The octopus burnup and criticality code system
International Nuclear Information System (INIS)
Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.
1996-01-01
The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)
The OCTOPUS burnup and criticality code system
International Nuclear Information System (INIS)
Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de
1996-06-01
The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)
Integrated burnup calculation code system SWAT
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Hirakawa, Naohiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwasaki, Tomohiko
1997-11-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. It enables us to analyze the burnup problem using neutron spectrum depending on environment of irradiation, combining SRAC which is Japanese standard thermal reactor analysis code system and ORIGEN2 which is burnup code widely used all over the world. SWAT makes effective cross section library based on results by SRAC, and performs the burnup analysis with ORIGEN2 using that library. SRAC and ORIGEN2 can be called as external module. SWAT has original cross section library on based JENDL-3.2 and libraries of fission yield and decay data prepared from JNDC FP Library second version. Using these libraries, user can use latest data in the calculation of SWAT besides the effective cross section prepared by SRAC. Also, User can make original ORIGEN2 library using the output file of SWAT. This report presents concept and user`s manual of SWAT. (author)
Systemization of burnup sensitivity analysis code. 2
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2005-02-01
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For
Accuracy assessment of a new Monte Carlo based burnup computer code
International Nuclear Information System (INIS)
El Bakkari, B.; ElBardouni, T.; Nacir, B.; ElYounoussi, C.; Boulaich, Y.; Meroun, O.; Zoubair, M.; Chakir, E.
2012-01-01
Highlights: ► A new burnup code called BUCAL1 was developed. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► Validation of BUCAL1 was done by code to code comparison using VVER-1000 LEU Benchmark Assembly. ► Differences from BM value were found to be ± 600 pcm for k ∞ and ±6% for the isotopic compositions. ► The effect on reactivity due to the burnup of Gd isotopes is well reproduced by BUCAL1. - Abstract: This study aims to test for the suitability and accuracy of a new home-made Monte Carlo burnup code, called BUCAL1, by investigating and predicting the neutronic behavior of a “VVER-1000 LEU Assembly Computational Benchmark”, at lattice level. BUCAL1 uses MCNP tally information directly in the computation; this approach allows performing straightforward and accurate calculation without having to use the calculated group fluxes to perform transmutation analysis in a separate code. ENDF/B-VII evaluated nuclear data library was used in these calculations. Processing of the data library is performed using recent updates of NJOY99 system. Code to code comparisons with the reported Nuclear OECD/NEA results are presented and analyzed.
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
International Nuclear Information System (INIS)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi; Kaneko, Toshiyuki.
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code 'MULTI-KENO' and the routine for the burnup calculation of the one dimensional burnup code 'UNITBURN'. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Burnup code for fuel assembly by Monte Carlo code. MKENO-BURN
Energy Technology Data Exchange (ETDEWEB)
Naito, Yoshitaka; Suyama, Kenya; Masukawa, Fumihiro; Matsumoto, Kiyoshi; Kurosawa, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Toshiyuki
1996-12-01
The evaluation of neutron spectrum is so important for burnup calculation of the heterogeneous geometry like recent BWR fuel assembly. MKENO-BURN is a multi dimensional burnup code that based on the three dimensional monte carlo neutron transport code `MULTI-KENO` and the routine for the burnup calculation of the one dimensional burnup code `UNITBURN`. MKENO-BURN analyzes the burnup problem of arbitrary regions after evaluating the neutron spectrum and making one group cross section in three dimensional geometry with MULTI-KENO. It enables us to do three dimensional burnup calculation. This report consists of general description of MKENO-BURN and the input data. (author)
Burnup calculation code system COMRAD96
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu
1997-06-01
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, `Cross Section Treatment`, `Generation and Depletion Calculation`, and `Post Process`. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the {gamma} Spectrum on a terminal. This report is the general description and user`s manual of COMRAD96. (author)
Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
Suyama, Kenya; Masukawa, Fumihiro; Ido, Masaru; Enomoto, Masaki; Takyu, Shuiti; Hara, Toshiharu.
1997-06-01
COMRAD was one of the burnup code system developed by JAERI. COMRAD96 is a transfered version of COMRAD to Engineering Work Station. It is divided to several functional modules, 'Cross Section Treatment', 'Generation and Depletion Calculation', and 'Post Process'. It enables us to analyze a burnup problem considering a change of neutron spectrum using UNITBURN. Also it can display the γ Spectrum on a terminal. This report is the general description and user's manual of COMRAD96. (author)
High burnup models in computer code fair
International Nuclear Information System (INIS)
Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.
1997-01-01
An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs
Fuel analysis code FAIR and its high burnup modelling capabilities
International Nuclear Information System (INIS)
Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.
1995-01-01
A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs
Systemization of burnup sensitivity analysis code (2) (Contract research)
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2008-08-01
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion
Revised SWAT. The integrated burnup calculation code system
Energy Technology Data Exchange (ETDEWEB)
Suyama, Kenya; Mochizuki, Hiroki [Department of Fuel Cycle Safety Research, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Kiyosumi, Takehide [The Japan Research Institute, Ltd., Tokyo (Japan)
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Revised SWAT. The integrated burnup calculation code system
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki; Kiyosumi, Takehide
2000-07-01
SWAT is an integrated burnup code system developed for analysis of post irradiation examination, transmutation of radioactive waste, and burnup credit problem. This report shows an outline and a user's manual of revised SWAT. This revised SWAT includes expansion of functions, increasing supported machines, and correction of several bugs reported from users of previous SWAT. (author)
Restructuring of burnup sensitivity analysis code system by using an object-oriented design approach
International Nuclear Information System (INIS)
Kenji, Yokoyama; Makoto, Ishikawa; Masahiro, Tatsumi; Hideaki, Hyoudou
2005-01-01
A new burnup sensitivity analysis code system was developed with help from the object-oriented technique and written in Python language. It was confirmed that they are powerful to support complex numerical calculation procedure such as reactor burnup sensitivity analysis. The new burnup sensitivity analysis code system PSAGEP was restructured from a complicated old code system and reborn as a user-friendly code system which can calculate the sensitivity coefficients of the nuclear characteristics considering multicycle burnup effect based on the generalized perturbation theory (GPT). A new encapsulation framework for conventional codes written in Fortran was developed. This framework supported to restructure the software architecture of the old code system by hiding implementation details and allowed users of the new code system to easily calculate the burnup sensitivity coefficients. The framework can be applied to the other development projects since it is carefully designed to be independent from PSAGEP. Numerical results of the burnup sensitivity coefficient of a typical fast breeder reactor were given with components based on GPT and the multicycle burnup effects on the sensitivity coefficient were discussed. (authors)
SWAT3.1 - the integrated burnup code system driving continuous energy Monte Carlo codes MVP and MCNP
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki; Takada, Tomoyuki; Ryufuku, Susumu; Okuno, Hiroshi; Murazaki, Minoru; Ohkubo, Kiyoshi
2009-05-01
Integrated burnup calculation code system SWAT is a system that combines neutronics calculation code SRAC,which is widely used in Japan, and point burnup calculation code ORIGEN2. It has been used to evaluate the composition of the uranium, plutonium, minor actinides and the fission products in the spent nuclear fuel. Based on this idea, the integrated burnup calculation code system SWAT3.1 was developed by combining the continuous energy Monte Carlo code MVP and MCNP, and ORIGEN2. This enables us to treat the arbitrary fuel geometry and to generate the effective cross section data to be used in the burnup calculation with few approximations. This report describes the outline, input data instruction and several examples of the calculation. (author)
Methods of RECORD, an LWR fuel assembly burnup code
International Nuclear Information System (INIS)
Skardhamar, T.; Naess, H.K.
1982-06-01
The RECORD computer code is a detailed rector physics code for performing efficient LWR fuel assembly calculations, taking into account most of the features found in BWR and PWR fuel designs. The code calculates neutron spectrum, reaction rates and reactivity as a function of fuel burnup, and it generates the few-group data required for use in full scale core simulation and fuel management calculations. The report describes the methods of the RECORD computer code and the basis for fundamental models selected, and gives a review of code qualifications against measured data. (Auth. /RF)
BEAVRS full core burnup calculation in hot full power condition by RMC code
International Nuclear Information System (INIS)
Liu, Shichang; Liang, Jingang; Wu, Qu; Guo, JuanJuan; Huang, Shanfang; Tang, Xiao; Li, Zeguang; Wang, Kan
2017-01-01
Highlights: • TMS and thermal scattering interpolation were developed to treat cross sections OTF. • Hybrid coupling system was developed for HFP burnup calculation of BEAVRS benchmark. • Domain decomposition was applied to handle memory problem of full core burnup. • Critical boron concentration with burnup by RMC agrees with the benchmark results. • RMC is capable of multi-physics coupling for simulations of nuclear reactors in HFP. - Abstract: Monte Carlo method can provide high fidelity neutronics analysis of different types of nuclear reactors, owing to its advantages of the flexible geometry modeling and the use of continuous-energy nuclear cross sections. However, nuclear reactors are complex systems with multi-physics interacting and coupling. MC codes can couple with depletion solver and thermal-hydraulics (T/H) codes simultaneously for the “transport-burnup-thermal-hydraulics” coupling calculations. MIT BEAVRS is a typical “transport-burnup-thermal-hydraulics” coupling benchmark. In this paper, RMC was coupled with sub-channel code COBRA, equipped with on-the-fly temperature-dependent cross section treatment and large-scale detailed burnup calculation based on domain decomposition. Then RMC was applied to the full core burnup calculations of BEAVRS benchmark in hot full power (HFP) condition. The numerical tests show that domain decomposition method can achieve the consistent results compared with original version of RMC while enlarging the computational burnup regions. The results of HFP by RMC agree well with the reference values of BEAVRS benchmark and also agree well with those of MC21. This work proves the feasibility and accuracy of RMC in multi-physics coupling and lifecycle simulations of nuclear reactors.
Evolution of the ELESTRES code for application to extended burnups
International Nuclear Information System (INIS)
Tayal, M.; Ranger, A.; Singhal, N.; Mak, R.
1990-01-01
The computer code ELESTRES is frequently used at Atomic Energy of Canada Limited to assess the integrity of CANDU fuel under normal operating conditions. The code also provides initial conditions for evaluating fuel behaviour during high-temperature transients. This paper describes recent improvements in the code in the areas of pellet expansion and of fission gas release. Both of these are very important considerations in ensuring fuel integrity at extended burnups. Firstly, in calculations of pellet expansion, the code now accounts for the effect of thermal stresses on the volume of gas bubbles at the boundaries of UO 2 grains. This has a major influence on the expansion of the pellet during power-ramps. Secondly, comparisons with data showed that the previous fission gas package significantly underpredicted the fission gas release at high burnups. This package has now been improved via modifications to the following modules: distance between neighbouring bubbles on grain boundaries; diffusivity; and thermal conductivity. The predictions of the revised version of the code show reasonable agreement with measurements of ridge strains and of fission gas release. An illustrative example demonstrates that the code can be used to identify a fuel design that would: reduce the sheath stresses at circumferential ridges by a factor of 2-10; and keep the gas pressure at very high burnups to below the coolant pressure
Directory of Open Access Journals (Sweden)
Gholamzadeh Zohreh
2014-12-01
Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view
A high burnup model developed for the DIONISIO code
Soba, A.; Denis, A.; Romero, L.; Villarino, E.; Sardella, F.
2013-02-01
A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO2 fuels in LWR conditions, predict the radial distribution of power density, burnup, and concentration of diverse nuclides within the pellet. The balance equations of all the isotopes involved in the fission process are solved in a simplified manner, and the one-group effective cross sections of all of them are obtained as functions of the radial position in the pellet, burnup, and enrichment in 235U. In this work, the subroutines are described and the results of the simulations performed with DIONISIO are presented. The good agreement with the data provided in the FUMEX II/III NEA data bank can be easily recognized.
LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory
International Nuclear Information System (INIS)
Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.
1976-04-01
The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
Korkmaz, M.E.; Agar, O.; Yigit, M.
2013-01-01
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed 232 Th and mixed 233 U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
Burnup calculations using serpent code in accelerator driven thorium reactors
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, M.E.; Agar, O. [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Physics Dept.; Yigit, M. [Aksaray Univ. (Turkey). Physics Dept.
2013-07-15
In this study, burnup calculations have been performed for a sodium cooled Accelerator Driven Thorium Reactor (ADTR) using the Serpent 1.1.16 Monte Carlo code. The ADTR has been designed for burning minor actinides, mixed {sup 232}Th and mixed {sup 233}U fuels. A solid Pb-Bi spallation target in the center of the core is used and sodium as coolant. The system is designed for a heating power of 2 000 MW and for an operation time of 600 days. For burnup calculations the Advanced Matrix Exponential Method CRAM (Chebyshev Rational Approximation Method) and different nuclear data libraries (ENDF7, JEF2.2, JEFF3.1.1) were used. The effective multiplication factor change from 0.93 to 0.97 for different nuclear data libraries during the reactor operation period. (orig.)
A multi-platform linking code for fuel burnup and radiotoxicity analysis
Cunha, R.; Pereira, C.; Veloso, M. A. F.; Cardoso, F.; Costa, A. L.
2014-02-01
A linking code between ORIGEN2.1 and MCNP has been developed at the Departamento de Engenharia Nuclear/UFMG to calculate coupled neutronic/isotopic results for nuclear systems and to produce a large number of criticality, burnup and radiotoxicity results. In its previous version, it evaluated the isotopic composition evolution in a Heat Pipe Power System model as well as the radiotoxicity and radioactivity during lifetime cycles. In the new version, the code presents features such as multi-platform execution and automatic results analysis. Improvements made in the code allow it to perform simulations in a simpler and faster way without compromising accuracy. Initially, the code generates a new input for MCNP based on the decisions of the user. After that, MCNP is run and data, such as recoverable energy per prompt fission neutron, reaction rates and keff, are automatically extracted from the output and used to calculate neutron flux and cross sections. These data are then used to construct new ORIGEN inputs, one for each cell in the core. Each new input is run on ORIGEN and generates outputs that represent the complete isotopic composition of the core on that time step. The results show good agreement between GB (Coupled Neutronic/Isotopic code) and Monteburns (Automated, Multi-Step Monte Carlo Burnup Code System), developed by the Los Alamos National Laboratory.
Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki
1997-11-01
EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.
ALEPH 1.1.2: A Monte Carlo burn-up code
International Nuclear Information System (INIS)
Haeck, W.; Verboomen, B.
2006-01-01
In the last 40 years, Monte Carlo particle transport has been applied to a multitude of problems such as shielding and medical applications, to various types of nuclear reactors, . . . The success of the Monte Carlo method is mainly based on its broad application area, on its ability to handle nuclear data not only in its most basic but also most complex form (namely continuous energy cross sections, complex interaction laws, detailed energy-angle correlations, multi-particle physics, . . . ), on its capability of modeling geometries from simple 1D to complex 3D, . . . There is also a current trend in Monte Carlo applications toward high detail 3D calculations (for instance voxel-based medical applications), something for which deterministic codes are neither suited nor performant as to computational time and precision. Apart from all these fields where Monte Carlo particle transport has been applied successfully, there is at least one area where Monte Carlo has had limited success, namely burn-up and activation calculations where the time parameter is added to the problem. The concept of Monte Carlo burn-up consists of coupling a Monte Carlo code to a burn-up module to improve the accuracy of depletion and activation calculations. For every time step the Monte Carlo code will provide reaction rates to the burn-up module which will return new material compositions to the Monte Carlo code. So if static Monte Carlo particle transport is slow, then Monte Carlo particle transport with burn-up will be even slower as calculations have to be performed for every time step in the problem. The computational issues to perform accurate Monte Carlo calculations are however continuously reduced due to improvements made in the basic Monte Carlo algorithms, due to the development of variance reduction techniques and due to developments in computer architecture (more powerful processors, the so-called brute force approach through parallel processors and networked systems
Modelling of high burnup structure in UO2 fuel with the RTOP code
International Nuclear Information System (INIS)
Likhanskii, V.; Zborovskii, V.; Evdokimov, I.; Kanyukova, V.; Sorokin, A.
2008-01-01
The present work deals with self-consistent physical approach aimed to derive the criterion of fuel restructuring avoiding correlations. The approach is based on study of large over pressurized bubbles formation on dislocations, at grain boundaries and in grain volume. At first, stage of formation of bubbles non-destroyable by fission fragments is examined using consistent modelling of point defects and fission gas behavior near dislocation and in grain volume. Then, evolution of formed large non-destroyable bubbles is considered using results of the previous step as initial values. Finally, condition of dislocation loops punching by sufficiently large over pressurized bubbles is regarded as the criterion of fuel restructuring onset. In the present work consideration of large over pressurized bubbles evolution is applied to modelling of the restructuring threshold depending on temperature, burnup and grain size. Effect of grain size predicted by the model is in qualitative agreement with experimental observations. Restructuring threshold criterion as an analytical function of local burnup and fuel temperature is derived and compared with HBRP project data. To predict rim-layer width formation depending on fuel burnup and irradiation conditions the model is implemented into the mechanistic fuel performance code RTOP. Calculated dependencies give upper estimate for the width of restructured region. Calculations show that one needs to consider temperature distribution within pellet which depends on irradiation history in order to model rim-structure formation
Fuel burnup calculation of Ghana MNSR using ORIGEN2 and REBUS3 codes.
Abrefah, R G; Nyarko, B J B; Fletcher, J J; Akaho, E H K
2013-10-01
Ghana Research Reactor-1 core is to be converted from HEU fuel to LEU fuel in the near future and managing the spent nuclear fuel is very important. A fuel depletion analysis of the GHARR-1 core was performed using ORIGEN2 and REBUS3 codes to estimate the isotopic inventory at end-of-cycle in order to help in the design of an appropriate spent fuel cask. The results obtained for both codes were consistent for U-235 burnup weight percent and Pu-239 build up as a result of burnup. Copyright © 2013 Elsevier Ltd. All rights reserved.
New high burnup fuel models for NRC's licensing audit code, FRAPCON
International Nuclear Information System (INIS)
Lanning, D.D.; Beyer, C.E.; Painter, C.L.
1996-01-01
Fuel behavior models have recently been updated within the U.S. Nuclear Regulatory Commission steady-state FRAPCON code used for auditing of fuel vendor/utility-codes and analyses. These modeling updates have concentrated on providing a best estimate prediction of steady-state fuel behavior up to the maximum burnup level s of current data (60 to 65 GWd/MTU rod-average). A decade has passed since these models were last updated. Currently, some U.S. utilities and fuel vendors are requesting approval for rod-average burnups greater than 60 GWd/MTU; however, until these recent updates the NRC did not have valid fuel performance models at these higher burnup levels. Pacific Northwest Laboratory (PNL) has reviewed 15 separate effects models within the FRAPCON fuel performance code (References 1 and 2) and identified nine models that needed updating for improved prediction of fuel behavior at high burnup levels. The six separate effects models not updated were the cladding thermal properties, cladding thermal expansion, cladding creepdown, fuel specific heat, fuel thermal expansion and open gap conductance. Comparison of these models to the currently available data indicates that these models still adequately predict the data within data uncertainties. The nine models identified as needing improvement for predicting high-burnup behavior are fission gas release (FGR), fuel thermal conductivity (accounting for both high burnup effects and burnable poison additions), fuel swelling, fuel relocation, radial power distribution, fuel-cladding contact gap conductance, cladding corrosion, cladding mechanical properties and cladding axial growth. Each of the updated models will be described in the following sections and the model predictions will be compared to currently available high burnup data
Validation of a new continuous Monte Carlo burnup code using a Mox fuel assembly
International Nuclear Information System (INIS)
El bakkari, B.; El Bardouni, T.; Merroun, O.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Chakir, E.
2009-01-01
The reactivity of nuclear fuel decreases with irradiation (or burnup) due to the transformation of heavy nuclides and the formation of fission products. Burnup credit studies aim at accounting for fuel irradiation in criticality studies of the nuclear fuel cycle (transport, storage, etc...). The principal objective of this study is to evaluate the potential capabilities of a newly developed burnup code called 'BUCAL1'. BUCAL1 differs in comparison with other burnup codes as it does not use the calculated neutron flux as input to other computer codes to generate the nuclide inventory for the next time step. Instead, BUCAL1 directly uses the neutron reaction tally information generated by MCNP for each nuclide of interest to determine the new nuclides inventory. This allows the full capabilities of MCNP to be incorporated into the calculation and a more accurate and robust analysis to be performed. Validation of BUCAL1 was processed by code-to-code comparisons using predictions of several codes from the NEA/OCED. Infinite multiplication factors (k ∞ ) and important fission product and actinide concentrations were compared for a MOX core benchmark exercise. Results of calculations are analysed and discussed.
Energy Technology Data Exchange (ETDEWEB)
Behler, Matthias; Hannstein, Volker; Kilger, Robert; Moser, Franz-Eberhard; Pfeiffer, Arndt; Stuke, Maik
2014-06-15
In order to account for the reactivity-reducing effect of burn-up in the criticality safety analysis for systems with irradiated nuclear fuel (''burnup credit''), numerical methods to determine the enrichment and burnup dependent nuclide inventory (''burnup code'') and its resulting multiplication factor k{sub eff} (''criticality code'') are applied. To allow for reliable conclusions, for both calculation systems the systematic deviations of the calculation results from the respective true values, the bias and its uncertainty, are being quantified by calculation and analysis of a sufficient number of suitable experiments. This quantification is specific for the application case under scope and is also called validation. GRS has developed a methodology to validate a calculation system for the application of burnup credit in the criticality safety analysis for irradiated fuel assemblies from pressurized water reactors. This methodology was demonstrated by applying the GRS home-built KENOREST burnup code and the criticality calculation sequence CSAS5 from SCALE code package. It comprises a bounding approach and alternatively a stochastic, which both have been exemplarily demonstrated by use of a generic spent fuel pool rack and a generic dry storage cask, respectively. Based on publicly available post irradiation examination and criticality experiments, currently the isotopes of uranium and plutonium elements can be regarded for.
New approach to derive linear power/burnup history input for CANDU fuel codes
International Nuclear Information System (INIS)
Lac Tang, T.; Richards, M.; Parent, G.
2003-01-01
The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)
Development of base technology for high burnup PWR fuel improvement Volume 1 and 2
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Eun; Lee, Sang Hee; Bae, Seong Man [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Chung, Jin Gon; Chung, Sun Kyo; Kim, Sun Du [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Kim, Jae Won; Chung, Sun Kyo; Kim, Sun Du [Korea Nuclear Fuel Development Inst., Seoul (Korea, Republic of)
1995-12-31
Development of base technology for high burnup nuclear fuel -Development of UO{sub 2} pellet manufacturing technology -Improvement of fuel rod performance code -Improvement of plenum spring design -Study on the mechanical characteristics of fuel cladding -Organization of fuel failure mechanism Establishment of next stage R and D program (author). 226 refs., 100 figs.
Analysis of recent post irradiation tests by Japanese and French burnup analysis code systems
International Nuclear Information System (INIS)
Iwasaki, Tomohiko; Hiraizumi, Hiroaki; Youinou, Gilles
2002-01-01
Benchmark problem based on Japanese Post Irradiation Experiment (PIE) data was analyzed by Japanese burnup analysis code and French one under the cooperative research program between the Japanese University Association (JUA) in Japan and Commissariat a l'Enegie Atomique (CEA) in France. Significant discrepancies over 10% were found between the Japanese and French results for 238 Pu, 243 Am, 244 Cm, 125 Sb, 154 Eu, 134 Cs and 144 Ce. It is supposed that the difference of C/E for 243 Am and 244 Cm between Japanese results and French ones is due to the (n,gamma) reaction of 242m Am. For 125 Sb and 154 Eu, the C/E values are improved by using new cross section and fission yield libraries. (author)
Development of a Burnup Module DECBURN Based on the Krylov Subspace Method
Energy Technology Data Exchange (ETDEWEB)
Cho, J. Y.; Kim, K. S.; Shim, H. J.; Song, J. S
2008-05-15
This report is to develop a burnup module DECBURN that is essential for the reactor analysis and the assembly homogenization codes to trace the fuel composition change during the core burnup. The developed burnup module solves the burnup equation by the matrix exponential method based on the Krylov Subspace method. The final solution of the matrix exponential is obtained by the matrix scaling and squaring method. To develop DECBURN module, this report includes the followings as: (1) Krylov Subspace Method for Burnup Equation, (2) Manufacturing of the DECBURN module, (3) Library Structure Setup and Library Manufacturing, (4) Examination of the DECBURN module, (5) Implementation to the DeCART code and Verification. DECBURN library includes the decay constants, one-group cross section and the fission yields. Examination of the DECBURN module is performed by manufacturing a driver program, and the results of the DECBURN module is compared with those of the ORIGEN program. Also, the implemented DECBURN module to the DeCART code is applied to the LWR depletion benchmark and a OPR-1000 pin cell problem, and the solutions are compared with the HELIOS code to verify the computational soundness and accuracy. In this process, the criticality calculation method and the predictor-corrector scheme are introduced to the DeCART code for a function of the homogenization code. The examination by a driver program shows that the DECBURN module produces exactly the same solution with the ORIGEN program. DeCART code that equips the DECBURN module produces a compatible solution to the other codes for the LWR depletion benchmark. Also the multiplication factors of the DeCART code for the OPR-1000 pin cell problem agree to the HELIOS code within 100 pcm over the whole burnup steps. The multiplication factors with the criticality calculation are also compatible with the HELIOS code. These results mean that the developed DECBURN module works soundly and produces an accurate solution
iBEST: a program for burnup history estimation of spent fuels based on ORIGEN-S
Directory of Open Access Journals (Sweden)
Do-Yeon Kim
2015-08-01
Full Text Available In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator, that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems.
iBEST: a program for burnup history estimation of spent fuels based on ORIGEN-S
International Nuclear Information System (INIS)
Kim, Do Yeon; Hong, Ser Gi; Ahn, Gil Hoon
2015-01-01
In this paper, we describe a computer program, iBEST (inverse Burnup ESTimator), that we developed to accurately estimate the burnup histories of spent nuclear fuels based on sample measurement data. The burnup history parameters include initial uranium enrichment, burnup, cooling time after discharge from reactor, and reactor type. The program uses algebraic equations derived using the simplified burnup chains of major actinides for initial estimations of burnup and uranium enrichment, and it uses the ORIGEN-S code to correct its initial estimations for improved accuracy. In addition, we newly developed a stable bisection method coupled with ORIGEN-S to correct burnup and enrichment values and implemented it in iBEST in order to fully take advantage of the new capabilities of ORIGEN-S for improving accuracy. The iBEST program was tested using several problems for verification and well-known realistic problems with measurement data from spent fuel samples from the Mihama-3 reactor for validation. The test results show that iBEST accurately estimates the burnup history parameters for the test problems and gives an acceptable level of accuracy for the realistic Mihama-3 problems
Development of a fuel rod thermal-mechanical analysis code for high burnup fuel
International Nuclear Information System (INIS)
Owaki, M.; Ikatsu, N.; Ohira, K.; Itagaki, N.
2001-01-01
The thermal-mechanical analysis code for high burnup BWR fuel rod has been developed by NFI. The irradiation data accumulated up to the assembly burnup of 55 GWd/t in commercial BWRs were adopted for the modeling. In the code, pellet thermal conductivity degradation with burnup progress was considered. Effects of the soluble FPs, irradiation defects and porosity increase due to RIM effect were taken into the model. In addition to the pellet thermal conductivity degradation, the pellet swelling due to the RIM porosity was studied. The modeling for the high burnup effects was also carried out for (U, Gd)O 2 and MOX fuel. The thermal conductivities of all pellet types, UO 2 , (U, Gd)O 2 and (U, Pu)O 2 pellets, are expressed by the same form of equation with individual coefficient γ in the code. The pellet center temperature was calculated using this modeling code, and compared with measured values for the code verification. The pellet center temperature calculated using the thermal conductivity degradation model agreed well with the measured values within ±150 deg. C. The influence of rim porosity on pellet center temperature is small, and the temperature increase in only 30 deg. C at 75 GWd/t and 200 W/cm. The pellet center temperature of MOX fuel was also calculated, and it was found that the pellet center temperature of MOX fuel with 10wt% PuO 2 is about 60 deg. C higher than UO 2 fuel at 75 GWd/t and 200 W/cm. (author)
Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes
Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.
2017-02-01
International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.
MONTEBURNS 2.0: An Automated, Multi-Step Monte Carlo Burnup Code System
International Nuclear Information System (INIS)
2007-01-01
A - Description of program or function: MONTEBURNS Version 2 calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. MONTEBURNS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with a radioactive decay and burnup code. Highlights on changes to Version 2 are listed in the transmittal letter. Along with other minor improvements in MONTEBURNS Version 2, the option was added to use CINDER90 instead of ORIGEN2 as the depletion/decay part of the system. CINDER90 is a multi-group depletion code developed at LANL and is not currently available from RSICC, nor from the NEA Databank. This MONTEBURNS release was tested with various combinations of CCC-715/MCNPX 2.4.0, CCC-710/MCNP5, CCC-700/MCNP4C, CCC-371/ORIGEN2.2, ORIGEN2.1 and CINDER90. Perl is required software and is not included in this distribution. MCNP, ORIGEN2, and CINDER90 are not included. The following changes have been made: 1) An increase in the number of removal group information that must be provided for each material in each step in the feed input file. 2) The capability to use CINDER90 instead of ORIGEN2.1 as the depletion/decay part of the code. 3) ORIGEN2.2 can also be used instead of ORIGEN2.1 in Monteburns. 4) The correction of including the capture cross sections to metastable as well as ground states if applicable for an isotope (i.e. Am-241 and Am-243 in particular). 5) The ability to use a MCNP input file that has a title card starting with 'm' (this was a bug in the first version of Monteburns). 6) A decrease in run time for cases involving decay-only steps (power of 0.0). Monteburns does not run MCNP to calculate cross sections for a step unless it is an irradiation step. 7) The ability to change the cross section libraries used each step. If different cross section libraries are desired for multiple steps. 8
International Nuclear Information System (INIS)
Serov, I.V.; Hoogenboom, J.E.
1993-07-01
The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or 'BuCo' material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP)
International Nuclear Information System (INIS)
Cetnar, Jerzy
2014-01-01
The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)
Calculation of effect of burnup history on spent fuel reactivity based on CASMO5
International Nuclear Information System (INIS)
Li Xiaobo; Xia Zhaodong; Zhu Qingfu
2015-01-01
Based on the burnup credit of actinides + fission products (APU-2) which are usually considered in spent fuel package, the effect of power density and operating history on k ∞ was studied. All the burnup calculations are based on the two-dimensional fuel assembly burnup program CASMO5. The results show that taking the core average power density of specified power plus a bounding margin of 0.0023 to k ∞ , and taking the operating history of specified power without shutdown during cycle and between cycles plus a bounding margin of 0.0045 to k ∞ can meet the bounding principle of burnup credit. (authors)
A simple formula for local burnup based on constant relative reaction rate per nuclei
Yuan, Cenxi; Wang, Xuming; Chen, Shengli
2015-01-01
A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of $^{235}$U, $^{238}$U, and $^{239}$Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reactio...
Shi, Xue-Ming; Peng, Xian-Jue
2016-09-01
Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.
Energy Technology Data Exchange (ETDEWEB)
Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Han, Sun Ho; Ha, Yeong Keong; Song, Kyu Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2008-11-15
Destructive methods were used for the burnup determination of a PWR nuclear fuel irradiated to a high burnup in power reactors. The total burnup was determined from a measurement of the Nd and Cs isotope burnup monitors. The methods included U, Pu, {sup 148}Nd, {sup 145}Nd+{sup 146}Nd, total of the Nd isotopes, and {sup 133}Cs and {sup 137}Cs determinations by the isotope dilution mass spectrometric method (IDMS) by using quadrupole spikes ({sup 233}U, {sup 242}Pu, {sup 150}Nd and {sup 133}Cs). The methods involved two sequential anion exchange resin (AG 1X8 and 1X4) separation procedures and a Cs purification with a cation exchange resin (AG 50WX4) separation procedure. The effective fission yield was calculated from the weighted fission yields averaged over the irradiation period. The results obtained by the Nd and Cs isotopes from the mass spectrometric measurement were compared with those by the ORIGEN code.
A comparison study of the 1MeV triton burn-up in JET using the HECTOR and SOCRATE codes
International Nuclear Information System (INIS)
Gorini, G.; Kovanen, M.A.
1988-01-01
The burn-up of the 1MeV tritons in deuterium plasmas has been measured in JET for various plasma conditions. To interpret these measurements the containment, slowing down and burn-up of fast tritons needs to be modelled with a reasonable accuracy. The numerical code SOCRATE has been written for this specific purpose and a second code, HECTOR, has been adapted to study the triton burn-up problem. In this paper we compare the results from the two codes in order to exclude possible errors in the numerical models, to assess their accuracy and to study the sensitivity of the calculation to various physical effects. (author)
Burnup analysis of the VVER-1000 reactor using thorium-based fuel
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, Mehmet E.; Agar, Osman; Bueyueker, Eylem [Karamanoglu Mehmetbey Univ., Karaman (Turkey). Faculty of Kamil Ozdag Science
2014-12-15
This paper aims to investigate {sup 232}Th/{sup 233}U fuel cycles in a VVER-1000 reactor through calculation by computer. The 3D core geometry of VVER-1000 system was designed using the Serpent Monte Carlo 1.1.19 Code. The Serpent Code using parallel programming interface (Message Passing Interface-MPI), was run on a workstation with 12-core and 48 GB RAM. {sup 232}Th/{sup 235}U/{sup 238}U oxide mixture was considered as fuel in the core, when the mass fraction of {sup 232}Th was increased as 0.05-0.1-0.2-0.3-0.4 respectively, the mass fraction of {sup 238}U equally was decreased. In the system, the calculations were made for 3 000 MW thermal power. For the burnup analyses, the core is assumed to deplete from initial fresh core up to a burnup of 16 MWd/kgU without refuelling considerations. In the burnup calculations, a burnup interval of 360 effective full power days (EFPDs) was defined. According to burnup, the mass changes of the {sup 232}Th, {sup 233}U, {sup 238}U, {sup 237}Np, {sup 239}Pu, {sup 241}Am and {sup 244}Cm were evaluated, and also flux and criticality of the system were calculated in dependence of the burnup rate.
BASHAN: A few-group three-dimensional diffusion code with burnup and fuel management features
International Nuclear Information System (INIS)
Pearce, D.F.
1970-12-01
The diffusion equation for a two or three-dimensional, two-group or multi-group downscatter problem is solved by conventional finite difference techniques. An x-y-z geometry is assumed with an 'in-channel' mesh point representation. Options are available which allow representation of a soluble poison dispersed throughout the reactor, and also absorber rods in specified channels. The power distribution and multiplication factor k eff are calculated and a point rating map is used to advance the irradiation at each mesh point by a specified time-step so that burnup is followed. Fuel changes may be made so that radial shuffling and axial shuffling fuel management schemes can be studies. The code has been written in FORTRAN S2 for an IBM 7030 (STRETCH) computer which, with a fast store of 80,000 locations, allows problems of up to 15,000 mesh points to be dealt with. Conversion to FORTRAN IV for IBM 360 has now been completed. (author)
International Nuclear Information System (INIS)
Suyama, K.; Mochizuki, H.; Okuno, H.; Miyoshi, Y.
2004-01-01
This paper provides validation results of SWAT2, the revised version of SWAT, which is a code system combining point burnup code ORIGEN2 and continuous energy Monte Carlo code MVP, by the analysis of post irradiation examinations (PIEs). Some isotopes show differences of calculation results between SWAT and SWAT2. However, generally, the differences are smaller than the error of PIE analysis that was reported in previous SWAT validation activity, and improved results are obtained for several important fission product nuclides. This study also includes comparison between an assembly and a single pin cell geometry models. (authors)
International Nuclear Information System (INIS)
Korkmaz, Mehmet E.; Agar, Osman
2014-01-01
In this research, we investigated the burnup characteristics and the conversion of fertile 232 Th into fissile 233 U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning 232 Th fuel (fuel pin 1) and 233 U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.
Energy Technology Data Exchange (ETDEWEB)
Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)
2014-06-15
In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.
Directory of Open Access Journals (Sweden)
MEHMET E. KORKMAZ
2014-06-01
Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.
Tag gas burnup based on three-dimensional FTR analysis
International Nuclear Information System (INIS)
Kidman, R.B.
1976-01-01
Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified
International Nuclear Information System (INIS)
Akishina, E.P.; Kostenko, B.F.; Ivanov, V.V.
2003-01-01
A new method of research in spatial structures that result from uranium dioxide burning in nuclear reactors of modern atomic plants is suggested. The method is based on the presentation of images of the mentioned structures in the form of the working field of a cellular automaton (CA). First, it has allowed one to extract some important quantitative characteristics of the structures directly from the micrographs of the uranium fuel surface. Secondly, the CA has been found out to allow one to formulate easily the dynamics of the evolution of the studied structures in terms of such micrograph elements as spots, spots' boundaries, cracks, etc. Relation has been found between the dynamics and some exactly solvable models of the theory of cellular automata, in particular, the Ising model and the vote model. This investigation gives a detailed description of some CA algorithms which allow one to perform the fuel surface image processing and to model its evolution caused by burnup or chemical etching. (author)
Automated generation of burnup chain for reactor analysis applications
International Nuclear Information System (INIS)
Tran, Viet-Phu; Tran, Hoai-Nam; Yamamoto, Akio; Endo, Tomohiro
2017-01-01
This paper presents the development of an automated generation of burnup chain for reactor analysis applications. Algorithms are proposed to reevaluate decay modes, branching ratios and effective fission product (FP) cumulative yields of a given list of important FPs taking into account intermediate reactions. A new burnup chain is generated using the updated data sources taken from the JENDL FP decay data file 2011 and Fission yields data file 2011. The new burnup chain is output according to the format for the SRAC code system. Verification has been performed to evaluate the accuracy of the new burnup chain. The results show that the new burnup chain reproduces well the results of a reference one with 193 fission products used in SRAC. Burnup calculations using the new burnup chain have also been performed based on UO 2 and MOX fuel pin cells and compared with a reference chain th2cm6fp193bp6T.
International Nuclear Information System (INIS)
Min Ku Jeon; Chang Hwa Lee; Jung Hoon Choi; In Hak Cho; Kweon Ho Kang; Hwan-Seo Park; Geun Il Park; Chang Je Park
2013-01-01
The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125 Sb, 125m Te, and 55 Fe are the major sources of radioactivity. On the other hand, 93m Nb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94 Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94 Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository. (author)
Theory analysis and simple calculation of travelling wave burnup scheme
International Nuclear Information System (INIS)
Zhang Jian; Yu Hong; Gang Zhi
2012-01-01
Travelling wave burnup scheme is a new burnup scheme that breeds fuel locally just before it burns. Based on the preliminary theory analysis, the physical imagine was found. Through the calculation of a R-z cylinder travelling wave reactor core with ERANOS code system, the basic physical characteristics of this new burnup scheme were concluded. The results show that travelling wave reactor is feasible in physics, and there are some good features in the reactor physics. (authors)
International Nuclear Information System (INIS)
Valach, M.; Zymak, J.; Svoboda, R.
1997-01-01
This paper presents the development status of the computer codes for the WWER fuel elements thermomechanical behavior modelling under high burnup conditions at the Nuclear Research Institute Rez. The accent is given on the analysis of the results from the parametric calculations, performed by the programmes PIN-W and RODQ2D, rather than on their detailed theoretical description. Several new optional correlations for the UO2 thermal conductivity with degradation effect caused by burnup were implemented into the both codes. Examples of performed calculations document differences between previous and new versions of both programmes. Some recommendations for further development of the codes are given in conclusion. (author). 6 refs, 9 figs
Activity ratio measurement and burnup analysis for high burnup PWR fuels
International Nuclear Information System (INIS)
Sato, Shunsuke; Nauchi, Yasushi; Hayakawa, Takehito; Kimura, Yasuhiko; Suyama, Kenya
2015-01-01
Applying burnup credit to spent fuel transportation and storage system is beneficial. To take burnup credit to criticality safety design for a spent fuel transportation cask and storage rack, the burnup of target fuel assembly based on core management data must be confirmed by experimental methods. Activity ratio method, in which measured the ratio of the activity of a nuclide to that of another, is one of the ways to confirm burnup history. However, there is no public data of gamma-ray spectrum from high burnup fuels and validation of depletion calculation codes is not sufficient in the evaluation of the burnup or nuclide inventories. In this study, applicability evaluation of activity ratio method was carried out for high burnup fuel samples taken from PWR lead use assembly. In the gamma-ray measurement experiments, energy spectrum was taken in the Reactor Fuel Examination Facility (RFEF) of Japan Atomic Energy Agency (JAEA), and 134 Cs/ 137 Cs and 154 Eu/ 137 Cs activity ratio were obtained. With the MVP-BURN code, the activity ratios were calculated by depletion calculation tracing the operation history. As a result, 134 Cs/ 137 Cs and 154 Eu/ 137 Cs activity ratios for UO 2 fuel samples show good agreements and the activity ratio method has good applicability to high burnup fuels. 154 Eu/ 134 Cs activity ratio for Gd 2 O 3 +UO 2 fuels also shows good agreements between calculation results and experimental results as well as the activity ratio for UO 2 fuels. It also becomes clear that it is necessary to pay attention to not only burnup but also axial burnup distribution history when confirming the burnup of UO 2 +Gd 2 O 3 fuel with 134 Cs/ 137 Cs activity ratios. (author)
Estimation of the Fuel Depletion Code Bias and Uncertainty in Burnup-Credit Criticality Analysis
International Nuclear Information System (INIS)
Kim, Jong Woon; Cho, Nam Zin; Lee, Sang Jin; Bae, Chang Yeal
2006-01-01
In the past, criticality safety analyses for commercial light-water-reactor (LWR) spent nuclear fuel (SNF) storage and transportation canisters assumed the spent fuel to be fresh (unirradiated) fuel with uniform isotopic compositions. This fresh-fuel assumption provides a well-defined, bounding approach to the criticality safety analysis that eliminates concerns related to the fuel operating history, and thus considerably simplifies the safety analysis. However, because this assumption ignores the inherent decrease in reactivity as a result of irradiation, it is very conservative. The concept of taking credit for the reduction in reactivity due to fuel burnup is commonly referred to as burnup credit. Implementation of burnup credit requires the computational prediction of the nuclide inventories (compositions) for the dominant fissile and absorbing nuclide species in spent fuel. In addition to that, the bias and uncertainty in the predicted concentration of all nuclides used in the analysis be established by comparisons of calculated and measured radiochemical assay data. In this paper, three methods for considering the bias and uncertainty will be reviewed. The estimated bias and uncertainty that the results of 3rd method are presented
Burnup simulations of different fuel grades using the MCNPX Monte Carlo code
Directory of Open Access Journals (Sweden)
Asah-Opoku Fiifi
2014-01-01
Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.
Husnayani, I.; Udiyani, P. M.; Bakhri, S.; Sunaryo, G. R.
2018-02-01
Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.
Directory of Open Access Journals (Sweden)
Cenxi Yuan
2016-01-01
Full Text Available A simple and analytical formula is suggested to solve the problems of the local burnup and the isotope distributions. The present method considers two extreme conditions of neutrons penetrating the fuel rod. Based on these considerations, the formula is obtained to calculate the reaction rates of 235U, 238U, and 239Pu and straightforward the local burnup and the isotope distributions. Starting from an initial burnup level, the parameters of the formula are fitted to the reaction rates given by a Monte Carlo (MC calculation. Then the present formula independently gives very similar results to the MC calculation from the starting to high burnup level but takes just a few minutes. The relative reaction rates are found to be almost independent of the radius (except (n,γ of 238U and the burnup, providing a solid background for the present formula. A more realistic examination is also performed when the fuel rods locate in an assembly. A combination of the present formula and the MC calculation is expected to have a nice balance between the numerical accuracy and time consumption.
International Nuclear Information System (INIS)
Khvostov, G.; Zimmermann, M.A.; Sugiyama, T.; Fuketa, T.
2008-01-01
The LS-1 RIA-simulating test conducted in NSRR with high-burnup fuel base irradiated in the BWR KKL to a pellet burn-up of 69 MWd/kgU is analysed using the FALCON fuel behaviour code coupled with the advanced GRSW-A model for fission gas release and swelling of the uranium dioxide fuels. A reasonable agreement of the results of calculation with the measured value of fission gas release in the failed fuel rod, amounting to about 23 % of total gas generated in the pellets is obtained. The mechanisms able to result in as significant fission gas release as measured by the post-test-examination are analysed with the model and discussed in consideration of the predicted initial micro-structural state of the fuel after the base irradiation. The minor role of gaseous swelling in the early failure of the LS-1 test-fuel-rod is shown by calculation, which is due to premature brittle cracking of the highly hydrated cladding under the conditions of the Room-Temperature capsule used in the LS-1 test. However, a significant potential impact of the gaseous swelling on cladding strain-stress conditions during the RIA is shown by the calculation when assuming a sufficient residual cladding ductility in the hypothetical test with the same parameters as in LS-1, but under conditions of the High- Temperature-High-Pressure capsule. (authors)
Energy Technology Data Exchange (ETDEWEB)
Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-08-01
PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)
International Nuclear Information System (INIS)
Sanders, T.L.
1989-01-01
The fresh fuel design basis provides some margin of safety, i.e., criticality safety is almost independent of loading operations if fuel designs do not change significantly over the next 40 years. However, the design basis enrichment for future nuclear fuel will most likely vary with time. As a result, it cannot be guaranteed that the perceived passivity of the concept will be maintained over the life cycle of a future cask system. Several options are available to ensure that the reliability of a burnup credit system is comparable to or greater than that of a system based on a fresh fuel assumption. Criticality safety and control reliability could increase with burnup credit implementation. The safety of a burnup credit system could be comparable to that for a system based on the fresh fuel assumption. A burnup credit philosophy could be implemented without any cost-benefit tradeoff. A burnup credit design basis could result in a significant reduction in total system risk as well as economic benefits. These reductions occur primarily as a result of increased cask capacities and, thus, fewer shipments. Fewer shipments also result in fewer operations over the useful life of a cask, and opportunities for error decrease. The system concept can be designed such that only benefits occur. These benefits could include enhanced criticality safety and the overall reliability of cask operations, as well as system risk and economic benefits. Thus, burnup credit should be available as an alternative for the criticality design of spent fuel shipping casks
International Nuclear Information System (INIS)
Cepraga, D.; Boeriu, St.; Gheorghiu, E.; Cristian, I.; Patrulescu, I.; Cimporescu, D.; Ciuvica, P.; Velciu, E.
1975-01-01
The calculation system DACC-5 is a zero-dimensional reactor physics code used to calculate the criticality and burn-up of light-water reactors. The code requires as input essential extensive reactor parameters (fuel rod radius, water density, etc.). The nuclear constants (intensive parameters) are calculated with a five-group model (2 thermal and 3 fast groups). A fitting procedure is systematically employed to reduce the computation time of the code. Zero-dimensional burn-up calculations are made in an automatic way. Part one of the paper contains the code physical model and computer structure. Part two of the paper will contain tests of DACC-5 credibility for different light-water power lattices
International Nuclear Information System (INIS)
Shin, Hee Sung; Suyama, Kenya; Mochizuki, Hiroki; Okuno, Hiroshi; Nomura, Yasushi
2001-01-01
The isotopic composition calculations were performed for 26 spent fuel samples from the Obrigheim PWR reactor and 55 spent fuel samples from 7 PWR reactors using the SAS2H module of the SCALE4.4 code system with 27, 44 and 238 group cross-section libraries and the SWAT code system with the 107 group cross-section library. For the analyses of samples from the Obrigheim PWR reactor, geometrical models were constructed for each of SCALE4.4/SAS2H and SWAT. For the analyses of samples from 7 PWR reactors, the geometrical model already adopted in the SCALE/SAS2H was directly converted to the model of SWAT. The four kinds of calculation results were compared with the measured data. For convenience, the ratio of the measured to calculated values was used as a parameter. When the ratio is less than unity, the calculation overestimates the measurement, and the ratio becomes closer to unity, they have a better agreement. For many important nuclides for burnup credit criticality safety evaluation, the four methods applied in this study showed good coincidence with measurements in general. More precise observations showed, however: (1) Less unity ratios were found for Pu-239 and -241 for selected 16 samples out of the 26 samples from the Obrigheim reactor (10 samples were deselected because their burnups were measured with Cs-137 non-destructive method, less reliable than Nd-148 method the rest 16 samples were measured with); (2) Larger than unity ratios were found for Am-241 and Cm-242 for both the 16 and 55 samples; (3) Larger than unity ratios were found for Sm-149 for the 55 samples; (4) SWAT was generally accompanied by larger ratios than those of SAS2H with some exceptions. Based on the measured-to-calculated ratios for 71 samples of a combined set in which 16 selected samples and 55 samples were included, the correction factors that should be multiplied to the calculated isotopic compositions were generated for a conservative estimate of the neutron multiplication factor
Estimating NIRR-1 burn-up and core life time expectancy using the codes WIMS and CITATION
Yahaya, B.; Ahmed, Y. A.; Balogun, G. I.; Agbo, S. A.
The Nigeria Research Reactor-1 (NIRR-1) is a low power miniature neutron source reactor (MNSR) located at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria Nigeria. The reactor went critical with initial core excess reactivity of 3.77 mk. The NIRR-1 cold excess reactivity measured at the time of commissioning was determined to be 4.97 mk, which is more than the licensed range of 3.5-4 mk. Hence some cadmium poison worth -1.2 mk was inserted into one of the inner irradiation sites which act as reactivity regulating device in order to reduce the core excess reactivity to 3.77 mk, which is within recommended licensed range of 3.5 mk and 4.0 mk. In this present study, the burn-up calculations of the NIRR-1 fuel and the estimation of the core life time expectancy after 10 years (the reactor core expected cycle) have been conducted using the codes WIMS and CITATION. The burn-up analyses carried out indicated that the excess reactivity of NIRR-1 follows a linear decreasing trend having 216 Effective Full Power Days (EFPD) operations. The reactivity worth of top beryllium shim data plates was calculated to be 19.072 mk. The result of depletion analysis for NIRR-1 core shows that (7.9947 ± 0.0008) g of U-235 was consumed for the period of 12 years of operating time. The production of the build-up of Pu-239 was found to be (0.0347 ± 0.0043) g. The core life time estimated in this research was found to be 30.33 years. This is in good agreement with the literature
International Nuclear Information System (INIS)
Yoriyaz, H.
1986-01-01
In this work a spatial burnup scheme and feedback effects has been implemented into the FERM ( 'Finite Element Response Matrix' )program. The spatially dependent neutronic parameters have been considered in three levels: zonewise calculation, assembly wise calculation and pointwise calculation. Flux and power distributions and the multiplication factor were calculated and compared with the results obtained by CITATIOn program. These comparisons showed that processing time in the Ferm code has been hundred of times shorter and no significant difference has been observed in the assembly average power distribution. (Author) [pt
An analysis of water reactor burnup data with the METHUSELAH II code
International Nuclear Information System (INIS)
Floyd, M.; Hicks, D.
1964-10-01
The METHUSELAH II code has been used to predict long term reactivity and isotopic changes in the YANKEE, Dresden and NRX reactors. In general it is shown that there is a satisfactory measure of agreement and the first core lives of YANKEE and Dresden appear well predicted. However there are discrepancies in the isotopic composition of the plutonium formed which appear to be correlated with the degree of hardness of the reactor spectrum. It is demonstrated that plausible changes in nuclear data could reduce the discrepancies. (author)
Incorporation of the variation in conductivity with burnup in the stability of code predictive LAPUR
International Nuclear Information System (INIS)
Escriba, A.; Munoz-cobo, J. L.; Merino, R.; Melara, J.; Albendea, M.
2013-01-01
In the field of nuclear safety, the analysis of the stability of boiling water reactors is one of the biggest challenges for researchers. LAPUR code that allows to obtain the parameters of stability of the plant (Decay rate and frequency), being one of the programs used by IBERDROLA can be used for these calculations. With the collaboration of the research group TIN of the Polytechnic University of Valencia, a model of loss of conductivity of uranium has joined with the burned LAPUR. This update allows you to play the phenomenon in a more realistic way. This improvement has been validated and verified contrasting results with reference values.
COGEMA/TRANSNUCLEAIRE's experience with burnup credit
International Nuclear Information System (INIS)
Chanzy, Y.; Guillou, E.
1998-01-01
Facing a continuous increase in the fuel enrichments, COGEMA and TRANSNUCLEAIRE have implemented step by step a burnup credit programme to improve the capacity of their equipment without major physical modification. Many authorizations have been granted by the French competent authority in wet storage, reprocessing and transport since 1981. As concerns transport, numerous authorizations have been validated by foreign competent authorities. Up to now, those authorizations are restricted to PWR Fuel type assemblies made of enriched uranium. The characterization of the irradiated fuel and the reactivity of the systems are evaluated by calculations performed with well qualified French codes developed by the CEA (French Atomic Energy Commission): CESAR as a depletion code and APPOLO-MORET as a criticality code. The authorizations are based on the assurance that the burnup considered is met on the least irradiated part of the fuel assemblies. Besides, the most reactive configuration is calculated and the burnup credit is restricted to major actinides only. This conservative approach allows not to take credit for any axial profile. On the operational side, the procedures have been reevaluated to avoid misloadings and a burnup verification is made before transport, storage and reprocessing. Depending on the level of burnup credit, it consists of a qualitative (go/no-go) verification or of a quantitative measurement. Thus the use of burnup credit is now a common practice in France and Germany and new improvements are still in progress: extended qualifications of the codes are made to enable the use of six selected fission products in the criticality evaluations. (author)
Burn-up measurements coupling gamma spectrometry and neutron measurement
International Nuclear Information System (INIS)
Toubon, H.; Pin, P.; Lebrun, A.; Oriol, L.; Saurel, N.; Gain, T.
2006-01-01
The need to apply for burn-up credit arises with the increase of the initial enrichment of nuclear fuel. When burn-up credit is used in criticality safety studies, it is often necessary to confirm it by measurement. For the last 10 years, CANBERRA has manufactured the PYTHON system for such measurements. However, the method used in the PYTHON itself uses certain reactor data to arrive at burn-up estimates. Based on R and D led by CEA and COGEMA in the framework of burn-up measurement for burn-up credit and safeguards applications, CANBERRA is developing the next generation of burn-up measurement device. This new product, named SMOPY, is able to measure burn-up of any kind of irradiated fuel assembly with a combination of gamma spectrometry and passive neutron measurements. The measurement data is used as input to the CESAR depletion code, which has been developed and qualified by CEA and COGEMA for burn-up credit determinations. In this paper, we explain the complementary nature of the gamma and neutron measurements. In addition, we draw on our previous experience from PYTHON system and from COGEMA La Hague to show what types of evaluations are required to qualify the SMOPY system, to estimate its uncertainties, and to detect discrepancies in the fuel data given by the reactor plant to characterize the irradiated fuel assembly. (authors)
International Nuclear Information System (INIS)
Gulliford, J.; Edge, J.A.; Gracey, J.; Harris, N.
2003-01-01
A new criticality safety assessment based on Actinide-Only Burnup Credit has been developed to cover operations in BNFL's Thermal Oxide Reprocessing Plant (THORP). Reduction of the gadolinium concentration leads to significant reduction in active waste volumes. Detailed description of the methodology was presented at ICNC 1999 and the basic components of the approved safety case have remained unchanged from those proposed then. This paper presents a brief summary of the new methodology, and describes further analyses carried out to quantify additional safety margins. These additional margins are not credited in the derivation of the operating limits, but provide further evidence of the fault tolerance inherent in the new regime. As part of the arrangements to monitor the overall performance of the plant and instrumentation under the new regime, various analyses of plant data are made, including 'on-line' cross checks of measured versus expected fuel parameters (i.e. in addition to the checks on Residual Enrichment). Statistical analyses of data are made and compared with similar data from earlier batches. A summary of analyses made on some of the early fuel batches is presented here. A summary of the likely further development in the Burnup Credit methodology is given in this paper. (author)
Development of methods for burn-up calculations for LWR's
International Nuclear Information System (INIS)
Jaschik, W.
1978-01-01
This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de
Detailed Burnup Calculations for Testing Nuclear Data
Leszczynski, F.
2005-05-01
A general method (MCQ) has been developed by introducing a microscopic burnup scheme that uses the Monte Carlo calculated fluxes and microscopic reaction rates of a complex system and a depletion code for burnup calculations as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy-dependent cross-section libraries and full 3D geometry of the system can be input for the calculations. The resulting predictions for the system at successive burnup time steps are thus based on a calculation route where both geometry and cross sections are accurately represented, without geometry simplifications and with continuous energy data, providing an independent approach for benchmarking other methods and nuclear data of actinides, fission products, and other burnable absorbers. The main advantage of this method over the classical deterministic methods currently used is that the MCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burnup codes adopted until now are the widely used MCNP and ORIGEN codes, but other codes can be used also. For using this method, there is need of a well-known set of nuclear data for isotopes involved in burnup chains, including burnable poisons, fission products, and actinides. For fixing the data to be included in this set, a study of the present status of nuclear data is performed, as part of the development of the MCQ method. This study begins with a review of the available cross-section data of isotopes involved in burnup chains for power and research nuclear reactors. The main data needs for burnup calculations are neutron cross sections, decay constants, branching ratios, fission energy, and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross
Transnucleaire's experience with burnup credit in transport operations
International Nuclear Information System (INIS)
Malesys, P.
2001-01-01
Facing a continued increase in fuel enrichment values, Transnucleaire has progressively implemented a burnup credit programme in order to maintain or, where possible, to improve the capacity of its transport packagings without physical modification. Many package design approvals, based on a notion of burnup credit, have been granted by the French competent authority for transport since the early eighties, and many of these approvals have been validated by foreign competent authorities. Up to now, these approvals are restricted to fuel assemblies made of enriched uranium and irradiated in pressurized water reactors (PWR). The characterization of the irradiated fuel and the reactivity of the package are evaluated by calculation, performed using qualified French codes developed by the CEA (Commisariat a l'Energie Atomique/French Atomic Energy Commission): CESAR as a depletion code and APOLO-MORET as a criticality code. The approvals are based on the hypothesis that the burnup considered is that applied on the least irradiated region of the fuel assemblies, the conservative approach being not to take credit for any axial profile of burnup along the fuel assembly. The most reactive configuration is calculated and the burnup credit is also restricted to major actinides only. On the operational side and in compliance with regulatory requirements, verification is made before transport, in order to meet safety objectives as required by the transport regulations. Besides a review of documentation related to the irradiation history of each fuel assembly, it consists of either a qualitative (go/no-go) verification or of a quantitative measurement, depending on the level of burnup credit. Thus the use of burnup credit is now a common practice with Transnucleaire's packages, particularly in France and Germany. New improvements are still in progress and qualifications of the calculation code are now well advanced, which will allow in the near future the use of six selected
Cell verification of parallel burnup calculation program MCBMPI based on MPI
International Nuclear Information System (INIS)
Yang Wankui; Liu Yaoguang; Ma Jimin; Wang Guanbo; Yang Xin; She Ding
2014-01-01
The parallel burnup calculation program MCBMPI was developed. The program was modularized. The parallel MCNP5 program MCNP5MPI was employed as neutron transport calculation module. And a composite of three solution methods was used to solve burnup equation, i.e. matrix exponential technique, TTA analytical solution, and Gauss Seidel iteration. MPI parallel zone decomposition strategy was concluded in the program. The program system only consists of MCNP5MPI and burnup subroutine. The latter achieves three main functions, i.e. zone decomposition, nuclide transferring and decaying, and data exchanging with MCNP5MPI. Also, the program was verified with the pressurized water reactor (PWR) cell burnup benchmark. The results show that it,s capable to apply the program to burnup calculation of multiple zones, and the computation efficiency could be significantly improved with the development of computer hardware. (authors)
東條, 匡志; tojo, masashi
2007-01-01
In this study, a BWR core calculation method is developed. The continuous energy Monte Carlo burn-up calculation code is newly applied to BWR assembly calculations of production level. The applicability of the present new calculation method is verified through the tracking-calculation of commercial BWR.The mechanism and quantitative effects of the error propagations, the spatial discretization and of the temperature distribution in fuel pellet on the Monte Carlo burn-up calculations are clari...
International Nuclear Information System (INIS)
Neuber, J.C.; Kuehl, H.
2001-01-01
This paper describes the experience gained in Germany in implementing burnup credit in wet storage and dry transport systems of spent PWR, BWR, and MOX fuel. It gives a survey of the levels of burnup credit presently used, the regulatory status and activities planned, the fuel depletion codes and criticality calculation codes employed, the verification methods used for validating these codes, the modeling assumptions made to ensure that the burnup credit criticality analysis is based on a fuel irradiation history which leads to bounding neutron multiplication factors, and the implementation of procedures used for fuel loading verification. (author)
Thermal-mechanical analyses of fuel rods in extended burnup cycles
International Nuclear Information System (INIS)
Rajan, S.R.; Sheppard, K.D.
1984-01-01
The purpose of this work was to investigate analytically the impact of low-leakage extended burnup cycling schemes on fuel performance in pressurized water reactors (PWRs). The thermal-mechanical analysis was done with the COMETHE code. Power histories from various fuel cycling schemes were imposed on a single fuel design, and the behaviour of rod internal pressure, fuel centerline temperature, and susceptibility to PCI-induced failure assessed as a function of burnup. These estimates were made both for base load operating histories as well as power histories that included load-follow operations. The high burnup schemes were found to have potential, though not severe, increases in internal pin pressure and cladding hoop stresses. Load follow operations are not expected to degrade fuel performance at conventional burnups, but tend to aggravate the situation of tensile cladding hoop stress at higher burnups. (author)
International Nuclear Information System (INIS)
1998-04-01
The criticality safety analysis of spent fuel systems has traditionally assumed that the fuel is fresh. This results in significant conservatism in the calculated value of the system's reactivity. Improved calculational methods allows one to take credit for the reactivity reduction associated with fuel burnup, hence reducing the analysis conservatism while maintaining an adequate criticality safety margin. Motivation for using burnup credit in criticality safety applications is generally based on economic considerations. Although economics may be a primary factor in deciding to use burnup credit, other benefits may be realized. Many of the additional benefits of burnup credit that are not strictly economic, may be considered to contribute to public health and safety, and resource conservation and environmental quality. Interest in the implementation of burnup credit has been shown by many countries. A summary of the information gathered by the IAEA about ongoing activities and regulatory status of burnup credit in different countries is included. Burnup credit implementation introduces new parameters and effects that should be addressed in the criticality analysis (e.g., axial and radial burnup shapes, fuel irradiation history, and others). Analysis of these parameters introduces new variations as well as the uncertainties, that should be considered in the safety assessment of the system. Also, the need arises to validate the isotopic composition that results from a depletion calculation, as well as to extend the current validation range of criticality codes to cover spent fuel. The use of burnup credit implies a verification of the fuel burnup before loading for transport, storage, disposal, or reprocessing each assembly, to make sure that the burnup level achieved complies with the criteria established. Methods and procedures used in different countries are described in this report
International Nuclear Information System (INIS)
Kim, Jung Do; Gil, Choong Sub; Lee, Jong Tai; Hwang, Won Guk
1992-01-01
A one-group cross section data base of the ORIGEN2 computer code was developed for research reactor applications. For this, ENDF/B-IV and -V data were processed using the NJOY code system into 69-group data. The burnup dependent weighting spectra for KMRR were calculated with the WIMS-KAERI computer code, and then the 69-group data were collapsed to one-group using the spectra. The ORlGEN2-predicted burnup-dependent actinide compositions of KMRR spent fuel using the newly developed data base show a good agreement with the results of detailed multigroup transport calculation. In addition, the burnup characteristics of KMRR spent fuel was analyzed with the new data base. (Author)
State of art in FE-based fuel performance codes
International Nuclear Information System (INIS)
Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun
2013-01-01
Fuel performance codes approximate this complex behavior using an axisymmetric, axially-stacked, one-dimensional radial representation to save computation cost. However, the need for improved modeling of PCMI and, particularly, the importance of multidimensional capability for accurate fuel performance simulation has been identified as safety margin decreases. Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed
Energy Technology Data Exchange (ETDEWEB)
Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenter [Institute for Nuclear Technology and Energy Systems, University of Stuttgart, Pfaffenwaldring 31, 70569 Stuttgart (Germany)
2008-07-01
The graphite moderated Modular High Temperature Pebble Bed Reactor enables very flexible loading strategies and is one candidate of the Generation IV reactors. For this reactor fuel cycles with high burnup (about 600 MWd/kg HM) based on plutonium (Pu) and minor actinides (MA) fuel will be investigated. The composition of this fuel is defined in the EU-PuMA-project which aims the reduction of high level waste. There exist nearly no neutronic full core calculations for this fuel composition with high burnup. Two methods (deterministic and Monte Carlo) will be used to determine the neutronics in a full core. The detailed results will be compared with respect to the influence on criticality and safety related parameters. (authors)
International Nuclear Information System (INIS)
Meier, Astrid; Bernnat, Wolfgang; Lohnert, Guenter
2008-01-01
The graphite moderated Modular High Temperature Pebble Bed Reactor enables very flexible loading strategies and is one candidate of the Generation IV reactors. For this reactor fuel cycles with high burnup (about 600 MWd/kg HM) based on plutonium (Pu) and minor actinides (MA) fuel will be investigated. The composition of this fuel is defined in the EU-PuMA-project which aims the reduction of high level waste. There exist nearly no neutronic full core calculations for this fuel composition with high burnup. Two methods (deterministic and Monte Carlo) will be used to determine the neutronics in a full core. The detailed results will be compared with respect to the influence on criticality and safety related parameters. (authors)
COREBN; A Core burn-up calculation module for SRAC2006
奥村 啓介
2007-01-01
COREBN is an auxiliary code of the SRAC system for multi-dimensional core burn-up calculation based on the diffusion theory and interpolation of macroscopic cross-sections tabulated to local parameters such as burn-up degree, moderator temperature and so on. The macroscopic cross-sections are prepared by cell burn-up calculations with the collision probability method of SRAC. SRAC and COREBN have wide applicability for various types of cell and core geometries. They have been used mainly for ...
Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor
International Nuclear Information System (INIS)
Nguyen Phuoc Lan; Do Quang Binh
2016-01-01
In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)
VVER-1000 burnup credit benchmark (CB5). New results evaluation
International Nuclear Information System (INIS)
Manolova, M.; Mihaylov, N.; Prodanova, R.
2008-01-01
The validation of depletion codes is an important task in spent fuel management, especially for burnup credit application in criticality safety analysis of spent fuel facilities. Because of lack of well documented experimental data for VVER-1000, the validation could be made on the basis of code intercomparison based on the numerical benchmark problems. Some years ago a VVER-1000 burnup credit benchmark (CB5) was proposed to the AER research community and the preliminary results from three depletion codes were compared. In the paper some new results for the isotopic concentrations of twelve actinides and fifteen fission products calculated by the depletion codes SCALE5.1, WIMS9, SCALE4.4 and NESSEL-NUKO are compared and evaluated. (authors)
Core burn-up calculation method of JRR-3
International Nuclear Information System (INIS)
Kato, Tomoaki; Yamashita, Kiyonobu
2007-01-01
SRAC code system is utilized for core burn-up calculation of JRR-3. SRAC code system includes calculation modules such as PIJ, PIJBURN, ANISN and CITATION for making effective cross section and calculation modules such as COREBN and HIST for core burn-up calculation. As for calculation method for JRR-3, PIJBURN (Cell burn-up calculation module) is used for making effective cross section of fuel region at each burn-up step. PIJ, ANISN and CITATION are used for making effective cross section of non-fuel region. COREBN and HIST is used for core burn-up calculation and fuel management. This paper presents details of NRR-3 core burn-up calculation. FNCA Participating countries are expected to carry out core burn-up calculation of domestic research reactor by SRAC code system by utilizing the information of this paper. (author)
Energy Technology Data Exchange (ETDEWEB)
Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)
Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio
Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik; Suzuki, Mitsutoshi
2014-09-01
Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of thereactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to runthe analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor typeas a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.
TRIGA fuel burn-up calculations and its confirmation
Energy Technology Data Exchange (ETDEWEB)
Khan, R., E-mail: rustamzia@yahoo.co [Vienna University of Technology (TU Wien)/Atominstitute (ATI), Stadionallee 2, A-1020, Vienna (Austria); Karimzadeh, S.; Boeck, H. [Vienna University of Technology (TU Wien)/Atominstitute (ATI), Stadionallee 2, A-1020, Vienna (Austria)
2010-05-15
The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.
Comparative study on plutonium and MA recycling in equilibrium burnup and standard burnup of PWR
International Nuclear Information System (INIS)
Waris, Abdul; Kurniadi, Rizal; Su'ud, Zaki; Permana, Sidik
2005-01-01
The equilibrium burnup model is a powerful method since its can handle all possible generated nuclides in any nuclear system. Moreover, this method is a simple time independent method. Hence the equilibrium burnup method could be very useful for evaluating and forecasting the characteristics of any nuclear fuel cycle, even the strange one, e.g. all nuclides are confined in the reactor. However, this method needs to be verified since the method is not a standard tool. The present study aimed to compare the characteristics of plutonium recycling and plutonium and minor actinides (MA) recycling in PWR with the equilibrium burnup and the standard burnup. In order to become more comprehensive study, an influence of moderator-to-fuel volume ratio (MFR) changes by changing the pin-pitch of fuel cell has been evaluated. The MFR ranges from 0.5 to 4.0. For the equilibrium burnup we used equilibrium cell-burnup code. We have employed 1368 nuclides in the equilibrium calculation with 129 of them are heavy metals (HMs). For standard burnup, SRAC2002 code has been utilized with 26 HMs and 66 fission products (FPs). The JENDL 3.2 library has been employed for both burnup schemes. The uranium, plutonium and MA vector, which resulted from the equilibrium burnup are directly used as fuel input composition for the standard burnup calculation. Both burnup results demonstrate that plutonium recycling and plutonium and MA recycling can be conducted safer in tight lattice core. They are also show the similar trend in neutron spectrum, which become harder with the increasing number of recycled heavy nuclides as well as the decreasing of the MFR values. However, there are some discrepancy on the effective multiplication factor and the conversion ratio, especially for the reactor core for MFR ≥ 2.0. (author)
International Nuclear Information System (INIS)
Geller, L.; Goldstein, L.; Franks, W.A.
1986-01-01
This paper reviews some of the considerations utilities must evaluate when going to higher discharge burnups. The advantages and disadvantages of higher discharge burnups are described, as well as a consistent approach for evaluating optimum discharge burnup and its comparison to current practice. When an analysis is performed over the life of the plant, the design of the terminal cycles has significant impact on the lifetime savings from higher burnups. Designs for high burnup cycles have a greater average inventory value in the core. As one goes to higher burnup, there is a greater likelihood of discarding a larger value in unused fuel unless the terminal cycles are designed carefully. This effect can be large enough in some cases to wipe out the lifetime cost savings relative to operating with a higher discharge burnup cycle
Development of Monte Carlo-based pebble bed reactor fuel management code
International Nuclear Information System (INIS)
Setiadipura, Topan; Obara, Toru
2014-01-01
Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed
State of art in FE-based fuel performance codes
International Nuclear Information System (INIS)
Kim, Hyo Chan; Yang, Yong Sik; Kim, Dae Ho; Bang, Je Geon; Kim, Sun Ki; Koo, Yang Hyun
2013-01-01
Finite element (FE) method that is reliable and proven solution in mechanical field has been introduced into fuel performance codes for multidimensional analysis. The present state of the art in numerical simulation of FE-based fuel performance predominantly involves 2-D axisymmetric model and 3-D volumetric model. The FRAPCON and FRAPTRAN own 1.5-D and 2-D FE model to simulate PCMI and cladding ballooning. In 2-D simulation, the FALCON code, developed by EPRI, is a 2-D (R-Z and R-θ) fully thermal-mechanically coupled steady-state and transient FE-based fuel behavior code. The French codes TOUTATIS and ALCYONE which are 3-D, and typically used to investigate localized behavior. In 2008, the Idaho National Laboratory (INL) has been developing multidimensional (2-D and 3-D) nuclear fuel performance code called BISON. In this paper, the current state of FE-based fuel performance code and their models are presented. Based on investigation into the codes, requirements and direction of development for new FE-based fuel performance code can be discussed. Based on comparison of models in FE-based fuel performance code, status of art in the codes can be discussed. A new FE-based fuel performance code should include typical pellet and cladding models which all codes own. In particular, specified pellet and cladding model such as gaseous swelling and high burnup structure (HBS) model should be developed to improve accuracy of code as well as consider AC condition. To reduce computation cost, the approximated gap and the optimized contact model should be also developed. Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena, occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. This multiphysics behavior is often tightly coupled, a well known example being the thermomechanical behavior. Adding to this complexity, important aspects of fuel behavior are inherently
Issues for effective implementation of burnup credit
International Nuclear Information System (INIS)
Parks, C.V.; Wagner, J.C.
2001-01-01
In the United States, burnup credit has been used in the criticality safety evaluation for storage pools at pressurized water reactors (PWRs) and considerable work has been performed to lay the foundation for use of burnup credit in dry storage and transport cask applications and permanent disposal applications. Many of the technical issues related to the basic physics phenomena and parameters of importance are similar in each of these applications. However, the nuclear fuel cycle in the United States has never been fully integrated and the implementation of burnup credit to each of these applications is dependent somewhat on the specific safety bases developed over the history of each operational area. This paper will briefly review the implementation status of burnup credit for each application area and explore some of the remaining issues associated with effective implementation of burnup credit. (author)
Lattice cell burnup calculation
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1977-01-01
Accurate burnup prediction is a key item for design and operation of a power reactor. It should supply information on isotopic changes at each point in the reactor core and the consequences of these changes on the reactivity, power distribution, kinetic characters, control rod patterns, fuel cycles and operating strategy. A basic stage in the burnup prediction is the lattice cell burnup calculation. This series of lectures attempts to give a review of the general principles and calculational methods developed and applied in this area of burnup physics
Non destructive assay of nuclear LEU spent fuels for burnup credit application
International Nuclear Information System (INIS)
Lebrun, A.; Bignan, G.
2001-01-01
Criticality safety analysis devoted to spent fuel storage and transportation has to be conservative in order to be sure no accident will ever happen. In the spent fuel storage field, the assumption of freshness has been used to achieve the conservative aspect of criticality safety procedures. Nevertheless, after being irradiated in a reactor core, the fuel elements have obviously lost part of their original reactivity. The concept of taking into account this reactivity loss in criticality safety analysis is known as Burnup credit. To be used, Burnup credit involves obtaining evidence of the reactivity loss with a Burnup measurement. Many non destructive assays (NDA) based on neutron as well as on gamma ray emissions are devoted to spent fuel characterization. Heavy nuclei that compose the fuels are modified during irradiation and cooling. Some of them emit neutrons spontaneously and the link to Burnup is a power link. As a result, burn-up determination with passive neutron measurement is extremely accurate. Some gamma emitters also have interesting properties in order to characterize spent fuels but the convenience of the gamma spectrometric methods is very dependent on characteristics of spent fuel. In addition, contrary to the neutron emission, the gamma signal is mostly representative of the peripheral rods of the fuels. Two devices based on neutron methods but combining different NDA methods which have been studied in the past are described in detail: 1. The PYTHON device is a combination of a passive neutron measurement, a collimated total gamma measurement, and an online depletion code. This device, which has been used in several Nuclear Power Plants in western Europe, gives the average Burnup within a 5% uncertainty and also the extremity Burnup, 2. The NAJA device is an automatic device that involves three nuclear methods and an online depletion code. It is designed to cover the whole fuel assembly panel (Active Neutron Interrogation, Passive Neutron
International Nuclear Information System (INIS)
Sanders, T.L.
1988-01-01
The proposed use of burnup credit in spent fuel cask design and operation represents a departure from current regulatory practice, and creates technical issues that ultimately must be resolved for the concept to be implemented. Issues related to specific technical considerations can generally be resolved conclusively. However, an underlying perception may still exist that the use of burnup credit compromises criticality safety. In practice, individual casks are designed to satisfy regulatory requirements in a generally conservative manner. The designer's application of the regulatory requirements involves some engineering judgement, as does the regulator's implementation of them. This does not have an adverse effect on safety, but does make it difficult to objectively compare new or alternative designs and/or operating approaches. 5 refs., 7 figs., 2 tabs
International Nuclear Information System (INIS)
Lanning, D.D.; Beyer, C.E.; Painter, C.L.
1997-12-01
This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs
Energy Technology Data Exchange (ETDEWEB)
Lanning, D.D.; Beyer, C.E.; Painter, C.L.
1997-12-01
This volume describes the fuel rod material and performance models that were updated for the FRAPCON-3 steady-state fuel rod performance code. The property and performance models were changed to account for behavior at extended burnup levels up to 65 Gwd/MTU. The property and performance models updated were the fission gas release, fuel thermal conductivity, fuel swelling, fuel relocation, radial power distribution, solid-solid contact gap conductance, cladding corrosion and hydriding, cladding mechanical properties, and cladding axial growth. Each updated property and model was compared to well characterized data up to high burnup levels. The installation of these properties and models in the FRAPCON-3 code along with input instructions are provided in Volume 2 of this report and Volume 3 provides a code assessment based on comparison to integral performance data. The updated FRAPCON-3 code is intended to replace the earlier codes FRAPCON-2 and GAPCON-THERMAL-2. 94 refs., 61 figs., 9 tabs.
Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis
International Nuclear Information System (INIS)
2011-01-01
ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost
Technical Data to Justify Full Burnup Credit in Criticality Safety Licensing Analysis
Energy Technology Data Exchange (ETDEWEB)
Enercon Services, Inc.
2011-03-14
ENERCON's understanding of the difficult issues related to obtaining and analyzing additional cross section test data to support Full Burnup Credit. A PIRT (Phenomena Identification and Ranking Table) analysis was performed by ENERCON to evaluate the costs and benefits of acquiring different types of nuclear data in support of Full Burnup Credit. A PIRT exercise is a formal expert elicitation process with the final output being the ranking tables. The PIRT analysis (Table 7-4: Results of PIRT Evaluation) showed that the acquisition of additional Actinide-Only experimental data, although beneficial, was associated with high cost and is not necessarily needed. The conclusion was that the existing Radiochemical Assay (RCA) data plus the French Haut Taux de Combustion (HTC)2 and handbook Laboratory Critical Experiment (LCE) data provide adequate benchmark validation for Actinide-Only Burnup Credit. The PIRT analysis indicated that the costs and schedule to obtain sufficient additional experimental data to support the addition of 16 fission products to Actinide-Only Burnup Credit to produce Full Burnup Credit are quite substantial. ENERCON estimates the cost to be $50M to $100M with a schedule of five or more years. The PIRT analysis highlights another option for fission product burnup credit, which is the application of computer-based uncertainty analyses (S/U - Sensitivity/Uncertainty methodologies), confirmed by the limited experimental data that is already available. S/U analyses essentially transform cross section uncertainty information contained in the cross section libraries into a reactivity bias and uncertainty. Recent work by ORNL and EPRI has shown that a methodology to support Full Burnup Credit is possible using a combination of traditional RCA and LCE validation plus S/U validation for fission product isotopics and cross sections. Further, the most recent cross section data (ENDF/B-VII) can be incorporated into the burnup credit codes at a reasonable cost
A new approach to make collapsed cross section for burnup calculation of subcritical system
International Nuclear Information System (INIS)
Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao
2008-01-01
A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)
International Nuclear Information System (INIS)
Khvostov, Grigori; Novikov, Vladimir; Medvedev, Anatoli; Bogatyr, Serguey
2005-01-01
An advanced model GRSWEL-A for fission gas behavior and micro-structural evolutions in Light Water Reactor (LWR) fuels was developed for and embedded in the START-3 fuel performance code. This paper represents the physical basis and verification of the model with emphasis on analysis of High Burn-up Structure (HBS), which is generally ascribed to a so-called rim-layer of high burn-up fuel pellets. Specifically, the issues of microscopic polygonization, loss of matrix fission gas, growth of fuel porosity and fission gas release are highlighted. The effects of HBS on total fission gas release, temperature distribution in the pellet, pellet swelling and permanent strain of the cladding are considered in the appropriate section of the paper by means of comparative and sensitivity analysis with the use of both modeling and available experimental data. In all the cases, an accounting for the present effects is found to be an important integral part of thorough analysis of LWR fuel behavior. Aside from the description of current capabilities of modeling, some priority directions of further improvement are outlined. (author)
Sensitivity theory for reactor burnup analysis based on depletion perturbation theory
International Nuclear Information System (INIS)
Yang, Wonsik.
1989-01-01
The large computational effort involved in the design and analysis of advanced reactor configurations motivated the development of Depletion Perturbation Theory (DPT) for general fuel cycle analysis. The work here focused on two important advances in the current methods. First, the adjoint equations were developed for using the efficient linear flux approximation to decouple the neutron/nuclide field equations. And second, DPT was extended to the constrained equilibrium cycle which is important for the consistent comparison and evaluation of alternative reactor designs. Practical strategies were formulated for solving the resulting adjoint equations and a computer code was developed for practical applications. In all cases analyzed, the sensitivity coefficients generated by DPT were in excellent agreement with the results of exact calculations. The work here indicates that for a given core response, the sensitivity coefficients to all input parameters can be computed by DPT with a computational effort similar to a single forward depletion calculation
MTR core loading pattern optimization using burnup dependent group constants
Directory of Open Access Journals (Sweden)
Iqbal Masood
2008-01-01
Full Text Available A diffusion theory based MTR fuel management methodology has been developed for finding superior core loading patterns at any stage for MTR systems, keeping track of burnup of individual fuel assemblies throughout their history. It is based on using burnup dependent group constants obtained by the WIMS-D/4 computer code for standard fuel elements and control fuel elements. This methodology has been implemented in a computer program named BFMTR, which carries out detailed five group diffusion theory calculations using the CITATION code as a subroutine. The core-wide spatial flux and power profiles thus obtained are used for calculating the peak-to-average power and flux-ratios along with the available excess reactivity of the system. The fuel manager can use the BFMTR code for loading pattern optimization for maximizing the excess reactivity, keeping the peak-to-average power as well as flux-ratio within constraints. The results obtained by the BFMTR code have been found to be in good agreement with the corresponding experimental values for the equilibrium core of the Pakistan Research Reactor-1.
A SAS2H/KENO-V methodology for 3D fuel burnup analysis
International Nuclear Information System (INIS)
Milosevic, M.; Greenspan, E.; Vujic, J.
2002-01-01
An efficient methodology for 3D fuel burnup analysis of LWR reactors is described in this paper. This methodology is founded on coupling Monte Carlo method for 3D calculation of node power distribution, and transport method for depletion calculation in ID Wigner-Seitz equivalent cell for each node independently. The proposed fuel burnup modeling, based on application of SCALE-4.4a control modules SAS2H and KENO-V.a is verified for the case of 2D x-y model of IRIS 15 x 15 fuel assembly (with reflective boundary condition) by using two well benchmarked code systems. The one is MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility code, and the second is KENO-V.a/ORIGEN2.1 code system recently developed by authors of this paper. The proposed SAS2H/KENO-V.a methodology was applied for 3D burnup analysis of IRIS-1000 benchmark.44 core. Detailed k sub e sub f sub f and power density evolution with burnup are reported. (author)
Actinide-only burnup credit methodology for PWR spent nuclear fuel
International Nuclear Information System (INIS)
Lancaster, D.B.; Fuentes, E.; Kang, C.; Rahimi, M.
1998-01-01
A conservative methodology is presented that would allow taking credit for burnup in the criticality safety analysis of spent nuclear fuel (SNF) packages. The method is based on the assumption that the isotopic concentration in the SNF and cross sections of each isotope for which credit is taken must be supported by validation experiments. The method allows credit for the changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps: 1. Validate a computer code system to calculate isotopic concentrations of spent nuclear fuel created during burnup in the reactor core and subsequent decay. 2. Validate a computer code system to predict the subcritical multiplication factor, k eff , of a spent nuclear fuel package by use of UO 2 and UO 2 /Puo 2 critical experiments. 3. Establish conditions for the SNF (depletion analysis) and package (criticality analysis) which bounds k eff . 4. Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). 5. Verify by measurement that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. (author)
High Burnup Fuel Performance and Safety Research
Energy Technology Data Exchange (ETDEWEB)
Bang, Je Keun; Lee, Chan Bok; Kim, Dae Ho (and others)
2007-03-15
The worldwide trend of nuclear fuel development is to develop a high burnup and high performance nuclear fuel with high economies and safety. Because the fuel performance evaluation code, INFRA, has a patent, and the superiority for prediction of fuel performance was proven through the IAEA CRP FUMEX-II program, the INFRA code can be utilized with commercial purpose in the industry. The INFRA code was provided and utilized usefully in the universities and relevant institutes domesticallly and it has been used as a reference code in the industry for the development of the intrinsic fuel rod design code.
Modelling fission gas release at high burnup
International Nuclear Information System (INIS)
Loesoenen, Pekka
1999-01-01
High burn-up phenomena in LWR/HWR UO 2 fuel pellets important for fission gas release were modelled. The degradation of the thermal conductivity of UO 2 was accounted for 1) with a burnup -dependent factor in the phonon term of the corresponding equation and 2) with a correlation describing the increase in the porosity at the pellet rim as a function of local burnup and radial position. The model was tested against IFA-432 and IFA-429 data. It was found out that the degradation of the thermal conductivity in the phonon term is perhaps not a function of the local burnup only, but the irradiation temperature may play an important role, too. The burnup as a function of the pellet radius has to be known to determine the local thermal conductivity. A model for this was picked up from the literature, but a new estimation of a few empirical fitting parameters was performed with hundreds of data points from the OECD/NEA data base and from the literature. The model predicts reliably the radial burnup profile and the fission gas generation across the pellet in typical LWR and HWR fuels. The thermal release and the athermal release from the pellet rim were modelled separately. The model for the rim release is a function of the temperature history and the local burnup. The rim release and the thermal release can occur at the same radial position of the pellets simultaneously, which is accounted for in the calculation of the total release. The model for the rim release in in agreement with the latest experimental findings, but the tuning of the model parameters is yet to be done. However, the fraction of the rim structured fuel and the excessive porosity in the rim structure in isothermal irradiation as a function of the burnup was predicted by using typical model parameters (author) (ml)
A PWR Thorium Pin Cell Burnup Benchmark
Energy Technology Data Exchange (ETDEWEB)
Weaver, Kevan Dean; Zhao, X.; Pilat, E. E; Hejzlar, P.
2000-05-01
As part of work to evaluate the potential benefits of using thorium in LWR fuel, a thorium fueled benchmark comparison was made in this study between state-of-the-art codes, MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations. The MOCUP runs were done individually at MIT and INEEL, using the same model but with some differences in techniques and cross section libraries. Eigenvalue and isotope concentrations were compared on a PWR pin cell model up to high burnup. The eigenvalue comparison as a function of burnup is good: the maximum difference is within 2% and the average absolute difference less than 1%. The isotope concentration comparisons are better than a set of MOX fuel benchmarks and comparable to a set of uranium fuel benchmarks reported in the literature. The actinide and fission product data sources used in the MOCUP burnup calculations for a typical thorium fuel are documented. Reasons for code vs code differences are analyzed and discussed.
Time step length versus efficiency of Monte Carlo burnup calculations
International Nuclear Information System (INIS)
Dufek, Jan; Valtavirta, Ville
2014-01-01
Highlights: • Time step length largely affects efficiency of MC burnup calculations. • Efficiency of MC burnup calculations improves with decreasing time step length. • Results were obtained from SIE-based Monte Carlo burnup calculations. - Abstract: We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy
International Nuclear Information System (INIS)
Nakamura, Takehiko; Yoshinaga, Makio
2000-11-01
Pulse irradiation tests of irradiated fuel are performed in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under Reactivity Initiated Accident Conditions (RIA). The severity of the RIA is represented by energy deposition or peak fuel enthalpy during the power excursion. In case of the irradiated fuel tests, the energy deposition varies depending both on the amounts and distribution of residual fissile and neutron absorbing fission products generated during the base irradiation. Thus, proper fuel burnup characterization, especially for low enriched commercial fuels, is important, because plutonium (Pu) takes a large part of fissile and its generation depends on the neutron spectrum during the base irradiation. Fuel burnup calculations were conducted with ORIGEN2, RODBURN and SWAT codes for the BWR fuels tested in the NSRR. The calculation results were compared with the measured isotope concentrations and used for the NSRR neutron calculations to evaluate energy depositions of the test fuel. The comparison of the code calculations and the measurements revealed that the neutron spectrum change due to difference in void fraction altered Pu generation and energy deposition in the NSRR tests considerably. With the properly evaluated neutron spectrum, the combined burnup and NSRR neutron calculation gave reasonably good evaluation of the energy deposition. The calculations provided radial distributions of the fission product accumulation during the base irradiation and power distribution during the NSRR pulse irradiation, which were important for the evaluation of both burnup characteristics and fission gas release behavior. (author)
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.; DeHart, M.D.
2000-03-01
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ``end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified.
Review of Axial Burnup Distribution Considerations for Burnup Credit Calculations
International Nuclear Information System (INIS)
Wagner, J.C.; DeHart, M.D.
2000-01-01
This report attempts to summarize and consolidate the existing knowledge on axial burnup distribution issues that are important to burnup credit criticality safety calculations. Recently released Nuclear Regulatory Commission (NRC) staff guidance permits limited burnup credit, and thus, has prompted resolution of the axial burnup distribution issue. The reactivity difference between the neutron multiplication factor (keff) calculated with explicit representation of the axial burnup distribution and keff calculated assuming a uniform axial burnup is referred to as the ''end effect.'' This end effect is shown to be dependent on many factors, including the axial-burnup profile, total accumulated burnup, cooling time, initial enrichment, assembly design, and the isotopics considered (i.e., actinide-only or actinides plus fission products). Axial modeling studies, efforts related to the development of axial-profile databases, and the determination of bounding axial profiles are also discussed. Finally, areas that could benefit from further efforts are identified
Conservative axial burnup distributions for actinide-only burnup credit
International Nuclear Information System (INIS)
Kang, C.; Lancaster, D.
1997-11-01
Unlike the fresh fuel approach, which assumes the initial isotopic compositions for criticality analyses, any burnup credit methodology must address the proper treatment of axial burnup distributions. A straightforward way of treating a given axial burnup distribution is to segment the fuel assembly into multiple meshes and to model each burnup mesh with the corresponding isotopic compositions. Although this approach represents a significant increase in modeling efforts compared to the uniform average burnup approach, it can adequately determine the reactivity effect of the axial burnup distribution. A major consideration is what axial burnup distributions are appropriate for use in light of many possible distributions depending on core operating conditions and histories. This paper summarizes criticality analyses performed to determine conservative axial burnup distributions. The conservative axial burnup distributions presented in this paper are included in the Topical Report on Actinide-Only Burnup Credit for Pressurized Water Reactor Spent Nuclear Fuel Packages, Revision 1 submitted in May 1997 by the US Department of Energy (DOE) to the US Nuclear Regulatory Commission (NRC). When approved by NRC, the conservative axial burnup distributions may be used to model PWR spent nuclear fuel for the purpose of gaining actinide only burnup credit
Implementation of LT codes based on chaos
International Nuclear Information System (INIS)
Zhou Qian; Li Liang; Chen Zengqiang; Zhao Jiaxiang
2008-01-01
Fountain codes provide an efficient way to transfer information over erasure channels like the Internet. LT codes are the first codes fully realizing the digital fountain concept. They are asymptotically optimal rateless erasure codes with highly efficient encoding and decoding algorithms. In theory, for each encoding symbol of LT codes, its degree is randomly chosen according to a predetermined degree distribution, and its neighbours used to generate that encoding symbol are chosen uniformly at random. Practical implementation of LT codes usually realizes the randomness through pseudo-randomness number generator like linear congruential method. This paper applies the pseudo-randomness of chaotic sequence in the implementation of LT codes. Two Kent chaotic maps are used to determine the degree and neighbour(s) of each encoding symbol. It is shown that the implemented LT codes based on chaos perform better than the LT codes implemented by the traditional pseudo-randomness number generator. (general)
International Nuclear Information System (INIS)
Chipsham, E.; Jarvis, O.N.; Sadler, G.
1989-01-01
Triton burnup measurements have been made at JET using time-integrated copper activation and time-resolved silicon detector techniques. The results confirm the classical nature of both the confinement and the slowing down of the 1 MeV tritons in a plasma. (author) 8 refs., 3 figs
Analysis of burnup credit on spent fuel storage
International Nuclear Information System (INIS)
Matsumura, T.; Sasahara, A.
1999-01-01
Chemical analyses were carried out on high burnup UO 2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins. Measured data of the composition of nuclides from 234 U to 242 Pu were used for evaluation of ORIGEN-2/82 code. Criticality calculations were executed for the casks which were being designed to store 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for (1) axial and horizontal profiles of burnup, and void history (BWR), (2) operational histories such as control rod insertion history, BPR insertion history and others, and (3) calculational accuracy of ORIGEN-2/82 code on the composition of nuclides. Present evaluation shows that introduction of burnup credit has a substantial merit in criticality safety analysis of the cask, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for present reactivity bias evaluation and showed a possibility of simplifying the reactivity bias evaluation in burnup credit. Finally, adapting procedures of burnup credit such as the burnup meter were evaluated. (author)
Energy Technology Data Exchange (ETDEWEB)
Hernandez, J.L.; Alonso, G.; Perusquia, R.; Montes, J.L.; Hernandez, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jlhm@nuclear.inin-mx
2003-07-01
An evaluation of the capacity of the COREMASTER-Presto code, to evaluate generically the burnt of the control bars in the Laguna Verde reactors plant (CLV) is made. It was found that the code only reports burnt values of the control rods in MWD/TM, in spite of having with a second order polynomial model, for the conversion to remainder of the Boron-10 (B-10). It was observed that said model is adequate only for burnt smaller to 45,000 MWD/TM. To evaluate the burnt of the control rods it was reproduced the balance cycle of 18 months for the CLV, executing Cm-Presto during 13 consecutive cycles. First without rod burnt, taking this as the base case. Later on, cases with 1, 2 and up to 13 cycles with rod burnt were generated. When comparing results it was observed that the control rods pattern it loses reactivity lineally with the burnt one. By each 10 G Wd/T of burnt of the nucleus it is decreased the reactivity of the pattern rods {approx} 1 pcm in hot condition and of {approx} 20 pcm in cold condition. When burning three cycles those rods more burnt reached the 13,900 MWD/TM, equivalent to 36% of B-10 reduction, near value to 34% proposed by aging in the one lost study of B-10. It was observed that Cm-Presto it doesn't burn the superior node of the control rods when these are completely extracted. A one big lost of B-10, of the order of 50%, it represents only a decrease of 11% of the reactivity value of the rod. One can affirm that even when it is strongly decreased the content of B-10, the rod is continue considering as a black absorber, that is to say, thermal neutron that enters in the neutron rod that is absorbed. (Author)
Energy Technology Data Exchange (ETDEWEB)
DeHart, M.D.
1996-05-01
Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports.
International Nuclear Information System (INIS)
DeHart, M.D.
1996-05-01
Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For example, the spent fuel composition must be adequately characterized and the criticality analysis model can be complicated by the need to consider axial burnup variations. Parametric analyses are needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models must be evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. The purpose of this report is to describe analyses and evaluations performed in order to demonstrate the effect physical parameters and modeling assumptions have on the criticality analysis of spent fuel. The analyses in this report include determination and ranking of the most important actinides and fission products; study of the effect of various depletion scenarios on subsequent criticality calculations; establishment of trends in neutron multiplication as a function of fuel enrichment, burnup, cooling time- and a parametric and modeling evaluation of three-dimensional effects (e.g., axially varying burnup and temperature/density effects) in a conceptual cask design. The sensitivity and parametric evaluations were performed with the consideration of two different burnup credit approaches: (1) only actinides in the fuel are considered in the criticality analysis, and (2) both actinides and fission products are considered. Calculations described in this report were performed using the criticality and depletion sequences available in the SCALE code system and the SCALE 27-group burnup library. Although the results described herein do not constitute a validation of SCALE for use in spent fuel analysis, independent validation efforts have been completed and are described in other reports
Analysis of BWR high burnup fuel in LOCA conditions
International Nuclear Information System (INIS)
Garcia Sedano, Pablo; Dey Navarro, Jose Manuel; Gallego Cabezon, Ines; Orive Moreno, Raul
2004-01-01
High Burnup Fuel Behaviour has been growing in importance since middle 80's when pellet microstructure changes (rim effect) and cladding oxidation rates increase were observed. Later on, Cadarache reactivity tests revealed cladding integrity failures below safety limits. These phenomena, occurred at high burnup, stressed the necessity of having a wide experimental data base that would allow to dispose non-extrapolated data of material properties submitted to higher burnups than 40000 MWd/TM and data of new materials at the same time. One of the objectives of the EPRI's Fuel Reliability Program is to establish the bases for the licensing of nuclear fuel to burnup levels beyond the current licensed value of 62 GWd/MTU rod average burnup. The technical bases to support those high burnup levels are being developed. One of the licensing points of concern is the behaviour of the high burnup fuel in LOCA conditions. To respond to this concern a series of LOCA experiments are being performed at Argonne National Laboratory using fuel rods from Limerick NPP at 57 GWd/TM and H.B. Robinson at 67 GWd/MTU. When the ANL tests have been finished, a conservative Peak Cladding Temperature/ Equivalent Cladding Reacted (PCT/ECR) limit will be determine from the residual ductility tests to be applied to the high burnup fuel. This makes necessary to determine the behaviour of the high burnup fuel in LOCA conditions and to determine the available safety margin. In licensing LOCA calculations, corresponding to present core designs and future core designs, the calculated PCT and ECR values as a function of the fuel burnup could be used to determine the relative severity of LOCA for the high burnup fuel. This report presents the LOCA analyses performed by IBERDROLA (Spanish utility), using results from the Cofrentes NPP (BWR-6) LOCA evaluations. (authors)
Analysis on burn-up behaviors for accelerator-driven sub-critical facility
International Nuclear Information System (INIS)
Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao
2000-01-01
An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91
Alloy development for high burnup cladding (PWR)
Energy Technology Data Exchange (ETDEWEB)
Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1999-04-01
An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.
Alloy development for high burnup cladding (PWR)
International Nuclear Information System (INIS)
Hahn, R.; Jeong, Y. H.; Baek, K. H.; Kim, S. J.; Choi, B. K.; Kim, J.M.
1999-04-01
An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs
Reliability-Based Code Calibration
DEFF Research Database (Denmark)
Faber, M.H.; Sørensen, John Dalsgaard
2003-01-01
. Thereafter the code calibration problem is presented in its principal decision theoretical form and it is discussed how acceptable levels of failure probability (or target reliabilities) may be established. Furthermore suggested values for acceptable annual failure probabilities are given for ultimate...
Nuclide analysis in high burnup fuel samples irradiated in Vandellos 2
Energy Technology Data Exchange (ETDEWEB)
Zwicky, H.U., E-mail: hans-urs.zwicky@bluewin.c [Zwicky Consulting GmbH, Chilacherstr. 17, 5236 Remigen (Switzerland); Low, J.; Granfors, M. [Studsvik Nuclear AB, 611 82 Nykoeping (Sweden); Alejano, C.; Conde, J.M. [Consejo de Seguridad Nuclear, Justo Dorado 11, 28040 Madrid (Spain); Casado, C.; Sabater, J.; Lloret, M.; Quecedo, M. [ENUSA, Santiago Rosinol 12, 28040 Madrid (Spain); Gago, J.A. [ENRESA, Emilio Vargas 7, 28043 Madrid (Spain)
2010-07-01
In the framework of a high burnup fuel demonstration programme, rods with an enrichment of 4.5% {sup 235}U were operated to a rod average burnup of about 70 MWd/kgU in the Spanish Vandellos 2 pressurised water reactor. The rods were sent to hot cells and used for different research projects. This paper describes the isotopic composition measurements performed on samples of those rods, and the analysis of the measurement results based on comparison against calculated values. The fraction and composition of fission gases released to the rod free volume was determined for two of the rods. About 8% of Kr and Xe produced by fission were released. From the isotopic composition of the gases, it could be concluded that the gases were not preferentially released from the peripheral part of the fuel column. Local burnup and isotopic content of gamma emitting nuclides were determined by quantitatively evaluating axial gamma scans of the full rods. Nine samples were cut at different axial levels from three of the rods and analysed in two campaigns. More than 50 isotopes of 16 different elements were assessed, most of them by Inductively Coupled Plasma Mass Spectrometry after separation with High Performance Liquid Chromatography. In general, these over 400 data points gave a consistent picture of the isotopic content of irradiated fuel as a function of burnup. Only in a few cases, the analysis provided unexpected results that seem to be wrong, in most cases due to unidentified reasons. Sample burnup analysis was performed by comparing experimental isotopic abundances of uranium and plutonium composition as well as neodymium isotopic concentrations with corresponding CASMO based data. The results were in agreement with values derived independently from gamma scanning and from core design data and plant operating records. Measured isotope abundances were finally assessed using the industry standard SAS2H sequence of the SCALE code system. This exercise showed good agreement
Models for fuel rod behaviour at high burnup
Energy Technology Data Exchange (ETDEWEB)
Jernkvist, Lars O.; Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park, Uppsala (Sweden)
2004-12-01
This report deals with release of fission product gases and irradiation-induced restructuring in uranium dioxide nuclear fuel. Waterside corrosion of zirconium alloy clad tubes to light water reactor fuel rods is also discussed. Computational models, suitable for implementation in the FRAPCON-3.2 computer code, are proposed for these potentially life-limiting phenomena. Hence, an integrated model for the calculation or thermal fission gas release by intragranular diffusion, gas trapping in grain boundaries, irradiation-induced re-solution, grain boundary saturation, and grain boundary sweeping in UO{sub 2} fuel, under time varying temperature loads, is formulated. After a brief review of the status of thermal fission gas release modelling, we delineate the governing equations for the aforementioned processes. Grain growth kinetic modelling is briefly reviewed and pertinent data on grain growth of high burnup fuel obtained during power ramps in the Third Risoe Fission Gas Release Project are evaluated. Sample computations are performed, which clearly show the connection between fission gas release and gram growth as a function of time at different isotherms. Models are also proposed for the restructuring of uranium dioxide fuel at high burnup, the so-called rim formation, and its effect on fuel porosity build-up, fuel thermal conductivity and fission gas release. These models are assessed by use of recent experimental data from the High Burnup Rim Project, as well as from post irradiation examinations of high-burnup fuel, irradiated in power reactors. Moreover, models for clad oxide growth and hydrogen pickup in PWRs, applicable to Zircaloy-4, ZIRLO or M5 cladding, are formulated, based on recent in-reactor corrosion data for high-burnup fuel rods. Our evaluation of these data indicates that the oxidation rate of ZIRLO-type materials is about 20% lower than for standard Zircaloy-4 cladding under typical PWR conditions. Likewise, the oxidation rate of M5 seems to be
A simple gamma spectrometry method for evaluating the burnup of MTR-type HEU fuel elements
Makmal, T.; Aviv, O.; Gilad, E.
2016-10-01
A simple method for the evaluation of the burnup of a materials testing reactor (MTR) fuel element by gamma spectrometry is presented. The method was applied to a highly enriched uranium MTR nuclear fuel element that was irradiated in a 5 MW pool-type research reactor for a total period of 34 years. The experimental approach is based on in-situ measurements of the MTR fuel element in the reactor pool by a portable high-purity germanium detector located in a gamma cell. To corroborate the method, analytical calculations (based on the irradiation history of the fuel element) and computer simulations using a dedicated fuel cycle burnup code ORIGEN2 were performed. The burnup of the MTR fuel element was found to be 52.4±8.8%, which is in good agreement with the analytical calculations and the computer simulations. The method presented here is suitable for research reactors with either a regular or an irregular irradiation regime and for reactors with limited infrastructure and/or resources. In addition, its simplicity and the enhanced safety it confers may render this method suitable for IAEA inspectors in fuel element burnup assessments during on-site inspections.
Analysis of burnup and isotopic compositions of BWR 9 x 9 UO2 fuel assemblies
International Nuclear Information System (INIS)
Suzuki, M.; Yamamoto, T.; Ando, Y.; Nakajima, T.
2012-01-01
In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO 2 fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for 238 Pu, 144 Nd, 145 Nd, 146 Nd, 148 Nd, 134 Cs, 154 Eu, 152 Sm, 154 Gd, and 157 Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)
A semi-empirical model for the formation and depletion of the high burnup structure in UO2
Pizzocri, D.; Cappia, F.; Luzzi, L.; Pastore, G.; Rondinella, V. V.; Van Uffelen, P.
2017-04-01
In the rim zone of UO2 nuclear fuel pellets, the combination of high burnup and low temperature drives a microstructural change, leading to the formation of the high burnup structure (HBS). In this work, we propose a semi-empirical model to describe the formation of the HBS, which embraces the polygonisation/recrystallization process and the depletion of intra-granular fission gas, describing them as inherently related. For this purpose, we performed grain-size measurements on samples at radial positions in which the restructuring was incomplete. Based on these new experimental data, we infer an exponential reduction of the average grain size with local effective burnup, paired with a simultaneous depletion of intra-granular fission gas driven by diffusion. The comparison with currently used models indicates the applicability of the herein developed model within integral fuel performance codes.
International Nuclear Information System (INIS)
Nguyen Minh Tuan; Pham Quang Huy; Tran Tri Vien; Trang Cao Su; Tran Quoc Duong; Dang Tran Thai Nguyen
2013-01-01
The fuel burn-up is an important parameter needed to be monitored and determined during a reactor operation and fuel management. The fuel burn-up can be calculated using computer codes and experimentally measured. This work presents the theory and experimental method applied to determine the burn-up of the irradiated and 36% enriched VVR-M2 fuel type assemblies of Dalat reactor. The method is based on measurement of Cs-137 absolute specific activity using gamma spectrometer. Designed measuring system consists of a collimator tube, high purity Germanium detector (HPGe) and associated electronics modules and online computer data acquisition system. The obtained results of measurement are comparable with theoretically calculated results. (author)
Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor
Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.
2015-09-01
Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
Energy Technology Data Exchange (ETDEWEB)
Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)
2008-04-15
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
International Nuclear Information System (INIS)
DeHart, M.D.; Parks, C.V.; Brady, M.C.
1996-06-01
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155
OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results
Energy Technology Data Exchange (ETDEWEB)
DeHart, M.D.
1993-01-01
Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.
OECD/NEA burnup credit calculational criticality benchmark Phase I-B results
Energy Technology Data Exchange (ETDEWEB)
DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)
1996-06-01
In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
International Nuclear Information System (INIS)
Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.
2008-01-01
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files
International Nuclear Information System (INIS)
Suyama, K.; Uchida, Y.; Kashima, T.; Ito, T.; Miyaji, T.
2016-01-01
Criticality control of damaged nuclear fuel is one of the key issues in the decommissioning operation of the Fukushima Daiichi Nuclear Power Station accident. The average isotopic composition of spent nuclear fuel as a function of burn-up is required in order to evaluate criticality parameters of the mixture of damaged nuclear fuel with other materials. The NEA Expert Group on Burn-up Credit Criticality (EGBUC) has organised several international benchmarks to assess the accuracy of burn-up calculation methodologies. For BWR fuel, the Phase III-B benchmark, published in 2002, was a remarkable landmark that provided general information on the burn-up properties of BWR spent fuel based on the 8x8 type fuel assembly. Since the publication of the Phase III-B benchmark, all major nuclear data libraries have been revised; in Japan from JENDL-3.2 to JENDL-4, in Europe from JEF-2.2 to JEFF-3.1 and in the US from ENDF/B-VI to ENDF/B-VII.1. Burn-up calculation methodologies have been improved by adopting continuous-energy Monte Carlo codes and modern neutronics calculation methods. Considering the importance of the criticality control of damaged fuel in the Fukushima Daiichi Nuclear Power Station accident, a new international burn-up calculation benchmark for the 9 x 9 STEP-3 BWR fuel assemblies was organised to carry out the inter-comparison of the averaged isotopic composition in the interest of the burnup credit criticality safety community. Benchmark specifications were proposed and approved at the EGBUC meeting in September 2012 and distributed in October 2012. The deadline for submitting results was set at the end of February 2013. The basic model for the benchmark problem is an infinite two-dimensional array of BWR fuel assemblies consisting of a 9 x 9 fuel rod array with a water channel in the centre. The initial uranium enrichment of fuel rods without gadolinium is 4.9, 4.4, 3.9, 3.4 and 2.1 wt% and 3.4 wt% for the rods using gadolinium. The burn-up conditions are
International Nuclear Information System (INIS)
Feltus, M.A.
1995-01-01
This paper addresses the advantages and disadvantages of using very high fuel burnup, reinsertion, and low leakage designs in advanced fuel cycle light water reactor cores as a technique to reduce vessel fluence, and total volume of spent fuel discharged into the waste management stream. The results demonstrate how to attain practical high burnup core designs using the Penn State Fuel Management Package (PSFMP, i.e., LEOPARD, MCRAC, ADMARC, OPHAL computer codes) (Levine, 1992). The PSFMP was used to scope out fuel management strategies, that can be verified with a direct comparison between the PSFMP and CASMO-3/SIMULATE-3 (Smith, 1989) results. This paper focuses on the practical use of such advanced fuel designs to: (a) achieve very high discharge burnups, (b) produce low leakage at the periphery, (c) have 24 month cycles, and (d) maintain safety margins and peak power levels, based on using the Haling power distribution as a target. Evaluations of practical and optimal extended burnup core designs, using the PSFMP, will show that very high burnup core designs are not only attainable, but are most cost-effective and beneficial to the environment in terms of waste reduction. (author)
Edge-based perceptual image coding.
Niu, Yi; Wu, Xiaolin; Shi, Guangming; Wang, Xiaotian
2012-04-01
We develop a novel psychovisually motivated edge-based low-bit-rate image codec. It offers a compact description of scale-invariant second-order statistics of natural images, the preservation of which is crucial to the perceptual quality of coded images. Although being edge based, the codec does not explicitly code the edge geometry. To save bits on edge descriptions, a background layer of the image is first coded and transmitted, from which the decoder estimates the trajectories of significant edges. The edge regions are then refined by a residual coding technique based on edge dilation and sequential scanning in the edge direction. Experimental results show that the new image coding technique outperforms the existing ones in both objective and perceptual quality, particularly at low bit rates.
Monte Carlo burnup simulation of the TAKAHAMA-3 benchmark experiment
International Nuclear Information System (INIS)
Dalle, Hugo M.
2009-01-01
High burnup PWR fuel is currently being studied at CDTN/CNEN-MG. Monte Carlo burnup code system MONTEBURNS is used to characterize the neutronic behavior of the fuel. In order to validate the code system and calculation methodology to be used in this study the Japanese Takahama-3 Benchmark was chosen, as it is the single burnup benchmark experimental data set freely available that partially reproduces the conditions of the fuel under evaluation. The burnup of the three PWR fuel rods of the Takahama-3 burnup benchmark was calculated by MONTEBURNS using the simplest infinite fuel pin cell model and also a more complex representation of an infinite heterogeneous fuel pin cells lattice. Calculations results for the mass of most isotopes of Uranium, Neptunium, Plutonium, Americium, Curium and some fission products, commonly used as burnup monitors, were compared with the Post Irradiation Examinations (PIE) values for all the three fuel rods. Results have shown some sensitivity to the MCNP neutron cross-section data libraries, particularly affected by the temperature in which the evaluated nuclear data files were processed. (author)
Triton burnup measurements in KSTAR using a neutron activation system
Jo, Jungmin; Cheon, MunSeong; Kim, Jun Young; Rhee, T.; Kim, Junghee; Shi, Yue-Jiang; Isobe, M.; Ogawa, K.; Chung, Kyoung-Jae; Hwang, Y. S.
2016-11-01
Measurements of the time-integrated triton burnup for deuterium plasma in Korea Superconducting Tokamak Advanced Research (KSTAR) have been performed following the simultaneous detection of the d-d and d-t neutrons. The d-d neutrons were measured using a 3He proportional counter, fission chamber, and activated indium sample, whereas the d-t neutrons were detected using activated silicon and copper samples. The triton burnup ratio from KSTAR discharges is found to be in the range 0.01%-0.50% depending on the plasma conditions. The measured burnup ratio is compared with the prompt loss fraction of tritons calculated with the Lorentz orbit code and the classical slowing-down time. The burnup ratio is found to increase as plasma current and classical slowing-down time increase.
Influence of FIMA burnup on actinides concentrations in PWR reactors
Directory of Open Access Journals (Sweden)
Oettingen Mikołaj
2016-01-01
Full Text Available In the paper we present the study on the dependence of actinides concentrations in the spent nuclear fuel on FIMA burnup. The concentrations of uranium, plutonium, americium and curium isotopes obtained in numerical simulation are compared with the result of the post irradiation assay of two spent fuel samples. The samples were cut from the fuel rod irradiated during two reactor cycles in the Japanese Ohi-2 Pressurized Water Reactor. The performed comparative analysis assesses the reliability of the developed numerical set-up, especially in terms of the system normalization to the measured FIMA burnup. The numerical simulations were preformed using the burnup and radiation transport mode of the Monte Carlo Continuous Energy Burnup Code – MCB, developed at the Department of Nuclear Energy, Faculty of Energy and Fuels of AGH University of Science and Technology.
Economic benefits of increased discharge burnup for PWR
International Nuclear Information System (INIS)
Liu Dingqin
1991-09-01
Increased discharge burnup brings a great deal of benefits to the utilities. The total fuel cost and its fraction in different fuel cycle activities have been calculated at different discharge burnup level and given specific conditions by using DQUECO code developed by author himself for the Qinshan NPP1 and Daya Bay NPP on a 12 month cycle. It is also pointed out that increasing burnup from 33.0 GWd/tU to 40.7 GWd/tU for the Daya Bay NPP and increasing burnup from 24.0 GWd/tU to 32.0 GWd/tU for the Qinshan NPP1 are not only technically possible, but also economically beneficial
International Nuclear Information System (INIS)
2018-01-01
The 'Burnup Credit' Working Group was established in 1997 to examine the various parameters, such as the irradiation conditions, the burnup profile and the nuclides (actinides and fission products), to be taken into consideration in the criticality studies that take credit from burnup. This report offers an overview of the work that has been completed or agreed under this framework. It presents the group findings on the following topics: - the axial distribution of nuclides or the axial burnup profile; - methods for validating the actual burnup and its axial distribution; - the calculation of nuclide concentrations after irradiation; - the calculation methods that will be used to determine the effective multiplication factor for systems containing used fuel assemblies. This document gathers together the work carried out by the French Burnup Credit Working Group; it is not a guide validating a particular method for taking burnup credit into account. All of the findings presented here may serve as a basis in industry for defining a method to take account of burnup credit in criticality studies; any industrial body effectively adopting such a method will also be responsible for defining it, based on its knowledge of the used fuel assemblies and the configuration to be addressed. This document forms a collection of the work completed by the Working Group up to 1 January 2007 but does not necessarily reflect ongoing work in the various institutes. (authors)
International Nuclear Information System (INIS)
Haj Hassan, H.; Ghazi, N.; Hainoun, A.
2007-01-01
The codes WIMSD-4 and BORGLES - part of the MTR-PC code package- have been applied to prepare the microscopic cross section library for the main elements of MNSR core for 6 neutron energy groups. The generated library was utilized from the 3D code CITATION to perform the calculation of fuel burn up and depletion including the identification of main fission products and its effects on the multiplication factor. In this regard some modifications have been introduced to the subroutine NUCY in CITATION to incorporate estimating the concentration of the related actinides and fission products. The burn up results indicated that the core life time of MNSR is being mainly estimated by Sm-149 following by Gd-157 and Cd-113. The accumulation of these actinides during 100 continuous operation days caused a reduction of ca. 2 mk for the excess reactivity. This result seems to be in good agreement with the available empirical value of 1.8 mk which relates to the whole discontinuous operation period of the reactor since its start and up to now. The calculation procedure simulates the sporadic operation with an adequate continuous operation period. This approximation is valid for the long lived actinides that mainly dictate the core life time. However, it is an overestimation for the concentration of short lived radioactive products like Xe-135. In the framework of this analysis the possibility of replacement of current MNSR fuel through low enriched fuels has been explored for two the fuel types U02-Mg and U3Si-Al. The results indicate that the first type (UO2-Mg) realize the criticality conditions with low enrichment of 20%, whereas the second type (U3Si-Al) required increasing the uranium enrichment up to 33%. For both fuel types the contribution of plutonium isotopes on the criticality has been also evaluated. Additionally, the influence of mixing burnable absorbers (Gd-113, Cd- 113) with the fresh fuels was investigated to identify their long-term control effect on the
International Nuclear Information System (INIS)
Leege, P.F.A. de; Li, J.M.; Kloosterman, J.L.
1995-04-01
This users' guide gives a description of the functionality and the requested input of the SAS6 code sequence which can be used to perform burnup and criticality calculations based on functional modules from the SCALE-4 code system and libraries. The input file for the SAS6 control module is very similar to that of the other SAS and CSAS control modules available in the SCALE-4 system. Especially the geometry input of SAS6 is quite similar to that of SAS2H. However, the functionality of SAS6 is different from that of SAS2H. The geometry of the reactor lattice can be treated in more detail because use is made of the two-dimensional lattice code WIMS-D/IRI (An adapted version of WIMS-D/4) instead of the one-dimensional transport code XSDRNPM-S. Also the neutron absorption and production rates of nuclides not explicitly specified in the input can be accounted for by six pseudo nuclides. Furthermore, the centre pin can be subdivided into maximal 10 zones to improve the burnup calculation of the centre pin and to obtain more accurate k-infinite values for the assembly. Also the time step specification is more flexible than in the SAS2H sequence. (orig.)
Burnup degree measuring device
International Nuclear Information System (INIS)
Tone, Tatsuo.
1994-01-01
A fixing stand on which a nuclear fuel support and a detector guide support stand are disposed is placed at a predetermined position in a fuel storage pool. Spent nuclear fuels stored in the pool are set in the nuclear fuel support. A closed end of a detector guide tube is secured to the detector guide support stand. The radiation detector is inserted to the detector guide tube, and it is disposed at a predetermined position near the nuclear fuel support. Radiation detection signals detected by the radiation detector are sent to a measuring device for measuring a burnup degree of spent nuclear fuels disposed to the outside of the pool by way of cables. Since the radiation detector is placed near the spent nuclear fuels only upon measurement of the burnup degree, radiation injuries of the radiation detector and the cables are reduced. Further, since the radiation detector and the cables are kept from contact with pool water, radiation decontamination upon maintenance and inspection is not necessary, to facilitate a calibration operation. (I.N.)
Simulation of integral local tests with high-burnup fuel
International Nuclear Information System (INIS)
Gyori, G.
2011-01-01
The behaviour of nuclear fuel under LOCA conditions may strongly depend on the burnup-dependent fuel characteristics, as it has been indicated by recent integral experiments. Fuel fragmentation and the associated fission gas release can influence the integral fuel behaviour, the rod rupture and the radiological release. The TRANSURANUS fuel performance code is a proper tool for the consistent simulation of burnup-dependent phenomena during normal operation and the thermo-mechanical behaviour of the fuel rod in a subsequent accident. The code has been extended with an empirical model for micro-cracking induced FGR and fuel fragmentation and verified against integral LOCA tests of international projects. (author)
Optimum burnup of BAEC TRIGA research reactor
International Nuclear Information System (INIS)
Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque
2013-01-01
Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor
Analysis of rim effect in high burnup (U, Gd)O2 fuel
International Nuclear Information System (INIS)
Kameyama, Takanori; Matsumura, Tetsuo; Kinoshita, Mikiyasu
1992-01-01
Extending burnup of LWR fuel is efficient to reduce the fuel cycle cost and the number of spent fuels. Gadolinia will be mixed in LWR fuels to control the initial reactivity of the high burnup assembly because gadolinia is one of the good burnable poisons of neutron. Rim effect in (U,Gd)O 2 fuel was analyzed by the detailed burnup analysis code VIMBURN. The rim effect in (U,Gd)O 2 is more significant than that in UO 2 fuel below 20 MWd/kgU and the difference of the rim effects in both fuels decreases as burnup proceeds above 20 MWd/kgU. The rim effects in (U,Gd)O 2 and UO 2 fuels are in the same level at high burnup of 80 MWd/kgU when the rim structure forms. The burnup rate of (U,Gd)O 2 rod is less than that of UO 2 rods surrounding the (U,Gd)O 2 rod in the assembly due to neutron absorption of gadolinium. Accordingly local burnup at the peripheral region in (U,Gd)O 2 fuel reaches the high burnup of 80 MWd/kgU a little after the local burnup in UO 2 fuel does. Therefore, it can be predicted that the impact of rim structure on the fuel behaviour in (U,Gd)O 2 fuel is as much as in UO 2 fuel at high burnup of the assembly. (author)
The adequacy of methods used for the approval of high burnup core loading
International Nuclear Information System (INIS)
Sonnenburg, H.G.
2002-01-01
New fuel assembly designs and new core loading strategies are foreseen by most utilities, optimising the use of nuclear fuel in nuclear power plants. Increasing the burn-up to high values above 50 MWd/kg affects the fuel and cladding conditions, which could have safety relevant consequences. It is the task of the safety authorities to assess the impact of these changes with respect to compliance with safety regulations. Usually this assessment is based on code analyses which contain models developed at a time when the burn-up was significantly lower. Because the high burn-up is accompanied with the development of new phenomena like the rim effect on fuel pellets, the codes' models need to be revised for the representation of these new phenomena. The objective of this paper is to present a review of the knowledge base of the fuel phenomena under high-burn-up conditions as seen from safety aspects. The safety relevant fuel rod phenomena will be discussed. It will further provide an assessment of the limitations of the methodologies so far applied in the context of LOCA and RIA transients. The recently started research activities in Germany to improve the methodologies will be presented. (author)
Analysis of temperature reactivity coefficient at very high burn-ups for pressurized water reactor
International Nuclear Information System (INIS)
Yu Shihe; Dai Xiang; Cao Xinrong
2012-01-01
In the high burn-up core, as the initial enrichment increases, the fast/thermal flux ratio also increases. The harder neutron spectrum influences the temperature reactivity coefficients. In this paper, a very high burn-up core was designed, very high burn-up levels was achieved using higher enrichments and various feed assembly and loading pattern options. The CASMO-4/SIMULATE-3 code system is used to model the high burn-up core and calculate temperature reactivity coefficient for the burn-up more than 60 GWD/T. The results show that the hardening of the neutron spectrum leads to more negative moderator temperature coefficients at high burn-ups irrespective of whether or not there is burnable poison; the is little variation with fuel temperature coefficient. (authors)
Performance of code 'FAIR' in IAEA CRP on FUMEX
International Nuclear Information System (INIS)
Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Kakodkar, A.
1996-01-01
A modern fuel performance analysis code FAIR has been developed for analysing high burnup fuel pins of water/heavy water cooled reactors. The code employs finite element method for modelling thermo mechanical behaviour of fuel pins and mechanistic models for modelling various physical and chemical phenomena affecting the behaviour of nuclear reactor fuel pins. High burnup affects such as pellet thermal conductivity degradation, enhanced fission gas release and radial flux redistribution are incorporated in the code FAIR. The code FAIR is capable of performing statistical analysis of fuel pins using Monte Carlo technique. The code is implemented on BARC parallel processing system ANUPAM. The code has recently participated in an International Atomic Energy Agency (IAEA) coordinated research program (CRP) on fuel modelling at extended burnups (FUMEX). Nineteen agencies from different countries participated in this exercise. In this CRP, spread over a period of three years, a number of high burnup fuel pins irradiated at Halden reactor are analysed. The first phase of the CRP is a blind code comparison exercise, where the computed results are compared with experimental results. The second phase consists of modifications to the code based on the experimental results of first phase and statistical analysis of fuel pins. The performance of the code FAIR in this CRP has been very good. The present report highlights the main features of code FAIR and its performance in the IAEA CRP on FUMEX. 14 refs., 5 tabs., ills
Conceptual cask design with burnup credit
International Nuclear Information System (INIS)
Lee, Seong Hee; Ahn, Joon Gi; Hwang, Hae Ryong
2003-01-01
Conceptual design has been performed for a spent fuel transport cask with burnup credit and a neutron-absorbing material to maximize transportation capacity. Both fresh and burned fuel are assumed to be stored in the cask and boral and borated stainless steel are selected for the neutron-absorbing materials. Three different sizes of cask with typical 14, 21 and 52 PWR fuel assemblies are modeled and analyzed with the SCALE 4.4 code system. In this analysis, the biases and uncertainties through validation calculations for both isotopic predictions and criticality calculation for the spent fuel have been taken into account. All of the reactor operating parameters, such as moderator density, soluble boron concentration, fuel temperature, specific power, and operating history, have been selected in a conservative way for the criticality analysis. Two different burnup credit loading curves are developed for boral and borated stainless steel absorbing materials. It is concluded that the spent fuel transport cask design with burnup credit is feasible and is expected to increase cask payloads. (author)
Directory of Open Access Journals (Sweden)
Jung Suk Kim
2015-12-01
Full Text Available The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233U, 242Pu, 150Nd, and 133Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.
The dependence of the global neutronic parameters on the fuel burnup for CANDU SEU43 core
International Nuclear Information System (INIS)
Balaceanu, V.; Pavelescu, M.
2010-01-01
In order to reduce the total fuel costs for the CANDU reactors, mainly by extending the fuel burnup limits, some fuel bundle concepts have been developed in different CANDU owner countries. Therefore, in our Institute the SEU43 (Slightly Enriched Uranium with 43 fuel elements) project was started in early '90s. The neutronic behavior analysis of the CANDU core with SEU43 fuel was an important step in our project design. The objective of this paper is to highline an analysis of the neutronic behavior of the CANDU SEU43 core with the fuel burnup. More exactly, the study refers to the dependence of some global neutronic parameters, mainly the reactivity, on the fuel burnup. Two types of CANDU core were taken into consideration: reference core (without any reactivity devices) and perturbed core (with a strong reactivity system inserted). The considered reactivity system is the Mechanical Control Absorber (MCA) one. The performed parameters are: k eff. values, the MCA reactivity worth and flux distributions. The fuel bundles in the core are SEU43, with the fuel enrichment in U 235 of 0.96% and at nominal power. For the fuel burnup the values are: 0.00 GWd/tU (fresh fuel); 8.00 GWd/tU and 25.00 GWd/tU. For reaching this objective, a global neutronic calculation system named WIMSPIJXYZ LEGENTR is used. Starting from a 69-groups ENDF/B-V based library, this system uses three transport codes: (1) the standard lattice-cell code WIMS, for generating macroscopic cross sections in supercell option and also for burnup calculations; (2) the PIJXYZ code for 3D simulation of the MCA reactivity devices and the 3D correction of the macroscopic cross sections; (3) the LEGENTR 3D transport code for estimating global neutronic parameters (CANDU core). The analysis of the neutronic parameters consists of comparing the obtained results with the similar results calculated with the DRAGON and DIREN codes. This comparison shows a good agreement between these results. (orig.)
Taking burnup credit for interim storage and transportation system for BWR fuels
International Nuclear Information System (INIS)
Yoshioka, K.I.
2000-01-01
In the back-end issues of nuclear fuel cycle, selection of reprocessing or one-through is a big issue. For both of the cases, a reasonable interim storage and transportation system is required. This study proposes an advanced practical monitoring and evaluation system. The system features the followings: (l) Storage racks and transportation casks taking credit for burnup. (2) A burnup estimation system using a compact monitor with Cd- Te detectors and fission chambers. (3) A neutron emission-rate evaluation methodology, especially important for high burnup MOX fuels. (4) A nuclear materials management system for safeguards. Current storage system and transport casks are designed on the basis of a fresh fuel assumption. The assumption is too conservative. Taking burnup credit gives a reasonable design while keeping conservatism. In order to establish a reasonable burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of some modules such as TGBLA, ORIGEN, CITATION, MCNP and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. The code takes operational history such as, power density, void fraction into account. This code is applied to the back-end issues for a more accurate design of a storage and a transportation system. The ORIGEN code is well-known one-point isotope depletion code. In the calculation system, the code calculates isotope compositions using libraries generated from the TGBLA code. The CITATION code, the MCNP code, and the KENO code are three dimensional diffusion code, continuous energy Monte Carlo code, discrete energy Monte Carlo code, respectively. Those codes calculate k- effective of the storage and transportation systems using isotope compositions generated from the ORIGEN code. The CITATION code and the KENO code are usually used for practical designs. The MCNP code is used for reference
PetriCode: A Tool for Template-Based Code Generation from CPN Models
DEFF Research Database (Denmark)
Simonsen, Kent Inge
2014-01-01
levels of abstraction. The elements of the models are annotated with code generation pragmatics enabling PetriCode to use a template-based approach to generate code while keeping the models uncluttered from implementation artefacts. PetriCode is the realization of our code generation approach which has...
Optimal interference code based on machine learning
Qian, Ye; Chen, Qian; Hu, Xiaobo; Cao, Ercong; Qian, Weixian; Gu, Guohua
2016-10-01
In this paper, we analyze the characteristics of pseudo-random code, by the case of m sequence. Depending on the description of coding theory, we introduce the jamming methods. We simulate the interference effect or probability model by the means of MATLAB to consolidate. In accordance with the length of decoding time the adversary spends, we find out the optimal formula and optimal coefficients based on machine learning, then we get the new optimal interference code. First, when it comes to the phase of recognition, this study judges the effect of interference by the way of simulating the length of time over the decoding period of laser seeker. Then, we use laser active deception jamming simulate interference process in the tracking phase in the next block. In this study we choose the method of laser active deception jamming. In order to improve the performance of the interference, this paper simulates the model by MATLAB software. We find out the least number of pulse intervals which must be received, then we can make the conclusion that the precise interval number of the laser pointer for m sequence encoding. In order to find the shortest space, we make the choice of the greatest common divisor method. Then, combining with the coding regularity that has been found before, we restore pulse interval of pseudo-random code, which has been already received. Finally, we can control the time period of laser interference, get the optimal interference code, and also increase the probability of interference as well.
Credit to fuel burnup for criticality safety evaluations in Spain
International Nuclear Information System (INIS)
Conde, J.M.; Recio, M.
1998-01-01
The status of development of burnup credit for criticality safety analyses in Spain is described in this paper. Ongoing activities in the country in this field, both national and international, are resumed. Burnup credit is currently being applied to wet storage of PWR fuel, and credit to integral burnable absorbers is given for BWR fuel storage. It is envisaged to apply burnup credit techniques to the new generation of transport casks now in the design phase. The analysis methodologies submitted for the analyses of PWR and BWR fuel wet storage are outlined. Analysis characteristics specific to burnup credit are described, namely the need to increase the experimental data to allow for a more detailed validation of the depletion codes, and of the criticality codes when applied to spent fuel. Reactivity effects that arise in burnup credit analysis, such as axial and radial effects, fuel irradiation history and others are revised. The methods used to address them in the approved methodologies are outlined. Finally, the regulatory approach used to accept these new analytical methodologies is described. (author)
Designing and Improving Code-based Cryptosystems
Meziani, Mohammed
2014-01-01
In modern cryptography, the security of the most secure cryptographic primitives is based on hard problems coming from number theory such as the factorization and the discrete logarithm problem.However, being mainly based on the intractability of those problems seems to be risky. In 1994, Peter Shor showed how these two problems can be solved in polynomial time using a quantum computer. In contrast, crypttographic primitives based on problems from coding theory are believed to resistquant...
Analysis of burnup credit on spent fuel transport / storage casks - estimation of reactivity bias
International Nuclear Information System (INIS)
Mat sumura, T.; Sasahara, A.; Takei, M.; Takekawa, T.; Kagehira, K.; Nicolaou, G.; Betti, M.
1998-01-01
Chemical analyses of high burnup UO2 (65 GWd/t) and MOX (45 GWd/t) spent fuel pins were carried out. Measured data of nuclides' composition from U234 to P 242 were used for evaluation of ORIGEN-2/82 code and a nuclear fuel design code (NULIF). Critically calculations were executed for transport and storage casks for 52 BWR or 21 PWR spent fuel assemblies. The reactivity biases were evaluated for axial and horizontal profiles of burnup, and historical void fraction (BWR), operational histories such as control rod insertion history, BPR insertion history and others, and calculational accuracy of ORIGEN-2/82 on nuclides' composition. This study shows that introduction of burnup credit has a large merit in criticality safety analysis of casks, even if these reactivity biases are considered. The concept of equivalent uniform burnup was adapted for the present reactivity bias evaluation and showed the possibility of simplifying the reactivity bias evaluation in burnup credit. (authors)
Wavelet based multicarrier code division multiple access ...
African Journals Online (AJOL)
This paper presents the study on Wavelet transform based Multicarrier Code Division Multiple Access (MC-CDMA) system for a downlink wireless channel. The performance of the system is studied for Additive White Gaussian Noise Channel (AWGN) and slowly varying multipath channels. The bit error rate (BER) versus ...
Improvement of burnup analysis for pebble bed reactors with an accumulative fuel loading scheme
International Nuclear Information System (INIS)
Simanullang, Irwan Liapto; Obara, Toru
2015-01-01
Given the limitations of natural uranium resources, innovative nuclear power plant concepts that increase the efficiency of nuclear fuel utilization are needed. The Pebble Bed Reactor (PBR) shows some potential to achieve high efficiency in natural uranium utilization. To simplify the PBR concept, PBR with an accumulation fuel loading scheme was introduced and the Fuel Handling System (FHS) removed. In this concept, the pebble balls are added little by little into the reactor core until the pebble balls reach the top of the reactor core, and all pebble balls are discharged from the core at the end of the operation period. A code based on the MVP/MVP-BURN method has been developed to perform an analysis of a PBR with the accumulative fuel loading scheme. The optimum fuel composition was found using the code for high burnup performance. Previous efforts provided several motivations to improve the burnup performance: First, some errors in the input code were corrected. This correction, and an overall simplification of the input code, was implemented for easier analysis of a PBR with the accumulative fuel loading scheme. Second, the optimum fuel design had been obtained in the infinite geometry. To improve the optimum fuel composition, a parametric survey was obtained by varying the amount of Heavy Metal (HM) uranium per pebble and the degree of uranium enrichment. Moreover, an entire analysis of the parametric survey was obtained in the finite geometry. The results show that improvements in the fuel composition can lead to more accurate analysis with the code. (author)
International Nuclear Information System (INIS)
Pop-Jordanov, J.
1974-01-01
One of the major problems in burnup studies is the reasonably fast and accurate calculation of the space-and-energy dependent neutron flux and reaction rates for realistic power reactor fuel geometries and compositions, and its optimal integration in the global reactor calculations. The scope of the present research was to develop improved methods trying to satisfy the above requirements. In the epithermal region, simple and efficient approximation is proposed which allows the analytical solution for the space dependence of the spherical harmonics flux moments, and hence the derivation of the recurrence relations between he flux moments at successive lethargy pivotal points. A new matrix formalism to invert the coefficient matrix of band structure resulted in a reduce computer time and memory demands. The research on epithermal region is finalized in computing programme SPLET, which calculates the space-lethargy distribution of the spherical harmonics neutron flux moments, and the related integral quantities as reaction rates and resonance integrals. For partial verification of the above methods a Monte Carlo procedure was developed. Using point-wise representation of variables, a flexible and fast convergent integral transport method SEPT i developed. Expanding the neutron source and flux in finite series of arbitrary polynomials, the space-and-energy dependent integral transport equation is transformed into a general linear algebraic form, which is solved numerically. A simple and efficient procedure for deriving multipoint equations and constructing matrix is proposed and examined, and no unwanted oscillations were noticed. The energy point method was combined with the spherical harmonics method as well. A multi zone few-group program SPECTAR for global reactor calculations was developed. For testing, the flux distribution, neutron leakage and effective multiplication factor for the PWR reactor of the power station San Onofre were calculated. In order to verify
Using Laguerre polynomials to compute the matrix exponential in burnup calculations
Energy Technology Data Exchange (ETDEWEB)
She, D.; Zhu, A.; Wang, K. [Dept. of Engineering Physics, Tsinghua Univ., Beijing, 100084 (China)
2012-07-01
An essential part of burnup analysis is to solve the burnup equations. The burnup equations can be regarded as a first-order linear system and solved by means of matrix exponential methods. Because of its large spectrum, it is difficult to compute the exponential of the burnup matrix. Conventional methods of computing the matrix exponential, such as the truncated Taylor expansion and the Pade approximation, are not applicable to burnup calculations. Recently the Chebyshev Rational Approximation Method (CRAM) has been applied to solve burnup matrix exponential and shown to be robust and accurate. However, the main defect of CRAM is that its coefficients are not easy to obtain. In this paper, an orthogonal polynomial expansion method, called Laguerre Polynomial Approximation Method (LPAM), is proposed to compute the matrix exponential in burnup calculations. The polynomial sequence of LPAM can be easily computed in any order and thus LPAM is quite convenient to be utilized into burnup codes. Two typical test cases with the decay and cross-section data taken from the standard ORIGEN 2.1 libraries are calculated for validation, against the reference results provided by CRAM of 14 order. Numerical results show that, LPAM is sufficiently accurate for burnup calculations. The influences of the parameters on the convergence of LPAM are also discussed. (authors)
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)
Image content authentication based on channel coding
Zhang, Fan; Xu, Lei
2008-03-01
The content authentication determines whether an image has been tampered or not, and if necessary, locate malicious alterations made on the image. Authentication on a still image or a video are motivated by recipient's interest, and its principle is that a receiver must be able to identify the source of this document reliably. Several techniques and concepts based on data hiding or steganography designed as a means for the image authentication. This paper presents a color image authentication algorithm based on convolution coding. The high bits of color digital image are coded by the convolution codes for the tamper detection and localization. The authentication messages are hidden in the low bits of image in order to keep the invisibility of authentication. All communications channels are subject to errors introduced because of additive Gaussian noise in their environment. Data perturbations cannot be eliminated but their effect can be minimized by the use of Forward Error Correction (FEC) techniques in the transmitted data stream and decoders in the receiving system that detect and correct bits in error. This paper presents a color image authentication algorithm based on convolution coding. The message of each pixel is convolution encoded with the encoder. After the process of parity check and block interleaving, the redundant bits are embedded in the image offset. The tamper can be detected and restored need not accessing the original image.
FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG
Directory of Open Access Journals (Sweden)
Tukiran Surbakti
2017-12-01
Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.
Measuring Modularity in Open Source Code Bases
Directory of Open Access Journals (Sweden)
Roberto Milev
2009-03-01
Full Text Available Modularity of an open source software code base has been associated with growth of the software development community, the incentives for voluntary code contribution, and a reduction in the number of users who take code without contributing back to the community. As a theoretical construct, modularity links OSS to other domains of research, including organization theory, the economics of industry structure, and new product development. However, measuring the modularity of an OSS design has proven difficult, especially for large and complex systems. In this article, we describe some preliminary results of recent research at Carleton University that examines the evolving modularity of large-scale software systems. We describe a measurement method and a new modularity metric for comparing code bases of different size, introduce an open source toolkit that implements this method and metric, and provide an analysis of the evolution of the Apache Tomcat application server as an illustrative example of the insights gained from this approach. Although these results are preliminary, they open the door to further cross-discipline research that quantitatively links the concerns of business managers, entrepreneurs, policy-makers, and open source software developers.
Modelling of fission gas swelling in the high burnup UO2 fuel
International Nuclear Information System (INIS)
Kim, Dae Ho; Lee, Chan Bock; Bang, Je Gun; Jung, Yeon Ho
1999-06-01
Discharge burnup of the fuel in LWR has been increased to improve the fuel economy, and currently the high burnup fuel of over 70 MWd/kg U-rod avg. is being developed by the fuel vendors worldwide. At high burnup, thermal / mechanical properties of the fuel is known to change and new phenomenon could arise. This report describes the model development on fission gas swelling in high burnup UO 2 fuel. For the low burnup fuel, swelling only by the solid fission products has been considered in the fuel performance analysis. However, at high burnup fuel, swelling by fission gas bubbles can not be neglected anymore. Therefore, fission gas swelling model which can predict bubble swelling of the high burnup UO 2 fuel during the steady-state and the transient conditions in LWR was developed. Based on the bubble growth model, the empirical fission gas swelling model was developed as function of burnup, time and temperature. The model showed that fuel bubble swelling would be proportional to the burnup by the power of 1.157 and to the time by the power of 0.157. Comparison of the model prediction with the measured fission gas swelling data under the various burnup and temperature conditions showed that the model would predict the measured data reasonably well. (author). 20 refs., 8 tabs., 17 figs
Accuracy considerations for Chebyshev rational approximation method (CRAM) in Burnup calculations
Energy Technology Data Exchange (ETDEWEB)
Pusa, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)
2013-07-01
The burnup equations can in principle be solved by computing the exponential of the burnup matrix. However, due to the difficult numerical characteristics of burnup matrices, the problem is extremely stiff and the matrix exponential solution has previously been considered infeasible for an entire burnup system containing over a thousand nuclides. It was recently discovered by the author that the eigenvalues of burnup matrices are generally located near the negative real axis, which prompted introducing the Chebyshev rational approximation method (CRAM) for solving the burnup equations. CRAM can be characterized as the best rational approximation on the negative real axis and it has been shown to be capable of simultaneously solving an entire burnup system both accurately and efficiently. In this paper, the accuracy of CRAM is further studied in the context of burnup equations. The approximation error is analyzed based on the eigenvalue decomposition of the burnup matrix. It is deduced that the relative accuracy of CRAM may be compromised if a nuclide concentration diminishes significantly during the considered time step. Numerical results are presented for two test cases, the first one representing a small burnup system with 36 nuclides and the second one a full a decay system with 1531 nuclides. (authors)
International Nuclear Information System (INIS)
Timofeeva, O.A.; Kurakin, K.U.
2006-01-01
The report deals with fast and thermal neutron flows distribution in structural elements of WWER-1000 fuel assembly and absorbing rods, determination of absorbing isotope burn-up and worth variation in WWER reactor control and protection system rods. Simulation of absorber rod burn-up is provided using code package SAPPHIRE 9 5 end RC W WER allowing detailed description of the core segment spatial model. Maximum burn-up of absorbing rods and respective worth variation of control and protection system rods is determined on the basis of a number of calculations considering known characteristics of fuel cycles (Authors)
EVOLUT - a computer program for fast burnup evaluation
International Nuclear Information System (INIS)
Craciunescu, T.; Dobrin, R.; Stamatescu, L.; Alexa, A.
1999-01-01
EVOLUT is a computer program for burnup evaluation. The input data consist on the one hand of axial and radial gamma-scanning profiles (for the experimental evaluation of the number of nuclei of a fission product - the burnup monitor - at the end of irradiation) and on the other hand of the history of irradiation (the time length and values proportional to the neutron flux for each step of irradiation). Using the equation of evolution of the burnup monitor the flux values are iteratively adjusted, by a multiplier factor, until the calculated number of nuclei is equal to the experimental one. The flux values are used in the equation of evolution of the fissile and fertile nuclei to determine the fission number and consequently the burnup. EVOLUT was successfully used in the analysis of several hundreds of CANDU and TRIGA-type fuel rods. We appreciate that EVOLUT is a useful tool in the burnup evaluation based on gamma spectrometry measurements. EVOLUT can be used on an usual AT computer and in this case the results are obtained in a few minutes. It has an original and user-friendly graphical interface and it provides also output in script MATLAB files for graphical representation and further numerical analysis. The computer program needs simple data and it is valuable especially when a large number of burnup analyses are required quickly. (authors)
ALEPH: An optimal approach to Monte Carlo burn-up
International Nuclear Information System (INIS)
Verboomen, B.
2007-01-01
The incentive of creating Monte Carlo burn-up codes arises from its ability to provide the most accurate locally dependent spectra and flux values in realistic 3D geometries of any type. These capabilities linked with the ability to handle nuclear data not only in its most basic but also most complex form (namely continuous energy cross sections, detailed energy-angle correlations, multi-particle physics, etc.) could make Monte Carlo burn-up codes very powerful, especially for hybrid and advanced nuclear systems (like for instance Accelerator Driven Systems). Still, such Monte Carlo burn-up codes have had limited success mainly due to the rather long CPU time required to carry out very detailed and accurate calculations, even with modern computer technology. To work around this issue, users often have to reduce the number of nuclides in the evolution chains or to consider either longer irradiation time steps and/or larger spatial burn-up cells, jeopardizing the accuracy of the calculation in all cases. There should always be a balance between accuracy and what is (reasonably) achievable. So when the Monte Carlo simulation time is as low as possible and if calculating the cross sections and flux values required for the depletion calculation takes little or no extra time compared to this simulation time, then we can actually be as accurate as we want. That is the optimum situation for Monte Carlo burn-up calculations.The ultimate goal of this work is to provide the Monte Carlo community with an efficient, flexible and easy to use alternative for Monte Carlo burn-up and activation calculations, which is what we did with ALEPH. ALEPH is a Monte Carlo burn-up code that uses ORIGEN 2.2 as a depletion module and any version of MCNP or MCNPX as the transport module. For now, ALEPH has been limited to updating microscopic cross section data only. By providing an easy to understand user interface, we also take away the burden from the user. For the user, it is as if he is
Analysis of the effect of UO{sub 2} high burnup microstructure on fission gas release
Energy Technology Data Exchange (ETDEWEB)
Jernkvist, Lars Olof; Massih, Ali [Quantum Technologies AB, Uppsala Science Park (Sweden)
2002-10-01
This report deals with high-burnup phenomena with relevance to fission gas release from UO{sub 2} nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas.
Analysis of the effect of UO2 high burnup microstructure on fission gas release
International Nuclear Information System (INIS)
Jernkvist, Lars Olof; Massih, Ali
2002-10-01
This report deals with high-burnup phenomena with relevance to fission gas release from UO 2 nuclear fuel. In particular, we study how the fission gas release is affected by local buildup of fissile plutonium isotopes and fission products at the fuel pellet periphery, with subsequent formation of a characteristic high-burnup rim zone micro-structure. An important aspect of these high-burnup effects is the degradation of fuel thermal conductivity, for which prevalent models are analysed and compared with respect to their theoretical bases and supporting experimental data. Moreover, the Halden IFA-429/519.9 high-burnup experiment is analysed by use of the FRAPCON3 computer code, into which modified and extended models for fission gas release are introduced. These models account for the change in Xe/Kr-ratio of produced and released fission gas with respect to time and space. In addition, several alternative correlations for fuel thermal conductivity are implemented, and their impact on calculated fission gas release is studied. The calculated fission gas release fraction in IFA-429/519.9 strongly depends on what correlation is used for the fuel thermal conductivity, since thermal release dominates over athermal release in this particular experiment. The conducted calculations show that athermal release processes account for less than 10% of the total gas release. However, athermal release from the fuel pellet rim zone is presumably underestimated by our models. This conclusion is corroborated by comparisons between measured and calculated Xe/Kr-ratios of the released fission gas
Energy Technology Data Exchange (ETDEWEB)
Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik
2017-06-15
The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.
Determination of axial profit performed burnup credit by SCALE 4.3-system
International Nuclear Information System (INIS)
Miro, R.; Verdu, G.; Munoz-Cobo, J. L.
1998-01-01
SCALE 4.3 is a modular code system designed for realizing standard computational analysis for licensing evaluation. Since now, spent fuel storage pools criticality analysis have been done considering this fuel as fresh, with its maximum enrichment. With burnup credit we can obtain cheaper and compact configurations. The procedure for calculating a spent fuel storage consists of a burnup calculation plus a criticality calculation. We can perform a conservative approximation for the burnup calculations using 1-D results, but, besides the geometry configurations for the 3-D criticality calculation. we need an appropriate approximation to model the burnup axial variation. We assume that for a burnup profile set, the most conservative profile is between the lower and the upper range of this profile, set. We consider only combinations of the maximum and minimum burnup in each axial region, for each burnup range. This gives an estimation of the different burnup shapes effect and the general characteristics of the most conservative shape. (Author) 6 refs
Power excursion analysis for high burnup cores
Energy Technology Data Exchange (ETDEWEB)
Diamond, D.J.; Neymotin, L.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)
1996-02-01
A study was undertaken of power excursions in high burnup cores. There were three objectives in this study. One was to identify boiling water reactor (BWR) and pressurized water reactor (PWR) transients in which there is significant energy deposition in the fuel. Another was to analyze the response of BWRs to the rod drop accident (RDA) and other transients in which there is a power excursion. The last objective was to investigate the sources of uncertainty in the RDA analysis. In a boiling water reactor, the events identified as having significant energy deposition in the fuel were a rod drop accident, a recirculation flow control failure, and the overpressure events; in a pressurized water reactor, they were a rod ejection accident and boron dilution events. The RDA analysis was done with RAMONA-4B, a computer code that models the space- dependent neutron kinetics throughout the core along with the thermal hydraulics in the core, vessel, and steamline. The results showed that the calculated maximum fuel enthalpy in high burnup fuel will be affected by core design, initial conditions, and modeling assumptions. The important uncertainties in each of these categories are discussed in the report.
Energy Technology Data Exchange (ETDEWEB)
Gulliford, N.T.; Hanlon, D. [AEA Technology, Winfrith (United Kingdom)
1996-09-01
Measurements have been made on samples of spent PWR, BWR and MOX fuel in the DIMPLE reactor to validate codes and data used to predict reactivity loss with burnup. This work has been carried out as part of an International Collaborative Programme known as CERES and includes the participation of the UK, France and the USA. Measurements have also been made on thirteen of the major fission product absorbers to validate neutron cross-section data for these isotopes. Comparisons with calculations based on AEA Technology`s WIMS Reactor Physics Code using the JEF2.2 data library show excellent agreement for a wide range of irradiated fuel samples. (author)
Strategies for Application of Isotopic Uncertainties in Burnup Credit
Energy Technology Data Exchange (ETDEWEB)
Gauld, I.C.
2002-12-23
Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the potential capacity of spent fuel transport and storage casks employing burnup credit. Methods that can provide a more accurate and realistic estimate of the uncertainty may enable increased spent fuel cask capacity and fewer casks needing to be transported, thereby reducing regulatory burden on licensee while maintaining safety for transporting spent fuel. This report surveys several different best-estimate strategies for considering the effects of nuclide uncertainties in burnup-credit analyses. The potential benefits of these strategies are illustrated for a prototypical burnup-credit cask design. The subcritical margin estimated using best-estimate methods is discussed in comparison to the margin estimated using conventional bounding methods of uncertainty propagation. To quantify the comparison, each of the strategies for estimating uncertainty has been performed using a common database of spent fuel isotopic assay measurements for pressurized-light-water reactor fuels and predicted nuclide concentrations obtained using the current version of the SCALE code system. The experimental database applied in this study has been significantly expanded to include new high-enrichment and high-burnup spent fuel assay data recently published for a wide range of important burnup-credit actinides and fission products. Expanded rare earth fission-product measurements performed at the Khlopin Radium Institute in Russia that contain the only known publicly-available measurement for {sup 103
Energy information data base: report number codes
International Nuclear Information System (INIS)
1979-09-01
Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used
Energy information data base: report number codes
Energy Technology Data Exchange (ETDEWEB)
None
1979-09-01
Each report processed by the US DOE Technical Information Center is identified by a unique report number consisting of a code plus a sequential number. In most cases, the code identifies the originating installation. In some cases, it identifies a specific program or a type of publication. Listed in this publication are all codes that have been used by DOE in cataloging reports. This compilation consists of two parts. Part I is an alphabetical listing of report codes identified with the issuing installations that have used the codes. Part II is an alphabetical listing of installations identified with codes each has used. (RWR)
Random Coding Bounds for DNA Codes Based on Fibonacci Ensembles of DNA Sequences
2008-07-01
Highway, Suite 1204, Arlington, VA 22202-4302, and to the Office of Management and Budget, Paperwork Reduction Project (0704-0188) Washington, DC...COVERED (From - To) 6 Jul 08 – 11 Jul 08 4. TITLE AND SUBTITLE RANDOM CODING BOUNDS FOR DNA CODES BASED ON FIBONACCI ENSEMBLES OF DNA SEQUENCES...sequences which are generalizations of the Fibonacci sequences. 15. SUBJECT TERMS DNA Codes, Fibonacci Ensembles, DNA Computing, Code Optimization 16
Energy Technology Data Exchange (ETDEWEB)
Leege, P.F.A. de; Li, J.M.; Kloosterman, J.L.
1995-04-01
This users` guide gives a description of the functionality and the requested input of the SAS6 code sequence which can be used to perform burnup and criticality calculations based on functional modules from the SCALE-4 code system and libraries. The input file for the SAS6 control module is very similar to that of the other SAS and CSAS control modules available in the SCALE-4 system. Especially the geometry input of SAS6 is quite similar to that of SAS2H. However, the functionality of SAS6 is different from that of SAS2H. The geometry of the reactor lattice can be treated in more detail because use is made of the two-dimensional lattice code WIMS-D/IRI (An adapted version of WIMS-D/4) instead of the one-dimensional transport code XSDRNPM-S. Also the neutron absorption and production rates of nuclides not explicitly specified in the input can be accounted for by six pseudo nuclides. Furthermore, the centre pin can be subdivided into maximal 10 zones to improve the burnup calculation of the centre pin and to obtain more accurate k-infinite values for the assembly. Also the time step specification is more flexible than in the SAS2H sequence. (orig.).
Burnup credit implementation in WWER spent fuel management systems: Status and future aspects
International Nuclear Information System (INIS)
Manolova, M.
1998-01-01
This paper describes the motivation for possible burnup credit implementation in WWER spent fuel management systems in Bulgaria. The activities being done are described, namely: the development and verification of a 3D few-group diffusion burnup model; the application of the KORIGEN code for evaluation of WWER fuel nuclear inventory during reactor core lifetime and after spent fuel discharge; using the SCALE modular system (PC Version 4.1) for criticality safety analyses of spent fuel storage facilities. Future plans involving such important tasks as validation and verification of computer systems and libraries for WWER burnup credit analysis are shown. (author)
Incorporating Code-Based Software in an Introductory Statistics Course
Doehler, Kirsten; Taylor, Laura
2015-01-01
This article is based on the experiences of two statistics professors who have taught students to write and effectively utilize code-based software in a college-level introductory statistics course. Advantages of using software and code-based software in this context are discussed. Suggestions are made on how to ease students into using code with…
Taking burnup credit for interim storage and transportation system for BWR fuels
International Nuclear Information System (INIS)
Yoshioka, Ken-ichi; Ando, Y.; Kumanomido, H.; Sasaki, T.; Mitsuhashi, I.; Ueda, M.
2001-01-01
In order to establish a realistic burnup credit design system, a calculation system has been developed for determining isotope compositions, burnup, and criticality. The calculation system consists of several modules such as TGBLA, ORIGEN, CITATION, MCNP, and KENO. The TGBLA code is a fuel design code for LWR fuels developed in TOSHIBA Corporation. A compact measurement system for a fuel assembly has been being developed to meet requirements for the burnup determination, the neutron emission-rate evaluation, and the nuclear materials management. For a spent MOX fuel, a neutron emission rate measurement method has been being developed. The system consists of Cd-Te detectors and / or fission chambers. Some model calculations were carried out for the latest design BWR fuels. The effect of taking burnup credit for a transportation cask is shown. (authors)
Technical development on burn-up credit for spent LWR fuels
International Nuclear Information System (INIS)
Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori
2000-10-01
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)
Technical Development on Burn-up Credit for Spent LWR Fuel
Energy Technology Data Exchange (ETDEWEB)
Gauld, I.C.
2001-12-26
Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.
Password Authentication Based on Fractal Coding Scheme
Directory of Open Access Journals (Sweden)
Nadia M. G. Al-Saidi
2012-01-01
Full Text Available Password authentication is a mechanism used to authenticate user identity over insecure communication channel. In this paper, a new method to improve the security of password authentication is proposed. It is based on the compression capability of the fractal image coding to provide an authorized user a secure access to registration and login process. In the proposed scheme, a hashed password string is generated and encrypted to be captured together with the user identity using text to image mechanisms. The advantage of fractal image coding is to be used to securely send the compressed image data through a nonsecured communication channel to the server. The verification of client information with the database system is achieved in the server to authenticate the legal user. The encrypted hashed password in the decoded fractal image is recognized using optical character recognition. The authentication process is performed after a successful verification of the client identity by comparing the decrypted hashed password with those which was stored in the database system. The system is analyzed and discussed from the attacker’s viewpoint. A security comparison is performed to show that the proposed scheme provides an essential security requirement, while their efficiency makes it easier to be applied alone or in hybrid with other security methods. Computer simulation and statistical analysis are presented.
Mean-based neural coding of voices.
Andics, Attila; McQueen, James M; Petersson, Karl Magnus
2013-10-01
The social significance of recognizing the person who talks to us is obvious, but the neural mechanisms that mediate talker identification are unclear. Regions along the bilateral superior temporal sulcus (STS) and the inferior frontal cortex (IFC) of the human brain are selective for voices, and they are sensitive to rapid voice changes. Although it has been proposed that voice recognition is supported by prototype-centered voice representations, the involvement of these category-selective cortical regions in the neural coding of such "mean voices" has not previously been demonstrated. Using fMRI in combination with a voice identity learning paradigm, we show that voice-selective regions are involved in the mean-based coding of voice identities. Voice typicality is encoded on a supra-individual level in the right STS along a stimulus-dependent, identity-independent (i.e., voice-acoustic) dimension, and on an intra-individual level in the right IFC along a stimulus-independent, identity-dependent (i.e., voice identity) dimension. Voice recognition therefore entails at least two anatomically separable stages, each characterized by neural mechanisms that reference the central tendencies of voice categories. Copyright © 2013 Elsevier Inc. All rights reserved.
Three Methods for Occupation Coding Based on Statistical Learning
Directory of Open Access Journals (Sweden)
Gweon Hyukjun
2017-03-01
Full Text Available Occupation coding, an important task in official statistics, refers to coding a respondent’s text answer into one of many hundreds of occupation codes. To date, occupation coding is still at least partially conducted manually, at great expense. We propose three methods for automatic coding: combining separate models for the detailed occupation codes and for aggregate occupation codes, a hybrid method that combines a duplicate-based approach with a statistical learning algorithm, and a modified nearest neighbor approach. Using data from the German General Social Survey (ALLBUS, we show that the proposed methods improve on both the coding accuracy of the underlying statistical learning algorithm and the coding accuracy of duplicates where duplicates exist. Further, we find defining duplicates based on ngram variables (a concept from text mining is preferable to one based on exact string matches.
Status of burnup credit implementation and research in Switzerland
International Nuclear Information System (INIS)
Grimm, P.
2001-01-01
Burnup credit has recently been approved by the Swiss licensing authority for the spent-fuel storage pool of a PWR plant for fuel exceeding the originally licensed initial enrichment. The criticality safety assessment is based on a configuration consisting of a small number (approximately a reload batch) of fresh assemblies surrounded by assemblies having a burnup corresponding to the minimum value in the top 1 m section after one cycle of irradiation. The allowable initial enrichment in this configuration is about 0.5% higher than for all fresh fuel. A central storage facility for all types of radioactive wastes from Switzerland, including cask storage of spent fuel assemblies is being commissioned presently. The first applications for licenses for casks to be used in this facility have been submitted. Credit for burnup has not been requested in these applications (conforming to the original licenses of the casks in their countries of origin), but utilities are interested in burnup credit for fuel with higher initial enrichments. Reactivity worth measurements as well as chemical assays of spent fuel samples in the LWR-PROTEUS facility at PSI are in detailed planning currently. The experiments, scheduled to start in 2001, will be performed in cooperation with the Swiss utilities and their fuel vendors. Although the focus of interest of these partners is on validation of in-core fuel management tools, the same experiments are also applicable to burnup credit, and contacts with further potential partners interested in this field are underway. (author)
Development and verification of Monte Carlo burnup calculation system
International Nuclear Information System (INIS)
Ando, Yoshihira; Yoshioka, Kenichi; Mitsuhashi, Ishi; Sakurada, Koichi; Sakurai, Shungo
2003-01-01
Monte Carlo burnup calculation code system has been developed to evaluate accurate various quantities required in the backend field. From the Actinide Research in a Nuclear Element (ARIANE) program, by using, the measured nuclide compositions of fuel rods in the fuel assemblies irradiated in the commercial Netherlands BWR, the analyses have been performed for the code system verification. The code system developed in this paper has been verified through analysis for MOX and UO2 fuel rods. This system enables to reduce large margin assumed in the present criticality analysis for LWR spent fuels. (J.P.N.)
The CORSYS neutronics code system
International Nuclear Information System (INIS)
Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.
1994-01-01
The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs
International Nuclear Information System (INIS)
Szuta, M.; Dąbrowski, L.
2013-01-01
Crossing the experimental critical fuel temperature dependent on burn-up, an onset of fission gas burst release is observed. This observed phenomena can be explained by assumption that the fission gas immobilization in the uranium dioxide irradiated to a fluency of greater than 10 19 fissions/cm 3 is mainly due to radiation induced chemical activity. Application of the “ab initio” method show that the bond energy of Xenon and Krypton is equal to –1.23 eV, and –3.42 eV respectively. Assuming further that the gas chemically bound can be released mainly in the process of re-crystallization and modifying the differential equation of Ainscough of grain growth by including the burn-up dependence and the experimental data of limiting grain size in function of the fuel temperature for the un-irradiated and irradiated fuel we can re-construct the experimental curve of Vitanza. (authors)
Ultrasonic measurement of high burn-up fuel elastic properties
International Nuclear Information System (INIS)
Laux, D.; Despaux, G.; Augereau, F.; Attal, J.; Gatt, J.; Basini, V.
2006-01-01
The ultrasonic method developed for the evaluation of high burn-up fuel elastic properties is presented hereafter. The objective of the method is to provide data for fuel thermo-mechanical calculation codes in order to improve industrial nuclear fuel and materials or to design new reactor components. The need for data is especially crucial for high burn-up fuel modelling for which the fuel mechanical properties are essential and for which a wide range of experiments in MTR reactors and high burn-up commercial reactor fuel examinations have been included in programmes worldwide. To contribute to the acquisition of this knowledge the LAIN activity is developing in two directions. First one is development of an ultrasonic focused technique adapted to active materials study. This technique was used few years ago in the EdF laboratory in Chinon to assess the ageing of materials under irradiation. It is now used in a hot cell at ITU Karlsruhe to determine the elastic moduli of high burnup fuels from 0 to 110 GWd/tU. Some of this work is presented here. The second on going programme is related to the qualification of acoustic sensors in nuclear environments, which is of a great interest for all the methods, which work, in a hostile nuclear environment
Protograph-Based Raptor-Like Codes
Divsalar, Dariush; Chen, Tsung-Yi; Wang, Jiadong; Wesel, Richard D.
2014-01-01
Theoretical analysis has long indicated that feedback improves the error exponent but not the capacity of pointto- point memoryless channels. The analytic and empirical results indicate that at short blocklength regime, practical rate-compatible punctured convolutional (RCPC) codes achieve low latency with the use of noiseless feedback. In 3GPP, standard rate-compatible turbo codes (RCPT) did not outperform the convolutional codes in the short blocklength regime. The reason is the convolutional codes for low number of states can be decoded optimally using Viterbi decoder. Despite excellent performance of convolutional codes at very short blocklengths, the strength of convolutional codes does not scale with the blocklength for a fixed number of states in its trellis.
Quantum BCH Codes Based on Spectral Techniques
International Nuclear Information System (INIS)
Guo Ying; Zeng Guihua
2006-01-01
When the time variable in quantum signal processing is discrete, the Fourier transform exists on the vector space of n-tuples over the Galois field F 2 , which plays an important role in the investigation of quantum signals. By using Fourier transforms, the idea of quantum coding theory can be described in a setting that is much different from that seen that far. Quantum BCH codes can be defined as codes whose quantum states have certain specified consecutive spectral components equal to zero and the error-correcting ability is also described by the number of the consecutive zeros. Moreover, the decoding of quantum codes can be described spectrally with more efficiency.
Burnup studies of the subcritical fusion-driven in-zinerator
International Nuclear Information System (INIS)
Persson, C. M.; Gudowski, W.; Venneri, F.
2007-01-01
A fusion-driven subcritical core, 'In-Zinerator', has been proposed for nuclear waste transmutation [1]. In this concept, a powerful Z-pinch neutron source will produce pulses of 14 MeV neutrons that multiply in a surrounding subcritical core consisting of spent fuel from the LWR fuel cycle or from deep burn high temperature reactors. The proposed design has pulse frequency 0.1 Hz and a thermal power of 3 GWth. The Z-pinch fusion experiment is located at Sandia Laboratories, USA, and can today fire once a day. However, investigations have been made how to increase the frequency to several fires per minute. Each fire yields 300 MJ corresponding to 1020 neutrons per pulse. The source chamber will in the In-Zinerator concept be surrounded by spent fuel to reach an effective multiplication factor, k e ff, of 0.97. The core will be cooled by liquid lead. In this paper, the burnup of different fuel compositions in the In-Zinerator will be studied as function of initial k e ff. The Monte Carlo based continuous energy burnup code MCB [2][3]will be used. References: [1] B.B. Cipiti, Fusion Transmutation of Waste and the Role of the In-Zinerator in the Nuclear Fuel Cycle, Sandia Report SAND2006-3522, Sandia National Laboratories, USA, 2006. [2] J. Cetnar, J Wallenius and W Gudowski, MCB: A continuous energy Monte-Carlo burnup simulation code, Actinide and fission product partitioning and transmutation, Proc. of the Fifth Int. Information Exchange Meeting, Mol, Belgium, 25-27 November 1998, 523, OECD/NEA, 1998. [3] http://www.nea.fr/abs/html/nea-1643.html
Issues related to criticality safety analysis for burnup credit applications
International Nuclear Information System (INIS)
DeHart, M.D.; Parks, C.V.
1995-01-01
Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design
ABB PWR fuel design for high burnup
International Nuclear Information System (INIS)
Nilsson, S.; Jourdain, P.; Limback, M.; Garde, A.M.
1998-01-01
Corrosion, hydriding and irradiation induced growth of a based materials are important factors for the high burnup performance of PWR fuel. ABB has developed a number of Zr based alloys to meet the need for fuel that enables operation to elevated burnups. The materials include composition and processing optimised Zircaloy 4 (OPTIN TM ) and Zircaloy 2 (Zircaloy 2P), as well as advanced Zr based alloys with chemical compositions outside the composition specified for Zircaloy. The advanced alloys are either used as Duplex or as single component claddings. The Duplex claddings have an inner component of Zircaloy and an outer layer of Zr with small additions of alloying elements. ABB has furthermore improved the dimensional stability of the fuel assembly by developing stiffer and more bow resistant guide tubes while debris related fuel failures have been eliminated from ABB fuel by introducing the Guardian TM grid. Intermediate flow mixers that improve the thermal hydraulic performance and the dimensional stability of the fuel has also been developed within ABB. (author)
Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4
International Nuclear Information System (INIS)
Tombakoglu, M.; Cecen, Y.
2001-01-01
In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)
Influence of porosity formation on irradiated UO2 fuel thermal conductivity at high burnup
Roostaii, B.; Kazeminejad, H.; Khakshournia, S.
2016-10-01
Based on the existing low temperature high burnup gaseous swelling model for UO2 fuel, the matrix swelling terms are calculated and the formation of total volume porosity up to burnup of 120 MWd/KgU is computed. The irradiated UO2 thermal conductivity model based on the Maxwell-Eucken correlation for porosity factor is selected as a case study and the calculation of porosity evolution with burnup is carried out. It is shown that taking into account the formation of porosity with burnup compared to the case with constant porosity equal to as-fabricated value leads to a decrease in the UO2 fuel thermal conductivity up to 15% at high burnup values of 120 MWd/kgU. Results of the calculations are also compared with the available experimental data and good agreement was found. The conducted parametric study clearly demonstrated the impact of the key parameters on the results of the present investigation.
Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.
2018-05-01
One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.
SENSITIVITY AND UNCERTAINTY ANALYSIS OF COMMERCIAL REACTOR CRITICALS FOR BURNUP CREDIT
International Nuclear Information System (INIS)
Radulescu, Georgeta; Mueller, Don; Wagner, John C.
2009-01-01
The purpose of this study is to provide insights into the neutronic similarities that may exist between a generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the type of CRC state-points that may be applicable for validation of burnup credit criticality safety calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-section sensitivity and uncertainty analysis methods developed at Oak Ridge National Laboratory and the TSUNAMI set of tools in the SCALE code system as a means to investigate system similarity on an integral and nuclide-reaction specific level. The results indicate that, except for the fresh fuel core configuration, all analyzed CRC state-points are either highly similar, similar, or marginally similar to a generic cask containing spent nuclear fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. Based on the integral system parameter, C k , approximately 30 of the 40 CRC state-points are applicable to validation of burnup credit in the generic cask containing typical spent fuel assemblies with burnups ranging from 10 to 60 GWd/MTU. The state-points providing the highest similarity (C k > 0.95) were attained at or near the end of a reactor cycle. The C k values are dominated by neutron reactions with major actinides and hydrogen, as the sensitivities of these reactions are much higher than those of the minor actinides and fission products. On a nuclide-reaction specific level, the CRC state-points provide significant similarity for most of the actinides and fission products relevant to burnup credit. A comparison of energy-dependent sensitivity profiles shows a slight shift of the CRC K eff sensitivity profiles toward higher energies in the thermal region as compared to the K eff sensitivity profile of the generic cask. Parameters representing
Investigation of very high burnup UO{sub 2} fuels in Light Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Cappia, Fabiola
2017-03-27
fuel mechanical properties and their relationship with the local microstructure at high burnup has been recognised, being one of the factors influencing Pellet-Cladding Mechanical Interaction (PCMI). The knowledge of the fuel mechanical properties has also fundamental importance to assess the mechanical integrity of the spent fuel during the back end of the fuel cycle. In this context, the scope of this work was twofold. The first task was the experimental study of the fuel microhardness and Young's modulus in high burnup UO{sub 2} fuels and their relationship with the local porosity, which has a major impact on their variation. Moreover, assessment of the accumulation of the decay damage during storage and its influence on the fuel microhardness has been carried out, in the framework of safety studies on the back end of the fuel cycle at high burnup. The second task consisted in the evaluation of the porosity and pore size distribution evolution in high burnup fuel, with particular focus on the HBS porosity. The experimental relationship between the high burnup fuel Young's modulus and local porosity obtained through combination of acoustic microscopy and microindentation measurements has been compared to the material property correlations commonly used in fuel performance codes, which are based on data from characterization of unirradiated UO{sub 2}. The investigation has revealed that the relationship is similar for non-irradiated and irradiated material, but in the latter case an additional factor that takes into account the Young's modulus decrease due to burnup accumulation has to be included in the correlation to match the experimental values. First analysis of the fuel microhardness as a function of the accumulated decay damage has shown that fuel microhardness does not significantly increase when the dose due to the additional decay damage accumulated during storage reaches ∼ 0.1 dpa, in agreement with what observed in unirradiated {sup 238}Pu
Comparison of analysis methods for burnup credit applications
International Nuclear Information System (INIS)
Sanders, T.L.; Brady, M.C.; Renier, J.P.; Parks, C.V.
1989-01-01
The current approach used for the development and certification of spent fuel storage and transport casks requires an assumption of fresh fuel isotopics in the criticality safety analysis. However, it has been shown that there is a considerable reactivity reduction when the isotopics representative of the depleted (or burned) fuel are used in a criticality analysis. Thus, by taking credit for the burned state of the fuel (i.e., burnup credit), a cask designer could achieve a significant increase in payload. Accurate prediction of k eff for spent fuel arrays depends both on the criticality safety analysis and the prediction of the spent fuel isotopics via a depletion analysis. Spent fuel isotopics can be obtained from detailed multidimensional reactor analyses, e.g. the code PDQ, or from point reactor burnup models. These reactor calculations will help verify the adequacy of the isotopics and determine Δk eff biases for various analysis assumptions (with and without fission products, actinide absorbers, burnable poison rods, etc.). New software developed to interface PDQ multidimensional isotopics with KENO V.a reactor and cask models is described. Analyses similar to those performed for the reactor cases are carried out with a representative burnup credit cask model using the North Anna fuel. This paper presents the analysis methodology that has been developed for evaluating the physics issues associated with burnup credit. It is applicable in the validation and characterization of fuel isotopics as well as in determining the influence of various analysis assumptions in terms of δk eff . The methodology is used in the calculation of reactor restart criticals and analysis of a typical burnup credit cask
A Framework for Reverse Engineering Large C++ Code Bases
Telea, Alexandru; Byelas, Heorhiy; Voinea, Lucian
2009-01-01
When assessing the quality and maintainability of large C++ code bases, tools are needed for extracting several facts from the source code, such as: architecture, structure, code smells, and quality metrics. Moreover, these facts should be presented in such ways so that one can correlate them and
A Framework for Reverse Engineering Large C++ Code Bases
Telea, Alexandru; Byelas, Heorhiy; Voinea, Lucian
2008-01-01
When assessing the quality and maintainability of large C++ code bases, tools are needed for extracting several facts from the source code, such as: architecture, structure, code smells, and quality metrics. Moreover, these facts should be presented in such ways so that one can correlate them and
Energy Technology Data Exchange (ETDEWEB)
Santos, Nadia Rodrigues dos, E-mail: nadia.santos@ifrj.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: malu@ien.gov.br, E-mail: zrlima@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2015-07-01
In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)
Improvements on burnup chain model and group cross section library in the SRAC system
International Nuclear Information System (INIS)
Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.
1992-01-01
Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)
A survey of previous and current industry-wide efforts regarding burnup credit
International Nuclear Information System (INIS)
Jones, R.H.
1989-01-01
Sandia has examined the matter of burnup credit from the perspective of physics, logistics, risk, and economics. A limited survey of the nuclear industry has been conducted to get a feeling for the actual application of burnup credit. Based on this survey, it can be concluded that the suppliers of spent fuel storage and transport casks are in general agreement that burnup credit offers the potential for improvements in cask efficiency without increasing the risk of accidental criticality. The actual improvement is design-specific but limited applications have demonstrated that capacity increases in the neighborhood of 20 percent are not unrealistic. A number of these vendors acknowledge that burnup credit has not been reduced to practice in cask applications and suggest that operational considerations may be more important to regulatory acceptance than to the physics. Nevertheless, the importance of burnup credit to the nuclear industry as a cask design and analysis tool has been confirmed by this survey
TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES
International Nuclear Information System (INIS)
DOE
1997-01-01
A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k eff , of a spent nuclear fuel package. Fifty-seven UO 2 , UO 2 /Gd 2 O 3 , and UO 2 /PuO 2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k eff (which can be a function of the trending parameters) such that the biased k eff , when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection
Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1
Energy Technology Data Exchange (ETDEWEB)
None, None
1997-04-01
A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package
Topical Report on Actinide-Only Burnup Credit for PWR Spent Nuclear Fuel Packages. Revision 2
Energy Technology Data Exchange (ETDEWEB)
None, None
1998-09-01
The objective of this topical report is to present to the NRC for review and acceptance a methodology for using burnup credit in the design of criticality control systems for PWR spent fuel transportation packages, while maintaining the criticality safety margins and related requirements of 10 CFR Part 71 and 72. The proposed methodology consists of five major steps as summarized below: (1) Validate a computer code system to calculate isotopic concentrations in SNF created during burnup in the reactor core and subsequent decay. (2) Validate a computer code system to predict the subcritical multiplication factor, keff, of a spent nuclear fuel package. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). and (5) Verify that SNF assemblies meet the package loading criteria and confirm proper fuel assembly selection prior to loading. (This step is required but the details are outside the scope of this topical report.) When reviewed and accepted by the NRC, this topical report will serve as a criterion document for criticality control analysts and will provide steps for the use of actinide-only burnup credit in the design of criticality control systems. The NRC-accepted burnup credit methodology will be used by commercial SNF storage and transportation package designers. Design-specific burnup credit criticality analyses will be defined, developed, and documented in the Safety Analysis Report (SAR) for each specific storage or transportation package that uses burnup credit. These SARs will then be submitted to the NRC for review and approval. This topical report is expected to be referenced in a number of storage and transportation cask applications to be submitted by commercial cask and canister designers to the NRC. Therefore, NRC acceptance of this topical report will result in increased efficiency of the
Application of burnup credit concept to transport
International Nuclear Information System (INIS)
Futamura, Yoshiaki; Nakagome, Yoshihiro.
1994-01-01
For the design and safety assessment of the casks for transporting spent fuel, the fuel contained in them has been assumed to be new fuel. The reason is, it was difficult to evaluate the variation of the reactivity of fuel, and the research on the affecting factors and the method of measuring burnup were not much advanced. Recently, high burnup fuel has been adopted, and initial degree of enrichment rose. The research has been advanced for pursuing the economy of the casks for spent fuel, and burnup credit has become applicable to their design and safety assessment. As the result, the containing capacity increases by about 20%. When burnup credit is considered, it is necessary to confirm accurately the burnup of spent fuel. The burnup dependence of the concentration of fissile substances and neutron emissivity, the coolant void dependence of the concentration of fissile substances, and the relation of neutron multiplication rate with initial degree of enrichment or burnup are discussed. The conceptual design of casks considering burnup credit and its assessment, the merit, problem and the countermeasures to it when burnup credit is introduced are described. (K.I.)
Phenomena and parameters important to burnup credit
International Nuclear Information System (INIS)
Parks, C.V.; Dehart, M.D.; Wagner, J.C.
2001-01-01
Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water- reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the United States and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given. (author)
Methodology for burnup credit application for WWER-440 reactors in Slovakia
International Nuclear Information System (INIS)
Chrapciak, V.
2006-01-01
Improved calculational methods allow one to take credit for the reactivity reduction associated with fuel burnup. This means reducing the analysis conservative while maintaining an adequate criticality safety margin. Application of burnup credit requires knowledge of the reactivity state of the irradiated fuel for which burnup credit is taken. The isotopic inventory and reactivity has to by calculated with validated codes. By depletion calculation is necessary to take into account: 1) List of used nuclides 2) Cooling time 3) Core parameters by irradiation - Fuel temperature - Moderator temperature/density - Soluble boron - Specific power and operating history 4) Axial and horizontal burnup profiles In this paper is numerical evaluation of impact of above parameters on reactivity for WWER-440 fuel (enrichment 3.6%) in condition of wet storage presented. The bounding conditions are defined. The SCALE 5.0 system is used for depletion and criticality calculation (Authors)
New burnup calculation of TRIGA IPR-R1 reactor
International Nuclear Information System (INIS)
Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.
2015-01-01
The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)
Energy Technology Data Exchange (ETDEWEB)
Belo, Thiago F.; Fiel, Joao Claudio B., E-mail: thiagofbelo@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)
2015-07-01
Nuclear reactor core analysis involves neutronic modeling and the calculations require problem dependent nuclear data generated with few neutron energy groups, as for instance the neutron cross sections. The methods used to obtain these problem-dependent cross sections, in the reactor calculations, generally uses nuclear computer codes that require a large processing time and computational memory, making the process computationally very expensive. Presently, analysis of the macroscopic cross section, as a function of nuclear parameters, has shown a very distinct behavior that cannot be represented by simply using linear interpolation. Indeed, a polynomial representation is more adequate for the data parameterization. To provide the cross sections of rapidly and without the dependence of complex systems calculations, this work developed a set of parameterized cross sections, based on the Tchebychev polynomials, by fitting the cross sections as a function of nuclear parameters, which include fuel temperature, moderator temperature and density, soluble boron concentration, uranium enrichment, and the burn-up. In this study is evaluated the problem-dependent about fission, scattering, total, nu-fission, capture, transport and absorption cross sections for a typical PWR fuel element reactor, considering burn-up cycle. The analysis was carried out with the SCALE 6.1 code package. The results of comparison with direct calculations with the SCALE code system and also the test using project parameters, such as the temperature coefficient of reactivity and fast fission factor, show excellent agreements. The differences between the cross-section parameterization methodology and the direct calculations based on the SCALE code system are less than 0.03 percent. (author)
Sandia National Laboratories analysis code data base
Energy Technology Data Exchange (ETDEWEB)
Peterson, C.W.
1994-11-01
Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.
Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1
Directory of Open Access Journals (Sweden)
Muhammad Atta
2011-01-01
Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.
International Nuclear Information System (INIS)
Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji
1999-03-01
As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.
1986-09-01
Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)
Optimal Reliability-Based Code Calibration
DEFF Research Database (Denmark)
Sørensen, John Dalsgaard; Kroon, I. B.; Faber, Michael Havbro
1994-01-01
Calibration of partial safety factors is considered in general, including classes of structures where no code exists beforehand. The partial safety factors are determined such that the difference between the reliability for the different structures in the class considered and a target reliability...
International Nuclear Information System (INIS)
Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan
2015-01-01
Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes
Burnup measurement study and prototype development in HTR-PM
International Nuclear Information System (INIS)
Yan Weihua; Zhang Zhao; Xiao Zhigang; Zhang Liguo
2014-01-01
In a pebble-bed core which employs the multi-pass scheme, it is mandatory to determine the burnup of each pebble after the pebble has been extracted from the core in order to determine whether its design burnup has been reached or whether it has to be reinserted into the core again. The burnup of the fuel pebbles can be determined by measuring the activity of 137 Cs with an HPGe detector because of their good correspondence, which is independent of the irradiation history in the core. Based on experiments and Geant4 simulation, the correction factor between the fuel and calibration source was derived by using the efficiency transfer method. By optimizing spectrum analysis algorithm and parameters, the relative standard deviation of the 137 Cs activity can be still controlled below 3.0% despite of the presence of interfering peaks. On the foundation of the simulation and experiment research, a complete solution for burnup measurement system in HTR-PM is provided. (authors)
A simplified burnup calculation strategy with refueling in static molten salt reactor
International Nuclear Information System (INIS)
Srivastava, A.K.; Gupta, Anurag; Krishnani, P.D.
2015-01-01
Molten Salt Reactors, by nature can be refuelled and reprocessed online. Thus, a simulation methodology has to be developed which can consider online refueling and reprocessing aspect of the reactor. To cater such needs a simplified burnup calculation strategy to account for refueling and removal of molten salt fuel at any desired burnup has been identified in static molten salt reactor in batch mode as a first step of way forward. The features of in-house code ITRAN has been explored for such calculations. The code also enables us to estimate the reactivity introduced in the system due to removal of any number of considered nuclides at any burnup. The effect of refueling fresh fuel and removal of burned fuel has been studied in batch mode with in-house code ITRAN. The effect of refueling and burnup on change in reactivity per day has been analyzed. The analysis of removal of 233 Pa at a particular burnup has been carried out. The similar analysis has been performed for some other nuclides also. (author)
Plutonium and Minor Actinides Recycling in Standard BWR using Equilibrium Burnup Model
Directory of Open Access Journals (Sweden)
Abdul Waris
2008-03-01
Full Text Available Plutonium (Pu and minor actinides (MA recycling in standard BWR with equilibrium burnup model has been studied. We considered the equilibrium burnup model as a simple time independent burnup method, which can manage all possible produced nuclides in any nuclear system. The equilibrium burnup code was bundled with a SRAC cell-calculation code to become a coupled cell-burnup calculation code system. The results show that the uranium enrichment for the criticality of the reactor, the amount of loaded fuel and the required natural uranium supply per year decrease for the Pu recycling and even much lower for the Pu & MA recycling case compared to those of the standard once-through BWR case. The neutron spectra become harder with the increasing number of recycled heavy nuclides in the reactor core. The total fissile rises from 4.77% of the total nuclides number density in the reactor core for the standard once-through BWR case to 6.64% and 6.72% for the Plutonium recycling case and the Pu & MA recycling case, respectively. The two later data may become the main basis why the required uranium enrichment declines and consequently diminishes the annual loaded fuel and the required natural uranium supply. All these facts demonstrate the advantage of plutonium and minor actinides recycling in BWR.
Simulation of high burn-up fuel cladding and its safety assessment under LOCA condition
International Nuclear Information System (INIS)
Park, Dong Jun; Won, Sung Bin; Choi, Byoung Kwon; Park, Jeong Yong; Koo, Yang Hyun
2011-01-01
Current LOCA safety criteria was established in the beginning of 1970s and based on the results obtained from non-irradiated Zircaloy-4 claddings. Because of major advantages in fuel-cycle costs, reactor operation, and waste management, the increase in fuel discharge burn-up is current worldwide trend in the nuclear industry. As the fuel burn-up increases, various phenomena unexpected have been reported due to changes in the condition of reactor operation and in-core environment. Since, it should be considered whether the current Loss-of-coolant accident (LOCA) criteria is suitable for high burn-up fuel cladding or not. In addition, many fuel vendors have recently developed new cladding alloys superior to Zircaloy-4 cladding. The performance of these advanced cladding alloys under LOCA, especially at high burn-up, is not well understood at this time. To better understand high burn-up effects and commercialize new cladding alloys, study of LOCA-related behavior of various types of high burn-up fuel cladding and their data base is essentially required. In this background, postulated LOCA test has been carried out with prehydrided Zircaloy-4 cladding as a surrogate for high burn-up cladding and the relevant results obtained are discussed
RIM formation and its effect on fission gas release in water reactor fuels at high burnup
International Nuclear Information System (INIS)
Viswanadham, C.S.; Sah, D.N.
2005-01-01
Water Reactor Fuels irradiated to burnup of more than 45 MWD/kg were found by several investigators to contain a porous region near the pellet periphery with submicroscopic grain sizes and a local burnup higher than the average fuel burnup. The resonance capture of neutrons in the rim region results in higher plutonium concentrations there, which results in more fissions, more accumulated burnup and more defects in the microstructure. The exact mechanism of the formation of the rim microstructure is still not very clear. This so-called RIM effect results in different properties in the rim region compared to the rest of the fuel pellet. The thermal conductivity of the rim region is significantly degraded, thereby resulting in increases in the fuel center temperature and higher fission gas release. The modelling of the effects of the rim on fuel properties and fission gas release so far has been largely empirical in nature. A threshold burnup is usually identified beyond which the RIM formation starts, and then a growth rate of the rim with burnup is assumed. This paper describes the current understanding of the Rim Effect and its consequences on fission gas release in water reactor fuels. The approach of modelling rim effect in computer code PROFESS is also described. (author)
Modification in the CITATION computer code: change of microscopic cross sections by zone
International Nuclear Information System (INIS)
Yamaguchi, M.; Kosaka, N.
1983-01-01
Some modifications done in the CITATION computer code are presented, aiming to calculate the accumulated burnup for each reactor zone in each step of burnup and allow changing the microscopic cross sections for each zone in accordance to the burnup accumulated after each step of burnup. Some input data were put in the computer code. The alterations were tested and the results were compared with and without modifications. (E.G.) [pt
TRANSFORM DOMAIN SLICE BASED DISTRIBUTED VIDEO CODING
Directory of Open Access Journals (Sweden)
SAMIR BELHOUARI
2011-10-01
Full Text Available Distributed video coding depends heavily on the virtual channel model. Due to the limitations of the side information estimation one stationary model does not properly describe the virtual channel. In this work the correlation noise is modelled per slice to obtain location-specific correlation noise model. The resulting delay from the lengthy Slepian-Wolf (SW codec input is also reduced by reducing the length of SW codec input. The proposed solution does not impose any extra complexity, it utilizes the existing resources. The results presented here support the proposed algorithm.
Model biases in high-burnup fast reactor simulations
International Nuclear Information System (INIS)
Touran, N.; Cheatham, J.; Petroski, R.
2012-01-01
A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k eff , power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)
International Nuclear Information System (INIS)
Mayson, R.T.H.; Gunston, K.J.
1999-01-01
Since 1991 BNFL has made a significant investment in the development of the burn-up credit method and the application to its operations. It has recently demonstrated that using this method for the THORP dissolvers, it is possible to justify operating safety with reduced neutron poison concentrations and this has now been submitted to the regulators. The continued challenges the criticality safety community is facing are to show that we are not reducing safety levels because we are using burn-up credit. The burn-up credit method that has been developed can be summarized as follows. It consists of performing reactivity calculations for irradiated fuel using compositions generated by and inventory prediction code, generally in order to determine the limiting burn-up required for that fuel in a particular environment. In addition, it has always been envisaged that a confirmatory measurement of burn-up would be required to be made prior to certain operations such as the sharing of fuel into a dissolver. The burn-up credit method therefore relies upon three key components of inventory prediction, reactivity calculation code and the quantification and verification of burn-up. (J.P.N.)
International Nuclear Information System (INIS)
Campolina, D. de A. M.; Lima, C.P.B.; Veloso, M.A.F.
2013-01-01
For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95. percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input. Particularly it was shown that during the burnup, the variances when considering all the parameters uncertainties is equivalent to the sum of variances if the parameter uncertainties are sampled separately
Recommendations for Addressing Axial Burnup in the PWR Burnup Credit Analyses
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.
2002-10-23
This report presents studies performed to support the development of a technically justifiable approach for addressing the axial-burnup distribution in pressurized-water reactor (PWR) burnup-credit criticality safety analyses. The effect of the axial-burnup distribution on reactivity and proposed approaches for addressing the axial-burnup distribution are briefly reviewed. A publicly available database of profiles is examined in detail to identify profiles that maximize the neutron multiplication factor, k{sub eff}, assess its adequacy for PWR burnup credit analyses, and investigate the existence of trends with fuel type and/or reactor operations. A statistical evaluation of the k{sub eff} values associated with the profiles in the axial-burnup-profile database was performed, and the most reactive (bounding) profiles were identified as statistical outliers. The impact of these bounding profiles on k{sub eff} is quantified for a high-density burnup credit cask. Analyses are also presented to quantify the potential reactivity consequence of loading assemblies with axial-burnup profiles that are not bounded by the database. The report concludes with a discussion on the issues for consideration and recommendations for addressing axial burnup in criticality safety analyses using burnup credit for dry cask storage and transportation.
International Nuclear Information System (INIS)
Sanders, T.L.; Lake, W.H.
1989-01-01
It is possible to develop an optimal strategy for implementing burnup credit in spent fuel transport casks. For transport, the relative risk is rapidly reduced if additional pre-transport controls such as a cavity dryness verifications are conducted prior to transport. Some other operational and design features that could be incorporated into a burnup credit cask strategy are listed. These examples represent many of the system features and alternatives already available for use in developing a broadly based criticality safety strategy for implementing burnup credit in the design and operation of spent fuel transport casks. 4 refs., 1 tab
Validation of SCALE-4 for burnup credit applications
International Nuclear Information System (INIS)
Bowman, S.M.; DeHart, M.D.; Parks, C.V.
1995-01-01
In the past, a criticality analysis of PWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. If credit is allowed for fuel burnup in the design of casks that are used in the transport of spent light water reactor fuel to a repository, the increase in payload can lead to a significant reduction in the cost of transport and a potential reduction in the risk to the public. A portion of the work has been performed at ORNL in support of the US DOE efforts to demonstrate a validation approach for criticality safety methods to be used in burnup credit cask design. To date, the SCALE code system developed at ORNL has been the primary computational tool used by DOE to investigate technical issues related to burnup credit. The ANSI/ANS-8.1 criticality safety standard requires validation and benchmarking of the calculational methods used in evaluating criticality safety limits for applications outside reactors by correlation against critical experiments that are applicable. Numerous critical experiments for fresh PWR-type fuel in storage and transport configurations exist and can be used as part of a validation database. However, there are no critical experiments with burned PWR-type fuel in storage and transport configurations. As an alternative, commercial reactors offer an excellent source of measured critical configurations. The results reported demonstrate the ability of the ORNL SCALE-4 methodology to predict a value of k eff very close to the known value of 1.0, both for fresh fuel criticals and for the more complex reactor criticals. Beyond these results, additional work in the determination of biases and uncertainties is necessary prior to use in burnup credit applications
Türkmen, Mehmet; Zabunoğlu, Okan H.
2012-10-01
Use of U and U-Th fuels in CANDU type of reactors (CANDU-6 and ACR-700) on the once-through nuclear fuel cycle is investigated. Based on the unit-cell approximation with the homogeneous-bundle/core model, utilizing the MONTEBURNS code, burnup computations are performed; discharge burnups are determined and expressed as functions of 235U and Th fractions, when applicable. Natural Uranium Requirement (and Saving) and Nuclear Resource Utilization are calculated for varying fuel compositions. Results are analyzed to observe the effects of 235U and Th fractions, thus to reach conclusions about use of Th in CANDU-6 and ACR-700 on the once-through cycle.
MCOR - Monte Carlo depletion code for reference LWR calculations
Energy Technology Data Exchange (ETDEWEB)
Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)
2011-04-15
Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan
2011-01-01
Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations
Energy Technology Data Exchange (ETDEWEB)
Noori-Kalkhoran, Omid; Ahangari, R. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research school; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Faculty of Engineering
2017-03-15
In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 reactor. Parallel processing of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of core at different days. PARCS V2.7?has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. A Parallel processing algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and Control rod worth have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective full Power Days (EFPDs) in some steps. Results have been compared with Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR) results. The results show great similarity and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.
Energy Technology Data Exchange (ETDEWEB)
Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-02-01
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
International Nuclear Information System (INIS)
Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya
2002-02-01
The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)
Burnup credit issues in transportation and storage
International Nuclear Information System (INIS)
Brady, M.C.; Sanders, T.L.; Seager, K.D.; Lake, W.H.
1992-01-01
Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the US experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed
Burnup credit issues in transportation and storage
International Nuclear Information System (INIS)
Brady, M.C.; Sanders, T.L.; Seager, K.D.; Lake, W.H.
1993-01-01
Reliance on the reduced reactivity of spent fuel for criticality control during transportation and storage is referred to as burnup credit. This concept has attracted international interest and is being actively pursued in the United States in the development of a new generation of transport casks. An overview of the U.S. experience in developing a methodology to implement burnup credit in an integrated approach to transport cask design is presented in this paper. Specifically, technical issues related to the analysis, validation and implementation of burnup credit are identified and discussed. (author)
Windows user-friendly code package development for operation of research reactors
International Nuclear Information System (INIS)
Hoang Anh Tuan
1998-01-01
The content of the project was to developed: 1. MS Windows interface to spectral codes like THERMOS, PEACO-COLLIS, GRACE and burn-up code. 2. MS Windows C-language burn-up diffusion hexagonal lattice code. The overall scope of the project was to develop a PC-based MS Windows code package for operation of Dalat research reactor. Various problems relating to neutronic physics like thermalization, resonance treatment, fast spectral treatment, change of isotopic concentration during burn-up time as well as burn-up distribution in the reactor core are considered in parallel to application of informatics technique. The developing process is a subject of the concept of user-friendly interface between end-users and the code package. High level input features through system of icon, menu, dialog box with regard to Common User Access (CUA) convention and sophisticated graphical output in MS Windows environment was used. The user-computer interface is also enhanced by using both keyboard and mouse, which creates a very natural manner for end-user. (author)
Context based Coding of Quantized Alpha Planes for Video Objects
DEFF Research Database (Denmark)
Aghito, Shankar Manuel; Forchhammer, Søren
2002-01-01
In object based video, each frame is a composition of objects that are coded separately. The composition is performed through the alpha plane that represents the transparency of the object. We present an alternative to MPEG-4 for coding of alpha planes that considers their specific properties. Co....... Comparisons in terms of rate and distortion are provided, showing that the proposed coding scheme for still alpha planes is better than the algorithms for I-frames used in MPEG-4.......In object based video, each frame is a composition of objects that are coded separately. The composition is performed through the alpha plane that represents the transparency of the object. We present an alternative to MPEG-4 for coding of alpha planes that considers their specific properties...
Design and analytic evaluation of a rim effect reduction type LWR fuel for extending burnup
International Nuclear Information System (INIS)
Matsumura, Tetsuo; Kameyama, Takanori; Kinoshita, Motoyasu
1991-01-01
We have designed a new concept fuel design 'Rim effect reduction type fuel' which has thin natural UO 2 layer on surface of a UO2 pellet. Our neutronic analyses with ANRB code show this fuel design can reduce rim effect (burnup at plelet rim) by about 30 GWd/t comparing a normal fuel. It is known that a high burnup fuel has different microstructure from as-fabricated one at fuel rim (which is called as rim region) due to rim effect. Therefore this fuel design can expect smaller rim region than a normal fuel. Our fuel performance analyses with EIMUS code show this fuel design can reduce fuel center temperature at high burnup if thermal conductivity of fuel pellet decreases with burnup in inverse proportion. However, this fuel design increases fuel center temperature at low and middle burnup than a normal fuel due to increase of thermal power density at pellet center. Additionally Irradiation experiment of this fuel design can be considered to offer important data which make clear the relation between rim effect and fuel performance. (author)
International Nuclear Information System (INIS)
Monteleone, S.
1998-03-01
This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) human reliability analysis and human performance evaluation; (2) technical issues related to rulemakings; (3) risk-informed, performance-based initiatives; and (4) high burn-up fuel research
Determination of burnup grade of fuel plates by gamma spectrometry
International Nuclear Information System (INIS)
Terremoto, Luis A.A.; Zeituni, Carlos A.; Perrotta, Jose A.
1999-01-01
This work describes absolute burnup measurements on spent MTR fuel elements by means of non-destructive gamma-ray spectroscopy which correlates activities of radioactive fission products with the fissioned mass of 235 U. Experiments based on such method were performed at the storage pool area of the IEA-R1 research reactor. The obtained results were compared with calculational ones based on neutronics. (author)
Burn-up TRIGA Mark II benchmark experiment
International Nuclear Information System (INIS)
Persic, A.; Ravnik, M.; Zagar, T.
1998-01-01
Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)
Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet
Amaya, Masaki; Nakamura, Jinichi; Nagase, Fumihisa; Fuketa, Toyoshi
2011-07-01
The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens' theory and reported thermal conductivities of unirradiated (U, Pu) O 2 and irradiated UO 2 pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.
Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.
2014-06-01
For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.
Energy Technology Data Exchange (ETDEWEB)
White, J.R.
1980-08-01
A generalized depletion perturbation formulation based on the quasi-static method for solving realistic multicycle reactor depletion problems is developed and implemented within the VENTURE/BURNER modular code system. The present development extends the original formulation derived by M.L. Williams to include nuclide discontinuities such as fuel shuffling and discharge. This theory is first described in detail with particular emphasis given to the similarity of the forward and adjoint quasi-static burnup equations. The specific algorithm and computational methods utilized to solve the adjoint problem within the newly developed DEPTH (Depletion Perturbation Theory) module are then briefly discussed. Finally, the main features and computational accuracy of this new method are illustrated through its application to several representative reactor depletion problems.
Development of high burnup nuclear fuel technology
International Nuclear Information System (INIS)
Suk, Ho Chun; Kang, Young Hwan; Jung, Jin Gone; Hwang, Won; Park, Zoo Hwan; Ryu, Woo Seog; Kim, Bong Goo; Kim, Il Gone
1987-04-01
The objectives of the project are mainly to develope both design and manufacturing technologies for 600 MWe-CANDU-PHWR-type high burnup nuclear fuel, and secondly to build up the foundation of PWR high burnup nuclear fuel technology on the basis of KAERI technology localized upon the standard 600 MWe-CANDU- PHWR nuclear fuel. So, as in the first stage, the goal of the program in the last one year was set up mainly to establish the concept of the nuclear fuel pellet design and manufacturing. The economic incentives for high burnup nuclear fuel technology development are improvement of fuel utilization, backend costs plant operation, etc. Forming the most important incentives of fuel cycle costs reduction and improvement of power operation, etc., the development of high burnup nuclear fuel technology and also the research on the incore fuel management and safety and technologies are necessary in this country
Progress in extending burnup of LWR fuel
International Nuclear Information System (INIS)
Freshley, M.D.
1986-04-01
Progress in increasing burnup of LWR fuel has been and continues to be made. Initially, LWR fuels were designed to achieve a burnup of about 33 GWd/tu for PWRs and about 28 GWd/tu for BWRs. Current warranties are about 36 GWd/tu and 31 GWd/tu for PWRs and BWRs, respectively. Present optimum extended burnups (batch average) are typically about 50 GWd/tu and 45 GWd/tu for PWRs and BWRs on 12-month cycles, respectively, and about 10% higher for 18-month cycles. Thus, the goal of research and development programs of the recent past has been to achieve these burnup levels reliably with maximum duty cycle flexibility
Burnup analysis of the power reactor, 2
International Nuclear Information System (INIS)
Ezure, Hideo
1975-09-01
In burnup analysis of JPDR-1 with FLARE, it was found to have problems. The program FLORA was developed for solution of the problems. By their bench mark tests FLORA was found to be useful for three-dimensional thermal-hydro-dynamic analysis of BWRs. It was applied to analysis of the burnup of JPDR-1. The input data and option of FLORA were corrected on referring to the results of gammer probe tests for JPDR-1. The void, source and burnup distributions were calculated each month during the operation. The burnup distribution in three assemblies revealed by a destructive test agrees better with that by FLORA than by FLARE. It was shown that the distortion of power distribution around the control rods by FLORA was smaller and closer to that by the gammer probe tests than by FLARE, and the connector of fuel assemblies and the plugs in the reflector had much influence on the power distribution. (auth.)
Energy Technology Data Exchange (ETDEWEB)
Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)
2016-08-15
Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.
Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up
DEFF Research Database (Denmark)
Carlsen, H.
1980-01-01
Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...
Unfolding code for neutron spectrometry based on neural nets technology
International Nuclear Information System (INIS)
Ortiz R, J. M.; Vega C, H. R.
2012-10-01
The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Neural Networks have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This unfolding code called Neutron Spectrometry and Dosimetry by means of Artificial Neural Networks was designed in a graphical interface under LabVIEW programming environment. The core of the code is an embedded neural network architecture, previously optimized by the R obust Design of Artificial Neural Networks Methodology . The main features of the code are: is easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 Lil(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, only seven rate counts measurement with a Bonner spheres spectrometer are required for simultaneously unfold the 60 energy bins of the neutron spectrum and to calculate 15 dosimetric quantities, for radiation protection porpoises. This code generates a full report in html format with all relevant information. (Author)
Burnup of fusion produced tritons and 3He ions in PLT and PDX
International Nuclear Information System (INIS)
Heidbrink, W.W.; Chrien, R.E.; Strachan, J.D.
1982-09-01
The d(d,p)t and d(d,n) 3 He fusion reactions produce 1 MeV tritons and 0.8 MeV 3 He ions which can subsequently undergo d(t,n)α and d( 3 He,p)α fusion reactions. The magnitude of this triton and 3 He ion burnup was measured on the PLT and PDX tokamaks by detection of the 14 MeV neutron and 15 MeV proton emission. In discharges with B/sub phi/ greater than or equal to 2 T, the measured 3 He burnup agrees well with predictions based on classical theories of ion confinement and slowing down, while the triton burnup was about four times lower than theoretically predicted. In discharges with weaker toroidal fields, the burnup of both ions fell by more than a factor of ten
Pore pressure and swelling in the rim region of LWR high burnup UO2 fuel
International Nuclear Information System (INIS)
Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Sohn, Dong-Seong
2001-01-01
Based on measured rim characteristics of LWR high burnup UO 2 fuel, the pressure of rim pores and the additional pellet swelling due to rim formation have been modeled. Using the assumption that the number of Xe atoms retained in the rim pores is the same as that which is depleted from the rim matrix, excessive pore pressure is derived as a function of temperature, pellet average burnup and pore radius. The rim pores with small radii are calculated to be highly overpressurized at high burnup. Comparison with experimental data shows that, while the pellet swelling obtained with best-estimate rim width is underpredicted, the one calculated with conservative rim width agrees well with the measured data for rim burnups between 50 and 65 GWd/tU. On the other hand, the measured swelling at 85 GWd/tU is about in-between the two calculations
Selection of Component Codes for Turbo Coding Based on Convergence properties
DEFF Research Database (Denmark)
Andersen, Jakob Dahl
1999-01-01
The turbo decoding is a sub-optimal decoding, i.e. it is not a maximum likelihood decoding. It is important to be aware of this fact when the parameters for the scheme are chosen. This goes especially for the selection of component codes, where the selection often has been based solely on the per......The turbo decoding is a sub-optimal decoding, i.e. it is not a maximum likelihood decoding. It is important to be aware of this fact when the parameters for the scheme are chosen. This goes especially for the selection of component codes, where the selection often has been based solely...... on the performance at high SNR's. We will show that it is important to base the choice on the performance at low SNR's, i.e. the convergence properties, as well. Further, the study of the performance with different component codes may lead to an understanding of the convergence process in the turbo codes....
Nuclide Importance and the Steady-State Burnup Equation
International Nuclear Information System (INIS)
Sekimoto, Hiroshi; Nemoto, Atsushi
2000-01-01
Conventional methods for evaluating some characteristic values of nuclides relating to burnup in a given neutron spectrum are reviewed in a mathematically systematic way, and a new method based on the importance theory is proposed. In this method, these characteristic values of a nuclide are equivalent to the importances of the nuclide. By solving the equation adjoint to the steady-state burnup equation with a properly chosen source term, the importances for all nuclides are obtained simultaneously.The fission number importance, net neutron importance, fission neutron importance, and absorbed neutron importance are evaluated and discussed. The net neutron importance is a measure directly estimating neutron economy, and it can be evaluated simply by calculating the fission neutron importance minus the absorbed neutron importance, where only the absorbed neutron importance depends on the fission product. The fission neutron importance and absorbed neutron importance are analyzed separately, and detailed discussions of the fission product effects are given for the absorbed neutron importance
Energy Technology Data Exchange (ETDEWEB)
Marshall, William BJ J [ORNL; Ade, Brian J [ORNL; Bowman, Stephen M [ORNL; Gauld, Ian C [ORNL; Ilas, Germina [ORNL; Mertyurek, Ugur [ORNL; Radulescu, Georgeta [ORNL
2015-01-01
Oak Ridge National Laboratory and the United States Nuclear Regulatory Commission have initiated a multiyear project to investigate application of burnup credit for boiling-water reactor (BWR) fuel in storage and transportation casks. This project includes two phases. The first phase (1) investigates applicability of peak reactivity methods currently used in spent fuel pools (SFPs) to storage and transportation systems and (2) evaluates validation of both reactivity (k_{eff}) calculations and burnup credit nuclide concentrations within these methods. The second phase will focus on extending burnup credit beyond peak reactivity. This paper documents the first phase, including an analysis of lattice design parameters and depletion effects, as well as both validation components. Initial efforts related to extended burnup credit are discussed in a companion paper. Peak reactivity analyses have been used in criticality analyses for licensing of BWR fuel in SFPs over the last 20 years. These analyses typically combine credit for the gadolinium burnable absorber present in the fuel with a modest amount of burnup credit. Gadolinium burnable absorbers are used in BWR assemblies to control core reactivity. The burnable absorber significantly reduces assembly reactivity at beginning of life, potentially leading to significant increases in assembly reactivity for burnups less than 15–20 GWd/MTU. The reactivity of each fuel lattice is dependent on gadolinium loading. The number of gadolinium-bearing fuel pins lowers initial lattice reactivity, but it has a small impact on the burnup and reactivity of the peak. The gadolinium concentration in each pin has a small impact on initial lattice reactivity but a significant effect on the reactivity of the peak and the burnup at which the peak occurs. The importance of the lattice parameters and depletion conditions are primarily determined by their impact on the gadolinium depletion. Criticality code validation for BWR burnup
Novel power saving architecture for FBG based OCDMA code generation
Osadola, Tolulope B.; Idris, Siti K.; Glesk, Ivan
2013-10-01
A novel architecture for generating incoherent, 2-dimensional wavelength hopping-time spreading optical CDMA codes is presented. The architecture is designed to facilitate the reuse of optical source signal that is unused after an OCDMA code has been generated using fiber Bragg grating based encoders. Effective utilization of available optical power is therefore achieved by cascading several OCDMA encoders thereby enabling 3dB savings in optical power.
Reconstruction of pin burnup characteristics from nodal calculations in hexagonal geometry
International Nuclear Information System (INIS)
Yang, W.S.; Finck, P.J.; Khalil, H.S.
1990-01-01
A reconstruction method has been developed for recovering pin burnup characteristics from fuel cycle calculations performed in hexagonal-z geometry using the nodal diffusion option of the DIF3D/REBUS-3 code system. Intra-modal distributions of group fluxes, nuclide densities, power density, burnup, and fluence are efficiently computed using polynomial shapes constrained to satisfy nodal information. The accuracy of the method has been tested by performing several numerical benchmark calculations and by comparing predicted local burnups to values measured for experimental assemblies in EBR-11. The results indicate that the reconstruction methods are quite accurate, yielding maximum errors in power and nuclide densities that are less than 2% for driver assemblies and typically less than 5% for blanket assemblies. 14 refs., 2 figs., 5 tabs
International Nuclear Information System (INIS)
Aggarwal, S.K.; Jaison, P.G.; Telmore, V.M.; Shah, R.V.; Sant, V.L.; Sasibhushan, K.; Parab, A.R.; Alamelu, D.
2010-03-01
Burn-up was determined experimentally using thermal ionization mass spectrometry for two samples from ThO 2 bundles irradiated in KAPS-2. This involved quantitative dissolution of the irradiated fuel samples followed by separation and determination of Th, U and a stable fission product burn-up monitor in the dissolved fuel solution. Stable fission product 148 Nd was used as a burn-up monitor for determining the number of fissions. Isotope Dilution-Thermal Ionisation Mass Spectrometry (ID-TIMS) using natural U, 229 Th and enriched 142 Nd as spikes was employed for the determination of U, Th and Nd, respectively. Atom % fission values of 1.25 ± 0.03 were obtained for both the samples. 232 U content in 233 U determined by alpha spectrometry was about 500 ppm and this was higher by a factor of 5 compared to the theoretically predicted value by ORIGEN-2 code. (author)
Effect of burnup history by moderator density on neutron-physical characteristics of WWER-1000 core
International Nuclear Information System (INIS)
Ovdiienko, I.; Kuchin, A.; Khalimonchuk, V.; Ieremenko, M.
2011-01-01
Results of assessment of burnup history effect by moderator density on neutron physical characteristics of WWER-1000 core are presented on example of stationary fuel loading with Russian design fuel assembly TWSA and AER benchmark for Khmelnitsky NPP that was proposed by TUV and SSTC NRC at nineteenth symposium. Assessment was performed by DYN3D code and cross section library sets generated by HELIOS code. Burnup history was taken into account by preparing of numerous cross section sets with different isotopic composition each of which was obtained by burning under different moderator density. For analysis of history effect 20 cross section sets were prepared for each fuel assembly corresponded to each of 20 axial layers of reactor core model for DYN3D code. Four fuel cycles were modeled both for stationary fuel loading with TWSA and AER benchmark for Khmelnitsky NPP to obtain steady value of error due to neglect of burnup history effect. Main attention of study was paid to effect of burnup history by moderator density to axial power distribution. Results of study for AER benchmark were compared with experimental values of axial power distribution for fuel assemblies of first, second, third and fourth year operation. (Authors)
Energy Technology Data Exchange (ETDEWEB)
Escriba, A.; Munoz-cobo, J. L.; Merino, R.; Melara, J.; Albendea, M.
2013-07-01
In the field of nuclear safety, the analysis of the stability of boiling water reactors is one of the biggest challenges for researchers. LAPUR code that allows to obtain the parameters of stability of the plant (Decay rate and frequency), being one of the programs used by IBERDROLA can be used for these calculations. With the collaboration of the research group TIN of the Polytechnic University of Valencia, a model of loss of conductivity of uranium has joined with the burned LAPUR. This update allows you to play the phenomenon in a more realistic way. This improvement has been validated and verified contrasting results with reference values.
Compositional Model Based Fisher Vector Coding for Image Classification.
Liu, Lingqiao; Wang, Peng; Shen, Chunhua; Wang, Lei; Hengel, Anton van den; Wang, Chao; Shen, Heng Tao
2017-12-01
Deriving from the gradient vector of a generative model of local features, Fisher vector coding (FVC) has been identified as an effective coding method for image classification. Most, if not all, FVC implementations employ the Gaussian mixture model (GMM) as the generative model for local features. However, the representative power of a GMM can be limited because it essentially assumes that local features can be characterized by a fixed number of feature prototypes, and the number of prototypes is usually small in FVC. To alleviate this limitation, in this work, we break the convention which assumes that a local feature is drawn from one of a few Gaussian distributions. Instead, we adopt a compositional mechanism which assumes that a local feature is drawn from a Gaussian distribution whose mean vector is composed as a linear combination of multiple key components, and the combination weight is a latent random variable. In doing so we greatly enhance the representative power of the generative model underlying FVC. To implement our idea, we design two particular generative models following this compositional approach. In our first model, the mean vector is sampled from the subspace spanned by a set of bases and the combination weight is drawn from a Laplace distribution. In our second model, we further assume that a local feature is composed of a discriminative part and a residual part. As a result, a local feature is generated by the linear combination of discriminative part bases and residual part bases. The decomposition of the discriminative and residual parts is achieved via the guidance of a pre-trained supervised coding method. By calculating the gradient vector of the proposed models, we derive two new Fisher vector coding strategies. The first is termed Sparse Coding-based Fisher Vector Coding (SCFVC) and can be used as the substitute of traditional GMM based FVC. The second is termed Hybrid Sparse Coding-based Fisher vector coding (HSCFVC) since it
Investigation of Burnup Credit Issues in BWR Fuel
International Nuclear Information System (INIS)
Broadhead, B.L.; DeHart, M.D.
1999-01-01
Calculations for long-term-disposal criticality safety of spent nuclear fuel requires the application of burnup credit because of the large mass of fissile material that will be present in the repository. Burnup credit calculations are based on depletion calculations that provide a conservative estimate of spent fuel contents, followed by criticality calculations to assess the value of keff for a spent fuel cask or a fuel configuration under a variety of probabilistically derived events. In order to ensure that the depletion calculation is conservative, it is necessary to both qualify and quantify assumptions that can be made in depletion models used to characterize spent fuel. Most effort in the United States this decade has focused on burnup issues related to pressurized-water reactors. However, requirements for the permanent disposal of fuel from boiling-water reactors has necessitated development of methods for prediction of spent fuel contents for such fuels. Concomitant with such analyses, validation is also necessary. This paper provides a summary of initial efforts at the Oak Ridge National Laboratory to better understand and validate spent fuel analyses for boiling-water-reactor fuel
A neutron spectrum unfolding code based on iterative procedures
International Nuclear Information System (INIS)
Ortiz R, J. M.; Vega C, H. R.
2012-10-01
In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)
WWER expert system for fuel failure analysis using the RTOP-CA code
International Nuclear Information System (INIS)
Likhanskii, V.; Evdokimov, I.; Sorokin, A.; Khromov, A.; Kanukova, V.; Apollonova, O.; Ugryumov, A.
2008-01-01
The computer expert system for fuel failure analysis of WWER during operation is presented. The diagnostics is based on the measurement of specific activity of reference nuclides in reactor primary coolant and application of a computer code for the data interpretation. The data analysis includes an evaluation of tramp uranium mass in reactor core, detection of failures by iodine and caesium spikes, evaluation of burnup of defective fuel. Evaluation of defective fuel burnup was carried out by applying the relation of caesium nuclides activity in spikes and relations of activities of gaseous fission products for steady state operational conditions. The method of burnup evaluation of defective fuel by use of fission gas activity is presented in details. The neural-network analysis is performed for determination of failed fuel rod number and defect size. Results of the expert system application are illustrated for several fuel campaigns on operating WWER NPPs. (authors)
International Nuclear Information System (INIS)
Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; Sterbentz, James W.
2014-01-01
Highlights: • The burnup of irradiated AGR-1 TRISO fuel was analyzed using gamma spectrometry. • The burnup of irradiated AGR-1 TRISO fuel was also analyzed using mass spectrometry. • Agreement between experimental results and neutron physics simulations was excellent. - Abstract: AGR-1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR-1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non-destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR-1 experiment. Two methods for evaluating burnup by gamma spectrometry were developed, one based on the Cs-137 activity and the other based on the ratio of Cs-134 and Cs-137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA (fissions per initial heavy metal atom) for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can be determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP-MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma
Color image coding based on recurrent iterated function systems
Kim, Kwon; Park, Rae-Hong
1998-02-01
This paper proposes a color image coding method based on recurrent iterated function systems (RIFSs). To encode a set of multispectral images, we apply a RIFS to multiset data consisting of three images. In the proposed method, the mappings not only between blocks within an individual spectral image but also between blocks of different spectral images are performed with contraction constraint. Simulation results show that the fractal coding based on the RIFS is useful for encoding concurrently a set of images by describing the similarity existing between a pair of images. In addition, the proposed color coding method can be applied to subband images and moving image sequences consisting of a set of images having similar gray patterns.
Production ready feature recognition based automatic group technology part coding
Energy Technology Data Exchange (ETDEWEB)
Ames, A.L.
1990-01-01
During the past four years, a feature recognition based expert system for automatically performing group technology part coding from solid model data has been under development. The system has become a production quality tool, capable of quickly the geometry based portions of a part code with no human intervention. It has been tested on over 200 solid models, half of which are models of production Sandia designs. Its performance rivals that of humans performing the same task, often surpassing them in speed and uniformity. The feature recognition capability developed for part coding is being extended to support other applications, such as manufacturability analysis, automatic decomposition (for finite element meshing and machining), and assembly planning. Initial surveys of these applications indicate that the current capability will provide a strong basis for other applications and that extensions toward more global geometric reasoning and tighter coupling with solid modeler functionality will be necessary.
Lossless Image Compression Based on Multiple-Tables Arithmetic Coding
Directory of Open Access Journals (Sweden)
Rung-Ching Chen
2009-01-01
Full Text Available This paper is intended to present a lossless image compression method based on multiple-tables arithmetic coding (MTAC method to encode a gray-level image f. First, the MTAC method employs a median edge detector (MED to reduce the entropy rate of f. The gray levels of two adjacent pixels in an image are usually similar. A base-switching transformation approach is then used to reduce the spatial redundancy of the image. The gray levels of some pixels in an image are more common than those of others. Finally, the arithmetic encoding method is applied to reduce the coding redundancy of the image. To promote high performance of the arithmetic encoding method, the MTAC method first classifies the data and then encodes each cluster of data using a distinct code table. The experimental results show that, in most cases, the MTAC method provides a higher efficiency in use of storage space than the lossless JPEG2000 does.
Development of burnup dependent fuel rod model in COBRA-TF
Yilmaz, Mine Ozdemir
The purpose of this research was to develop a burnup dependent fuel thermal conductivity model within Pennsylvania State University, Reactor Dynamics and Fuel Management Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account first, the degradation of fuel thermal conductivity with high burnup; and second, the fuel thermal conductivity dependence on the Gadolinium content for both UO2 and MOX fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 fuel rods and Duriez/Modified NFI Model for MOX fuel rods were incorporated into CTF and fuel centerline predictions were compared against Halden experimental test data and FRAPCON-3.4 predictions to validate the burnup dependent fuel thermal conductivity model in CTF. Experimental test cases from Halden reactor fuel rods for UO2 fuel rods at Beginning of Life (BOL), through lifetime without Gd2O3 and through lifetime with Gd 2O3 and a MOX fuel rod were simulated with CTF. Since test fuel rod and FRAPCON-3.4 results were based on single rod measurements, CTF was run for a single fuel rod surrounded with a single channel configuration. Input decks for CTF were developed for one fuel rod located at the center of a subchannel (rod-centered subchannel approach). Fuel centerline temperatures predicted by CTF were compared against the measurements from Halden experimental test data and the predictions from FRAPCON-3.4. After implementing the new fuel thermal conductivity model in CTF and validating the model with experimental data, CTF model was applied to steady state and transient calculations. 4x4 PWR fuel bundle configuration from Purdue MOX benchmark was used to apply the new model for steady state and transient calculations. First, one of each high burnup UO2 and MOX fuel rods from 4x4 matrix were selected to carry out single fuel rod calculations and fuel centerline temperatures predicted by CTF/TORT-TD were compared against CTF /TORT-TD /FRAPTRAN
Parallel GPU implementation of PWR reactor burnup
International Nuclear Information System (INIS)
Heimlich, A.; Silva, F.C.; Martinez, A.S.
2016-01-01
Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.
Status of burnup credit implementation in Switzerland
International Nuclear Information System (INIS)
Grimm, P.
1998-01-01
Burnup credit is currently not used for the storage of spent fuel in the reactor pools in Switzerland, but credit is taken for integral burnable absorbers. Interest exists to take credit of burnup in future for the storage in a central away-from-reactor facility presently under construction. For spent fuel transports to foreign reprocessing plants the regulations of the receiving countries must be applied in addition to the Swiss licensing criteria. Burnup credit has been applied by one Swiss PWR utility for such transports in a consistent manner with the licensing practice in the receiving countries. Measurements of reactivity worths of small spent fuel samples in a Swiss zero-power research reactor are at an early stage of planning. (author)
Development of the point-depletion code DEPTH
Energy Technology Data Exchange (ETDEWEB)
She, Ding [Department of Engineering Physics, Tsinghua University, Beijing 100084 (China); Wang, Kan, E-mail: wangkan@mail.tsinghua.edu.cn [Department of Engineering Physics, Tsinghua University, Beijing 100084 (China); Yu, Ganglin [Department of Engineering Physics, Tsinghua University, Beijing 100084 (China)
2013-05-15
Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code.
ISOTOPIC MODEL FOR COMMERCIAL SNF BURNUP CREDIT
International Nuclear Information System (INIS)
A.H. Wells
2004-01-01
The purpose of this report is to demonstrate a process for selecting bounding depletion parameters, show that they are conservative for pressurized water reactor (PWR) and boiling water reactor (BWR) spent nuclear fuel (SNF), and establish the range of burnup for which the parameters are conservative. The general range of applicability is for commercial light water reactor (LWR) SNF with initial enrichments between 2.0 and 5.0 weight percent 235 U and burnups between 10 and 50 gigawatt-day per metric ton of uranium (GWd/MTU)
FURNACE-J, 2-D Diffusion Burnup for Fast Reactors from JAERI Fast-Set
International Nuclear Information System (INIS)
Ikawa, Koji
1984-01-01
1 - Nature of physical problem solved: FURNACEJ is a two-dimensional diffusion-burnup code for use in the detailed burnup analysis of fast reactors. The code is an extension code of the FURNACE. There exists no essential difference between FURNACE and FURNACEJ. However, the latter can deal with JAERI-Fast-Set as its cross section library, while the former is designed to use ABBN set. Additionally, in FURNACEJ, group-dependent and -independent transverse buckling of each region can be computed and punched on cards, if desired. This is prepared for users so as to use them as input data for detailed two-dimensional x-y calculations. 2 - Restrictions on the complexity of the problem: Only r-z geometry is available
Burnup verification tests with the FORK measurement system-implementation for burnup credit
International Nuclear Information System (INIS)
Ewing, R.I.
1994-01-01
Verification measurements may be used to help ensure nuclear criticality safety when burnup credit is applied to spent fuel transport and storage systems. The FORK system measures the passive neutron and gamma-ray emission from spent fuel assemblies while in the storage pool. It was designed at Los Alamos National Laboratory for the International Atomic Energy Agency safeguards program and is well suited to verify burnup and cooling time records at commercial Pressurized Water Reactor (PWR) sites. This report deals with the application of the FORK system to burnup credit operations
Image coding based on maximum entropy partitioning for identifying ...
Indian Academy of Sciences (India)
A new coding scheme based on maximum entropy partitioning is proposed in our work, particularly to identify the improbable intensities related to different emotions. The improbable intensities when used as a mask decode the facial expression correctly, providing an effectiveplatform for future emotion categorization ...
Rhodes, Gillian; Burton, Nichola; Jeffery, Linda; Read, Ainsley; Taylor, Libby; Ewing, Louise
2018-05-01
Individuals with autism spectrum disorder (ASD) can have difficulty recognizing emotional expressions. Here, we asked whether the underlying perceptual coding of expression is disrupted. Typical individuals code expression relative to a perceptual (average) norm that is continuously updated by experience. This adaptability of face-coding mechanisms has been linked to performance on various face tasks. We used an adaptation aftereffect paradigm to characterize expression coding in children and adolescents with autism. We asked whether face expression coding is less adaptable in autism and whether there is any fundamental disruption of norm-based coding. If expression coding is norm-based, then the face aftereffects should increase with adaptor expression strength (distance from the average expression). We observed this pattern in both autistic and typically developing participants, suggesting that norm-based coding is fundamentally intact in autism. Critically, however, expression aftereffects were reduced in the autism group, indicating that expression-coding mechanisms are less readily tuned by experience. Reduced adaptability has also been reported for coding of face identity and gaze direction. Thus, there appears to be a pervasive lack of adaptability in face-coding mechanisms in autism, which could contribute to face processing and broader social difficulties in the disorder. © 2017 The British Psychological Society.
File compression and encryption based on LLS and arithmetic coding
Yu, Changzhi; Li, Hengjian; Wang, Xiyu
2018-03-01
e propose a file compression model based on arithmetic coding. Firstly, the original symbols, to be encoded, are input to the encoder one by one, we produce a set of chaotic sequences by using the Logistic and sine chaos system(LLS), and the values of this chaotic sequences are randomly modified the Upper and lower limits of current symbols probability. In order to achieve the purpose of encryption, we modify the upper and lower limits of all character probabilities when encoding each symbols. Experimental results show that the proposed model can achieve the purpose of data encryption while achieving almost the same compression efficiency as the arithmetic coding.
FARO base case post-test analysis by COMETA code
Energy Technology Data Exchange (ETDEWEB)
Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)
1995-09-01
The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.
Energy Technology Data Exchange (ETDEWEB)
Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik
1998-03-01
EXOTIC-7 was the first in-pile test with {sup 6}Li-enriched (50%) lithium orthosilicate (Li{sub 4}SiO{sub 4}) pebbles and with DEMO representative Li-burnup. Post irradiation examinations of the Li{sub 4}SiO{sub 4} have been performed at the Forschungszentrum Karlsruhe (FZK), mainly to investigate the tritium release kinetics as well as the effect of Li-burnup and/or contact with beryllium during irradiation. The release rate of Li{sub 4}SiO{sub 4} from pure Li{sub 4}SiO{sub 4} bed of capsule 28.1-1 is characterized by a broad main peak at about 400degC and by a smaller peak at about 800degC, and that from the mixed beds of capsule 28.2 and 26.2-1 shows again these two peaks, but most of the tritium is now released from the 800degC peak. This shift of release from low to high temperature may be due to the higher Li-burnup and/or due to contact with Be during irradiation. Due to the very difficult interpretation of the in-situ tritium release data, residence times have been estimated on the basis of the out-of-pile tests. The residence time for Li{sub 4}SiO{sub 4} from caps. 28.1-1 irradiated at 10% Li-burnup agrees quite well with that of the same material irradiated at Li-burnup lower than 3% in the EXOTIC-6 experiment. In spite of the observed shift in the release peaks from low to high temperature, also the residence time for Li{sub 4}SiO{sub 4} from caps. 26.2-1 irradiated at 13% Li-burnup agrees quite well with the data from EXOTIC-6 experiment. On the other hand, the residence time for Li{sub 4}SiO{sub 4} from caps. 28.2 (Li-burnup 18%) is about a factor 1.7-3.8 higher than that for caps. 26.2-1. Based on these data on can conclude that up to 13% Li-burnup neither the contact with beryllium nor the Li-burnup have a detrimental effect on the tritium release of Li{sub 4}SiO{sub 4} pebbles, but at 18% Li-burnup the residence time is increased by about a factor three. (J.P.N.)
Investigation of Burnup Credit Modeling Issues Associated with BWR Fuel
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.
2000-10-12
Although significant effort has been dedicated to the study of burnup-credit issues over the past decade, U.S. studies to-date have primarily focused on spent pressurized-water-reactor (PWR) fuel. The current licensing approach taken by the U.S. Department of Energy for burnup credit in transportation seeks approval for PWR fuel only. Burnup credit for boiling-water-reactor (BWR) fuel has not yet been formally sought. Burnup credit for PWR fuel was pursued first because: (1) nearly two-thirds (by mass) of the total discharged commercial spent fuel in the United States is PWR fuel, (2) it can substantially increase the fuel assembly capacity with respect to current designs for PWR storage and transportation casks, and (3) fuel depletion in PWRs is generally less complicated than fuel depletion in BWRs. However, due to international needs, the increased enrichment of modern BWR fuels, and criticality safety issues related to permanent disposal within the United States, more attention has recently focused on spent BWR fuel. Specifically, credit for fuel burnup in the criticality safety analysis for long-term disposal of spent nuclear fuel enables improved design efficiency, which, due to the large mass of fissile material that will be stored in the repository, can have substantial financial benefits. For criticality safety purposes, current PWR storage and transportation canister designs employ flux traps between assemblies. Credit for fuel burnup will eliminate the need for these flux traps, and thus, significantly increase the PWR assembly capacity (for a fixed canister volume). Increases in assembly capacity of approximately one-third are expected. In contrast, current BWR canister designs do not require flux traps for criticality safety, and thus, are already at their maximum capacity in terms of physical storage. Therefore, benefits associated with burnup credit for BWR storage and transportation casks may be limited to increasing the enrichment capacity and
Hybrid Video Coding Based on Bidimensional Matching Pursuit
Directory of Open Access Journals (Sweden)
Lorenzo Granai
2004-12-01
Full Text Available Hybrid video coding combines together two stages: first, motion estimation and compensation predict each frame from the neighboring frames, then the prediction error is coded, reducing the correlation in the spatial domain. In this work, we focus on the latter stage, presenting a scheme that profits from some of the features introduced by the standard H.264/AVC for motion estimation and replaces the transform in the spatial domain. The prediction error is so coded using the matching pursuit algorithm which decomposes the signal over an appositely designed bidimensional, anisotropic, redundant dictionary. Comparisons are made among the proposed technique, H.264, and a DCT-based coding scheme. Moreover, we introduce fast techniques for atom selection, which exploit the spatial localization of the atoms. An adaptive coding scheme aimed at optimizing the resource allocation is also presented, together with a rate-distortion study for the matching pursuit algorithm. Results show that the proposed scheme outperforms the standard DCT, especially at very low bit rates.
Impact on burnup performance of coated particle fuel design in pebble bed reactor with ROX fuel
International Nuclear Information System (INIS)
Ho, Hai Quan; Obara, Toru
2015-01-01
The pebble bed reactor (PBR), a kind of high-temperature gas-cooled reactor (HTGR), is expected to be among the next generation of nuclear reactors as it has excellent passive safety features, as well as online refueling and high thermal efficiency. Rock-like oxide (ROX) fuel has been studied at the Japan Atomic Energy Agency (JAEA) as a new once-through type fuel concept. Rock-like oxide used as fuel in a PBR can be expected to achieve high burnup and improve chemical stabilities. In the once-through fuel concept, the main challenge is to achieve as high a burnup as possible without failure of the spent fuel. The purpose of this study was to investigate the impact on burnup performance of different coated fuel particle (CFP) designs in a PBR with ROX fuel. In the study, the AGR-1 Coated Particle design and Deep-Burn Coated Particle design were used to make the burnup performance comparison. Criticality and core burnup calculations were performed by MCPBR code using the JENDL-4.0 library. Results at equilibrium showed that the two reactors utilizing AGR-1 Coated Particle and Deep-Burn Coated Particle designs could be critical with almost the same multiplication factor k eff . However, the power peaking factor and maximum power per fuel ball in the AGR-1 coated particle design was lower than that of Deep-Burn coated particle design. The AGR-1 design also showed an advantage in fissions per initial fissile atoms (FIFA); the AGR-1 coated particle design produced a higher FIFA than the Deep-Burn coated particle design. These results suggest that the difference in coated particle fuel design can have an effect on the burnup performance in ROX fuel. (author)
Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base
Savage, B.; Snoke, J. A.
2017-12-01
The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years
GRADSPMHD: A parallel MHD code based on the SPH formalism
Vanaverbeke, S.; Keppens, R.; Poedts, S.
2014-03-01
We present GRADSPMHD, a completely Lagrangian parallel magnetohydrodynamics code based on the SPH formalism. The implementation of the equations of SPMHD in the “GRAD-h” formalism assembles known results, including the derivation of the discretized MHD equations from a variational principle, the inclusion of time-dependent artificial viscosity, resistivity and conductivity terms, as well as the inclusion of a mixed hyperbolic/parabolic correction scheme for satisfying the ∇ṡB→ constraint on the magnetic field. The code uses a tree-based formalism for neighbor finding and can optionally use the tree code for computing the self-gravity of the plasma. The structure of the code closely follows the framework of our parallel GRADSPH FORTRAN 90 code which we added previously to the CPC program library. We demonstrate the capabilities of GRADSPMHD by running 1, 2, and 3 dimensional standard benchmark tests and we find good agreement with previous work done by other researchers. The code is also applied to the problem of simulating the magnetorotational instability in 2.5D shearing box tests as well as in global simulations of magnetized accretion disks. We find good agreement with available results on this subject in the literature. Finally, we discuss the performance of the code on a parallel supercomputer with distributed memory architecture. Catalogue identifier: AERP_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AERP_v1_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 620503 No. of bytes in distributed program, including test data, etc.: 19837671 Distribution format: tar.gz Programming language: FORTRAN 90/MPI. Computer: HPC cluster. Operating system: Unix. Has the code been vectorized or parallelized?: Yes, parallelized using MPI. RAM: ˜30 MB for a
International Nuclear Information System (INIS)
Simanullang, Irwan Liapto; Obara, Toru
2017-01-01
Highlights: • Burnup performance using ROX fuel in PBR with accumulative fuel loading scheme was analyzed. • Initial excess reactivity was suppressed by reducing 235 U enrichment in the startup condition. • Negative temperature coefficient was achieved in all condition of PBR with accumulative fuel loading scheme using ROX fuel. • Core lifetime of PBR with accumulative fuel loading scheme using ROX fuel was shorter than with UO 2 fuel. • In PBR with accumulative fuel loading scheme using ROX fuel, achieved discharged burnup can be as high as that for UO 2 fuel. - Abstract: The Japan Atomic Energy Agency (JAEA) has proposed rock-like oxide (ROX) fuel as a new, once-through type fuel concept. Here, burnup performance using ROX fuel was simulated in a pebble bed reactor with an accumulative fuel loading scheme. The MVP-BURN code was used to simulate the burnup calculation. Fuel of 5 g-HM/pebble with 20% 235 U enrichment was selected as the optimum composition. Discharged burnup could reach up to 218 GWd/t, with a core lifetime of about 8.4 years. However, high excess reactivity occurred in the initial condition. Initial fuel enrichment was therefore reduced from 20% to 4.65% to counter the initial excess reactivity. The operation period was reduced by the decrease of initial fuel enrichment, but the maximum discharged burnup was 198 GWd/t. Burnup performance of ROX fuel in this reactor concept was compared with that of UO 2 fuel obtained previously. Discharged burnup for ROX fuel in the PBR with an accumulative fuel loading scheme was as high as UO 2 fuel. Maximum power density could be lowered by introducing ROX fuel compared to UO 2 fuel. However, PBR core lifetime was shorter with ROX fuel than with UO 2 fuel. A negative temperature coefficient was achieved for both UO 2 and ROX fuels throughout the operation period.
VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup
International Nuclear Information System (INIS)
2002-01-01
1 - Description of program or function: The VENTURE program solves the usual neutronics eigenvalue, adjoint, fixed source, and criticality search problems. It treats up to three dimensions, maps power density, and does first-order perturbation analysis at the macroscopic cross section level. The BURNER code solves the nuclide chain equations to estimate the nuclide concentrations and burnup at the end of an exposure time or after a shutdown period. This package is based on the CCC-459/BOLD VENTURE IV code system developed at Oak Ridge National Laboratory. In January 1989 the University of Cincinnati contributed the first VENTURE-PC package to RSICC's collection. It was a subset of the mainframe version consisting of the VENTURE and BURNER modules plus several processing modules. VENTURE-PC was distributed as CCC-459 until July 1997 when a new version (with updated source code compatible with newer FORTRAN-77 compilers, some revisions, and extensions to solve much larger problems) was contributed by Argonne National Laboratory. The principle code modules included in the VENTURE-PC system are: VENTURE: Multigroup neutronics finite-difference diffusion theory. BURNER: Depletion calculation for reactor core analysis. Other modules within VENTURE-PC are: DVENTR: Venture input processor; DCRSPR: Neutron cross section processor; DUTLIN: Control file (CNTRL) input processor; DCMACR: Citation format cross section input processor; CRXSPR: Cross section processor; DENMAN: Fuel repositioning module. In August of 1999, Argonne again contributed an updated version of the code which overcomes problem size constraints caused by binary record length limits inherent to the Fortran 90 compiler. The need for long records is detected and avoided by sub-blocking them. Also, the latest Fortran 95 compiler offers substantial speed gains on the newest processors. The source code is updated to be compatible with either Fortran 90 or Fortran 95. In August 2002, the package was updated with
WWER-1000 Burnup Credit Benchmark (CB5)
International Nuclear Information System (INIS)
Manolova, M.A.
2002-01-01
In the paper the specification of WWER-1000 Burnup Credit Benchmark first phase (depletion calculations), given. The second phase - criticality calculations for the WWER-1000 fuel pin cell, will be given after the evaluation of the results, obtained at the first phase. The proposed benchmark is a continuation of the WWER benchmark activities in this field (Author)
Burnup credit effect on proposed cask payloads
International Nuclear Information System (INIS)
Hall, I.K.
1989-01-01
The purpose of the Cask Systems Development Program (CSDP) is to develop a variety of cask systems which will allow safe and economical movement of commercial spent nuclear fuel and high-level waste from the generator to the Federal repository or Monitored Retrievable Storage (MRS) facility. Program schedule objectives for the initial phase of the CSDP include the development of certified spent fuel cask systems by 1995 to support Office of Civilian Radioactive Waste Management shipments from the utilities beginning in the late 1990s. Forty-nine proposals for developing a family of spent fuel casks were received and comparisons made. General conclusions that can be drawn from the comparisons are that (1) the new generation of casks will have substantially increased payloads in comparison to current casks, and (2) an even greater payload increase may be achievable with burnup credit. The ranges in the payload estimates do not allow a precise separation of the payload increase attributable to the proposed allowance of fuel burnup credit, as compared wilt the no-burnup-credit case. The beneficial effects of cask payload increases on overall costs and risks of transporting spent fuel are significant; therefore further work aimed toward taking advantage of burnup credit is warranted
Transient behaviour of high burnup fuel
International Nuclear Information System (INIS)
1996-01-01
The main subjects of the meeting were the discussion of regulatory background, integral tests and analysis, plant calculations, separate-effect test and analysis, concerning high burnup phenomena during RIA accidents in reactors, especially LWR, BWR and PWR type reactors. 32 papers were abstracted and indexed individually for the INIS database. (R.P.)
Energy Technology Data Exchange (ETDEWEB)
Kramer, K J; Latkowski, J F; Abbott, R P; Boyd, J K; Powers, J J; Seifried, J E
2008-10-24
Lawrence Livermore National Laboratory is currently developing a hybrid fusion-fission nuclear energy system, called LIFE, to generate power and burn nuclear waste. We utilize inertial confinement fusion to drive a subcritical fission blanket surrounding the fusion chamber. It is composed of TRISO-based fuel cooled by the molten salt flibe. Low-yield (37.5 MJ) targets and a repetition rate of 13.3 Hz produce a 500 MW fusion source that is coupled to the subcritical blanket, which provides an additional gain of 4-8, depending on the fuel. In the present work, we describe the neutron transport and nuclear burnup analysis. We utilize standard analysis tools including, the Monte Carlo N-Particle (MCNP) transport code, ORIGEN2 and Monteburns to perform the nuclear design. These analyses focus primarily on a fuel composed of depleted uranium not requiring chemical reprocessing or enrichment. However, other fuels such as weapons grade plutonium and highly-enriched uranium are also under consideration. In addition, we have developed a methodology using {sup 6}Li as a burnable poison to replace the tritium burned in the fusion targets and to maintain constant power over the lifetime of the engine. The results from depleted uranium analyses suggest up to 99% burnup of actinides is attainable while maintaining full power at 2GW for more than five decades.
High burnup rim project. (IV) Threshold burnup of rim structure formation
International Nuclear Information System (INIS)
Sonoda, T.; Matzke, Hj.; Kinoshita, M.
1999-01-01
High burnup extension of LWR fuel is progressing to reduce the amount of total process flow in the nuclear fuel cycle and eventually to reduce the fuel cycle costs. As a result, the local burnup is now exceeding the anticipated range of the UO 2 fuel that was investigated in the great time of the 1960's. A 'new phenomenon', a crystallographic re-structuring, is commonly observed at the rim area of high burnup fuel pellets in LWRs, and also in FBRs to some extent. The objectives of the High Burnup Rim Project (HBRP) are to identify the conditions of the rim structure formation as functions of burnup and temperature, and to investigate physical and chemical properties of fuels following this re-structuring. After the irradiation, the rods were transported to the Institute for Transuranium Elements (ITU), Karlsruhe for Post Irradiation Examinations (PIE). This report shows recent progress of PIE, and discusses threshold burnup and temperature of the rim structure formation (author) (ml)
Directory of Open Access Journals (Sweden)
Ramzan Naeem
2007-01-01
Full Text Available An efficient approach for joint source and channel coding is presented. The proposed approach exploits the joint optimization of a wavelet-based scalable video coding framework and a forward error correction method based on turbo codes. The scheme minimizes the reconstructed video distortion at the decoder subject to a constraint on the overall transmission bitrate budget. The minimization is achieved by exploiting the source rate distortion characteristics and the statistics of the available codes. Here, the critical problem of estimating the bit error rate probability in error-prone applications is discussed. Aiming at improving the overall performance of the underlying joint source-channel coding, the combination of the packet size, interleaver, and channel coding rate is optimized using Lagrangian optimization. Experimental results show that the proposed approach outperforms conventional forward error correction techniques at all bit error rates. It also significantly improves the performance of end-to-end scalable video transmission at all channel bit rates.
Directory of Open Access Journals (Sweden)
Naeem Ramzan
2007-03-01
Full Text Available An efficient approach for joint source and channel coding is presented. The proposed approach exploits the joint optimization of a wavelet-based scalable video coding framework and a forward error correction method based on turbo codes. The scheme minimizes the reconstructed video distortion at the decoder subject to a constraint on the overall transmission bitrate budget. The minimization is achieved by exploiting the source rate distortion characteristics and the statistics of the available codes. Here, the critical problem of estimating the bit error rate probability in error-prone applications is discussed. Aiming at improving the overall performance of the underlying joint source-channel coding, the combination of the packet size, interleaver, and channel coding rate is optimized using Lagrangian optimization. Experimental results show that the proposed approach outperforms conventional forward error correction techniques at all bit error rates. It also significantly improves the performance of end-to-end scalable video transmission at all channel bit rates.
The PHREEQE Geochemical equilibrium code data base and calculations
International Nuclear Information System (INIS)
Andersoon, K.
1987-01-01
Compilation of a thermodynamic data base for actinides and fission products for use with PHREEQE has begun and a preliminary set of actinide data has been tested for the PHREEQE code in a version run on an IBM XT computer. The work until now has shown that the PHREEQE code mostly gives satisfying results for specification of actinides in natural water environment. For U and Np under oxidizing conditions, however, the code has difficulties to converge with pH and Eh conserved when a solubility limit is applied. For further calculations of actinide and fission product specification and solubility in a waste repository and in the surrounding geosphere, more data are needed. It is necessary to evaluate the influence of the large uncertainties of some data. A quality assurance and a check on the consistency of the data base is also needed. Further work with data bases should include: an extension to fission products, an extension to engineering materials, an extension to other ligands than hydroxide and carbonate, inclusion of more mineral phases, inclusion of enthalpy data, a control of primary references in order to decide if values from different compilations are taken from the same primary reference and contacts and discussions with other groups, working with actinide data bases, e.g. at the OECD/NEA and at the IAEA. (author)
Huffman-based code compression techniques for embedded processors
Bonny, Mohamed Talal
2010-09-01
The size of embedded software is increasing at a rapid pace. It is often challenging and time consuming to fit an amount of required software functionality within a given hardware resource budget. Code compression is a means to alleviate the problem by providing substantial savings in terms of code size. In this article we introduce a novel and efficient hardware-supported compression technique that is based on Huffman Coding. Our technique reduces the size of the generated decoding table, which takes a large portion of the memory. It combines our previous techniques, Instruction Splitting Technique and Instruction Re-encoding Technique into new one called Combined Compression Technique to improve the final compression ratio by taking advantage of both previous techniques. The instruction Splitting Technique is instruction set architecture (ISA)-independent. It splits the instructions into portions of varying size (called patterns) before Huffman coding is applied. This technique improves the final compression ratio by more than 20% compared to other known schemes based on Huffman Coding. The average compression ratios achieved using this technique are 48% and 50% for ARM and MIPS, respectively. The Instruction Re-encoding Technique is ISA-dependent. It investigates the benefits of reencoding unused bits (we call them reencodable bits) in the instruction format for a specific application to improve the compression ratio. Reencoding those bits can reduce the size of decoding tables by up to 40%. Using this technique, we improve the final compression ratios in comparison to the first technique to 46% and 45% for ARM and MIPS, respectively (including all overhead that incurs). The Combined Compression Technique improves the compression ratio to 45% and 42% for ARM and MIPS, respectively. In our compression technique, we have conducted evaluations using a representative set of applications and we have applied each technique to two major embedded processor architectures
Edge-preserving Intra Depth Coding based on Context-coding and H.264/AVC
DEFF Research Database (Denmark)
Zamarin, Marco; Salmistraro, Matteo; Forchhammer, Søren
2013-01-01
Depth map coding plays a crucial role in 3D Video communication systems based on the “Multi-view Video plus Depth” representation as view synthesis performance is strongly affected by the accuracy of depth information, especially at edges in the depth map image. In this paper an efficient algorithm...... for edge-preserving intra depth compression based on H.264/AVC is presented. The proposed method introduces a new Intra mode specifically targeted to depth macroblocks with arbitrarily shaped edges, which are typically not efficiently represented by DCT. Edge macroblocks are partitioned into two regions...... performance. Experiments show that the proposed Intra mode can improve view synthesis performance: average Bjøntegaard bit rate savings of 25% have been reported over a standard H.264/AVC Intra coder....
Trellis and turbo coding iterative and graph-based error control coding
Schlegel, Christian B
2015-01-01
This new edition has been extensively revised to reflect the progress in error control coding over the past few years. Over 60% of the material has been completely reworked, and 30% of the material is original. Convolutional, turbo, and low density parity-check (LDPC) coding and polar codes in a unified framework. Advanced research-related developments such as spatial coupling. A focus on algorithmic and implementation aspects of error control coding.
Fuel performance analysis code 'FAIR'
International Nuclear Information System (INIS)
Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.
1994-01-01
For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs
Automatic Structure-Based Code Generation from Coloured Petri Nets
DEFF Research Database (Denmark)
Kristensen, Lars Michael; Westergaard, Michael
2010-01-01
Automatic code generation based on Coloured Petri Net (CPN) models is challenging because CPNs allow for the construction of abstract models that intermix control flow and data processing, making translation into conventional programming constructs difficult. We introduce Process-Partitioned CPNs....... The viability of our approach is demonstrated by applying it to automatically generate an Erlang implementation of the Dynamic MANET On-demand (DYMO) routing protocol specified by the Internet Engineering Task Force (IETF)....
Determination of IRT-2M fuel burnup by gamma spectrometry.
Koleška, Michal; Viererbl, Ladislav; Marek, Milan; Ernest, Jaroslav; Šunka, Michal; Vinš, Miroslav
2016-01-01
A spectrometric system was developed for evaluating spent fuel in the LVR-15 research reactor, which employs highly enriched (36%) IRT-2M-type fuel. Such system allows the measurement of detailed fission product profiles. Within these measurements, nuclides such as (137)Cs, (134)Cs, (144)Ce, (106)Ru and (154)Eu may be detected in fuel assemblies with different cooling times varying between 1.67 and 7.53 years. Burnup calculations using the MCNPX Monte Carlo code data showed good agreement with measurements, though some discrepancies were observed in certain regions. These discrepancies are attributed to the evaluation of irradiation history, reactor regulation pattern and buildup schemes. Copyright © 2015 Elsevier Ltd. All rights reserved.
Simulation of triton burn-up in JET plasmas
Energy Technology Data Exchange (ETDEWEB)
Loughlin, M.J.; Balet, B.; Jarvis, O.N.; Stubberfield, P.M. [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking
1994-07-01
This paper presents the first triton burn-up calculations for JET plasmas using the transport code TRANSP. Four hot ion H-mode deuterium plasmas are studied. For these discharges, the 2.5 MeV emission rises rapidly and then collapses abruptly. This phenomenon is not fully understood but in each case the collapse phase is associated with a large impurity influx known as the ``carbon bloom``. The peak 14 MeV emission occurs at this time, somewhat later than that of the 2.5 MeV neutron peak. The present results give a clear indication that there are no significant departures from classical slowing down and spatial diffusion for tritons in JET plasmas. (authors). 7 refs., 3 figs., 1 tab.
Energy Technology Data Exchange (ETDEWEB)
Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)
2005-07-01
We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.
MINTEQ2 geochemical code: provisionary organic data base
International Nuclear Information System (INIS)
Morrey, J.R.; Krupka, K.M.; Dove, F.H.
1985-10-01
Organic components in aqueous radioactive chemical sources, surface waters, and ground waters could substantially alter the mobility of radioactive and other important nonradioactive elements released from a defense waste disposal system. It is therefore important to be able to predict, as accurately as possible, the effects of selected organic components on the solubilities of radionuclides and important nonradioactive elements. The geochemical code MINTEQ2 can be used to assess solubilities provided that appropriate thermochemical data for organic and inorganic aqueous species and solids are available for its use. The code accepts an assemblage of gaseous and solid phases in contact with an aqueous phase and calculates the thermochemical equilibrium between these phases. Unlike typical hydrologic flow and transport codes where the data base is entirely site specific (i.e., parameters particular to the specific site), MINTEQ2 requires an additional generic thermochemical data base. This report discusses the addition of provisionary organic reactions and associated equilibrium constants to the generic data base that can be used by MINTEQ2 in scoping calculations or preliminary performance assessments
International Nuclear Information System (INIS)
Joseph, M.; Karunasagar, D.; Saha, B.
1996-01-01
Burn-up an important parameter during evaluation of the performance of any nuclear fuel. Among the various techniques available, the preferred one for its determination is based on accurate measurement of a suitable fission product monitor and the residual heavy elements. Since isotopes of rare earth elements are generally used as burn-up monitors, conditions were standardized for rapid separation (within 15 minutes) of light rare earths using high performance liquid chromatography based on either anion exchange (Partisil 10 SAX) in methanol-nitric acid medium or by cation exchange on a reverse phase column (Spherisorb 5-ODS-2 or Supelcosil LC-18) dynamically modified with 1-octane sulfonate or camphor-10-sulfonic acid (β). Both these methods were assessed for separation of individual fission product rare earths from their mixtures. A new approach has been examined in detail for rapid assay of neodymium, which appears promising for faster and accurate measurement of burn-up. (author)
International Nuclear Information System (INIS)
Iqbal, M.; Mehmood, T.; Ayazuddin, S.K.; Salahuddin, A.; Pervez, S.
2001-01-01
Two independent experimental methods have been used for the comparative study of fuel burnup measurement in low enriched uranium, plate type research reactor. In the first method a gamma ray activity ratio method was employed. An experimental setup was established for gamma ray scanning using prior calibrated high purity germanium detector. The computer software KORIGEN gave the theoretical support. In the second method reactivity difference technique was used. At the same location in the same core configuration the fresh and burned fuel element's reactivity worth was estimated. For theoretical estimated curve, group cross-sections were generated using computer code WIMS-D/4, and three dimensional modeling was made by computer code CITATION. The measured burnup of different fuel elements using these methods were found to be in good agreement
Model for evolution of grain size in the rim region of high burnup UO2 fuel
Xiao, Hongxing; Long, Chongsheng; Chen, Hongsheng
2016-04-01
The restructuring process of the high burnup structure (HBS) formation in UO2 fuel results in sub-micron size grains that accelerate the fission gas swelling, which will raise some concern over the safety of extended the nuclear fuel operation life in the reactor. A mechanistic and engineering model for evolution of grain size in the rim region of high burnup UO2 fuel based on the experimental observations of the HBS in the literature is presented. The model takes into account dislocations evolution under irradiation and the grain subdivision occur successively at increasing local burnup. It is assumed that the original driving force for subdivision of grain in the HBS of UO2 fuel is the production and accumulation of dislocation loops during irradiation. The dislocation loops can also be annealed through thermal diffusion when the temperature is high enough. The capability of this model is validated by the comparison with the experimental data of temperature threshold of subdivision, dislocation density and sub-grain size as a function of local burnup. It is shown that the calculated results of the dislocation density and subdivided grain size as a function of local burnup are in good agreement with the experimental results.
Estimation of pore pressure in the rim region of high burnup UO2 fuel
International Nuclear Information System (INIS)
Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong
1999-01-01
An attempt has been made to estimate the pore pressure in the rim region of high burnup UO 2 fuel as a function of rim burnup using the measured rim width, average porosity and pore density in the rim region. First, a linear relationship is developed based on measured rim burnup and rim width. Second, fraction of fission gas retained in the grain boundary of rim region is estimated. Third, total pores in the rim is calculated from the measured pore density in the rim region. Finally using the assumption that all the pores in the rim have the same size of 1.2μm, pore pressure is calculated from the equation of state for ideal gas. An estimated pore pressure of about 60 to 80 MPa for the rim burnup of 90 GWd/tU appears to be in reasonable agreement with other value given in a literature that pore pressure at 800 K become 90-210 MPa for pellet average burnup of 80 GWd/tU
Optical image encryption based on real-valued coding and subtracting with the help of QR code
Deng, Xiaopeng
2015-08-01
A novel optical image encryption based on real-valued coding and subtracting is proposed with the help of quick response (QR) code. In the encryption process, the original image to be encoded is firstly transformed into the corresponding QR code, and then the corresponding QR code is encoded into two phase-only masks (POMs) by using basic vector operations. Finally, the absolute values of the real or imaginary parts of the two POMs are chosen as the ciphertexts. In decryption process, the QR code can be approximately restored by recording the intensity of the subtraction between the ciphertexts, and hence the original image can be retrieved without any quality loss by scanning the restored QR code with a smartphone. Simulation results and actual smartphone collected results show that the method is feasible and has strong tolerance to noise, phase difference and ratio between intensities of the two decryption light beams.
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
Depletion of gadolinium burnable poison in a PWR assembly with high burnup fuel
Energy Technology Data Exchange (ETDEWEB)
Refeat, Riham Mahmoud [Nuclear and Radiological Regulatory Authority (NRRA), Cairo (Egypt). Safety Engineering Dept.
2015-12-15
A tendency to increase the discharge burnup of nuclear fuel for Advanced Pressurized Water Reactors (PWR) has been a characteristic of its operation for many years. It will be able to burn at very high burnup of about 70 GWd/t with UO{sub 2} fuels. The U-235 enrichment must be higher than 5 %, which leads to the necessity of using an extremely efficient burnable poison like Gadolinium oxide. Using gadolinium isotope is significant due to its particular depletion behavior (''Onion-Skin'' effect). In this paper, the MCNPX2.7 code is used to calculate the important neutronic parameters of the next generation fuels of PWR. K-infinity, local peaking factor and fission rate distributions are calculated for a PWR assembly which burn at very high burnup reaching 70 GWd/t. The calculations are performed using the recently released evaluated Gadolinium cross section data. The results obtained are close to those of a LWR next generation fuel benchmark problem. This demonstrates that the calculation scheme used is able to accurately model a PWR assembly that operates at high burnup values.
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
Okumura, Keisuke; Kugo, Teruhiko; Kaneko, Kunio; Tsuchihashi, Keichiro
2007-02-01
The SRAC is a code system applicable to neutronics analysis of a variety of reactor types. Since the publication of the second version of the users manual (JAERI-1302) in 1986 for the SRAC system, a number of additions and modifications to the functions and the library data have been made to establish a comprehensive neutronics code system. The current system includes major neutron data libraries (JENDL-3.3, JENDL-3.2, ENDF/B-VII, ENDF/B-VI.8, JEFF-3.1, JEF-2.2, etc.), and integrates five elementary codes for neutron transport and diffusion calculation; PIJ based on the collision probability method applicable to 16 kind of lattice models, S N transport codes ANISN(1D) and TWOTRN(2D), diffusion codes TUD(1D) and CITATION(multi-D). The system also includes an auxiliary code COREBN for multi-dimensional core burn-up calculation. (author)
Probabilistic Decision Based Block Partitioning for Future Video Coding
Wang, Zhao
2017-11-29
In the latest Joint Video Exploration Team development, the quadtree plus binary tree (QTBT) block partitioning structure has been proposed for future video coding. Compared to the traditional quadtree structure of High Efficiency Video Coding (HEVC) standard, QTBT provides more flexible patterns for splitting the blocks, which results in dramatically increased combinations of block partitions and high computational complexity. In view of this, a confidence interval based early termination (CIET) scheme is proposed for QTBT to identify the unnecessary partition modes in the sense of rate-distortion (RD) optimization. In particular, a RD model is established to predict the RD cost of each partition pattern without the full encoding process. Subsequently, the mode decision problem is casted into a probabilistic framework to select the final partition based on the confidence interval decision strategy. Experimental results show that the proposed CIET algorithm can speed up QTBT block partitioning structure by reducing 54.7% encoding time with only 1.12% increase in terms of bit rate. Moreover, the proposed scheme performs consistently well for the high resolution sequences, of which the video coding efficiency is crucial in real applications.
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.
Compact Hilbert Curve Index Algorithm Based on Gray Code
Directory of Open Access Journals (Sweden)
CAO Xuefeng
2016-12-01
Full Text Available Hilbert curve has best clustering in various kinds of space filling curves, and has been used as an important tools in discrete global grid spatial index design field. But there are lots of redundancies in the standard Hilbert curve index when the data set has large differences between dimensions. In this paper, the construction features of Hilbert curve is analyzed based on Gray code, and then the compact Hilbert curve index algorithm is put forward, in which the redundancy problem has been avoided while Hilbert curve clustering preserved. Finally, experiment results shows that the compact Hilbert curve index outperforms the standard Hilbert index, their 1 computational complexity is nearly equivalent, but the real data set test shows the coding time and storage space decrease 40%, the speedup ratio of sorting speed is nearly 4.3.
DCT/DST-based transform coding for intra prediction in image/video coding.
Saxena, Ankur; Fernandes, Felix C
2013-10-01
In this paper, we present a DCT/DST based transform scheme that applies either the conventional DCT or type-7 DST for all the video-coding intra-prediction modes: vertical, horizontal, and oblique. Our approach is applicable to any block-based intra prediction scheme in a codec that employs transforms along the horizontal and vertical direction separably. Previously, Han, Saxena, and Rose showed that for the intra-predicted residuals of horizontal and vertical modes, the DST is the optimal transform with performance close to the KLT. Here, we prove that this is indeed the case for the other oblique modes. The optimal choice of using DCT or DST is based on intra-prediction modes and requires no additional signaling information or rate-distortion search. The DCT/DST scheme presented in this paper was adopted in the HEVC standardization in March 2011. Further simplifications, especially to reduce implementation complexity, which remove the mode-dependency between DCT and DST, and simply always use DST for the 4 × 4 intra luma blocks, were adopted in the HEVC standard in July 2012. Simulation results conducted for the DCT/DST algorithm are shown in the reference software for the ongoing HEVC standardization. Our results show that the DCT/DST scheme provides significant BD-rate improvement over the conventional DCT based scheme for intra prediction in video sequences.
SWAT2: The improved SWAT code system by incorporating the continuous energy Monte Carlo code MVP
International Nuclear Information System (INIS)
Mochizuki, Hiroki; Suyama, Kenya; Okuno, Hiroshi
2003-01-01
SWAT is a code system, which performs the burnup calculation by the combination of the neutronics calculation code, SRAC95 and the one group burnup calculation code, ORIGEN2.1. The SWAT code system can deal with the cell geometry in SRAC95. However, a precise treatment of resonance absorptions by the SRAC95 code using the ultra-fine group cross section library is not directly applicable to two- or three-dimensional geometry models, because of restrictions in SRAC95. To overcome this problem, SWAT2 which newly introduced the continuous energy Monte Carlo code, MVP into SWAT was developed. Thereby, the burnup calculation by the continuous energy in any geometry became possible. Moreover, using the 147 group cross section library called SWAT library, the reactions which are not dealt with by SRAC95 and MVP can be treated. OECD/NEA burnup credit criticality safety benchmark problems Phase-IB (PWR, a single pin cell model) and Phase-IIIB (BWR, fuel assembly model) were calculated as a verification of SWAT2, and the results were compared with the average values of calculation results of burnup calculation code of each organization. Through two benchmark problems, it was confirmed that SWAT2 was applicable to the burnup calculation of the complicated geometry. (author)
Steady State and Transient Fuel Rod Performance Analyses by Pad and Transuranus Codes
International Nuclear Information System (INIS)
Slyeptsov, O.; Slyeptsov, S.; Kulish, G.; Ostapov, A.; Chernov, I.
2013-01-01
The report performed under IAEA research contract No.15370/L2 describes the analysis results of WWER and PWR fuel rod performance at steady state operation and transients by means of PAD and TRANSURANUS codes. The code TRANSURANUS v1m1j09 developed by Institute for of Transuranium Elements (ITU) was used based on the Licensing Agreement N31302. The code PAD 4.0 developed by Westinghouse Electric Company was utilized in the frame of the Ukraine Nuclear Fuel Qualification Project for safety substantiation for the use of Westinghouse fuel assemblies in the mixed core of WWER-1000 reactor. The experimental data for the Russian fuel rod behavior obtained during the steady-state operation in the WWER-440 core of reactor Kola-3 and during the power transients in the core of MIR research reactor were taken from the IFPE database of the OECD/NEA and utilized for assessing the codes themselves during simulation of such properties as fuel burnup, fuel centerline temperature (FCT), fuel swelling, cladding strain, fission gas release (FGR) and rod internal pressure (RIP) in the rod burnup range of (41 - 60) GWD/MTU. The experimental data of fuel behavior at steady-state operation during seven reactor cycles presented by AREVA for the standard PWR fuel rod design were used to examine the code FGR model in the fuel burnup range of (37 - 81) GWD/MTU. (author)
Energy Technology Data Exchange (ETDEWEB)
Ilas, Germina [ORNL; Gauld, Ian C [ORNL
2011-01-01
This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.
A Modal Expansion Equilibrium Cycle Perturbation Method for Optimizing High Burnup Fast Reactors
Touran, Nicholas W.
This dissertation develops a simulation tool capable of optimizing advanced nuclear reactors considering the multiobjective nature of their design. An Enhanced Equilibrium Cycle (EEC) method based on the classic equilibrium method is developed to evaluate the response of the equilibrium cycle to changes in the core design. Advances are made in the consideration of burnup-dependent cross sections and dynamic fuel performance (fission gas release, fuel growth, and bond squeeze-out) to allow accuracy in high-burnup reactors such as the Traveling Wave Reactor. EEC is accelerated for design changes near a reference state through a new modal expansion perturbation method that expands arbitrary flux perturbations on a basis of λ-eigenmodes. A code is developed to solve the 3-D, multigroup diffusion equation with an Arnoldi-based solver that determines hundreds of the reference flux harmonics and later uses these harmonics to determine expansion coefficients required to approximate the perturbed flux. The harmonics are only required for the reference state, and many substantial and localized perturbations from this state are shown to be well-approximated with efficient expressions after the reference calculation is performed. The modal expansion method is coupled to EEC to produce the later-in-time response of each design perturbation. Because the code determines the perturbed flux explicitly, a wide variety of core performance metrics may be monitored by working within a recently-developed data management system called the ARMI. Through ARMI, the response of each design perturbation may be evaluated not only for the flux and reactivity, but also for reactivity coefficients, thermal hydraulics parameters, economics, and transient performance. Considering the parameters available, an automated optimization framework is designed and implemented. A non-parametric surrogate model using the Alternating Conditional Expectation (ACE) algorithm is trained with many design
Application of scale-4 depletion/criticality sequences in burnup credit analyses
International Nuclear Information System (INIS)
Brady, M.C.
1991-01-01
The concept of allowing reactivity credit for the transmuted state of spent fuel complicates the criticality analysis by requiring the specification of the fuel mixture to potentially include large numbers of isotopes representative of the fuel conditions. These conditions include the initial enrichment, local or average burnup conditions depending on the analysis approach, and the post-shutdown cooling time. In the development of an analysis methodology to evaluate spent fuel shipping and transport casks (flasks) based on this burnup credit, commercial reactor critical configurations were evaluated as potential experimental spent fuel criticals. This paper describes how the SCALE-4 depletion sequences (SAS2H), the cross-section processing sequence (CSASN), and the criticality module (KENO V.a) were used to evaluate these reactor criticals. A description of a newly developed sequence for linking SAS2H calculated burnup-dependent isotopics to KENO V.a mixing tables [SAS2H Nuclide Inventories for KENO Runs (SNIKR)] is also included
Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code
International Nuclear Information System (INIS)
Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.
2004-01-01
Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials
Rhodes, Gillian; Jeffery, Linda; Boeing, Alexandra; Calder, Andrew J.
2013-01-01
Despite the discovery of body-selective neural areas in occipitotemporal cortex, little is known about how bodies are visually coded. We used perceptual adaptation to determine how body identity is coded. Brief exposure to a body (e.g., anti-Rose) biased perception toward an identity with opposite properties (Rose). Moreover, the size of this…
A network coding based routing protocol for underwater sensor networks.
Wu, Huayang; Chen, Min; Guan, Xin
2012-01-01
Due to the particularities of the underwater environment, some negative factors will seriously interfere with data transmission rates, reliability of data communication, communication range, and network throughput and energy consumption of underwater sensor networks (UWSNs). Thus, full consideration of node energy savings, while maintaining a quick, correct and effective data transmission, extending the network life cycle are essential when routing protocols for underwater sensor networks are studied. In this paper, we have proposed a novel routing algorithm for UWSNs. To increase energy consumption efficiency and extend network lifetime, we propose a time-slot based routing algorithm (TSR).We designed a probability balanced mechanism and applied it to TSR. The theory of network coding is introduced to TSBR to meet the requirement of further reducing node energy consumption and extending network lifetime. Hence, time-slot based balanced network coding (TSBNC) comes into being. We evaluated the proposed time-slot based balancing routing algorithm and compared it with other classical underwater routing protocols. The simulation results show that the proposed protocol can reduce the probability of node conflicts, shorten the process of routing construction, balance energy consumption of each node and effectively prolong the network lifetime.
Local coding based matching kernel method for image classification.
Directory of Open Access Journals (Sweden)
Yan Song
Full Text Available This paper mainly focuses on how to effectively and efficiently measure visual similarity for local feature based representation. Among existing methods, metrics based on Bag of Visual Word (BoV techniques are efficient and conceptually simple, at the expense of effectiveness. By contrast, kernel based metrics are more effective, but at the cost of greater computational complexity and increased storage requirements. We show that a unified visual matching framework can be developed to encompass both BoV and kernel based metrics, in which local kernel plays an important role between feature pairs or between features and their reconstruction. Generally, local kernels are defined using Euclidean distance or its derivatives, based either explicitly or implicitly on an assumption of Gaussian noise. However, local features such as SIFT and HoG often follow a heavy-tailed distribution which tends to undermine the motivation behind Euclidean metrics. Motivated by recent advances in feature coding techniques, a novel efficient local coding based matching kernel (LCMK method is proposed. This exploits the manifold structures in Hilbert space derived from local kernels. The proposed method combines advantages of both BoV and kernel based metrics, and achieves a linear computational complexity. This enables efficient and scalable visual matching to be performed on large scale image sets. To evaluate the effectiveness of the proposed LCMK method, we conduct extensive experiments with widely used benchmark datasets, including 15-Scenes, Caltech101/256, PASCAL VOC 2007 and 2011 datasets. Experimental results confirm the effectiveness of the relatively efficient LCMK method.
International Nuclear Information System (INIS)
Suyama, Kenya; Mochizuki, Hiroki
2006-01-01
The value of the burnup is one of the most important parameters of samples taken by post-irradiation examination (PIE). Generally, it is evaluated by the Neodymium-148 method. Precise evaluation of the burnup value requires: (1) an effective fission yield of 148 Nd; (2) neutron capture reactions of 147 Nd and 148 Nd; (3) a conversion factor from fissions per initial heavy metal to the burnup unit GWd/t. In this study, the burnup values of the PIE data from Mihama-3 and Genkai-1 PWRs, which were taken by the Japan Atomic Energy Research Institute, were re-evaluated using more accurate corrections for each of these three items. The PIE data were then re-analyzed using SWAT and SWAT2 code systems with JENDL-3.3 library. The re-evaluation of the effective fission yield of 148 Nd has an effect of 1.5-2.0% on burnup values. Considering the neutron capture reactions of 147 Nd and 148 Nd removes dependence of C/E values of 148 Nd on the burnup value. The conversion factor from FIMA(%) to GWd/t changes according to the burnup value. Its effect on the burnup evaluation is small for samples having burnup of larger than 30 GWd/t. The analyses using the corrected burnup values showed that the calculated 148 Nd concentrations and the PIE data is approximately 1%, whereas this was 3-5% in prior analyses. This analysis indicates that the burnup values of samples from Mihama-3 and Genkai-1 PWRs should be corrected by 2-3%. The effect of re-evaluation of the burnup value on the neutron multiplication factor is an approximately 0.6% change in PIE samples having the burnup of larger than 30 GWd/t. Finally, comparison between calculation results using a single pin-cell model and an assembly model is carried out. Because the results agreed with each other within a few percent, we concluded that the single pin-cell model is suitable for the analysis of PIE samples and that the underestimation of plutonium isotopes, which occurred in the previous analyses, does not result from a geometry
Nuclear component design ontology building based on ASME codes
International Nuclear Information System (INIS)
Bao Shiyi; Zhou Yu; He Shuyan
2005-01-01
The adoption of ontology analysis in the study of concept knowledge acquisition and representation for the nuclear component design process based on computer-supported cooperative work (CSCW) makes it possible to share and reuse numerous concept knowledge of multi-disciplinary domains. A practical ontology building method is accordingly proposed based on Protege knowledge model in combination with both top-down and bottom-up approaches together with Formal Concept Analysis (FCA). FCA exhibits its advantages in the way it helps establish and improve taxonomic hierarchy of concepts and resolve concept conflict occurred in modeling multi-disciplinary domains. With Protege-3.0 as the ontology building tool, a nuclear component design ontology based ASME codes is developed by utilizing the ontology building method. The ontology serves as the basis to realize concept knowledge sharing and reusing of nuclear component design. (authors)
Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor
Directory of Open Access Journals (Sweden)
T. Setiadipura
2015-04-01
Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble
Preliminary neutronic design of high burnup OTTO cycle pebble bed reactor
International Nuclear Information System (INIS)
Setiadipura, T.; Zuhair; Irwanto, D.
2015-01-01
The pebble bed type High Temperature Gas-cooled Reactor (HTGR) is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR) which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO) cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM) loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble. (author)
Safety analysis of MOX fuels by fuel performance code
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-12-01
Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)
Code cases for implementing risk-based inservice testing in the ASME OM code
Energy Technology Data Exchange (ETDEWEB)
Rowley, C.W.
1996-12-01
Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices.
Directory of Open Access Journals (Sweden)
Kępisty Grzegorz
2015-09-01
Full Text Available In this paper, we compare the methodology of different time-step models in the context of Monte Carlo burnup calculations for nuclear reactors. We discuss the differences between staircase step model, slope model, bridge scheme and stochastic implicit Euler method proposed in literature. We focus on the spatial stability of depletion procedure and put additional emphasis on the problem of normalization of neutron source strength. Considered methodology has been implemented in our continuous energy Monte Carlo burnup code (MCB5. The burnup simulations have been performed using the simplified high temperature gas-cooled reactor (HTGR system with and without modeling of control rod withdrawal. Useful conclusions have been formulated on the basis of results.
DNA as a Binary Code: How the Physical Structure of Nucleotide Bases Carries Information
McCallister, Gary
2005-01-01
The DNA triplet code also functions as a binary code. Because double-ring compounds cannot bind to double-ring compounds in the DNA code, the sequence of bases classified simply as purines or pyrimidines can encode for smaller groups of possible amino acids. This is an intuitive approach to teaching the DNA code. (Contains 6 figures.)
International Nuclear Information System (INIS)
Jayalal, M.L.; Ramachandran, Suja; Rathakrishnan, S.; Satya Murty, S.A.V.; Sai Baba, M.
2015-01-01
Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the
Gröbner Bases, Coding, and Cryptography
Sala, Massimiliano; Perret, Ludovic
2009-01-01
Coding theory and cryptography allow secure and reliable data transmission, which is at the heart of modern communication. This book offers a comprehensive overview on the application of commutative algebra to coding theory and cryptography. It analyzes important properties of algebraic/geometric coding systems individually.
International Nuclear Information System (INIS)
Haggard, D.L.; Tanner, J.E.
1997-06-01
Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure 134/137 Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. 134/137 Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from 134 Cs and the 662 keV peak from 137 Cs/ 137m Ba. Assembly No. 2 134/137 Cs ratio measured on 4-15-92 was 0.19. The measured 134/137 Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of 134 Cs and 137 Cs. The measured 134/137 Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of 134 Cs and 8620--11646 curies of 137 Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies 134 Cs and 8973 curies 137 Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured 134 Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured 137 Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material
Energy Technology Data Exchange (ETDEWEB)
Haggard, D.L.; Tanner, J.E.
1997-06-01
Nondestructive assay (NDA) gamma spectroscopy techniques were used to measure {sup 134/137}Cs ratios on nine PuAl Mark 42 fuel assemblies. The purpose of the ratio measurement was to confirm theoretical burnup calculations. {sup 134/137}Cs ratios were determined from the measured activity based on corrected net peak area counts for the 605 keV peak from {sup 134}Cs and the 662 keV peak from {sup 137}Cs/{sup 137m}Ba. Assembly No. 2 {sup 134/137}Cs ratio measured on 4-15-92 was 0.19. The measured {sup 134/137}Cs ratio was decay corrected to be 2.11 on 8-1-84 based on the half lives of {sup 134}Cs and {sup 137}Cs. The measured {sup 134/137}Cs ratio range was 1.90--2.14 for all nine assemblies. These measured values compare to a theoretical ratio of 1.7 on 8-1-84 determined by burnup calculations. Total cesium curie content was also requested and determined using the NDA direct measurements. Gamma spectral data were measured on the nine sectioned Mark 42 fuel assemblies. Measured cesium curie content, decay corrected to 8-1-84, ranged from 18170--24480 curies of {sup 134}Cs and 8620--11646 curies of {sup 137}Cs. Theoretical cesium curie content of 8-1-84 was 15200 curies {sup 134}Cs and 8973 curies {sup 137}Cs. Direct assay cesium ratio is 12% to 26% higher than the predicted ratio of 1.7. The measured {sup 134}Cs data indicate between 20%--61% more activity than that predicted by the burnup code, whereas the measured {sup 137}Cs activity is between 4% less to 30% more than the predicted activity. This information may be used to address issues concerning criticality safety, storage, and shipping of this type of material.
The Impact of Operating Parameters and Correlated Parameters for Extended BWR Burnup Credit
Energy Technology Data Exchange (ETDEWEB)
Ade, Brian J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Marshall, William B. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bowman, Stephen M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2017-06-01
Applicants for certificates of compliance for spent nuclear fuel (SNF) transportation and dry storage systems perform analyses to demonstrate that these systems are adequately subcritical per the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For pressurized water reactor (PWR) SNF, these analyses may credit the reduction in assembly reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during power operation. This credit for reactivity reduction during depletion is commonly referred to as burnup credit (BUC). US Nuclear Regulatory Commission (NRC) staff review BUC analyses according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks.
International Nuclear Information System (INIS)
Ikeda, H.; Kikuchi, T.; Ono, S.
2001-01-01
FPAC (fuel performance analysis code) is a thermal and mechanical design code developed by Nuclear Fuel Industries, Ltd. (NFI). NFI has been utilizing FPAC to design the commercial PWR fuel rods, which are manufactured by NFI, or to analyze fuel rod's irradiation behavior. Japanese PWR utilities are planning to increase the discharged burnup limit of fuel assembly to 55GWd/t. Therefore, NFI has improved FPAC to enable to be used to more accurately estimate fuel rod behavior at high burnup. The integral assessment of improved FPAC has been verified from available data including measured centerline temperatures, PIE data and so on, and they cover burnup up to approximately 90GWd/t. The predicted results obtained from the improved FPAC show good agreement with measured data. (author)
Development of multi-physics code systems based on the reactor dynamics code DYN3D
Energy Technology Data Exchange (ETDEWEB)
Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)
2011-07-15
The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)
Development of multi-physics code systems based on the reactor dynamics code DYN3D
International Nuclear Information System (INIS)
Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo
2011-01-01
The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)
Criticality implications of extended fuel burnup
International Nuclear Information System (INIS)
Eng, R.
1985-01-01
The advantages (and disadvantages) of extended fuel burnup to the operating nuclear utility are well documented. Niagara Mohawk found that increasing the refueling interval from 12 months to 24 months led to a 60% increase in full power days per refueling outage day. Today's incentives include decreasing spent fuel accumulation, increasing plant capacity factors through longer cycles, reducing uranium ore requirements, reducing radiation exposures of workers during refueling, reduced disposal requirements, reduced number of heat-up and cooldown transients, reduced plant security burden (during refueling outages), and reduced manpower time for regulatory review
Burnup characteristics of binary breeder reactors
International Nuclear Information System (INIS)
Dias, A.F.; Nascimento, J.A. do; Ishiguro, Y.
1983-01-01
Burnup calculations of a binary breeder reactor have been done for two cases of fueling. In one case the U 233 /TH fueled inner core and the Pu/U-fueled outer core have the same number of fuel assemblies. In the other case two outermost rings in the inner core are Pu/U-fueled. The second case is considered for an initial phase of thorim cycle introduction when the supply of U 233 could be limited. Results show an efficient breeding on the thorium cycle in both cases. (Author) [pt
Optical fingerprint recognition based on local minutiae structure coding.
Yi, Yao; Cao, Liangcai; Guo, Wei; Luo, Yaping; Feng, Jianjiang; He, Qingsheng; Jin, Guofan
2013-07-15
A parallel volume holographic optical fingerprint recognition system robust to fingerprint translation, rotation and nonlinear distortion is proposed. The optical fingerprint recognition measures the similarity by using the optical filters of multiplexed holograms recorded in the holographic media. A fingerprint is encoded into multiple template data pages based on the local minutiae structure coding method after it is adapted for the optical data channel. An improved filter recording time schedule and a post-filtering calibration technology are combined to suppress the calculating error from the large variations in data page filling ratio. Experimental results tested on FVC2002 DB1 and a forensic database comprising 270,216 fingerprints demonstrate the robustness and feasibility of the system.
Energy Technology Data Exchange (ETDEWEB)
Pusa, M.; Leppaenen, J. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)
2012-07-01
The Chebyshev Rational Approximation Method (CRAM) has been recently introduced by the authors for solving the burnup equations with excellent results. This method has been shown to be capable of simultaneously solving an entire burnup system with thousands of nuclides both accurately and efficiently. The method was prompted by an analysis of the spectral properties of burnup matrices and it can be characterized as the best rational approximation on the negative real axis. The coefficients of the rational approximation are fixed and have been reported for various approximation orders. In addition to these coefficients, implementing the method only requires a linear solver. This paper describes an efficient method for solving the linear systems associated with the CRAM approximation. The introduced direct method is based on sparse Gaussian elimination where the sparsity pattern of the resulting upper triangular matrix is determined before the numerical elimination phase. The stability of the proposed Gaussian elimination method is discussed based on considering the numerical properties of burnup matrices. Suitable algorithms are presented for computing the symbolic factorization and numerical elimination in order to facilitate the implementation of CRAM and its adoption into routine use. The accuracy and efficiency of the described technique are demonstrated by computing the CRAM approximations for a large test case with over 1600 nuclides. (authors)
Benchmarking burnup reconstruction methods for dynamically operated research reactors
Energy Technology Data Exchange (ETDEWEB)
Sternat, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Charlton, William S. [Univ. of Nebraska, Lincoln, NE (United States). National Strategic Research Institute; Nichols, Theodore F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
2016-03-01
The burnup of an HEU fueled dynamically operated research reactor, the Oak Ridge Research Reactor, was experimentally reconstructed using two different analytic methodologies and a suite of signature isotopes to evaluate techniques for estimating burnup for research reactor fuel. The methods studied include using individual signature isotopes and the complete mass spectrometry spectrum to recover the sample’s burnup. The individual, or sets of, isotopes include ^{148}Nd, ^{137}Cs+^{137}Ba, ^{139}La, and ^{145}Nd+^{146}Nd. The storage documentation from the analyzed fuel material provided two different measures of burnup: burnup percentage and the total power generated from the assembly in MWd. When normalized to conventional units, these two references differed by 7.8% (395.42GWd/MTHM and 426.27GWd/MTHM) in the resulting burnup for the spent fuel element used in the benchmark. Among all methods being evaluated, the results were within 11.3% of either reference burnup. The results were mixed in closeness to both reference burnups; however, consistent results were achieved from all three experimental samples.
Modeling of PWR fuel at extended burnup
International Nuclear Information System (INIS)
Dias, Raphael Mejias
2016-01-01
This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)
Improvement of QR Code Recognition Based on Pillbox Filter Analysis
Directory of Open Access Journals (Sweden)
Jia-Shing Sheu
2013-04-01
Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.
Finnish contribution to the CB4 burnup credit benchmark
International Nuclear Information System (INIS)
Wasastjerna, F.
2001-01-01
The CB4 phase of the WWER burnup credit benchmark series studies the effect of flat and realistic axial burnup profiles on the multiplication factor of a conceptual WWER cask loaded with spent fuel. The benchmark was calculated at VTT Energy with MCNP4C, using mainly ENDF/B-V1 cross sections. According to the calculation results the effect of the axial homogenization on the k eff estimate is complex. At low burnups the use of a axial profile overestimates k eff but at high burnups the reverse is the case. Ignoring fission products leads to conservative k eff and the effect of axial homogenization on the multiplication factor is similar to a reduction of the burnup (Authors)
Burnup credit applications in a high-capacity truck cask
International Nuclear Information System (INIS)
Boshoven, J.K.
1992-09-01
General Atomics (GA) has designed two legal weight truck (LWT) casks, the GA-4 and GA-9, to carry four pressurized-water-reactor (PWR) and nine boiling-water-reactor (BWR) fuel assemblies, respectively. GA plans to submit applications for certification to the US Nuclear Regulatory Commission (NRC) for the two casks in mid-1993. GA will include burnup credit analysis in the Safety Analysis Report for Packaging (SARP) for the GA-4 Cask. By including burnup credit in the criticality safety analysis for PWR fuels with initial enrichments above 3% U-235, public and occupation risks are reduced and cost savings are realized. The GA approach to burnup credit analysis incorporates the information produced in the US Department of Energy Burnup Credit Program. This paper describes the application of burnup credit to the criticality control design of the GA-4 Cask
Progress on RMC: a Monte Carlo neutron transport code for reactor analysis
International Nuclear Information System (INIS)
Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin
2011-01-01
This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)
Radiochemical Assays of Irradiated VVER-440 Fuel for Use in Spent Fuel Burnup Credit Activities
International Nuclear Information System (INIS)
Jardine, L J
2005-01-01
The objective of this spent fuel burnup credit work was to study and describe a VVER-440 reactor spent fuel assembly (FA) initial state before irradiation, its operational irradiation history and the resulting radionuclide distribution in the fuel assembly after irradiation. This work includes the following stages: (1) to pick out and select a specific spent (irradiated) FA for examination; (2) to describe the FA initial state before irradiation; (3) to describe the irradiation history, including thermal calculations; (4) to examine the burnup distribution of select radionuclides along the FA height and cross-section; (5) to examine the radionuclide distributions; (6) to determine the Kr-85 release into the plenum; (7) to select and prepare FA rod specimens for destructive examinations; (8) to determine the radionuclide compositions, isotope masses and burnup in the rod specimens; and (9) to analyze, document and process the results. The specific workscope included the destructive assay (DA) of spent fuel assembly rod segments with an ∼38.5 MWd/KgU burnup from a single VVER-440 fuel assembly from the Novovorenezh reactor in Russia. Based on irradiation history criteria, four rods from the fuel assembly were selected and removed from the assembly for examination. Next, 8 sections were cut from the four rods and sent for destructive analysis of radionuclides by radiochemical analyses. The results were documented in a series of seven reports over a period of ∼1 1/2 years
Evaluation of Cross-Section Sensitivities in Computing Burnup Credit Fission Product Concentrations
International Nuclear Information System (INIS)
Gauld, I.C.
2005-01-01
U.S. Nuclear Regulatory Commission Interim Staff Guidance 8 (ISG-8) for burnup credit covers actinides only, a position based primarily on the lack of definitive critical experiments and adequate radiochemical assay data that can be used to quantify the uncertainty associated with fission product credit. The accuracy of fission product neutron cross sections is paramount to the accuracy of criticality analyses that credit fission products in two respects: (1) the microscopic cross sections determine the reactivity worth of the fission products in spent fuel and (2) the cross sections determine the reaction rates during irradiation and thus influence the accuracy of predicted final concentrations of the fission products in the spent fuel. This report evaluates and quantifies the importance of the fission product cross sections in predicting concentrations of fission products proposed for use in burnup credit. The study includes an assessment of the major fission products in burnup credit and their production precursors. Finally, the cross-section importances, or sensitivities, are combined with the importance of each major fission product to the system eigenvalue (k eff ) to determine the net importance of cross sections to k eff . The importances established the following fission products, listed in descending order of priority, that are most likely to benefit burnup credit when their cross-section uncertainties are reduced: 151 Sm, 103 Rh, 155 Eu, 150 Sm, 152 Sm, 153 Eu, 154 Eu, and 143 Nd
Modeling of Pore Coarsening in the Rim Region of High Burn-up UO2 Fuel
Directory of Open Access Journals (Sweden)
Hongxing Xiao
2016-08-01
Full Text Available An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS of irradiated UO2 fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO2 fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO2 fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.
Analysis of bubble pressure in the rim region of high burnup PWR fuel
Energy Technology Data Exchange (ETDEWEB)
Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-02-01
Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)
Kinetic Monte Carlo Potts Model for Simulating a High Burnup Structure in UO2
International Nuclear Information System (INIS)
Oh, Jae-Yong; Koo, Yang-Hyun; Lee, Byung-Ho
2008-01-01
A Potts model, based on the kinetic Monte Carlo method, was originally developed for magnetic domain evolutions, but it was also proposed as a model for a grain growth in polycrystals due to similarities between Potts domain structures and grain structures. It has modeled various microstructural phenomena such as grain growths, a recrystallization, a sintering, and so on. A high burnup structure (HBS) is observed in the periphery of a high burnup UO 2 fuel. Although its formation mechanism is not clearly understood yet, its characteristics are well recognized: The HBS microstructure consists of very small grains and large bubbles instead of original as-sintered grains. A threshold burnup for the HBS is observed at a local burnup 60-80 Gwd/tM, and the threshold temperature is 1000-1200 .deg. C. Concerning a energy stability, the HBS can be created if the system energy of the HBS is lower than that of the original structure in an irradiated UO 2 . In this paper, a Potts model was implemented for simulating the HBS by calculating system energies, and the simulation results were compared with the HBS characteristics mentioned above
Preserving Envelope Efficiency in Performance Based Code Compliance
Energy Technology Data Exchange (ETDEWEB)
Thornton, Brian A. [Thornton Energy Consulting (United States); Sullivan, Greg P. [Efficiency Solutions (United States); Rosenberg, Michael I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Baechler, Michael C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2015-06-20
The City of Seattle 2012 Energy Code (Seattle 2014), one of the most progressive in the country, is under revision for its 2015 edition. Additionally, city personnel participate in the development of the next generation of the Washington State Energy Code and the International Energy Code. Seattle has pledged carbon neutrality by 2050 including buildings, transportation and other sectors. The United States Department of Energy (DOE), through Pacific Northwest National Laboratory (PNNL) provided technical assistance to Seattle in order to understand the implications of one potential direction for its code development, limiting trade-offs of long-lived building envelope components less stringent than the prescriptive code envelope requirements by using better-than-code but shorter-lived lighting and heating, ventilation, and air-conditioning (HVAC) components through the total building performance modeled energy compliance path. Weaker building envelopes can permanently limit building energy performance even as lighting and HVAC components are upgraded over time, because retrofitting the envelope is less likely and more expensive. Weaker building envelopes may also increase the required size, cost and complexity of HVAC systems and may adversely affect occupant comfort. This report presents the results of this technical assistance. The use of modeled energy code compliance to trade-off envelope components with shorter-lived building components is not unique to Seattle and the lessons and possible solutions described in this report have implications for other jurisdictions and energy codes.
International Nuclear Information System (INIS)
Nomura, Yasushi; Okuno, Hiroshi; Miyoshi, Yoshinori
2004-03-01
Firstly, concerning the methods to set burnup for depletion calculation linked with criticality safety evaluation taking burnup credit into consideration, the upper 50 cm averaged burnups approved by regulations in European countries and USA are comparatively evaluated. Secondary, errors produced by different shapes of axial spent fuel burnup distribution assumed for criticality calculation, bias errors associated with depletion calculation compensated by correction factors applied to calculated nuclide isotopic composition, and statistic errors exerted by variation of irradiation history parameters used as input data for depletion calculation, are separately evaluated by performing criticality analyses with the spent fuel transport cask model of OECD/NEA Burnup Credit Criticality Benchmark. As a result, methods are proposed to set equivalent burnups reduced from a given burnup so as to compensate these errors to obtain criticality calculation results on the conservative side. Finally, 'Equivalent Uniform Burnup' and Equivalent Initial Enrichment' which are derived by incorporating these errors synthetically, are described to mention possibility of their common usage irrespective of difference in spent fuel transport cask specification. (author)
The burnup dependence of light water reactor spent fuel oxidation
Energy Technology Data Exchange (ETDEWEB)
Hanson, B.D.
1998-07-01
Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies
Efficacy of Code Optimization on Cache-based Processors
VanderWijngaart, Rob F.; Chancellor, Marisa K. (Technical Monitor)
1997-01-01
The current common wisdom in the U.S. is that the powerful, cost-effective supercomputers of tomorrow will be based on commodity (RISC) micro-processors with cache memories. Already, most distributed systems in the world use such hardware as building blocks. This shift away from vector supercomputers and towards cache-based systems has brought about a change in programming paradigm, even when ignoring issues of parallelism. Vector machines require inner-loop independence and regular, non-pathological memory strides (usually this means: non-power-of-two strides) to allow efficient vectorization of array operations. Cache-based systems require spatial and temporal locality of data, so that data once read from main memory and stored in high-speed cache memory is used optimally before being written back to main memory. This means that the most cache-friendly array operations are those that feature zero or unit stride, so that each unit of data read from main memory (a cache line) contains information for the next iteration in the loop. Moreover, loops ought to be 'fat', meaning that as many operations as possible are performed on cache data-provided instruction caches do not overflow and enough registers are available. If unit stride is not possible, for example because of some data dependency, then care must be taken to avoid pathological strides, just ads on vector computers. For cache-based systems the issues are more complex, due to the effects of associativity and of non-unit block (cache line) size. But there is more to the story. Most modern micro-processors are superscalar, which means that they can issue several (arithmetic) instructions per clock cycle, provided that there are enough independent instructions in the loop body. This is another argument for providing fat loop bodies. With these restrictions, it appears fairly straightforward to produce code that will run efficiently on any cache-based system. It can be argued that although some of the important
Manufacturing Data Uncertainties Propagation Method in Burn-Up Problems
Directory of Open Access Journals (Sweden)
Thomas Frosio
2017-01-01
Full Text Available A nuclear data-based uncertainty propagation methodology is extended to enable propagation of manufacturing/technological data (TD uncertainties in a burn-up calculation problem, taking into account correlation terms between Boltzmann and Bateman terms. The methodology is applied to reactivity and power distributions in a Material Testing Reactor benchmark. Due to the inherent statistical behavior of manufacturing tolerances, Monte Carlo sampling method is used for determining output perturbations on integral quantities. A global sensitivity analysis (GSA is performed for each manufacturing parameter and allows identifying and ranking the influential parameters whose tolerances need to be better controlled. We show that the overall impact of some TD uncertainties, such as uranium enrichment, or fuel plate thickness, on the reactivity is negligible because the different core areas induce compensating effects on the global quantity. However, local quantities, such as power distributions, are strongly impacted by TD uncertainty propagations. For isotopic concentrations, no clear trends appear on the results.
Energy Technology Data Exchange (ETDEWEB)
Toubon, H. [Cogema, 78 - Saint Quentin en Yvelines (France); Guillou, E. [Cogema Etablissement de la Hague, D/SQ/SMT, 50 - Beaumont Hague (France); Cousinou, P. [CEA Fontenay aux Roses, Inst. de Protection et de Surete Nucleaire, 92 (France); Barbry, F. [CEA Valduc, Inst. de Protection et de Surete Nucleaire, 21 - Is sur Tille (France); Grouiller, J.P.; Bignan, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France)
2001-07-01
Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis.
International Nuclear Information System (INIS)
Toubon, H.; Guillou, E.; Cousinou, P.; Barbry, F.; Grouiller, J.P.; Bignan, G.
2001-01-01
Burn-up credit can be defined as making allowance for absorbent radioactive isotopes in criticality studies, in order to optimise safety margins and avoid over-engineering of nuclear facilities. As far as the COGEMA Group is concerned, the three fields in which burn-up credit proves to be an advantage are the transport of spent fuel assemblies, their interim storage in spent fuel pools and reprocessing. In the case of transport, burn-up credit means that cask size do not need to be altered, despite an increase in the initial enrichment of the fuel assemblies. Burn-up credit also makes it possible to offer new cask designs with higher capacity. Burn-up credit means that fuel assemblies with a higher initial enrichment can be put into interim storage in existing facilities and opens the way to the possibility of more compact ones. As far as reprocessing is concerned, burn-up credit makes it possible to keep up current production rates, despite an increase in the initial enrichment of the fuel assemblies being reprocessed. In collaboration with the French Atomic Energy Commission and the Institute for Nuclear Safety and Protection, the COGEMA Group is participating in an extensive experimental programme and working to qualify criticality and fuel depletion computer codes. The research programme currently underway should mean that by 2003, allowance will be made for fission products in criticality safety analysis
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)
Optimalisation Of Oxide Burn-Up Enhanced For RSG-Gas Core
International Nuclear Information System (INIS)
Tukiran; Sembiring, Tagor Malem
2000-01-01
Strategy of fuel management of the RSG-Gas core has been changed from 6/1 to 5/1 pattern so the evaluation of fuel management is necessary to be done. The aim of evaluation is to look for the optimal fuel management so that the fuel can be stayed longer in the core and finally can save cost of operation. Using Batan-EQUIL-2D code did the evaluation of fuel management with 5/1 pattern. The result of evaluation is used to choose which one is more advantage without break the safety margin which is available in the Safety Analysis Report (SAR) firstly, the fuel management was calculated with core excess reactivity of 9,2% criteria. Secondly, fuel burn-up maximum of 56% criteria and the last, fuel burn-up maximum of 64% criteria. From the result of fuel management calculation of the RSG-Gas equilibrium core can be concluded that the optimal RSG-Gas equilibrium core with 5/1 pattern is if the fuel burn-up maximum 64% and the energy in a cycle of operation is 715 MWD. The fuel can be added one more step in the core without break any safety margin. It means that the RSG-Gas equilibrium core can save fuel and cost reduction
Some Thermodynamic Features of Uranium-Plutonium Nitride Fuel in the Course of Burnup
Rusinkevich, A. A.; Ivanov, A. S.; Belov, G. V.; Skupov, M. V.
2017-12-01
Calculation studies on the effect of carbon and oxygen impurities on the chemical and phase compositions of nitride uranium-plutonium fuel in the course of burnup are performed using the IVTANTHERMO code. It is shown that the number of moles of UN decreases with increasing burnup level, whereas UN1.466, UN1.54, and UN1.73 exhibit a considerable increase. The presence of oxygen and carbon impurities causes an increase in the content of the UN1.466, UN1.54 and UN1.73 phases in the initial fuel by several orders of magnitude, in particular, at a relatively low temperature. At the same time, the presence of impurities abruptly reduces the content of free uranium in unburned fuel. Plutonium in the considered system is contained in form of Pu, PuC, PuC2, Pu2C3, and PuN. Plutonium carbides, as well as uranium carbides, are formed in small amounts. Most of the plutonium remains in the form of nitride PuN, whereas unbound Pu is present only in the areas with a low burnup level and high temperatures.
Highly parallel line-based image coding for many cores.
Peng, Xiulian; Xu, Jizheng; Zhou, You; Wu, Feng
2012-01-01
Computers are developing along with a new trend from the dual-core and quad-core processors to ones with tens or even hundreds of cores. Multimedia, as one of the most important applications in computers, has an urgent need to design parallel coding algorithms for compression. Taking intraframe/image coding as a start point, this paper proposes a pure line-by-line coding scheme (LBLC) to meet the need. In LBLC, an input image is processed line by line sequentially, and each line is divided into small fixed-length segments. The compression of all segments from prediction to entropy coding is completely independent and concurrent at many cores. Results on a general-purpose computer show that our scheme can get a 13.9 times speedup with 15 cores at the encoder and a 10.3 times speedup at the decoder. Ideally, such near-linear speeding relation with the number of cores can be kept for more than 100 cores. In addition to the high parallelism, the proposed scheme can perform comparatively or even better than the H.264 high profile above middle bit rates. At near-lossless coding, it outperforms H.264 more than 10 dB. At lossless coding, up to 14% bit-rate reduction is observed compared with H.264 lossless coding at the high 4:4:4 profile.
IAEA Consultancy on Technical Influence of High Burnup UOX and MOX LWR Fuel on Spent Fuel Management
International Nuclear Information System (INIS)
Einziger, R.E.; Lovasic, Z.
2015-01-01
This paper reviews the results of the International Atomic Energy Agency (IAEA) project investigating the influence of high burnup and mixed-oxide (MOX) fuels, from water power reactors, on SFM. These data will provide information on the impacts, regarding SFM, for those countries operating light-water reactors (LWR) and heavy-water reactors (HWR)s with zirconium alloy-clad uranium dioxide (UOX) fuels, that are considering the use of higher burnup UOX or the introduction of reprocessing and MOX fuels. The mechanical designs of lower burnup UOX and higher burnup UOX or MOX fuel are very similar, but some of the properties (e.g., higher fuel rod internal pressures; higher decay heat; higher specific activity; and degraded cladding mechanical properties of higher burnup UOX and MOX spent fuels) may potentially significantly affect the behavior of the fuel after irradiation. The effects of these property changes on wet and dry storage, transportation, reprocessing, refabrication of fuel, and final disposal were evaluated, based on regulatory, safety, and operational considerations. Political and strategic considerations were not taken into account since relative importance of technical, economic and strategic considerations vary from country to country. There will also be an impact of these fuels on issues like non-proliferation, safeguards, and sustainability, but because of the complexity of factors affecting those issues, they are only briefly discussed. The advantages and drawbacks of using high burnup UOX or MOX, for each applicable issue in each stage of the back end of the fuel cycle, were evaluated and are discussed. Although, in theory, higher burnup fuel and MOX fuels mean a smaller quantity of spent fuel, the potential need for some changes in design of spent fuel storage, transportation, handling, reprocessing, refabrication, and disposal will have to be balanced with the benefits of their use. (author)
De Matteis, Sara; Jarvis, Deborah; Young, Heather; Young, Alan; Allen, Naomi; Potts, James; Darnton, Andrew; Rushton, Lesley; Cullinan, Paul
2017-03-01
Objectives The standard approach to the assessment of occupational exposures is through the manual collection and coding of job histories. This method is time-consuming and costly and makes it potentially unfeasible to perform high quality analyses on occupational exposures in large population-based studies. Our aim was to develop a novel, efficient web-based tool to collect and code lifetime job histories in the UK Biobank, a population-based cohort of over 500 000 participants. Methods We developed OSCAR (occupations self-coding automatic recording) based on the hierarchical structure of the UK Standard Occupational Classification (SOC) 2000, which allows individuals to collect and automatically code their lifetime job histories via a simple decision-tree model. Participants were asked to find each of their jobs by selecting appropriate job categories until they identified their job title, which was linked to a hidden 4-digit SOC code. For each occupation a job title in free text was also collected to estimate Cohen's kappa (κ) inter-rater agreement between SOC codes assigned by OSCAR and an expert manual coder. Results OSCAR was administered to 324 653 UK Biobank participants with an existing email address between June and September 2015. Complete 4-digit SOC-coded lifetime job histories were collected for 108 784 participants (response rate: 34%). Agreement between the 4-digit SOC codes assigned by OSCAR and the manual coder for a random sample of 400 job titles was moderately good [κ=0.45, 95% confidence interval (95% CI) 0.42-0.49], and improved when broader job categories were considered (κ=0.64, 95% CI 0.61-0.69 at a 1-digit SOC-code level). Conclusions OSCAR is a novel, efficient, and reasonably reliable web-based tool for collecting and automatically coding lifetime job histories in large population-based studies. Further application in other research projects for external validation purposes is warranted.
International Nuclear Information System (INIS)
Endo, T.; Sato, S.; Yamamoto, A.
2012-01-01
Average burnup of damaged fuels loaded in Fukushima Dai-ichi reactors is estimated, using the 134 Cs/ 137 Cs ratio method for measured radioactivities of 134 Cs and 137 Cs in contaminated soils within the range of 100 km from the Fukushima Dai-ichi nuclear power plants. As a result, the measured 134 Cs/ 137 Cs ratio from the contaminated soil is 0.996±0.07 as of March 11, 2011. Based on the 134 Cs/ 137 Cs ratio method, the estimated burnup of damaged fuels is approximately 17.2±1.5 [GWd/tHM]. It is noted that the numerical results of various calculation codes (SRAC2006/PIJ, SCALE6.0/TRITON, and MVP-BURN) are almost the same evaluation values of 134 Cs/ 137 Cs ratio with same evaluated nuclear data library (ENDF-B/VII.0). The void fraction effect in depletion calculation has a major impact on 134 Cs/ 137 Cs ratio compared with the differences between JENDL-4.0 and ENDF-B/VII.0. (authors)
Smoothing-Based Relative Navigation and Coded Aperture Imaging
Saenz-Otero, Alvar; Liebe, Carl Christian; Hunter, Roger C.; Baker, Christopher
2017-01-01
This project will develop an efficient smoothing software for incremental estimation of the relative poses and velocities between multiple, small spacecraft in a formation, and a small, long range depth sensor based on coded aperture imaging that is capable of identifying other spacecraft in the formation. The smoothing algorithm will obtain the maximum a posteriori estimate of the relative poses between the spacecraft by using all available sensor information in the spacecraft formation.This algorithm will be portable between different satellite platforms that possess different sensor suites and computational capabilities, and will be adaptable in the case that one or more satellites in the formation become inoperable. It will obtain a solution that will approach an exact solution, as opposed to one with linearization approximation that is typical of filtering algorithms. Thus, the algorithms developed and demonstrated as part of this program will enhance the applicability of small spacecraft to multi-platform operations, such as precisely aligned constellations and fractionated satellite systems.
Supporting Situated Learning Based on QR Codes with Etiquetar App: A Pilot Study
Camacho, Miguel Olmedo; Pérez-Sanagustín, Mar; Alario-Hoyos, Carlos; Soldani, Xavier; Kloos, Carlos Delgado; Sayago, Sergio
2014-01-01
EtiquetAR is an authoring tool for supporting the design and enactment of situated learning experiences based on QR tags. Practitioners use etiquetAR for creating, managing and personalizing collections of QR codes with special properties: (1) codes can have more than one link pointing at different multimedia resources, (2) codes can be updated…
Investigation and basic evaluation for ultra-high burnup fuel cladding material
International Nuclear Information System (INIS)
Ioka, Ikuo; Nagase, Fumihisa; Futakawa, Masatoshi; Kiuchi, Kiyoshi
2001-03-01
In ultra-high burnup of the power reactor, it is an essential problem to develop the cladding with excellent durability. First, development history and approach of the safety assessment of Zircaloy for the high burnup fuel were summarized in the report. Second, the basic evaluation and investigation were carried out on the material with high practicability in order to select the candidate materials for the ultra-high burnup fuel. In addition, the basic research on modification technology of the cladding surface was carried out from the viewpoint of the addition of safety margin as a cladding. From the development history of the zirconium alloy including the Zircaloy, it is hard to estimate the results of in-pile test from those of the conventional corrosion test (out-pile test). Therefore, the development of the new testing technology that can simulate the actual environment and the elucidation of the corrosion-controlling factor of the cladding are desired. In cases of RIA (Reactivity Initiated Accident) and LOCA (Loss of Coolant Accident), it seems that the loss of ductility in zirconium alloys under heavy irradiation and boiling of high temperature water restricts the extension of fuel burnup. From preliminary evaluation on the high corrosion-resistance materials (austenitic stainless steel, iron or nickel base superalloys, titanium alloy, niobium alloy, vanadium alloy and ferritic stainless steel), stabilized austenitic stainless steels with a capability of future improvement and high-purity niobium alloys with a expectation of the good corrosion resistance were selected as candidate materials of ultra-high burnup cladding. (author)
Burnup performance of OTTO cycle pebble bed reactors with ROX fuel
International Nuclear Information System (INIS)
Ho, Hai Quan; Obara, Toru
2015-01-01
Highlights: • A 300 MW t Small Pebble Bed Reactor with Rock-like oxide fuel is proposed. • Using ROX fuel can achieve high discharged burnup of spent fuel. • High geological stability can be expected in direct disposal of the spent ROX fuel. • The Pebble Bed Reactor with ROX fuel can be critical at steady state operation. • All the reactor designs have a negative temperature coefficient. - Abstract: A pebble bed high-temperature gas-cooled reactor (PBR) with rock-like oxide (ROX) fuel was designed to achieve high discharged burnup and improve the integrity of the spent fuel in geological disposal. The MCPBR code with a JENDL-4.0 library, which developed the analysis of the Once-Through-Then-Out (OTTO) cycle in PBR, was used to perform the criticality and burnup analysis. Burnup calculations for eight cases were carried out for both ROX fuel and a UO 2 fuel reactor with different heavy-metal loading conditions. The effective multiplication factor of all cases approximately equalled unity in the equilibrium condition. The ROX fuel reactor showed lower FIFA than the UO 2 fuel reactor at the same heavy-metal loading, about 5–15%. However, the power peaking factor and maximum power per fuel ball in the ROX fuel core were lower than that of UO 2 fuel core. This effect makes it possible to compensate for the lower-FIFA disadvantage in a ROX fuel core. All reactor designs had a negative temperature coefficient that is needed for the passive safety features of a pebble bed reactor
MPEC-4 Based Intelligent Coding for Plume Video Sequences
National Research Council Canada - National Science Library
Park, Gean
1999-01-01
.... Among those technologies, MPEG-4 codec can multiplex a set of independently coded, arbitrarily shaped video objects and transmit it through either fixed or variable rate channels such as internet...
The commercial impact of burnup increase
International Nuclear Information System (INIS)
Fenzlein, C.; Schricker, W.
2002-01-01
Deregulation has a dramatic effect on competition in the electricity markets. This will lead to a continued pressure on the prices in virtually all areas of the nuclear fuel cycle and will encourage further optimization, technical and technological progress and innovations with respect to further cost reductions of power production. The permission of direct disposal, in Germany legally granted in 1994 as an alternative to the reprocessing path, made possible cost savings and has consequently resulted in a decline of reprocessing prices. In addition, suppliers as well as operators are making considerable efforts to reduce the disposal costs fraction by optimizing disposal technologies and concepts. The increase of discharge has essentially contributed to the reduction the disposal cost fraction. Compared to former scenarios, the economic potential of burn-up increase is decreasing
Energy Technology Data Exchange (ETDEWEB)
Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Araya, Fumimasa; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1999-12-01
As a part of conceptual investigation on advanced light water reactors for the future, a light water reactor with the high burn-up of 100 GWd/t, the long cycle operation of 3 years and the full MOX core is being studied, aiming at the improvement on economical aspects, the reduction of the spent fuel production, the utilization of Plutonium and so forth. The present report summarizes investigation on PWR-type reactors. The core with the increased moderation of the moderator-to-fuel volume ratio of 2.6 {approx} 3.0 has been proposed be such a core that accomplishes requirements mentioned above. Through the neutronic and the thermo-hydrodynamic evaluation, the performances of the core have been evaluated. Also, the safety designing is underway considering the reactor system with the passive safety features. (author)
GPU-based parallel clustered differential pulse code modulation
Wu, Jiaji; Li, Wenze; Kong, Wanqiu
2015-10-01
Hyperspectral remote sensing technology is widely used in marine remote sensing, geological exploration, atmospheric and environmental remote sensing. Owing to the rapid development of hyperspectral remote sensing technology, resolution of hyperspectral image has got a huge boost. Thus data size of hyperspectral image is becoming larger. In order to reduce their saving and transmission cost, lossless compression for hyperspectral image has become an important research topic. In recent years, large numbers of algorithms have been proposed to reduce the redundancy between different spectra. Among of them, the most classical and expansible algorithm is the Clustered Differential Pulse Code Modulation (CDPCM) algorithm. This algorithm contains three parts: first clusters all spectral lines, then trains linear predictors for each band. Secondly, use these predictors to predict pixels, and get the residual image by subtraction between original image and predicted image. Finally, encode the residual image. However, the process of calculating predictors is timecosting. In order to improve the processing speed, we propose a parallel C-DPCM based on CUDA (Compute Unified Device Architecture) with GPU. Recently, general-purpose computing based on GPUs has been greatly developed. The capacity of GPU improves rapidly by increasing the number of processing units and storage control units. CUDA is a parallel computing platform and programming model created by NVIDIA. It gives developers direct access to the virtual instruction set and memory of the parallel computational elements in GPUs. Our core idea is to achieve the calculation of predictors in parallel. By respectively adopting global memory, shared memory and register memory, we finally get a decent speedup.
Light a CANDLE. An innovative burnup strategy of nuclear reactors
International Nuclear Information System (INIS)
Sekimoto, Hiroshi
2005-11-01
CANDLE is a new burnup strategy for nuclear reactors, which stands for Constant Axial Shape of Neutron Flux, Nuclide Densities and Power Shape During Life of Energy Production. When this candle-like burnup strategy is adopted, although the fuel is fixed in a reactor core, the burning region moves, at a speed proportionate to the power output, along the direction of the core axis without changing the spatial distribution of the number density of the nuclides, neutron flux, and power density. Excess reactivity is not necessary for burnup and the shape of the power distribution and core characteristics do not change with the progress of burnup. It is not necessary to use control rods for the control of the burnup. This booklet described the concept of the CANDLE burnup strategy with basic explanations of excess neutrons and its specific application to a high-temperature gas-cooled reactor and a fast reactor with excellent neutron economy. Supplementary issues concerning the initial core and high burnup were also referred. (T. Tanaka)
Hardware-efficient bosonic quantum error-correcting codes based on symmetry operators
Niu, Murphy Yuezhen; Chuang, Isaac L.; Shapiro, Jeffrey H.
2018-03-01
We establish a symmetry-operator framework for designing quantum error-correcting (QEC) codes based on fundamental properties of the underlying system dynamics. Based on this framework, we propose three hardware-efficient bosonic QEC codes that are suitable for χ(2 )-interaction based quantum computation in multimode Fock bases: the χ(2 ) parity-check code, the χ(2 ) embedded error-correcting code, and the χ(2 ) binomial code. All of these QEC codes detect photon-loss or photon-gain errors by means of photon-number parity measurements, and then correct them via χ(2 ) Hamiltonian evolutions and linear-optics transformations. Our symmetry-operator framework provides a systematic procedure for finding QEC codes that are not stabilizer codes, and it enables convenient extension of a given encoding to higher-dimensional qudit bases. The χ(2 ) binomial code is of special interest because, with m ≤N identified from channel monitoring, it can correct m -photon-loss errors, or m -photon-gain errors, or (m -1 )th -order dephasing errors using logical qudits that are encoded in O (N ) photons. In comparison, other bosonic QEC codes require O (N2) photons to correct the same degree of bosonic errors. Such improved photon efficiency underscores the additional error-correction power that can be provided by channel monitoring. We develop quantum Hamming bounds for photon-loss errors in the code subspaces associated with the χ(2 ) parity-check code and the χ(2 ) embedded error-correcting code, and we prove that these codes saturate their respective bounds. Our χ(2 ) QEC codes exhibit hardware efficiency in that they address the principal error mechanisms and exploit the available physical interactions of the underlying hardware, thus reducing the physical resources required for implementing their encoding, decoding, and error-correction operations, and their universal encoded-basis gate sets.
The role of ORIGEN-S in the design of burnup credit spent fuel casks
International Nuclear Information System (INIS)
Brady, M.C.
1991-01-01
Current licensing practices for spent fuel pools, storage facilities, and transportation casks require a conservative ''fresh fuel assumption'' be used in the criticality analysis. Burnup credit refers to a new approach in criticality analyses for spent fuel handling systems in which reactivity credit is allowed for the depleted state of the fuel. Studies have shown that the increased cask capacities that can be achieved with burnup credit offer both economic and risk incentives. The US Department of Energy is currently sponsoring a program to develop analysis methodologies and establish a new generation of spent fuel casks using the principle of burnup credit. The key difference in this new approach is the necessity to accurately predict the isotopic composition of the spent fuel. ORIGEN-S was selected to satisfy this requirement because of the flexibility and user-friendly input offered via its usage in the Standardized Computer Analyses for Licensing and Evaluation (SCALE) code system. Specifically, through the Shielding Analysis Sequence 2H (SAS2H), ORIGEN-S is linked with cross-section processing codes and one-dimensional transport analyses to produce problem-specific cross-section data for the point-depletion calculation. The utility code COUPLE facilitates updating basic cross-section and fission-yield data for the calculations. This paper describes the fundamental role fulfilled by ORIGEN-S in the development of the analysis methodology, validation of the methods, definition of criticality safety margins and other licensing considerations in the design of a new generation of spent fuel casks. Particular emphasis is given to the performance of ORIGEN-S in comparisons with measurements of irradiated fuel compositions and in predicting isotopics for use in the calculation of reactor restart critical configurations that are performed as a part of the validation process
A guide to introducing burnup credit, preliminary version (English translation)
International Nuclear Information System (INIS)
Okuno, Hiroshi; Suyama, Kenya; Ryufuku, Susumu
2017-06-01
There is an ongoing discussion on the application of burnup credit to the criticality safety controls of facilities that treat spent fuels. With regard to such application of burnup credit in Japan, this document summarizes the current technical status of the prediction of the isotopic composition and criticality of spent fuels, as well as safety evaluation concerns and the current status of legal affairs. This report is an English translation of A Guide to Introducing Burnup Credit, Preliminary Version, originally published in Japanese as JAERI-Tech 2001-055 by the Nuclear Fuel Cycle Facility Safety Research Committee. (author)
Safety aspects related to burnup increase and mixed oxide fuel
International Nuclear Information System (INIS)
Thomas, W.
1992-01-01
The dominant factor presently limiting the fuel burnup is the response of the cladding hulls. To maintain the excellent record of very low fuel failure rates for increased burnups further technical development is underway and necessary. In the nuclear fuel cycle increased burnups lead to a remarkable reduction of spent fuel arisings and corresponding economic savings. Thermal recycling of plutonium presently provides an opportunity to reduce the rising accumulation of plutunium in a situation where there is no demand for this fissile material in Fast Breeder Reactors. (orig.) [de
BNFL assessment of methods of attaining high burnup MOX fuel
International Nuclear Information System (INIS)
Brown, C.; Hesketh, K.W.; Palmer, I.D.
1998-01-01
It is clear that in order to maintain competitiveness with UO 2 fuel, the burnups achievable in MOX fuel must be enhanced beyond the levels attainable today. There are two aspects which require attention when studying methods of increased burnups - cladding integrity and fuel performance. Current irradiation experience indicates that one of the main performance issues for MOX fuel is fission gas retention. MOX, with its lower thermal conductivity, runs at higher temperatures than UO 2 fuel; this can result in enhanced fission gas release. This paper explores methods of effectively reducing gas release and thereby improving MOX burnup potential. (author)
A MATLAB based 3D modeling and inversion code for MT data
Singh, Arun; Dehiya, Rahul; Gupta, Pravin K.; Israil, M.
2017-07-01
The development of a MATLAB based computer code, AP3DMT, for modeling and inversion of 3D Magnetotelluric (MT) data is presented. The code comprises two independent components: grid generator code and modeling/inversion code. The grid generator code performs model discretization and acts as an interface by generating various I/O files. The inversion code performs core computations in modular form - forward modeling, data functionals, sensitivity computations and regularization. These modules can be readily extended to other similar inverse problems like Controlled-Source EM (CSEM). The modular structure of the code provides a framework useful for implementation of new applications and inversion algorithms. The use of MATLAB and its libraries makes it more compact and user friendly. The code has been validated on several published models. To demonstrate its versatility and capabilities the results of inversion for two complex models are presented.
Mode-dependent templates and scan order for H.264/AVC-based intra lossless coding.
Gu, Zhouye; Lin, Weisi; Lee, Bu-Sung; Lau, Chiew Tong; Sun, Ming-Ting
2012-09-01
In H.264/advanced video coding (AVC), lossless coding and lossy coding share the same entropy coding module. However, the entropy coders in the H.264/AVC standard were original designed for lossy video coding and do not yield adequate performance for lossless video coding. In this paper, we analyze the problem with the current lossless coding scheme and propose a mode-dependent template (MD-template) based method for intra lossless coding. By exploring the statistical redundancy of the prediction residual in the H.264/AVC intra prediction modes, more zero coefficients are generated. By designing a new scan order for each MD-template, the scanned coefficients sequence fits the H.264/AVC entropy coders better. A fast implementation algorithm is also designed. With little computation increase, experimental results confirm that the proposed fast algorithm achieves about 7.2% bit saving compared with the current H.264/AVC fidelity range extensions high profile.
Advanced Corrosion-Resistant Zr Alloys for High Burnup and Generation IV Applications
International Nuclear Information System (INIS)
Arthur Motta; Yong Hwan Jeong; R.J. Comstock; G.S. Was; Y.S. Kim
2006-01-01
The objective of this collaboration between four institutions in the US and Korea is to demonstrate a technical basis for the improvement of the corrosion resistance of zirconium-based alloys in more extreme operating environments (such as those present in severe fuel duty, cycles high burnup, boiling, aggressive chemistry) and to investigate the feasibility (from the point of view of corrosion rate) of using advanced zirconium-based alloys in a supercritical water environment
Evaluation of the Frequency for Gas Sampling for the High Burnup Confirmatory Data Project
Energy Technology Data Exchange (ETDEWEB)
Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Marschman, Steven C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-05-01
This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations.
Performance of FSO-OFDM based on BCH code
Directory of Open Access Journals (Sweden)
Jiao Xiao-lu
2016-01-01
Full Text Available As contrasted with the traditional OOK (on-off key system, FSO-OFDM system can resist the atmospheric scattering and improve the spectrum utilization rate effectively. Due to the instability of the atmospheric channel, the system will be affected by various factors, and resulting in a high BER. BCH code has a good error correcting ability, particularly in the short-length and medium-length code, and its performance is close to the theoretical value. It not only can check the burst errors but also can correct the random errors. Therefore, the BCH code is applied to the system to reduce the system BER. At last, the semi-physical simulation has been conducted with MATLAB. The simulation results show that when the BER is 10-2, the performance of OFDM is superior 4dB compared with OOK. In different weather conditions (extension rain, advection fog, dust days, when the BER is 10-5, the performance of BCH (255,191 channel coding is superior 4~5dB compared with uncoded system. All in all, OFDM technology and BCH code can reduce the system BER.
Directory of Open Access Journals (Sweden)
Shang-Chien Wu
2018-02-01
Full Text Available This study proposes a new method of analyzing the burnup credit in boiling water reactor spent fuel assemblies against various operating parameters. The operating parameters under investigation include fuel temperature, axial burnup profile, axial moderator density profile, and control blade usage. In particular, the effects of variations in one and two operating parameters on the curve of effective multiplication factor (keff versus burnup (B are, respectively, the so-called single and compound effects. All the calculations were performed using SCALE 6.1 together with the Evaluated Nuclear Data Files, part B (ENDF/B-VII238-neutron energy group data library. Furthermore, two geometrical models were established based on the General Electric (GE14 10 × 10 boiling water reactor fuel assembly and the Generic Burnup-Credit (GBC-68 storage cask. The results revealed that the curves of keff versus B, due to single and compound effects, can be approximated using a first degree polynomial of B. However, the reactivity deviation (or changes of keff,Δk in some compound effects was not a summation of the all Δk resulting from the two associated single effects. This phenomenon is undesirable because it may to some extent affect the precise assessment of burnup credit. In this study, a general formula was thus proposed to express the curves of keff versus B for both single and compound effects.
BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup
International Nuclear Information System (INIS)
1998-01-01
cross section variation or correlation on nuclide concentrations is provided, but a temperature dependence is coded. Steady state condition with continuous fueling is established by a global iterative scheme that applies the criticality search scheme in the neutronics and models fuel movement directly in the exposure code. Time-dependent sensitivity data applies the forward march, reverse importance approach. The codes do not process data from the user input data stream allowing flexible task assignment along selected calculational paths. Multigroup cross section data are produced locally using the PSR-0063/AMPX II or CCC-0450/SCALE-2 code systems to produce resonance shielding (NITAWL) and cell weighted (XSDRN) microscopic cross sections. Locally, each code is compiled and loaded, and only one version is maintained in a quality assurance state in load module form. An on-line catalog procedure, installed with system support, provides job control instructions with nominal default of space allocation to files. Executing the catalog procedure makes the driver memory resident. The first user input data line must be the control module name used for the run. VENTURE-PC: The VENTURE module applies the finite-difference diffusion or a simple P1 approximation. VENTURE uses an outer-inner iteration scheme with several different data handling methods. Over-relaxation is applied to the inner and outer iterations, and succeeding flux iterates may be accelerated with the Chebychev process. - The BURNER code (module EXPOSURE) uses a difference formulation based on average generation rates; or a matrix exponential formulation to approximate the solution of the coupled burnup differential equations; or an explicit solution for simply coupled nuclide chains. Space dependence is included by working with zone averaged fluxes
Depth-based Multi-View 3D Video Coding
DEFF Research Database (Denmark)
Zamarin, Marco
-view video. Depth maps are typically used to synthesize the desired output views, and the performance of view synthesis algorithms strongly depends on the accuracy of depth information. In this thesis, novel algorithms for efficient depth map compression in MVD scenarios are proposed, with particular focus...... on edge-preserving solutions. In a proposed scheme, texture-depth correlation is exploited to predict surface shapes in the depth signal. In this way depth coding performance can be improved in terms of both compression gain and edge-preservation. Another solution proposes a new intra coding mode targeted...... to depth blocks featuring arbitrarily-shaped edges. Edge information is encoded exploiting previously coded edge blocks. Integrated in H.264/AVC, the proposed mode allows significant bit rate savings compared with a number of state-of-the-art depth codecs. View synthesis performances are also improved...
International Nuclear Information System (INIS)
Sawa, Kazuhiro; Fukuda, Kousaku; Acharya, R.
1995-01-01
This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for an irradiation test in a removable beryllium position of the High Flux Isotope Reactor(HFIR) at Oak Ridge National Laboratory. This test is being carried out under Annex 2 of the Arrangement between the U.S. Department of Energy and the Japan Atomic Energy Research Institute on Cooperation in Research and Development regarding High-Temperature Gas-cooled Reactors. The fuel used in the test is an advanced type. The advanced fuel was designed aiming at burnup of about 10%FIMA(% fissions per initial metallic atom) which was higher than that of the first charge fuel for the High Temperature Engineering Test Reactor(HTTR) and was produced in Japan. CACA-2, a heavy isotope and fission product concentration calculational code for experimental irradiation capsules, was used to determine time-dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries(HEATING) code was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body, which contains the fuel compacts, and of the primary pressure vessel were determined such that the requirements of running the fuel compacts at an average temperature less than 1250degC and of not exceeding a maximum fuel temperature of 1350degC were met throughout the four cycles of irradiation. The detail design of the capsule was carried out based on this analysis. (author)
3D Scan-Based Wavelet Transform and Quality Control for Video Coding
Directory of Open Access Journals (Sweden)
Parisot Christophe
2003-01-01
Full Text Available Wavelet coding has been shown to achieve better compression than DCT coding and moreover allows scalability. 2D DWT can be easily extended to 3D and thus applied to video coding. However, 3D subband coding of video suffers from two drawbacks. The first is the amount of memory required for coding large 3D blocks; the second is the lack of temporal quality due to the sequence temporal splitting. In fact, 3D block-based video coders produce jerks. They appear at blocks temporal borders during video playback. In this paper, we propose a new temporal scan-based wavelet transform method for video coding combining the advantages of wavelet coding (performance, scalability with acceptable reduced memory requirements, no additional CPU complexity, and avoiding jerks. We also propose an efficient quality allocation procedure to ensure a constant quality over time.
Nieh, Ta-Chun; Yang, Chao-Chin; Huang, Jen-Fa
2011-08-01
A complete complementary/prime/shifted prime (CPS) code family for the optical code-division multiple-access (OCDMA) system is proposed. Based on the ability of complete complementary (CC) code, the multiple-access interference (MAI) can be suppressed and eliminated via spectral amplitude coding (SAC) OCDMA system under asynchronous/synchronous transmission. By utilizing the shifted prime (SP) code in the SAC scheme, the hardware implementation of encoder/decoder can be simplified with a reduced number of optical components, such as arrayed waveguide grating (AWG) and fiber Bragg grating (FBG). This system has a superior performance as compared to previous bipolar-bipolar coding OCDMA systems.
Flowgen: Flowchart-based documentation for C + + codes
Kosower, David A.; Lopez-Villarejo, J. J.
2015-11-01
We present the Flowgen tool, which generates flowcharts from annotated C + + source code. The tool generates a set of interconnected high-level UML activity diagrams, one for each function or method in the C + + sources. It provides a simple and visual overview of complex implementations of numerical algorithms. Flowgen is complementary to the widely-used Doxygen documentation tool. The ultimate aim is to render complex C + + computer codes accessible, and to enhance collaboration between programmers and algorithm or science specialists. We describe the tool and a proof-of-concept application to the VINCIA plug-in for simulating collisions at CERN's Large Hadron Collider.
Experimental data bases useful for quantification of model uncertainties in best estimate codes
International Nuclear Information System (INIS)
Wilson, G.E.; Katsma, K.R.; Jacobson, J.L.; Boodry, K.S.
1988-01-01
A data base is necessary for assessment of thermal hydraulic codes within the context of the new NRC ECCS Rule. Separate effect tests examine particular phenomena that may be used to develop and/or verify models and constitutive relationships in the code. Integral tests are used to demonstrate the capability of codes to model global characteristics and sequence of events for real or hypothetical transients. The nuclear industry has developed a large experimental data base of fundamental nuclear, thermal-hydraulic phenomena for code validation. Given a particular scenario, and recognizing the scenario's important phenomena, selected information from this data base may be used to demonstrate applicability of a particular code to simulate the scenario and to determine code model uncertainties. LBLOCA experimental data bases useful to this objective are identified in this paper. 2 tabs
Annual report of nuclear code evaluation committee for fiscal 2000 year
Energy Technology Data Exchange (ETDEWEB)
NONE
2002-03-01
In this report, research results discussed in fiscal 2000 year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. In 2000, papers mainly on the three topics of (1) present status of burnup credit evaluation methods, (2) issues concerning convergence of criticality calculation and (3) estimation methods for errors associated with criticality calculation based on nuclear data covariance file, are presented and discussed. These results are sorted to grasp the present status of related technology and described in this report. (author)
Isotopic Bias and Uncertainty for Burnup Credit Applications
International Nuclear Information System (INIS)
J.M. Scaglione
2002-01-01
The application of burnup credit requires calculating the isotopic inventory of the irradiated fuel. The depletion calculation simulates the burnup of the fuel under reactor operating conditions. The result of the depletion analysis is the predicted isotopic composition, which is ultimately input to a criticality analysis to determine the system multiplication factor (k eff ). This paper demonstrates an approach for calculating the isotopic bias and uncertainty in k eff for commercial spent nuclear fuel burnup credit. This paper covers 74 different radiochemical assayed spent fuel samples from 22 different fuel assemblies that were irradiated in eight different pressurized water reactors (PWRs). The samples evaluated span an enrichment range of 2.556 wt% U-235 through 4.67 wt% U-235, and burnups from 6.92 GWd/MTU through 55.7 GWd/MTU
Impact of extended burnup on the nuclear fuel cycle
International Nuclear Information System (INIS)
1993-04-01
The Advisory Group Meeting was held in Vienna from 2 to 5 December 1991, to review, analyse, and discuss the effects of burnup extension in both light and heavy water reactors on all aspects of the fuel cycle. Twenty experts from thirteen countries participated in this meeting. There was consensus that both economic and environmental benefits are driving forces toward the achievement of higher burnups and that the present trend of burnup extension may be expected to continue. The extended burnup has been considered for the three main stages of the fuel cycle: the front end, in-reactor issues and the back end. Thirteen papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs
Parametric neutronic analyses related to burnup credit cask design
International Nuclear Information System (INIS)
Parks, C.V.
1989-01-01
The consideration of spent fuel histories (burnup credit) in the design of spent fuel shipping casks will result in cost savings and public risk benefits in the overall fuel transportation system. The purpose of this paper is to describe the depletion and criticality analyses performed in conjunction with and supplemental to the referenced analysis. Specifically, the objectives are to indicate trends in spent fuel isotopic composition with burnup and decay time; provide spent fuel pin lattice values as a function of burnup, decay time, and initial enrichment; demonstrate the variation of k eff for infinite arrays of spent fuel assemblies separated by generic cask basket designs (borated and unborated) of varying thicknesses; and verify the potential cask reactivity margin available with burnup credit via analysis with generic cask models
A novel chaotic encryption scheme based on arithmetic coding
International Nuclear Information System (INIS)
Mi Bo; Liao Xiaofeng; Chen Yong
2008-01-01
In this paper, under the combination of arithmetic coding and logistic map, a novel chaotic encryption scheme is presented. The plaintexts are encrypted and compressed by using an arithmetic coder whose mapping intervals are changed irregularly according to a keystream derived from chaotic map and plaintext. Performance and security of the scheme are also studied experimentally and theoretically in detail
Language-Based Security for Malicious Mobile Code
2007-09-30
Algol), and the Berkeley SDS-940 system employed object-code rewriting as part of its system pro- filer. More recently, the SPIN [5], Vino [52, 47], and...1993. [52] E. Yasuhiro, J. Gwertzman, M. Seltzer, C. Small, Keith A. Smith, and D. Tang. VINO : The 1994 fall harvest. Technical Report TR-34-94
Two-Factor Authentication System based on QR-Codes
Directory of Open Access Journals (Sweden)
Andrey Yunusovich Iskhakov
2014-09-01
Full Text Available The opportunity of two-factor authentication usage in the control systems and access management on the basis of Quick Response codes with one-time passwords is analyzed in the work. The mobile application is proposed to use as a software token.
A burn-up module coupling to an AMPX system
International Nuclear Information System (INIS)
Salvatore Duque, M.; Gomez, S.E.; Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.
1990-01-01
The Reactors and Neutrons Division of the Bariloche Atomic Center uses the AMPX system for the study of high conversion reactors (HCR). Such system allows to make neutronic calculations from the nuclear data library (ENDF/B-IV). The Nuclear Engineering career of the Balseiro Institute developed and implemented a burn-up module at a μ-cell level (BUM: Burn-up Module) which agrees with the requirement to be coupled to the AMPX system. (Author) [es
Burnup determination of a fuel element concerning different cooling times
International Nuclear Information System (INIS)
Henriquez, C.; Navarro, G.; Pereda, C.; Mutis, O.; Terremoto, Luis A.A.; Zeituni, Carlos A.
2002-01-01
In this work we report a complete set of measurements and some relevant results regarding the burnup process of a fuel element containing low enriched nuclear fuel. This fuel element was fabricated at the Plant of Fuel Elements of the Chilean Nuclear Energy Commission (CCHEN). Measurements were carried out using gamma-ray spectroscopy and the absolute burnup of the fuel element was determined. (author)
Calculation routes to determine burnup credit loading curves
International Nuclear Information System (INIS)
Neuber, J.S.
2007-01-01
The objective of the paper on hand is to describe the key steps of the calculation routes used for evaluating burnup credit loading curves and to discuss procedures which are adequate to estimate the biases and variances in the calculation routes. In addition, impacts of the formulation of bounding or conservative approaches on the estimates of these biases and variances as well as on the reactivity effects due to the non-uniformity of the burnup distribution within the fuel are discussed. (author)
A guide introducing burnup credit, preliminary version. Contract research
International Nuclear Information System (INIS)
2001-07-01
It is examined to take burnup credit into account for criticality safety control of facility treating spent fuel. This work is a collection of current technical status of predicting isotopic composition and criticality of spent fuel, points to be specially considered for safety evaluation, and current status of legal affairs for the purpose of applying burnup credit to the criticality safety evaluation of the facility treating spent fuel in Japan. (author)
Draft evaluation of the frequency for gas sampling for the high burnup confirmatory data project
Energy Technology Data Exchange (ETDEWEB)
Stockman, Christine T. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Alsaed, Halim A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2015-03-26
This report fulfills the M3 milestone M3FT-15SN0802041, “Draft Evaluation of the Frequency for Gas Sampling for the High Burn-up Storage Demonstration Project” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. This report provides a technically based gas sampling frequency strategy for the High Burnup (HBU) Confirmatory Data Project. The evaluation of: 1) the types and magnitudes of gases that could be present in the project cask and, 2) the degradation mechanisms that could change gas compositions culminates in an adaptive gas sampling frequency strategy. This adaptive strategy is compared against the sampling frequency that has been developed based on operational considerations. Gas sampling will provide information on the presence of residual water (and byproducts associated with its reactions and decomposition) and breach of cladding, which could inform the decision of when to open the project cask.
Trellises and Trellis-Based Decoding Algorithms for Linear Block Codes
Lin, Shu
1998-01-01
A code trellis is a graphical representation of a code, block or convolutional, in which every path represents a codeword (or a code sequence for a convolutional code). This representation makes it possible to implement Maximum Likelihood Decoding (MLD) of a code with reduced decoding complexity. The most well known trellis-based MLD algorithm is the Viterbi algorithm. The trellis representation was first introduced and used for convolutional codes [23]. This representation, together with the Viterbi decoding algorithm, has resulted in a wide range of applications of convolutional codes for error control in digital communications over the last two decades. There are two major reasons for this inactive period of research in this area. First, most coding theorists at that time believed that block codes did not have simple trellis structure like convolutional codes and maximum likelihood decoding of linear block codes using the Viterbi algorithm was practically impossible, except for very short block codes. Second, since almost all of the linear block codes are constructed algebraically or based on finite geometries, it was the belief of many coding theorists that algebraic decoding was the only way to decode these codes. These two reasons seriously hindered the development of efficient soft-decision decoding methods for linear block codes and their applications to error control in digital communications. This led to a general belief that block codes are inferior to convolutional codes and hence, that they were not useful. Chapter 2 gives a brief review of linear block codes. The goal is to provide the essential background material for the development of trellis structure and trellis-based decoding algorithms for linear block codes in the later chapters. Chapters 3 through 6 present the fundamental concepts, finite-state machine model, state space formulation, basic structural properties, state labeling, construction procedures, complexity, minimality, and
Modelling the high burnup UO2 structure in LWR fuel
International Nuclear Information System (INIS)
Lassmann, K.; Walker, C.T.; Laar, J. van de; Lindstroem, F.
1995-01-01
The concept of a burnup threshold for the formation of the high burnup UO 2 structure (HBS) is supported by experimental data, which also reveal that a transition zone exists between the normal UO 2 structure and the fully developed HBS. From the analysis of radial xenon profiles measured by EPMA a threshold burnup is obtained in the range 60-75 GW d/t U. The lower value is considered to be the threshold for the onset of the HBS and the higher value the threshold for the fully developed HBS. Xenon depletion in the transition zone and the fully developed HBS can be described by a simple model. At local burnups above 120 GW d/t U the xenon generated is in equilibrium with the xenon lost to the fission gas pores and the concentration does not fall below 0.25 wt%. The TRANSURANUS burnup model TUBRNP predicts reasonably well the penetration of the HBS and the associated xenon depletion up to a cross section average burnup of approximately 70 GW d/t U. (orig.)
Galactic Cosmic Ray Event-Based Risk Model (GERM) Code
Cucinotta, Francis A.; Plante, Ianik; Ponomarev, Artem L.; Kim, Myung-Hee Y.
2013-01-01
This software describes the transport and energy deposition of the passage of galactic cosmic rays in astronaut tissues during space travel, or heavy ion beams in patients in cancer therapy. Space radiation risk is a probability distribution, and time-dependent biological events must be accounted for physical description of space radiation transport in tissues and cells. A stochastic model can calculate the probability density directly without unverified assumptions about shape of probability density function. The prior art of transport codes calculates the average flux and dose of particles behind spacecraft and tissue shielding. Because of the signaling times for activation and relaxation in the cell and tissue, transport code must describe temporal and microspatial density of functions to correlate DNA and oxidative damage with non-targeted effects of signals, bystander, etc. These are absolutely ignored or impossible in the prior art. The GERM code provides scientists data interpretation of experiments; modeling of beam line, shielding of target samples, and sample holders; and estimation of basic physical and biological outputs of their experiments. For mono-energetic ion beams, basic physical and biological properties are calculated for a selected ion type, such as kinetic energy, mass, charge number, absorbed dose, or fluence. Evaluated quantities are linear energy transfer (LET), range (R), absorption and fragmentation cross-sections, and the probability of nuclear interactions after 1 or 5 cm of water equivalent material. In addition, a set of biophysical properties is evaluated, such as the Poisson distribution for a specified cellular area, cell survival curves, and DNA damage yields per cell. Also, the GERM code calculates the radiation transport of the beam line for either a fixed number of user-specified depths or at multiple positions along the Bragg curve of the particle in a selected material. The GERM code makes the numerical estimates of basic
3-D shape measurement based on complementary Gray-code light
Zhang, Qican; Su, Xianyu; Xiang, Liqun; Sun, Xuezhen
2012-04-01
A combination of phase-shift with Gray-code light projection into a three-dimensional (3-D) measurement system has been exploited to digitalize 3-D shape information of a tested object, even with a discontinuous surface. Unfortunately, the phase unwrapping will fall into an error, when an improper value of Gray coding is caused by mistake at the partial boundary of two adjacent binary words. To this end, a new complementary Gray-code method is proposed in this paper as well as the corresponding phase-unwrapping method. This problem of phase unwrapping could be cleverly solved by projecting an additional Gray-code pattern to extend this code and using the different and complementary boundary locations of the traditional and additional codes. The results of computer simulation and experiment confirm that this proposed method based on complementary Gray-code can reliably reconstruct the nature phase distribution of the tested object with only one extra fringe pattern.
Propagation of Statistical and Nuclear Data Uncertainties in Monte-Carlo Burn-up Calculations
García Herranz, Nuria; Cabellos de Francisco, Oscar Luis; Sanz Gonzalo, Javier; Juan Ruiz, Jesús; Kuijper, Jim C.
2008-01-01
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP–ACAB system, which comb...
An empirical formulation to describe the evolution of the high burnup structure
Lemes, Martín; Soba, Alejandro; Denis, Alicia
2015-01-01
In the present work the behavior of fuel pellets for LWR power reactors in the high burnup range (average burnup higher than about 45 MWd/kgU) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup. Gradually, a new microstructure develops in that ring, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behavior. It is generally accepted that the evolution of porosity in the high burnup structure (HBS) is determinant of the retention capacity of the fission gases rejected from the fuel matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Although the mechanisms governing the microstructural transformation have not been completely elucidated yet, some empirical expressions can be given, and this is the intention of the present work, for representing the main physical parameters. Starting from several works published in the open literature, some mathematical expressions were developed to describe the behavior and progress of porosity at local burnup values ranging from 60 to 300 MWd/kgU. The analysis includes the interactions of different orders between pores, the growth of the pore radius by capturing vacancies, the evolution of porosity, pore number density and overpressure within the closed pores, the inventory of fission gas dissolved in the matrix and retained in the pores. The model is mathematically expressed by a system of non-linear differential equations. In the present work, results of this calculation scheme are compared with experimental data available in
Security Concerns and Countermeasures in Network Coding Based Communications Systems
DEFF Research Database (Denmark)
Talooki, Vahid; Bassoli, Riccardo; Roetter, Daniel Enrique Lucani
2015-01-01
This survey paper shows the state of the art in security mechanisms, where a deep review of the current research and the status of this topic is carried out. We start by introducing network coding and its variety applications in enhancing current traditional networks. In particular, we analyze two...... key protocol types, namely, state-aware and stateless protocols, specifying the benefits and disadvantages of each one of them. We also present the key security assumptions of network coding (NC) systems as well as a detailed analysis of the security goals and threats, both passive and active....... This paper also presents a detailed taxonomy and a timeline of the different NC security mechanisms and schemes reported in the literature. Current proposed security mechanisms and schemes for NC in the literature are classified later. Finally a timeline of these mechanism and schemes is presented....
Directory of Open Access Journals (Sweden)
Isaac Caicedo-Castro
2014-01-01
Full Text Available This paper presents CodeRAnts, a new recommendation method based on a collaborative searching technique and inspired on the ant colony metaphor. This method aims to fill the gap in the current state of the matter regarding recommender systems for software reuse, for which prior works present two problems. The first is that, recommender systems based on these works cannot learn from the collaboration of programmers and second, outcomes of assessments carried out on these systems present low precision measures and recall and in some of these systems, these metrics have not been evaluated. The work presented in this paper contributes a recommendation method, which solves these problems.
QR code based noise-free optical encryption and decryption of a gray scale image
Jiao, Shuming; Zou, Wenbin; Li, Xia
2017-03-01
In optical encryption systems, speckle noise is one major challenge in obtaining high quality decrypted images. This problem can be addressed by employing a QR code based noise-free scheme. Previous works have been conducted for optically encrypting a few characters or a short expression employing QR codes. This paper proposes a practical scheme for optically encrypting and decrypting a gray-scale image based on QR codes for the first time. The proposed scheme is compatible with common QR code generators and readers. Numerical simulation results reveal the proposed method can encrypt and decrypt an input image correctly.
Directory of Open Access Journals (Sweden)
Ricky Akbar
2017-09-01
Full Text Available Perusahaan Listrik Negara (PLN has implemented a smart electricity system or prepaid electricity. The customers pay the electricity voucher first before use the electricity. The token contained in electricity voucher that has been purchased by the customer is inserted into the Meter Prabayar (MPB installed in the location of customers. When a customer purchases a voucher, it will get a receipt that contains all of the customer's identity and the 20-digit of voucher code (token to be entered into MPB as a substitute for electrical energy credit. Receipts obtained by the customer is certainly vulnerable to loss, or hijacked by unresponsible parties. In this study, authors designed and develop an android based application by utilizing QR code technology as a replacement for the receipt of prepaid electricity credit which contains the identity of the customer and the 20-digit voucher code. The application is developed by implemented waterfall methodology. The implementation process of the waterfall methods used, are (1 analysis of functional requirement of the system by conducting a preliminary study and data collection based on field studies and literature, (2 system design by using UML diagrams and Business Process Model Notation (BPMN and Entity Relationship diagram (ERD, (3 design implementation by using OOP (Object Oriented programming technique. Web application is developed by using laravel PHP framework and database MySQL while mobile application is developed by using B4A (4 developed system is tested by using blackbox method testing. Final result of this research is a Web and mobile applications for the sale of electricityvoucher by QR Code technology.
Resistor-logic demultiplexers for nanoelectronics based on constant-weight codes.
Kuekes, Philip J; Robinett, Warren; Roth, Ron M; Seroussi, Gadiel; Snider, Gregory S; Stanley Williams, R
2006-02-28
The voltage margin of a resistor-logic demultiplexer can be improved significantly by basing its connection pattern on a constant-weight code. Each distinct code determines a unique demultiplexer, and therefore a large family of circuits is defined. We consider using these demultiplexers for building nanoscale crossbar memories, and determine the voltage margin of the memory system based on a particular code. We determine a purely code-theoretic criterion for selecting codes that will yield memories with large voltage margins, which is to minimize the ratio of the maximum to the minimum Hamming distance between distinct codewords. For the specific example of a 64 × 64 crossbar, we discuss what codes provide optimal performance for a memory.
Energy Technology Data Exchange (ETDEWEB)
Panka, Istvan; Hegyi, Gyoergy; Maraczy, Csaba; Temesvari, Emese [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.
2017-11-15
The best-estimate KARATE code system has been widely used for core design calculations and simulations of slow transients of VVER reactors. Recently there has been an increasing need for assessing the uncertainties of such calculations by propagating the basic input uncertainties of the models through the full calculation chain. In order to determine the uncertainties of quantities of interest during the burnup, the statistical version of the KARATE code system has been elaborated. In the first part of the paper, the main features of the new code system are discussed. The applied statistical method is based on Monte-Carlo sampling of the considered input data taking into account mainly the covariance matrices of the cross sections and/or the technological uncertainties. In the second part of the paper, only the uncertainties of cross sections are considered and an equilibrium cycle related to a VVER-440 type reactor is investigated. The burnup dependence of the uncertainties of some safety related parameters (e.g. critical boron concentration, rod worth, feedback coefficients, assembly-wise radial power and burnup distribution) are discussed and compared to the recently used limits.
Energy Technology Data Exchange (ETDEWEB)
Liang, Jingang; Wang, Kan; Qiu, Yishu [Dept. of Engineering Physics, LiuQing Building, Tsinghua University, Beijing (China); Chai, Xiao Ming; Qiang, Sheng Long [Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-06-15
Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC) codes in accomplishing pin-wise three-dimensional (3D) full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC) code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Energy Technology Data Exchange (ETDEWEB)
Wagner, J.C.
2001-08-02
This report proposes and documents a computational benchmark problem for the estimation of the additional reactivity margin available in spent nuclear fuel (SNF) from fission products and minor actinides in a burnup-credit storage/transport environment, relative to SNF compositions containing only the major actinides. The benchmark problem/configuration is a generic burnup credit cask designed to hold 32 pressurized water reactor (PWR) assemblies. The purpose of this computational benchmark is to provide a reference configuration for the estimation of the additional reactivity margin, which is encouraged in the U.S. Nuclear Regulatory Commission (NRC) guidance for partial burnup credit (ISG8), and document reference estimations of the additional reactivity margin as a function of initial enrichment, burnup, and cooling time. Consequently, the geometry and material specifications are provided in sufficient detail to enable independent evaluations. Estimates of additional reactivity margin for this reference configuration may be compared to those of similar burnup-credit casks to provide an indication of the validity of design-specific estimates of fission-product margin. The reference solutions were generated with the SAS2H-depletion and CSAS25-criticality sequences of the SCALE 4.4a package. Although the SAS2H and CSAS25 sequences have been extensively validated elsewhere, the reference solutions are not directly or indirectly based on experimental results. Consequently, this computational benchmark cannot be used to satisfy the ANS 8.1 requirements for validation of calculational methods and is not intended to be used to establish biases for burnup credit analyses.
International Nuclear Information System (INIS)
Eberle, R.; Heins, L.; Sontheimer, F.
1997-01-01
The proof that fuel rods will safely withstand all loads arising from inpile service conditions is generally achieved through the assessment of a number of design criteria by using a conservative analysis methodology in conjunction with design limits ''on the safe side''. The classical approach is the application of a fuel rod code to the Worst Case which is defined by the combination of most unfavorable conditions and assumptions with respect to the criterion under consideration. As it is evident that the deterministic construction of such Worst Cases imply an (unknown but) intuitively very high degree of conservatism, it is not surprising that this will develop to cause problems the more demanding fuel insertion conditions have to be anticipated (increased burnup, high efficiency loading schemes, etc.). A certain relief can be gained form cautious revisions of single design limits based on grown performance experience. But this increase of knowledge allows as well to change the established deterministic ''go/no-go'' conception into a better differentiating assessment methodology by which the quantification of the implied conservatism and the remaining design margins is possible: the Probabilistic Design Methodology (PDM). Principles and elements of the PDM are described. An essential prerequisite is a best-estimate fuel rod code which incorporates the latest state of knowledge about potential performance limiting phenomena (e.g. burnup degradation of fuel oxide thermal conductivity) as Siemens/KWU's CARO-E does. An example is given how input distributions for rod data and model parameters transfer into a frequency distribution of maximum rod internal pressure, and indications are given how this is to be interpreted in view of a probabilistically re-formulated design criterion. The PDM provides a realistic conservative assessment of design criteria and will thus recover design margins for increasingly aggravated loading conditions. (author). 9 refs, 9 figs, 2 tabs
Zhang, Yun; Kwong, Sam; Wang, Xu; Yuan, Hui; Pan, Zhaoqing; Xu, Long
2015-07-01
In this paper, we propose a machine learning-based fast coding unit (CU) depth decision method for High Efficiency Video Coding (HEVC), which optimizes the complexity allocation at CU level with given rate-distortion (RD) cost constraints. First, we analyze quad-tree CU depth decision process in HEVC and model it as a three-level of hierarchical binary decision problem. Second, a flexible CU depth decision structure is presented, which allows the performances of each CU depth decision be smoothly transferred between the coding complexity and RD performance. Then, a three-output joint classifier consists of multiple binary classifiers with different parameters is designed to control the risk of false prediction. Finally, a sophisticated RD-complexity model is derived to determine the optimal parameters for the joint classifier, which is capable of minimizing the complexity in each CU depth at given RD degradation constraints. Comparative experiments over various sequences show that the proposed CU depth decision algorithm can reduce the computational complexity from 28.82% to 70.93%, and 51.45% on average when compared with the original HEVC test model. The Bjøntegaard delta peak signal-to-noise ratio and Bjøntegaard delta bit rate are -0.061 dB and 1.98% on average, which is negligible. The overall performance of the proposed algorithm outperforms those of the state-of-the-art schemes.
Simonaitis, Linas; McDonald, Clement J
2009-10-01
The utility of National Drug Codes (NDCs) and drug knowledge bases (DKBs) in the organization of prescription records from multiple sources was studied. The master files of most pharmacy systems include NDCs and local codes to identify the products they dispense. We obtained a large sample of prescription records from seven different sources. These records carried a national product code or a local code that could be translated into a national product code via their formulary master. We obtained mapping tables from five DKBs. We measured the degree to which the DKB mapping tables covered the national product codes carried in or associated with the sample of prescription records. Considering the total prescription volume, DKBs covered 93.0-99.8% of the product codes from three outpatient sources and 77.4-97.0% of the product codes from four inpatient sources. Among the in-patient sources, invented codes explained 36-94% of the noncoverage. Outpatient pharmacy sources rarely invented codes, which comprised only 0.11-0.21% of their total prescription volume, compared with inpatient pharmacy sources for which invented codes comprised 1.7-7.4% of their prescription volume. The distribution of prescribed products was highly skewed, with 1.4-4.4% of codes accounting for 50% of the message volume and 10.7-34.5% accounting for 90% of the message volume. DKBs cover the product codes used by outpatient sources sufficiently well to permit automatic mapping. Changes in policies and standards could increase coverage of product codes used by inpatient sources.
Model-based coding of facial images based on facial muscle motion through isodensity maps
So, Ikken; Nakamura, Osamu; Minami, Toshi
1991-11-01
A model-based coding system has come under serious consideration for the next generation of image coding schemes, aimed at greater efficiency in TV telephone and TV conference systems. In this model-based coding system, the sender's model image is transmitted and stored at the receiving side before the start of the conversation. During the conversation, feature points are extracted from the facial image of the sender and are transmitted to the receiver. The facial expression of the sender facial is reconstructed from the feature points received and a wireframed model constructed at the receiving side. However, the conventional methods have the following problems: (1) Extreme changes of the gray level, such as in wrinkles caused by change of expression, cannot be reconstructed at the receiving side. (2) Extraction of stable feature points from facial images with irregular features such as spectacles or facial hair is very difficult. To cope with the first problem, a new algorithm based on isodensity lines which can represent detailed changes in expression by density correction has already been proposed and good results obtained. As for the second problem, we propose in this paper a new algorithm to reconstruct facial images by transmitting other feature points extracted from isodensity maps.
Quantum secret sharing based on quantum error-correcting codes
International Nuclear Information System (INIS)
Zhang Zu-Rong; Liu Wei-Tao; Li Cheng-Zu
2011-01-01
Quantum secret sharing(QSS) is a procedure of sharing classical information or quantum information by using quantum states. This paper presents how to use a [2k − 1, 1, k] quantum error-correcting code (QECC) to implement a quantum (k, 2k − 1) threshold scheme. It also takes advantage of classical enhancement of the [2k − 1, 1, k] QECC to establish a QSS scheme which can share classical information and quantum information simultaneously. Because information is encoded into QECC, these schemes can prevent intercept-resend attacks and be implemented on some noisy channels. (general)
High burnup fuel onset conditions in dry storage. Prediction of EOL rod internal pressure
Energy Technology Data Exchange (ETDEWEB)
Feria, F.; Herranz, L.E.
2015-07-01
During dry storage, cladding resistance to failure can be affected by several degrading mechanisms like creep or hydrides radial reorientation. The driving force of these effects is the stress at which the cladding is submitted. The maximum stress in the cladding is determined by the end-of-reactor-life (EOL) rod internal pressure, PEOL, at the maximum temperature attained during dry storage. Thus, PEOL sets the initial conditions of storage for potential time-dependent changes in the cladding. Based on FRAPCON-3.5 calculations, the aim of this work is to analyse the PEOL of a PWR fuel rod irradiated to burnups greater than 60 GWd/tU, where limited information is available. In order to be conservative, demanding irradiation histories have been used with a peak linear power of 44 kW/m. FRAPCON-3.5 results show an increasing exponential trend of PEOL with burnup, from which a simple correlation has been derived. The comparison with experimental data found in the literature confirms the enveloping nature of the predicted curve. Based on that, a conservative prediction of cladding stress in dry storage has been obtained. The comparison with a critical stress threshold related to hydrides embrittlement seems to point out that this issue should not be a concern at burnups below 65 GWd/tU. (Author)
International Nuclear Information System (INIS)
Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh
2014-01-01
The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects
Energy Technology Data Exchange (ETDEWEB)
Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)
2014-05-01
The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.
International Nuclear Information System (INIS)
Hocking, W.H.; Szostak, F.J.
1999-09-01
An investigation of the composition of the metallic inclusions in CANDU fuel, which contain Mo, Tc, Ru, Rh and Pd, has been conducted as a function of burnup by wavelength dispersive X-ray (WDX) microanalysis. Quantitative measurements were performed on micrometer sized particles embedded in thin sections of fuel using elemental standards and the ZAF method. Because the fission yields of the noble metals change with burnup, as a consequence of a shift from almost entirely 235 U fission to mainly 239 Pu fission, their inventories were calculated from the fuel power histories using the WIMS-Origin code for comparison with experiment. Contrary to expectations that the oxygen potential would be buffered by progressive Mo oxidation, little evidence was obtained for reduced incorporation of Mo in the noble-metal particles at high burnup. These surprising results are discussed with respect to the oxygen balance in irradiated CANDU fuels and the likely intrinsic and extrinsic sinks for excess oxygen. (author)
A novel QC-LDPC code based on the finite field multiplicative group for optical communications
Yuan, Jian-guo; Xu, Liang; Tong, Qing-zhen
2013-09-01
A novel construction method of quasi-cyclic low-density parity-check (QC-LDPC) code is proposed based on the finite field multiplicative group, which has easier construction, more flexible code-length code-rate adjustment and lower encoding/decoding complexity. Moreover, a regular QC-LDPC(5334,4962) code is constructed. The simulation results show that the constructed QC-LDPC(5334,4962) code can gain better error correction performance under the condition of the additive white Gaussian noise (AWGN) channel with iterative decoding sum-product algorithm (SPA). At the bit error rate (BER) of 10-6, the net coding gain (NCG) of the constructed QC-LDPC(5334,4962) code is 1.8 dB, 0.9 dB and 0.2 dB more than that of the classic RS(255,239) code in ITU-T G.975, the LDPC(32640,30592) code in ITU-T G.975.1 and the SCG-LDPC(3969,3720) code constructed by the random method, respectively. So it is more suitable for optical communication systems.
Warped Discrete Cosine Transform-Based Low Bit-Rate Block Coding Using Image Downsampling
Directory of Open Access Journals (Sweden)
Ertürk Sarp
2007-01-01
Full Text Available This paper presents warped discrete cosine transform (WDCT-based low bit-rate block coding using image downsampling. While WDCT aims to improve the performance of conventional DCT by frequency warping, the WDCT has only been applicable to high bit-rate coding applications because of the overhead required to define the parameters of the warping filter. Recently, low bit-rate block coding based on image downsampling prior to block coding followed by upsampling after the decoding process is proposed to improve the compression performance for low bit-rate block coders. This paper demonstrates that a superior performance can be achieved if WDCT is used in conjunction with image downsampling-based block coding for low bit-rate applications.
A Compressed Sensing-Based Low-Density Parity-Check Real-Number Code
Directory of Open Access Journals (Sweden)
Zaixing He
2013-09-01
Full Text Available In this paper, we propose a novel low-density parity-check real-number code, based on compressed sensing. A real-valued message is encoded by a coding matrix (with more rows than columns and transmitted over an erroneous channel, where sparse errors (impulsive noise corrupt the codeword. In the decoding procedure, we apply a structured sparse (low-density parity-check matrix, the Permuted Block Diagonal matrix, to the corrupted output, and the errors can be corrected by solving a compressed sensing problem. A compressed sensing algorithm, Cross Low-dimensional Pursuit, is used to decode the code by solving this compressed sensing problem. The proposed code has high error correction performance and decoding efficiency. The comparative experimental results demonstrate both advantages of our code. We also apply our code to cryptography.
Managing Widely Disparate Code Bases Through Automation of Continuous Integration and Deployment
McLaughlin, B. D.; Joshi, T.
2013-12-01
NASA EOSDIS tools, services, and service endpoints are widely dispersed across different sub-agencies and sub-organizations. Each of these entities has a different set of skills and widely varying codebases. Some produce sophisticated, well-tested, stable and deployable code, while others are struggling to meet stringent requirements with limited resources. This disparity makes the process of partnering with and deploying code onto the Earthdata platform (https://earthdata.nasa.gov) difficult, even at times impossible. The Earthdata Code Collaborative (ECC) is a project repository and code hosting facility that addresses this problem directly through a three-tiered approach: 1. Provide a standardized set of testing and automation tools for all hosted projects. 2. Regularly report on bugs and features as well as testing coverage and success through Web-based tools. 3. Directly pipeline projects from the ECC into the Earthdata production environment. This session will explain the architecture behind the ECC, including the custom software and 3rd party tools used. It will also detail the process by which decisions were and are being made to arrive at a fully-automated suite of tools and tests that allow any code base to quickly improve its quality and become a candidate for Earthdata inclusion. The session is oriented towards developers, managers, and team members involved in the process of developing, testing, deploying, and ensuring the quality of a code base, whether that code base be tens of millions of lines of code or simply hundreds.
ZZ PWR-AXBUPRO-GKN, Measured Axial Burnup Profiles, NPP Neckarewstheim
International Nuclear Information System (INIS)
Neuber, Jens-Christian; Lamprecht, Thomas
1999-01-01
-GKN2K contains a sample of 850 Axial Burnup Shapes released by Nuclear Power Plant Neckarwestheim II, Germany, on May 03, 2000 through Siemens AG Power Generation. All of these shapes belong to one and the same fuel assembly type, namely the Siemens Konvoi fuel assembly type FOCUS (TM). For this fuel assembly type the shapes were gathered from the cycles 5 through 12 of NPP Neckarwestheim II. All the shapes refer to EOCs. The shapes are derived from in-core 3D power density distribution measurements based on flux measurements. At 28 fuel assembly positions the flux data are monitored at 32 equidistant axial nodes. Thus, one has a total of 896 measuring points These measurements are performed every fourteenth day. The measurements are performed with the aid of the Siemens/KWU's Aeroball System which has the advantage of monitoring simultaneously all the axial nodes. The high spatial resolution and the high frequency of the measurement campaigns as well as the accuracy of the measurement result in shapes of outstanding quality. For instance, the spatial resolution suffices to discriminate the flux dips caused by the presence of the spacer grids. What regards the end effect, the presence of spacer grids in the ends of the fuel zone should attract one's attention. The fuel assemblies to which the axial shapes under examination refer have had initial enrichments of 3.8 wt.-% and 4.0 wt.-% U-235. For the benchmark the initial enrichment is assumed to be 4.0 wt.-%
Sonar target localization based on spike coded spectrograms
Directory of Open Access Journals (Sweden)
Bertrand FONTAINE
2006-12-01
Full Text Available Target location is coded into the pattern of spikes that run up the auditory nerve to the bat's brain. Realistic scenes containing multiple, closely spaced, reflectors give rise to complex echo signals consisting of multiple filtered copies of the bat's own vocalisation. Some of this filtering is due to the directivity of the bat’s reception system i.e., the outer ears, and some of it is due to sound absorption and the reflection process. The analysis below concentrates on the conspicuous ridges (notches these filter operations give rise to in the time-frequency representation of the echo as produced by the bat's inner ear. Assuming multiple threshold detecting neurons for each frequency channel it is shown how the distribution of spike times within the generated spike bursts is linked to the presence and characteristics of these notches. A neural network decoding the spike bursts in terms of target location is described.
Directory of Open Access Journals (Sweden)
H. Prashantha Kumar
2011-09-01
Full Text Available Low density parity check (LDPC codes are capacity-approaching codes, which means that practical constructions exist that allow the noise threshold to be set very close to the theoretical Shannon limit for a memory less channel. LDPC codes are finding increasing use in applications like LTE-Networks, digital television, high density data storage systems, deep space communication systems etc. Several algebraic and combinatorial methods are available for constructing LDPC codes. In this paper we discuss a novel low complexity algebraic method for constructing regular LDPC like codes derived from full rank codes. We demonstrate that by employing these codes over AWGN channels, coding gains in excess of 2dB over un-coded systems can be realized when soft iterative decoding using a parity check tree is employed.
A Secure Network Coding-based Data Gathering Model and Its Protocol in Wireless Sensor Networks
Directory of Open Access Journals (Sweden)
Qian Xiao
2012-09-01
Full Text Available To provide security for data gathering based on network coding in wireless sensor networks (WSNs, a secure network coding-based data gathering model is proposed, and a data-privacy preserving and pollution preventing (DPPaamp;PP protocol using network coding is designed. DPPaamp;PP makes use of a new proposed pollution symbol selection and pollution (PSSP scheme based on a new obfuscation idea to pollute existing symbols. Analyses of DPPaamp;PP show that it not only requires low overhead on computation and communication, but also provides high security on resisting brute-force attacks.
Disposal criticality analysis methodology's principal isotope burnup credit
International Nuclear Information System (INIS)
Doering, T.W.; Thomas, D.A.
2001-01-01
This paper presents the burnup credit aspects of the United States Department of Energy Yucca Mountain Project's methodology for performing criticality analyses for commercial light-water-reactor fuel. The disposal burnup credit methodology uses a 'principal isotope' model, which takes credit for the reduced reactivity associated with the build-up of the primary principal actinides and fission products in irradiated fuel. Burnup credit is important to the disposal criticality analysis methodology and to the design of commercial fuel waste packages. The burnup credit methodology developed for disposal of irradiated commercial nuclear fuel can also be applied to storage and transportation of irradiated commercial nuclear fuel. For all applications a series of loading curves are developed using a best estimate methodology and depending on the application, an additional administrative safety margin may be applied. The burnup credit methodology better represents the 'true' reactivity of the irradiated fuel configuration, and hence the real safety margin, than do evaluations using the 'fresh fuel' assumption. (author)
Problems associated with high burnup of VVER reactor fuel
International Nuclear Information System (INIS)
Reshetnikov, F.G.; Golovin, I.S.; Bibilashvili, Yu.K.; Solyany, V.I.
1981-01-01
One of the principal direction of improving the characteristics of the thermal power reactor fuel cycle is to increase the burn-up of fuel in fuel elements. So in future for VVER-1000 elements the planned burn-up of fuel must be up to 40000-50000 MW-day/t U. The realization of those parameters would permit a substantial decrease in the consumption of natural uranium in the open fuel cycle, a considerable reduction of the load on fuel element fabrication and reprocessing plants, which will favourably affect the whole economics of the fuel - power cycle. However, the position of the optimum of the fuel component of the cost depending on burn-up is determined not only by the economy of uranium, the cost of fuel element fabrication processes, uranium enrichment and the chemical reprocessing of burnt fuel, but also by the provision of the required safety of high burn-up fuel elements. Thus, scientists and designing engineers face the problem of designing serviceable and reliable thermal power reactor fuel elements intended for longer service life and higher burn-ups and ensuring the safety of the whole reactor plant. This paper deals with some of the aspects of this most complicated problem for the fuel elements of VVER type only
Advances in Metallic Fuels for High Burnup and Actinide Transmutation
Energy Technology Data Exchange (ETDEWEB)
Hayes, S. L.; Harp, J. M.; Chichester, H. J. M.; Fielding, R. S.; Mariani, R. D.; Carmack, W. J.
2016-10-01
Research and development activities on metallic fuels in the US are focused on their potential use for actinide transmutation in future sodium fast reactors. As part of this application, there is a desire to demonstrate a multifold increase in burnup potential. A number of metallic fuel design innovations are under investigation with a view toward significantly increasing the burnup potential of metallic fuels, since higher discharge burnups equate to lower potential actinide losses during recycle. Promising innovations under investigation include: 1) lowering the fuel smeared density in order to accommodate the additional swelling expected as burnups increase, 2) utilizing an annular fuel geometry for better geometrical stability at low smeared densities, as well as the potential to eliminate the need for a sodium bond, and 3) minor alloy additions to immobilize lanthanide fission products inside the metallic fuel matrix and prevent their transport to the cladding resulting in fuel-cladding chemical interaction. This paper presents results from these efforts to advance metallic fuel technology in support of high burnup and actinide transmutation objectives. Highlights include examples of fabrication of low smeared density annular metallic fuels, experiments to identify alloy additions effective in immobilizing lanthanide fission products, and early postirradiation examinations of annular metallic fuels having low smeared densities and palladium additions for fission product immobilization.
Achieving High Burnup Targets With Mox Fuels: Techno Economic Implications
International Nuclear Information System (INIS)
Clement Ravi Chandar, S.; Sivayya, D.N.; Puthiyavinayagam, P.; Chellapandi, P.
2013-01-01
For a typical MOX fuelled SFR of power reactor size, Implications due to higher burnup have been quantified. Advantages: – Improvement in the economy is seen upto 200 GWd/ t; Disadvantages: – Design changes > 150 GWd/ t bu; – Need for 8/ 16 more fuel SA at 150/ 200 GWd/ t bu; – Higher enrichment of B 4 C in CSR/ DSR at higher bu; – Reduction in LHR may be required at higher bu; – Structural material changes beyond 150 GWd/ t bu; – Reprocessing point of view-Sp Activity & Decay heat increase. Need for R & D is a must before increasing burnup. bu- refers burnup. Efforts to increase MOX fuel burnup beyond 200 GWd/ t may not be highly lucrative; • MOX fuelled FBR would be restricted to two or four further reactors; • Imported MOX fuelled FBRs may be considered; • India looks towards launching metal fuel FBRs in the future. – Due to high Breeding Ratio; – High burnup capability
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
Chassie, G.G.; Tochaie, M.; Xu, Z.
2010-01-01
This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)
International Nuclear Information System (INIS)
2001-07-01
The Technical Committee Meeting (TCM) included separate sessions on the specific topics of fuel thermal performance and fission product retention. On thermal performance, it is apparent that the capability exists to measure conductivity in high burnup fuel either by out-of-pile measurement or by instrumentation of test reactor rods. State-of-the-art modelling codes contain models for the conductivity degradation process, and hence adequate predictions of fuel temperature are achievable. Concerning fission product release, it is clear that many groups around the world are actively investigating the subject, with experimental and modelling programmes being pursued. However, a general consensus on the exact mechanisms of gas release and related gas bubble swelling has yet to emerge, even at medium burnup levels. Fission gas phenomena, not only the release to open volumes, but the whole sequence of processes taking place prior to this, need to be modelled in any modern fuel performance code. The presence of gaseous fission products may generate rapid fuel swelling during power transients, and this can cause PCI and rod failure. At high burnups, the quantity of released gases could give rise to pressures exceeding the safe limits. Modelling of pellet-cladding interaction (PCI) effects during transient operation is also an active area of study for many groups. In some situations a purely empirical approach to failure modelling can be justified, while for other applications a more detailed mechanistic approach is required. Another aspect of cladding modelling which was featured at the TCM concerned corrosion and hydriding. Although this issue can be the main life-limiting factor on fuel duty, it is apparent that modelling methods, and the experimental measurement techniques that underpin them, are adequate. A session was included on MOX fuel modelling. Substantial programmes of work, especially by the MOX vendors, appear to be underway to bring the level of understanding
Verification calculations for the WWER version of the TRANSURANUS code
International Nuclear Information System (INIS)
Elenkov, D.; Boneva, S.; Georgieva, M.; Georgiev, S.; Schubert, A.; Van Uffelen, P.
2006-01-01
The paper presents part of the work performed in the study project 'Research and Development for Licensing of Nuclear Fuel in Bulgaria'. The main objective of the project is to provide assistance for solving technical questions of the fuel licensing process in Bulgaria. One important issue is the extension of the predictive capabilities of fuel performance codes for Russian-type WWER reactors. In the last decade, a series of international projects has been based on the TRANSURANUS fuel performance code: Specific models for WWER fuel have been developed and implemented in the code in the late 90's. In 2000-2003, basic verification work was done by using experimental data of nuclear fuel irradiated in WWER-440 reactors. While the present paper focuses on the analysis of WWER-1000 standard fuel under normal operating conditions, the above study project covers additional tasks: 1) Post-irradiation calculations of ramp tests performed in the DR3 test reactor of the Risoe National Laboratory (five instrumented fuel rods of the Risoe 3 dataset contained in the IFPE database) using the TRANSURANUS code; 2) Compilation of cross-section libraries for isotope evolution calculations in WWER-440 and WWER-1000 fuel assemblies using the ORIGEN-S code; 3) Analysis of current situation and needs for an extension of the curriculum in Nuclear Engineering at the Technical University of Sofia. In this paper the post-irradiation calculations of steady-state irradiation experiments with nuclear fuel for Russian-type WWER-1000 reactors, using the latest release of the TRANSURANUS code (v1m1j03)are presented. Regarding a comprehensive verification of modern fuel performance codes, the burn-up region above 40 MWd/kgU is of increasing importance. A number of new phenomena emerge at high fuel burn-up, implying the need for enlarged databases of postirradiation examinations (PIE). For one fuel assembly irradiated in a WWER-1000 reactor with a rod discharge burn-up between 50 and 55 MWd
Substep methods for burnup calculations with Bateman solutions
International Nuclear Information System (INIS)
Isotalo, A.E.; Aarnio, P.A.
2011-01-01
Highlights: → Bateman solution based depletion requires constant microscopic reaction rates. → Traditionally constant approximation is used for each depletion step. → Here depletion steps are divided to substeps which are solved sequentially. → This allows piecewise constant, rather than constant, approximation for each step. → Discretization errors are almost completely removed with only minor slowdown. - Abstract: When material changes in burnup calculations are solved by evaluating an explicit solution to the Bateman equations with constant microscopic reaction rates, one has to first predict the development of the reaction rates during the step and then further approximate these predictions with their averages in the depletion calculation. Representing the continuously changing reaction rates with their averages results in some error regardless of how accurately their development was predicted. Since neutronics solutions tend to be computationally expensive, steps in typical calculations are long and the resulting discretization errors significant. In this paper we present a simple solution to reducing these errors: the depletion steps are divided to substeps that are solved sequentially, allowing finer discretization of the reaction rates without additional neutronics solutions. This greatly reduces the discretization errors and, at least when combined with Monte Carlo neutronics, causes only minor slowdown as neutronics dominates the total running time.
Coding and Billing in Surgical Education: A Systems-Based Practice Education Program.
Ghaderi, Kimeya F; Schmidt, Scott T; Drolet, Brian C
Despite increased emphasis on systems-based practice through the Accreditation Council for Graduate Medical Education core competencies, few studies have examined what surgical residents know about coding and billing. We sought to create and measure the effectiveness of a multifaceted approach to improving resident knowledge and performance of documenting and coding outpatient encounters. We identified knowledge gaps and barriers to documentation and coding in the outpatient setting. We implemented a series of educational and workflow interventions with a group of 12 residents in a surgical clinic at a tertiary care center. To measure the effect of this program, we compared billing codes for 1 year before intervention (FY2012) to prospectively collected data from the postintervention period (FY2013). All related documentation and coding were verified by study-blinded auditors. Interventions took place at the outpatient surgical clinic at Rhode Island Hospital, a tertiary-care center. A cohort of 12 plastic surgery residents ranging from postgraduate year 2 through postgraduate year 6 participated in the interventional sequence. A total of 1285 patient encounters in the preintervention group were compared with 1170 encounters in the postintervention group. Using evaluation and management codes (E&M) as a measure of documentation and coding, we demonstrated a significant and durable increase in billing with supporting clinical documentation after the intervention. For established patient visits, the monthly average E&M code level increased from 2.14 to 3.05 (p patients the monthly average E&M level increased from 2.61 to 3.19 (p educational and workflow interventions, which improved resident coding and billing of outpatient clinic encounters. Using externally audited coding data, we demonstrate significantly increased rates of higher complexity E&M coding in a stable patient population based on improved documentation and billing awareness by the residents. Copyright
Multiple frequencies sequential coding for SSVEP-based brain-computer interface.
Directory of Open Access Journals (Sweden)
Yangsong Zhang
Full Text Available BACKGROUND: Steady-state visual evoked potential (SSVEP-based brain-computer interface (BCI has become one of the most promising modalities for a practical noninvasive BCI system. Owing to both the limitation of refresh rate of liquid crystal display (LCD or cathode ray tube (CRT monitor, and the specific physiological response property that only a very small number of stimuli at certain frequencies could evoke strong SSVEPs, the available frequencies for SSVEP stimuli are limited. Therefore, it may not be enough to code multiple targets with the traditional frequencies coding protocols, which poses a big challenge for the design of a practical SSVEP-based BCI. This study aimed to provide an innovative coding method to tackle this problem. METHODOLOGY/PRINCIPAL FINDINGS: In this study, we present a novel protocol termed multiple frequencies sequential coding (MFSC for SSVEP-based BCI. In MFSC, multiple frequencies are sequentially used in each cycle to code the targets. To fulfill the sequential coding, each cycle is divided into several coding epochs, and during each epoch, certain frequency is used. Obviously, different frequencies or the same frequency can be presented in the coding epochs, and the different epoch sequence corresponds to the different targets. To show the feasibility of MFSC, we used two frequencies to realize four targets and carried on an offline experiment. The current study shows that: 1 MFSC is feasible and efficient; 2 the performance of SSVEP-based BCI based on MFSC can be comparable to some existed systems. CONCLUSIONS/SIGNIFICANCE: The proposed protocol could potentially implement much more targets with the limited available frequencies compared with the traditional frequencies coding protocol. The efficiency of the new protocol was confirmed by real data experiment. We propose that the SSVEP-based BCI under MFSC might be a promising choice in the future.
Preparation of higher-actinide burnup and cross section samples
International Nuclear Information System (INIS)
Adair, H.L.; Kobisk, E.H.; Quinby, T.C.; Thomas, D.K.; Dailey, J.M.
1981-01-01
A joint research program involving the United States and the United Kingdom was instigated about four years ago for the purpose of studying burnup of higher actinides using in-core irradiation in the fast reactor at Dounreay, Scotland. Simultaneously, determination of cross sections of a wide variety of higher actinide isotopes was proposed. Coincidental neutron flux and energy spectral measurements were to be made using vanadium encapsulated dosimetry materials in the immediate region of the burnup and cross section samples. The higher actinide samples chosen for the burnup study were 241 Am and 244 Cm in the forms of Am 2 O 3 , Cm 2 O 3 , and Am 6 Cm(RE) 7 O 21 , where (RE) represents a mixture of lanthanide sesquioxides. It is the purpose of this paper to describe technology development and its application in the preparation of the fuel specimens and the cross section specimens that are being used in this cooperative program
Increased fuel burn-up and fuel cycle equilibrium
International Nuclear Information System (INIS)
Debes, M.
2001-01-01
Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)
Validation issues for depletion and criticality analysis in burnup credit
International Nuclear Information System (INIS)
Parks, C.V.; Broadhead, B.L.; Dehart, M.D.; Gauld, I.C.
2001-01-01
This paper reviews validation issues associated with implementation of burnup credit in transport, dry storage, and disposal. The issues discussed are ones that have been identified by one or more constituents of the United States technical community (national laboratories, licensees, and regulators) that have been exploring the use of burnup credit. There is not necessarily agreement on the importance of the various issues, which sometimes is what creates the issue. The broad issues relate to the paucity of available experimental data (radiochemical assays and critical experiments) covering the full range and characteristics of spent nuclear fuel in away-from-reactor systems. The paper will also introduce recent efforts initiated at Oak Ridge National Laboratory (ORNL) to provide technical information that can help better assess the value of different experiments. The focus of the paper is on experience with validation issues related to use of burnup credit for transport and dry storage applications. (author)
Application of grammar-based codes for lossless compression of digital mammograms
Li, Xiaoli; Krishnan, Srithar; Ma, Ngok-Wah
2006-01-01
A newly developed grammar-based lossless source coding theory and its implementation was proposed in 1999 and 2000, respectively, by Yang and Kieffer. The code first transforms the original data sequence into an irreducible context-free grammar, which is then compressed using arithmetic coding. In the study of grammar-based coding for mammography applications, we encountered two issues: processing time and limited number of single-character grammar G variables. For the first issue, we discover a feature that can simplify the matching subsequence search in the irreducible grammar transform process. Using this discovery, an extended grammar code technique is proposed and the processing time of the grammar code can be significantly reduced. For the second issue, we propose to use double-character symbols to increase the number of grammar variables. Under the condition that all the G variables have the same probability of being used, our analysis shows that the double- and single-character approaches have the same compression rates. By using the methods proposed, we show that the grammar code can outperform three other schemes: Lempel-Ziv-Welch (LZW), arithmetic, and Huffman on compression ratio, and has similar error tolerance capabilities as LZW coding under similar circumstances.
An investigation into the signals leakage from a smartcard based on different runtime code
CSIR Research Space (South Africa)
Frieslaar, I
2015-09-01
Full Text Available September 2015 at the Arabella Hotel & Spa in the heart of the Kogelberg Biosphere Reserve near Hermanus, Western Cape, South Africa An Investigation into the Signals Leakage From a Smartcard based on Different Runtime Code Ibraheem Frieslaar...
Interactive Consistency Algorithms Based on Voting and Error-Correding Codes
Krol, Th.
1995-01-01
This paper presents a new class of synchronous deterministic non authenticated algorithms for reaching interactive consistency (Byzantine agreement). The algorithms are based on voting and error correcting codes and require considerably less data communication than the original algorithm, whereas
Auto Code Generation for Simulink-Based Attitude Determination Control System
MolinaFraticelli, Jose Carlos
2012-01-01
This paper details the work done to auto generate C code from a Simulink-Based Attitude Determination Control System (ADCS) to be used in target platforms. NASA Marshall Engineers have developed an ADCS Simulink simulation to be used as a component for the flight software of a satellite. This generated code can be used for carrying out Hardware in the loop testing of components for a satellite in a convenient manner with easily tunable parameters. Due to the nature of the embedded hardware components such as microcontrollers, this simulation code cannot be used directly, as it is, on the target platform and must first be converted into C code; this process is known as auto code generation. In order to generate C code from this simulation; it must be modified to follow specific standards set in place by the auto code generation process. Some of these modifications include changing certain simulation models into their atomic representations which can bring new complications into the simulation. The execution order of these models can change based on these modifications. Great care must be taken in order to maintain a working simulation that can also be used for auto code generation. After modifying the ADCS simulation for the auto code generation process, it is shown that the difference between the output data of the former and that of the latter is between acceptable bounds. Thus, it can be said that the process is a success since all the output requirements are met. Based on these results, it can be argued that this generated C code can be effectively used by any desired platform as long as it follows the specific memory requirements established in the Simulink Model.
GRILLIX. A 3D turbulence code for magnetic fusion devices based on a field line map
International Nuclear Information System (INIS)
Stegmeir, Andreas Korbinian
2015-01-01
The complex geometry in the scrape-off layer of tokamaks poses problems to existing turbulence codes. The usually employed field aligned coordinates become ill defined at the separatrix. Therefore the parallel code GRILLIX was developed, which is based on a field line map. This allows simulations in additional complex geometries, especially across the separatrix. A new discretisation, based on the support operator method, for the highly anisotropic diffusion was developed and applied to a simple turbulence model (Hasegawa-Wakatani).
Burnup measurements with the Los Alamos fork detector
International Nuclear Information System (INIS)
Bosler, G.E.; Rinard, P.M.
1991-01-01
The fork detector system can determine the burnup of spent-fuel assemblies. It is a transportable instrument that can be mounted permanently in a spent-fuel pond near a loading area for shipping casks, or be attached to the storage pond bridge for measurements on partially raised spent-fuel assemblies. The accuracy of the predicted burnup has been demonstrated to be as good as 2% from measurements on assemblies in the United States and other countries. Instruments have also been developed at other facilities throughout the world using the same or different techniques, but with similar accuracies. 14 refs., 2 figs., 2 tabs
Fission gas release from fuel at high burnup
International Nuclear Information System (INIS)
Meyer, R.O.; Beyer, C.E.; Voglewede, J.C.
1978-03-01
The release of fission gas from fuel pellets at high burnup is reviewed in the context of the safety analysis performed for reactor license applications. Licensing actions are described that were taken to correct deficient gas release models used in these safety analyses. A correction function, which was developed by the Nuclear Regulatory Commission staff and its consultants, is presented. Related information, which includes some previously unpublished data, is also summarized. The report thus provides guidance for the analysis of high burnup gas release in licensing situations
Fast image restoration method based on coded exposure and vibration detection
He, Lirong; Cui, Guangmang; Feng, Huajun; Xu, Zhihai; Li, Qi; Chen, Yueting
2015-10-01
Fast image restoration method is proposed for vibration image deblurring based on coded exposure and vibration detection. The criterion of the code sequence selection is discussed in detail, and several factors are considered to search for the optimal coded exposure sequence. The blurred vibration image is obtained by the coded exposure technique. Meanwhile, the vibration track information of the camera is detected by a fiber-optic gyroscope. The point spread function (PSF) is estimated using a statistical method with the selected code sequence and vibration track information. Finally, the blurred image is quickly restored with the estimated PSF through a direct inverse filtering method. Simulation experiments are conducted to test the performance of the approach with different vibration forms. A real imaging system is constructed to verify the effectiveness of the proposed algorithm. Experimental results show that the presented algorithm could yield better subjective experiences and superior objective evaluation values.
Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria
2017-07-01
In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.
Modeling of pore coarsening in the rim region of high burn-up UO{sub 2} fuel
Energy Technology Data Exchange (ETDEWEB)
Xiao, Hongxing; Long, Chong Sheng [Science and Technology on Reactor Fuel and Materials Laboratory, Nuclear Power Institute of China, Chengdu (China)
2016-08-15
An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated UO{sub 2} fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated UO{sub 2} fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of UO{sub 2} fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.
Study on the sensitivity of Self-Powered Neutron Detectors (SPND) and its change due to burn-up
International Nuclear Information System (INIS)
Cho, Gyuseong; Lee, Wanno; Yoon, Jeong-Hyoun.
1996-01-01
Self-Powered Neutron Detectors (SPND) are currently used to estimate the power generation distribution and fuel burn-up in several nuclear power reactors in Korea. While they have several advantages such as small size, low cost, and relatively simple electronics required in conjunction with its usage, it has some intrinsic problems of the low level of output current, a slow response time, the rapid change of sensitivity which makes it difficult to use for a long term. In this paper, Monte Carlo simulation was accomplished to calculate the escape probability as a function of the birth position for the typical geometry of rhodium-based SPNDs. Using the simulation result, the burn-up profile of rhodium number density and the neutron sensitivity is calculated as a function of burn-up time in the reactor. The sensitivity of the SPND decreases non-linearly due to the high absorption cross-section and the non-uniform burn-up of rhodium in the emitter rod. The method used here can be applied to the analysis of other types of SPNDs and will be useful in the optimum design of new SPNDs for long-term usage. (author)
Fabrication of simulated mid-burnup CANDU fuel in the RFFL
International Nuclear Information System (INIS)
Dimayuga, F.C.
1997-01-01
Prediction of the coefficient of void reactivity for CANDU reactor cores is key to modeling postulated loss-of-coolant scenarios to support safety analyses. To reduce the uncertainty in these predictions, computer codes used to generate the predictions have to be well validated for cores at equilibrium burnup. To this end, a COG-funded project was launched to resume mixed oxide (MOX) fuel fabrication operations in the mothballed Recycle Fuel Fabrication Laboratory (RFFL) at CRL, and produce (U,Pu)O 2 fuel simulating mid-burnup CANDU fuel for physics testing in the ZED-2 reactor. In August 1996, rehabilitation of the RFFL was completed, and MOX operations were resumed in the facility. An up-to-date description of the RFFL, including the upgraded safety systems and process equipment, is presented. An overview of the fabrication campaign to produce 37 MOX fuel bundles for ZED-2 tests is given. The fabrication process used to manufacture the fuel from the starting powders to the finished elements and bundles is summarized. Fabrication data including production throughputs and inspection results is discussed. (author)
Burnup credit validation of SCALE-4 using light water reactor criticals
International Nuclear Information System (INIS)
Bowman, S.M.; Hermann, O.W.; Brady, M.C.
1993-01-01
The ANSI/ANS 8.1 criticality safety standard recommends validation and benchmarking of the calculational methods used in evaluating criticality safety limits for away-from-reactor applications. The lack of critical experiments with burned light-water-reactor (LWR) fuel in racks or in casks necessitates the validation of burnup credit methods by comparison to LWR core criticals. These are relevant benchmarks because they test a methodology's ability to predict spent fuel isotopics and to evaluate the reactivity effects of heterogeneities and strong absorbers. Data are available to perform analyses at precise state points. The US Department of Energy Burnup Credit Program has sponsored analysis of selected reactor core critical configurations from commercial pressurized-water-reactors (PWRs) in order to validate an appropriate analysis methodology. The initial methodology used the SCALE-4 code system to analyze a set of reactor critical configurations from Virginia Power's Surry and North Anna reactors. The methodology has since been revised to simplify both the data requirements and the calculational procedure for the criticality analyst. This revised methodology is validated here by comparison to three reactor critical configurations from Tennessee Valley Authority's Sequoyah Unit 2 Cycle 3 and two from Virginia Power's Surry Unit 1 Cycle 2
Simulation of the effects of the extend fuel rod burn-up under LOCA scenario
Energy Technology Data Exchange (ETDEWEB)
Gomes, Daniel de Souza; Silva, Antonio Teixeira; Abe, Alfredo, E-mail: dsgomes@ipen.br, E-mail: teixiera@ipen.br, E-mail: ayabe@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Giovedi, Claudia; Martins, Marcelo Ramos, E-mail: claudia.giovedi@labrisco.usp.br, E-mail: mrmartin@usp.br [Universidade de Sao Paulo (POLI/USP), Sao Paulo, SP (Brazil). Lab. de Analise, Avaliacao e Gerenciamento de Risco
2015-07-01
Due to the high burn-up imposed to the nuclear fuel in the last recent years, new challenges become important, including a deep review of the fuel performance under accident conditions. In this sense, available data in the open literature show that some experiments were carried out in order to study the behavior of fuel rods under LOCA (Loss of Coolant Accident) scenario. For instance, a series of experiments, designated IFA-650 series, performed in the Halden reactor in 2010 present data related to zircaloy fuel rods submitted to LOCA conditions. In the tests were addressed issues such as fuel fragmentation, relocation and dispersal for an extended irradiation cycle. In the studied case (IFA-650.5), the LOCA scenario was evaluated after a burnup of 83.4 MWd/kg. The aim of this paper is to compare the experimental data to the fuel performance obtained applying the codes FRAPCON and FRAPTRAN. Different phenomena were evaluated, such as ballooning, burst, cladding oxidation and fuel relocation. Also, the cladding metallurgical phase transformation was considered. The obtained results reproduced in a good way the experimental data, showing that the adopted models are representative of the observed phenomena. (author)
A heterogeneous model for burnup calculation in high temperature gas-cooled reactors
International Nuclear Information System (INIS)
Perfetti, C. M.; Angahie, S.; Baxter, A.; Ellis, C.
2008-01-01
A high resolution MCNPX model is developed to simulate nuclear design characteristics and fuel cycle features of High Temperature Gas-Cooled Reactors. Contrary to the conventional approach in the MCNPX model, fuel regions containing TRISO particles are not homogenized. A cube corner distribution approximation is used to directly model randomly dispersed TRISO fuel particles in a graphite matrix. The universe filling technique is used cover the entire range of fuel particles in the core. The heterogeneous MCNPX model is applied to simulate and analyze the complete fuel cycle of the General Atomics Plutonium-Consumption Modular Helium Reactor (PC-MHR). The PC-MHR reactor design is a variation of the General Atomic MHR design and is designed for the consumption or burning of excess Russian weapons plutonium. The MCNPX burnup calculation of the PC-MHR includes the simulation of a 260 effective full-power day fuel cycle at 600 MWt. Results of the MCNPX calculations suggest that during 260 effective full-power day cycle, 40% reduction in the whole core Pu-239 inventory could be achieved. Results of heterogeneous MCNPX burnup calculations in PC-MHR are compared with results of deterministically calculated values obtained from DIF3D codes. For the 260 effective full-power day cycle, the difference in mass Pu-239 mass reduction calculation using heterogeneous MCNPX and homogeneous DIF3D models is 6%. The difference in MCNPX and DIF3D calculated results for higher actinides are mostly higher than 6%. (authors)
Fuel burnup analysis of the TRIGA Mark II reactor at the University of Pavia
International Nuclear Information System (INIS)
Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto
2016-01-01
Highlights: • A fuel evolution model for a TRIGA Mark II reactor has been developed. • Reproduction of nearly 50 years of reactor operation. • The model was used to predict the best reactor reconfiguration. • Reactor life was extended without adding fresh fuel elements. - Abstract: A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper.
Corrosion behaviour of Zircaloy 4 fuel cans for high burnup in EdF PWRs
International Nuclear Information System (INIS)
Blat, M.; Kerrec, O.; Bourgoin, J.; Vrignaud, E.; Amanrich, H.
1994-01-01
Uniform corrosion of fuel cladding could be a limitation for burn-up enhancement. First, the oxide thickness measured on fuel cladding for high burn-up has been compared to the prediction of the EDF code, CYRANO 2E. A comparative metallurgical characterization has been also performed on samples which were oxidized in pile and in autoclave. Then, laboratories studies have been launched for a better understanding of the corrosion mechanisms. A reflection was proposed on the two main theoretical concepts proposed for these mechanisms. Their kinetics could be controlled by transfers in liquid medium (electrolyte) or in solid medium (compact oxide). For the first topic, a nanoscopic characterization of the oxide is in progress, using Atomic Force Microscope. The first results are presented. In the second case, an electrochemical approach (impedance spectroscopy and voltametry) is developed in our laboratories. The obtained results could give some new keys in order to understand the influence of some parameters (alloys composition, coolant chemistry,...). (authors). 7 figs., 1 tab., 7 refs
Directory of Open Access Journals (Sweden)
Jingang Liang
2016-06-01
Full Text Available Because of prohibitive data storage requirements in large-scale simulations, the memory problem is an obstacle for Monte Carlo (MC codes in accomplishing pin-wise three-dimensional (3D full-core calculations, particularly for whole-core depletion analyses. Various kinds of data are evaluated and quantificational total memory requirements are analyzed based on the Reactor Monte Carlo (RMC code, showing that tally data, material data, and isotope densities in depletion are three major parts of memory storage. The domain decomposition method is investigated as a means of saving memory, by dividing spatial geometry into domains that are simulated separately by parallel processors. For the validity of particle tracking during transport simulations, particles need to be communicated between domains. In consideration of efficiency, an asynchronous particle communication algorithm is designed and implemented. Furthermore, we couple the domain decomposition method with MC burnup process, under a strategy of utilizing consistent domain partition in both transport and depletion modules. A numerical test of 3D full-core burnup calculations is carried out, indicating that the RMC code, with the domain decomposition method, is capable of pin-wise full-core burnup calculations with millions of depletion regions.
Kavuluru, Ramakanth; Han, Sifei; Harris, Daniel
2017-01-01
Diagnosis codes are extracted from medical records for billing and reimbursement and for secondary uses such as quality control and cohort identification. In the US, these codes come from the standard terminology ICD-9-CM derived from the international classification of diseases (ICD). ICD-9 codes are generally extracted by trained human coders by reading all artifacts available in a patient’s medical record following specific coding guidelines. To assist coders in this manual process, this paper proposes an unsupervised ensemb