01-01

    This paper is devoted to the modelling of unsteady state mechanical and thermo-physical behaviour of high burn-up VVER fuel at a power ramp. The contribution of the processes related to the kinetics of fission gas to the consequences of pellet-clad mechanical interaction is analysed by the example of integral VVER-440 rod 9 from the R7 experimental series, with a pellet burn-up in the active part at around 60 MWd/kgU. This fuel rod incurred ramp testing with a ramp value ΔW 1 ∼ 250 W/cm in the MIR research reactor. The experimentally revealed residual deformation of the clad by 30-40 microns in the 'hottest' portion of the rod, reaching a maximum linear power of up to 430 W/cm, is numerically justified on the basis of accounting for the unsteady state swelling and additional degradation of fuel thermal conductivity due to temperature-induced formation and development of gaseous porosity within the grains and on the grain boundaries. The good prediction capability of the START-3 code, coupled with the advanced model of fission gas related processes, with regard to the important mechanical (residual deformation of clad, pellet-clad gap size, central hole filling), thermal physical (fission gas release) and micro-structural (profiles of intra-granular concentration of the retained fission gas and fuel porosity across a pellet) consequences of the R7 test is shown. (authors)

  12. Partially closed fuel cycle of WWER-440

    International Nuclear Information System (INIS)

    Darilek, P.; Sebian, V.; Necas, V.

    2002-01-01

    Position of nuclear energy at the energy sources competition is characterised briefly. Multi-tier transmutation system is outlined out as effective back-end solution and consequently as factor that can increase nuclear energy competitiveness. LWR and equivalent WWER are suggested as a first tier reactors. Partially closed fuel cycle with combined fuel assemblies is briefed. Main back-end effects are characterised (Authors)

  13. The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

    International Nuclear Information System (INIS)

    Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

    2010-01-01

    LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

  14. Living PSA program for VVER 440/213 in the Czech Republic

    International Nuclear Information System (INIS)

    Husak, S.; Patrik, M.

    2000-01-01

    The paper presents an overview of a Living PSA concept in the Czech Republic for the VVER 440/213 NPP Dukovany unit. The first step of PSA program was a Level 1 basic study for Unit No. 1 which was completed in 1995. The main objective of the study was to determine the risk level of full power operation and its contributors as well as to reveal the weak points of the plant. Living PSA program for a Level 1 study has been afterwards established as a framework for all activities related to risk assessment and risk based decision-making support in NPP Dukovany. The basic parts of the project are: a management of PSA models and studies to implement design and procedures, modifications or new data inputs from data collection; continuous improvement based of new analyses, experiments or more detailed models; an extensions of the scope (external events, all plant operating modes, other sources of radioactive releases). The Living PSA program in NPP Dukovany provides basis for three kinds of PSA activities: risk assessment applications, risk monitoring and risk assessment of operational. (author)

  15. Development and application of Siton, a new fuel cycle simulation code

    International Nuclear Information System (INIS)

    Brolly, Aron; Szieberth, Mate; Halasz, Mate; Nagy, Lajos; Feher, Sandor

    2015-01-01

    As the result of the co-operation between the Centre for Energy Research (EK) and the Institute of Nuclear Techniques (NTI) a new fuel cycle simulation code called SITON was developed. Physical model of the code takes into account six facilities of the nuclear fuel cycle namely material stocks, spent fuel interim storages, plants for uranium enrichment, fuel fabrication, spent fuel reprocessing and reactors. Facilities can be linked in a flexible manner and their number is not limited. Lag time of the facilities and cooling time of the spent fuel, which are the two main parameters to introduce lag time into the fuel cycle, are taken into account. Material transfer between the facilities is modelled in a discrete manner tracking 52 nuclides and their short-lived decay daughters. Composition of the discharged fuel is determined by means of burn-up tables except for the 2400 MWth design of gas cooled fast reactor (GFR2400) which has a separate burn-up module developed at the NTI. To demonstrate the capabilities of SITON introduction of a GFR2400 into the Hungarian reactor park using the legacy spent fuel of the four presently operating VVER-440 units was simulated. 2040 was assumed as the commissioning date of the GFR2400 and recycling of its fuel was started as soon as possible. It was found that the plutonium content of the legacy spent fuel is sufficient to the start-up of only one GFR2400. There is an intermediate period between the commissioning of the reactor and the recycling of its first discharged fuel. Plutonium need of this period can be covered by the legacy spent fuel if the cooling time of the spent GFR2400 fuel is 2 years. If the cooling time is 5 years there will be a lack of plutonium in this period. To counterbalance this lack an EPR was started before the GFR2400 and its spent fuel was accumulated and reprocessed. Cooling time of the spent EPR fuel was also varied. Finally, an EPR only scenario is presented using two EPRs as a reference case

  16. Perspectives for practical application of the combined fuel kernels in VVER-type reactors

    International Nuclear Information System (INIS)

    Baranov, V.; Ternovykh, M.; Tikhomirov, G.; Khlunov, A.; Tenishev, A.; Kurina, I.

    2011-01-01

    The paper considers the main physical processes that take place in fuel kernels under real operation conditions of VVER-type reactors. Main attention is given to the effects induced by combinations of layers with different physical properties inside of fuel kernels on these physical processes. Basic neutron-physical characteristics were calculated for some combined fuel kernels in fuel rods of VVER-type reactors. There are many goals in development of the combined fuel kernels, and these goals define selecting the combinations and compositions of radial layers inside of the kernels. For example, the slower formation of the rim-layer on outer surface of the kernels made of enriched uranium dioxide can be achieved by introduction of inner layer made of natural or depleted uranium dioxide. Other potential goals (lower temperature in the kernel center, better conditions for burn-up of neutron poisons, better retention of toxic materials) could be reached by other combinations of fuel compositions in central and peripheral zones of the fuel kernels. Also, the paper presents the results obtained in experimental manufacturing of the combined fuel pellets. (authors)

  17. Inspection qualification as a tool to risk based ET ISI of VVER type SG tubes

    International Nuclear Information System (INIS)

    Horacek, L.

    2002-01-01

    A Pilot study on Eddy current inspection qualification of VVER 440 steam generator tubes, discussed in this paper, followed the ENIQ methodology principles and covered briefly the assumed scope of ET qualification, relevant elaborated qualification documents, known ISI limitations and a review of input information on component and defects determined for Eddy current inspection qualification of VVER 440 steam generator tubes. The information includes the fabrication of the test blocks with SG tube segments provided by intended defect simulations of realistic SCC type and basic data on the realistic SCC type defects manufacturing technology. Lessons learned from the development of manufacturing technology of SSC type of defects, regional blind tests, elaboration of the preliminary technical justification for Eddy current automated inspections, potential optimisation of inspection procedures, laboratory and practical open trials are summarised in the paper. The results of the Pilot study also especially in relation to POD curve being determined seem to be useful for practical operational ISI programme and Risk informed ISI decisions and the establishment of plugging criteria of VVER 440 and VVER 1000 type steam generator tubes. (orig.)

  18. Enhancing VVER annular proliferation resistance fuel with minor actinides

    International Nuclear Information System (INIS)

    Chang, G. S.

    2007-01-01

    to the reactivity control of the systems into which they are incorporated. In the study, a typical pressurized water reactor (PWR) VVER-1000 annular fuel unit lattice cell model with UO 2 fuel pins will be used to investigate the effectiveness of minor actinide reduction approach (MARA) for enhancing proliferation resistance and improving the fuel cycle performance in the intermediate term goal for future nuclear energy systems

  19. Validation of containment thermal hydraulic computer codes for VVER reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiri Macek; Lubomir Denk [Nuclear Research Institute Rez plc Thermal-Hydraulic Analyses Department CZ 250 68 Husinec-Rez (Czech Republic)

    2005-07-01

    Full text of publication follows: The Czech Republic operates 4 VVER-440 units, two VVER-1000 units are being finalized (one of them is undergoing commissioning). Thermal-hydraulics Department of the Nuclear Research Institute Rez performs accident analyses for these plants using a number of computer codes. To model the primary and secondary circuits behaviour the system codes ATHLET, CATHARE, RELAP, TRAC are applied. Containment and pressure-suppression system are modelled with COCOSYS and MELCOR codes, the reactor power calculations (point and space-neutron kinetics) are made with DYN3D, NESTLE and CDF codes (FLUENT, TRIO) are used for some specific problems.An integral part of the current Czech project 'New Energy Sources' is selection of a new nuclear source. Within this and the preceding projects financed by the Czech Ministry of Industry and Trade and the EU PHARE, the Department carries and has carried out the systematic validation of thermal-hydraulic and reactor physics computer codes applying data obtained on several experimental facilities as well as the real operational data. One of the important components of the VVER 440/213 NPP is its containment with pressure suppression system (bubble condenser). For safety analyses of this system, computer codes of the type MELCOR and COCOSYS are used in the Czech Republic. These codes were developed for containments of classic PWRs or BWRs. In order to apply these codes for VVER 440 systems, their validation on experimental facilities must be performed.The paper provides concise information on these activities of the NRI and its Thermal-hydraulics Department. The containment system of the VVER 440/213, its functions and approaches to solution of its safety is described with definition of acceptance criteria. A detailed example of the containment code validation on EREC Test facility (LOCA and MSLB) and the consequent utilisation of the results for a real NPP purposes is included. An approach to

  20. Large eddy simulation of a fuel rod subchannel

    International Nuclear Information System (INIS)

    Mayer, Gusztav

    2007-01-01

    In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

  1. Ambition to reach zero level failure in VVER 1000 with russian fuel

    International Nuclear Information System (INIS)

    Mečíř, V.

    2015-01-01

    The purpose of “The Zero Failure Level Project” is to bring to real operation of VVER 1000 units the dream of all utilities such as problem free and cost effective operation. This essentially turns into requirement on failure free fuel operation. At the same time the general requirements such as safety, cost effectiveness, operational flexibility, fuel cycle and fuel flexibility need to be satisfied. Several specific tasks were performed and many of them are still in process. Specific failure tree was developed in a format, which allows step by step failure tree improvement. Fuel types and its modifications, taking into account manufacturing conditions, were specified. In parallel with fuel types classification, real operational conditions were evaluated based on approximately 280 parameters by fuel assembly design features, operational procedures and practices and about 250 reactor unit parameters. As a result of this stage, groups of units with similar fuel operational conditions should be revealed and experience sharing database created. It is also recognized a need for consistent methods of operational data and data from pool side fuel assembly inspection. In the area of Foreign Material Exclusion activities closer cooperation between utility and supplier should be established including foreign material classification and improvement in root cause investigation

  2. Issues in the use of Weapons-Grade MOX Fuel in VVER-1000 Nuclear Reactors: Comparison of UO2 and MOX Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J.

    2005-05-27

    The purpose of this report is to quantify the differences between mixed oxide (MOX) and low-enriched uranium (LEU) fuels and to assess in reasonable detail the potential impacts of MOX fuel use in VVER-1000 nuclear power plants in Russia. This report is a generic tool to assist in the identification of plant modifications that may be required to accommodate receiving, storing, handling, irradiating, and disposing of MOX fuel in VVER-1000 reactors. The report is based on information from work performed by Russian and U.S. institutions. The report quantifies each issue, and the differences between LEU and MOX fuels are described as accurately as possible, given the current sources of data.

  3. Development of WWER-440 fuel. Use of fuel assemblies of 2-nd and 3-rd generations with increased enrichment

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Ananev, U.; Baranov, A.; Kukushkin, U.

    2009-01-01

    The problem of increasing the power of units at NPPs with WWER-440 is of current importance. There are all the necessary prerequisites for the above-stated problem as a result of updating the design of fuel assemblies and codes. The decrease of power peaking factor in the core is achieved by using profiled fuel assemblies, fuel-integrated burning absorber, FAs with modernized docking unit, modern codes, which allows decreasing conservatism of RP safety substantiation. A wide range of experimental studies of fuel behaviour has been performed which has reached burn-up of (50-60) MW·day/kgU in transition and emergency conditions, post-reactor studies of fuel assemblies, fuel rods and fuel pellets with a 5-year operating period have been performed, which prove high reliability of fuel, presence of a large margin in the fuel pillar, which helps reactor operation at increased power. The results of the work performed on introduction of 5-6 fuel cycles show that the ultimate fuel state on operability in WWER-440 reactors is far from being achieved. Neutron-physical and thermal-hydraulic characteristics of the cores of working power units with RP V-213 are such that actual (design and measured) power peaking factors on fuel assemblies and fuel rods, as a rule, are smaller than the maximum design values. This factor is a real reserve for power forcing. There is experience of operating Units 1, 2, 4 of the Kola NPP and Unit 2 of the Rovno NPP at increased power. Units of the Loviisa NPP are operated at 109 % power. During transfer to work at increased power it is reasonable to use fuel assemblies with increased height of the fuel pillar, which allows decreasing medium linear power distribution. Further development of the 2-nd generation fuel assembly design and consequent transition to working fuel assemblies of the 3-rd generation provides significant improvement of fuel consumption under the conditions of WWER-440 reactors operation with more continuous fuel cycles and

  4. Tightening unit EZ 250 for VVER 1000 type reactor pressure vessel head flange joints

    International Nuclear Information System (INIS)

    Ruchar, Miloslav; Nadenik, Tomas; Kroj, Ludek

    2010-01-01

    The programme of flange joints tightening by seals made of expanded graphite for VVER 440 and VVER 1000 reactor head flange joints is highlighted, and tightening units of row EZ 650 and EZ 650 TK and KNI for VVER 440 reactor head flange joints and EZ 250 tightening unit for VVER 1000 reactor head flange joints are described in detail. The main advantages of electronically controlled tightening units include: Precise and uniform compression of the gasket during the tightening procedure; Automated solution to the graphite relaxing problem during tightening; Possibility of diagnosis of the thread status of the joints being tightened; Alleviation of operator's tough work; Shorter time of tensioning associated with a lower collective doses; Quick preparation of tightening procedure report from the data measured; Calibration renewal is possible in advance at time of unit storage without the need to place it on a real joint. (P.A.)

  5. Analysis of core damage frequency: Nuclear power plant Dukovany, VVER/440 V-213 Unit 1, internal events. Volume 1: Main report

    International Nuclear Information System (INIS)

    Pugila, W.J.

    1994-01-01

    This report presents the final results from the Level 1 probabilistic safety assessment (PSA) for the Dukovany VVER/440 V-213 nuclear power plant, Unit 1. Section 1.1 describes the objectives of this study. Section 1.2 discusses the approach that was used for completing the Dukovany PSA. Section 1.3 summarizes the results of the PSA. Section 1.4 provides a comparison of the results of the Dukovany PSA with the results of other PSAs for different types of reactors worldwide. Section 1.5 summarizes the conclusions of the Dukovany PSA

  6. The results of postirradiation examinations of VVER-1000 and VVER-440 fuel rods

    Science.gov (United States)

    Dubrovin, K. P.; Ivanov, E. G.; Strijov, P. N.; Yakovlev, V. V.

    1991-02-01

    The paper presents the results of postirradiation examination of the fuel rods having different fuel-cladding gaps, pellet densities, pellet inner diameters and so on. The fuel rods were irradiated in the material science reactor (MR) of the Kurchatov Institute of Atomic Energy and at 4 unit of the Novo-Voronezh nuclear powerplant. Some data on fission gas release and rod geometry and compared with computer code predictions.

  7. Feasibility and usefulness of reconstructing obsolete power blocks of VVER-440 reactors

    International Nuclear Information System (INIS)

    Kirichenko, A.M.; Krushenik, S.D.; Sigal, M.V.; Kustov, V.P.

    1993-01-01

    At the present time, in Russia and in the East European countries there are atomic power stations with first-generation VVER-440 reactors built according to specification which no longer satisfy the more rigorous modern safety standards. Among these power stations are, in particular, the Novovoronezh and the Armenian Atomic Power Station and two blocks of the Kola Atomic Power Station. The search for technical solutions for modernizing these power blocks is complicated because two conditions which are hard to reconcile must be fulfilled: an acceptable safety level must be obtained and the rebuilding must be economically justifiable (particularly since the time of operation of a power block until its standard service life is over is short). Research work undertaken in the All-Union Scientific Research Institute of Atomic Power Stations has shown that one way of overcoming these difficulties may involve changing the operating conditions of the reactor assembly to a less demanding mode of operation. This solution implies an economically justified minimum of structural improvements, provides the required safety level, and prolongs the service life of the power block. The reduction of the thermal power, and consequently, the necessary transfer of a power block to another option

  8. Core design experience of WWER-440 reactors when they working on increased power level

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.; Melenchuk, I.

    2015-01-01

    The Kola NPP continues commercial operation of 2nd generation fuel (FA-2) and trial operation of 3rd generation fuel (FA-3), which has a number of design features providing the best operational characteristics. This report gives the results of VVER-440 core operation with FA-2 and FA-3 with enrichment increased up to 4.87%, and at the power level uprated to 107% of nominal power level. Brief analysis of obtained data is carried out. Peculiarities and techniques of developing loading patterns with new types of nuclear fuel for operation at the uprated power level are reviewed. (authors)

  9. BNFL Springfields Fuel Division

    International Nuclear Information System (INIS)

    Tarkiainen, S.; Plit, H.

    1998-01-01

    The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

  10. Fuel rod behaviour at high burnup WWER fuel cycles

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

    2003-01-01

    The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

  11. Advanced designs of VVER reactor plant

    International Nuclear Information System (INIS)

    Mokhov, V.A.

    2010-01-01

    The history of VVER reactors, current challenges and approaches to the challenges are highlighted. The VVER-1200 reactor of 3+ generation for AES-2006 units are under construction at the Leningrad 2 nuclear power plant (LNPP-2). The main parameters are listed and details are presented of the vessel, steam generator, and improved fuel. The issue of the NPP safety is discussed. Additional topics include the MIR-1200 reactor unit, VVER-600, and VVER-SCP (Generation 4). (P.A.)

  12. Stress corrosion cracking (Standard Astm G 30-90) in stainless steel 08X18H10T of swimming-pool that contain nuclear fuel in reactors V.V.E.R.-440

    International Nuclear Information System (INIS)

    Zamora R, L.; Herrera, V.

    1998-01-01

    The standard recommended practice for making and using 'U' bend stress corrosion test specimens; Designation G30-90 has been used as a laboratory tool to study the susceptibility of austenitic stainless steels and the other materials of test of intergranular stress corrosion cracking (IGSCC). The experiment has been development in a similar conditions of the chemical regime, the swimming-pool that containing nuclear fuel in borated water reactors VVER-440 in general this cladding by two films, one of carbon steel (04T26) and other with austenitic stainless steel 08X18HT (similar type 321) stabilized with titanium, the thickness of filler metals was to 4 to 8 mm. The specimens was prepare one plate with this characteristics, the welding was put in the part central with the following measurements of 160x15x5 mm. The specimens strips bent approximately 180 degrees around radius of curvature of R=14.5 mm and ε 1 = 17.2% and maintained in this plastically deformed condition during the test. And then preparing metallographically and exposure in environment of 12 and 40 gr./l of H 3 BO 3 70 Centigrade with or noting contaminants of NaCl. The results showed the initial cracks. (Author)

  13. Power release estimation inside of a fuel pin neighbouring a WWER-440 control rod

    International Nuclear Information System (INIS)

    Mikus, J.

    2006-01-01

    This work presents an estimation of the control rod (CR) influence in the WWER-440 core on the power release inside of a fuel pin neighbouring CR, that can have some consequences due to possible static and cyclic loads, for example fuel pin / fuel assembly bowing. For this purpose detailed (usual) axial power distribution measurements were performed in a WWER-440 type core on the light water, zero-power research reactor LR-0 in fuel pins near to an authentic CR model at zero boron concentration in moderator, modelling the conditions at the end of fuel cycle. To demonstrate the CR influence on power distribution inside of one fuel pin neighbouring CR, results of above measurements were used for estimation of the: 1) Axial power distribution inside of the investigated fuel pin in both opposite positions on its pellets surface that are situated to- and outwards CR and corresponding gradient of the (r, z) - power distribution in above opposite positions and 2) Azimuthal power distributions on pellet surface of the investigated fuel pin in horizontal planes at selected axial coordinates. Similar information can be relevant from the viewpoint of the fuel pin failures occurrence investigation

  14. Operational benchmark for VVER-1000, unit 6, Kozloduy NPP

    International Nuclear Information System (INIS)

    Apostolov, T.; Petrov, B.

    1999-01-01

    Benchmark calculations have been carried out using the 3D nodal code TRAPEZ. Global neutron-physics characteristics of the VVER-1000 core, Kozloduy NPP Unit 6, have been determined taking into account the real loading patterns and operational history of the first three cycles. The code TRLOAD has been used to perform the fuel reloading between any two cycles. The reactor and components descriptions as well as material compositions are given. The results presented include the critical boric acid concentration, the radial power distribution, the axial power distribution for the maximum overload assembly, and the burnup distribution at three different moments during each cycle. Calculated values have been compared with measured data. It is shown that the results obtained by the TRAPEZ code are in good agreement with the experimental data. The information presented could serve as a test case for validation of code packages designed for analyzing the steady-state operation of VVERs. (author)

  15. Typical design/qualification acceptance criteria for newly installed pipelines and equipment components of VVER-type NPPs

    International Nuclear Information System (INIS)

    Masopust, R.

    2003-01-01

    This paper describes in general the typical design/qualification acceptance criteria and seismic acceptance criteria in particular that are applicable for important to safety newly installed pipelines and equipment components of VVER-type already existing NPPs, specifically during the design verification phase of this newly installed equipment. These criteria are currently used for VVER 440-213 and VVER 1000 NPPs in Czech Republic and in Slovakia. The similar criteria are also used in Hungary. (author)

  16. Engineering Margin Factors Used in the Design of the VVER Fuel Cycles

    Science.gov (United States)

    Lizorkin, M. P.; Shishkov, L. K.

    2017-12-01

    The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

  17. Containment leak-tightness enhancement at VVER 440 NPPs

    International Nuclear Information System (INIS)

    Prandorfy, M.

    2000-01-01

    The hermetic compartments of WWER 440 NPPs fulfil the function of the containment used at NPPs all over the world. The purpose of the containment is to protect the NPP personnel against radioactive impact as well as to prevent radioactive leakage to the. environ ent during a lost of coolant accident. Leak-tightness enhancement in NPPs with WWER 440/213 and WWER 440/230 reactors is an important safety issue. New procedures, measures and methods were adopted at NPPs in Mochovce, Jaslovske Bohunice, Dukovany and PAKS for leak identification and sealing works performed by VUEZ Levice. (authors)

  18. Hydrogen mixing analyses for a VVER containment.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J.J.; Kostka, P.; Techy, Z.

    2002-02-25

    Hydrogen combustion may represent a threat to containment integrity in a VVER-440/213 plant owing to the combination of high pressure and high temperature. A study has been carried out using the GASFLOW 2.1 three-dimensional CFD code to evaluate the hydrogen distribution in the containment during a beyond design basis accident. The VVER-440/213 containment input model consists of two 3D blocks connected via one-dimensional (1D) ducts. One 3D block contains the reactor building and the accident localization tower with the suppression pools. Another 3D block models the air traps. 1D ducts represent the check valves connecting the accident localization tower with the air traps. The VVER pressure suppression system, called ''bubbler condenser,'' was modeled as a distributed heat sink with water thermodynamic properties. This model accounts for the energy balance. However, it is not currently possible to model dynamic phenomena associated with the water pools (e.g., vent clearing, level change). The GASFLOW 2.1 calculation gave detailed results for the spatial distribution of thermal-hydraulic parameters and gas concentrations. The range and trend of the parameters are reasonable and valuable. There are particularly interesting circulation patterns around the steam generators, in the bubbler tower and other primary system compartments. In case of the bubbler tower, concentration and temperature contour plots show an inhomogeneous distribution along the height and width, changing during the accident. Hydrogen concentrations also vary within primary system compartments displaying lower as well as higher (up to 13-20% and higher) values in some nodes. Prediction of such concentration distributions was not previously possible with lumped parameter codes. GASFLOW 2.1 calculations were compared with CONTAIN 1.2 (lumped parameter code) results. Apart from the qualitatively similar trends, there are, for the time being, quantitative differences between the

  19. Water chemistry experiences with VVERs at Kudankulam

    International Nuclear Information System (INIS)

    Rout, D.; Upadhyaya, T.C.; Ravindranath; Selvinayagam, P.; Sundar, R.S.

    2015-01-01

    Kudankulam Nuclear Power Project - 1 and 2 (Kudankulam NPP - 1 and 2) are pressurised water cooled VVERs of 1000 MWe each. Kudankulam NPP Unit - 1 is presently on its first cycle of operation and Kudankulam NPP Unit - 2 is on the advanced stage of commissioning with the successful completion of hot run related Functional tests. Water Chemistry aspects during various phases of commissioning of Kudankulam NPP Unit - 1 such as Hot Run, Boric acid flushing, initial fuel Loading (IFL), First approach to Criticality (FAC) are discussed. The main objectives of the use of controlled primary water chemistry programme during the hot functional tests are reviewed. The importance of the relevant water chemistry parameters were ensured to have the quality of the passive layer formed on the primary coolant system surfaces. The operational experiences during the 1 st cycle of operation of primary water chemistry, radioactivity transport and build-up are presented. The operational experience of some VVER units in the field of the primary water chemistry, radioactivity transport and build-up are presented as a comparison to VVER at Kudankulam NPP. The effects of the initial passivated layer formed on metal surfaces during hot run, activated corrosion products levels in the primary coolant under controlled water chemistry regime and the contamination/radiation situation are discussed. This report also includes the water chemistry related issues of secondary water systems. (author)

  20. Analysis of WWER-440 fuel performance under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Gunduz, Oe; Koese, S; Akbas, T [Atomenerjisi Komisyonu, Ankara (Turkey); Colak, Ue [Ankara Nuclear Research and Training Center (Turkey)

    1994-12-31

    FRAPCON-2 code originally developed for LWR fuel behaviour simulation is used to analyse the WWER-440 fuel rod behaviour at normal operational conditions. The code is capable of utilizing different models for mechanical analysis and gas release calculations. Heat transfer calculations are accomplished through a collocation technique by the method of weighted residuals. Temperature and burnup element properties are evaluated using MATPRO package. As the material properties of Zr-1%Nb used as cladding in WWER-440s are not provided in the code, Zircaloy-4 is used as a substitute for Zr-1%Nb. Mac-Donald-Weisman model is used for gas release calculation. FRACAS-1 and FRACAS-2 models are used in the mechanical calculations. It is assumed that the reactor was operated for 920 days (three consecutive cycles), the burnup being 42000 Mwd/t U. Results of the fuel rod behaviour analysis are given for three axial nodes: bottom node, central node and top node. The variations of the following characteristic fuel rod parameters are studied through the prescribed power history: unmoved gap thickness, gap heat transfer coefficient, fuel axial elongation, cladding axial elongation, fuel centerline temperature and ZrO-thickness at cladding surface. The value of each parameter is calculated as a function of the effective power days for the three nodes by using FRACAS-1 and FRACAS-2 codes for comparison.The results show that calculations with deformable pellet approximation with FRACAS-II model could provide better information for the behaviour of a typical fuel rod. Calculations indicate that fuel rod failure is not observed during the operation. All fuel rod parameters investigated are found to be within the safety limits. It is concluded, however, that for better assessment of reactor safety these calculations should be extended for transient conditions such as LOCA. 1 tab., 10 figs., 4 refs.

  1. Risk assessment basis for WWER-440 spent nuclear fuel

    International Nuclear Information System (INIS)

    Lascek, M.; Necas, V.; Darilek, P.

    2000-01-01

    The most problematic part of nuclear fuel cycle is its back end. Various high level waste management are available or under development (final disposal of spent assemblies in deep repository, reprocessing, partitioning, transmutation,...). Application of any method is connected with production of characteristic high level waste (amount, radio-toxicity, form,...) as well as various risk level for the environment and mankind. Strategy selection should be based on risk analysis also. The paper deals with assessment of risk, that is associated with WWER-440 spent fuel inventory. In order to evaluate the risk, the accumulated amount of the radioactive inventory is calculated and the decay of the long-lived radionuclides is computed by ORIGEN code. Analysis is oriented on calculation of hazard indexes for assessing the relative hazards of actinides, toxic and long-lived radionuclides. (Authors)

  2. The effect of the volumetric heat source distribution of the fuel pellet on the minimum DNBR ratio

    International Nuclear Information System (INIS)

    Hordosy, G.; Kereszturi, A.; Maroti, L.; Trosztel, I.

    1995-01-01

    The radial power distribution in a VVER-440 type fuel assembly is strongly non-uniform as a result of the water-gap between the shrouds and the moderator filled central tube. Consequently, it can be expected that the power density inside a single fuel rod is inhomogeneous, as well. In the paper the methodology and the results of coupled thermohydraulic and neutronic calculations are presented. The objective of the analysis was the investigation of the heat source distribution and the determination of the possible extent of the power non-uniformity in a corner rod which has always the highest peaking factor in a VVER-440 type assembly. The results of the analysis revealed that there can be a strong non-uniformity of power distribution inside a fuel pellet, and the effect depends first of all on the general assembly conditions, while the local subchannel parameters have only a slight influence on the pellet heat source distribution. (author)

  3. Calculation of spatial weighting functions for ex-core detectors of VVER-440 reactors by Monte Carlo method

    International Nuclear Information System (INIS)

    Berki, T.

    2003-01-01

    The signal of ex-core detectors depends not only on the total power of a reactor but also on the power distribution. The spatial weighting function establishes correspondence between the power distribution and the detector signal. The weighting function is independent of the power distribution. The weighting function is used for detector-response analyses, for example in the case of rod-drop experiments. (1) The paper describes the calculation and analysis of the weighting function of a VVER-440. The three-dimensional Monte Carlo code MCNP is used for the evaluation. Results from forward and adjoint calculations are compared. The effect of the change in the concentration of boric acid is also investigated. The evaluation of the spatial weighting function is a fixed-source neutron transport problem, which can be solved much faster by adjoint calculation, however forward calculations provide more detailed results. It is showed that the effect of boric acid upon the weighting function is negligible. (author)

  4. 3D analysis of the reactivity insertion accident in VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Abdullayev, A. M.; Zhukov, A. I.; Slyeptsov, S. M. [NSC Kharkov Inst. for Physics and Technology, 1, Akademicheskaya Str., Kharkov 61108 (Ukraine)

    2012-07-01

    Fuel parameters such as peak enthalpy and temperature during rod ejection accident are calculated. The calculations are performed by 3D neutron kinetics code NESTLE and 3D thermal-hydraulic code VIPRE-W. Both hot zero power and hot full power cases were studied for an equilibrium cycle with Westinghouse hex fuel in VVER-1000. It is shown that the use of 3D methodology can significantly increase safety margins for current criteria and met future criteria. (authors)

  5. Completion of the VVER 440/213 NPP Mochovce incorporation enhanced safety features

    International Nuclear Information System (INIS)

    Charbonneau, S.; Eckert, G.

    1996-01-01

    The cooperation between the western countries and the countries of ex-eastern block in the field of nuclear safety is recent and still limited. The main reasons for this situation are limited or non existent capabilities of these countries for financing as well as non acceptable legal conditions concerning the third party nuclear liability in this part of Europe. Nevertheless, Framatome and Siemens associated in the consortium named EUCOM, have signed in April 1996 the contract of about 100 million US dollars with Slovak electricity company (SLOVENSKE ELEKTRARNE-SE) for upgrading the Units 1 and 2 of Mochovce Nuclear Power Plant according to the western safety standards. This is the first important project involving west-european companies in the modernisation of Russian type of pressurized water reactor (VVER 440/213). The consortium will cooperate with other partners involved in the project: Slovak, Czech and Russian. The financing of the project will be provided mainly form Slovak and Czech sources. The safety upgrading will be financed through French and German buyer credits. French company Electricite de France (EDF) will be the consultant for SE. The safety upgrading measures have been elaborated taking into account the recommendation of Vienna International Atomic Energy Agency (IAEA) and the evaluation of the safety realised by RISKAUDIT, the common organization of German and French safety authorities (GSR and IPSN). Hence all guaranties have been taken to fulfil the western safety criteria for Nuclear Power Plant Mochovce. (author)

  6. Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, Sandor; Lipcsei, Sandor [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research - MTA

    2017-09-15

    Our aim was to develop a method based on noise diagnostics for the estimation of the moderator temperature coefficient of reactivity (MTC) for the Paks VVER-440 units in normal operation. The method requires determining core average neutron flux and temperature fluctuations. The circulation period of the primary coolant, transfer properties of the steam generators, as well as the source and the propagation of the temperature perturbations and the proportions of the perturbation components were investigated in order to estimate the feedback caused by the circulation of the primary coolant. Finally, after developing the new MTC estimator, determining its frequency range and optimal parameters, trends were produced based on an overall evaluation of measurements made with standard instrumentation during a one-year-long period at Paks NPP.

  7. Experimental support of WWER-440 fuel reliability and serviceability at high burnup

    Energy Technology Data Exchange (ETDEWEB)

    Smirnov, A; Ivanov, V; Pnyushkin, A [Nauchno-Issledovatel` skij Inst. Atomnykh Reaktorov, Dimitrovgrad (Russian Federation); Tzibulya, V [AO Mashinostroitelnij Zavod Electrostal (Russian Federation); Kolosovsky, V; Bibilashvili, Yu [Vsesoyuznyj Nauchno-Issledovatel` skij Inst. Neorganicheskikh Materialov, Moscow (Russian Federation); Dubrovin, K [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    Results from post-reactor examination of two WWER-440 fuel assemblies spent at the Kola NPP Unit 3 during 4 and 5 fuel cycles are presented. The fuel assembly states and their serviceability allowance are estimated experimentally at the RIAR hot laboratory and studied by non-destructive and destructive methods. The following parameters are examined: fuel assembly overall dimensions change; fuel element diameter change; fuel element cladding corrosion and hydriding; fuel element cladding mechanical properties; fission gas release from fuel and gas pressure; fuel macro- and microstructure. it has been found that the maximum fuel burnup of fuel assemblies No. 1 and No.2 achieved is 58.3 and 64.0 MWd/kg, respectively. The mechanical fuel pellets-cladding interaction has been observed at the average fuel burnup above 45 MWd/kg that occurred with increasing the local cladding diameter at the areas of pellets end arrangement (bamboo stick). The gas release linearly increases at the range 2.7% per 10 MWd/kg within burnup of 43-60 MWd/kg. 9 figs., 3 refs.

  8. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool

    International Nuclear Information System (INIS)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L.

    2007-01-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  9. Development of a VVER-1000 core loading pattern optimization program based on perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2012-01-01

    Highlights: ► We use perturbation theory to find an optimum fuel loading pattern in a VVER-1000. ► We provide a software for in-core fuel management optimization. ► We consider two objectives for our method (perturbation theory). ► We show that perturbation theory method is very fast and accurate for optimization. - Abstract: In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. Two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain the fuel integrity. Because of the numerous possible patterns of fuel assemblies in the reactor core, finding the best configuration is so important and challenging. Different techniques for optimization of fuel loading pattern in the reactor core have been introduced by now. In this study, a software is programmed in C language to find an order of the fuel loading pattern of a VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process launches by considering an initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. The results on a typical VVER-1000 reactor reveal that the method could reach to a pattern with an allowed radial power peaking factor and increases the cycle length 1.1 days, as well.

  10. Development of anti-debris filter for WWER-440 working fuel assembly

    International Nuclear Information System (INIS)

    Kolosovsky, V.; Aksyonov, P.; Kukushkin, Y.; Molchanov, V.; Kolobaev, A.

    2006-01-01

    Mechanical damaging of the fuel rod claddings caused by debris is one of the main reasons for fuel assembly failures. The paper focuses on the program and results of experimental and design activities carried out by Russian organizations relating to the development and investigation of operational characteristics of anti-debris filters for WWER-440 working fuel assemblies. Lead working fuel assemblies equipped with anti-debris filters have been loaded in the core of Kola-2 NPP. The results obtained can be used for making the decision concerning the application of anti-debris filter for WWER-440 working fuel assemblies with the purpose of enhancing their debris-resistance properties. (authors)

  11. Research on loading pattern optimization for VVER reactor

    International Nuclear Information System (INIS)

    Tran Viet Phu; Nguyen Thi Mai Huong; Nguyen Huu Tiep; Ta Duy Long; Tran Vinh Thanh; Tran Hoai Nam

    2017-01-01

    A study on fuel loading pattern optimization of a VVER reactor was performed. In this study, a core physics simulator was developed based on a multi-group diffusion theory for the use in the problem of fuel loading optimization of VVER reactors. The core simulator could handle the triangular meshes of the core and the computational speed is fast. Verification of the core simulator was confirmed against a benchmark problem of a VVER-1000 reactor. Several optimization methods such as DS, SA, TS and a combination of them were investigated and implemented in coupling with the core simulator. Calculations was performed for optimizing the fuel loading pattern of the core using these methods based on a benchmark core model in comparison with the reference core. Comparison among these methods have shown that a combination of SA+TS is the most effective for the problem of fuel loading pattern optimization. Advanced methods are being researched continuously. (author)

  12. Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Savander, V. I.; Shumskiy, B. E., E-mail: borisshumskij@yandex.ru [National Research Nuclear University MEPhI (Russian Federation); Pinegin, A. A. [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

  13. Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1996-12-01

    The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

  14. A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

    International Nuclear Information System (INIS)

    Bezrukov, Y.A.; Schekoldin, V.I.; Zaitsev, S.I.; Churkin, A.N.; Lisenkov, E.A.

    2016-01-01

    The paper covers a brief review of reflooding studies performed in different countries and the relevant tests performed in OKB GIDROPRESS (Russia) are discussed in more detail. The OKB GIDROPRESS test facility simulates the primary circuit of the VVER-440 reactor, with a full-scale fuel assembly (FA) mockup as the core simulator. The VVER core reflooding was studied in a FA mockup containing 126 fuel rod simulators with axial power peaking. The experiments were performed for two types of flooding. The first type is top flooding of the empty (steamed) FA mockup. The second type is bottom flooding of the FA mockup with level of boiling water. The test parameters are as follows: the range of the supplied power to the bundle is from 40 to 320 kW, the cooling water flow rate is from 0.04 to 1.1 kg/s, the maximum temperature of the fuel rod simulator is 800 C. degrees and the linear heat flux is from 0.1 to 1.0 kW/m. The test results were used for computer code validation, especially for the TRAP package code. The experiments show that as the transverse dimension of the FA mockup increases, the flow choking of the water supplied from the top by the steam flow significantly decreases

  15. Characteristics and properties of cladding tubes for VVER-1000 higher Uranium content fuel rods

    International Nuclear Information System (INIS)

    Peregud, M.; Markelov, A.; Novikov, V.; Gusev, A.; Konkov, V.; Pimenov, Y.; Agapitov, V.; Shtutsa, M.

    2009-01-01

    To improve the fuel cycle economics and to further increase the VVER fuel usability the work programme is under way to design novel improved fuel, fuel rods and fuel assemblies. Longer FA operation time that is needed to increase the fuel burnup and the related design developments of novel fuel assemblies resulted not only in changing types and sizes of Zirconium items and fuel assembly components but also altered the requirements placed on their technical characteristics. To use fuel rods having a larger charge of fuel, to improve their behaviour in LOCA, to reduce fuel rod damage ability during assembling the work was carried out to perfect the characteristics of both the cladding (reduced wall thickness and more rigid tolerances for geometry) and its material. To meet the more rigid requirements for the geometry dimensions of cladding tubes an improved process flow sheet has been designed and employed for their fabrication and also the finishing treatment of tube surfaces has been improved. The higher and stable properties of the cladding materials were managed through using the special purity in terms of Hafnium Zirconium (not higher than 100 ppm Hf) as a base of the E110 alloy and maintaining within the valid specifications for the alloy the optimized contents of Oxygen and Iron at the levels of (600 - 990) ppm and (250 - 700) ppm, respectively. The work was under way in 2004 - 2008 years; during this period the technology and materials science solutions were mastered that were phased-in introduced into the production of the cladding tubes for the fuels loaded into the of the Kalinin NPP Unit 1

  16. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

    Matal, O.; Sousek, P.; Simo, T.; Lehota, M.; Lipka, J.; Slugen, V.

    2000-01-01

    Defects in feed water distributing system due to corrosion-erosion effects have been observed at many VVER 440 steam generators (SG). Therefore analysis of defects origin and consequently design development and testing of a new feed water distributing system were performed. System tests in-situ supported by calculations and comparison of measured and calculated data were focused on demonstration of long term reliable operation, definition of water flow and water chemical characteristics at the SG secondary side and their measurements and study of dynamic characteristics needed for the innovated feed water distributing system seismic features approval. The innovated feed water distributing system was installed in the SGs of two VVER units already. (author)

  17. Time versus frequency domain calculation of the main building complex of the VVER 440/213 NPP PAKS

    International Nuclear Information System (INIS)

    Katona, T.; Ratkai, S.; Halbritter, A.; Krutzik, N.J.; Schuetz, W.

    1995-01-01

    Various dynamic analyses were conducted for the main building complex of the VVER 440/213 PAKS in order to determine the dynamic response and assess the aseismic capacity of this nuclear power plant. Different types of mathematical models for idealizing the soil and the building structures were used. The main goal of the study presented here was to demonstrate the effects of different procedures for consideration of soil-structure interaction on the dynamic response of the structures mentioned above. The analyses were based on appropriate mathematical models of the coupled vibration structures (reactor building, turbine hall, intermediate building structures) and the layered soil. On the basis of this study, it can be concluded that substructure models using frequency-independent impedances and cut-off of modal damping usually provide conservative results. Complex models which allow the soil-soil and the structure or by frequency-dependent impedances) provide more accurate results. The latter approach results in more efficient designs which are not only safe but also economical. (author). 7 refs., 15 figs

  18. Recent results of three-dimensional CFD simulations of coolant mixing in VVER-440/213 reactor pressure vessel

    International Nuclear Information System (INIS)

    Kiss, B.; Boros, I.; Aszodi, A.

    2008-01-01

    The Budapest University of Technology and Economics, Institute of Nuclear Techniques has been working since 2001 on the three-dimensional CFD model of the reactor pressure vessel of the VVER-440 type reactor. During this time period - due to the development of the available computational capacity - a very complex and detailed model of the RPV has been developed. The aim of the construction of the new model is to describe further internal structures of the RPV (e.g. correct modeling of brake tubes, or internals in the upper mixing chamber) and to perform an extensive sensitivity analysis on the different modeling and calculation parameters (e.g. porous region models vs. detailed modeling, or n different turbulence models). The new model can be applied for steady state calculation during normal operational condition and for different transient analyses as well. One interesting application is the participation in a planned benchmark exercise on the start-up of the sixth main coolant pump, which is aimed to compare the capabilities of mixing models of one-dimensional system codes with the results of CFD simulation. (authors)

  19. Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Loetsch, T.; Khalimonchuk, V.; Kuchin, A.

    2009-01-01

    In the framework of a project supported by the German BMU the code DYN3D should be further validated and verified. During the work a lack of a benchmark on core burnup calculations for VVER-1000 reactors was noticed. Such a benchmark is useful for validating and verifying the whole package of codes and data libraries for reactor physics calculations including fuel assembly modelling, fuel assembly data preparation, few group data parametrisation and reactor core modelling. The benchmark proposed specifies the core loading patterns of burnup cycles for a VVER-1000 reactor core as well as a set of operational data such as load follow, boron concentration in the coolant, cycle length, measured reactivity coefficients and power density distributions. The reactor core characteristics chosen for comparison and the first results obtained during the work with the reactor physics code DYN3D are presented. This work presents the continuation of efforts of the projects mentioned to estimate the accuracy of calculated characteristics of VVER-1000 reactor cores. In addition, the codes used for reactor physics calculations of safety related reactor core characteristics should be validated and verified for the cases in which they are to be used. This is significant for safety related evaluations and assessments carried out in the framework of licensing and supervision procedures in the field of reactor physics. (authors)

  20. Evaluation of efficiency of axial profiling in WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Ananjev, Yu. A.; Kurakin, K. Yu.; Artemov, V.G.; Ivanov, A.S.

    2005-01-01

    The present report deals with consideration of fuel enrichment axial profiling in WWER-440 assemblies. The study is performed on improving the effectiveness of fuel utilization using the example of implementing the axial profiling in the assemblies of the second generation. For simulation of fuel loadings the computer code package SAPFIR 9 5 and RC is used that allows for correct consideration of specific features of assemblies design changes. The methodical approach to assessment of effectiveness of implementing the axial profiling is considered with the use of capabilities of the mentioned code package. In conclusion the recommendations are given on using the fuel enrichment axial profiling in WWER-440 assemblies (Authors)

  1. The evolution of the fuel cycle in the Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Bajgl, J [Jaderna Elektrarna, Dukovany (Czech Republic)

    1994-12-31

    The ten-year operational experience of four WWER-440 units in Dukovany NPP with a total number of 35 cycles is outlined. The strategy of fuel reloading has been changed from out-in schemes to low-leakage patterns. The linear pin power limitation will be introduced. The main physical limits and conditions for the NPP operation are listed. The main goal is to go to a full 4-year fuel cycle in which the burnup will be about 40 Mwd/kg U. 6 tabs.

  2. State of the VVER-1000 spent U-Gd fuel rods based on the results of post-irradiation examinations

    International Nuclear Information System (INIS)

    Shevlyakov, G.; Zvir, E.; Strozhuk, A.; Polenok, V.; Sidorenko, O.; Volkova, I.; Nikitin, O.

    2015-01-01

    The present paper is devoted to post-irradiation examinations (PIE) of U-Gd fuel rods with different geometry of the fuel pellets irradiated as part of the VVER-1000 fuel assembly. As evidenced by their PIE data, they did not exhaust their service life based on the main parameters (geometrical dimensions, corrosion state, and release of fission product gases). (author)

  3. Evaluation of the applicability of cladding deformation model in RELAP5/MOD3.2 code for VVER-1000 fuel

    International Nuclear Information System (INIS)

    Vorob'ev, Yu.; Zhabin, O.

    2015-01-01

    Applicability of cladding deformation model in RELAP5/MOD3.2 code is analyzed for VVER-1000 fuel cladding from Zr+1%Nb alloy. Experimental data and calculation model of fuel assembly channel of the core are used for this purpose. The model applicability is tested for the cladding temperature range from 600 to 1200 deg C and pressure range from 1 to 12 MPa. Evaluation results demonstrate limited applicability of built-in RELAP5/MOD3.2 cladding deformation model to the estimation of Zr+1%Nb cladding rupture conditions. The limitations found shall be considered in application of RELAP5/MOD3.2 cladding deformation model in the design-basis accident analysis of VVER reactors

  4. The Plinius/Colima CA-U3 test on fission-product aerosol release over a VVER-type corium pool; L'essai Plinius/Colima CA-U3 sur le relachement des aerosols de produits de fission au-dessus d'un bain de corium de type VVER

    Energy Technology Data Exchange (ETDEWEB)

    Journeau, Ch.; Piluso, P.; Correggio, P.; Godin-Jacqmin, L

    2007-07-01

    In a hypothetical case of severe accident in a PWR type VVER-440, a complex corium pool could be formed and fission products could be released. In order to study aerosols release in terms of mechanisms, kinetics, nature or quantity, and to better precise the source term of VVER-440, a series of experiments have been performed in the Colima facility and the test Colima CA-U3 has been successfully performed thanks to technological modifications to melt a prototypical corium at 2760 C degrees. Specific instrumentation has allowed us to follow the evolution of the corium melt and the release, transport and deposition of the fission products. The main conclusions are: -) there is a large release of Cr, Te, Sr, Pr and Rh (>95%w), -) there is a significant release of Fe (50%w), -) there is a small release of Ba, Ce, La, Nb, Nd and Y (<90%w), -) there is a very small release of U in proportion (<5%w) but it is one of the major released species in mass, and -) there is no release of Zr. The Colima experimental results are consistent with previous experiments on irradiated fuels except for Ba, Fe and U releases. (A.C.)

  5. Control of selected VVER components life time SKODA JS a.s. experience

    International Nuclear Information System (INIS)

    Zdebor, J.; Pribulla, E.

    2005-01-01

    Experience from the operation of nuclear power plants with type VVER reactors has shown that the life time management of a number of nuclear reactor components is technically as well as with respect to safety substantiated even at the time which exceeds their originally designed life time. To accept such solution it was necessary to develop and implement a number of programs based on which it was possible to evaluate the actual condition of monitored equipment. It is a condition allowing to adopt solution concerning the possibility of their further operation. Experience from the manufacture of 21 sets of reactor equipment for VVER 440 and 3 sets VVER 1000 has been gathered in SKODA JS a.s. which have been completed by experience from service activities performed at operated nuclear power plants. Analyses of manufacturing data and in-service inspection results completed by a lot of laboratory tests have become the basis for the development of selected VVER components life time management programs. The paper focuses on the life time management of those VVER components which SKODA JS a.s. has most experience with. (authors)

  6. Trends and results in In-Core management for the Kozloduy NPP WWER-440 reactors

    International Nuclear Information System (INIS)

    Haralampieva, Tz.; Antov, A.; Georgieva, N.; Spasova, V.

    2001-01-01

    The paper presents the experience gained during the design and operation of the last fuel cycles of the four WWER-440/V-230 units at Kozloduy NPP. High efficiency and economy of the fuel utilization requires very precise procedures for fuel in-core management, including calculations and analyses for reloading scheme design, compared with results from operational measurements and fuel cycle efficiency. The paper describes the main stages of implementation of advanced fuel assemblies in the Kozloduy NPP WWER-440 reactors. New advanced fuel has been implemented after the completion of comprehensive neutron-physical, thermal-hydraulic and thermal-mechanical analyses by using advanced computer codes. As a general task of the fuel cycle improvements it is pointed the increasing of the final fuel burnup and decreasing of the number of spent fuel assemblies. Series of calculations and analyses, related to the introducing of the advanced fuel assemblies and improvement of the fuel cycle characteristics have been carried out to guarantee the safe operation and fuel reliability

  7. DRAGON analysis of MOX fueled VVER cell benchmarks

    International Nuclear Information System (INIS)

    Marleau, G.; Foissac, F.

    2002-01-01

    The computational unit-cell benchmarks problems for LEU and MOX fueled VVER-1000 ('water-water energetic reactor') have been analyzed using the code DRAGON with ENDF/B-V and ENDF/B-VI based WIMS-AECL cross section libraries. The results obtained were compared with those generated using the SAS2H module of the SCALE-4.3 computational code system and with the code HELIOS. Good agreements between DRAGON and HELIOS were obtained when the ENDF/B-VI based library was considered while the ENDF/B-V DRAGON results were generally closer to those obtained using SAS2H. This study was useful for the verification of the DRAGON code and confirms that HELIOS and DRAGON have a similar behavior when compatible cross sections library are used. (author)

  8. System analysis of nuclear safety of VVER reactor with MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, A.D.; Zharkov, V.P.; Suslov, I.R. [Russia, Moscow Malaya Krasnoselskaya St. (Russian Federation); Boyarinov, V.F.; Kevrolev, V.V.; Tchibinyaev, A.V.; Tsibulskiy, V.F. [RRC KI, Russia, Moscow (Russian Federation); Kochurov, B.P. [ITEP, Russia, Moscow (Russian Federation); Giovanni, B. [NFPSC, FRAMATOME (France)

    2005-07-01

    The report presents a short summary of the results achieved in the ISTC (International Science and Technology Center) project 'System analysis of nuclear safety of VVER reactor with MOX fuel' (April 2005). The studies within the project are of a systematic character and include the solutions of 15 tasks. The report gives an overview of the major blocks of these tasks: neutron transport equation solution; calculations of isotopic vectors, analysis of the impact of uncertainties on predicted reactor functionals. The calculation methods, the verification results and the corresponding codes are briefly described. (authors)

  9. CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment

    Energy Technology Data Exchange (ETDEWEB)

    Heitsch, Matthias, E-mail: Matthias.Heitsch@ec.europa.e [Institute for Energy, Joint Research Centre, PO Box 2, 1755 ZG Petten (Netherlands); Huhtanen, Risto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Techy, Zsolt [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Fry, Chris [Serco, Winfrith Technology Centre, Dorchester, Dorset DT2 8DH (United Kingdom); Kostka, Pal [VEIKI Institute for Electric Power Research Co., PO Box 80, H-1251 Budapest (Hungary); Niemi, Jarto [VTT Technical Research Centre of Finland, PO Box 1000, FI-02044 VTT (Finland); Schramm, Berthold [Gesellschaft fuer Anlagen- und Reaktorsicherheit, GRS mbH, Schwertnergasse 1, 50667 Koeln (Germany)

    2010-02-15

    In the PHARE project 'Hydrogen Management for the VVER440/213' (HU2002/000-632-04-01), CFD (Computational Fluid Dynamics) calculations using GASFLOW, FLUENT and CFX were performed for the Paks NPP (Nuclear Power Plant), modelling a defined severe accident scenario which involves the release of hydrogen. The purpose of this work is to demonstrate that CFD codes can be used to model gas movement inside a containment during a severe accident. With growing experience in performing such analyses, the results encourage the use of CFD in assessing the risk of losing containment integrity as a result of hydrogen deflagrations. As an effective mitigation measure in such a situation, the implementation of catalytic recombiners is planned in the Paks NPP. In order to support these plans both unmitigated and recombiner-mitigated simulations were performed. These are described and selected results are compared. The codes CFX and FLUENT needed refinement to their models of wall and bulk steam condensation in order to be able to fully simulate the severe accident under consideration. Several CFD codes were used in parallel to model the same accident scenario in order to reduce uncertainties in the results. Previously it was considered impractical to use CFD codes to simulate a full containment subject to a severe accident extending over many hours. This was because of the expected prohibitive computing times and missing physical capabilities of the codes. This work demonstrates that, because of developments in the capabilities of CFD codes and improvements in computer power, these calculations have now become feasible.

  10. Application of the SCANAIR code for VVER RIA conditions - Boron dilution accident

    International Nuclear Information System (INIS)

    Arffman, A.; Cazalis, B.

    2010-01-01

    This paper consists of two parts. In part A, RIA pulse tests conducted at the Russian BIGR reactor are being analysed at IRSN with SCANAIR V6 fuel performance code as a part of the code validation for VVER fuel. Recently a new version of the SCANAIR code was made available to VTT Technical Research Centre of Finland, and part B of the paper covers the introduction of the code version at VTT by a calculation of a hypothetical boron dilution accident in a VVER-440 power reactor. Concerning part A, it appears that the SCANAIR V6 version, including a BIGR/NSRR heat transfer model, validated by Japanese NSRR experiments, and a Norton viscoplastic clad mechanical behaviour, is able to simulate the rod thermal behaviour in BIGR tests. Concerning the clad mechanics, it has been seen that a pellet swelling model is able to simulate the average rod deformation. Nonetheless, the current clad creep model associated with the free volume equilibrium assumption is not suited to predict the maximum clad deformation and the possible post DNB rod failure because they do not simulate local balloons. Furthermore, it has been shown that the clad deformation is strongly dependent on transient gas transfer. Concerning part B, a boron dilution accident previously calculated with SCANAIR V2 was recalculated with SCANAIR V6. A limited amount of result parameters were compared with the results of VTT's neutronics code TRAB. Divergence problems encountered previously when reaching the DNB limit were not present anymore. Fuel and cladding temperatures produced by SCANAIR were in good agreement with those calculated with TRAB

  11. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    International Nuclear Information System (INIS)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K.

    2008-01-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of FAs of state-of-the-art designs. This technique is employed in the VVER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

  12. The Procedure for Determination of Special Margin Factors to Account for a Bow of the VVER-1000 Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tsyganov, Sergey V.; Marin, Stanislav V.; Shishkov, Lev K. [Russian Research Center ' Kurchatov Institute' , 1., Kurchatov sq., 123182 Moscow (Russian Federation)

    2008-07-01

    Starting from 1980's, the problem of bow of the VVER-1000 reactor FAs and the effect of that on the operational safety is being discussed. At the initial period of time, the extension of time for dropping control rods of the control and protection system associated with this bow posed the highest threat. Later on, new more rigid structures were developed for FAs that eliminated the problems of control rods. However, bow of the VVER-1000 reactor FAs is observed up to now. The scale of this bow reduced significantly but it still effects safety. Even a minor bow available may lead to the noticeable increase of power of individual fuel pins associated with the local variation of the coolant amount. This effect must be taken into account on designing fuel loadings to eliminate the exceeding of set limitations. The introduction of additional special margins is the standard method for taking this effect into account. The present paper describes the conservative technique for the assessment of additional margins for bow of FAs of state-of-the-art designs. This technique is employed in the VVER-1000 reactor designing. The chosen conservatism degree is discussed as well as the method for its assurance and acceptable ways for its slackening. The example of the margin evaluation for the up-to-date fuel loading is given. (authors)

  13. Analysis of noncondensable effect during small break transient in VVER-440 geometry with CATHARE V1.3L. Preliminary results

    International Nuclear Information System (INIS)

    Sarrette, C.

    1996-11-01

    The report presents a study of the transport and dissolution-release of non-condensable gas into the fluid of the primary loop for the VVER-440 geometry. The analysis has been done using a new model developed for the CATHARE thermal hydraulic code. Results are presented, obtained from calculations of small break loss-of-coolant (SBLOCA) accidents for the Loviisa nuclear power plant (NPP) geometry. The influence of nitrogen dissolved in the water of the accumulators of the emergency core coolant system (ECCS) on natural circulation is discussed. Possibilities of formation of nitrogen bubbles in the main vessels upper plenum, top of the downcomer, steam generators collectors, and upper structures of RCP's are investigated. First results show that there is potentiality for interruption, mainly due to the presence of nitrogen in the top of the downcomer and the upper parts of the RCP's. These preliminary results should be confirmed by carrying out calculations now prematurely stopped for numerical reasons. (8 refs.)

  14. Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Szabados, L.; Ezsoel, G.; Perneczky, L. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

    2001-07-01

    Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

  15. Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

    International Nuclear Information System (INIS)

    Szabados, L.; Ezsoel, G.; Perneczky, L.

    2001-01-01

    Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

  16. The prediction problems of VVER fuel element cladding failure theory

    International Nuclear Information System (INIS)

    Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D.

    2016-01-01

    Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element (FE) cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly (FA) loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.

  17. Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Horak, W.C.

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

  18. Probabilistic assessments of fuel performance

    International Nuclear Information System (INIS)

    Kelppe, S.; Ranta-Puska, K.

    1998-01-01

    The probabilistic Monte Carlo Method, coupled with quasi-random sampling, is applied for the fuel performance analyses. By using known distributions of fabrication parameters and real power histories with their randomly selected combinations, and by making a large number of ENIGMA code calculations, one expects to find out the state of the whole reactor fuel. Good statistics requires thousands of runs. A sample case representing VVER-440 reactor fuel indicates relatively low fuel temperatures and mainly athermal fission gas release if any. The rod internal pressure remains typically below 2.5 MPa, which leaves a large margin to the system pressure of 12 MPa Gap conductance, an essential parameter in the accident evaluations, shows no decrease from its start-of-life value. (orig.)

  19. Steam oxidation of Zr 1% Nb clads of VVER fuels in high temperature

    International Nuclear Information System (INIS)

    Solyanyj, V.I.; Bibilashvili, Yu.K.; Dranenko, V.V.; Levin, A.Ya.; Izrajlevskij, L.B.; Morozov, A.M.

    1984-01-01

    In a wide range of accident conditions processes of clad corrosion effected by steam are rather intensive and in many respects influence the safety of NPP and the after-accident dismantling of a reactor core. This paper discusses the results of comprehensive studies into corrosion behaviour of Zr 1%Nb clads of VVER-type fuels at high temperatures. These studies are a continuation of previous work and the base for the design modelling of corrosion processes

  20. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    The report provides a summary of the design of the canister for final disposal of nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  1. Reloading pattern optimization of VVER-1000 reactors in transient cycles using genetic algorithm

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • The genetic algorithm (GA) and the innovative weighting factors method were used. • The coupling of WIMSD5-B and CITATION-LDI2 neutronic codes with the thermohydraulic WERL code was employed. • Optimization of reloading patterns was carried out in two states. • First an arrangement with satisfactory excess reactivity and the flattest power distribution was searched. • Second, it is tried to obtain an arrangement with satisfactory safety threshold and the maximum K_e_f_f. - Abstract: The present paper proposes application of the genetic algorithm (GA) and the innovative weighting factor method to optimize the reloading pattern of Bushehr VVER-1000 reactor in the second cycle. To estimate the composition of fuel assemblies remaining from the first cycle and precisely calculate the objective parameters of each reloading pattern in the second cycle, coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic section was employed. Optimization of the reloading patterns was carried out in two states. To meet the mentioned objective, with application of the weighting factor method in the first state, the type and quantity of the loadable fresh assemblies were determined to enable the reactor core to maintain the core criticality over the entire cycle length. Afterwards, the genetic algorithm was used to optimize the reloading pattern of the reactor to obtain an arrangement with flat radial power distribution. In the second state, the optimization algorithm was free to select the type and number of fresh fuel assemblies to be able to search for an arrangement with the maximum effective multiplication factor and the safe power peaking factor. In addition, in order to ensure the safety and desirability of the proposed patterns in both states, a time-dependent examination of the thermo-neutronic behavior of the reactor core was carried out during the second cycle. With consideration of the new

  2. Comparison of problems and experience of core operation with distorted fuel element assemblies in VVER-1000 and PWR reactors

    International Nuclear Information System (INIS)

    Afanas'ev, A.

    1999-01-01

    The main reactors leading to distortion of fuel element assemblies during reactor operation were studied. A series of actions which compensate this effect was proposed. Criteria of operation limitation in VVER-1000 and PWR reactors are described

  3. Mechanical fragmentation of nuclear reactor fuel assemblies by the double cutting method

    International Nuclear Information System (INIS)

    Voitsekhovskii, B.V.; Istomin, V.L.; Mitrofanov, V.V.

    1995-01-01

    A method is described for cutting a spent fuel assembly with straight shears into pieces of a prescribed size. The method does not require separation of the casing and the lattices. The double cutting method is briefly described, and experiments designed for cutting BN-350 and VVER-440 fuel assemblies are outlined. The testing showed that the cutting method was suitable for mechanical polarization of fuel assemblies. The investigations led to the development of turnkey industrial equipment for cutting spent fuel assemblies of different geometries with a maximum size up to 170 mm. 6 refs., 8 figs., 1 tab

  4. Development of a new WWER-440 fuel design

    International Nuclear Information System (INIS)

    Coucil, D.; Totev, T.

    1998-01-01

    In March 1996 British Nuclear Fuel Limited signed a contract with Imatran Voima and Paks Nuclear Power Plant to design, develop, license and supply 5 Lead Test Assemblies to the WWER-440 reactor at Loviisa in Finland. In June 1998 the manufacture of these 5 assemblies (4 fixed assemblies and 1 follower assembly) was completed. The fuel is expected to be loaded into Loviisa Unit 2 reactor during the shutdown scheduled for September of this year. (Authors)

  5. Economical aspects of multiple plutonium and uranium recycling in VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, P.N.; Bobrov, E.A.; Dudnikov, A.A.; Teplov, P.S. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2016-09-15

    The basic strategy of Russian Nuclear Energy development is the formation of the closed fuel cycle based on fast breeder and thermal reactors, as well as the solution of problems of spent nuclear fuel accumulation and availability of resources. Three options of multiple Pu and U recycling in VVER reactors are considered in this work. Comparison of MOX and REMIX fuel recycling approaches for the closed fuel cycle involving thermal reactors is presented. REMIX fuel is supposed to be fabricated from non-separated mixture of uranium and plutonium obtained in spent fuel reprocessing with further makeup by enriched U. These options make it possible to recycle several times the total amount of Pu and U obtained from spent fuel. The main difference is the full or partial fuel loading of the core by assemblies with recycled Pu. The third option presents the concept of heterogeneous arrangement of fuel pins made of enriched uranium and MOX in one fuel assembly. It should be noted that fabrication of all fuel assemblies with Pu requires the use of expensive manufacturing technology. These three options of core loading can be balanced with respect to maximum Pu and U involvement in the fuel cycle. Various physical and economical aspects of Pu and U multiple recycling for selected options are considered in this work.

  6. Application of Integral Ex-Core and Differential In-Core Neutron Measurements for Adjustment of Fuel Burn-Up Distributions in VVER-1000

    Science.gov (United States)

    Borodkin, Pavel G.; Borodkin, Gennady I.; Khrennikov, Nikolay N.

    2010-10-01

    The paper deals with calculational and semi-analytical evaluations of VVER-1000 reactor core neutron source distributions and their influence on measurements and calculations of the integral through-vessel neutron leakage. Time-integrated neutron source distributions used for DORT calculations were prepared by two different approaches based on a) calculated fuel burn-up (standard routine procedure) and b) in-core measurements by means of SPD & TC (new approach). Taking into account that fuel burn-up distributions in operating VVER may be evaluated now by analytical methods (calculations) only it is needed to develop new approaches for testing and correction of calculational evaluations. Results presented in this paper allow to consider a reverse task of alternative estimation of fuel burn-up distributions. The approach proposed is based on adjustment (fitting) of time-integrated neutron source distributions, and hence fuel burn-up patterns in some part of reactor core, on the base of ex-core neutron leakage measurement, neutron-physical calculation and in-core SPD & TC measurement data.

  7. Improvement of operation efficiency for WWER-440 and WWER-1000 for TRIGON fuel assembly design features

    Energy Technology Data Exchange (ETDEWEB)

    Silberstein, A [European WWER Fuels GmbH, Lyon (France)

    1994-12-31

    TRIGON 440 and TRIGON 1000 fuel assemblies and their assembly matching counterparts are described. Their role in increasing the efficiency of WWER reactors is stressed. Special attention is paid to their design features as well as calibrated means of predicting behaviour under irradiation from light water reactor core operation. They reduce the fuel cycle cost as a result of the reduced need for natural uranium which have to be enriched and of the smaller number of fuel assemblies which have to be fabricated, stored or reprocessed. The improved control assemblies bring comfort to the plant operator due to intrinsic progress in safety with respect to accidental situation, trouble-free behaviour and long time utilization in the reactor. 14 figs.

  8. Results of operation of VVER-1000 FAs manufactured at PJSC NCCP

    International Nuclear Information System (INIS)

    Davidov, D.; Brovkin, O.; Bezborodov, Y.

    2015-01-01

    Fuel Assemblies manufactured at PJSC NCCP are in operation at 27 VVER-1000 power units at 11 NPPs in Russia, Ukraine, Bulgaria, China, Iran and India. Basic results of operation of PJSC NCCP VVER-1000 FAs during 2007-2014 are presented. The operation results confirm the design characteristics of fuel, i.e.: average fuel burnup up to 55 MW*day/kgU in FAs; safe and reliable FA operation, with low leaking rate (in the order of 10-6). The achieved operation characteristics of TVSA and TVS-2M Fuel Assemblies prove the quality, reliability and competitiveness of FAs manufactured at PJSC NCCP

  9. Calculation of the fuel composition and the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor during the initial startup and the first cycle using the WIMSD5-B, CITATION-LDI2 and WERL codes

    International Nuclear Information System (INIS)

    Rahmani, Yashar; Pazirandeh, Ali; Ghofrani, Mohammad B.; Sadighi, Mostafa

    2013-01-01

    Highlights: ► In this paper, the changes of the thermo-neutronic parameters of a VVER 1000 reactor were studied during the first cycle. ► The coupling of neutronic and thermo-hydraulic codes was utilized. ► A computational program (WERL code) was designed to calculate the temperature distribution of the reactor core. ► To estimate the concentration of the released gaseous fission products, the Weisman model was used. ► The results of this study enjoyed the desirable accuracy. - Abstract: In this paper, the concentrations of fission products and fuel isotopes as well as the changes of the thermo-neutronic parameters of the Bushehr’s VVER-1000 reactor were studied during the initial startup and the first cycle. In order to perform the time-dependent cell calculations and obtain the concentration of fuel elements, the WIMSD5-B code was used. Besides, by utilizing the CITATION-LDI2 code, the effective multiplication factor and the thermal power distribution of the reactor were calculated. A computer program (WERL code) was designed in order to perform accurate calculation of the temperature distribution of the reactor core. For this purpose, the Ross–Stoute, Weisman, and Lee–Kesler models were used for calculating of the gap conductance coefficient, fission gas release and gap pressure, respectively. The results demonstrated that in designing the startup process, in addition to the role considered for overcoming the power defects and in preparing the required conditions for performing the safety-assurance tests, the flattening of the reactor’s power must be taken into account. Comparison between the results of this modeling and the final safety analysis report of this reactor showed that the results presented in this paper are satisfactorily accurate

  10. Comparative of fuel cycle cost for light water nuclear power plants

    International Nuclear Information System (INIS)

    Kocic, A.; Dimitrijevic, Z.

    1978-01-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  11. The corrosion and corrosion mechanical properties evaluation for the LBB concept in VVERs

    Energy Technology Data Exchange (ETDEWEB)

    Ruscak, M.; Chvatal, P.; Karnik, D.

    1997-04-01

    One of the conditions required for Leak Before Break application is the verification that the influence of corrosion environment on the material of the component can be neglected. Both the general corrosion and/or the initiation and, growth of corrosion-mechanical cracks must not cause the degradation. The primary piping in the VVER nuclear power plant is made from austenitic steels (VVER 440) and low alloy steels protected with the austenitic cladding (VVER 1000). Inspection of the base metal and heterogeneous weldments from the VVER 440 showed that the crack growth rates are below 10 m/s if a low oxygen level is kept in the primary environment. No intergranular cracking was observed in low and high oxygen water after any type of testing, with constant or periodic loading. In the framework of the LBB assessment of the VVER 1000, the corrosion and corrosion mechanical properties were also evaluated. The corrosion and corrosion mechanical testing was oriented predominantly to three types of tests: stress corrosion cracking tests corrosion fatigue tests evaluation of the resistance against corrosion damage. In this paper, the methods used for these tests are described and the materials are compared from the point of view of response on static and periodic mechanical stress on the low alloyed steel 10GN2WA and weld metal exposed in the primary circuit environment. The slow strain rate tests and static loading of both C-rings and CT specimens were performed in order to assess the stress corrosion cracking characteristics. Cyclic loading of CT specimens was done to evaluate the kinetics of the crack growth under periodical loading. Results are shown to illustrate the approaches used. The data obtained were evaluated also from the point of view of comparison of the influence of different structure on the stress corrosion cracking appearance. The results obtained for the base metal and weld metal of the piping are presented here.

  12. The most extensive reconstruction of nuclear power plant with VVER 440/V230 reactor

    International Nuclear Information System (INIS)

    Ferenc, M.

    2000-01-01

    The nuclear power plant V-1 Bohunice consists of two VVER-440 units with V-230 reactors. Unit 1 was commissioned in 1978 and Unit 2 in 1980. Large experience and knowledge from the operation of previous units with V-230 reactors were incorporated into the V-1 design, which resulted in a higher level of safety and operational reliability of these units. The Siemens company which won an international bidding process developed these basic goals for the Gradual Upgrading into the so called Basic Engineering (BE). For the implementation of the Gradual Upgrading in line with the BE, Rekon consortium was established consisting of Siemens and VUJE. The implementation of the Gradual Upgrading is scheduled for the time period of 1996 - 2000. Siemens was responsible for the upgrading strategy - based on the approved results of the basic engineering phase and the PSAR, the engineering and realization of all I and C improvements, and also for the seismic upgrade. VUJE's responsibility covered the detailed engineering and implementation of mechanical, electrical and civil part of upgrading measures as well as overall organisation and evaluation of verification tests. The consortium awarded contracts for final planning and design, installation services and commissioning to other Slovakian subcontractors in order to ensure the largest possible local content. The gradual reconstruction of the V-1 Bohunice with V230 reactors represents a comprehensive reconstruction of safety-related systems and equipment. Following its completion, the units will be operated with a safety level accepted internationally. (author)

  13. Several perspectives on water-chemical cycles for nuclear power stations equipped with type VVER and RBMK reactors

    International Nuclear Information System (INIS)

    Mamet, A.P.; Mamet, V.A.; Pashevich, V.I.; Nazarenko, P.N.

    1982-01-01

    Water-chemical cycles for loops I and II of VVER reactors are discussed. These cycles are mixed ammonia-sodium with a variable concentration of boric acid and ammonia hydrazine with a pH factor of 9.1 +/- 0.1. New water-chemical cycles are considered for use in both existing and new nuclear power plants. Application of these new water-chemical cycles showed produce a significant improvement in operating conditions of nuclear power plants. Upon accumulation of sufficient operating experience with these cycles, it should be possible to raise the issue of revising applicable standard documentation

  14. Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Alekseev, Pavel N.; Balanin, Andrey L.; Dudnikov, Anatoly A.; Fomichenko, Petr A.; Nevinitsa, Vladimir A.; Frolov, Aleksey A.; Lubina, Anna S.; Sedov, Aleksey A.; Subbotin, Aleksey S.; Blandinsky, Viktor Yu. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)

    2015-09-15

    A subcritical molten salt reactor is proposed for minor actinides (separated from spent fuel VVER-1000 light water reactor) incineration and for {sup 233}U conversion from {sup 232}Th. Here the subcritical molten salt reactor with fuel composition of heavy nuclide fluorides in molten LiF - NaF - KF salt and with external neutron source, based on 1 GeV proton accelerator and molten salt cooled tungsten target is considered. The paper presents the results of parametrical analysis of equilibrium nuclide composition of molten salt reactor with minor actinides feed in dependence of core dimensions, average neutron flux and external neutron source intensity. Reactor design is defined; requirements to external neutron source are posed; heavy nuclides equilibrium and fuel cycle main parameters are calculated.

  15. Modeling of WWER-440 Fuel Pin Behavior at Extended Burn-up

    International Nuclear Information System (INIS)

    El-Koliel, M.S.; Abou-Zaid, A.A.; El-Kafas, A.A.

    2004-01-01

    Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWER's as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased to 60 to 70 Mwd/kg U. The change in the fuel radial power distribution as a function of fuel burn up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. In this paper, the radial burn-up and fissile products distributions of WWER-440 UO 2 fuel pin were evaluated using MCNP 4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted fission gas release calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin cell well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. a computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented

  16. In-core nuclear fuel management optimization of VVER1000 using perturbation theory

    International Nuclear Information System (INIS)

    Hosseini, Mohammad; Vosoughi, Naser

    2011-01-01

    In-core nuclear fuel management is one of the most important concerns in the design of nuclear reactors. The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor in order to extract the maximum energy, and keeping the local power peaking factor lower than a predetermined value to maintain fuel integrity. Because of the numerous possible patterns of the fuel assemblies in the reactor core, finding the best configuration is so important and complex. Different methods for optimization of fuel loading pattern in the core have been introduced so far. In this study, a software is programmed in C ⧣ language to find an order of the fuel loading pattern of the VVER-1000 reactor core using the perturbation theory. Our optimization method is based on minimizing the radial power peaking factor. The optimization process lunches by considering the initial loading pattern and the specifications of the fuel assemblies which are given as the input of the software. It shall be noticed that the designed algorithm is performed by just shuffling the fuel assemblies. The obtained results by employing the mentioned method on a typical reactor reveal that this method has a high precision in achieving a pattern with an allowable radial power peaking factor. (author)

  17. Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    International Nuclear Information System (INIS)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina

    2017-01-01

    The 7 th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7 th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7 th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  18. Assessment of the recovery annealing efficiency for VVER-1000 materials' structure reset and lifetime extension

    International Nuclear Information System (INIS)

    Gurovich, B.; Kuleshova, E.; Prikhodko, K.; Fedotova, S.

    2011-01-01

    The results of the VVER-1000 reactor pressure vessels welds studies based on the surveillance specimens sets have revealed a high embrittlement rate of steel with high nickel content compared with predicted embrittlement determined from the Russian Guide. For these critical vessels further safe operation (even during design service life) is not allowed without additional measures (recovery annealing of the VVER-1000 welds as earlier for VVER- 440). The reason is that the rate of high nickel VVER-1000 welds embrittlement is significantly higher than that is for base metal. In order to solve a problem of VVER-1000 lifetime extension recovery annealing validation and accelerated reirradiation of specimens for prolonged operation period estimation after annealing were necessary. In this work comparison of electron-microscopy fine structure studies and fractographic studies of Charpy specimens fracture surface of the VVER-1000 high nickel welds in different states were carried out. It allows estimation of the recovery annealing effect on steels structure and its behavior at further operation. It is shown that both secondary and primary irradiation causes alike radiation-induced fine structure changes: dislocation loops and nano-size precipitates. Recovery annealing leads to full dislocation loops dissolution and significant nano-size precipitates solution but not to the initial values. The rate of radiation defects and radiation-induced precipitates accumulation at reirradiation weld after recovery annealing is lower than at primary irradiation and determine the lower secondary embrittlement rate of VVER-1000 weld. (authors)

  19. ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

    Energy Technology Data Exchange (ETDEWEB)

    Skutnik, Steven E. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    2017-06-19

    processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.

  20. Inspection qualification programme for VVER reactors and review of round robin test results

    International Nuclear Information System (INIS)

    Horacek, L.; Zdarek, J.

    1998-01-01

    Experience obtained, especially from in-service inspections of VVER 440-type reactor pressure vessels and from the Czech round test trials with international participation of ultrasonic teams, has highlighted the need for an in-service inspection qualification programme in the Czech Republic focused on NDT procedures, equipment and personnel. Recently, several national and international regional projects included in the PHARE programme (projects 4.1.2/93 and 1.02/94), briefly described, have been initiated. These projects are to cover step by step the programme of the in-service inspection qualification in view of technical justification as well as of practical assessment-performance demonstration-for all the main VVER-type primary circuit components. (orig.)

  1. Basket criticality design of a dual purpose cask for VVER 1000 spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rezaeian, Mahdi [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Kamali, Jamshid [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of)

    2016-12-15

    Dual purpose cask technology is one of the most prominent options for interim storage of spent fuels following their removal from reactors. Criticality safety of the spent fuel assemblies are ensured by design of the basket within these casks. In this study, a set of criticality design calculations of a dual purpose cask for 12 VVER 1000 spent fuel assemblies of Bushehr nuclear power plant were carried out. The basket material of borated stainless steel with 0.5 to 2.5 wt% of boron and Boral (Al-B{sub 4}C) with 1.5 to 40 wt% of boron carbide, were investigated and the minimum required receptacle pitch of the basket was determined. Using the calculated receptacle pitch of the basket, the minimum required diameter of the cavity could be established.

  2. Evaluation gives the activity inventory the nuclear fuel irradiated and its radioactive waste

    International Nuclear Information System (INIS)

    Rodriguez Gual, Maritza

    1998-01-01

    The present work has as objectives to give a quantitative evaluation to the activity that possesses the nuclear fuel for 3,6% enrichment with a burnt one the 33 000 NWd/Tu proposed for the Juragua Nuclear Power Plant . In this work the method is used I calculate ORIGEN2. Obtained results are presented and they are compared with other calculations carried out in reactors type VVER-440

  3. Experimental research on safety assurance of advanced WWER fuel cycles

    International Nuclear Information System (INIS)

    Krainov, Ju.; Kukushkin, Ju.

    2002-01-01

    The paper presents the results of experimental investigations on substantiation of implementation of a modernized butt joint for the WWER-440 reactor, carried out in the critical test facility 'P' in the RRC 'Kurchatov Institute'. The comparison results of the calculation and experimental data obtained in the physical startup of Volgodonsk NPP-1 with the WWER-1000 are also given. In the implementation of four-year fuel cycle in the WWER-440 with the average enrichment of fuel makeup 3.82% it was solved to conduct experimental research of power distribution in the vicinity of control rod butt junction. Moreover, it was assumed that adequate actions should be applied to eliminate inadmissible power jumps, if necessary. It is not available to measure their values in NPP conditions. Therefore, the power distribution near the butt joint was studied in a 19-rod bank installed in the critical test facility 'P' first for the normal design of the joint when surrounding fuel assemblies enrichment goes up. Then a set of calculation and tests was fulfilled to optimize a butt junction design. On the base of this research the composition of a butt junction was advanced by placing Hf plates into the junction. The effectiveness of modernized butt joint design was experimentally confirmed. In Volgodonsk NPP-1 with WWER-1000 the four-year fuel cycle is being implemented. During the physical startup of the reactor the measurements of the reactivity effects and coefficients were measured at the minimum controlled flux level, and the parameters of a number of critical states were recorded. The data obtained were compared with the calculation. The validity of the certified code package for forecasting the neutronic characteristics of WWER-1000 cores in the implementation of a four year fuel cycle has been supported (Authors)

  4. The influence of changes in the VVER-1000 fuel assembly shape during operation on the power density distribution

    Energy Technology Data Exchange (ETDEWEB)

    Shishkov, L. K., E-mail: Shishkov-LK@nrcki.ru; Gorodkov, S. S.; Mikailov, E. F.; Sukhino-Homenko, E. A.; Sumarokova, A. S., E-mail: Sumarokova-AS@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2016-12-15

    A new approach to calculation of the coefficients of sensitivity of the fuel pin power to deviations in gap sizes between fuel assemblies of the VVER-1000 reactor during its operation is proposed. It is shown that the calculations by the MCU code should be performed for a full-size model of the core to take the interference of the gap influence into account. In order to reduce the conservatism of calculations, the coolant density and coolant temperature feedbacks should be taken into account, as well as the fuel burnup.

  5. Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2001-01-01

    An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)

  6. Evaluation of an experiment modelling heat transfer from the melt pool for use in VVER 440/213 reactors

    International Nuclear Information System (INIS)

    Skop, J.

    2003-12-01

    The strategy of confining core melt within the reactor vessel is among promising strategies to mitigate severe accidents of VVER 440/213 reactors. This strategy consists in residual heat removal from the melt by external vessel cooling from the outside, using water from the flooded reactor downcomer. This approach can only be successful if the critical heat flux on the external vessel surface is not exceeded. This can be assessed based on the parameters of heat transfer from the core melt pool in the conditions of natural circulation within the pool. Those parameters are the subject of the report. A basic description of the terms and physical basis of the strategy of confining core melt inside the vessel is given in Chapter 2, which also briefly explains similarity theory, based on which the results obtained on experimental facilities, using simulation materials, can be related to the actual situation inside a real reactor. Chapter 3 presents an overview of experimental work addressing the characteristics of heat transfer from the core melt pool in natural circulation conditions and a description of the experimental facilities. An overview of the results emerging from the experiments and their evaluation with respect to their applicability to reactors in Czech nuclear power plants are given in Chapter 4

  7. Start-up of a cold loop in a VVER-440, the 7{sup th} AER benchmark calculation with HEXTRAN-SMABRE-PORFLO

    Energy Technology Data Exchange (ETDEWEB)

    Hovi, Ville; Taivassalo, Veikko; Haemaelaeinen, Anitta; Raety, Hanna; Syrjaelahti, Elina [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    The 7{sup th} dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7{sup th} AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7{sup th} AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.

  8. Construction of the Plant RT-2 as a way for solving the problem of VVER-1000 spent fuel management in Russia

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Lyubtsev, R.I.; Egorov, N.N.; Lebedev, V.A.; Revenko, Y.A.; Fedosov, Y.G.; Dubrovskii, V.M.

    1993-01-01

    Nuclear power in the Russian Federation in the future will be based on the VVER-1000 and it's modifications. To manage the spent fuels from this plant, the Plant RT-2 was designed to process the spent fuel. Plant construction was started in 1984 and stopped in 1989 due to economic difficulties. The necessity of the continuation of the plant is discussed

  9. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    International Nuclear Information System (INIS)

    Vitkova, M.; Manolova, M.; Stefanova, S.; Simeonova, V.; Passage, G.; Lassmann, K.

    1994-01-01

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: 1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; 2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; 3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs

  10. WWER-440 fuel rod performance analysis with PIN-Micro and TRANSURANUS codes

    Energy Technology Data Exchange (ETDEWEB)

    Vitkova, M; Manolova, M; Stefanova, S; Simeonova, V; Passage, G [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika; Kharalampieva, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Lassmann, K [European Atomic Energy Community, Karlsruhe (Germany). European Inst. for Transuranium Elements

    1994-12-31

    PIN-micro and TRANSURANUS codes were used to analyse the WWER-440 fuel rod behaviour at normal operation conditions. Two highest loaded fuel rods of the fuel assemblies irradiated in WWER-440 with different power histories were selected. A set of the most probable average values of all geometrical and technological parameters were used. A comparison between PIN-micro and TRANSURANUS codes was performed using identical input data. The results for inner gas pressure, gap size, local linear heat rate, fuel central temperature and fission gas release as a function of time calculated for the selected fuel rods are presented. The following conclusions were drawn: (1) The PIN-micro code predicts adequately the thermal and mechanical behaviour of the two fuel rods; (2) The comparison of the results obtained by PIN-micro and TRANSURANUS shows a reasonable agreement and the discrepancies could be explained by the lack of thoroughly WWER oriented verification of TRANSURANUS; (3) The advanced TRANSURANUS code could be successfully applied for WWER fuel rod thermal and mechanical analysis after incorporation of all necessary WWER specific material properties and models for the Zr+1%Nb cladding, for the fuel rod as a whole and after validation against WWER experimental and operational data. 1 tab., 10 figs., 10 refs.

  11. Using of the Serpent code based on the Monte-Carlo method for calculation of the VVER-1000 fuel assembly characteristics

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2016-12-01

    Full Text Available The description of calculation scheme of fuel assembly for preparation of few-group characteristics is considered with help of Serpent code. This code uses the Monte-Carlo method and energy continuous microscopic data libraries. Serpent code is devoted for calculation of fuel assembly characteristics, burnup calculations and preparation of few-group homogenized macroscopic cross-sections. The results of verification simulations in comparison with other codes (WIMS, HELIOS, NESSEL etc., which are used for neutron-physical analysis of VVER type fuel, are presented.

  12. VVER-440 control rod follower induced local power peaking computational benchmark: MCNP and KARATE solutions

    International Nuclear Information System (INIS)

    Hegyi, G.; Hordosy, G.; Kereszturi, A.; Maraczy, C.; Temesvari, E.

    2009-01-01

    In this paper the linear pin power calculation in the KARATE-440 code system are presented. The calculation results show that: 1) The coupler mathematical benchmark was solved by the KARATE code system using the same methods and approximations as in case of NPP applications. 2) Without Hf plate in the fuel pin row next to the problematic coupler section pronounced local power peak arises. 4) Though the underprediction of the peak by KARATE-SADR is 4%, it is in accordance with the applied uncertainty of KARATE-SADR. 5) The application of Hf plate solves the problem.

  13. Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

  14. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    Science.gov (United States)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  15. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  16. Used mixed oxide fuel reprocessing at RT-1 plant

    Energy Technology Data Exchange (ETDEWEB)

    Kolupaev, D.; Logunov, M.; Mashkin, A.; Bugrov, K.; Korchenkin, K. [FSUE PA ' Mayak' , 30, Lenins str, Ozersk, 460065 (Russian Federation); Shadrin, A.; Dvoeglazov, K. [ITCP ' PRORYV' , 2/8 Malaya Krasmoselskay str, Moscow, 107140 (Russian Federation)

    2016-07-01

    Reprocessing of the mixed uranium-plutonium spent nuclear fuel of the BN-600 reactor was performed at the RT-1 plant twice, in 2012 and 2014. In total, 8 fuel assemblies with a burn-up from 73 to 89 GW day/t and the cooling time from 17 to 21 years were reprocessed. The reprocessing included the stages of dissolution, clarification, extraction separation of U and Pu with purification from the fission products, refining of uranium and plutonium at the relevant refining cycles. Dissolution of the fuel composition of MOX used nuclear fuel (UNF) in nitric acid solutions in the presence of fluoride ion has occurred with the full transfer of actinides into solution. Due to the high content of Pu extraction separation of U and Pu was carried out on a nuclear-safe equipment designed for the reprocessing of highly enriched U spent nuclear fuel and Pu refining. Technological processes of extraction, separation and refining of actinides proceeded without deviations from the normal mode. The output flow of the extraction outlets in their compositions corresponded to the regulatory norms and remained at the level of the compositions of the streams resulting from the reprocessing of fuel types typical for the RT-1 plant. No increased losses of Pu into waste have been registered during the reprocessing of BN-600 MOX UNF an compare with VVER-440 uranium UNF reprocessing. (authors)

  17. Effect of increased fuel exploitation on the main characteristics of spent WWER 440 fuel

    International Nuclear Information System (INIS)

    Zib, A.

    2001-01-01

    The article deals with the effect of a higher fuel exploitation on the main characteristics (particularly radioactivity and decay heat power) of spent WWER 440 fuel. The main characteristics were calculated by using the Origen code. The study was implemented as a three-stage process. In the first stage, the radioactivity and residual thermal power time evolution values were calculated for the 'typical fuel', i. e. fuel assembly with initial enrichment of 3.6% U-235, 3 years in reactor, and burnup of 30 MWd/kg U. In the second stage, ceteris paribus radioactivity and thermal power analyses of sensitivity to changes in the fuel burnup, initial fuel enrichment, and time in reactor were carried out for the typical fuel assembly. In the third stage, the effect of changes in all three variables was investigated for fuel assemblies possessing parameters that approach those applied at the Dukovany NPP. The effect of a higher fuel exploitation on the interim fuel storage is also mentioned. (author)

  18. Neutronics feasibility of using Gd2O3 particles in VVER-1000 fuel assembly

    International Nuclear Information System (INIS)

    Hoang Van Khanh; Hoang Thanh Phi Hung; Tran Hoai Nam

    2016-01-01

    Neutronics feasibility of using Gd 2 O 3 particles for controlling excess reactivity of VVER-1000 fuel assembly has been investigated. The motivation is that the use of Gd 2 O 3 particles would increase the thermal conductivity of the UO 2 +Gd 2 O 3 fuel pellet which is one of the desirable characteristics for designing future high burnup fuel. The calculation results show that the Gd 2 O 3 particles with the diameter of 60 µm could control the reactivity similarly to that of homogeneous mixture with the same amount of Gd 2 O 3 . The power densities at the fuel pin with Gd 2 O 3 particles increase by about 10-11%, leading to the decrease of the power peak and a slightly flatter power distribution. The power peak appears at the periphery pins at the beginning of burnup process which is decreased by 0.9 % when using Gd 2 O 3 particles. Further work and improvement are being planned to optimize the high power peaking at the beginning of burnup. (author)

  19. The FARC fuel archive of WWER

    International Nuclear Information System (INIS)

    Zizin, M.N.; Parfenova, N.A.; Proselkov, V.N.; Shishkov, L.K.

    1998-01-01

    The principles of organisation are explained and the structure of the FARC fuel archive for WWER reactors is circumscribed. The objective of archive is accumulation of fuel data, data storage and obtaining of fuel using characteristics. The working version of fuel archive on 01 July 1998 is realised, in which the data tables for fuel assemblies for 169 WWER-440 cycles and 35 WWER-1000 cycles are stored. There are two different versions of fuel archive - for WWER-440 (FARC) and for WWER-1000 (FARC1000). A structure of some tables and the texts of programs for them differ. (Authors)

  20. Severe accident experiments on PLINIUS platform. Results of first experiments on COLIMA facility related to VVER-440. Presentation of planned VULCANO and KROTOS tests

    International Nuclear Information System (INIS)

    Piluso, P.; Boccaccio, E.; Bonnet, J.-M.; Journeau, C.; Fouquart, P.; Magallon, D.; Ivanov, I.; Mladenov, I.; Kalchev, S.; Grudev, P.; Alsmeyer, H.; Fluhrer, B.; Leskovar, M.

    2005-01-01

    In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture of nuclear fuel (UO 2 + Fission Products), metallic or oxidized cladding + steel, called c orium , of highly refractory oxides (UO 2 , ZrO 2 ) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the substrate decomposition products (generally oxides such as SiO 2 , Al 2 O 3 , CaO, Fe 2 O 3 ). The French Atomic Energy Commission (CEA) has launched a R and D programme aimed at providing the tools for improving the mastering of severe accidents. It encompasses the development of models and codes, performance of experiments in simulant and prototypic materials and the analysis of international experiments. The experiments with prototypic corium (i.e. material containing depleted UO 2 ) are performed in the PLINIUS experimental platform at CEA Cadarache. It comprises the VULCANO facility for 50-100 kg tests (corium-material interactions, corium solidification etc.), the COLIMA facility for smaller scale (∼1 kg) experiments, the VITI facility for corium properties measurement and the KROTOS facility for corium-water interaction (a few kg). In the framework of the 5 th European Framework Programme, free trans-national access to these facilities has been offered to EU and Associated States researchers. For the first PLINIUS access, COLIMA experiments have been conducted with a Bulgarian Team (TU/SOFIA, BAS/INRNE and NPP/KOZLODUY). This series of tests was devoted to experimental studies on fission products release and corium behaviour in the late phase in a hypothetic case of severe accident in a PWR type VVER-440. The COLIMA experimental results are consistent with previous experiments on irradiated fuels (VERCORS, PHEBUS) with small differences for some fission products and show new results for the remaining corium. For the second visit, scientific users from FZK in Germany were selected to validate the COMET core

  1. Development of some operations in technological flowsheet for spent VVER fuel reprocessing at a pilot plant

    International Nuclear Information System (INIS)

    Lazarev, L.N.; Galkin, B.Ya; Lyubtsev, R.I.; Romanovskii, V.N.; Velikhov, E.P.

    1981-01-01

    The fuel reprocessing pilot plants for high active materials would permit the study and development or particular processing steps and flowsheet variations; in some cases, these experimental installations realize on a small scale practically all technological chains of large reprocessing plants. Such a fuel reprocessing pilot plant with capacity of 3 kg U/d has been built at V. G. Khlopin Radium Institute. The pilot plant is installed in the hot cell of radiochemical compartment, and is composed of the equipments for fuel element cutting and dissolving, the preparation of feed solution (clarification, correction), extraction reprocessing and the production of uranium, plutonium and neptunium concentrates, the complex processing of liquid and solid wastes and a special unit for gas purification and analysis. In the last few years, a series of experiments have been carried out on the reprocessing of spent VVER fuel. (J.P.N.)

  2. Features of fuel performance at high fuel burnups

    International Nuclear Information System (INIS)

    Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

    2001-01-01

    Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

  3. Shut-down margin study for the next generation VVER-1000 reactor including 13 x 13 hexagonal annular assemblies

    International Nuclear Information System (INIS)

    Faghihi, Farshad; Mirvakili, S. Mohammad

    2011-01-01

    Highlights: → Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated. → The MCNP-5 code is run for many cases with different core burn up at various core temperatures. → There is a substantial drop in SDM in the case of annular fuel for the same power level. → SDM for our proposed VVER-1000 annular pins is calculated for specific average fuel burn up values at the BOC, MOC, and EOC. - Abstract: Shut-Down Margin (SDM) for the next generation annular fuel core of typical VVER-1000, 13 x 13 assemblies are calculated as the main aim of the present research. We have applied the MCNP-5 code for many cases with different values of core burn up at various core temperatures, and therefore their corresponding coolant densities and boric acid concentrations. There is a substantial drop in SDM in the case of annular fuel for the same power level. Specifically, SDM for our proposed VVER-1000 annular pins is calculated when the average fuel burn up values at the BOC, MOC, and EOC are 0.531, 11.5, and 43 MW-days/kg-U, respectively.

  4. Evaluation of carbon-14 life cycle in reactors VVER-1000

    International Nuclear Information System (INIS)

    Lysakova, Katerina; Neumann, Jan; Vonkova, Katerina

    2012-09-01

    This work is aimed at the evaluation of carbon-14 life cycle in light water reactors VVER-1000. Carbon-14 is generated as a side product in different systems of nuclear reactors and has been an issue not only in radioactive waste management but mainly in release into the environment in the form of gaseous effluents. The principal sources of this radionuclide are in primary cooling water and fuel. Considerable amount of C-14 is generated by neutron reactions with oxygen 17 O and nitrogen 14 N present in water coolant and fuel. The reaction likelihood and consequently volume of generated radioisotope depends on several factors, especially on the effective cross-section, concentrations of parent elements and conditions of power plant operating strategies. Due to its long half-life and high capability of integration into the environment and thus into the living species, it is very important to monitor the movement of carbon-14 in all systems of nuclear power plant and to manage its release out of NPP. The dominant forms of radioactive carbon-14 are the hydrocarbons owing to the combinations with hydrogen used for absorption of radiolytic oxygen. These organic compounds, such as formaldehyde, methyl alcohol, ethyl alcohol and formic acid can be mostly retained on ion exchange resins used in the system for purifying primary cooling water. The gaseous carbon compounds (CH 4 and CO 2 ) are released into the atmosphere via the ventilation systems of NPP. Based on the information and data obtained from different sources, it has been designed a balance model of possible carbon-14 pathways throughout the whole NPP. This model includes also mass balance model equations for each important node in system and available sampling points which will be the background for further calculations. This document is specifically not to intended to describe the best monitoring program attributes or technologies but rather to provide evaluation of obtained data and find the optimal way to

  5. Molten salt fueled nuclear facility with steam-and gas turbine cycles of heat transformation

    International Nuclear Information System (INIS)

    Ananich, P.I.; Bunin, E.N.; Kazazyan, V.T.; Nemtsev, V.A.; Sikorin, S.N.

    2001-01-01

    The molten salt fueled nuclear facilities with fuel circulating in the primary circuit have a series of the potential advantages in comparison with the traditional thermal and fast reactors with solid fuel elements. These advantages are ensured by the possibility to receive effective neutron balance in the core, minimum margin reactivity, more deep fuel burnup, unbroken correctness of the fuel physical and chemical properties and by low prices of the fuel cycle. The neutron and thermal-physical calculations of the various variants of the MSFNF with steam-water and gas turbine power circuits and their technical and economical comparison are carried out in this article. Calculations of molten salt nuclear power plant with gas turbine power circuit have been carried out using chemically reacting working body ''nitrin'' (N304 + 1%NO). The molten salt fueled reactors with the thermal power near of 2300 MW with two fuel compositions have been considered. The base variant has been taken the design of NPP with VVER NP-1000 when comparing the results of the calculations. Its economical performances are presented in prices of 1990. The results of the calculations show that it is difficult to determine the advantages of any one of the variants considered in a unique fashion. But NPP with MSR possesses large reserves in the process of optimization of cycle and energy equipment parameters that can improve its technical and economical performances sufficiently. (author)

  6. CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.

    2002-01-01

    During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

  7. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  8. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  9. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  10. Plasma, a plant safety monitoring and assessment system for VVER-440 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hornaes, A.; Hulsund, J. E. [Institutt for energiteknikk (IFE), OECD Halden Reactor Project, Halden (Norway); Lipcsei, S.; Major, Cs.; Racz, A.; Vegh, J. [KFKI, Atomic Energy Research Institute, Budapest (Hungary); Eiler, J. [Paks, Nuclear Power Plant Ltd, Paks (Hungary)

    1999-05-15

    The objective with the Plant Safety Monitoring and Assessment System (PLASMA) is to develop an operator support system to support the execution of new symptom-based Emergency Operating Procedures for application in VVER reactors, with the Paks NPP in Hungary as the target plant. Many of the VVER reactors are rewriting their EOPs to comply more with Western standards of symptom-based EOPs. In this connection it is desirable to improve the data validation, information integration and presentation for operators when executing the EOPs. The entry-point to a symptom-oriented procedure is defined by the occurrence of a well-defined reactor operation status, with all its symptoms. However, the application of the EOF benefits from an operator support system, which performs plant status and symptom identification reliably and accurately. The development of the PLASMA system is a joint venture between Institutt for energiteknikk (IFE) and KFKI with the NPP Paks as the target plant. The project has been initiated and partly funded by the Science and Technology Agency (STA), Japan through the OECD NEA assistance program. In Hungary, considerable effort has concentrated on the safety reassessment of the Paks NPP and new EOPs are being written, but no comprehensive Operator Support System (OSS) for plant safety assessment is installed. Some safety parameter display functions are incorporated into diverse operator support systems, but an online 'plant safety monitoring and assessment system' is still missing. The present project comprises designing, constructing, testing and installing such an OSS, which to a great extent could support plant operators in their safety assessment work (author) (ml)

  11. Fuel utilization experience in Bohunice NPP and regulatory requirements for implementation of progressive fuel management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Patenyi, V [Nuclear Regulatory Authority, Bratislava (Slovakia); Darilek, P; Majercik, J [Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

    1994-12-31

    The experience gained in fuel utilization and the basic requirements for fuel licensing in the Slovak NPPs is described. The original project of WWER-440 reactors supposes 3-year fuel cycle with cycle length of about 320 full power days (FPD). Since 1984 it was reduced to 290 FPD. Based on the experience of other countries, a 4-year fuel cycle utilization started in 1987. It is illustrated with data from the Bohunice NPP units. Among 504 fuel assemblies left for the fourth burnup cycle no leakage was observed. The mean burnup achieved in the different units varied from 33.1 to 38.5 Mwd/kg U. The new fuel assemblies used are different from the recent ones in construction, thermohydraulics, water-uranium ratio, enrichment and material design. To meet the safety criteria, regulatory requirements for exploitation of new fuel in WWER-440 were formulated by the Nuclear Regulatory Authority of Slovak Republic. 1 tab., 5 refs.

  12. Nuclear fuel for VVER reactors. Actual state and trends

    International Nuclear Information System (INIS)

    Molchanov, V.

    2009-01-01

    The implementation of 2nd generation of FA with the following performance characteristics: 1) Average FP failure factor during operation is less than 1x10 -6 ; 2) Fuel burn-up: up to 60 GWxd/tHM; 3) 5-year fuel cycle; 4) Unit thermal power uprate up to 110% N nom ; 5) Load Follow Mode (97,5±2,5% and 100-75-100% N nom ) is presented. A new generation of FA (TVSA and TVS-2) is developed, implemented and successfully operated at Russian, Ukrainian and Bulgarian NPPs providing: Safe and reliable operation cycle of 6 years; Assembly burn-up - up to 60 GWxd/tHM; Load Follow Mode. For satisfaction of customer needs regarding nuclear fuel performance the following development of FA design is under way: Power uprate up to 104% N nom and Fuel cycles length up to 18 months

  13. Regulatory use the classification security systems of I and C in VVER type reactors

    International Nuclear Information System (INIS)

    Ilizastegui Perez, F.

    1998-01-01

    Presently work the author proposes a classification to the system I and C to the VVER 440 type reactor in categories the regulatory control with a view to establishing the degree to the attention that the regulator should pay to these systems, leaving the importance that have the same ones for the security the installation, during the execution the works that are carried out with this equipment in the stages construction, setting in service and exploitation

  14. Top-Level Software for VVER-1000 In-core Monitoring System under Implementation of Expanded Nuclear Fuel Diversification Program in Ukraine

    International Nuclear Information System (INIS)

    Khalimonchuk, V.A.

    2015-01-01

    The paper considers the possibility and expediency of developing mathematical software for VVER-1000 ICMS in Ukraine. This mathematical software is among the most important conditions for implementation of the expanded nuclear fuel diversification program. The top-level software is to be developed based on SSTC own studies in the development of codes for power distribution recovery, which were successfully used previously for RBMK-1000 safety analysis

  15. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  16. AER Working Group B activities in 2009

    International Nuclear Information System (INIS)

    Darilek, P.

    2009-01-01

    Regular meeting of Core Design Group was organized by SKODA JS a.s. in Plzen (Czech Republic) during the period of 4 to 6 May 2009. There were presented altogether 17 participants from 7 member organizations and 7 presentations were read. Presented papers covered topics as follows: 1) Two presentations dealt with upgrade of calculation and display tools. 2) Three presentations were devoted to benchmark calculations. 3) Two presentations informed about gradual improvement of fuel assembly and cycle for VVER-440 reactors

  17. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  18. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Sepold, L.

    1995-06-01

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [de

  19. Three-dimensional analysis of the coolant flow characteristics in the fuel assemblies of VVER-1000 reactors

    International Nuclear Information System (INIS)

    Dinh Van Thin; Tran Thi Nhan

    2015-01-01

    Computational Fluid Dynamics (CFD) is a widely used method around the world for complex flow and heat industrial problems. In this paper, the coolant flow parameters were investigated in subchannels of VVER-1000 reactor’s fuel assemblies by ANSYS V14.5 programme. The different mesh solutions and turbulence models were carried out to deal with the water flow problems such as velocity distribution, streamline, temperature and pressure change as well as the hydraulic resistances of the spacer grids. The obtained results are good agreement with the measured values and the published reports from other authors. (author)

  20. Dry spent fuel storage facility at Kozloduy Nuclear Power Plant

    International Nuclear Information System (INIS)

    Goehring, R.; Stoev, M.; Davis, N.; Thomas, E.

    2004-01-01

    The Dry Spent Fuel Storage Facility (DSF) is financed by the Kozloduy International Decommissioning Support Fund (KIDSF) which is managed by European Bank for Reconstruction and Development (EBRD). On behalf of the Employer, the Kozloduy Nuclear Power Plant, a Project Management Unit (KPMU) under lead of British Nuclear Group is managing the contract with a Joint Venture Consortium under lead of RWE NUKEM mbH. The scope of the contract includes design, manufacturing and construction, testing and commissioning of the new storage facility for 2800 VVER-440 spent fuel assemblies at the KNPP site (turn-key contract). The storage technology will be cask storage of CONSTOR type, a steel-concrete-steel container. The licensing process complies with the national Bulgarian regulations and international rules. (authors)

  1. The estimation of the control rods absorber burn-up during the VVER-1000 operation

    Energy Technology Data Exchange (ETDEWEB)

    Bolshagin, Sergey N.; Gorodkov, Sergey S.; Sukhino-Khomenko, Evgeniya A. [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation)

    2013-09-15

    The isotopic composition of the control rods absorber changes under the neutron flux influence, so the control rods efficiency can decrease. In the VVER-1000 control rods boron carbide and dysprosium titanate are used as absorbing materials. In boric part the efficiency decreases due to the {sup 10}B isotope burn-up. Dysprosium isotopes turn into other absorbing isotopes, so the absorbing properties of dysprosium part decrease to a lesser degree. Also the control rod's shells may be deformed as a consequence of boron carbide radiation swelling. This fact should be considered in substantiation of control rods durability. For the estimation of the control rods absorber burn-up two models are developed: VVER-1000 3-D fuel assembly with control rods partially immersed (imitation of the control rods operation in the working group) and VVER-1000 3-D fuel assembly with control rods, located at the upper limit switch (imitation of the control rods operation in groups of the emergency shutdown system). (orig.)

  2. Validation experience with the core calculation program karate

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Hordosy, G.; Kereszturi, A.; Makai, M.; Maraczy, Cs.

    1995-01-01

    A relatively fast and easy-to-handle modular code system named KARATE-440 has been elaborated for steady-state operational calculations of VVER-440 type reactors. It is built up from cell, assembly and global calculations. In the frame of the program neutron physical and thermohydraulic process of the core at normal startup, steady and slow transient can be simulated. The verification and validation of the global code have been prepared recently. The test cases include mathematical benchmark and measurements on operating VVER-440 units. Summary of the results, such as startup parameters, boron letdown curves, radial and axial power distributions of some cycles of Paks NPP is presented. (author)

  3. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  4. Criticality safety and shielding analysis of WWER-440 fuel configurations

    International Nuclear Information System (INIS)

    Christoskov, I.

    2008-01-01

    An overview is made of some studies performed on the criticality safety and radiation shielding analysis of irradiated WWER-440 fuel storage and handling configurations. The analytical tools are based on the SCALE 4.4a code system, in combination with the TORT discrete ordinates transport code and the BUGLE-96 cross-sections library. The accuracy of some important results is assessed through comparison with independent evaluations and with measurement data. (author)

  5. Fast Reactor Fuel Cycle Cost Estimates for Advanced Fuel Cycle Studies

    International Nuclear Information System (INIS)

    Harrison, Thomas

    2013-01-01

    Presentation Outline: • Why Do I Need a Cost Basis?; • History of the Advanced Fuel Cycle Cost Basis; • Description of the Cost Basis; • Current Work; • Fast Reactor Fuel Cycle Applications; • Sample Fuel Cycle Cost Estimate Analysis; • Future Work

  6. Optimization of fuel core loading pattern design in a VVER nuclear power reactors using Particle Swarm Optimization (PSO)

    International Nuclear Information System (INIS)

    Babazadeh, Davood; Boroushaki, Mehrdad; Lucas, Caro

    2009-01-01

    The two main goals in core fuel loading pattern design optimization are maximizing the core effective multiplication factor (K eff ) in order to extract the maximum energy, and keeping the local power peaking factor (P q ) lower than a predetermined value to maintain fuel integrity. In this research, a new strategy based on Particle Swarm Optimization (PSO) algorithm has been developed to optimize the fuel core loading pattern in a typical VVER. The PSO algorithm presents a simple social model by inspiration from bird collective behavior in finding food. A modified version of PSO algorithm for discrete variables has been developed and implemented successfully for the multi-objective optimization of fuel loading pattern design with constraints of keeping P q lower than a predetermined value and maximizing K eff . This strategy has been accomplished using WIMSD and CITATION calculation codes. Simulation results show that this algorithm can help in the acquisition of a new pattern without contravention of the constraints.

  7. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  8. Power peak in vicinity of WWER-440 control rod

    International Nuclear Information System (INIS)

    Mikus, J.

    2003-01-01

    The measurements of axial power (fission density) distribution have been performed by means of gamma activity determination (gamma scanning method) of the irradiated fuel pins detecting gamma quanta in the La peak area - 1596.5 keV of their selected parts having 20 mm length in the axial coordinates range 50-950 mm with a 10 mm step using a rectangular collimator (dimensions 20x10 mm). Two NaI(Tl) scintillation crystals (one as a monitor) with diameter of 40 mm were used, each of them in Pb shielding (thickness 150 mm). The obtained results enlarge the available 'power peaking database' and enable validation of the codes also in important case of zero-boron concentration that corresponds to the end of WWER-440 fuel cycle. This validation can improve the reliability of the calculation results of the power distribution in WWER-440 cores, re-loading schemes etc

  9. Some reactor properties of the new designed nuclear fuels after neutron irradiation

    International Nuclear Information System (INIS)

    Bajan, M.; Necas, V.

    2013-01-01

    The main goal of this paper was perform the optimisation of the fuel assemblies from the profiling point of view as well as the enrichment of individual rods in such a way that the power peaking factor is steady as possible and also the stock of reactivity for six year fuel cycle. For this reason the limit for maximum fuel rod enrichment was increased to 5.95%. The power in the individual rods is the factor, which can limit the total reactor's power, it is very important to minimise the power peaking factor as possible. At the first the power peaking factor of selected fuel assemblies used in VVER-440 reactor were investigated and from results was based perspective designs which was divided into four parts according to the position of pins with gadolinium burnable absorber and according to the shroudless design. From every part the most perspective fuel assembly was chosen. The results are shown in the Fig. 7. The best result is using the shroudless design. As the second best design is fuel assembly with three gadolinium rods in the middle of the assembly. The power peaking factor unsteadiness is much lower as the reference fuel assembly Gd-2. Also it was demonstrate that the increase of enrichment to 5.95% is perspective, because in several designs the difference in enrichment in individual pins was 1% "2'3"5U. Considering only the present allowed value (max 5%) it would not be possible to reach such good power peaking factor and the reactivity sufficient for 6-years fuel cycle. Profiling optimisation together with modernization of structural changes of assembly was achieved the low power peaking factor unsteadiness in individual pins and higher average enrichment of "2"3"5U. So the optimisation can be summarized as very prosperous and perspective. (authors)

  10. Validation of computer codes and modelling methods for giving proof of nuclear saefty of transport and storage of spent VVER-type nuclear fuels. Part 1. Purposes and goals of the project. Final report

    International Nuclear Information System (INIS)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-01-01

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k eff is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [de

  11. Comparative of fuel cycle cost for light water nuclear power plants; Uporedna analiza cene gorivnog ciklusa lakovodne nuklearne elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Kocic, A; Dimitrijevic, Z [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1978-07-01

    Starting from ost general fuel cycle scheme for light water reactors this article deals with conceptual differences of BWR, PWR and WWER as well as with the influence of certain phases of fuel cycle on economic parameters of an equivalent 1000 MWe reactor using a computer program CENA /1/ and typical parameters of each reactor type. An analysis of two particular power plants 628 MWe and 440 MWe WWER by means of the same program is given in the second part of this paper taking into account the differences of in-core fuel management. This second approach is especially interesting for the economy of the power plant itself in the period of planning. (author)

  12. Behaviour of a VVER-1000 fuel element with boron carbide/steel absorber tested under severe fuel damage conditions in the CORA facility (Results of experiment CORA-W2)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.; Sepold, L.

    1994-10-01

    The 'Severe Fuel Damage' (SFD) experiments of the Kernforschungszentrum Karlsruhe (KfK), Federal Republic of Germany, were carried out in the out-of-pile facility 'CORA' as part of the international Severe Fuel Damage (SFD) research. The experimental program was set up to provide information on the failure mechanisms of Light Water Reactor (LWR) fuel elements in a temperature range from 1200 C to 2000 C and in few cases up to 2400 C. Between 1987 and 1992 a total of 17 CORA experiments with two different bundle configurations, i.e. PWR (Pressurized Water Reactor) and BWR (Boiling Water Reactor) bundles were performed. These assemblies represented 'Western-type' fuel elements with the pertinent materials for fuel, cladding, grid spacer, and absorber rod. At the end of the experimental program two VVER-1000 specific tests were run in the CORA facility with identical objectives but with genuine VVER-type materials. The experiments, designated CORA-W1 and CORA-W2 were conducted on February 18, 1993 and April 21, 1993, respectively. Test bundle CORA-W1 was without absorber material whereas CORA-W2 contained one absorber rod (boron carbide/steel). As in the earlier CORA tests the test bundles were subjected to temperature transients of a slow heatup rate in a steam environment. The transient phases of the tests were initiated with a temperature ramp rate of 1 K/s. With these conditions a so-called small-break LOCA was simulated. The temperature escalation due to the exothermal zircon/niobium-steam reaction started at about 1200 C, leading the bundles to maximum temperatures of approximately 1900 C. The thermal response of bundle CORA-W2 is comparable to that of CORA-W1. In test CORA-W2, however, the temperature front moved faster from the top to the bottom compared to test CORA-W1 [de

  13. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kupca, L [VUJE Inc. (Slovakia)

    2012-07-01

    In the Slovak Republic are under operation 6 units (4 in the Jaslovske Bohunice site, and 2 in the Mochovce), 2 units are under construction in Mochovce site. All units are WWER-440 type. The fresh fuel is imported from the Russian Federation. The spent fuel assemblies are stored in wet conditions in Bohunice Interim Storage Spent Fuel Facility (SFIS). By 15 July 2008, there were 8413 assemblies in SFIS. The objectives are: 1) Wet AR storage of spent fuel from the NPP Bohunice and Mochovce: Surveillance of conditions for spent fuel storage in the at-reactor (AR) storage pools of both NPP's (characteristics of pool water, corrosion product data); Visual control of storage pool components; Evaluation of storage conditions with respect to long-term stability (corrosion of fuel cladding, structural materials); 2) Wet SFIS storage at Bohunice: Measurement of spent fuel conditions during the long-term wet storage, activity data in the storage casks and amount of crud; Surveillance program for SFIS structural materials.

  14. Source convergence problems in the application of burnup credit for WWER-440 fuel

    International Nuclear Information System (INIS)

    Hordosy, Gabor

    2003-01-01

    The problems in Monte Carlo criticality calculations caused by the slow convergence of the fission source are examined on an example. A spent fuel storage cask designed for WWER-440 fuel used a sample case. The influence of the main parameters of the calculations is investigated including the initial fission source. A possible strategy is proposed to overcome the difficulties associated by the slow source convergence. The advantage of the proposed strategy that it can be implemented using the standard MCNP features. (author)

  15. Nuclear fuel cycle system analysis

    International Nuclear Information System (INIS)

    Ko, W. I.; Kwon, E. H.; Kim, S. G.; Park, B. H.; Song, K. C.; Song, D. Y.; Lee, H. H.; Chang, H. L.; Jeong, C. J.

    2012-04-01

    The nuclear fuel cycle system analysis method has been designed and established for an integrated nuclear fuel cycle system assessment by analyzing various methodologies. The economics, PR(Proliferation Resistance) and environmental impact evaluation of the fuel cycle system were performed using improved DB, and finally the best fuel cycle option which is applicable in Korea was derived. In addition, this research is helped to increase the national credibility and transparency for PR with developing and fulfilling PR enhancement program. The detailed contents of the work are as follows: 1)Establish and improve the DB for nuclear fuel cycle system analysis 2)Development of the analysis model for nuclear fuel cycle 3)Preliminary study for nuclear fuel cycle analysis 4)Development of overall evaluation model of nuclear fuel cycle system 5)Overall evaluation of nuclear fuel cycle system 6)Evaluate the PR for nuclear fuel cycle system and derive the enhancement method 7)Derive and fulfill of nuclear transparency enhancement method The optimum fuel cycle option which is economical and applicable to domestic situation was derived in this research. It would be a basis for establishment of the long-term strategy for nuclear fuel cycle. This work contributes for guaranteeing the technical, economical validity of the optimal fuel cycle option. Deriving and fulfillment of the method for enhancing nuclear transparency will also contribute to renewing the ROK-U.S Atomic Energy Agreement in 2014

  16. Definition of the seventh dynamic AER benchmark-WWER-440 pressure vessel coolant mixing by re-connection of an isolated loop

    International Nuclear Information System (INIS)

    Kotsarev, A.; Lizorkin, M.; Petrin, R.

    2010-01-01

    The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

  17. Fuel cycle management

    International Nuclear Information System (INIS)

    Herbin, H.C.

    1977-01-01

    The fuel cycle management is more and more dependent on the management of the generation means among the power plants tied to the grid. This is due mainly because of the importance taken by the nuclear power plants within the power system. The main task of the fuel cycle management is to define the refuelling pattern of the new and irradiated fuel assemblies to load in the core as a function of: 1) the differences which exist between the actual conditions of the core and what was expected for the present cycle, 2) the operating constraints and the reactor availability, 3) the technical requirements in safety and the technological limits of the fuel, 4) the economics. Three levels of fuel cycle management can be considered: 1) a long term management: determination of enrichments and expected cycle lengths, 2) a mid term management whose aim corresponds to the evaluation of the batch to load within the core as a function of both: the next cycle length to achieve and the integrated power history of all the cycles up to the present one, 3) a short term management which deals with the updating of the loaded fuel utilisations to take into account the operation perturbations, or with the alteration of the loading pattern of the next batch to respect unexpected conditions. (orig.) [de

  18. Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

    International Nuclear Information System (INIS)

    Stefanova, A.E.; Groudev, P.P.; Atanasova, B.P.

    2009-01-01

    This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies-the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA. To perform this investigation it has been used MELCOR 'input model' for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding. It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety). Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP

  19. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G.

    1999-01-01

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without re-enrichment, the plutonium as conventional Mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  20. CANDU fuel-cycle vision

    International Nuclear Information System (INIS)

    Boczar, P.G

    1998-05-01

    The fuel-cycle path chosen by a particular country will depend on a range of local and global factors. The CANDU reactor provides the fuel-cycle flexibility to enable any country to optimize its fuel-cycle strategy to suit its own needs. AECL has developed the CANFLEX fuel bundle as the near-term carrier of advanced fuel cycles. A demonstration irradiation of 24 CANFLEX bundles in the Point Lepreau power station, and a full-scale critical heat flux (CHF) test in water are planned in 1998, before commercial implementation of CANFLEX fuelling. CANFLEX fuel provides a reduction in peak linear element ratings, and a significant enhancement in thermalhydraulic performance. Whereas natural uranium fuel provides many advantages, the use of slightly enriched uranium (SEU) in CANDU reactors offers even lower fuel-cycle costs and other benefits, such as uprating capability through flattening the channel power distribution across the core. Recycled uranium (RU) from reprocessing spent PWR fuel is a subset of SEU that has significant economic promise. AECL views the use of SEU/RU in the CANFLEX bundle as the first logical step from natural uranium. High neutron economy enables the use of low-fissile fuel in CANDU reactors, which opens up a spectrum of unique fuel-cycle opportunities that exploit the synergism between CANDU reactors and LWRs. At one end of this spectrum is the use of materials from conventional reprocessing: CANDU reactors can utilize the RU directly without reenrichment, the plutonium as conventional mixed-oxide (MOX) fuel, and the actinide waste mixed with plutonium in an inert-matrix carrier. At the other end of the spectrum is the DUPIC cycle, employing only thermal-mechanical processes to convert spent LWR fuel into CANDU fuel, with no purposeful separation of isotopes from the fuel, and possessing a high degree of proliferation resistance. Between these two extremes are other advanced recycling options that offer particular advantages in exploiting the

  1. Increase in VVER type reactor critical heat fluxes due to placing the mixing grids

    International Nuclear Information System (INIS)

    Bezrukov, Y.; Lisenkov, E.; Vasilchenko, I.

    2011-01-01

    The report deals with the results of studies of critical heat fluxes (CHF) on the models of VVER type reactor fuel assembly models equipped with the 'Vihr' intensifiers-grids. The models are the seven-rod bundles with the uniform and non-uniform axial power that correspond to two periods of FA operation i.e. beginning of cycle and end of cycle. The experiments performed showed that the mixing grids of this type are capable of increasing the FA burnout power. The power ascension rate depends on both coolant pressure and steam quality value in the CHF point. Placing the mixing grids in the bundle upper spans results in shifting the point of DNB occurrence downward along the FA height. The experimental data obtained will be used to develop the correlations for determining the CHF in the FA equipped with the mixing grids. (authors)

  2. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  3. Fuel cycle cost uncertainty from nuclear fuel cycle comparison

    International Nuclear Information System (INIS)

    Li, J.; McNelis, D.; Yim, M.S.

    2013-01-01

    This paper examined the uncertainty in fuel cycle cost (FCC) calculation by considering both model and parameter uncertainty. Four different fuel cycle options were compared in the analysis including the once-through cycle (OT), the DUPIC cycle, the MOX cycle and a closed fuel cycle with fast reactors (FR). The model uncertainty was addressed by using three different FCC modeling approaches with and without the time value of money consideration. The relative ratios of FCC in comparison to OT did not change much by using different modeling approaches. This observation was consistent with the results of the sensitivity study for the discount rate. Two different sets of data with uncertainty range of unit costs were used to address the parameter uncertainty of the FCC calculation. The sensitivity study showed that the dominating contributor to the total variance of FCC is the uranium price. In general, the FCC of OT was found to be the lowest followed by FR, MOX, and DUPIC. But depending on the uranium price, the FR cycle was found to have lower FCC over OT. The reprocessing cost was also found to have a major impact on FCC

  4. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  5. Minimization of the fission product waste by using thorium based fuel instead of uranium dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Galahom, A. Abdelghafar, E-mail: Agalahom@yahoo.com

    2017-04-01

    This research discusses the neutronic characteristics of VVER-1200 assembly fueled with five different fuel types based on thorium. These types of fuel based on mixing thorium as a fertile material with different fissile materials. The neutronic characteristics of these fuels are investigated by comparing their neutronic characteristics with the conventional uranium dioxide fuel using the MCNPX code. The objective of this study is to reduce the production of long-lived actinides, get rid of plutonium component and to improve the fuel cycle economy while maintaining acceptable values of the neutronic safety parameters such as moderator temperature coefficient, Doppler coefficient and effective delayed neutrons (β). The thorium based fuel has a more negative Doppler coefficient than uranium dioxide fuel. The moderator temperature coefficient (MTC) has been calculated for the different proposed fuels. Also, the fissile inventory ratio has been calculated at different burnup step. The use of Th-232 as a fertile material instead of U-238 in a nuclear fuel is the most promising fuel in VVER-1200 as it is the ideal solution to avoid the production of more plutonium components and long-lived minor actinides. The reactor grade plutonium accumulated in light water reactor with burnup can be recycled by mixing it with Th-232 to fuel the VVER-1200 assembly. The concentrations of Xe-135 and Sm-151 have been investigated, due to their high thermal neutron absorption cross section.

  6. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  7. Nuclear Fuel Cycle Information System. A directory of nuclear fuel cycle facilities. 2009 ed

    International Nuclear Information System (INIS)

    2009-04-01

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities, published online as part of the Integrated Nuclear Fuel Cycle Information System (iNFCIS: http://www-nfcis.iaea.org/). This is the fourth hardcopy publication in almost 30 years and it represents a snapshot of the NFCIS database as of the end of 2008. Together with the attached CD-ROM, it provides information on 650 civilian nuclear fuel cycle facilities in 53 countries, thus helping to improve the transparency of global nuclear fuel cycle activities

  8. Nuclear-fuel-cycle education: Module 1. Nuclear fuel cycle overview

    International Nuclear Information System (INIS)

    Eckhoff, N.D.

    1981-07-01

    This educational module is an overview of the nuclear-fule-cycle. The overview covers nuclear energy resources, the present and future US nuclear industry, the industry view of nuclear power, the International Nuclear Fuel Cycle Evaluation program, the Union of Concerned Scientists view of the nuclear-fuel-cycle, an analysis of this viewpoint, resource requirements for a model light water reactor, and world nuclear power considerations

  9. Large-scale fuel cycle centres

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The US Nuclear Regulatory Commission (NRC) has considered the nuclear energy centre concept for fuel cycle plants in the Nuclear Energy Centre Site Survey 1975 (NECSS-75) Rep. No. NUREG-0001, an important study mandated by the US Congress in the Energy Reorganization Act of 1974 which created the NRC. For this study, the NRC defined fuel cycle centres as consisting of fuel reprocessing and mixed-oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle centre sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000-300,000MW(e). The types of fuel cycle facilities located at the fuel cycle centre permit the assessment of the role of fuel cycle centres in enhancing the safeguard of strategic special nuclear materials - plutonium and mixed oxides. Siting fuel cycle centres presents a smaller problem than siting reactors. A single reprocessing plant of the scale projected for use in the USA (1500-2000t/a) can reprocess fuel from reactors producing 50,000-65,000MW(e). Only two or three fuel cycle centres of the upper limit size considered in the NECSS-75 would be required in the USA by the year 2000. The NECSS-75 fuel cycle centre evaluation showed that large-scale fuel cycle centres present no real technical siting difficulties from a radiological effluent and safety standpoint. Some construction economies may be achievable with fuel cycle centres, which offer opportunities to improve waste-management systems. Combined centres consisting of reactors and fuel reprocessing and mixed-oxide fuel fabrication plants were also studied in the NECSS. Such centres can eliminate shipment not only of Pu but also mixed-oxide fuel. Increased fuel cycle costs result from implementation of combined centres unless the fuel reprocessing plants are commercial-sized. Development of Pu-burning reactors could reduce any economic penalties of combined centres. The need for effective fissile

  10. Future fuel cycles

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1980-01-01

    A fuel cycle must offer both financial and resource savings if it is to be considered for introduction into Ontario's nuclear system. The most promising alternative CANDU fuel cycles are examined in the context of both of these factors over a wide range of installed capacity growth rates and economic assumptions, in order to determine which fuel cycle, or cycles, should be introduced, and when. It is concluded that the optimum path for the long term begins with the prompt introduction of the low-enriched-uranium fuel cycle. For a wide range of conditions, this cycle remains the optimum throughout the very long term. Conditions of rapid nuclear growth and very high uranium price escalation rates warrant the supersedure of the low-enriched-uranium cycle by either a plutonium-topped thorium cycle or plutonium recycle, beginning between 2010 and 2025. It is also found that the uranium resource position is sound in terms of both known resources and production capability. Moreover, introduction of the low-enriched-uranium fuel cycle and 1250 MWe reactor units will assure the economic viability of nuclear power until at least 2020, even if uranium prices increase at a rate of 3.5% above inflation. The interrelationship between these two conclusions lies in the tremendous incentive for exploration which will occur if the real uranium price escalation rate is high. From a competitive viewpoint, nuclear power can withstand increases in the price of uranium. However, such increases will likely further expand the resource base, making nuclear an even more reliable energy source. (auth)

  11. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

  12. Primary water chemistry for NPP with VVER-TOI

    International Nuclear Information System (INIS)

    Susakin, S.N.; Brykov, S.I.; Zadonsky, N.V.; Bystrova, O.S.

    2012-09-01

    Nowadays within the framework of development of the nuclear power industry in Russia the VVER-TOI reactor is under designing (Standard optimized design). The given design provides for improvement of operation safety level, of technical-economic, operational and load-follow characteristics, and for the raise of competitive capacity of reactor plant and NPP as a whole. In VVER-TOI reactor plant design the primary water chemistry has been improved considering operation experience of VVER reactor plants and a possibility of RP operation under load-follow modes from the viewpoint of meeting the following requirements: - suppression of generation of oxidizing radiolytic products under power operation; - assurance of corrosion resistance of structural materials of equipment and pipelines throughout the NPP design service life; - minimization of deposits on surfaces of the reactor core fuel rods and on heat exchange surface of steam generators; - minimization of accumulation of activated corrosion products; - minimization of the amount of radioactive processing waste. In meeting these requirements an important role is devoted to suppression of generation of oxidizing radiolytic products owing to accumulation of hydrogen in the primary coolant. At NPP with VVER-1000 reactor the ammonia-potassium water chemistry is used wherein the hydrogen accumulation is provided at the expense of ammonia proportioning. Usage of ammonia leads to generation of additional amount of radioactive processing waste and to increased irregularity of maintaining the water chemistry under the daily load-follow modes. In VVER TOI design the primary water chemistry is improved by replacing the proportioning of ammonia with the proportioning of gaseous hydrogen. Different process schemes were considered that provide for a possibility of hydrogen accumulation and maintaining owing to direct proportioning of gaseous hydrogen. The obtained results showed that transition to the potassium water chemistry

  13. Operation results of 3-rd generation nuclear fuel WWER-440 in initial period

    International Nuclear Information System (INIS)

    Adeev, V.; Panov, A.

    2011-01-01

    On unit 4 of Kola NPP trial operation of 3-rd generation's fuel began in 2010. Fuel assemblies of 3-rd generation (FA-3) have a number of design features that provide better operational characteristics. Concise description of a design and the basic advantages of fuel of 3-rd generation are described in articles. Increasing of efficiency of nuclear fuel usage will be achieved by reduction of the parasitic capture of thermal neutrons in constructional materials (weight of zirconium is reduced), optimization of uranium-water relation (increase in fuel elements step), increasing of uranium loading (usage of fuel pellets with increased diameter and without central hole in them). By results of trial operation mass transition to use of given type of assemblies in WWER-440 is possible. This report presents the basic outcomes of the trial operation, a brief survey of the obtained data. The basic characteristics of the reactor core with fuel of 3-rd generation are resulted in work. (authors)

  14. Large-scale fuel cycle centers

    International Nuclear Information System (INIS)

    Smiley, S.H.; Black, K.M.

    1977-01-01

    The United States Nuclear Regulatory Commission (NRC) has considered the nuclear energy center concept for fuel cycle plants in the Nuclear Energy Center Site Survey - 1975 (NECSS-75) -- an important study mandated by the U.S. Congress in the Energy Reorganization Act of 1974 which created the NRC. For the study, NRC defined fuel cycle centers to consist of fuel reprocessing and mixed oxide fuel fabrication plants, and optional high-level waste and transuranic waste management facilities. A range of fuel cycle center sizes corresponded to the fuel throughput of power plants with a total capacity of 50,000 - 300,000 MWe. The types of fuel cycle facilities located at the fuel cycle center permit the assessment of the role of fuel cycle centers in enhancing safeguarding of strategic special nuclear materials -- plutonium and mixed oxides. Siting of fuel cycle centers presents a considerably smaller problem than the siting of reactors. A single reprocessing plant of the scale projected for use in the United States (1500-2000 MT/yr) can reprocess the fuel from reactors producing 50,000-65,000 MWe. Only two or three fuel cycle centers of the upper limit size considered in the NECSS-75 would be required in the United States by the year 2000 . The NECSS-75 fuel cycle center evaluations showed that large scale fuel cycle centers present no real technical difficulties in siting from a radiological effluent and safety standpoint. Some construction economies may be attainable with fuel cycle centers; such centers offer opportunities for improved waste management systems. Combined centers consisting of reactors and fuel reprocessing and mixed oxide fuel fabrication plants were also studied in the NECSS. Such centers can eliminate not only shipment of plutonium, but also mixed oxide fuel. Increased fuel cycle costs result from implementation of combined centers unless the fuel reprocessing plants are commercial-sized. Development of plutonium-burning reactors could reduce any

  15. Economics analysis of fuel cycle cost of fusion–fission hybrid reactors based on different fuel cycle strategies

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2015-01-15

    Highlights: • Economics analysis of fuel cycle cost of FFHRs is carried out. • The mass flows of different fuel cycle strategies are established based on the equilibrium fuel cycle model. • The levelized fuel cycle costs of different fuel cycle strategies are calculated, and compared with current once-through fuel cycle. - Abstract: The economics analysis of fuel cycle cost of fusion–fission hybrid reactors has been performed to compare four fuel cycle strategies: light water cooled blanket burning natural uranium (Strategy A) or spent nuclear fuel (Strategy B), sodium cooled blanket burning transuranics (Strategy C) or minor actinides (Strategy D). The levelized fuel cycle costs (LFCC) which does not include the capital cost, operation and maintenance cost have been calculated based on the equilibrium mass flows. The current once-through (OT) cycle strategy has also been analyzed to serve as the reference fuel cycle for comparisons. It is found that Strategy A and Strategy B have lower LFCCs than OT cycle; although the LFCC of Strategy C is higher than that of OT cycle when the uranium price is at its nominal value, it would become comparable to that of OT cycle when the uranium price reaches its historical peak value level; Strategy D shows the highest LFCC, because it needs to reprocess huge mass of spent nuclear fuel; LFCC is sensitive to the discharge burnup of the nuclear fuel.

  16. Enhancement of Training Capabilities in VVER Technology Through Establishment of VVER Training Academy

    International Nuclear Information System (INIS)

    Ilieva, M.; Miteva, R.

    2015-01-01

    Education and training (E&T) have always been key factor to the sustainability of the nuclear industry. With regard to E&T it is still the challenge to raise the interest of qualified young people of studies and professions related to nuclear technologies. CORONA Project is established to provide a special purpose structure for training and for gathering the existing and generating new knowledge in the VVER area as well as to contribute to transnational mobility and lifelong learning amongst VVER operating countries. CORONA Project consists of two parts: CORONA I (2011–2014) “Establishment of a regional centre of competence for VVER technology and Nuclear Applications”, co-financed by the EC Framework Programme 7 and CORONA II “Enhancement of training capabilities in VVER technology through establishment of VVER training academy”, co-financed by the EURATOM 2014-2015 Working programme of HORIZON 2020. The project is focused on development of training schemes for VVER nuclear professionals, subcontractors, students and for non-nuclear specialists working in support of nuclear applications as civil engineers, physical protection employees, government employees, secondary school teachers, journalists. Safety culture and soft skills training are incorporated as an integral part of all training schemes because they require continuous consideration. It is vital for the acceptance of nuclear energy by the public and for the safe performance of the nuclear installations. CORONA II project is to proceed with the development of state-of-the-art virtual training centre — CORONA Academy. This objective will be realised through networking between universities, research organizations, regulatory bodies, industry and any other organizations involved in the application of nuclear science, ionising radiation and nuclear safety. It will bring together the most experienced trainers and will allow trainees from different locations to access the needed knowledge on demand

  17. Development of the temperature field at the WWER-440 core outlet monitoring system and application of the data analyses methods

    International Nuclear Information System (INIS)

    Spasova, V.; Georgieva, N.; Haralampieva, Tz.

    2001-01-01

    On-line internal reactor monitoring by 216 thermal couples, located at the reactor core outlet, is carried out during power operation of WWER-440 Units 1 and 2 at Kozloduy NPP. Automatic monitoring of technology process is performed by IB-500MA, which collects and performs initial data processing (discrediting and conversion of analogue signals into digital mode). The paper also presents the results and analyses of power distribution monitoring during the past 21-th and current 22-th fuel cycle at Kozloduy NPP, Unit 1 by using archiving system capacity and related software. The possibility to perform operational assessment and analysis of power distribution in the reactor core in each point of the fuel cycle is checked by comparison of the neutron-physical calculation results with reactor coolant system parameters. Paper shows that the processing and analysis of accumulated significant amount of data in the archive files increases accuracy and reliability of power distribution monitoring in the reactor core in each moment of the fuel cycle of WWER-440 reactors at Kozloduy NPP

  18. Status and prospects of the core surveillance system SCORPIO-VVER in Czech Republic and Slovakia

    International Nuclear Information System (INIS)

    Molnar, J.; Vocka, R.

    2011-01-01

    The SCORPIO-VVER core monitoring system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (Czech Republic) and two units of Bohunice NPP (Slovak Republic) replacing the original Russian VK3 system. By both Czech and Slovak nuclear regulatory bodies it was licensed as a Technical Specification Surveillance tool. Since it's first installation, the development of SCORPIO-VVER system continues along with the changes in WWER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The latest most significant changes were done in connection with implementation of a new digital I and C system, loading of the optimized gadolinium bearing Gd2 fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions, in design and methodology of the limit and technical specifications checking (implementation of the on-line shutdown margin calculation to the system) and improvements in the predictive part of the system (Strategy Generator). After the currently finished upgrades (Upgrade 2 at EBO Slovakia in 2009, Upgrade 5 at EDU Czech Republic in 2010) the SCORPIO-VVER is still in focus of Central European nuclear power plants with the road map of modification and implementation up to 2015. In April 2011 the Upgrade 3 at EBO Slovakia has been contracted to support the changed reactor technical specification and for support of the new type of fuel planned to load in 2012. During the summer of 2011 the discussions started with EDU NPP in Czech Republic regarding to the future development of the SCORPIO-VVER system up to 2015. Parallel with the support of current installations at NPPs the project of new installations is ongoing. During

  19. Verifiable Fuel Cycle Simulation Model (VISION): A Tool for Analyzing Nuclear Fuel Cycle Futures

    International Nuclear Information System (INIS)

    Jacobson, Jacob J.; Piet, Steven J.; Matthern, Gretchen E.; Shropshire, David E.; Jeffers, Robert F.; Yacout, A.M.; Schweitzer, Tyler

    2010-01-01

    The nuclear fuel cycle consists of a set of complex components that are intended to work together. To support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system such as how the system would respond to each technological change, a series of which moves the fuel cycle from where it is to a postulated future state. The system analysis working group of the United States research program on advanced fuel cycles (formerly called the Advanced Fuel Cycle Initiative) is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION. For example, VISION users can now change yearly the selection of separation or reactor technologies, the performance characteristics of those technologies, and/or the routing of material among separation and reactor types - with the model still operating on a PC in <5 min.

  20. Economic comparison of fusion fuel cycles

    International Nuclear Information System (INIS)

    Brereton, S.J.; Kazimi, M.S.

    1987-01-01

    The economics of the DT, DD, and DHe fusion fuel cycles are evaluated by comparison on a consistent basis. The designs for the comparison employ HT-9 structure and helium coolant; liquid lithium is used as the tritium breeding material for the DT fuel cycle. The reactors are pulsed, superconducting tokamaks, producing 1200 MW of electric power. The DT and DD designs scan a range of values of plasma beta, assuming first stability scaling laws. The results indicate that on a purely economic basis, the DT fuel cycle is superior to both of the advanced fuel cycles. Geometric factors, materials limitations, and plasma beta were seen to have an impact on the Cost of Electricity (COE). The economics for the DD fuel cycle are more strongly affected by these parameters than is the DT fuel cycle. Fuel costs are a major factor in determining the COE for the DHe fuel cycle. Based on costs directly attributable to the fuel cycle, the DT fuel cycle appears most attractive. Technological advances, improved understanding of physics, or strides in advanced energy conversion schemes may result in altering the economic ranking of the fuel cycles indicated here. 7 refs., 6 figs., 2 tabs

  1. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

  2. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  3. Nuclear-fuel-cycle costs. Consolidated Fuel-Reprocessing Program

    International Nuclear Information System (INIS)

    Burch, W.D.; Haire, M.J.; Rainey, R.H.

    1981-01-01

    The costs for the back-end of the nuclear fuel cycle, which were developed as part of the Nonproliferation Alternative Systems Assessment Program (NASAP), are presented. Total fuel-cycle costs are given for the pressurized-water reactor once-through and fuel-recycle systems, and for the liquid-metal fast-breeder-reactor system. These calculations show that fuel-cycle costs are a small part of the total power costs. For breeder reactors, fuel-cycle costs are about half that of the present once-through system. The total power cost of the breeder-reactor system is greater than that of light-water reactor at today's prices for uranium and enrichment

  4. Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    Panayotov, I.; Mihaylov, N.; Ilieva, K.; Kirilova, D.; Manolova, M.

    2009-01-01

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

  5. Implementation of safety parameter display system at VVER-440 NPPs

    International Nuclear Information System (INIS)

    Manninen, T.

    1997-01-01

    Furnishing WWER-440 nuclear power plant units with a safety parameter display system (SPDS) fulfilling the requirements of internationally recognized standards and guidelines has been ranked high on the lists of proposed safety improvement projects. Technically such an SPDS system can be implemented either as a separate stand-alone system or as a more or less closely integrated part of a process information system of the plant unit. In the paper examples of these approaches are presented. Functionally all these examples include the well proven SPDS concept developed by IVO Power Engineering Ltd, Finland. The functional design basis, the general requirements for the system platform, experience with implementation and expansion possibilities of the systems are discussed. (author)

  6. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  7. Assessment of spent WWER-440 fuel performance under long-term storage conditions

    Energy Technology Data Exchange (ETDEWEB)

    Takats, F [TS Enercon Kft. (Hungary)

    2012-07-01

    Paks Nuclear Power Plant is the only NPP in Hungary. It has four WWER-440 type reactor units. The fresh fuel is imported from Russia so far. The spent fuel assemblies were shipped back to Russia until 1997 after about 6 years cooling at the plant. A dry storage facility (MVDS type) has been constructed and is operational since then. By 1 January 2008, there were 5107 assemblies in dry storage. The objectives are: 1) Wet AR storage of spent fuel from the NPP Paks: Measurements of conditions for spent fuel storage in the at-reactor (AR) storage pools of Paks NPP (physical and chemical characteristics of pool water, corrosion product data); Measurements and visual control of storage pool component characteristics; Evaluation of storage characteristics and conditions with respect to long-term stability (corrosion of fuel cladding, construction materials); 2) Dry AFR storage at Paks NPP: Calculation and measurement of spent fuel conditions during the transfer from the storage pool to the modular vault dry storage (MVDS) on the site; Calculation and measurement of spent fuel conditions during the preparation of fuel for dry storage (drying process), such as crud release, activity build-up; Measurement of spent fuel conditions during the long-term dry storage, activity data in the storage tubes and amount of crud.

  8. VVER-440 loading patterns optimization using ATHENA code

    International Nuclear Information System (INIS)

    Katovsky, K.; Sustek, J.; Bajgl, J.; Cada, R.

    2009-01-01

    In this paper the Czech optimization state-of-the-art, new code system development goals and OPAL optimization system are briefly mentioned. The algorithms, maths, present status and future developments of the ATHENA code are described. A calculation exercise of the Dukovany NPP cycles, on increased power using ATHENA, starting with on-coming 24th cycle (303 FPD) continuing with 25th (322 FPD), and 26th (336 FPD); for all cycles K R ≤1.54 is presented

  9. Analysis of coolant flow in central tube of WWER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Zsiros, G.; Toth, S.; Attila Aszodi, A.

    2011-01-01

    Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

  10. Validation of Advanced Computer Codes for VVER Technology: LB-LOCA Transient in PSB-VVER Facility

    Directory of Open Access Journals (Sweden)

    A. Del Nevo

    2012-01-01

    Full Text Available The OECD/NEA PSB-VVER project provided unique and useful experimental data for code validation from PSB-VVER test facility. This facility represents the scaled-down layout of the Russian-designed pressurized water reactor, namely, VVER-1000. Five experiments were executed, dealing with loss of coolant scenarios (small, intermediate, and large break loss of coolant accidents, a primary-to-secondary leak, and a parametric study (natural circulation test aimed at characterizing the VVER system at reduced mass inventory conditions. The comparative analysis, presented in the paper, regards the large break loss of coolant accident experiment. Four participants from three different institutions were involved in the benchmark and applied their own models and set up for four different thermal-hydraulic system codes. The benchmark demonstrated the performances of such codes in predicting phenomena relevant for safety on the basis of fixed criteria.

  11. Effect of a dual-purpose cask payload increment of spent fuel assemblies from VVER 1000 Bushehr Nuclear Power Plant on basket criticality

    International Nuclear Information System (INIS)

    Rezaeian, M.; Kamali, J.

    2017-01-01

    Dual-purpose casks can be utilized for dry interim storage and transportation of the highly radioactive spent fuel assemblies (SFAs) of Bushehr Nuclear Power Plant (NPP). Criticality safety analysis was carried out using the MCNP code for the cask containing 12, 18, or 19 SFAs. The basket materials of borated stainless steel and Boral (Al-B_4C) were investigated, and the minimum required receptacle pitch of the basket was determined. - Highlights: • Criticality safety analysis for a dual purpose cask was carried out. • The basket material of borated stainless steel and Boral were investigated. • Minimum receptacle pitch was determined for 12, 18, or 19 VVER 1000 spent fuel assemblies.

  12. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  13. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  14. Part 5. Fuel cycle options

    International Nuclear Information System (INIS)

    Lineberry, M.J.; McFarlane, H.F.; Amundson, P.I.; Goin, R.W.; Webster, D.S.

    1980-01-01

    The results of the FBR fuel cycle study that supported US contributions to the INFCE are presented. Fuel cycle technology is reviewed from both generic and historical standpoints. Technology requirements are developed within the framework of three deployment scenarios: the reference international, the secured area, and the integral cycle. Reprocessing, fabrication, waste handling, transportation, and safeguards are discussed for each deployment scenario. Fuel cycle modifications designed to increase proliferation defenses are described and assessed for effectiveness and technology feasibility. The present status of fuel cycle technology is reviewed and key issues that require resolution are identified

  15. Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

    2003-07-01

    The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

  16. Benefits of barrier fuel on fuel cycle economics

    International Nuclear Information System (INIS)

    Crowther, R.L.; Kunz, C.L.

    1988-01-01

    Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect of fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs

  17. Extended fuel cycle length

    International Nuclear Information System (INIS)

    Bruyere, M.; Vallee, A.; Collette, C.

    1986-09-01

    Extended fuel cycle length and burnup are currently offered by Framatome and Fragema in order to satisfy the needs of the utilities in terms of fuel cycle cost and of overall systems cost optimization. We intend to point out the consequences of an increased fuel cycle length and burnup on reactor safety, in order to determine whether the bounding safety analyses presented in the Safety Analysis Report are applicable and to evaluate the effect on plant licensing. This paper presents the results of this examination. The first part indicates the consequences of increased fuel cycle length and burnup on the nuclear data used in the bounding accident analyses. In the second part of this paper, the required safety reanalyses are presented and the impact on the safety margins of different fuel management strategies is examined. In addition, systems modifications which can be required are indicated

  18. Some aspects of nuclear fuel utilisation at Ukrainian's NPPs during last two years

    International Nuclear Information System (INIS)

    Ieremenko, M.; Bilodid, Y.; Ovdiienko, Y.

    2011-01-01

    In the first part of the report the brief characteristic of the realized fuel cycles on Ukraine's NPPs, types of loaded fuel is described. Experience of new fuel type implementation are present (FA Second Generation for WWER-440. Westinghouse FA for WWER-1000). Next issue of report is the some of problem with fuel utilisation (leakage FA, problem with C b calculation and other). And the last issue of report is presentation of future new fuel implementation (WWER-440/1000). (authors)

  19. Practical introduction of thorium fuel cycles

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1982-01-01

    The pracitcal introduction of throrium fuel cycles implies that thorium fuel cycles compete economically with uranium fuel cycles in economic nuclear power plants. In this study the reactor types under consideration are light water reactors (LWRs), heavy water reactors (HWRs), high-temperature gas-cooled reactors (HTGRs), and fast breeder reactors (FBRs). On the basis that once-through fuel cycles will be used almost exclusively for the next 20 or 25 years, introduction of economic thorium fuel cycles appears best accomplished by commercial introduction of HTGRs. As the price of natural uranium increases, along with commercialization of fuel recycle, there will be increasing incentive to utilize thorium fuel cycles in heavy water reactors and light water reactors as well as in HTGRs. After FBRs and fuel recycle are commercialized, use of thorium fuel cycles in the blanket of FBRs appears advantageous when fast breeder reactors and thermal reactors operate in a symbiosis mode (i.e., where 233 U bred in the blanket of a fast breeder reactor is utilized as fissile fuel in thermal converter reactors)

  20. Analysis of the VVER Standard Problem INSC-PSBV1 '11% Coolant Leak from Upper Plenum' with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Melikhov, O.; Melikhov, V.; Parfenov, Yu.; Gavritenkova, O.; Lipatov, I.; Elkin, I.; Bayless, P.

    2004-01-01

    Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than to the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in

  1. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  2. Advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    Green, R.E.; Boczar, P.G.

    1990-04-01

    This paper re-examines the rationale for advanced nuclear fuel cycles in general, and for CANDU advanced fuel cycles in particular. The traditional resource-related arguments for more uranium nuclear fuel cycles are currently clouded by record-low prices for uranium. However, the total known conventional uranium resources can support projected uranium requirements for only another 50 years or so, less if a major revival of the nuclear option occurs as part of the solution to the world's environmental problems. While the extent of the uranium resource in the earth's crust and oceans is very large, uncertainty in the availability and price of uranium is the prime resource-related motivation for advanced fuel cycles. There are other important reasons for pursuing advanced fuel cycles. The three R's of the environmental movement, reduce, recycle, reuse, can be achieved in nuclear energy production through the employment of advanced fuel cycles. The adoption of more uranium-conserving fuel cycles would reduce the amount of uranium which needs to be mined, and the environmental impact of that mining. Environmental concerns over the back end of the fuel cycle can be mitigated as well. Higher fuel burnup reduces the volume of spent fuels which needs to be disposed of. The transmutation of actinides and long-lived fission products into short-lived fission products would reduce the radiological hazard of the waste from thousands to hundreds of years. Recycling of uranium and/or plutonium in spent fuel reuses valuable fissile material, leaving only true waste to be disposed of. Advanced fuel cycles have an economical benefit as well, enabling a ceiling to be put on fuel cycle costs, which are

  3. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu.N.; Gabaraev, B.A.

    2002-01-01

    Full text: The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-gas toppings permits through reducing power of ageing reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation Application of the steam-gas topping permits: extend the service life of ageing VVER-440 reactor by 10...15 years; use the turbine and other NPP balance-of-plant equipment at full power; increase the efficiency of combined cycle up to 48% and more; enhance the safety of NPP operation; utilize NPP balance-of-plant equipment after reactor decommissioning; perform the cost-effective operation in maneuvering modes; increase capacity factor of the plant. The construction of pilot project on the basis of the VK-50 reactor will allow not only to demonstrate new technology but also to attain appreciable economic effect including that obtained due to using the available reserves of the NPP turbine. (author)

  4. The fuel cycle scoping system

    International Nuclear Information System (INIS)

    Dooley, G.D.; Malone, J.P.

    1986-01-01

    The Fuel Cycle Scoping System (FCSS) was created to fill the need for a scoping tool which provides the utilities with the ability to quickly evaluate alternative fuel management strategies, tails assay choices, fuel fabrication quotes, fuel financing alternatives, fuel cycle schedules, and other fuel cycle perturbations. The FCSS was specifically designed for PC's that support dBASE-III(TM), a relational data base software system by Ashton-Tate. However, knowledge of dBASE-III is not necessary in order to utilize the FCSS. The FCSS is menu driven and can be utilized as a teaching tool as well as a scoping tool

  5. Economic evaluation of fast reactor fuel cycling

    International Nuclear Information System (INIS)

    Hu Ping; Zhao Fuyu; Yan Zhou; Li Chong

    2012-01-01

    Economic calculation and analysis of two kinds of nuclear fuel cycle are conducted by check off method, based on the nuclear fuel cycling process and model for fast reactor power plant, and comparison is carried out for the economy of fast reactor fuel cycle and PWR once-through fuel cycle. Calculated based on the current price level, the economy of PWR one-through fuel cycle is better than that of the fast reactor fuel cycle. However, in the long term considering the rising of the natural uranium's price and the development of the post treatment technology for nuclear fuels, the cost of the fast reactor fuel cycle is expected to match or lower than that of the PWR once-through fuel cycle. (authors)

  6. Optimization of the fuel cycle

    International Nuclear Information System (INIS)

    Kidd, S.W.; Balu, K.; Boczar, P.G.; Krebs, W.D.

    1999-01-01

    The nuclear fuel cycle can be optimized subject to a wide range of criteria. Prime amongst these are economics, sustainability of resources, environmental aspects, and proliferation-resistance of the fuel cycle. Other specific national objectives will also be important. These criteria, and their relative importance, will vary from country to country, and with time. There is no single fuel cycle strategy that is optimal for all countries. Within the short term, the industry is attached to dominant thermal reactor technologies, which themselves have two main variants, a cycle closed by reprocessing of spent fuel and subsequent recycling and a once through one where spent fuel is stored in advance of geological disposal. However, even with current technologies, much can be done to optimize the fuel cycles to meet the relevant criteria. In the long term, resource sustainability can be assured for centuries through the use of fast breeder reactors, supporting high-conversion thermal reactors, possibly also utilizing the thorium cycle. These must, however, meet the other key criteria by being both economic and safe. (author)

  7. Core designs for new VVER reactors and operational experience of immediate prototypes

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Mokhov, V.; Ryzhov, S.

    2011-01-01

    The paper covers the recent improvements analyzed in order to implement the enhanced core performances. AES-2006 reactor core design is considered from the point of view of its application and improvement in the planned VVER-TOI project and of the possibilities of using the basic engineering solutions for the cores with spectral control. The discussion of several types of mixing grids considered in the paper involves a preliminary assessment of their efficiency and the information on their introduction into pilot operation at the VVER-1000 Units. Special attention is given to the results of the operation of immediate prototypes (TVS-2 and TVS-2M) that corroborate the reliability of the design both with regard for the core geometrical stability and fuel cladding tightness

  8. Implementation of the neutronics model of HEXTRAN/HEXBU-3D into APROS for WWER calculations

    International Nuclear Information System (INIS)

    Rintala, J.

    2008-01-01

    A new three-dimensional nodal model for neutronics calculation is currently under implementation into APROS - Advanced PROcess Simulation environment - to conform the increasing accuracy requirements. The new model is based on an advanced nodal code HEXTRAN and its static version HEXBU-3D by VTT, Technical Research Centre of Finland. Currently the new APROS is under a testing programme. Later a systematic validation will be performed. In the first phase, a goal is to obtain a fully validated model for VVER-440 calculations. Thus, all the current test calculations are performed by using Loviisa NPP's VVER-440 model of APROS. In future, the model is planned to be applied for the calculations of VVER-1000 type reactors as well as in rectangular fuel geometry. The paper outlines first the general aspects of the method, and then the current situation of the implementation. Because of the identical model with the models of HEXTRAN and HEXBU-3D, the results in the test calculations are compared to the results of those. In the paper, results of two static test calculations are shown. Currently the model works well already in static analyses. Only minor problems with the control assemblies of VVER-440 type reactor still exist but the reasons are known and will be corrected in near future. Dynamical characteristics of the model are up to now tested only by some empirical tests. (author)

  9. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  10. Fuel cycle related parametric study considering long lived actinide production, decay heat and fuel cycle performances

    International Nuclear Information System (INIS)

    Raepsaet, X.; Damian, F.; Lenain, R.; Lecomte, M.

    2001-01-01

    One of the very attractive HTGR reactor characteristics is its highly versatile and flexible core that can fulfil a wide range of diverse fuel cycles. Based on a GTMHR-600 MWth reactor, analyses of several fuel cycles were carried out without taking into account common fuel particle performance limits (burnup, fast fluence, temperature). These values are, however, indicated in each case. Fuel derived from uranium, thorium and a wide variety of plutonium grades has been considered. Long-lived actinide production and total residual decay heat were evaluated for the various types of fuel. The results presented in this papers provide a comparison of the potential and limits of each fuel cycle and allow to define specific cycles offering lowest actinide production and residual heat associated with a long life cycle. (author)

  11. Advanced Fuel Cycle Economic Sensitivity Analysis

    Energy Technology Data Exchange (ETDEWEB)

    David Shropshire; Kent Williams; J.D. Smith; Brent Boore

    2006-12-01

    A fuel cycle economic analysis was performed on four fuel cycles to provide a baseline for initial cost comparison using the Gen IV Economic Modeling Work Group G4 ECON spreadsheet model, Decision Programming Language software, the 2006 Advanced Fuel Cycle Cost Basis report, industry cost data, international papers, the nuclear power related cost study from MIT, Harvard, and the University of Chicago. The analysis developed and compared the fuel cycle cost component of the total cost of energy for a wide range of fuel cycles including: once through, thermal with fast recycle, continuous fast recycle, and thermal recycle.

  12. Fuel-cycle cost comparisons with oxide and silicide fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1982-01-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

  13. Closing the fuel cycle

    International Nuclear Information System (INIS)

    Aycoberry, C.; Rougeau, J.P.

    1987-01-01

    The progressive implementation of some key nuclear fuel cycle capecities in a country corresponds to a strategy for the acquisition of an independant energy source, France, Japan, and some European countries are engaged in such strategic programs. In France, COGEMA, the nuclear fuel company, has now completed the industrial demonstration of the closed fuel cycle. Its experience covers every step of the front-end and of the back-end: transportation of spent fuels, storage, reprocessing, wastes conditioning. The La Hague reprocessing plant smooth operation, as well as the large investment program under active progress can testify of full mastering of this industry. Together with other French and European companies, COGEMA is engaged in the recycling industry, both for uranium through conversion of uranyl nitrate for its further reeichment, and for plutonium through MOX fuel fabrication. Reprocessing and recycling offer the optimum solution for a complete, economic, safe and future-oriented fuel cycle, hence contributing to the necessary development of nuclear energy. (author)

  14. Techniques and results of examination of fission product release from VVER fuel rods with artificial defects and a burnup of ∼60 MWd/kgU at the MIR loop facility

    International Nuclear Information System (INIS)

    Burukin, A.; Goryachev, A.; Ilyenko, S.; Izhutov, A.; Konyashov, V.; Shishin, V.; Shulimov, V.; Luzanova, L.; Miglo, V.

    2009-01-01

    Complex of equipment and several techniques for examination of radioactive fission product release from defective fuel rods were developed, prepared and tested at the PV-1 loop facility of the MIR reactor. During the first test, which was conducted at the PV-1 loop facility and aimed at testing of developed equipment and techniques, measurement of radioactive fission product release from an experimental re-fabricated fuel rod with a burnup of ∼60 MWd/kgU and an artificial defect was performed under design-basis steady-state operating conditions of the VVER-1000 reactor. PIE of all main parameters of the experimental defective fuel rod did not reveal any state peculiarities which could be caused by the artificial defect, i.e. fuel and cladding characteristics in the defect area did not differ from the initial ones (before testing) as well as their characteristics in areas distant from the defect; they are typical for fuel rods with a similar irradiation history in the VVER NPP. The gap in the experimental fuel rod was bridged due to close contact between fuel and cladding at increased fuel burnup; it can appreciable reduce release of radioactive fission products into the PV-1 primary coolant. This suggestion and quantitative characteristics of effect of gap bridging in a high-burnup fuel rod on radioactive fission product release should be investigated during the next tests performed at the PV-1 loop facility. Values of radioactive fission product release measured during the first test at the PV-1 loop facility in the MIR reactor will be used for development of an empirical engineering model in order to take into account high burnup effects and their impact on fission product release from fuel and defective fuel rods

  15. Quadratic reactivity fuel cycle model

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1985-01-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau 2 as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau 2 in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper

  16. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  17. The status of nuclear fuel cycle system analysis for the development of advanced nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kim, Seong Ki; Lee, Hyo Jik; Chang, Hong Rae; Kwon, Eun Ha; Lee, Yoon Hee; Gao, Fanxing [KAERI, Daejeon (Korea, Republic of)

    2011-11-15

    The system analysis has been used with different system and objectives in various fields. In the nuclear field, the system can be applied from uranium mining to spent fuel reprocessing or disposal which is called the nuclear fuel cycle. The analysis of nuclear fuel cycle can be guideline for development of advanced fuel cycle through integrating and evaluating the technologies. For this purpose, objective approach is essential and modeling and simulation can be useful. In this report, several methods which can be applicable for development of advanced nuclear fuel cycle, such as TRL, simulation and trade analysis were explained with case study

  18. Fuel pin bowing and related investigation of WWER-440 control rod influence on power release inside of neighbouring fuel pins

    International Nuclear Information System (INIS)

    Mikus, J.

    2005-01-01

    The purpose of this work consists in investigation of the WWER-440 control rod (CR) influence on space power distribution, especially from viewpoint of the values and gradient occurrence that could result in static and cyclic loads with some consequences, e.g. fuel pin bowing. As known, CR can cause power peaks in periphery fuel pins of adjacent operating assemblies because of the butt joint design of the absorbing adapter to the CR fuel part, that is, presence of the water cavity resulting in a flash up of thermal neutrons. As a consequence, beside well-known peaks in axial power distribution, above power gradients can occur inside of mentioned fuel pins. Because of complicated geometry and material composition of the CR, the detailed calculations concerning both above phenomena are complicated, too. Therefore it is useful to acquire appropriate experimental data to investigate mentioned influence and compare them with calculations. Since detailed power distributions cannot be obtained in the NPP, needed information is provided by means of experiments on research reactors. In case of measurements inside of fuel pins, special (e.g. track) detectors placed between fuel pellets are used. Such works are relatively complicated and time consuming, therefore an evaluation based on mathematical modelling and numerical approximation was proposed by means of that, and using measured power release in some selected fuel pins, information about power release inside of one of these fuel pins, can be obtained. For this purpose, an experiment on light water, zero-power research reactor LR-0 was realized and axial power distribution measurements were performed in a WWER-440 type core near to an authentic CR model. Application of the above evaluation method is demonstrated on one ''investigated'' fuel pin neighbouring CR by means of following results: 1. Axial power distribution inside of investigated fuel pin in two opposite positions on its pellets surface that are situated to

  19. Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

    2000-08-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect.

  20. Compatibility analysis of DUPIC fuel (part5) - DUPIC fuel cycle economics analysis

    International Nuclear Information System (INIS)

    Ko, Won Il; Choi, Hang Bok; Yang, Myung Seung

    2000-08-01

    This study examines the economics of the DUPIC fuel cycle using unit costs of fuel cycle components estimated based on conceptual designs. The fuel cycle cost (FCC) was calculated by a deterministic method in which reference values of fuel cycle components are used. The FCC was then analyzed by a Monte Carlo simulation to get the uncertainty of the FCC associated with the unit costs of the fuel cycle components. From the deterministic analysis on the one-batch equilibrium fuel cycle model, the DUPIC FCC was estimated to be 6.55-6.72 mills/kWh for proposed DUPIC fuel options, which is a little smaller than that of the once-through FCC by 0.04-0.28 mills/kWh. Considering the uncertainty (0.45-0.51 mills/kWh) of the FCC estimated by the Monte Carlo simulation method, the cost difference between the DUPIC and once-through fuel cycle is negligible. On the other hand, the material balance calculation has shown that the DUPIC fuel cycle can save natural uranium resources by -20% and reduce the spent fuel arising by -65%, compared with the once-through fuel cycle. In conclusion, the DUPIC fuel cycle possesses a strong advantage over the once-through fuel cycle from the viewpoint of the environmental effect

  1. Fuel cycle services

    International Nuclear Information System (INIS)

    Gruber, Gerhard J.

    1990-01-01

    TRIGA reactor operators are increasingly concerned about the back end of their Fuel Cycle due to a new environmental policy in the USA. The question how to close the Fuel Cycle will have to be answered by all operators sooner or later. Reprocessing of the TRIGA fuel elements is not available. Only long term storage and final disposal can be considered. But for such a storage or disposal a special treatment of the fuel elements and of course a final depository is necessary. NUKEM plans to undertake efforts to assist the TRIGA operators in this area. For that reason we need to know your special needs for today and tomorrow - so that potential processors can consider whether to offer these services on the market. (orig.)

  2. Physics of fusion-fuel cycles

    International Nuclear Information System (INIS)

    McNally, J.R. Jr.

    1981-01-01

    The evaluation of nuclear fusion fuels for a magnetic fusion economy must take into account the various technological impacts of the various fusion fuel cycles as well as the relative reactivity and the required β's and temperatures necessary for economic steady-state burns. This paper will review some of the physics of the various fusion fuel cycles (D-T, catalyzed D-D, D- 3 He, D- 6 Li, and the exotic fuels: 3 He 3 He and the proton-based fuels such as P- 6 Li, P- 9 Be, and P- 11 B) including such items as: (1) tritium inventory, burnup, and recycle, (2) neutrons, (3) condensable fuels and ashes, (4) direct electrical recovery prospects, (5) fissile breeding, etc. The advantages as well as the disadvantages of the different fusion fuel cycles will be discussed. The optimum fuel cycle from an overall standpoint of viability and potential technological considerations appears to be catalyzed D-D, which could also support smaller relatively clean, lean-D, rich- 3 He satellite reactors as well as fission reactors

  3. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P. G.; Fehrenbach, P. J.; Meneley, D. A.

    1996-01-01

    There are many reasons for countries embarking on a CANDU R program to start with the natural uranium fuel cycle. Simplicity of fuel design, ease of fabrication, and ready availability of natural uranium all help to localize the technology and to reduce reliance on foreign technology. Nonetheless, at some point, the incentives for using natural uranium fuel may be outweighed by the advantages of alternate fuel cycles. The excellent neutron economy, on-line refuelling, and simple fuel-bundle design provide an unsurpassed degree of fuel-cycle flexibility in CANDU reactors. The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a two- to three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than dose conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U. S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or

  4. Homogeneous Thorium Fuel Cycles in Candu Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R.; Magill, M. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada)

    2009-06-15

    The CANDU{sup R} reactor has an unsurpassed degree of fuel-cycle flexibility, as a consequence of its fuel-channel design, excellent neutron economy, on-power refueling, and simple fuel bundle [1]. These features facilitate the introduction and full exploitation of thorium fuel cycles in Candu reactors in an evolutionary fashion. Because thorium itself does not contain a fissile isotope, neutrons must be provided by adding a fissile material, either within or outside of the thorium-based fuel. Those same Candu features that provide fuel-cycle flexibility also make possible many thorium fuel-cycle options. Various thorium fuel cycles can be categorized by the type and geometry of the added fissile material. The simplest of these fuel cycles are based on homogeneous thorium fuel designs, where the fissile material is mixed uniformly with the fertile thorium. These fuel cycles can be competitive in resource utilization with the best uranium-based fuel cycles, while building up a 'mine' of U-233 in the spent fuel, for possible recycle in thermal reactors. When U-233 is recycled from the spent fuel, thorium-based fuel cycles in Candu reactors can provide substantial improvements in the efficiency of energy production from existing fissile resources. The fissile component driving the initial fuel could be enriched uranium, plutonium, or uranium-233. Many different thorium fuel cycle options have been studied at AECL [2,3]. This paper presents the results of recent homogeneous thorium fuel cycle calculations using plutonium and enriched uranium as driver fuels, with and without U-233 recycle. High and low burnup cases have been investigated for both the once-through and U-233 recycle cases. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). 1. Boczar, P.G. 'Candu Fuel-Cycle Vision', Presented at IAEA Technical Committee Meeting on 'Fuel Cycle Options for LWRs and HWRs', 1998 April 28 - May 01, also Atomic Energy

  5. Answering Key Fuel Cycle Questions

    International Nuclear Information System (INIS)

    Piet, S.J.; Dixon, B.W.; Bennett, R.G.; Smith, J.D.; Hill, R.N.

    2004-01-01

    Given the range of fuel cycle goals and criteria, and the wide range of fuel cycle options, how can the set of options eventually be narrowed in a transparent and justifiable fashion? It is impractical to develop all options. We suggest an approach that starts by considering a range of goals for the Advanced Fuel Cycle Initiative (AFCI) and then posits seven questions, such as whether Cs and Sr isotopes should be separated from spent fuel and, if so, what should be done with them. For each question, we consider which of the goals may be relevant to eventually providing answers. The AFCI program has both ''outcome'' and ''process'' goals because it must address both waste already accumulating as well as completing the fuel cycle in connection with advanced nuclear power plant concepts. The outcome objectives are waste geologic repository capacity and cost, energy security and sustainability, proliferation resistance, fuel cycle economics, and safety. The process objectives are rea diness to proceed and adaptability and robustness in the face of uncertainties

  6. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  7. Development of four-year fuel cycle based on the advanced fuel assembly with uranium-gadolinium fuel and its implementation to the operating WWER-440 units

    International Nuclear Information System (INIS)

    Lunin, G.; Novikov, A.; Pavlov, V.; Pavlovichev, P.; Filimonov, P.

    2000-01-01

    Over the past few years in Russia the investigations aimed at the increase of the reliability, safety and efficiency of operation of the WWER-1000 reactors as well as of its competitiveness in the world market were carried out. In the frame of these investigations the four-year fuel cycle, based on advanced fuel assemblies with zirconium alloy spacer grids and guide tubes and with fuel pellet having a reduced diameter of the central hole (1,5 mm), has been developed. For the compensation of a part of excess reactivity, Gd 2 O 3 integrated burnable absorbers are used. CPS absorbing rods contain a combine absorber (B 4 C + Dy 2 O 3 *TiO 2 ). A part of depleted fuel is located on the core periphery. The algorithms controlling the reactor power and power distribution have been updated. For checking of the solutions adopted and for verification of code package developed at the RRC 'Kurchatov Institute' the wide-scale experimental operation of advanced FA and its individual components is carried out. (Authors)

  8. National Policy on Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Soedyartomo, S.

    1996-01-01

    National policy on nuclear fuel cycle is aimed at attaining the expected condition, i.e. being able to support optimality the national energy policy and other related Government policies taking into account current domestic nuclear fuel cycle condition and the trend of international nuclear fuel cycle development, the national strength, weakness, thread and opportunity in the field of energy. This policy has to be followed by the strategy to accomplish covering the optimization of domestic efforts, cooperation with other countries, and or purchasing licences. These policy and strategy have to be broken down into various nuclear fuel cycle programmes covering basically assesment of the whole cycle, performing research and development of the whole cycle without enrichment and reprocessing being able for weapon, as well as programmes for industrialization of the fuel cycle stepwisery commencing with the middle part of the cycle and ending with the edge of the back-end of the cycle

  9. The thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.R.

    1977-01-01

    The utilization of the thorium fuel cycle has long since been considered attractive owing to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Rapidly increasing uranium prices, public reluctance for widespread Pu recycling and expected delays for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, problems associated with reprocessing and waste handling, particularly with re-fabrication by remote handling of 233 U, are certainly not appreciably more difficult than for Pu recycling. To divert from uranium as a nuclear energy source it seems worth while intensifying future efforts for closing the Th/ 233 U fuel cycle. HTGRs are particularly promising for economic application. However, further research and development activities should not concentrate on this reactor type alone. Light- and heavy-water-moderated reactors, and even future fast breeders, may just as well take advantage of a demonstrated thorium fuel cycle. (author)

  10. Fusion of eastern and western technology in VVER 1000 NPP upgrade

    International Nuclear Information System (INIS)

    Ubra, O.; Fleischhans, J.; Kveton, M.

    1997-01-01

    An extensive modernization program upgrading two units of VVER 1000 type of the Czech nuclear power plant (NPP) Temelin to meet the latest international standards is presented. The program is based primarily on combination of eastern and western technology and it has been implemented during plant construction. The NPP Temelin was originally designed according to the standards of the former Soviet Union. After a series of reviews in the 1990s, a decision was made by the Temelin management of upgrade the design of the plant, including the supply of fuel and instrumentation and control system by a western company. The adoption of western technology and practices has helped to solve a large number of IAEA safety issues related to design and operation of VVER 1000 NPP. Details on the current Temelin design and other related safety matters are presented

  11. Plutonium in an enduring fuel cycle

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1998-05-01

    Nuclear fuel cycles evolved over the past five decades have allowed many nations of the world to enjoy the benefits of nuclear energy, while contributing to the sustainable consumption of the world's energy resources. The nuclear fuel cycle for energy production suffered many traumas since the 1970s because of perceived risks of proliferation of nuclear weapons. However, the experience of the past five decades has shown that the world community is committed to safeguarding all fissile materials and continuing the use of nuclear energy resources. Decisions of a few nations to discard spent nuclear fuels in geologic formations are contrary to the goals of an enduring nuclear fuel cycle and sustainable development being pursued by the world community. The maintenance of an enduring nuclear fuel cycle is dependent on sensible management of all the resources of the fuel cycle, including spent fuels

  12. Fast breeder fuel cycle

    International Nuclear Information System (INIS)

    1978-09-01

    Basic elements of the ex-reactor part of the fuel cycle (reprocessing, fabrication, waste handling and transportation) are described. Possible technical and proliferation measures are evaluated, including current methods of accountability, surveillance and protection. The reference oxide based cycle and advanced cycles based on carbide and metallic fuels are considered utilizing conventional processes; advanced nonaqueous reprocessing is also considered. This contribution provides a comprehensive data base for evaluation of proliferation risks

  13. IAEA activities on nuclear fuel cycle 1997

    Energy Technology Data Exchange (ETDEWEB)

    Oi, N [International Atomic Energy Agency, Vienna (Austria). Nuclear Fuel Cycle and Materials Section

    1997-12-01

    The presentation discussing the IAEA activities on nuclear fuel cycle reviews the following issues: organizational charts of IAEA, division of nuclear power and the fuel cycle, nuclear fuel cycle and materials section; 1997 budget estimates; budget trends; the nuclear fuel cycle programme.

  14. IAEA activities on nuclear fuel cycle 1997

    International Nuclear Information System (INIS)

    Oi, N.

    1997-01-01

    The presentation discussing the IAEA activities on nuclear fuel cycle reviews the following issues: organizational charts of IAEA, division of nuclear power and the fuel cycle, nuclear fuel cycle and materials section; 1997 budget estimates; budget trends; the nuclear fuel cycle programme

  15. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

  16. The uranium-plutonium breeder reactor fuel cycle

    International Nuclear Information System (INIS)

    Salmon, A.; Allardice, R.H.

    1979-01-01

    All power-producing systems have an associated fuel cycle covering the history of the fuel from its source to its eventual sink. Most, if not all, of the processes of extraction, preparation, generation, reprocessing, waste treatment and transportation are involved. With thermal nuclear reactors more than one fuel cycle is possible, however it is probable that the uranium-plutonium fuel cycle will become predominant; in this cycle the fuel is mined, usually enriched, fabricated, used and then reprocessed. The useful components of the fuel, the uranium and the plutonium, are then available for further use, the waste products are treated and disposed of safely. This particular thermal reactor fuel cycle is essential if the fast breeder reactor (FBR) using plutonium as its major fuel is to be used in a power-producing system, because it provides the necessary initial plutonium to get the system started. In this paper the authors only consider the FBR using plutonium as its major fuel, at present it is the type envisaged in all, current national plans for FBR power systems. The corresponding fuel cycle, the uranium-plutonium breeder reactor fuel cycle, is basically the same as the thermal reactor fuel cycle - the fuel is used and then reprocessed to separate the useful components from the waste products, the useful uranium and plutonium are used again and the waste disposed of safely. However the details of the cycle are significantly different from those of the thermal reactor cycle. (Auth.)

  17. Standard model of WWER-440 fuel rod for Transuranus and its application for RELAP5 hot channel validation

    International Nuclear Information System (INIS)

    Hatala, B.; Cvan, M.

    2001-01-01

    Within the PECO European Commission project of 'Extension of the validation matrix of the TRANSURANUS code' is developed a generic model of WWER-440 fuel rod. The model is intended to be applied for both realistic and licensing, conservative analysis. For such an application the TRANSURANUS code would be complementary tool to generally used system codes, e.g. RELAP5, providing realistic, more detailed insight into processes and safety criteria, relevant to the fuel rod. The paper presents general description of the model for TRANSURANUS code, brief discussion of approaches used in TRANSURANUS and RELAP5 code safety analysis, accompanied with information about RELAP5 model (whole scope unit model, used for licensing analysis). The existing model for RELAP5 code for WWER-440/V-213 Bohunice V2 unit is checked and modified in hot channel part to allow transparent comparison with the TRANSURANUS code. The results from comparison calculations of the both codes are presented for fresh fuel and quasi steady state scenario and are in good agreement, almost identical. These results might be used as a basis for transient analysis

  18. Fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1984-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors

  19. Overview of the CANDU fuel handling system for advanced fuel cycles

    International Nuclear Information System (INIS)

    Koivisto, D.J.; Brown, D.R.

    1997-01-01

    Because of its neutron economies and on-power re-fuelling capabilities the CANDU system is ideally suited for implementing advanced fuel cycles because it can be adapted to burn these alternative fuels without major changes to the reactor. The fuel handling system is adaptable to implement advanced fuel cycles with some minor changes. Each individual advanced fuel cycle imposes some new set of special requirements on the fuel handling system that is different from the requirements usually encountered in handling the traditional natural uranium fuel. These changes are minor from an overall plant point of view but will require some interesting design and operating changes to the fuel handling system. Some preliminary conceptual design has been done on the fuel handling system in support of these fuel cycles. Some fuel handling details were studies in depth for some of the advanced fuel cycles. This paper provides an overview of the concepts and design challenges. (author)

  20. The thorium fuel cycle

    International Nuclear Information System (INIS)

    Merz, E.R.

    1977-01-01

    The utilization of the thorium fuel cycle has long since been considered attractive due to the excellent neutronic characteristics of 233 U, and the widespread and cheap thorium resources. Although the uranium ore as well as the separative work requirements are usually lower for any thorium-based fuel cycle in comparison to present uranium-plutonium fuel cycles of thermal water reactors, interest by nuclear industry has hitherto been marginal. Fast increasing uranium prices, public reluctance against widespread Pu-recycling and expected retardations for the market penetration of fast breeders have led to a reconsideration of the thorium fuel cycle merits. In addition, it could be learned in the meantime that problems associated with reprocessing and waste handling, but particularly with a remote refabrication of 233 U are certainly not appreciably more difficult than for Pu-recycling. This may not only be due to psychological constraints but be based upon technological as well as economical facts, which have been mostly neglected up till now. In order to diversify from uranium as a nuclear energy source it seems to be worthwhile to greatly intensify efforts in the future for closing the Th/ 233 U fuel cycle. HTGR's are particularly promising for economic application. However, further R and D activites should not be solely focussed on this reactor type alone. Light and heavy-water moderated reactors, as well as even fast breeders later on, may just as well take advantage of a demonstrated thorium fuel cycle. A summary is presented of the state-of-the-art of Th/ 233 U-recycling technology and the efforts still necessary to demonstrate this technology all the way through to its industrial application

  1. World nuclear fuel cycle requirements 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-10

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs.

  2. World nuclear fuel cycle requirements 1991

    International Nuclear Information System (INIS)

    1991-01-01

    The nuclear fuel cycle consists of mining and milling uranium ore, processing the uranium into a form suitable for generating electricity, ''burning'' the fuel in nuclear reactors, and managing the resulting spent nuclear fuel. This report presents projections of domestic and foreign requirements for natural uranium and enrichment services as well as projections of discharges of spent nuclear fuel. These fuel cycle requirements are based on the forecasts of future commercial nuclear power capacity and generation published in a recent Energy Information Administration (EIA) report. Also included in this report are projections of the amount of spent fuel discharged at the end of each fuel cycle for each nuclear generating unit in the United States. The International Nuclear Model is used for calculating the projected nuclear fuel cycle requirements. 14 figs., 38 tabs

  3. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR (Integral Fast Reactor) program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. (author)

  4. IFR fuel cycle demonstration in the EBR-II Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lineberry, M.J.; Phipps, R.D.; Rigg, R.H.; Benedict, R.W.; Carnes, M.D.; Herceg, J.E.; Holtz, R.E.

    1991-01-01

    The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase which includes completion of facility modifications, and installation and cold checkout of process equipment. This paper reviews the design and construction of the facility, the design and fabrication of the process equipment, and the schedule and initial plan for its operation. 5 refs., 4 figs

  5. DUPIC fuel cycle economics assessment (1)

    International Nuclear Information System (INIS)

    Choi, H. B.; Roh, G. H.; Kim, D. H.

    1999-04-01

    This is a state-of-art report that describes the current status of the DUPIC fuel cycle economics analysis conducted by the DUPIC fuel compatibility assessment group of the DUPIC fuel development project. For the DUPIC fuel cycle economics analysis, the DUPIC fuel compatibility assessment group has organized the 1st technical meeting composed of 8 domestic specialists from government, academy, industry, etc. and a foreign specialist of hot-cell design from TRI on July 16, 1998. This report contains the presentation material of the 1st technical meeting, published date used for the economics analysis and opinions of participants, which could be utilized for further DUPIC fuel cycle and back-end fuel cycle economics analyses. (author). 11 refs., 7 charts

  6. IFR fuel cycle

    International Nuclear Information System (INIS)

    Battles, J.E.; Miller, W.E.; Lineberry, M.J.; Phipps, R.D.

    1992-01-01

    The next major milestone of the IFR program is engineering-scale demonstration of the pyroprocess fuel cycle. The EBR-II Fuel Cycle Facility has just entered a startup phase, which includes completion of facility modifications and installation and cold checkout of process equipment. This paper reviews the development of the electrorefining pyroprocess, the design and construction of the facility for the hot demonstration, the design and fabrication of the equipment, and the schedule and initial plan for its operation

  7. Fuel cycle management in Finland

    International Nuclear Information System (INIS)

    Vaeyrynen, H.; Mikkola, I.

    1987-01-01

    Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

  8. Method for determining the outlet temperature of fuel assemblies unsupplied with thermometer in WWER-440 reactors

    International Nuclear Information System (INIS)

    Miko, S.; Kalya, Z.; Hamvas, I.

    1987-09-01

    The paper outlines a method for the evaluation of the outlet temperatures of fuel assemblies unsupplied with thermometer in WWER-440 reactors. The process is based on interpolation of directly measured assembly temperatures. A quantitative comparison of the errors of described algorithm to those of standard plant-computer interpolation rutine is also presented. (author)

  9. The DUPIC alternative for backend fuel cycle

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Boczar, P.; Sullivan, J.; Gadsby, R.D.

    1997-01-01

    The DUPIC fuel cycle was conceived as an alternative to the conventional fuel cycle backed options, with a view to multiple benefits expectable from burning spent PWR fuel again in CANDU reactors. It is based on the basic idea that the bulk of spent PWR fuel can be directly refabricated into a reusable fuel for CANDU of which high efficiency in neutron utilization would exhaustively burn the fissile remnants in the spent PWR fuel to a level below that of natural uranium. Such ''burn again'' strategy of the DUPIC fuel cycle implies that the spent PWR fuel will become CANDU fuel of higher burnup with relevant benefits such as spent PWR fuel disposition, saving of natural uranium fuel, etc. A salient feature of the DUPIC fuel cycle is neither the fissile content nor the bulk radioactivity is separated from the DUPIC mass flow which must be contained and shielded all along the cycle. This feature can be considered as a factor of proliferation resistance by deterrence against access to sensitive materials. It means also the requirement for remote systems technologies for DUPIC fuel operation. The conflicting aspects between better safeguardability and harder engineering problems of the radioactive fuel operation may be the important reason why the decades' old concept, since INFCE, of ''hot'' fuel cycle has not been pursued with much progress. In this context, the DUPIC fuel cycle could be a live example for development of proliferation resistant fuel cycle. As the DUPIC fuel cycle looks for synergism of fuel linkage from PWR to CANDU (or in broader sense LWR to HWR), Korea occupies a best position for DUPIC exercise with her unique strategy of reactor mix of both reactor types. But the DUPIC benefits can be extended to global bonus, expectable from successful development of the technology. (author)

  10. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  11. Fuel cycles using adulterated plutonium

    International Nuclear Information System (INIS)

    Brooksbank, R.E.; Bigelow, J.E.; Campbell, D.O.; Kitts, F.G.; Lindauer, R.B.

    1978-01-01

    Adjustments in the U-Pu fuel cycle necessitated by decisions made to improve the nonproliferation objectives of the US are examined. The uranium-based fuel cycle, using bred plutonium to provide the fissile enrichment, is the fuel system with the highest degree of commercial development at the present time. However, because purified plutonium can be used in weapons, this fuel cycle is potentially vulnerable to diversion of that plutonium. It does appear that there are technologically sound ways in which the plutonium might be adulterated by admixture with 238 U and/or radioisotopes, and maintained in that state throughout the fuel cycle, so that the likelihood of a successful diversion is small. Adulteration of the plutonium in this manner would have relatively little effect on the operations of existing or planned reactors. Studies now in progress should show within a year or two whether the less expensive coprocessing scheme would provide adequate protection (coupled perhaps with elaborate conventional safeguards procedures) or if the more expensive spiked fuel cycle is needed as in the proposed civex pocess. If the latter is the case, it will be further necessary to determine the optimum spiking level, which could vary as much as a factor of a billion. A very basic question hangs on these determinations: What is to be the nature of the recycle fuel fabrication facilities. If the hot, fully remote fuel fabrication is required, then a great deal of further development work will be required to make the full cycle fully commercial

  12. CORONA project -contribution to VVER nuclear education and training

    International Nuclear Information System (INIS)

    Ilieva, M.; Miteva, R.; Takov, T.

    2016-01-01

    CORONA Project is established to stimulate the transnational mobility and lifelong learning amongst VVER end users. The project aims to provide a special purpose structure for training of specialists and to maintain the nuclear expertise by gathering the existing and generating new knowledge in the VVER area. CORONA Project consists of two parts: CORONA I (2011-2014) ''Establishment of a regional center of competence for VVER technology and Nuclear Applications'', co-financed by the Framework Program 7 of the European Union (EU) and CORONA II (2015-2018) ''Enhancement of training capabilities in VVER technology through establishment of VVER training academy'', co-financed by HORIZON 2020, EURATOM 2014-2015. The selected form of the CORONA Academy, together with the online availability of the training opportunities will allow trainees from different locations to access the needed knowledge on demand. The project will target also new-comers in VVER community like Vietnam, Turkey, Belarus, etc. (authors)

  13. Dynamic Simulations of Advanced Fuel Cycles

    International Nuclear Information System (INIS)

    Piet, Steven J.; Dixon, Brent W.; Jacobson, Jacob J.; Matthern, Gretchen E.; Shropshire, David E.

    2011-01-01

    Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the U.S. Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe 'lessons learned' from dynamic simulations but attempt to answer the 'so what' question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

  14. CANDU fuel cycle options in Korea

    International Nuclear Information System (INIS)

    Boczar, P.G.; Fehrenbach, P.J.; Meneley, D.A.

    1996-04-01

    The easiest first step in CANDU fuel-cycle evolution may be the use of slightly enriched uranium (SEU), including recovered uranium from reprocessed LWR spent fuel. Relatively low enrichment (up to 1.2%) will result in a twoto three-fold reduction in the quantity of spent fuel per unit energy production, reductions in fuel-cycle costs, and greater flexibility in the design of new reactors. The CANFLEX (CANDU FLEXible) fuel bundle would be the optimal fuel carrier. A country that has both CANDU and PWR reactors can exploit the natural synergism between these two reactor types to minimize overall waste production, and maximize energy derived from the fuel. This synergism can be exploited through several different fuel cycles. A high burnup CANDU MOX fuel design could be used to utilize plutonium from conventional reprocessing or more advanced reprocessing options (such as co-processing). DUPIC (Direct Use of Spent PWR Fuel In CANDU) represents a recycle option that has a higher degree of proliferation resistance than does conventional reprocessing, since it uses only dry processes for converting spent PWR fuel into CANDU fuel, without separating the plutonium. Good progress is being made in the current KAERI, AECL, and U.S. Department of State program in demonstrating the technical feasibility of DUPIC. In the longer term, CANDU reactors offer even more dramatic synergistic fuel cycles with PWR or FBR reactors. If the objective of a national fuel-cycle program is the minimization of actinide waste or destruction of long-lived fission products, then studies have shown the superiority of CANDU reactors in meeting this objective. Long-term energy security can be assured either through the thorium cycle or through a CANDU 1 FBR system, in which the FBR would be operated as a 'fuel factory,' providing the fissile material to power a number of lower-cost, high efficiency CANDU reactors. In summary, the CANDU reactor's simple fuel design, high neutron economy, and on

  15. Overview of VVER water chemistry

    International Nuclear Information System (INIS)

    Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Sundar, R.S.

    2007-01-01

    Kudankulam Nuclear Power project is having twin units of 1000MWe of VVER type. This paper highlights the different analytical techniques that are followed to maintain the system chemistry within the technical specifications. This paper also briefs the different chemicals that are added to the systems and how they are monitored. Basic differences with respect to chemistry between a PHWR and VVER are also highlighted in this paper. (author)

  16. Investigation of neutron physical features of WWER-440 assembly containing differently enriched pins and Gd burnable poison

    International Nuclear Information System (INIS)

    Nemes, Imre

    2000-01-01

    In this paper different pin-distributions of WWER-440 fuel assembly are examined. Assemblies contain 3 Gd-doped pins (Hungarian design), 6 Gd-doped pins near the assembly corners (Russian design) and differently profiled U5-enrichment in different pins. The main neutron physical characteristics of this assemblies - as the function of burnup - are calculated using HELIOS code. The calculated parameters of different assembly designs are analyzed from the standpoint of fuel cycle economy and refueling design practice. (Authors)

  17. Recent IAEA activities on CANDU-PHWR fuels and fuel cycles

    International Nuclear Information System (INIS)

    Inozemtsev, V.; Ganguly, C.

    2005-01-01

    Pressurized Heavy Water Reactors (PHWR), widely known as CANDU, are in operation in Argentina, Canada, China, India, Pakistan, Republic of Korea and Romania and account for about 6% of the world's nuclear electricity production. The CANDU reactor and its fuel have several unique features, like horizontal calandria and coolant tubes, on-power fuel loading, thin-walled collapsible clad coated with graphite on the inner surface, very high density (>96%TD) natural uranium oxide fuel and amenability to slightly enriched uranium oxide, mixed uranium plutonium oxide (MOX), mixed thorium plutonium oxide, mixed thorium uranium (U-233) oxide and inert matrix fuels. Several Technical Working Groups (TWG) of IAEA periodically discuss and review CANDU reactors, its fuel and fuel cycle options. These include TWGs on water-cooled nuclear power reactor Fuel Performance and Technology (TWGFPT), on Nuclear Fuel Cycle Options and spent fuel management (TWGNFCO) and on Heavy Water Reactors (TWGHWR). In addition, IAEA-INPRO project also covers Advanced CANDU Reactors (ACR) and DUPIC fuel cycles. The present paper summarises the Agency's activities in CANDU fuel and fuel cycle, highlighting the progress during the last two years. In the past we saw HWR and LWR technologies and fuel cycles separate, but nowadays their interaction is obviously growing, and their mutual influence may have a synergetic character if we look at the world nuclear fuel cycle as at an integrated system where the both are important elements in line with fast neutron, gas cooled and other advanced reactors. As an international organization the IAEA considers this challenge and makes concrete steps to tackle it for the benefit of all Member States. (author)

  18. Economic aspects of Dukovany NPP fuel cycle

    International Nuclear Information System (INIS)

    Vesely, P.; Borovicka, M.

    2001-01-01

    The paper discusses some aspects of high burnup program implementation at Dukovany NPP and its influence on the fuel cycle costs. Dukovany internal fuel cycle is originally designed as a three years cycle of the Out-In-In fuel reloading patterns. These reloads are not only uneconomical but they additionally increased the radiation load of the reactor pressure vessel due to high neutron leakage typical for Out-In-In loading pattern. To avoid the high neutron leakage from the core a transition to 4-year fuel cycle is started in 1987. The neutron leakage from the core is sequentially decreased by insertion of older fuel assemblies at the core periphery. Other developments in fuel cycle are: 1) increasing of enrichment in control assemblies (3.6% of U-235); 2) improvement in fuel assembly design (reduce the assembly shroud thickness from 2.1 to 1.6 mm); 3) introduction of Zr spacer grid instead of stainless steel; 4) introduction of new type of assembly with profiled enrichment with average value of 3.82%. Due to increased reactivity of the new assemblies the transition to the partial 5-year fuel cycle is required. Typical fuel loading pattern for 3, 3.5, 4 and 5-year cycles are shown in the presented paper. An evaluation of fuel cost is also discussed by using comparative analysis of different fuel cycle options. The analysis shows that introduction of the high burnup program has decrease relative fuel cycle costs

  19. Fuel cycle oriented approach

    International Nuclear Information System (INIS)

    Petit, A.

    1987-01-01

    The term fuel cycle oriented approach is currently used to designate two quite different things: the attempt to consider all or part of a national fuel cycle as one material balance area (MBA) or to consider individual MBAs existing in a state while designing a unique safeguards approach for each and applying the principle of nondiscrimination to fuel cycles as a whole, rather than to individual facilities. The merits of such an approach are acceptability by the industry and comparison with the contemplated establishment of long-term criteria. The following points concern the acceptability by the industry: (1) The main interest of the industry is to keep an open international market and therefore, to have effective and efficient safeguards. (2) The main concerns of the industry regarding international safeguards are economic burden, intrusiveness, and discrimination. Answers to these legitimate concerns, which retain the benefits of a fuel cycle oriented approach, are needed. More specifically, the problem of reimbursing the operator the costs that he has incurred for the safeguards must be considered

  20. Analysis of possible fuel cycles

    International Nuclear Information System (INIS)

    Boehm, H.; Kessler, G.; Engelmann, P.; Maerkl, H.; Stoll, W.

    1978-01-01

    A brief survey is presented of the most important fuel cycles. A rough analysis of fuel cycles is attempted under the aspects of proliferation, status of technical feasibility, resource conservation and waste management and the most important criteria for such an analysis are discussed. Among the multitude of potential combinations of fuel cycles and types of reactors only a few have reached a level of technical feasibility which would make them eligible for commercial implementation within the next decade. However, if, for instance, the higher proliferation resistance of a specific fuel cycle is to be utilized to diminish the worldwide proliferation hazard, that cycle would first of all have to be introduced on an industrial scale as quickly as possible. The analysis shows that the reduction of the bazard of worldwide proliferation will continue to be the objective primarily of international agreements and measures taken in the political realm. (orig.) [de

  1. Feasibility study on tandem fuel cycle

    International Nuclear Information System (INIS)

    Han, P.S.; Suh, I.S.; Rim, C.S.; Kim, B.K.; Suh, K.S.; Ro, S.K.; Juhn, P.I.; Kim, S.Y.

    1983-01-01

    The objective of this feasibility study is to review and assess the current state of technology concerning the tandem fuel cycle. Based on the results from this study, a long-term development plan suitable for Korea has been proposed for this cycle, i.e., the PWR → CANDU tandem fuel cycle which used plutonium and uranium, recovered from spent PWR fuel by co-processing, as fuel material for CANDU reactors. (Author)

  2. French development program on fuel cycle

    International Nuclear Information System (INIS)

    Viala, M.; Bourgeois, M.

    1991-01-01

    The need to close the fuel cycle of fast reactors makes the development of the cycle installations (fuel fabrication, irradiated assembly conditioning before reprocessing, reprocessing and waste management) especially independent with the development of the reactor. French experience with the integrated cycle over a period of about 25 years, the tonnage of fuels fabricated (more than 100 t of mixed oxides) for the Rapsodie, Phoenix and SuperPhoenix reactors, and the tonnage of reprocessed fuel (nearly 30 t of plutonium fuel) demonstrate the control of the cycle operations. The capacities of the cycle installations in existence and under construction are largely adequate for presents needs, even including a new European EFR reactor. They include the Cadarache fuel fabrication complex, the La Hague UP2-800 reprocessing plant, and the Marcoule pilot facility. Short- and medium-term R and D programs are connected with fuel developments, with the primary objective of very high burnups. For the longer term and for a specific plant to reprocess fast reactor fuels, the programs could concern new fabrication and reprocessing systems and the study of the consequences of the reduction in fuel out-of-core time

  3. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  4. Proliferation resistance fuel cycle technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. S.; Ko, W. I

    1999-02-01

    The issues of dual use in nuclear technology are analysed for nuclear fuel cycle with special focus on uranium enrichment and spent fuel reprocessing which are considered as the most sensitive components in terms of vulnerability to diversion. Technical alternatives to mitigrate the vulnerability, as has been analysed in depth during the NASAP and INFCE era in the late seventies, are reviewed to characterize the DUPIC fuel cycle alternative. On the other hand, the new realities in nuclear energy including the disposition of weapon materials as a legacy of cold war are recast in an angle of nuclear proliferation resistance and safeguards with a discussion on the concept of spent fuel standard concept and its compliance with the DUPIC fuel cycle technology. (author)

  5. Thorium fuel cycle - Potential benefits and challenges

    International Nuclear Information System (INIS)

    2005-05-01

    There has been significant interest among Member States in developing advanced and innovative technologies for safe, proliferation resistant and economically efficient nuclear fuel cycles, while minimizing waste and environmental impacts. This publication provides an insight into the reasons for renewed interest in the thorium fuel cycle, different implementation scenarios and options for the thorium cycle and an update of the information base on thorium fuels and fuel cycles. The present TECDOC focuses on the upcoming thorium based reactors, current information base, front and back end issues, including manufacturing and reprocessing of thorium fuels and waste management, proliferation-resistance and economic issues. The concluding chapter summarizes future prospects and recommendations pertaining to thorium fuels and fuel cycles

  6. Sustainability Features of Nuclear Fuel Cycle Options

    Directory of Open Access Journals (Sweden)

    Stefano Passerini

    2012-09-01

    Full Text Available The nuclear fuel cycle is the series of stages that nuclear fuel materials go through in a cradle to grave framework. The Once Through Cycle (OTC is the current fuel cycle implemented in the United States; in which an appropriate form of the fuel is irradiated through a nuclear reactor only once before it is disposed of as waste. The discharged fuel contains materials that can be suitable for use as fuel. Thus, different types of fuel recycling technologies may be introduced in order to more fully utilize the energy potential of the fuel, or reduce the environmental impacts and proliferation concerns about the discarded fuel materials. Nuclear fuel cycle systems analysis is applied in this paper to attain a better understanding of the strengths and weaknesses of fuel cycle alternatives. Through the use of the nuclear fuel cycle analysis code CAFCA (Code for Advanced Fuel Cycle Analysis, the impact of a number of recycling technologies and the associated fuel cycle options is explored in the context of the U.S. energy scenario over 100 years. Particular focus is given to the quantification of Uranium utilization, the amount of Transuranic Material (TRU generated and the economics of the different options compared to the base-line case, the OTC option. It is concluded that LWRs and the OTC are likely to dominate the nuclear energy supply system for the period considered due to limitations on availability of TRU to initiate recycling technologies. While the introduction of U-235 initiated fast reactors can accelerate their penetration of the nuclear energy system, their higher capital cost may lead to continued preference for the LWR-OTC cycle.

  7. Synergistic fuel cycles of the future

    International Nuclear Information System (INIS)

    Meneley, D.A.; Dastur, A.R.

    1997-01-01

    Good neutron economy is the basis of the fuel cycle flexibility in the CANDU reactor. This paper describes the fuel cycle options available to the CANDU owner with special emphasis on resource conservation and waste management. CANDU fuel cycles with low initial fissile content operate with relatively high conversion ratio. The natural uranium cycle provides over 55 % of energy from the plutonium that is created during fuel life. Resource utilization is over 7 MWd/kg NU. This can be improved by slight enrichment (between 0.9 and 1.2 wt % U235) of the fuel. Resource utilization increases to 11 MWd/kg NU with the Slightly Enriched Uranium cycle. Thorium based cycles in CANDU operate at near-breeder efficiency. Obey provide attractive options when used with natural uranium or separated (reactor grade and weapons grade) plutonium as driver fuels. In the latter case, the energy from the U233 plus the initial plutonium content amounts to 3.4 GW(th).d/kg Pu-fissile. The same utilization is expected from the use of FBR plutonium in a CANDU thorium cycle. Extension of natural resource is achieved by the use of spent fuels in CANDU. The LWR/CANDU Tandem cycle leads to an additional 77 % of energy through the use of reprocessed LWR fuel (which has a fissile content of 1.6 wt %) in CANDU. Dry reprocessing of LWR fuel with the OREOX process (a more safeguardable alternative to the PUREX process) provides an additional 50 % energy. Uranium recovered (RU) from separation of plutonium contained in spent LWR fuel provides an additional 15 MWd/kg RU. CANDU's low fissile requirement provides the possibility, through the use of non-fertile targets, of extracting energy from the minor actinides contained in spent fuel. In addition to the resource utilization advantage described above, there is a corresponding reduction in waste arisings with such cycles. This is especially significant when separated plutonium is available as a fissile resource. (author)

  8. Nuclear plant analyzer program for Bulgaria

    International Nuclear Information System (INIS)

    Shier, W.; Kennett, R.

    1993-01-01

    An interactive nuclear plant analyzer(NPA) has been developed for use by the Bulgarian technical community in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. The current NPA includes models for a VVER-440 Model 230 and a VVER-1000 Model 320 and is operational on an IBM RISC6000 workstation. The RELAP5/MOD2 computer code has been used for the calculation of the reactor responses to the interactive commands initiated by the NPA operator. The interactive capabilities of the NPA have been developed to provide considerable flexibility in the plant actions that can be initiated by the operator. The current capabilities for both the VVER-440 and VVER-1000 models include: (1) scram initiation; (2) reactor coolant pump trip; (3) high pressure safety injection system initiation; (4) low pressure safety injection system initiation; (5) pressurizer safety valve opening; (6) steam generator relief/safety valve opening; (7) feedwater system initiation and trip; (8) turbine trip; and (9) emergency feedwater initiation. The NPA has the capability to display the results of the simulations in various forms that are determined by the model developer. Results displayed on the reactor mask are shown through the user defined, digital display of various plant parameters and through color changes that reflect changes in primary system fluid temperatures, fuel and clad temperatures, and the temperature of other metal structures. In addition, changes in the status of various components and systems can be initiated and/or displayed both numerically and graphically on the mask. This paper provides a description of the structure of the NPA, a discussion of the simulation models used for the VVER-440 and the VVER-1000, and an overview of the NPA capabilities. Typical results obtained using both simulation models will be discussed

  9. Safety enhancement concept for NPP of new generation with VVER reactors

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Kukhtevich, I.; Semashko, S.; Svetlov, S.; Solodovnikov, A.

    2004-01-01

    through a fuel pool during accidents accompanied by leaks of primary coolant. Moreover, if this is a case, no detention of corium within RPV is guaranteed during severe accidents. Anyhow, many of the useful properties of other passive systems adopted for NPP VVER-640 design may be used in high-powered NPP designs. In the course of selecting the configuration of NPP VVER-1500, it is advisable to take the latest NPP VVER-1000 designs as a basis. The design is proposed to be developed on a stage-by-stage scheme. In this context, the first stage would include the elaboration of a design of a prototype power unit on the basis of existing NPP VVER-1000 designs with the supplement of a passive containment heat removal system. Furthermore, the design shall provide for a possibility of implementing additional passive safety systems without drastic changes in the design. During further stages, it is necessary to finalize and approve a technique for assessment of probabilistic safety indices as regards inter-group common cause failures. The next step is a substantiation of implementation of additional passive safety systems and selection of their properties. Despite of the formidability of problems arising in the course of elaborating new NPP VVER-1500 designs, a holistic analysis of such designs which are under elaboration now both in Russia and abroad, as well as consideration of the latest safety requirements will create a sound basis for advanced technical approaches to safety assurance, thus allowing to reach the best safety indices in the world. (author)

  10. Nuclear fuel cycles : description, demand and supply estimates

    International Nuclear Information System (INIS)

    Gadallah, A.A.; Abou Zahra, A.A.; Hammad, F.H.

    1985-01-01

    This report deals with various nuclear fuel cycles description as well as the world demand and supply estimates of materials and services. Estimates of world nuclear fuel cycle requirements: nuclear fuel, heavy water and other fuel cycle services as well as the availability and production capabilities of these requirements, are discussed for several reactor fuel cycle strategies, different operating and under construction fuel cycle facilities in some industrialized and developed countries are surveyed. Various uncertainties and bottlenecks which are recently facing the development of some fuel cycle components are also discussed, as well as various proposals concerning fuel cycle back-end concepts. finally, the nuclear fuel cycles activities in some developing countries are reviewed with emphasis on the egyptian plans to introduce nuclear power in the country. 11 fig., 16 tab

  11. The Nuclear Fuel Cycle Information System

    International Nuclear Information System (INIS)

    1987-02-01

    The Nuclear Fuel Cycle Information System (NFCIS) is an international directory of civilian nuclear fuel cycle facilities. Its purpose is to identify existing and planned nuclear fuel cycle facilities throughout the world and to indicate their main parameters. It includes information on facilities for uranium ore processing, refining, conversion and enrichment, for fuel fabrication, away-from-reactor storage of spent fuel and reprocessing, and for the production of zirconium metal and Zircaloy tubing. NFCIS currently covers 271 facilities in 32 countries and includes 171 references

  12. Nuclear-fuel-cycle optimization: methods and modelling techniques

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1982-01-01

    This book present methods applicable to analyzing fuel-cycle logistics and optimization as well as in evaluating the economics of different reactor strategies. After an introduction to the phases of a fuel cycle, uranium cost trends are assessed in a global perspective. Subsequent chapters deal with the fuel-cycle problems faced by a power utility. The fuel-cycle models cover the entire cycle from the supply of uranium to the disposition of spent fuel. The chapter headings are: Nuclear Fuel Cycle, Uranium Supply and Demand, Basic Model of the LWR (light water reactor) Fuel Cycle, Resolution of Uncertainties, Assessment of Proliferation Risks, Multigoal Optimization, Generalized Fuel-Cycle Models, Reactor Strategy Calculations, and Interface with Energy Strategies. 47 references, 34 figures, 25 tables

  13. Determination of the Effectiveness of Control Rods in the VVER Reactor Fuel Assemblies; Determination de l'Efficacite des Barres de Reglage dans les Ensembles Combustibles du reacteur VVER; Opredelenie ehffektivnosti reguliruyushchikh sterzhnej v sborkakh reaktora VVEHR; Determinacion de la Eficacia de las Barras de Control en los Conjuntos de Elementos Combustibles del Reactor VVER

    Energy Technology Data Exchange (ETDEWEB)

    Semenov, V. N.; Lunin, G. L.; Komissarov, L. V.; Kamyshan, A. N.; Halizev, V. I.; Andrianov, G. Ja.; Voznesenskij, V. A.; Kuz' micheva, V. A.; Lebedev, V. I. [Ordena Lenina Institut Atomnoj Energii Im. I.V. Kurchatova, Moskva, SSSR (Russian Federation)

    1964-06-15

    The paper describes experiments done in homogeneous mock-ups of the fuel assemblies from the VVER Reactor (at one level of enrichment) to determine the effectiveness of absorbing systems comprising shim fuel assemblies or water cavities and of absorbing rods clad in jackets made of differing materials. The paper also gives data on some experiments that have been done in mock-ups of assemblies with differing levels of enrichment. These experiments make it possible to verify the methods used in calculation and to evaluate the prospects of using them for heterogeneous reactors. (author) [French] Le memoire decrit les experiences qui ont ete faites pour determiner l 'efficacite des absorbants contenus dans les barres de compensation, l'effet cavitaire et l 'efficacite des absorbants gaines de materiaux divers, au moyen d'assemblages homogenes de cartouches de combustible du reacteur VVER (reacteur de puissance ralenti et refroidi a l 'eau ayant le meme taux d'enrichissement. On y trouve en outre des donnees sur certaines experiences executees a l 'aide d'assemblages de cartouches de combustible taux d'enrichissement differents. Ces travaux permettent de verifier la methode de calcul et d'evaluer ses possibilites d'application aux reacteurs non homogenes. (author) [Spanish] Se describen en la memoria experimentos para determinar la eficacia de los materiales absorbentes contenidos en las barras de compensacion, el efecto de cavitacion y la eficacia de los materiales absorbentes revestidos de diversos materiales, realizados con ayuda de los conjuntos homogeneos de elementos combustibles del reactor VVER (reactor de potencia moderado y refrigerado por agua) con un solo grado de enriquecimiento. Ademas, se exponen datos sobre los experimentos efectuados con ayuda de conjuntos de grados de enriquecimientos; variados. Tales experimentos permiten verificar el metodo de calculo teorico, utilizad o y evaluar la posibilidad de aplicarlo a los reactores no homogeneos. (author

  14. Information about AER WG A on improvement, extension and validation of parametrized few-group libraries for VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Mikolas, P.

    2010-01-01

    Joint AER Working Group A on 'Improvement, extension and validation of parameterized few-group libraries for WWER-440 and WWER-1000' and AER Working Group B on 'Core design' nineteenth meeting was hosted by VUJE a. s. in Modra - Harmonia (Slovakia) during the period of 20. to 22. April 2010. There were present altogether 12 participants from 8 member organizations and 9 papers were presented (8 of them in written form). Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation. Presentations were devoted to some aspects of transport and diffusion calculations and to the benchmark dealing with WWER-1000 core periphery power tilt. Tamas Parko (co-authors Istvan Pos and Sandor Patai Szabo) described in his presentation 'Application of Discontinuity factors in C-PORCA 7 code', Radoslav Zajac (co-authors Petr Darilek and Vladimir Necas) spoke about 'Fast Reactor Nodalisation in HELIOS Code', Gabriel Farkas presented 'Calculation of Spatial Weighting Functions of Ex-Core Neutron Detectors for WWER-440 Using Monte Carlo Approach' and Daniel Sprinzl (co-authors Vaclav Krysl, Pavel Mikolas and Jiri Svarny) provided a definition of a benchmark in ' 'MIDICORE' WWER-1000 core periphery power tilt benchmark proposal'. (Author)

  15. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2009-12-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  16. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert

    2007-04-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 26 cost modules—24 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, and high-level waste.

  17. Advanced Fuel Cycle Cost Basis

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire; K. A. Williams; W. B. Boore; J. D. Smith; B. W. Dixon; M. Dunzik-Gougar; R. D. Adams; D. Gombert; E. Schneider

    2008-03-01

    This report, commissioned by the U.S. Department of Energy (DOE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the Advanced Fuel Cycle Initiative (AFCI) Program. The report describes the AFCI cost basis development process, reference information on AFCI cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for 25 cost modules—23 fuel cycle cost modules and 2 reactor modules. The cost modules were developed in the areas of natural uranium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste.

  18. MOCA, Criticality of VVER Reactor Hexagonal Fuel Assemblies

    International Nuclear Information System (INIS)

    KYNCL, Jan

    1994-01-01

    1 - Description of program or function: Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly. 2 - Method of solution: Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks: In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables. The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated. In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations. 3 - Restrictions on the complexity of the problem: In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and

  19. Fuel cycles for the 80's

    International Nuclear Information System (INIS)

    1980-01-01

    Papers presented at the American Nuclear Society's topical meeting on the fuel cycle are summarized. Present progress and goals in the areas of fuel fabrication, fuel reprocessing, spent fuel storage, accountability, and safeguards are reported. Present governmental policies which affect the fuel cycle are also discussed. Individual presentations are processed for inclusion in the Energy Data Base

  20. Fuel cycle centres

    International Nuclear Information System (INIS)

    Hagen, M.

    1977-01-01

    The concept of co-locating and integrating fuel cycle facilities at one site is discussed. This concept offers considerable advantages, especially in minimizing the amount of radioactive material to be transported on public roads. Safeguards and physical protection as relating to such an integrated system of facilities are analysed in detail, also industrial and commercial questions. An overall risk-benefit evaluation turns out to be in favour of fuel cycle centres. These centres seem to be specifically attractive with regard to the back end of the fuel cycle, including on-site disposal of radioactive wastes. The respective German approach is presented as an example. Special emphasis is given to the site selection procedures in this case. Time scale and cost for the implementation of this concept are important factors to be looked at. Since participation of governmental institutions in these centres seems to be indispensable their respective roles as compared to industry must be clearly defined. The idea of adjusting fuel cycle centres to regional rather than national use might be an attractive option, depending on the specific parameters in the region, though results of existing multinational ventures are inconclusive in this respect. Major difficulties might be expected e.g. because of different national safety regulations and standards as well as commercial conditions among partner countries. Public acceptance in the host country seems to be another stumbling block for the realization of this type of multinational facilities

  1. Nuclear fuel cycle, nuclear fuel makes the rounds: choosing a closed fuel cycle, nuclear fuel cycle processes, front-end of the fuel cycle: from crude ore to enriched uranium, back-end of the fuel cycle: the second life of nuclear fuel, and tomorrow: multiple recycling while generating increasingly less waste

    International Nuclear Information System (INIS)

    Philippon, Patrick

    2016-01-01

    France has opted for a policy of processing and recycling spent fuel. This option has already been deployed commercially since the 1990's, but will reach its full potential with the fourth generation. The CEA developed the processes in use today, and is pursuing research to improve, extend, and adapt these technologies to tomorrow's challenges. France has opted for a 'closed cycle' to recycle the reusable materials in spent fuel (uranium and plutonium) and optimise ultimate waste management. France has opted for a 'closed' nuclear fuel cycle. Spent fuel is processed to recover the reusable materials: uranium and plutonium. The remaining components (fission products and minor actinides) are the ultimate waste. This info-graphic shows the main steps in the fuel cycle currently implemented commercially in France. From the mine to the reactor, a vast industrial system ensures the conversion of uranium contained in the ore to obtain uranium oxide (UOX) fuel pellets. Selective extraction, purification, enrichment - key scientific and technical challenges for the teams in the Nuclear Energy Division (DEN). The back-end stages of the fuel cycle for recycling the reusable materials in spent fuel and conditioning the final waste-forms have reached maturity. CEA teams are pursuing their research in support of industry to optimise these processes. Multi-recycle plutonium, make even better use of uranium resources and, over the longer term, explore the possibility of transmuting the most highly radioactive waste: these are the challenges facing future nuclear systems. (authors)

  2. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    Energy Technology Data Exchange (ETDEWEB)

    Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  3. State of fuel rods spent in the VVER-1000 reactor up to a fuel burnup of 75 MW·Day/KgU

    International Nuclear Information System (INIS)

    Markov, D.; Zvir, E.; Polenok, V.; Zhitelev, V.; Strozhuk, A.; Volkova, I.

    2011-01-01

    The presented material contains the data on change in form, corrosion state and mechanical properties of fuel rod claddings, change in fuel structure and release of gaseous fission products (GFP) under the cladding. The results of PIEs of the VVER-1000 fuel rods with the high burnup of fuel (average value is 72.3 MW·day/kgU and maximum is 75 MW·day/kgU) carried out in JSC 'SSC RIAR' show that by the basic operational characteristics the lifetime of fuel rods with such burnup of fuel is not exhausted. The state of fuel rods is characterized by following key parameters. The fuel-to-cladding gap on the most part of the fuel meat is absent. With the burnup growth, diameter of the fuel rod increases due to fuel meat swelling. In so doing, the reverse strain achieves the values of 0.40-0.47 %. Ridges on the cladding are formed practically along the entire length of the fuel meat, average height of ridges makes up 25 μm, maximum - 40 μm. At burnups exceeding 55 MW·day/kgU, the rate of the fuel rod elongation is less than at low and average burnups. So if within a burnup range of 20-55 MW·day/kgU, the rate of the fuel rod elongation makes up about 0.330mm per 1 MW·day/kgU, at burnups exceeding 55 MW·day/kgU it is only 0.085mm per 1 MW·day/kgU. Corrosion state of the claddings of fuel rods with high burnup of fuel is satisfactory. The oxide film, as a rule, is uniform, dense, without cracks and exfoliation, its thickness on the external surface does not exceed 13 μm, while on the internal surface - 15 μm. Hydrogenation is insignificant, mass fraction of hydrogen does not exceed 0.01 %. Interaction of fuel rods with spacer grids does not result in significant fretting-corrosion. Based of the results of tests, short-term mechanical properties of the claddings of fuel rods with high burnup of fuel remain at high level. The state of fuel is characterized by absence of the fuel-to-cladding gap on the most part of the fuel meat, fuel is tightly fixed to the cladding

  4. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.-P.

    1995-01-01

    Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

  5. Development of FR fuel cycle in japan (1) development scope of fuel cycle technology

    International Nuclear Information System (INIS)

    Nakamura, H.; Funasaka, H.; Namekawa, T.

    2008-01-01

    A fast reactor (FR) cycle has a potential to realize a sustainable energy supply system that is harmonized with environment by fully recycling both uranium (U) and transuranium (TRU) elements. In Japan, a Feasibility Study on Commercialized FR Cycle Systems (FS) was launched in July 1999, and through two different study phases, a final report was presented in 2006. As a result of FS, a combined system of sodium-cooled FR with mixed-oxide (MOX) fuel, advanced aqueous reprocessing and simplified pelletizing fuel fabrication was considered to be most promising for commercialization. The advanced aqueous reprocessing system, which is called the New Extraction system for TRU recovery (NEXT), consists of a U crystallization process for the bulk of U recovery, a simplified solvent extraction process for residual U, plutonium (Pu) and neptunium (Np) without Pu partitioning and purification, and a process for recovering americium (Am) and curium (Cm) from the raffinate. The ratio of Pu/U concentration in the mother solution after crystallization is adequate for MOX fuel fabrication, and thus complicated powder mixing processes for adjusting Pu content in MOX fuel can be eliminated in the subsequent simplified fuel fabrication system. In this system, lubricant-mixing process can also be eliminated by adopting the advanced technology in which lubricant is coated on the inner surface of a die before fuel powder supply. Such a simplification could help us overcoming the difficulty to treat MA bearing fuel powders in a hot cell. Ministry of Education, Culture, Sports, Science and Technology (MEXT) reviewed these results of FS in 2006 and identified the most promising FR cycle concept proposed in the FS phase II study as a mainline choice for commercialization. According to such a governmental assessment, R and D activities of FR cycle systems were decided to be concentrated mainly to the innovative technology development for the mainline concept. The stage of R and D project was

  6. Advanced nuclear fuel cycles activities in IAEA

    International Nuclear Information System (INIS)

    Nawada, H.P.; Ganguly, C.

    2007-01-01

    Full text of publication follows. Of late several developments in reprocessing areas along with advances in fuel design and robotics have led to immense interest in partitioning and transmutation (P and T). The R and D efforts in the P and T area are being paid increased attention as potential answers to ever-growing issues threatening sustainability, environmental protection and non-proliferation. Any fuel cycle studies that integrate partitioning and transmutation are also known as ''advanced fuel cycles'' (AFC), that could incinerate plutonium and minor actinide (MA) elements (namely Am, Np, Cm, etc.) which are the main contributors to long-term radiotoxicity. The R and D efforts in developing these innovative fuel cycles as well as reactors are being co-ordinated by international initiatives such as Innovative Nuclear Power Reactors and Fuel Cycles (INPRO), the Generation IV International Forum (GIF) and the Global Nuclear Energy Partnership (GENP). For these advanced nuclear fuel cycle schemes to take shape, the development of liquid-metal-cooled reactor fuel cycles would be the most essential step for implementation of P and T. Some member states are also evaluating other concepts involving the use of thorium fuel cycle or inert-matrix fuel or coated particle fuel. Advanced fuel cycle involving novel partitioning methods such as pyrochemical separation methods to recover the transuranic elements are being developed by some member states which would form a critical stage of P and T. However, methods that can achieve a very high reduction (>99.5%) of MA and long-lived fission products in the waste streams after partitioning must be achieved to realize the goal of an improved protection of the environment. In addition, the development of MA-based fuel is also an essential and crucial step for transmutation of these transuranic elements. The presentation intends to describe progress of the IAEA activities encompassing the following subject-areas: minimization of

  7. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Patarin, L.

    2002-01-01

    This book treats of the different aspects of the industrial operations linked with the nuclear fuel, before and after its use in nuclear reactors. The basis science of this nuclear fuel cycle is chemistry. Thus a recall of the elementary notions of chemistry is given in order to understand the phenomena involved in the ore processing, in the isotope enrichment, in the fabrication of fuel pellets and rods (front-end of the cycle), in the extraction of recyclable materials (residual uranium and plutonium), and in the processing and conditioning of wastes (back-end of the fuel cycle). Nuclear reactors produce about 80% of the French electric power and the Cogema group makes 40% of its turnover at the export. Thus this book contains also some economic and geopolitical data in order to clearly position the stakes. The last part, devoted to the management of wastes, presents the solutions already operational and also the research studies in progress. (J.S.)

  8. Radkowsky Thorium Fuel Project

    International Nuclear Information System (INIS)

    Todosow, Michael

    2006-01-01

    In the early/mid 1990's Prof. Alvin Radkowsky, former chief scientist of the U.S. Naval Reactors program, proposed an alternate fuel concept employing thorium-based fuel for use in existing/next generation pressurized water reactors (PWRs). The concept was based on the use of a 'seed-blanket-unit' (SBU) that was a one-for-one replacement for a standard PWR assembly with a uranium-based central 'driver' zone, surrounded by a 'blanket' zone containing uranium and thorium. Therefore, the SBU could be retrofit without significant modifications into existing/next generation PWRs. The objective was to improve the proliferation and waste characteristics of the current once-through fuel cycle. The objective of a series of projects funded by the Initiatives for Proliferation Prevention program of the U.S. Department of Energy (DOE-IPP) - BNL-T2-0074,a,b-RU 'Radkowsky Thorium Fuel (RTF) Concept' - was to explore the characteristics and potential of this concept. The work was performed under several BNL CRADAs (BNL-C-96-02 and BNL-C-98-15) with the Radkowsky Thorium Power Corp./Thorium Power Inc. and utilized the technical and experimental capabilities in the Former Soviet Union (FSU) to explore the potential of this concept for implementation in Russian pressurized water reactors (VVERs), and where possible, also generate data that could be used for design and licensing of the concept for Western PWRs. The Project in Russia was managed by the Russian Research Center-?'Kurchatov Institute' (RRC-KI), and included several institutes (e.g., PJSC 'Electrostal', NPO 'LUCH' (Podolsk), RIINM (Bochvar Institute), GAN RF (Gosatomnadzor), Kalininskaja NPP (VVER-1000)), and consisted of the following phases: Phase-1 ($550K/$275K to Russia): The objective was to perform an initial review of all aspects of the concept (design, performance, safety, implementation issues, cost, etc.) to confirm feasibility/viability and identify any 'show-stoppers'; Phase-2 ($600K/$300K to Russia

  9. Fuel cycle and waste management: A perspective from British nuclear fuels plc

    International Nuclear Information System (INIS)

    Holmes, R.G.G.; Fairhall, G.A.; Robbins, R.A.

    1996-01-01

    The phrase fuel cycle and waste management implies two separate and distinct activities. British Nuclear Fuels plc (BNFL) has adopted a holistic approach to the fuel cycle that integrates the traditional fuel cycle activities of conversion to uranium hexafluoride, fuel fabrication, power generation, and reprocessing with waste arisings, its subsequent treatment, and disposal

  10. A fuel cycle cost study with HEU and LEU fuels

    International Nuclear Information System (INIS)

    Matos, J.E.; Freese, K.E.

    1985-01-01

    Fuel cycle costs are compared for a range of 235 U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  11. A fuel cycle cost study with HEU and LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [Argonne National Laboratory, Argonne, IL (United States)

    1985-07-01

    Fuel cycle costs are compared for a range of {sup 235}U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors. (author)

  12. Characteristics of fuel cycle waste

    International Nuclear Information System (INIS)

    Aquilina, C.A.; Everette, S.E.

    1982-01-01

    The Low-Level Waste Management System started in 1979 to describe and model the existing commercial low-level waste management system. The system description produced is based on the identification of the different elements making up both the fuel and non-fuel cycle and their relationships to each other. A systems model based on the system description can accurately reflect the flow of low-level waste from generator to disposal site and is only limited by the reliability of the information it uses. For both the fuel cycle and non-fuel cycle large quantities of information is required in order to allow the system to operate at its full potential. Work is ongoing to collect this information. Significant progress is being made in the fuel cycle area primarily because the majority of fuel cycle low-level radioactive waste is produced by commercial power reactor plant operations. The Low-Level Waste Management system is only beginning to derive the benefits to be obtained from an accurate low-level waste management information system. As data is verified and analyzed, results on a national as well as individual organization level will be gained. Comparisons to previous studies will be made. Accurate projections of waste volumes and activities to be produced, projected impacts of waste streams of treatment or management changes are only examples of information to be produced. 1 figure, 1 table

  13. Back end of an enduring fuel cycle

    International Nuclear Information System (INIS)

    Pillay, K.K.S.

    1998-03-01

    An enduring nuclear fuel cycle is an essential part of sustainable consumption, the process whereby world's riches are consumed in a responsible manner so that future generations can continue to enjoy at least some of them. In many countries, the goal of sustainable development has focused attention on the benefits of nuclear technologies. However, sustenance of the nuclear fuel cycle is dependent on sensible management of all the resources of the fuel cycle, including energy, spent fuels, and all of its side streams. The nuclear fuel cycle for energy production has suffered many traumas since the mid seventies. The common basis of technologies producing nuclear explosives and consumable nuclear energy has been a preoccupation for some, predicament for others, and a perception problem for many. It is essential to reestablish a reliable back end of the nuclear fuel cycle that can sustain the resource requirements of an enduring full cycle. This paper identifies some pragmatic steps necessary to reverse the trend and to maintain a necessary fuel cycle option for the future

  14. Fuel cycle based safeguards

    International Nuclear Information System (INIS)

    De Montmollin, J.M.; Higinbotham, W.A.; Gupta, D.

    1985-07-01

    In NPT safeguards the same model approach and absolute-quantity inspection goals are applied at present to all similar facilities, irrespective of the State's fuel cycle. There is a continuing interest and activity on the part of the IAEA in new NPT safeguards approaches that more directly address a State's nuclear activities as a whole. This fuel cycle based safeguards system is expected to a) provide a statement of findings for the entire State rather than only for individual facilities; b) allocate inspection efforts so as to reflect more realistically the different categories of nuclear materials in the different parts of the fuel cycle and c) provide more timely and better coordinated information on the inputs, outputs and inventories of nuclear materials in a State. (orig./RF) [de

  15. Fuel cycle parameters for strategy studies

    International Nuclear Information System (INIS)

    Archinoff, G.H.

    1979-05-01

    This report summarizes seven fuel cycle parameters (efficiency, specific power, burnup, equilibrium net fissile feed, equilibrium net fissile surplus, first charge fissile content, and whether or not fuel reprocessing is required) to be used in long-term strategy analyses of fuel cycles based on natural UO 2 , low enriched uranium, mixed oxides, plutonium topped thorium, uranium topped thorium, and the fast breeder oxide cycle. (LL)

  16. Alternative fuel cycles and non-proliferation aspects

    International Nuclear Information System (INIS)

    Kessler, G.

    1980-10-01

    The most important physical characteristics of the U/Pu and the Th/U fuel cycles and the technical data of the most significant converter reactors operating with Th/U fuel are outlined in the report. Near breeders as well as breeders with a thermal neutron spectrum are briefly discussed, and the potential of breeders with fast neutron spectra in the Th/U fuel is outlined. The essential criteria for the comparison of the alternative fuel cycles with the reference Pu/U cycle are the consumption of natural uranium, the numbers of U-233 producing and U-233 consuming converter reactors and the amounts of fission material transported and handled within the fuel cycle (reprocessing, refabrication). Although the alternative U/Th fuel cycles are feasible with some advantages and some disadvantages as compared to the reference U/Pu cycle, not much experience has so far been gathered with pilot plants of the fuel cycle. The respective status in reprocessing, refabrication and waste disposal is briefly discussed. Finally, a comparison of the risk potential inherent in secular storage is presented and questions of resistance to proliferation and of safeguards of the U/Th cycle are discussed

  17. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    International Nuclear Information System (INIS)

    Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.

    1997-01-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab

  18. Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code

    Energy Technology Data Exchange (ETDEWEB)

    Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)

    1997-08-01

    The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.

  19. Globalization of the nuclear fuel cycle impact of developments on fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Durpel, L.; Bertel, E. [OCDE-NEA, Nuclear Development Div., 92 - Issy-les-Moulineaux (France)

    1999-07-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the de-regulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to compete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economical perspective including environmental and social considerations. (authors)

  20. Globalisation of the nuclear fuel cycle - impact of developments on fuel management

    International Nuclear Information System (INIS)

    Durpel, L. van den; Bertel, E.

    2000-01-01

    Nuclear energy will have to cope more and more with a rapid changing environment due to economic competitive pressure and the deregulatory progress. In current economic environment, utilities will have to focus strongly on the reduction of their total generation costs, covering the fuel cycle costs, which are only partly under their control. Developments in the fuel cycle will be in the short-term rather evolutionary addressing the current needs of utilities. However, within the context of sustainable development and more and more inclusion of externalities in energy generation costs, more performing developments in the fuel cycle could become important and feasible. A life-cycle design approach of the fuel cycle will be requested in order to cover all factors in order to decrease significantly the nuclear energy generation cost to complete with other alternative fuels in the long-term. This paper will report on some of the trends one could distinguish in the fuel cycle with emphasis on cost reduction. OECD/NEA is currently conducting a study on the fuel cycle aiming to assess current and future nuclear fuel cycles according to the potential for further improvement of the full added-value chain of these cycles from a mainly technological and economic perspective including environmental and social considerations. (orig.) [de

  1. ITER fuel cycle systems layout

    International Nuclear Information System (INIS)

    Kveton, O.K.

    1990-10-01

    The ITER fuel cycle building (FCB) will contain the following systems: fuel purification - permeator based; fuel purification - molecular sieves; impurity treatment; waste water storage and treatment; isotope separation; waste water tritium extraction; tritium extraction from solid breeder; tritium extraction from test modules; tritium storage, shipping and receiving; tritium laboratory; atmosphere detritiation systems; fuel cycle control centre; tritiated equipment maintenance space; control maintenance space; health physics laboratory; access, access control and facilities. The layout of the FCB and the requirements for these systems are described. (10 figs.)

  2. WWER-440 reactor thermal power increase. Up-to-date approaches to substantiation of the core heat-engineering reliability

    International Nuclear Information System (INIS)

    Vasilchenko, I.; Lushin, V.; Zubtsov, D.

    2006-01-01

    Increasing the Units power is an urgent problem for nuclear power plants with WWER-440 reactors. Improving the fuel assembly designs and calculated codes creates all prerequisites to fulfil this purpose. The decrease in the core power peaking is reached by using the profiled fuel assemblies, burnable absorber integrated into the fuel, the FA with the modernized interface attachment, modern calculated codes that allows to reduce conservatism of the RP safety substantiation. A wide spectrum of experimental study of behaviour of the fuel having reached burn-up (50-60) MW days / kg U under the transients and accident conditions was carried out, the post-irradiated examination of the fuel assemblies, fuel rods and fuel pellets with four and five annual operating fuel cycle were performed as well and confirmed the high reliability of the fuel, the presence of large margins of the fuel stack state that contributes to reactor operation at the increased power. The results of the carried out experiments on implementing the five and six annual fuel cycles show that the limiting fuel state as to its serviceability in the WWER-440 reactors is far from being reached. Presently there is an experience of the increased power operation of Kola NPP, Units 1, 2, 4 and Rovno NPP, Unit 2. The Loviisa NPP Units are operated at 109 % power. The Russian experts had gained an experience in substantiating the core operation at 108 % power for Paks NPP, Unit 4. In this paper the additional conditions for increasing the power of the Kola NPP, Units 1 and 2 and the main results of substantiation of increase in power of the Paks NPP, Unit 4 up to 1485 MW are presented in details

  3. Estimation of material degradation of VVER-1000 baffle

    Science.gov (United States)

    Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

    2017-09-01

    The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  4. The evolving nuclear fuel cycle

    International Nuclear Information System (INIS)

    Gale, J.D.; Hanson, G.E.; Coleman, T.A.

    1993-01-01

    Various economics and political pressures have shaped the evolution of nuclear fuel cycles over the past 10 to 15 yr. Future trends will no doubt be similarly driven. This paper discusses the influences that long cycles, high discharge burnups, fuel reliability, and costs will have on the future nuclear cycle. Maintaining the economic viability of nuclear generation is a key issue facing many utilities. Nuclear fuel has been a tremendous bargain for utilities, helping to offset major increases in operation and maintenance (O ampersand M) expenses. An important factor in reducing O ampersand M costs is increasing capacity factor by eliminating outages

  5. Introducing advanced nuclear fuel cycles in Canada

    International Nuclear Information System (INIS)

    Duret, M.F.

    1978-05-01

    The ability of several different advanced fuel cycles to provide energy for a range of energy growth scenarios has been examined for a few special situations of interest in Canada. Plutonium generated from the CANDU-PHW operating on natural uranium is used to initiate advanced fuel cycles in the year 2000. The four fuel cycles compared are: 1) natural uranium in the CANDU-PHW; 2) high burnup thorium cycle in the CANDU-PHW; 3) self-sufficient thorium cycle in the CANDU-PHW; 4) plutonium-uranium cycle in a fast breeder reactor. The general features of the results are quite clear. While any plutonium generated prior to the introduction of the advanced fuel cycle remains, system requirements for natural uranium for each of the advanced fuel cycles are the same and are governed by the rate at which plants operating on natural uranium can be retired. When the accumulated plutonium inventory has been entirely used, natural uranium is again required to provide inventory for the advanced fuel cycle reactors. The time interval during which no uranium is required varies only from about 25 to 40 years for both thorium cycles, depending primarily on the energy growth rate. The breeder does not require the entire plutonium inventory produced and so would call for less processing of fuel from the PHW reactors. (author)

  6. Advanced nuclear fuel cycles and radioactive waste management

    International Nuclear Information System (INIS)

    2006-01-01

    This study analyses a range of advanced nuclear fuel cycle options from the perspective of their effect on radioactive waste management policies. It presents various fuel cycle options which illustrate differences between alternative technologies, but does not purport to cover all foreseeable future fuel cycles. The analysis extends the work carried out in previous studies, assesses the fuel cycles as a whole, including all radioactive waste generated at each step of the cycles, and covers high-level waste repository performance for the different fuel cycles considered. The estimates of quantities and types of waste arising from advanced fuel cycles are based on best available data and experts' judgement. The effects of various advanced fuel cycles on the management of radioactive waste are assessed relative to current technologies and options, using tools such as repository performance analysis and cost studies. (author)

  7. Simulation of Cycle-to-Cycle Variation in Dual-Fuel Engines

    KAUST Repository

    Jaasim, Mohammed

    2017-03-13

    Standard practices of internal combustion (IC) engine experiments are to conduct the measurements of quantities averaged over a large number of cycles. Depending on the operating conditions, the cycle-to-cycle variation (CCV) of quantities, such as the indicated mean effective pressure (IMEP) are observed at different levels. Accurate prediction of CCV in IC engines is an important but challenging task. Computational fluid dynamics (CFD) simulations using high performance computing (HPC) can be used effectively to visualize such 3D spatial distributions. In the present study, a dual fuel large engine is considered, with natural gas injected into the manifold accompanied with direct injection of diesel pilot fuel to trigger ignition. Multiple engine cycles in 3D are simulated in series as in the experiments to investigate the potential of HPC based high fidelity simulations to accurately capture the cycle to cycle variation in dual fuel engines. Open cycle simulations are conducted to predict the combined effect of the stratification of fuel-air mixture, temperature and turbulence on the CCV of pressure. The predicted coefficient of variation (COV) of pressure compared to the results from closed cycle simulations and the experiments.

  8. The economics of thorium fuel cycles

    International Nuclear Information System (INIS)

    James, R.A.

    1978-01-01

    The individual cost components and the total fuel cycle costs for natural uranium and thorium fuel cycles are discussed. The thorium cycles are initiated by using either enriched uranium or plutonium. Subsequent thorium cycles utilize recycled uranium-233 and, where necessary, either uranium-235 or plutonium as topping. A calculation is performed to establish the economic conditions under which thorium cycles are economically attractive. (auth)

  9. Innovation in the fuel cycle industry

    International Nuclear Information System (INIS)

    Lamorlette, Guy

    1998-01-01

    The fuel cycle industry will have to adapt to the production of new fuel and in the same time will have to improve its performance. Innovation will be a key factor of success. Innovation must be driven by the needs of the fuel cycle industry to achieve. The fuel cycle requirement of tomorrow, Innovative processes for mining high grade uranium, Innovative enrichment process, Sorting the pellets at Melox plant, Innovation in action, and Innovative waste management in la Hague are presented. A number of innovative solutions are already implemented and are in action on industrial facilities. As problems are becoming more and more tough to address, international cooperation will be required. The fuel cycle industry, as a part of the nuclear power industry, is committed to seek improvements through performance upgrade and innovation. (Cho. G. S.). 10 refs., 4 figs

  10. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-01-01

    The procurement and preparation of fuel for nuclear power reactors, followed by its recovery, processing and management subsequent to reactor discharge, are frequently referred to as the ''front end'' and ''back end'' of the nuclear fuel cycle. The facilities associated with these activities have an extensive and well-documented safety record accumulated over the past 50 years by technical experts and safety authorities. This information has enabled an in-depth analysis of the complete fuel cycle. Preceded by two previous editions in 1981 and 1993, this new edition of the Safety of the Nuclear Fuel Cycle represents the most up-to-date analysis of the safety aspects of the nuclear fuel cycle. It will be of considerable interest to nuclear safety experts, but also to those wishing to acquire extensive information about the fuel cycle more generally. (author)

  11. The safety of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    2005-10-01

    The procurement and preparation of fuel for nuclear power reactors, followed by its recovery, processing and management subsequent to reactor discharge, are frequently referred to as the 'front end' and 'back end' of the nuclear fuel cycle. The facilities associated with these activities have an extensive and well-documented safety record accumulated over the past 50 years by technical experts and safety authorities. This information has enabled an in-depth analysis of the complete fuel cycle. Preceded by two previous editions in 1981 and 1993, this new edition of The Safety of the Nuclear Fuel Cycle represents the most up-to-date analysis of the safety aspects of the nuclear fuel cycle. It will be of considerable interest to nuclear safety experts, but also to those wishing to acquire extensive information about the fuel cycle more generally. (author)

  12. Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

    International Nuclear Information System (INIS)

    Forsberg, C.; Miller, W.F.

    2013-01-01

    The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that states will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state

  13. Compact simulators for WWER-440 type nuclear power plants

    International Nuclear Information System (INIS)

    Vegh, E.; Janosy, J.S.

    1991-09-01

    This paper describes a compact simulator for VVER-440 type plants. Up till now three simulators have been delivered: to Paks Nuclear Power Plant (Hungary), and to Kola and Rovno NPPs (Soviet Union). In this compact simulator the modelling complexity of the plant is almost similar to that of a full-scope one apart from the Control Room being replaced by a Control Desk and four colour graphic display units. The simulation of the plant covers the whole operating range: from cold shut-down state to nominal power level. The simulator contains up to 32 different initial conditions. Moreover, every 5 minutes the simulator produces so called 'snapshots', i.e. disc images of the Data Base. Using these snapshots the instructor can go back in time (called backtracking) in order to repeat some previously passed events

  14. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  15. Measures for ensuring hydrogen fire and explosion safety for VVER-440/230

    International Nuclear Information System (INIS)

    Bezlepkin, V.; Semashko, S.; Svetlov, S.; Sidorov, V.; Ivkov, I.; Krylov, Yu.; Kukhtevich, V.

    2004-01-01

    This paper deals with the findings of calculation analysis as regards the release of mass, energy and hydrogen during beyond-design-basis accident (BDBA) at Kola NPP equipped with VVER-440 reactor (B-230 design) and in respect of distribution of hydrogen throughout NPP tight compartments. The analysis figures out the number and locations of passive catalytic hydrogen recombiners and of the sensors of the hydrogen concentration monitoring system. In order to prove the hydrogen safety of the design, it has been necessary to review accidents accompanied by maximum emissions (both peak and integral ones) of hydrogen into the tight area. During design-basis accident (DBA), no steam/zirconium reactions occur in the reactor core. Out of BDBA, the severe accidents with damage to the core accompanied oxidative reactions between zirconium and steel with emission of hydrogen are regarded as the most dangerous ones. Assessment of additional hydrogen sources shows that the contribution of such sources to the total amount of hydrogen that may emit during a severe accident is insignificant. Calculations have been made for the following scenarios of severe accidents, which seem to be the most important in terms of hydrogen safety analysis: - 20 mm leak from the primary circuit in combination with a failure of the emergency makeup system; - 500 mm PCP rupture in the vicinity of reactor inlet branch with bi-lateral leakage of coolant. Releases of mass and energy during the aforesaid scenarios, changes of medium parameters within the tight compartments and analysis of possible fire conditions have been analyzed by means of Russian computer codes RATEG/SVECHA/HEFEST, KUPOL-M and LIMITS. The said analysis shows that the large break accident (500 mm), i.e. PCP rupture in the vicinity of the reactor branch with bi-lateral leakage of coolant is of the keen interest in terms of hydrogen safety. This accident typifies powerful short-term release of hydrogen at a significantly lesser

  16. High-conversion HTRs and their fuel cycle

    International Nuclear Information System (INIS)

    Gutmann, H.; Hansen, U.; Larsen, H.; Price, M.S.T.

    1976-01-01

    The high-temperature reactors using graphite as structural core material and helium as coolant represent thermal reactor designs with a very high degree of neutron economy which, when using the thorium fuel cycle, offer, at least theoretically, the possibility of thermal breeding. Though this was already known from previous studies, the economic climate at that time was such that the achievement of high conversion ratios conflicted with the incentive for low fuel cycle costs. Consequently, thorium cycle conversion ratios of around 0.6 were found optimum, and the core and fuel element layout followed from the economic ground rules. The recent change in attitude, brought about partly by the slow process of realization of the limits to the earth's accessible high-grade uranium ore resources and more dramatically by the oil crisis, makes it necessary to concentrate attention again on the high conversion fuel cycles. This report discusses the principles of the core design and the fuel cycle layout for High Conversion HTRs (HCHTRs). Though most of the principles apply equally to HTRs of the pebble-bed and the prismatic fuel element design types, the paper concentrates on the latter. Design and fuel cycle strategies for the full utilization of the high conversion potential are compared with others that aim at easier reprocessing and the ''environmental'' fuel cycle. The paper concludes by discussing operating and fuel cycle characteristics and economics of HCHTRs, and how the latter impinge on the allowable price for uranium ore and the available uranium resources. (author)

  17. Fuel cycle cost analysis on molten-salt reactors

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro

    1976-01-01

    An evaluation is made of the fuel cycle costs for molten-salt reactors (MSR's), developed at Oak Ridge National Laboratory. Eight combinations of conditions affecting fuel cycle costs are compared, covering 233 U-Th, 235 U-Th and 239 Pu-Th fuels, with and without on-site continuous fuel reprocessing. The resulting fuel cycle costs range from 0.61 to 1.18 mill/kWh. A discussion is also given on the practicability of these fuel cycles. The calculations indicate that somewhat lower fuel cycle costs can be expected from reactor operation in converter mode on 235 U make-up with fuel reprocessed in batches every 10 years to avoid fission product precipitation, than from operation as 233 U-Th breeder with continuous reprocessing. (auth.)

  18. Primary water chemistry of VVERs-operating experience

    International Nuclear Information System (INIS)

    Kysela, Jan; Zmitko, Milan; Petrecky, Igor

    1998-01-01

    VVER units are operated in mixed boron-potassium-ammonia water chemistry. Several modifications of the water chemistry, differing in boron-potassium co-ordination and in the way how hydrogen concentration is produced and maintain in the coolant, is used. From the operational experience point of view VVER units do not show any significant problems connected with the primary coolant chemistry. The latest results indicate that dose rate levels are slowly returning to the former ones. An improvement of the radiation situation observed last two years is supported by the surface activity measurements. However, the final conclusion on the radiation situation can be made only after evaluation of the several following cycles. Further investigation is also needed to clarify a possible effect of modified water chemistry and shut-down chemistry on radioactivity build-up and dose rate level at Dukovany units. Structure materials composition has a significant effect on radiation situation in the units. It concerns mainly of cobalt content in SG material. There is no clear evidence of possible effect of the SG shut-down regimes on the radiation situation in the units even if the dose rate and surface activity data show wide spread for the individual reactor loops. (S.Y.)

  19. Nuclear Fuel Cycle Evaluation and Real Options

    Directory of Open Access Journals (Sweden)

    L. Havlíček

    2008-01-01

    Full Text Available The first part of this paper describes the nuclear fuel cycle. It is divided into three parts. The first part, called Front-End, covers all activities connected with fuel procurement and fabrication. The middle part of the cycle includes fuel reload design activities and the operation of the fuel in the reactor. Back-End comprises all activities ensuring safe separation of spent fuel and radioactive waste from the environment. The individual stages of the fuel cycle are strongly interrelated. Overall economic optimization is very difficult. Generally, NPV is used for an economic evaluation in the nuclear fuel cycle. However the high volatility of uranium prices in the Front-End, and the large uncertainty of both economic and technical parameters in the Back-End, make the use of NPV difficult. The real option method is able to evaluate the value added by flexibility of decision making by a company under conditions of uncertainty. The possibility of applying this method to the nuclear fuel cycle evaluation is studied. 

  20. Analysis of environmental friendliness of DUPIC fuel cycle

    International Nuclear Information System (INIS)

    Ko, Won Il; Kim, Ho Dong

    2001-07-01

    Some properties of irradiated DUPIC fuels are compared with those of other fuel cycles. It was indicated that the toxicity of the DUPIC option based on 1 GWe-yr is much smaller than those of other fuel cycle options, and is just about half the order of magnitude of other fuel cycles. From the activity analysis of 99 Tc and 237 Np, which are important to the long-term transport of fission products stored in geologic media, the DUPIC option, was being contained only about half of those other options. It was found from the actinide content estimation that the MOX option has the lowest plutonium arising based on 1 GWe-year and followed by the DUPIC option. However, fissile Pu content generated in the DUPIC fuel was the lowest among the fuel cycle options. From the analysis of radiation barrier in proliferation resistance aspect, the fresh DUPIC fuel can play a radiation barrier part, better than CANDU spent fuels as well as fresh MOX fuel. It is indicated that the DUPIC fuel cycle has the excellent resistance to proliferation, compared with an existing reprocessing option and CANDU once-through option. In conclusions, DUPIC fuel cycle would have good properties on environmental effect and proliferation resistance, compared to other fuel cycle cases

  1. Developing a computational tool for predicting physical parameters of a typical VVER-1000 core based on artificial neural network

    International Nuclear Information System (INIS)

    Mirvakili, S.M.; Faghihi, F.; Khalafi, H.

    2012-01-01

    Highlights: ► Thermal–hydraulics parameters of a VVER-1000 core based on neural network (ANN), are carried out. ► Required data for ANN training are found based on modified COBRA-EN code and then linked each other using MATLAB software. ► Based on ANN method, average and maximum temperature of fuel and clad as well as MDNBR of each FA are predicted. -- Abstract: The main goal of the present article is to design a computational tool to predict physical parameters of the VVER-1000 nuclear reactor core based on artificial neural network (ANN), taking into account a detailed physical model of the fuel rods and coolant channels in a fuel assembly. Predictions of thermal characteristics of fuel, clad and coolant are performed using cascade feed forward ANN based on linear fission power distribution and power peaking factors of FAs and hot channels factors (which are found based on our previous neutronic calculations). A software package has been developed to prepare the required data for ANN training which applies a modified COBRA-EN code for sub-channel analysis and links the codes using the MATLAB software. Based on the current estimation system, five main core TH parameters are predicted, which include the average and maximum temperatures of fuel and clad as well as the minimum departure from nucleate boiling ratio (MDNBR) for each FA. To get the best conditions for the considered ANNs training, a comprehensive sensitivity study has been performed to examine the effects of variation of hidden neurons, hidden layers, transfer functions, and the learning algorithms on the training and simulation results. Performance evaluation results show that the developed ANN can be trained to estimate the core TH parameters of a typical VVER-1000 reactor quickly without loss of accuracy.

  2. Overview of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Knief, R.A.

    1978-01-01

    The nuclear fuel cycle is substantially more complicated than the energy production cycles of conventional fuels because of the very low abundance of uranium 235, the presence of radioactivity, the potential for producing fissile nuclides from irradiation, and the risk that fissile materials will be used for nuclear weapons. These factors add enrichment, recycling, spent fuel storage, and safeguards to the cycle, besides making the conventional steps of exploration, mining, processing, use, waste disposal, and transportation more difficult

  3. IFR fuel cycle--pyroprocess development

    International Nuclear Information System (INIS)

    Laidler, J.J.; Miller, W.E.; Johnson, T.R.; Ackerman, J.P.; Battles, J.E.

    1992-01-01

    The Integral Fast Reactor (IFR) fuel cycle is based on the use of a metallic fuel alloy, with nominal composition U-2OPu-lOZr. In its present state of development, this fuel system offers excellent high-burnup capabilities. Test fuel has been carried to burnups in excess of 20 atom % in EBR-II irradiations, and to peak burnups over 15 atom % in FFTF. The metallic fuel possesses physical characteristics, in particular very high thermal conductivity, that facilitate a high degree of passive inherent safety in the IFR design. The fuel has been shown to provide very large margins to failure in overpower transient events. Rapid overpower transient tests carried out in the TREAT reactor have shown the capability to withstand up to 400% overpower conditions before failing. An operational transient test conducted in EBR-II at a power ramp rate of 0.1% per second reached its termination point of 130% of normal power without any fuel failures. The IFR metallic fuel also exhibits superior compatibility with the liquid sodium coolant. Equally as important as the performance advantages offered by the use of metallic fuel is the fact that this fuel system permits the use of an innovative reprocessing method, known as ''pyroprocessing,'' featuring fused-salt electrorefining of the spent fuel. Development of the IFR pyroprocess has been underway at the Argonne National Laboratory for over five years, and great progress has been made toward establishing a commercially-viable process. Pyroprocessing offers a simple, compact means for closure of the fuel cycle, with anticipated significant savings in fuel cycle costs

  4. High conversion HTRs and their fuel cycle

    International Nuclear Information System (INIS)

    Gutmann, H.; Hansen, U.; Larsen, H.; Price, M.S.T.

    1975-01-01

    This report discusses the principles of the core design and the fuel cycle layout for High Conversion HTRs (HCHTRs). Though most of the principles apply equally to HTRs of the pebble-bed and the prismatic fuel element design types, the paper concentrates on the latter. Design and fuel cycle strategies for the full utilisation of the high conversion potential are compared with others that aim at easier reprocessing and the 'environmental' fuel cycle. The paper concludes by discussing operating and fuel cycle characteristics and economics of HCHTRs, and how the latter impinge on the allowable price for uranium ore and the available uranium resources. (orig./UA) [de

  5. Romanian nuclear fuel cycle development

    International Nuclear Information System (INIS)

    Rapeanu, S.N.; Comsa, Olivia

    1998-01-01

    Romanian decision to introduce nuclear power was based on the evaluation of electricity demand and supply as well as a domestic resources assessment. The option was the introduction of CANDU-PHWR through a license agreement with AECL Canada. The major factors in this choice have been the need of diversifying the energy resources, the improvement the national industry and the independence of foreign suppliers. Romanian Nuclear Power Program envisaged a large national participation in Cernavoda NPP completion, in the development of nuclear fuel cycle facilities and horizontal industry, in R and D and human resources. As consequence, important support was being given to development of industries involved in Nuclear Fuel Cycle and manufacturing of equipment and nuclear materials based on technology transfer, implementation of advanced design execution standards, QA procedures and current nuclear safety requirements at international level. Unit 1 of the first Romanian nuclear power plant, Cernavoda NPP with a final profile 5x700 Mw e, is now in operation and its production represents 10% of all national electricity production. There were also developed all stages of FRONT END of Nuclear Fuel Cycle as well as programs for spent fuel and waste management. Industrial facilities for uranian production, U 3 O 8 concentrate, UO 2 powder and CANDU fuel bundles, as well as heavy water plant, supply the required fuel and heavy water for Cernavoda NPP. The paper presents the Romanian activities in Nuclear Fuel Cycle and waste management fields. (authors)

  6. Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

    International Nuclear Information System (INIS)

    Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

    2004-01-01

    The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)

  7. Economic Analysis of Several Nuclear Fuel Cycles

    International Nuclear Information System (INIS)

    Ko, Won Il; Gao, Fanxing; Kim, Sung Ki

    2012-01-01

    Economics is one of the essential criteria to be considered for the future deployment of the nuclear power. With regard to the competitive power market, the cost of electricity from nuclear power plants is somewhat highly competitive with those from the other electricity generations, averaging lower in cost than fossil fuels, wind, or solar. However, a closer look at the nuclear power production brings an insight that the cost varies within a wide range, highly depending on a nuclear fuel cycle option. The option of nuclear fuel cycle is a key determinant in the economics, and therefrom, a comprehensive comparison among the proposed fuel cycle options necessitates an economic analysis for thirteen promising options based on the material flow analysis obtained by an equilibrium model as specified in the first article (Modeling and System Analysis of Different Fuel Cycle Options for Nuclear Power Sustainability (I): Uranium Consumption and Waste Generation). The objective of the article is to provide a systematic cost comparison among these nuclear fuel cycles. The generation cost (GC) generally consists of a capital cost, an operation and maintenance cost (O and M cost), a fuel cycle cost (FCC), and a decontaminating and decommissioning (D and D) cost. FCC includes a frontend cost and a back-end cost, as well as costs associated with fuel recycling in the cases of semi-closed and closed cycle options. As a part of GC, the economic analysis on FCC mainly focuses on the cost differences among fuel cycle options considered and therefore efficiently avoids the large uncertainties of the Generation-IV reactor capital costs and the advanced reprocessing costs. However, the GC provides a more comprehensive result covering all the associated costs, and therefrom, both GC and FCC have been analyzed, respectively. As a widely applied tool, the levelized cost (mills/KWh) proves to be a fundamental calculation principle in the energy and power industry, which is particularly

  8. Status and prospects of WWER in-core fuel management activities

    International Nuclear Information System (INIS)

    Novikov, A.N.; Pavlov, V.I.; Pavlovichev, A.M.; Proselkov, V.N.; Saprykin, V.V.

    1994-01-01

    A short review is given of recent extensive calculational and experimental studies carried out in Russia and Bulgaria for WWER fuel cycle modernization. The main activities performed at Kola NPP, Novovoronezh NPP, Kozloduy NPP and Balakovo NPP are outlined. Based on experience gained, the following improvements in the fuel cycle have been introduced: 1) increased fuel burnup; 2) reduced natural uranium consumption and decreased amount of separation work per energy output unit; 3) increased efficiency of the reactor emergency protection; 4) reduced fast neutron flux onto the reactor vessel. The main characteristics of modernized fuel cycles of WWER-440 and WWER-1000 are presented. 4 tabs., 3 figs., 14 refs

  9. Status and prospects of WWER in-core fuel management activities

    Energy Technology Data Exchange (ETDEWEB)

    Novikov, A N; Pavlov, V I; Pavlovichev, A M; Proselkov, V N; Saprykin, V V [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)

    1994-12-31

    A short review is given of recent extensive calculational and experimental studies carried out in Russia and Bulgaria for WWER fuel cycle modernization. The main activities performed at Kola NPP, Novovoronezh NPP, Kozloduy NPP and Balakovo NPP are outlined. Based on experience gained, the following improvements in the fuel cycle have been introduced: (1) increased fuel burnup; (2) reduced natural uranium consumption and decreased amount of separation work per energy output unit; (3) increased efficiency of the reactor emergency protection; (4) reduced fast neutron flux onto the reactor vessel. The main characteristics of modernized fuel cycles of WWER-440 and WWER-1000 are presented. 4 tabs., 3 figs., 14 refs.

  10. Nuclear fuel utilization in Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Boyadzhiev, Z; Kharalampieva, Ts; Pejchinov, Ts [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

    1994-12-31

    An assessment of fuel utilization in Kozloduy NPP units 1-6 is made on the basis of operational data obtained for a total of 62 fuel cycles. Basic characteristics of core loading and operation conditions are given. SPPS-1 and BIPR-7 codes are used to calculate assembly-wise power distributions for different full power days of a given cycle and unit. The data are compared with the measured values of these quantities. The analysis performed shows that the core loading option chosen has led to efficient fuel utilization without violation of the nuclear safety criteria. For WWER-440 (Units 1 - 4) this is expressed in effective reduction of the reactor vessel irradiation, maintaining the design duration of the fuel cycles at a reduced number of assemblies by a factor 5 - 5-10%, utilizing fuel with higher enrichment and implementing the 4-year fuel cycle. For WWER-1000 the improvements lead to: adoption of the 3-year fuel cycle utilizing fuel with 4.4% initial enrichment, implementation of improved fuel with a new type of absorbers and more effective low-leakage core loading patterns. 10 tabs., 6 figs., 7 refs.

  11. The nuclear fuel cycle, an overview

    International Nuclear Information System (INIS)

    Ballery, J.L.; Cazalet, J.; Hagemann, R.

    1995-01-01

    Because uranium is widely distributed on the face of the Earth, nuclear energy has a very large potential as an energy source in view of future depletion of fossil fuel reserves. Also future energy requirements will be very sizeable as populations of developing countries are often growing and make the energy question one of the major challenges for the coming decades. Today, nuclear contributes some 340 GWe to the energy requirements of the world. Present and future nuclear programs require an adequate fuel cycle industry, from mining, refining, conversion, enrichment, fuel fabrication, fuel reprocessing and the storage of the resulting wastes. The commercial fuel cycle activities amount to an annual business in the 7-8 billions of US Dollars in the hands of a large number of industrial operators. This paper gives details about companies and countries involved in each step of the fuel cycle and about the national strategies and options chosen regarding the back end of the fuel cycle (waste storage and reprocessing). These options are illustrated by considering the policy adopted in three countries (France, United Kingdom, Japan) versed in reprocessing. (J.S.). 13 figs., 2 tabs

  12. Physics characteristics of CANDU cores with advanced fuel cycles

    International Nuclear Information System (INIS)

    Garvey, P.M.

    1985-01-01

    The current generation of CANDU reactors, of which some 20 GWE are either in operations or under construction worldwide, have been designed specifically for the natural uranium fuel cycle. The CANDU concept, due to its D 2 O coolant and moderator, on-power refuelling and low absorption structural materials, makes the most effective utilization of mined uranium of all currently commercialized reactors. An economic fuel cycle cost is also achieved through the use of natural uranium and a simple fuel bundle design. Total unit energy costs are achieved that allow this reactor concept to effectively compete with other reactor types and other forms of energy production. There are, however, other fuel cycles that could be introduced into this reactor type. These include the slightly enriched uranium fuel cycle, fuel cycles in which plutonium is recycled with uranium, and the thorium cycle in which U-233 is recycled. There is also a special range of fuel cycles that could utilize the spent fuel from LWR's. Two specific variants are a fuel cycle that only utilizes the spent uranium, and a fuel cycle in which both the uranium and plutonium are recycled into a CANDU. For the main part these fuel cycles are characterized by a higher initial enrichment, and hence discharge burnup, than the natural uranium cycle. For these fuel cycles the main design features of both the reactor and fuel bundle would be retained. Recently a detailed study of the use in a CANDU of mixed plutonium and uranium oxide fuel from an LWR has been undertaken by AECL. This study illustrates many of the generic technical issues associated with the use of Advanced Fuel Cycles. This paper will report the main findings of this evaluation, including the power distribution in the reactor and fuel bundle, the choice of fuel management scheme, and the impact on the control and safety characteristics of the reactor. These studies have not identified any aspects that significantly impact upon the introduction of

  13. Compound process fuel cycle concept

    International Nuclear Information System (INIS)

    Ikegami, Tetsuo

    2005-01-01

    Mass flow of light water reactor spent fuel for a newly proposed nuclear fuel cycle concept 'Compound Process Fuel Cycle' has been studied in order to assess the capacity of the concept for accepting light water reactor spent fuels, taking an example for boiling water reactor mixed oxide spent fuel of 60 GWd/t burn-up and for a fast reactor core of 3 GW thermal output. The acceptable heavy metal of boiling water reactor mixed oxide spent fuel is about 3.7 t/y/reactor while the burn-up of the recycled fuel is about 160 GWd/t and about 1.6 t/y reactor with the recycled fuel burn-up of about 300 GWd/t, in the case of 2 times recycle and 4 times recycle respectively. The compound process fuel cycle concept has such flexibility that it can accept so much light water reactor spent fuels as to suppress the light water reactor spent fuel pile-up if not so high fuel burn-up is expected, and can aim at high fuel burn-up if the light water reactor spent fuel pile-up is not so much. Following distinctive features of the concept have also been revealed. A sort of ideal utilization of boiling water reactor mixed oxide spent fuel might be achieved through this concept, since both plutonium and minor actinide reach equilibrium state beyond 2 times recycle. Changes of the reactivity coefficients during recycles are mild, giving roughly same level of reactivity coefficients as the conventional large scale fast breeder core. Both the radio-activity and the heat generation after 4 year cooling and after 4 times recycle are less than 2.5 times of those of the pre recycle fuel. (author)

  14. Effect of fuel burnup on the mechanical safety coefficients

    International Nuclear Information System (INIS)

    Plyashkevich, V.Ju.; Sidorenko, V.D.; Shishkov, L.K.

    2001-01-01

    )In the paper the results of studies of changes in the process of campaign 'disturbances' of local heat flux and local fuel burnup, resulting from the 'mechanical' deviations in the composition and geometrical characteristics of fuel rods from the nominal are given. As example, the WWER-440 fuel assembly with burnable poisons used in the five-year fuel cycle is considered. The effect of deviations in fuel enrichment, fuel content, gadolinium content and geometrical size was studied (Authors)

  15. Nuclear safety evaluation of the VVER 440, Type 213

    International Nuclear Information System (INIS)

    Urbancik, L.

    1997-01-01

    The supervisory activities of the State Office for Nuclear Safety at the Dukovany nuclear power plant are described. No event resulting in an inpermissible radioactivity leak into the environment occurred at the plant in 1996. From among the 76 failures and events having occurred, only 4 were classified as level 1 on the International Nuclear Event Scale. Changes in the technology of radioactive waste bituminization were proposed. The Interim Spent Fuel Storage Facility at the Dukovany site was in test operation in 1996. Selected physical parameters of this facility were monitored. Seven international transports of spent fuel were accomplished in 1996. The dose rates in the surroundings of the Dukovany plant are monitored constantly by a teledosimetric system operated by the nuclear power plant. Periodical sampling and radionuclide activity measurements in the environment are also performed. (M.D.)

  16. Fuel cycle centers revisited: Consolidation of fuel cycle activities in a few countries

    International Nuclear Information System (INIS)

    Kratzer, M.B.

    1996-01-01

    Despite varied expressions, the general impression remains that the international fuel cycle center concept, whatever its merits, is visionary. It also is quite possibly unattainable in light of strong national pressures toward independence and self-sufficiency in all things nuclear. Is the fuel cycle center an idea that has come and gone? Is it an idea whose time has not yet come? Or is it, as this paper suggests, an idea that has already arrived on the scene, attracting little attention or even acknowledgement of its presence? The difficult in answering this questions arises, in part, from the fact that despite its long and obvious appeal, there has been very little systematic analysis of the concept itself. Such obvious questions as how many and where fuel cycle centers should be located; what characteristics should the hot country or countries possess; and what are the institutional forms or features that endow the concept with enhanced proliferation protection have rarely been seriously and systematically addressed. The title of this paper focuses on limiting the geographic spread of fuel cycle facilities, and some may suggest that doing so does not necessarily call for any type of international or multinational arrangements applicable to those that exist. It is a premise of this paper, however, that a restriction on the number of countries possessing sensitive fuel cycle facilities necessarily involves some degree of multinationalization. This is not only because in every instance a nonproliferation pledge and international or multinational safeguards, or both, will be applied to the facility, but also because a restriction on the number of countries possessing these facilities implies that those in existence will serve a multinational market. This feature in itself is an important form of international auspices. Thus, the two concepts--limitation and multinationalization--if not necessarily one and the same, are at least de facto corollaries

  17. WWER-440 type reactor core

    International Nuclear Information System (INIS)

    Mizov, J.; Svec, P.; Rajci, T.

    1987-01-01

    Assemblies with patly spent fuel of enrichment within 5 and 36 MWd/kg U or lower than the maximum enrichment of freshly charged fuel are placed in at least one of the peripheral positions of each hexagonal sector of the WWER-440 reactor type core. This increases fuel availability and reduces the integral neutron dose to the reactor vessel. The duration is extended of the reactor campaign and/or the mean fuel enrichment necessary for the required duration of the period between refuellings is reduced. Thus, fuel costs are reduced by 1 up to 3%. The results obtained in the experiment are tabulated. (J.B.). 1 fig., 3 tabs

  18. Rapsodie: A closed fuel cycle

    International Nuclear Information System (INIS)

    Levallet, E.H.; Costa, L.; Mougniot, J.C.; Robin, J.

    1977-01-01

    The Fortissimo Version of the core of the RAPSODIE fast reactor produces 40 MWTh. Since its start up in May 1970 in the CEN-CADARACHE its availability has stayed around 85%. Some of the mixed oxyde fuel pins UO 2 - 30% PuO 2 have already reached 150.000 MWd/t. The reprocessing is done in the pilot plant located in the La Hague Center and the plutonium obtained has already been re-used in the reactor. The Rapsodie-Fortissimo cycle is therefore now a closed cycle. This cycle is quite representative of fast reactor cycle characteristics and thus provides a remarkable research and development tool for the study of fabrication, in-reactor performances, transport, storage and reprocessing. These studies concern in particular the evolution of fission products and heavy isotopes content in fuel which controls both reprocessing schemes and intensity of emitted radiations. A program for the analysis of irradiated fuel has been developed either using samples collected all along the cycle, or following the actual reprocessing subassemblies. A set of basic data and calculation models has been established with two objectives: to give a better interpretation of the experimental program on one hand, and to extrapolate these results to the fuel cycle of fast reactors in general on the other hand. The first results have been quite encouraging up to now [fr

  19. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Hornaes, A.; Bodal, T.; Sunde, S.

    1998-01-01

    The Institutt for energiteknikk has developed the core surveillance system SCORPIO, which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators, which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety, as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. The system has been implemented on western PWRs, but the basic concept is applicable to a wide range of reactors including VVERs. The main differences between VVERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a VVER version of SCORPIO has been done in co-operation with the Nuclear Research Institute Rez, and industry partners in the Czech Republic. The first system is installed at Dukovany NPP, where the Site Acceptance Test was completed 6. March 1998.(author)

  20. Back-end of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Choi, J.S.

    2002-01-01

    Current strategies of the back-end nuclear fuel cycles are: (1) direct-disposal of spent fuel (Open Cycle), and (2) reprocessing of the spent fuel and recycling of the recovered nuclear materials (Closed Cycle). The selection of these strategies is country-specific, and factors affecting selection of strategy are identified and discussed in this paper. (author)

  1. International nuclear fuel cycle fact book. Revision 6

    International Nuclear Information System (INIS)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2

  2. Rolls-Royce successful modernization of safety-critical Instrumentation and Control (I and C) equipment at the Dukovany VVER 440/213 Nuclear Power Plant, based on SPINLINE 3 platform

    International Nuclear Information System (INIS)

    Rebreyend, P.; Burel, J.P.; Spoc, J.; Karasek, A.

    2010-01-01

    Rolls-Royce has provided on-time delivery of a substantial safety-critical I and C overhaul for four Nuclear reactors operated by Czech Republic utility, CEZ a.s. This nine-year project is considered to be one of the largest I and C modernization projects in the world. The Dukovany VVER 440 I and C modernization project and its key success factors are profiled in this paper. The project is in the final stages with the last unit to be completed in 2009. Beginning in September 2000, the project is in compliance with the initial schedule. Rolls-Royce has been designing and manufacturing I and C solutions dedicated to the implementation of safety and safety-related functions in nuclear power plants (NPPs) for more than 30 years. Though the early solutions were non-software-based, since 1984 software-based solutions for safety I and C functions have been deployed in operating NPPs across France and 15 other countries. The Rolls-Royce platform is suitable for implementation of safety I and C functions in new NPPs, as well as in the modernization of safety equipment in existing plants. CEZ a.s. is a major electricity supplier for the national grid. At Dukovany, CEZ a.s. operates four units of VVER-440/213-type reactors producing one quarter of CEZ a.s. electricity production. The first of these units was connected to the grid in 1985. Since the year 2000, the nine-year modernization program has been underway at Dukovany, at a cost of more than 200 million Euros. The equipment replacement was implemented during regular, planned outages of the original equipment and systems. After an international bidding phase, CEZ a.s. awarded a contract to Skoda JS for general engineering and project management. Individual subcontracts were then signed between Skoda JS and a consortium between Rolls-Royce and Areva for modernization of the safety systems, including the Reactor Protection System (RPS), the Reactor Control System (RCS), and the Post-Accident Monitoring System (PAMS). Two

  3. Validation of computer codes and modelling methods for giving proof of nuclear safety of transport and storage of spent VVER-type nuclear fuels. Pt. 2. Criticality safety during transport and storage of spent VVER fuel elements. Final report; Einschaetzung von Rechenprogrammen und Methoden zum Nachweis der nuklearen Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoffen. T. 2. Kritikalitaetssicherheit bei Transport und Lagerung von WWER-Brennelementen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buechse, H.; Langowski, A.; Lein, M.; Nagel, R.; Schmidt, H.; Stammel, M.

    1995-03-15

    The report gives the results of investigations on the validation of computer codes used to prove nuclear safety during transport and storage of spent VVER - fuel of NPP Greifswald and Rheinsberg. Characteristics of typical spent fuel (nuclide concentration, neutron source strength, gamma spectrum, decay heat) - calculated with several codes - and dose rates (e.g. in the surrounding of a loaded spent fuel cask) - based on the different source terms - are presented. Differences and their possible reasons are discussed. The results show that despite the differences in the source terms all relevant health physics requirements are met for all cases of source term. The validation of the criticality code OMEGA was established by calculation of appr. 200 critical experiments of LWR fuel, including VVER fuel rod arrangements. The mean error of the effective multiplication factor k{sub eff} is -0,01 compared to the experiment for this area of applicability. Thus, the OMEGA error of 2% assumed in earlier works has turned out to be sufficiently conservative. (orig.) [Deutsch] Der Bericht enthaelt die Ergebnisse von Untersuchungen zur Validierung von Rechenprogrammen, welche zum Nachweis der nukelaren Sicherheit bei Transport und Lagerung von WWER-Kernbrennstoff der KKW Greifswald und Rheinsberg eingesetzt wurden. Es werden eine Charakteristik des abgebrannten Brennstoffs (Nuklidkonzentrationen, Neutronenquellstaerke, Gammaspektrum, Nachzerfallsleistung) - berechnet mit verschiedenen Programmen - und Ortsdosisleistungen (z.B. in der Umgebung eines Transportbehaelters) - basierend auf den verschiedenen Quelltermen - angegeben. Differenzen und Ursachen werden diskutiert. Die Ergebnisse zeigen, dass trotz der Differenzen in den Quelltermen alle strahlenschutztechnisch relevanten Aussagen unbeeinflusst bleiben. Fuer die Einschaetzung des Gueltigkeitsbereiches des Monte-Carlo-Programms OMEGA wurden ca. 200 kritische Experimente mit LWR-Brennstoff unter besonderer Beruecksichtigung

  4. Introductory remarks about the international fuel cycle

    International Nuclear Information System (INIS)

    Wolfe, B.

    1989-01-01

    The reason why nuclear power has promise is because of the promise of its fuel cycle. The fuel cycle is in fairly good shape and has demonstrated the characteristics of good economics, good general characterization, and good maintenance of the various parts of the fuel cycle. The thermal recycling of fuel is an area in which the economics have changed to the point that, at least in many parts of the world, it's no longer economical

  5. IAEA Activities in the Area of Fast Reactors and Related Fuels and Fuel Cycles

    International Nuclear Information System (INIS)

    Monti, S.; Basak, U.; Dyck, G.; Inozemtsev, V.; Toti, A.; Zeman, A.

    2013-01-01

    Summary: • The IAEA role to support fast reactors and associated fuel cycle development programmes; • Main IAEA activities on fast reactors and related fuel and fuel cycle technology; • Main IAEA deliverables on fast reactors and related fuel and fuel cycle technology

  6. Perspective decisions of WWER nuclear fuel: Implementation at Russian NPPs

    International Nuclear Information System (INIS)

    Molchanov, V.

    2003-01-01

    The scientific and technical policy pursued by JSC TVEL has managed to create a new generation of fuel assembly design on the basis of solutions tested at various units of Russian NPPs - Kola NPP, Kalinin NPP, Unit 1, Balakovo NPP Unit 1. The requirements set for the new generation nuclear fuel for WWER are: 1) High fuel burnup - up to 70 MWxdays/kgU; 2) Extended operation cycle - up to 6 years; 3) Increase of uranium charge to the core; 4) Increased lateral stability - bow not more than 7 mm; 5) High level of operating reliability - fuel rod leakage not worse than 10-5 1/year; 6) Demountable fuel assembly design. Post-irradiation examination results of fuel assemblies discharged from WWER-1000 reactors demonstrate that fuel rods have substantial reserve in general characteristics including that of dealing with planned burnup. In order to meet the requirements, trials are started for: implementation of rigid skeleton (WWER-1000); fuel column length extension (WWER-1000 and WWER-440); increase of UO 2 charge (WWER-1000 and WWER-440); enhancing of operational reliability and demountable design. It is concluded that the Russian nuclear fuel for WWER-type reactors is competitive and enables the implementation of state-of-the-art cost effective fuel cycles

  7. Regulation of fuel cycle facilities in the UK

    International Nuclear Information System (INIS)

    Ascroft-Hutton, H.W.

    2001-01-01

    The UK has facilities for the production of uranium hexafluoride, its enrichment, conversion into fuel and for the subsequent reprocessing of irradiated fuel and closure of the fuel cycle. All of these facilities must be licensed under UK legislation. HM Nuclear Installations Inspectorate has delegated powers to issue the licence and to attach any conditions it considers necessary in the interests of safety. The fuel cycle facilities in the UK have been licensed since 1971. This paper describes briefly the UK nuclear regulatory framework and the fuel cycle facilities involved. It considers the regulatory practices adopted together with similarities and differences between regulation of fuel cycle facilities and power reactors. The safety issues associated with the fuel cycle are discussed and NII's regulatory strategy for these facilities is set out. (author)

  8. Studying some regimes of the WWER-440 type reactor failed fuel element operation

    International Nuclear Information System (INIS)

    Aksenov, N.A.; Samsonov, B.V.; Sulaberidze, V.Sh.; Frej, A.K.

    1981-01-01

    The results of investigating the serviceability of experimental fuel elements close by type to that of the WWER-440 type reactor in the cans of which untightness in the form of small opening are made. The tests are carried out in the SM-2 reactor high temperature water loop at the temperature of 473 K, pressure of (1-2)x10 4 kPa, coolant flow rate of 3.7-5.5 m 3 /h. The analysis of the obtained results shows that the character of changes in the fission product (FP) activity in the circuit in a considerable extent is determined bt the thermal-optical conditions of the fuel element operation. If water in the gap between fuel and can does not boil, activity changes smoothly and bursts caused by increased FP release are observed only under transient conditions of reactor operation. In the presence of water boiling in the gap the FP release has of impulse character with the frequency determined besides the untightness dimension by free volume inside the fuel element can (with its increase the pulsation frequency increases). FP release from fuel is connected with their direct escape from an open surface. When water in the gap the FP release from the fuel element occurs practically immediately. Without boiling the FP delay in the gap is determined by their diffusion in a layer of water. The conclusion is drawn that the FP release from failed fuel elements may be reduced by eliminating the water boiling in the gap between the fuel and the can by means of the fuel element power or coolant temperature decrease

  9. International nuclear fuel cycle fact book. Revision 6

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, K.M.; Lakey, L.T.; Leigh, I.W.; Jeffs, A.G.

    1986-01-01

    The International Fuel Cycle Fact Book has been compiled in an effort to provide (1) an overview of worldwide nuclear power and fuel cycle programs and (2) current data concerning fuel cycle and waste management facilities, R and D programs and key personnel. Additional information on each country's program is available in the International Source Book: Nuclear Fuel Cycle Research and Development, PNL-2478, Rev. 2.

  10. Social awareness on nuclear fuel cycle

    International Nuclear Information System (INIS)

    Tanigaki, Toshihiko

    2006-01-01

    In the present we surveyed public opinion regarding the nuclear fuel cycle to find out about the social awareness about nuclear fuel cycle and nuclear facilities. The study revealed that people's image of nuclear power is more familiar than the image of the nuclear fuel cycle. People tend to display more recognition and concern towards nuclear power and reprocessing plants than towards other facilities. Comparatively speaking, they tend to perceive radioactive waste disposal facilities and nuclear power plants as being highly more dangerous than reprocessing plants. It is found also that with the exception of nuclear power plants don't know very much whether nuclear fuel cycle facilities are in operation in Japan or not. The results suggests that 1) the relatively mild image of the nuclear fuel cycle is the result of the interactive effect of the highly dangerous image of nuclear power plants and the less dangerous image of reprocessing plants; and 2) that the image of a given plant (nuclear power plant, reprocessing plant, radioactive waste disposal facility) is influenced by the fact of whether the name of the plant suggests the presence of danger or not. (author)

  11. Reference thorium fuel cycle

    International Nuclear Information System (INIS)

    Driggers, F.E.

    1978-08-01

    In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with 238 U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of 238 U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure

  12. Aspects of the fast reactors fuel cycle

    International Nuclear Information System (INIS)

    Zouain, D.M.

    1982-06-01

    The fuel cycle for fast reactors, is analysed, regarding the technical aspects of the developing of the reprocessing stages and the fuel fabrication. The environmental impact of LMFBRs and the waste management of this cycle are studied. The economic aspects of the fuel cycle, are studied too. Some coments about the Brazilian fast reactors programs are done. (E.G.) [pt

  13. Survey of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Zech, H.J.; Pickert, F.K.

    1975-01-01

    A brief outline of the technical aspects of the fuel cycle, starting from the mining of uranium up to fuel element fabrication, is followed by a more detailed description of the management of the outer fuel cycle. This includes the system of contracts and their reciprocal technical and chronological interdepence, as well as financial aspects, market conditions and trends. (RB) [de

  14. Ecological effects of fuel cycle activities

    Energy Technology Data Exchange (ETDEWEB)

    Barnthouse, L; Cada, G; Kroodsma, R; Shriner, D; Tolbert, V; Turner, R

    1994-07-01

    The purpose of this paper is to summarize the approach used to characterize ecological impacts of the coal fuel cycle. The same approach is used for many of the impacts in other fuel cycles as well. The principal analytical approach being used in the study is an accounting framework - that is, a series of matrices that map each phase of the fuel cycle to a suite of possible. emissions, each emission to a suite of impact categories, and each impact category to an external cost. This paper summarizes the ecological impacts of all phases of the coal fuel cycle, defines the ecological impact categories used in the study's 'accounting framework', and discusses alternative approaches to quantification. Externalities associated with CO{sub 2}-induced global climate change are beyond the scope of this paper and are not discussed.

  15. Ecological effects of fuel cycle activities

    International Nuclear Information System (INIS)

    Barnthouse, L.; Cada, G.; Kroodsma, R.; Shriner, D.; Tolbert, V.; Turner, R.

    1994-01-01

    The purpose of this paper is to summarize the approach used to characterize ecological impacts of the coal fuel cycle. The same approach is used for many of the impacts in other fuel cycles as well. The principal analytical approach being used in the study is an accounting framework - that is, a series of matrices that map each phase of the fuel cycle to a suite of possible. emissions, each emission to a suite of impact categories, and each impact category to an external cost. This paper summarizes the ecological impacts of all phases of the coal fuel cycle, defines the ecological impact categories used in the study's 'accounting framework', and discusses alternative approaches to quantification. Externalities associated with CO 2 -induced global climate change are beyond the scope of this paper and are not discussed

  16. Nuclear Fusion Fuel Cycle Research Perspectives

    International Nuclear Information System (INIS)

    Chung, Hongsuk; Koo, Daeseo; Park, Jongcheol; Kim, Yeanjin; Yun, Sei-Hun

    2015-01-01

    As a part of the International Thermonuclear Experimental Reactor (ITER) Project, we at the Korea Atomic Energy Research Institute (KAERI) and our National Fusion Research Institute (NFRI) colleagues are investigating nuclear fusion fuel cycle hardware including a nuclear fusion fuel Storage and Delivery System (SDS). To have a better knowledge of the nuclear fusion fuel cycle, we present our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). To have better knowledge of the nuclear fusion fuel cycle, we presented our research efforts not only on SDS but also on the Fuel Supply System (FS), Tokamak Exhaust Processing System (TEP), Isotope Separation System (ISS), and Detritiation System (DS). Our efforts to enhance the tritium confinement will be continued for the development of cleaner nuclear fusion power plants

  17. Innovative instrumentation for VVERs based in non-invasive techniques

    International Nuclear Information System (INIS)

    Jeanneau, H.; Favennec, J.M.; Tournu, E.; Germain, J.L.

    2000-01-01

    Nuclear power plants such as VVERs can greatly benefit from innovative instrumentation to improve plant safety and efficiency. In recent years innovative instrumentation has been developed for PWRs with the aim of providing additional measurements of physical parameters on the primary and secondary circuits: the addition of new instrumentation is made possible by using non-invasive techniques such as ultrasonics and radiation detection. These innovations can be adapted for upgrading VVERs presently in operation and also in future VVERs. The following innovative instrumentation for the control, monitoring or testing at VVERs is described: 1. instrumentation for more accurate primary side direct measurements (for a better monitoring of the primary circuit); 2. instrumentation to monitor radioactivity leaks (for a safer plant); 3. instrumentation-related systems to improve the plant efficiency (for a cheaper kWh)

  18. Serving the fuel cycle: preparing tomorrow's packagings

    International Nuclear Information System (INIS)

    Roland, V.

    2001-01-01

    The main fleet of transport packagings serving today the fuel cycle was born more than 20 years ago. Or was it they? The present paper will show that serving the fuel cycle by preparing tomorrow's logistics is actually an on-going process, rather than a rupture. We shall review the great packagings of the fuel cycle: In the front end, the major actors are the UF 4 , UF 6 , enriched UF 6 , UO 2 powders, fresh fuel packagings. In the back end of the fuel cycle, we find the dry transport casks of the TN-12, TN-17, TN-13, family and also the Excellox wet flasks. In the waste management, a whole fleet of containers, culminating in the TN Gemini, are available or being created. (author)

  19. Nuclear fuel cycle modelling using MESSAGE

    International Nuclear Information System (INIS)

    Guiying Zhang; Dongsheng Niu; Guoliang Xu; Hui Zhang; Jue Li; Lei Cao; Zeqin Guo; Zhichao Wang; Yutong Qiu; Yanming Shi; Gaoliang Li

    2017-01-01

    In order to demonstrate the possibilities of application of MESSAGE tool for the modelling of a Nuclear Energy System at the national level, one of the possible open nuclear fuel cycle options based on thermal reactors has been modelled using MESSAGE. The steps of the front-end and back-end of nuclear fuel cycle and nuclear reactor operation are described. The optimal structure for Nuclear Power Development and optimal schedule for introducing various reactor technologies and fuel cycle options; infrastructure facilities, nuclear material flows and waste, investments and other costs are demonstrated. (author)

  20. Nuclear fuel cycle information workshop

    International Nuclear Information System (INIS)

    1983-01-01

    This overview of the nuclear fuel cycle is divided into three parts. First, is a brief discussion of the basic principles of how nuclear reactors work; second, is a look at the major types of nuclear reactors being used and world-wide nuclear capacity; and third, is an overview of the nuclear fuel cycle and the present industrial capability in the US

  1. Proceeding of the Fifth Scientific Presentation on Nuclear Fuel Cycle: Development of Nuclear Fuel Cycle Technology in Third Millennium

    International Nuclear Information System (INIS)

    Suripto, A.; Sastratenaya, A.S.; Sutarno, D.

    2000-01-01

    The proceeding contains papers presented in the Fifth Scientific Presentation on Nuclear Fuel Element Cycle with theme of Development of Nuclear Fuel Cycle Technology in Third Millennium, held on 22 February in Jakarta, Indonesia. These papers were divided by three groups that are technology of exploration, processing, purification and analysis of nuclear materials; technology of nuclear fuel elements and structures; and technology of waste management, safety and management of nuclear fuel cycle. There are 35 papers indexed individually. (id)

  2. Survey of nuclear fuel cycle economics: 1970--1985

    International Nuclear Information System (INIS)

    Prince, B.E.; Peerenboom, J.P.; Delene, J.G.

    1977-03-01

    This report is intended to provide a coherent view of the diversity of factors that may affect nuclear fuel cycle economics through about 1985. The nuclear fuel cycle was surveyed as to past trends, current problems, and future considerations. Unit costs were projected for each step in the fuel cycle. Nuclear fuel accounting procedures were reviewed; methods of calculating fuel costs were examined; and application was made to Light Water Reactors (LWR) over the next decade. A method conforming to Federal Power Commission accounting procedures and used by utilities to account for backend fuel-cycle costs was described which assigns a zero net salvage value to discharged fuel. LWR fuel cycle costs of from 4 to 6 mills/kWhr (1976 dollars) were estimated for 1985. These are expected to reach 6 to 9 mills/kWr if the effect of inflation is included

  3. The nuclear fuel cycle versus the carbon cycle

    International Nuclear Information System (INIS)

    Ewing, R.C.

    2005-01-01

    Nuclear power provides approximately 17% of the world's electricity, which is equivalent to a reduction in carbon emissions of ∼0.5 gigatonnes (Gt) of C/yr. This is a modest reduction as compared with global emissions of carbon, ∼7 Gt C/yr. Most analyses suggest that in order to have a significant and timely impact on carbon emissions, carbon-free sources, such as nuclear power, would have to expand total production of energy by factors of three to ten by 2050. A three-fold increase in nuclear power capacity would result in a projected reduction in carbon emissions of 1 to 2 Gt C/yr, depending on the type of carbon-based energy source that is displaced. This three-fold increase utilizing present nuclear technologies would result in 25,000 metric tonnes (t) of spent nuclear fuel (SNF) per year, containing over 200 t of plutonium. This is compared to a present global inventory of approximately 280,000 t of SNF and >1,700 t of Pu. A nuclear weapon can be fashioned from as little as 5 kg of 239 Pu. However, there is considerable technological flexibility in the nuclear fuel cycle. There are three types of nuclear fuel cycles that might be utilized for the increased production of energy: open, closed, or a symbiotic combination of different types of reactor (such as, thermal and fast neutron reactors). The neutron energy spectrum has a significant effect on the fission product yield, and the consumption of long-lived actinides, by fission, is best achieved by fast neutrons. Within each cycle, the volume and composition of the high-level nuclear waste and fissile material depend on the type of nuclear fuel, the amount of burn-up, the extent of radionuclide separation during reprocessing, and the types of materials used to immobilize different radionuclides. As an example, a 232 Th-based fuel cycle can be used to breed fissile 233 U with minimum production of Pu. In this paper, I will contrast the production of excess carbon in the form of CO 2 from fossil fuels with

  4. CANDU fuel cycles - present and future

    International Nuclear Information System (INIS)

    Mooradian, A.J.

    1976-05-01

    The present commercially proven Canadian nuclear power system is based on a once-through natural uranium fuel cycle characterized by high uranium utilization and a high conversion efficiency. The cycle closes with secure retrievable storage of spent fuel. This cycle is based on a CANDU reactor concept which is now well understood. Both active and passive fuel storage options have been investigated and will be described in this paper. Future development of the CANDU system is focussed on conservation of uranium by plutonium and thorium recycle. The full exploitation of these options requires continued emphasis on neutron conservation, efficiency of extraction and fuel refabrication processes. The results of recent studies are discussed in this paper. (author)

  5. CFTSIM-ITER dynamic fuel cycle model

    International Nuclear Information System (INIS)

    Busigin, A.; Gierszewski, P.

    1998-01-01

    Dynamic system models have been developed for specific tritium systems with considerable detail and for integrated fuel cycles with lesser detail (e.g. D. Holland, B. Merrill, Analysis of tritium migration and deposition in fusion reactor systems, Proceedings of the Ninth Symposium Eng. Problems of Fusion Research (1981); M.A. Abdou, E. Vold, C. Gung, M. Youssef, K. Shin, DT fuel self-sufficiency in fusion reactors, Fusion Technol. (1986); G. Spannagel, P. Gierszewski, Dynamic tritium inventory of a NET/ITER fuel cycle with lithium salt solution blanket, Fusion Eng. Des. (1991); W. Kuan, M.A. Abdou, R.S. Willms, Dynamic simulation of a proposed ITER tritium processing system, Fusion Technol. (1995)). In order to provide a tool to understand and optimize the behavior of the ITER fuel cycle, a dynamic fuel cycle model called CFTSIM is under development. The CFTSIM code incorporates more detailed ITER models, specifically for the important isotope separation system, and also has an easier-to-use graphical interface. This paper provides an overview of CFTSIM Version 1.0. The models included are those with significant and varying tritium inventories over a test campaign: fueling, plasma and first wall, pumping, fuel cleanup, isotope separation and storage. An illustration of the results is shown. (orig.)

  6. Alternative fuel cycles

    International Nuclear Information System (INIS)

    Penn, W.J.

    1979-05-01

    Uranium resource utilization and economic considerations provide incentives to study alternative fuel cycles as future options to the PHWR natural uranium cycle. Preliminary studies to define the most favourable alternatives and their possible introduction dates are discussed. The important and uncertain components which influence option selection are reviewed, including nuclear capacity growth, uranium availability and demand, economic potential, and required technological developments. Finally, a summary of Ontario Hydro's program to further assess cycle selection and define development needs is given. (auth)

  7. Fuel cycle problems in fusion reactors

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1976-01-01

    Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known

  8. Status and development of the thorium fuel cycle

    International Nuclear Information System (INIS)

    Yi Weijing; Wei Renjie

    2003-01-01

    A perspective view of the thorium fuel cycle is provided in this paper. The advantages and disadvantages of the thorium fuel cycle are given and the development of thorium fuel cycle in several types of reactors is introduced. The main difficulties in developing the thorium fuel cycle lie in the reprocessing and disposal of the waste and its economy, and the ways tried by foreign countries to solve the problems are presented in the paper

  9. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    International Nuclear Information System (INIS)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A.; Elkin, I.V.; Pylev, S.S.

    2005-01-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  10. PSB-VVER counterpart experiment simulating a small cold leg break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V.N.; Melikhov, O.I.; Kapustin, A.V.; Lipatov, I.A.; Dremin, G.I.; Nikonov, S.M.; Rovnov, A.A. [Elektrogorsk Research and Engineering Center, EREC, Bezymiannaja Street, 6, Elektrogorsk, Moscow Region, 142530 (Russian Federation); Elkin, I.V.; Pylev, S.S. [NSI RRC ' Kurchatov Institute' , Kurchatov Sq., 1, Moscow, 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: An experiment simulating a small break LOCA has been performed in PSB-VVER facility, under PSB-VVER OECD Project. The test is intended to be a counterpart one to an experiment performed in the LOBI integral test facility. The objectives of the PSB-VVER small cold leg break test are: to study VVER-1000 thermal hydraulic response following a small break in the cold leg, to provide data for code assessment regarding phenomena indicated in the VVER-1000 code validation matrix and to study the scaling effect. The scenario for the PSB-VVER experiment has been designed taking the LOBI BL-34 test as reference. The ratio of primary system volumes (without volume of the pressurizer and the surge line) has been chosen to scale the reference experiment conditions and to generate the conditions of PSB-VVER cold leg break experiment. The resulting conditions are compared with the LOBI cold leg break test conditions by means of different counterpart test criteria. Comparing the two experiments in terms of the criteria shows that basic requirements to the counterpart test are fulfilled. A pretest analysis with RELAP5/MOD3.2 code has shown that the PSBVVER small break experiment is expected to show the same relevant phenomena and main events as the LOBI BL-34 test. The predicted PSB-VVER primary pressure is very close to that measured in the LOBI facility. The measured pressure in the PSB-VVER primary system has turned out to be very close to that registered in LOBI BL-34 test. This verifies the approach used for developing the conditions of the PSB-VVER counterpart test. The experiment results and the RELAP5/MOD3.2 pretest calculation are in good agreement. A posttest calculation of the experiment with RELAP5/MOD3.2 code has been performed in order to assess the codes capability to simulate the phenomena relevant to the test. The code has shown a reasonable prediction of the phenomena measured in the experiment. (authors)

  11. Fuel cycle technologies - The next 50 years

    International Nuclear Information System (INIS)

    Chamberlain, L.N.; Ion, S.E.; Patterson, J.

    1997-01-01

    World energy demands are set to increase through the next Millennium. As fossil fuel reserves fall and environmental concerns increase there is likely to be a growing dependence on nuclear and renewable sources for electricity generation. This paper considers some of the desirable attributes of the nuclear fuel cycle in the year 2050 and emphasises the importance of considering the whole of the fuel cycle in an integrated way - the concept of the 'holistic' fuel cycle. We then consider how some sectors of the fuel cycle will develop, through a number of multi- national contributions covering: enrichment, fuel, aqueous reprocessing, non-aqueous reprocessing, P and T, MOX, direct disposal, waste. Finally, we summarize some of the key technical and institutional challenges that lie ahead if nuclear power is going to play its part in ensuring that planet Earth is a safe and hospitable place to live. (author)

  12. Modeling the Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Jacobson, Jacob J.; Dunzik-Gougar, Mary Lou; Juchau, Christopher A.

    2010-01-01

    A review of existing nuclear fuel cycle systems analysis codes was performed to determine if any existing codes meet technical and functional requirements defined for a U.S. national program supporting the global and domestic assessment, development and deployment of nuclear energy systems. The program would be implemented using an interconnected architecture of different codes ranging from the fuel cycle analysis code, which is the subject of the review, to fundamental physical and mechanistic codes. Four main functions are defined for the code: (1) the ability to characterize and deploy individual fuel cycle facilities and reactors in a simulation, while discretely tracking material movements, (2) the capability to perform an uncertainty analysis for each element of the fuel cycle and an aggregate uncertainty analysis, (3) the inclusion of an optimization engine able to optimize simultaneously across multiple objective functions, and (4) open and accessible code software and documentation to aid in collaboration between multiple entities and facilitate software updates. Existing codes, categorized as annualized or discrete fuel tracking codes, were assessed according to the four functions and associated requirements. These codes were developed by various government, education and industrial entities to fulfill particular needs. In some cases, decisions were made during code development to limit the level of detail included in a code to ease its use or to focus on certain aspects of a fuel cycle to address specific questions. The review revealed that while no two of the codes are identical, they all perform many of the same basic functions. No code was able to perform defined function 2 or several requirements of functions 1 and 3. Based on this review, it was concluded that the functions and requirements will be met only with development of a new code, referred to as GENIUS.

  13. Back end of the fuel cycle

    International Nuclear Information System (INIS)

    Wolfe, B.; Lambert, R.W.

    1975-01-01

    At present, that portion of the nuclear fuel cycle involving reprocessing, waste management, and mixed-oxide fuel fabrication is in an unsettled state. Government regulatory requirements with respect to all aspects of the back end of the fuel cycle are still being formulated, and there is little positive experience on the operation of commercial reprocessing or mixed-oxide fabrication plants. In view of this unsettled situation, it will be difficult to meet the reprocessing and mixed-oxide fabrication needs of the next decade in the pattern previously anticipated. The costs in the back end of the fuel cycle are much higher than had been anticipated several years ago, a situation similar to that of almost all large endeavors in this country. On the other hand, the added costs are small relative to total power costs and do not affect the economic advantage of nuclear power as compared to other power sources. A rough economic analysis indicates that the question for the back end of the fuel cycle has changed from one of optimizing profitability to one of determining the most economic disposition of spent fuel. Long-term spent fuel storage is a practical and economically acceptable way to provide time for determining a sound course of action for the back end of the fuel cycle. Indeed, if one could count on a breeder economy before the end of the century, one possible course of action is to store light-water fuel until the plutonium can be used in breeders. However, for philosophical as well as practical reasons, it is important that the uncertainties in the course of action should be resolved as quickly as possible. Long-term storage should not be an excuse to delay resolution of the basic questions. (U.S.)

  14. Nuclear fuel cycle scenarios at CGNPC

    International Nuclear Information System (INIS)

    Xiao, Min; Zhou, Zhou; Nie, Li Hong; Mao, Guo Ping; Hao, Si Xiong; Shen, Kang

    2008-01-01

    Established in 1994, China Guangdong Nuclear Power Holding Co. (CGNPC) now owns two power stations GNPS and LNPS Phase I, with approximate 4000 MWe of installed capacity. With plant upgrades, advanced fuel management has been introduced into the two plants to improve the plant economical behavior with the high burnup fuel implemented. For the purpose of sustainable development, some preliminary studies on nuclear fuel cycle, especially on the back-end, have been carried out at CGNPC. According to the nuclear power development plan of China, the timing for operation and the capacity of the reprocessing facility are studied based on the amount of the spent fuel forecast in the future. Furthermore, scenarios of the fuel cycles in the future in China with the next generation of nuclear power were considered. Based on the international experiences on the spent fuel management, several options of spent fuel reprocessing strategies are investigated in detail, for example, MOX fuel recycling in light water reactor, especially in the current reactors of CGNPC, spent fuel intermediated storage, etc. All the investigations help us to draw an overall scheme of the nuclear fuel cycle, and to find a suitable road-map to achieve the sustainable development of nuclear power. (authors)

  15. Regional nuclear fuel cycle centers study project

    International Nuclear Information System (INIS)

    Bennett, L.; Catlin, R.G.; Meckoni, V.

    1977-01-01

    The concept of regional fuel cycle centers (RFCC) has attracted wide interest. The concept was endorsed by many countries in discussions at the General Conference of the International Atomic Energy Agency and at the General Assembly of the United Nations. Accordingly, in 1975, the IAEA initiated a detailed study of the RFCC concept. The Agency study has concentrated on what is referred to as the ''back-end'' of the fuel cycle because that is the portion which is currently problematic. The study covers transport, storage, processing and recycle activities starting from the time the spent fuel leaves the reactor storage pools and through all steps until the recycled fuel is in finished fuel elements and shipped to the reactor. A detailed evaluation of the specific features of large regional fuel cycle centers established on a multinational basis vis-a-vis smaller dispersed fuel cycle facilities set up on a national basis has been carried out. The methodology for assessment of alternative strategies for fuel storage, reprocessing, and recycling of plutonium has been developed, characteristic data on material flows and cost factors have been generated, and an analytic system has been developed to carry out such evaluations including appropriate sensitivity analysis. Studies in related areas on institutional and legal, organizational, environmental, materials control and other essential aspects have also been made. The material developed during the course of this Study would enable any group of interested Member States to examine and work out alternative strategies pertinent to their present and projected nuclear fuel cycle needs, as well as evolve institutional, legal and other appropriate frameworks or agreements for the establishment of fuel cycle centers on a multinational cooperative basis

  16. International Nuclear Fuel Cycle Fact Book

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, I.W.; Patridge, M.D.

    1991-05-01

    As the US Department of Energy (DOE) and DOE contractors have become increasingly involved with other nations in nuclear fuel cycle and waste management cooperative activities, a need has developed for a ready source of information concerning foreign fuel cycle programs, facilities, and personnel. This Fact Book was compiled to meet that need. The information contained in the International Nuclear Fuel Cycle Fact Book has been obtained from many unclassified sources: nuclear trade journals and newsletters; reports of foreign visits and visitors; CEC, IAEA, and OECN/NEA activities reports; not reflect any one single source but frequently represent a consolidation/combination of information.

  17. Fuel cycle cost comparisons with oxide and silicide fuels

    Energy Technology Data Exchange (ETDEWEB)

    Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

    1983-09-01

    This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data is presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

  18. The IFR modern nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hannum, W.H.

    1991-01-01

    Nuclear power is an essential component of the world's energy supply. The IFR program, by returning to fundamentals, offers a fresh approach to closing the nuclear fuel cycle. This closed fuel cycle represents the ultimate in efficient resource utilization and environmental accountability. 35 refs., 2 tabs

  19. Implications of alternative fuel cycles

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The United States is re-examining alternative fuel cycles and nuclear power strategies, and doubtful attempts are being made to justify the economics of the 'throw-away' fuel cycle. At an international forum on 'An acceptable nuclear energy future for the world' at Fort Lauderdale, Karl Cohen of General Electric and a leading authority on this topic put the implications into perspective. Extracts from his address are presented

  20. Gadolinia experience and design for PWR fuel cycles

    International Nuclear Information System (INIS)

    Stephenson, L. C.

    2000-01-01

    The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with