The code system COROUT: Radioactive inventory calculations
International Nuclear Information System (INIS)
The code system COROUT is devoted to the evaluation of nuclear reactor out-of-core radioactive inventory for the sake of the nuclear power plant decommissioning problem. The code includes calculations of the neutron flux distributions and activation kinetics in the consistent way. Only thermal neutrons are taken into consideration in the present code version. Code is divided into three steps. The first step prepares the necessary data file containing data on reactor geometry, core flux, reactor operational history and data on elements in the out-of-core zones. The main part of calculations are performed during the second step. Here the thermal neutron flux distribution in the out-of-core area is calculated for two-dimensional cylindrical geometry and the system of gain-loss equations and the activation kinetics is solved for the elements in the different out-of-core shells. The Vladimirov's method of iterations on the spatial grid is used for the neutron flux calculations. The kinetic equations are solved by the operational method. The change of neutron field due to activation during reactor campaign is taken into account. The third part of COROUT code system allows to prepare plots of flux and activity distribution for different shells. All steps could be initiated independently using the results stored at the previous steps. The code is destined for the personal computers and has been written on the base of 32-bit FORTRAN language for IBM PC. 4 refs, 6 figs, 1 tab
Burnup calculation code system COMRAD96
International Nuclear Information System (INIS)
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)
Integrated burnup calculation code system SWAT
International Nuclear Information System (INIS)
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)
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
Fast reactor nuclear physics parameters calculation code system 'EXPARAM'
International Nuclear Information System (INIS)
The calculation code system ''EXPARAM'' was designed to analyze the experimental results systematically measured at the fast critical assembly (FCA) in Tokai research establishment of JAERI. Some calculation codes developed independently in JAERI and in US research institutes were collected and arranged as the fast reactor physics calculation code system. The multi-group core calculation code and the perturbation calculation code based on the diffusion theory and transport theory calculate the physics parameters such as eigenvalue, reaction rate, Doppler reactivity worth and sodium void worth. The dynamic physics parameters such as prompt neutron lifetime and effective delayed neutron fraction are also calculated. Input and Output data of calculation codes are transferred to each other using a direct access file on UNIX computer system. (author)
SRAC2006: A comprehensive neutronics calculation code system
International Nuclear Information System (INIS)
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, SN 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)
Validation of IRBURN calculation code system through burnup benchmark analysis
International Nuclear Information System (INIS)
Assessment of the reactor fuel composition during the irradiation time, fuel management and criticality safety analysis require the utilization of a validated burnup calculation code system. In this work a newly developed burnup calculation code system, IRBURN, is introduced for the estimation and analysis of the fuel burnup in LWR reactors. IRBURN provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP4C as well as the versatile code for calculating the buildup and decay of nuclides in nuclear materials, ORIGEN2.1, along with other data processing and linking subroutines. This code has the capability of using different depletion calculation schemes. The accuracy and precision of the implemented algorithms to estimate the eigenvalue and spent fuel isotope concentrations are demonstrated by validation against reliable benchmark problem analyses. A comparison of IRBURN results with experimental data demonstrates that the code predicts the spent fuel concentrations within 10% accuracy. Furthermore, standard deviations of the average values for isotopic concentrations including IRBURN data decreases considerably in comparison with the same parameter excluding IRBURN results, except for a few sets of isotopes. The eigenvalue comparison between our results and the benchmark problems shows a good prediction of the k-inf values during the entire burnup history with the maximum difference of 1% at 100 MWd/kgU.
Methods and computer codes for nuclear systems calculations
Indian Academy of Sciences (India)
B P Kochurov; A P Knyazev; A Yu Kwaretzkheli
2007-02-01
Some numerical methods for reactor cell, sub-critical systems and 3D models of nuclear reactors are presented. The methods are developed for steady states and space–time calculations. Computer code TRIFON solves space-energy problem in (, ) systems of finite height and calculates heterogeneous few-group matrix parameters of reactor cells. These parameters are used as input data in the computer code SHERHAN solving the 3D heterogeneous reactor equation for steady states and 3D space–time neutron processes simulation. Modification of TRIFON was developed for the simulation of space–time processes in sub-critical systems with external sources. An option of SHERHAN code for the system with external sources is under development.
User effects on the transient system code calculations. Final report
International Nuclear Information System (INIS)
Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them
Simulator validation of calculation code in REDNET upgrade system
International Nuclear Information System (INIS)
The reactor data network (REDNET) is a computer-based data acquisition, display and archival system which acquires data from the National Research Universal (NRU) reactor's 'fuelled sites', and several experimental loop facilities in support of CANDU technology development (e.g., fuel, fuel behaviour, and materials research programs). The system supports the processing of data collected for subsequent display at the respective experimental facilities, and in the NRU control room. REDNET was installed in the 1980s based on the 1970s computer technology. The computer hardware is obsolete and spare parts are either extremely hard to find or are now unavailable. The Upgrade system is intended to replace the REDNET and eliminate the risk of losing the data acquisition of important experimental data needed in support of the CANDU Fuel Development Program. An important goal of the Upgrade system is to improve the accuracy in the measurement and calculation of thermal power. Calculations in REDNET are performed in FORTRAN code with some in-house macros. The same calculations are re-implemented in the Upgrade system in structured-text and function-block languages. To ensure that there is no deviation or loss of accuracy in the calculations of the Upgrade system compared to those in REDNET, software validation is performed on calculation code in the Upgrade system. The validation consists of a two-stage and three-point check (at ∼0%, 50% and ∼100% signal level) process for every data type and data point in the Upgrade system. This paper presents the purpose, the major tools and process, and the results of the validation. It is concluded, based on the validation results, that the Upgrade system achieves at least the same, and in many cases better, accuracy in all the calculations. (author)
Computer code for calculating reliability/availability of technical systems
International Nuclear Information System (INIS)
Three computer codes are reviewed, which can be applied to reliability analyses of technical systems. They are based on the fault tree and the laws of probability theory. The codes can be used for both non-repairable and repairable systems. The simulation code REMO 79 and the analytical code RELAV are based on the conception that a failure of system components is immediately detected and repaired. The model of the FUPRO2 code provides for failures to be detected and repaired only in periodic functional tests. Apart from code descriptions experience and far-reaching aspects resulting from modularization of the fault trees are summarized. (author)
Saphyr: a code system from reactor design to reference calculations
International Nuclear Information System (INIS)
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels
Saphyr: a code system from reactor design to reference calculations
Energy Technology Data Exchange (ETDEWEB)
Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)
2003-07-01
In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.
Recent transportation calculation code systems and their accuracy evaluation
International Nuclear Information System (INIS)
In the field of shielding design, many studies have been carried out for the development of radiation transportation calculation codes (transportation codes) including Monte Carlo codes. The present report outlines major transportation codes used in Japan for design of shielding. Major one-dimensional codes include ANISN (Sn), PALLAS-PL and SP-Br (direct integration) whili two-dimensional ones include DOT-3.5 and TWOTRAN-II. All these transportation codes have been developed on the basis of numerical solution to the Boltzmann's transportation equation. These codes are roughly divided into two groups: discrete ordinates type and Monte Carlo type. The former include Sn-type codes and direct integration type codes. Sn-type codes are currently used most widely. The accuracy and other features of a code should be tested before applysing it to practical shielding design. One of the techniques for this purpose is the benchmark method, which consists of benchmark tests and analysis of the test results. The possible overall error involved in calculations can be determined from the benchmark tests. (Nogami, K.)
Miniature neutron source reactor burnup calculations using IRBURN code system
International Nuclear Information System (INIS)
Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.
Revised SWAT. The integrated burnup calculation code system
International Nuclear Information System (INIS)
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
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)
Benchmark calculations by the nuclear criticality safety analysis code system JACS(MGCL, KENO-IV)
International Nuclear Information System (INIS)
Since 1980, as many as 1394 cases of benchmark calculations on criticality problems have been performed by the KENO-IV Monte Carlo calculation code with the MGCL cross section data library. The code system is a part of the criticality safety evaluation code system JACS developed at JAERI. The code validation results have been published in a series of JAERI-M reports and others. This report summarizes these results and the reliability of the code system systematically. The number of the calculated cases briefly described in this report together with their experimental systems and data are 502 for 17 kinds of homogeneous single-unit systems, 331 for 8 kinds of homogeneous multi-unit systems and 561 for 16 kinds of heterogeneous systems. Discussions and interpretations are made on the calculated keff's (neutron multiplication factors) with their bias errors. The factors related to the bias errors are confirmed together with their causes and trends. (author)
Burnup calculations using the ORIGEN code in the CONKEMO computing system
International Nuclear Information System (INIS)
This article describes the CONKEMO computing system for kinetic multigroup calculations of nuclear reactors and their physical characteristics during burnup. The ORIGEN burnup calculation code has been added to the system. The results of an international benchmark calculation are also presented. (author)
An integrated multi-functional neutronics calculation and analysis code system: VisualBUS
International Nuclear Information System (INIS)
Neutronics calculation and analysis are the bases of reactor physics design, radiation protection, fuel management optimization, nuclear safety analysis, etc. After surveying and evaluating the status and trend of development of neutronics calculation and analysis codes, a network-based integrated multi-functional neutronics calculation and analysis code system has been designed and developed for applications in fusion, fission and various hybrid systems based on the adoption of advanced neutronics calculating approaches and modern computer' software technologies. A series of benchmark tests and applications have shown the maturity and effectiveness of the system. This paper gives a brief overview about main technical features of the system, the benchmark tests and applications. (authors)
A transport based one-dimensional perturbation code for reactivity calculations in metal systems
Energy Technology Data Exchange (ETDEWEB)
Wenz, T.R.
1995-02-01
A one-dimensional reactivity calculation code is developed using first order perturbation theory. The reactivity equation is based on the multi-group transport equation using the discrete ordinates method for angular dependence. In addition to the first order perturbation approximations, the reactivity code uses only the isotropic scattering data, but cross section libraries with higher order scattering data can still be used with this code. The reactivity code obtains all the flux, cross section, and geometry data from the standard interface files created by ONEDANT, a discrete ordinates transport code. Comparisons between calculated and experimental reactivities were done with the central reactivity worth data for Lady Godiva, a bare uranium metal assembly. Good agreement is found for isotopes that do not violate the assumptions in the first order approximation. In general for cases where there are large discrepancies, the discretized cross section data is not accurately representing certain resonance regions that coincide with dominant flux groups in the Godiva assembly. Comparing reactivities calculated with first order perturbation theory and a straight {Delta}k/k calculation shows agreement within 10% indicating the perturbation of the calculated fluxes is small enough for first order perturbation theory to be applicable in the modeled system. Computation time comparisons between reactivities calculated with first order perturbation theory and straight {Delta}k/k calculations indicate considerable time can be saved performing a calculation with a perturbation code particularly as the complexity of the modeled problems increase.
A transport based one-dimensional perturbation code for reactivity calculations in metal systems
International Nuclear Information System (INIS)
A one-dimensional reactivity calculation code is developed using first order perturbation theory. The reactivity equation is based on the multi-group transport equation using the discrete ordinates method for angular dependence. In addition to the first order perturbation approximations, the reactivity code uses only the isotropic scattering data, but cross section libraries with higher order scattering data can still be used with this code. The reactivity code obtains all the flux, cross section, and geometry data from the standard interface files created by ONEDANT, a discrete ordinates transport code. Comparisons between calculated and experimental reactivities were done with the central reactivity worth data for Lady Godiva, a bare uranium metal assembly. Good agreement is found for isotopes that do not violate the assumptions in the first order approximation. In general for cases where there are large discrepancies, the discretized cross section data is not accurately representing certain resonance regions that coincide with dominant flux groups in the Godiva assembly. Comparing reactivities calculated with first order perturbation theory and a straight Δk/k calculation shows agreement within 10% indicating the perturbation of the calculated fluxes is small enough for first order perturbation theory to be applicable in the modeled system. Computation time comparisons between reactivities calculated with first order perturbation theory and straight Δk/k calculations indicate considerable time can be saved performing a calculation with a perturbation code particularly as the complexity of the modeled problems increase
International Nuclear Information System (INIS)
In a D-T burning fusion reactor, the radioactivity induced by the 14 MeV neutrons causes many problems. It limits personnel access to the reactor during shutdown, generates decay heat and produces radwastes. A code system THIDA had been developed in 1978 to calculate the radioactivity and dose rate around a fusion device. The THIDA system consisted of the followings: one- and two-dimensional discrete ordinates radiation transport codes; induced activity calculation code; three libraries for transmutation and decay chain data, transmutation cross sections and delayed gamma-ray emission data. The present report gives a complete description of THIDA-2, a new advanced version of the THIDA system which has the following major improvements: 1. Capability to treat three-dimensional calculation models by the use of a Monte Carlo transport code. 2. Accurate decay heat calculation following the transport of delayed gamma rays. 3. Simplification of the data input process by the use of free format scheme and closer coupling between the radiation transport codes and the induced activity calculation code. 4. Self-descriptive output format and additional plotter output. 5. Capability to calculate problems requiring larger core memory by the use of variable dimension. (author)
User effects on the thermal-hydraulic transient system code calculations
International Nuclear Information System (INIS)
Large thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are dealt by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. During these comparisons of the code results, there has been a continuous debate on the way how the code user influences the predicted system behaviour. This rather subjective element might become a crucial point with respect to the quantitative evaluation of the code uncertainties which is essential if the 'best estimate codes are used for licensing procedures'. The International Standard Problem Exercises (ISPs) proposed by the OECD-Committee for the Safety of Nuclear Installations (CSNI) and by IAEA (International Atomic Energy Agency) and thermalhydraulic code assessment activity undertaken by US.NRC under International Code Assessment and Application Program (ICAP) demonstrate the large effort put in this framework by organisations all over the world. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies the influence of the code user on the calculated results is directly accounted. In this paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under user effects. These reasons and user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. In summary
DNBR calculation in digital core protection system by a subchannel analysis code
International Nuclear Information System (INIS)
The DNBR calculation uncertainty and DNBR margin were evaluated in digital core protection system by a thermal-hydrualic subchannel analysis code MATRA. A simplified thermal-hydraulic code CETOP is used to calculate on-line DNBR in core protection system at a digital PWR. The DNBR tuning process against a best-estimate subchannel analysis code is required for CETOP to ensure accurate and conservative DNBR calculation but not necessary for MATRA. The DNBR calculations by MATRA and CETOP were performed for a large number of operating condition in Yonggwang nulcear units 3-4 where the digitial core protection system is initially implemented in Korea. MATRA resulted in a less negative mean value (i.e., reduce the overconservatism) and a somewhat larger standard deviation of the DNBR error. The uncertainty corrected minimum DNBR by MATRA was shown to be higher by 1.8% -9.9% that the CETOP DNBR
User effects on the thermal-hydraulic transient system code calculations
International Nuclear Information System (INIS)
Large thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are dealt by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. During these comparisons of the code results, there has been a continuous debate on the way how the code user influences the predicted system behaviour. This rather subjective element might become a crucial point with respect to the quantitative evaluation of the code uncertainties which is essential if the ''best estimate codes are used for licensing procedures''. The International Standard Problem Exercises (ISPs) proposed by the OECD-Committee for the Safety of Nuclear Installations (CSNI) and by IAEA (International Atomic Energy Agency) and thermalhydraulic code assessment activity undertaken by US.NRC under International Code Assessment and Application Program (ICAP) demonstrate the large effort put in this framework by organisations all over the world. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies the influence of the code user on the calculated results is directly accounted. In this paper, the results of the investigations on the user effects for the thermalhydraulic transient system codes will be presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under user effects. These reasons and user effects will be discussed in detail and general recommendations and conclusions will be presented to control and limit them. (authors). 15
DIST: a computer code system for calculation of distribution ratios of solutes in the purex system
Energy Technology Data Exchange (ETDEWEB)
Tachimori, Shoichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1996-05-01
Purex is a solvent extraction process for reprocessing the spent nuclear fuel using tri n-butylphosphate (TBP). A computer code system DIST has been developed to calculate distribution ratios for the major solutes in the Purex process. The DIST system is composed of database storing experimental distribution data of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}: DISTEX and of Zr(IV), Tc(VII): DISTEXFP and calculation programs to calculate distribution ratios of U(IV), U(VI), Pu(III), Pu(IV), Pu(VI), Np(IV), Np(VI), HNO{sub 3} and HNO{sub 2}(DIST1), and Zr(IV), Tc(VII)(DITS2). The DIST1 and DIST2 determine, by the best-fit procedures, the most appropriate values of many parameters put on empirical equations by using the DISTEX data which fulfill the assigned conditions and are applied to calculate distribution ratios of the respective solutes. Approximately 5,000 data were stored in the DISTEX and DISTEXFP. In the present report, the following items are described, 1) specific features of DIST1 and DIST2 codes and the examples of calculation 2) explanation of databases, DISTEX, DISTEXFP and a program DISTIN, which manages the data in the DISTEX and DISTEXFP by functions as input, search, correction and delete. and at the annex, 3) programs of DIST1, DIST2, and figure-drawing programs DIST1G and DIST2G 4) user manual for DISTIN. 5) source programs of DIST1 and DIST2. 6) the experimental data stored in the DISTEX and DISTEXFP. (author). 122 refs.
Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors
International Nuclear Information System (INIS)
The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)
Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning
International Nuclear Information System (INIS)
Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs
SWAT, Step-Wise Burnup Analysis Code System to Combine SRAC-95 Cell Calculation Code and ORIGEN2
International Nuclear Information System (INIS)
1 - Description of program or function: SWAT evaluates isotopic composition of spent nuclear fuel, especially for burnup credit issues by driving codes SRAC95 and ORIGEN2.1 or ORIGEN2. SWAT is an automated driver code system. At the initial development phase, it was constructed by combining source programs of SRAC and ORIGEN2. To overcome the problem associated with code updates, SWAT chose to use system function of UNIX operating system to execute SRAC95 and ORIGEN2. So that, SWAT is independent of development and modification of SRAC95 and ORIGEN2.1. In SWAT, ORIGEN2(82) or ORIGEN2.1 is used for burnup calculations using the matrix exponential method. An updated decay library is included in the distribution. SWAT uses SRAC95 for neutron spectrum and effective cross section calculation in 107 groups, using the collision probability method for given geometry and isotopic composition. One or two dimensional cell geometries are supported in SRAC95. NEA-1698/02: The main purpose of new package is to run SWAT on several machines not supported in previous package (IA64 under Linux, Windows with cygwin and Sun,...) and several commercial FORTRAN compiler (Intel, PGI, Fujitsu). 2 - Methods: In calculating the problem-dependent cross section in SWAT, the total burnup history is divided into 'burnup steps'. Power, boric acid concentration, temperature of each region, and void ratio of coolant are given as history data. For each burnup step, the neutron spectrum and effective cross section are evaluated by SRAC95 using the information given in previous burnup calculation and cell geometry information. The user can select geometry options for the collision probability method in SRAC95. 3 - Restrictions on the complexity of the problem: Resonance absorption calculation with ultra-fine group cross section can not be directly applicable for 2D geometry
Development of a MCNP–ORIGEN burn-up calculation code system and its accuracy assessment
International Nuclear Information System (INIS)
Highlights: • MCNP and ORIGEN are coupled to perform nuclides depletion and decay calculation. • Coupled system MCORE uses “modified predictor corrector” approach. • MCORE can use different depletion schemes and simulate fuel shuffling. • MCORE is assessed by a “VVER-1000 LEU Assembly Computational Benchmark”. • MCORE is also assessed by a fast reactor benchmark problem. - Abstract: An MCNP–ORIGEN burn-up calculation code system, named MCORE (MCNP and ORIGEN burn-up Evaluation code), is developed in this work. MCORE makes use of the Monte Carlo neutron and photon transport code MCNP4C and nuclides depletion and decay calculation code ORIGEN2.1. MCNP and ORIGEN are coupled by data processing and linking subroutines. In MCORE, a so called “modified predictor corrector” approach is used. MCORE provides the capability of using different depletion calculation schemes and simulating fuel shuffling. Total nuclide density changes in active cells are considered in MCORE. The validity and applicability of the developed code are tested by investigating and predicting the neutronic and isotopic behavior of a “VVER-1000 LEU Assembly Computational Benchmark” at lattice level and a “Physics of Plutonium Recycling” fast reactor at core level (OECD-NEA). The comparison results show that the MCORE code predicts the nuclide composition within 5% accuracy and k∞ within 800 pcm at the end of the burn-up for LEU assembly (40 MWD/kg HM). For a fast reactor, the results obtained by MCORE are in the range of reported results except for 243Am. In general, MCORE results show a good agreement with the benchmark values
Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system
Energy Technology Data Exchange (ETDEWEB)
Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi
1998-01-01
In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)
Benchmark calculations on nuclear characteristics of JRR-4 HEU core by SRAC code system
International Nuclear Information System (INIS)
The reduced enrichment program for the JRR-4 has been progressing based on JAERI's RERTR (Reduced Enrichment Research and Test Reactor) program. The SRAC (JAERI Thermal Reactor Standard Code System for Reactor Design and Analysis) is used for the neutronic design of the JRR-4 LEU Core. This report describes the benchmark calculations on the neutronic characteristics of the JRR-4 HEU Core in order to validate the calculation method. The benchmark calculations were performed on the various kind of neutronic characteristics such as excess reactivity, criticality, control rod worth, thermal neutron flux distribution, void coefficient, temperature coefficient, mass coefficient, kinetic parameters and poisoning effect by Xe-135 build up. As the result, it was confirmed that these calculated values are in satisfactory agreement with the measured values. Therefore, the calculational method by the SRAC was validated. (author)
IRACM : A code system to calculate induced radioactivity produced by ions and neutrons
International Nuclear Information System (INIS)
It is essential to estimate of radioactivity induced in accelerator components and samples bombarded by energetic ion beams and the secondary neutrons of high-energy accelerator facilities in order to reduce the amount of radioactive wastes and to minimize radiation exposure to personnel. A computer code system IRACM has been developed to estimate product nuclides and induced radioactivity in various radiation environments of accelerator facilities. Nuclide transmutation with incident particles of neutron, proton, deuteron, alpha, 12C, 14N, 16O, 20Ne and 40Ar can be computed for arbitrary multi-layer target system in a one-dimensional geometry. The code system consists of calculation modules and libraries including activation cross sections, decay data and photon emission data. The system can be executed in both FACOM-M780 mainframe and DEC workstations. (author)
Comparative calculations on selected two-phase flow phenomena using major PWR system codes
International Nuclear Information System (INIS)
In 1988 a comparative study on important features and models in six major best estimate thermal hydraulic codes for PWR systems was implemented (Comparison of thermal hydraulic safety codes for PWR Graham, Trotman, London, EUR 11522). It was a limitation of that study that the source codes themselves were not available but the comparison had to be based on the available documentation. In the present study, the source codes were available and the capability of four system codes to predict complex two-phase flow phenomena has been assessed. Two areas of investigation were selected: (a) pressurized spray phenomena; (b) boil-up phenomena in rod bundles. As regards the first area, experimental data obtained in 1972 on the Neptunus Facility (Delft University of Technology) were compared with the results of the calculations using Athlet, Cathare, Relap 5 and TRAC-PT1 and, concerning the second area, the results of two experimental facilities obtained in 1980 and 1985 on Thetis (UKEA) and Pericles (CEA-Grenoble) were considered
CEQCSY: a new code for chemical equilibrium calculation in multiphased systems
International Nuclear Information System (INIS)
As part of the CEC Chemval/mirage project, a method is presented for calculating the thermodynamic equilibrium state of a multiphase system, by minimizing its Gibbs free energy constrained by mass balances. Compared to the other algorithms available in the literature, the method has three main characteristics: - the sets of equations corresponding to the conditions of homogeneous and heterogeneous equilibria are simultaneously solved, - a mathematical criterion for bringing a new multicomponent phase in the system is rigorously demontrated. - It enables a detailed representation of the multisite solid solutions with constraints called site closure relation. The code CEQCSY (Chemical Equilibrium in Complex SYstem) uses this formalism, and works with the thermodynamic data base from the EQ3/6 code. This choice makes easier several compared tests with EQ6: quartz dissolution in water, water-atmospheric air equilibrium, theoretical re-equilibrium of seawater, hydrothermal alteration of granite including solid solutions. The test results demonstrate the high efficiency and velocity of the code CEQCSY, when working on equilibrium state of multiphase systems. This high velocity was the aim of this work, in order to couple with thermic, hydrodynamic or mechanic codes
MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA
International Nuclear Information System (INIS)
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations. A CD-ROM is attached as an appendix. (J.P.N.)
International Nuclear Information System (INIS)
The computer codes BROHR and SYSFIT are presented. Both codes are based on the first-order matrix formalism of ion optics. By means of the code BROHR the trajectories of ions and electrons inside of any inclined field accelerating tubes can be calculated. The influence of the stripping process at tandem accelerators is included by changing of the mass and the charge of the ions and by increasing the beam emittance. The code SYSFIT is used for calculation of any beam transport systems and of the transported beam. Special requested imaging properties can be realized by parameter variation. Calculated examples are given for both codes. (author)
Benchmark calculations by the thermal reactor standard nuclear design code system SRAC
International Nuclear Information System (INIS)
This report summarizes the present status of the thermal reactor standard nuclear design code system SRAC developed by the nuclear design working group of the JAERI thermal reactor standard code committee which was started on July 1978. Descriptions are given at first on the brief introduction and the process of development of the code system SRAC, and then, the several benchmark tests performed to evaluate the performance of the code system. The results show the good predictions of the experimental keff values of the critical facilities; TCA for LWR, JMTRC for JAERI MTR, DCA for the Japanese Advanced Thermal Reactor and SHE for VHTR. A trial to the IAEA benchmark calculations on the Reduction of uranium Enrichment of Research and Test Reactors yields satisfactory agreements with the results of ANL. Another test to evaluate the fast group constants was also attempted by tracing the fast reactor benchmark problems which have been used to evaluate nuclear data file in the FBR reactor physics field. (author)
International Nuclear Information System (INIS)
The Carmen code (theory and user's manual) is described. This code for assembly and core calculations uses diffusion theory (Citation), with feedback in the cross sections by zone due to the effects of burnup, water density, fuel temperature, Xenon and Samarium. The burnup calculation of a full cycle is solved in only an execution of Carmen, and in a reduced computer time. (auth.)
International Nuclear Information System (INIS)
A computer code system for fast calculation of activation and transmutation has been developed. The system consists of a driver code, cross-section libraries, flux libraries, a material library, and a decay library. The code is used to predict transmutations in a Ti-modified 316 stainless steel, a commercial ferritic alloy (HT9), and a V-15%Cr-5%Ti alloy in various magnetic fusion energy (MFE) test facilities and conceptual reactors
International Nuclear Information System (INIS)
Criticality calculation codes/code systems MCNP, MVP, SCALE and JACS, which are currently typically used in Japan for nuclear criticality safety evaluation, were benchmarked for so called dissolver-typed systems, i.e., fuel rod arrays immersed in fuel solution. The benchmark analyses were made for the evaluated critical experiments published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: one evaluation representing five critical configurations from heterogeneous core of low-enriched uranium dioxides at the Japan Atomic Energy Research Institute and two evaluations representing 16 critical configurations from heterogeneous core of mixed uranium and plutonium dioxides (MOXs) at the Battelle Pacific Northwest Laboratories of the U.S.A. The results of the analyses showed that the minimum values of the neutron multiplication factor obtained with MCNP, MVP, SCALE and JACS were 0.993, 0.990, 0.993, 0.972, respectively, which values are from 2% to 4% larger than the maximum permissible multiplication factor of 0.95. (author)
International Nuclear Information System (INIS)
There have been two versions of SWAT depending on details of its development history: the revised SWAT that uses the deterministic calculation code SRAC as a neutron transportation solver, and the SWAT3.1 that uses the continuous energy Monte Carlo code MVP or MCNP5 for the same purpose. It takes several hours, however, to execute one calculation by the continuous energy Monte Carlo code even on the super computer of the Japan Atomic Energy Agency. Moreover, two-dimensional burnup calculation is not practical using the revised SWAT because it has problems on production of effective cross section data and applying them to arbitrary fuel geometry when a calculation model has multiple burnup zones. Therefore, SWAT4.0 has been developed by adding, to SWAT3.1, a function to utilize the deterministic code SARC2006, which has shorter calculation time, as an outer module of neutron transportation solver for burnup calculation. SWAT4.0 has been enabled to execute two-dimensional burnup calculation by providing an input data template of SRAC2006 to SWAT4.0 input data, and updating atomic number densities of burnup zones in each burnup step. This report describes outline, input data instruction, and examples of calculations of SWAT4.0. (author)
Initial core calculation of 1 MW reactor TRIGA PUSPATI (RTP) using SRAC code system
International Nuclear Information System (INIS)
The 1 MWatt TRIGA PUSPATI Reactor (RTP) was located in Malaysian Institute for Nuclear Technology Research (MINT). This research reactor was from TRIGA MARK II type and was put into operation on 1983 and has reached its first criticality on 28 June 1982. Since then, this reactor has been used for various beam experiments, irradiation facilities, radioisotope production and education and training. The RTP uses three types of fuel elements, namely, 8.5wt%, 12wt% and 20wt% which enriched to about 20% of U-235 for all types. The RTP has four control rods which made up of boron carbide. It has cylindrical core but not in periodically in its lattice structure, which possibly locates 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector was made from high purity graphite. Because of this research reactor's power is relatively small compared to the power reactor; it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been made for several times. Until now, there are 11 configurations of the core and recently has achieved the 12th configuration. This paper will described the first core configuration calculation using SRAC code system which was first introduced in 2005 during the FNCA workshop. (author)
International Nuclear Information System (INIS)
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates. (authors)
Energy Technology Data Exchange (ETDEWEB)
Austregesilo, Henrique [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)]. E-mail: Henrique.Austregesilo@grs.de; Bals, Christine [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany); Trambauer, Klaus [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, 85748 Garching (Germany)
2007-09-15
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B{sub 4}C oxidation do not affect significantly the total calculated hydrogen release rates.
International Nuclear Information System (INIS)
In the frame of developmental assessment and code validation, a post-test calculation of the test QUENCH-07 was performed with ATHLET-CD. The system code ATHLET-CD is being developed for best-estimate simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation and fission products behaviour. The first step of the work was the simulation of the test QUENCH-07 applying the modelling options recommended in the code User's Manual (reference calculation). The global results of this calculation showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed experimentally. Results of this sensitivity analysis indicate that the main experimental measurements lay within the uncertainty range of the corresponding calculated values. Among the main contributors to the uncertainty of code results are the heat transfer coefficient due to forced convection to superheated steam-argon mixture, the thermal conductivity of the shroud isolation and the external heater rod resistance. Uncertainties on modelling of B4C oxidation do not affect significantly the total calculated hydrogen release rates
Owen, Albert K.
1987-01-01
A computer code was written which utilizes ray tracing techniques to predict the changes in position and geometry of a laser Doppler velocimeter probe volume resulting from refraction effects. The code predicts the position change, changes in beam crossing angle, and the amount of uncrossing that occur when the beams traverse a region with a changed index of refraction, such as a glass window. The code calculates the changes for flat plate, cylinder, general axisymmetric and general surface windows and is currently operational on a VAX 8600 computer system.
International Nuclear Information System (INIS)
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 keff (''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.
International Nuclear Information System (INIS)
The Institute for Reactor Safety (IRS) of the Research Centre of Karlsruhe (FZK) is involved in the qualification and further development of design and safety program systems for current LWR and innovative reactors. These investigations are focused not only on the study of the behavior of the whole nuclear power plant by means of system codes but also on the core behavior by means of detailed coupled neutronic and thermal hydraulic solutions. To improve the prediction of the safety margins based on local parameters the coupling of detailed neutronic (transport or Monte Carlo) and thermal hydraulic (subchannel or CFD codes) at the fuel pin level is required. At IRS/FZK a coupling on the sub channel code COBRA-TF with the neutron transport code THREEDANT has been realized within the Karlsruhe modular reactor calculation system KAPROS-E. This coupling scheme has been applied to determine the pin power of a German PWR fuel assembly obtaining promising results. In addition, this coupling scheme is being qualified by performing Monte Carlo calculations for selected fuel assembly conditions. In this contribution, the approach followed to couple 3D transport code THREEDANT with the thermal hydraulic subchannel code COBRA -TF as well as selected results for a PWR fuel assembly will be presented. In addition the qualification work using Monte Carlo solutions for the pin-by-pin simulation of the PWR fuel assembly will be given and discussed. (authors)
Decay heat experiment and validation of calculation code systems for fusion reactor
International Nuclear Information System (INIS)
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)
Decay heat experiment and validation of calculation code systems for fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)
The MARS15-based FermiCORD code system for calculation of the accelerator-induced residual dose
Grebe, A; Lu, T; Mokhov, N; Pronskikh, V
2016-01-01
The FermiCORD code system, a set of codes based on MARS15 that calculates the accelerator-induced residual doses at experimental facilities of arbitrary configurations, has been developed. FermiCORD is written in C++ as an add-on to Fortran-based MARS15. The FermiCORD algorithm consists of two stages: 1) simulation of residual doses on contact with the surfaces surrounding the studied location and of radionuclide inventories in the structures surrounding those locations using MARS15, and 2) simulation of the emission of the nuclear decay gamma-quanta by the residuals in the activated structures and scoring the prompt doses of these gamma-quanta at arbitrary distances from those structures. The FermiCORD code system has been benchmarked against similar algorithms based on other code systems and showed a good agreement. The code system has been applied for calculation of the residual dose of the target station for the Mu2e experiment and the results have been compared to approximate dosimetric approaches.
International Nuclear Information System (INIS)
Consideration is given of a technique and algorithms of constructing neutron trajectories in the Monte-Carlo method taking into account the data on adjoint transport equation solution. When simulating the transport part of transfer kernel the use is made of piecewise-linear approximation of free path length density along the particle motion direction. The approach has been implemented in programs within the framework of the BRAND code system. The importance is calculated in the multigroup P1-approximation within the framework of the DD-30 code system. The efficiency of the developed computation technique is demonstrated by means of solution of two model problems. 4 refs.; 2 tabs
The GRAPE code system for the calculation of precompound and compound nuclear reactions
International Nuclear Information System (INIS)
The statistical exciton model following the master-equation approach has been improved and extended for application as an evaluation tool of double-differential reaction cross sections at incident nucleon energies of 5 to 50 MeV. For this purpose the code system GRAPE has been developed. An important characteristic of the proposed model is that consistency with equilibrium models has been demanded for the summed exciton-state densities as well as for the particle and γ-ray emission cross sections. Consistency with the adopted state densities has also been imposed upon the internal transition rates. A survey of the theory is given and the structure of the GRYPHON code is described. This report also contains a users' manual for GRYPHON
International Nuclear Information System (INIS)
The RTP is a light-water moderated and pool-type TRIGA MARK II reactor with power capacity of 1MWt. It was built in 1979 and attained the first criticality on 28 June 1982. The RTP was designed mainly for neutron activation analysis, small angle neutron scattering, neutron radiography, radioisotope production, education and training purposes. It uses standard TRIGA fuel developed by General Atomic in which the zirconium hydride moderator is homogeneously combined with enriched uranium. It has a cylindrical core with which possibility of locating 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector is made of high purity graphite. Because of its relatively small power, it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been carried out several times. Until now, there were 12 configurations of the core, the most recent change being in July 2006. This paper will describe the RTP core calculation using the Monte Carlo MVP code system. VP is a general multi-purpose Monte Carlo code for neutron and photon transport calculation in order to have an accurate and fast Monte Carlo simulation of neutron and photon transport problems. The MVP Monte Carlo code calculation is based on the continuous energy method. This code is capable of adopting an accurate physics model, geometry description and variance reduction technique. When compared to the conventional scalar method, this code could achieve higher computation speed by up to a factor of 10 on the vector super-computer. The RTP core has been modelled using cylinder geometry along the z-coordinate geometry with the MVP code system while its material cross section data is calculated beforehand. The JENDL3.3 data library was used in the whole calculation. The objectives of the calculation are to calculate the multiplication factor values (keff), fission density and flux distribution from the tally data. The calculation also
International Nuclear Information System (INIS)
Full text of publication follows: The QUENCH fuel bundle experiments, performed at the Forschungszentrum Karlsruhe in Germany, aim to investigate the hydrogen source term and the bundle degradation during reflood of an overheated reactor core. The test QUENCH-07, in which the bundle was cooled from high temperatures by steam injected from the bottom, was the first experiment in this test series with a boron carbide absorber rod in the bundle. One major objective of this test was to provide information on the B4C/SS/Zry interactions, on the formation of gaseous reaction products during B4C oxidation and control rod degradation, and on the impact of control rod degradation on surrounding rods. In the general frame of developmental assessment and code validation, a post-test calculation of test QUENCH-07 complemented by an uncertainty analysis was performed with the code ATHLET-CD. The system code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management procedures. It applies the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with dedicated models for core degradation, including mechanical rod behaviour, zirconium and B4C oxidation, melting and relocation of metallic and ceramic components, and for the release and transport of fission products and aerosols. The first step of the work was the simulation of the QUENCH-07 experiment, applying the modeling options recommended in the code User's Manual (reference calculation). The global results of this calculation, mainly with respect to the hydrogen release rate and to the time evolution of bundle temperatures in different elevations, showed a good agreement with the measured data. This calculation was complemented by a sensitivity analysis in order to investigate the influence of a combined variation of code input parameters on the simulation of the main phenomena observed in the experiment. For this
Decay heat measurement on fusion reactor materials and validation of calculation code system
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)
International Nuclear Information System (INIS)
A general purpose user's version of the EGS4 code system has been developed to make EGS4 easily applicable to the safety analysis of nuclear fuel cycle facilities. One such application involves the determination of skyshine dose for a variety of photon sources. To verify the accuracy of the code, it was benchmarked with Kansas State University (KSU) photon skyshine experiment of 1977. The results of the simulation showed that this version of EGS4 would be appicable to the skyshine calculation. (author)
International Nuclear Information System (INIS)
Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The
International Nuclear Information System (INIS)
The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)
Comparison of energy deposition calculations by the LAHET code system with experimental results
International Nuclear Information System (INIS)
A comparison was performed between the energy deposition predicted by the LAHET code system (LCS) and experimental values for 800-, 1,000-, and 1,200-MeV protons on targets composed of beryllium, carbon, aluminum, iron, copper, lead, bismuth, and uranium. The lead, bismuth, and uranium targets showed agreement within ∼ 10% at locations throughout the targets, and the agreement of the total energy deposited over the axial length of the targets ranged from 1 to 18%. for the lighter materials, the agreement at locations throughout the target was within ∼25 %. No definable trend could be determined for the lighter materials because some LCS predictions were greater and some were less than the experimental results, and some showed very good agreement. Also, the LCS underpredicted the proton ranges for 800-MeV protons on iron, 800- and 1,000-MeV protons on copper, and 800- and 1,000-MeV protons on uranium
International Nuclear Information System (INIS)
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
International Nuclear Information System (INIS)
The new code system SUHAM-U-VVER unites two code systems: UNK and SUHAM-2D. It uses micro group (about 7000 groups) library of the UNK code and the Surface Harmonics method for solving the multigroup (up to 89 groups) neutron transport equation in all volume of a reactor core. The first variant of the code system SUHAM-U-VVER-01 having capacity to calculate from beginning to end the one state of two-dimensional layer of VVER-1000 reactor core is described. Selection of the number and boundaries of groups is carried out. Verification of new code system was carried out by the benchmark calculations of 16 states of 4 variants of VVER-1000 fuel assemblies with uranium and MOX fuel. Practically all the values calculated by SUHAM-U-VVER-01 are between the values calculated by reference codes
Energy Technology Data Exchange (ETDEWEB)
Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)
2005-07-01
The new code system SUHAM-U-VVER unites two code systems: UNK and SUHAM-2D. It uses micro group (about 7000 groups) library of the UNK code and the Surface Harmonics method for solving the multigroup (up to 89 groups) neutron transport equation in all volume of a reactor core. The first variant of the code system SUHAM-U-VVER-01 having capacity to calculate from beginning to end the one state of two-dimensional layer of VVER-1000 reactor core is described. Selection of the number and boundaries of groups is carried out. Verification of new code system was carried out by the benchmark calculations of 16 states of 4 variants of VVER-1000 fuel assemblies with uranium and MOX fuel. Practically all the values calculated by SUHAM-U-VVER-01 are between the values calculated by reference codes.
International Nuclear Information System (INIS)
In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Panka, Istvan; Kereszturi, Andras [Hungarian Academy of Sciences, Budapest (Hungary). Reactor Analysis Dept.
2013-09-15
In this paper the uncertainties of the neutronic calculations at core level - originating from the uncertainties of the basic nuclear data - are presented. The investigations have been made for a VVER-1000 core (Kozloduy-6) defined in the frame of the OECD NEA UAM benchmark. In the first part of the paper, the uncertainties of the effective multiplication factor, the assembly-wise radial power distribution, the axial power distribution and the rod worth are shown. After that the preliminary evaluation of the uncertainties of the neutron kinetic calculations are presented for a rod movement transient at HZP (Hot Zero Power) state, where the uncertainties of the time dependent core and assembly powers and the dynamic reactivity were evaluated. In both cases, we will see that the most important quantities - at core level and at HZP state - have a considerable uncertainty which is originating from the uncertainties of the basic cross section library in these investigations. (orig.)
Performance of large LWR system codes in calculating the steam-generator heat-transfer behavior
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This paper presents a series of modeling experiences and problems in simulating the thermal-hydraulic behavior of large PWR steam generators using the RELAP4 and RELAP5 computer codes. Sensitivity studies investigating the heat transfer characteristics of both once-through and U-tube steam generators are discussed. Suggestions and recommendations are given for effective use and potential future improvements of these codes
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The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
WIMS-ANL 4.0, Deterministic Code System for Lattice Calculation
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1 - Description of program or function: The WIMS-ANL code is an extension of the Winfrith WIMS-D4 code for lattice cell computations. This code has been tailored to address some of the problem areas encountered in dealing with research reactor fuels, experiment, reflector and control regions. The SUPERCELL option eliminates some of the limitations of the traditional SPECTROX solution and supports the solution of more complex geometries with a more detailed spatial mesh and multiple resonance materials. The code generates both macroscopic and microscopic cross sections in the ISOTXS format with any selected number of energy groups. The user can specify which fission product isotopes are to be explicitly included in the microscopic burnup dependent ISOTXS library. Fission product library data can be generated for use with the MCNP code and burnup dependent applications. The cross section library data provided are based on ENDF/B version VI (69 group) and V (69 and 172 group) data. A revised 172 group library based on ENDF/B-VI is being generated with newer data and additional isotopes. This library will be made available at a later time. The code is variably dimensioned so that other group structures could be used. The source code and output format have been completely revised to reflect current coding practices and to permit display of the results on typical desk top monitors. The content of the output displayed is completely under the user's control. 2 - Methods:The methods of solution in WIMS-ANL remain unchanged from those used in the original WIMS-D4 code with the same resonance treatment and a choice of collision probability and DSN solutions for the simple lattice cell. The SUPERCELL option provides for the selection of supporting auxiliary cells that might represent the various different elements and varying spectra of the final SUPERCELL model. The resonance treatments where applicable are carried out in the auxiliary cells. These data are combined in the
Plant transients under small abnormalities of FBR 'Monju' calculated by a plant system code
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The objectives of the present study are to analyze plant transients caused by small abnormalities and to find plant parameters by which operators can recognize these small abnormalities. In order to evaluate the plant transient during an abnormal situation in the water system using the plant system code NETFLOW++, the turbine and feedwater (FW) systems should be analyzed with good precision. The code is validated using the measured data at 'Monju'. Several abnormalities in the water system are candidates of the present study, e.g., FW control valve degradation, FW pump degradation, heat transfer degradation due to fouling on heat transfer tubes of the evaporator, loss-of-feedwater-heating, etc. All major components in the tertiary system are included in the calculation model such as the steam generators, the high-pressure turbine, the deaerator, the FW pump, the FW heaters, the FW control valves, the steam control valve, extraction lines and drainpipes. In case of a malfunction of a FW control valve resulting in low flow rate, a large temperature increase at the outlet of the evaporator is observed. On the other hand, a temperature decrease at the outlet of the evaporator occurs if heat transfer tubes in the evaporator have fouling. As a result of the calculations, it was determined that temperature at the outlet of the evaporator is a good indicator to detect abnormal situations. (author)
Development of Fast running DNBR Calculation Code
Energy Technology Data Exchange (ETDEWEB)
Kwon, Hyuk; Seo, K. W.; Kim, S. J.; Hwang, D. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2010-10-15
SMART core adopted a core protection(SCOPS) and a supervising system(SCOMS) to satisfy the SAFDL for AOO and normal operation. Generally, the criteria is limited to the DNBR limit so that the DNBR calculation module is required in the protection and the supervising system of core. There are CPU time limit and calculation robustness as some requirements of the DNBR calculation module in SCOPS and SCOMS caused by hardware limitations. The non-iterative few channel methods are needed to satisfy the requirements. Non-iterative numerical method is similar to the CETOP algorithm originated from ref. 1. The method is known as the non-iterative prediction and correction method. An optimum number of channels for core lumping model is selected as 4- channel which is same channel number of CETOP model. A compensation model of lumped channel is needed to ensure that the 4-channel thermal hydraulic field is nearly equivalent to that field of 1/8-core model that is calculated by MATRA-S. The code called FAST that is fast running DNBR calculation is developed to satisfy the requirements of CPU time and calculation robustness. Present paper is described of characteristics and calculation results of developed FAST code
Core-wide calculations by HELIOS code
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The transport method of HELIOS is called the CCCP method, because it is based on current coupling and collision probabilities. The system to be calculated consists of space elements that are coupled with each other and with the boundaries by interface currents. The angular dependence of the interface or coupling currents can be discretized in various ways. This is done by partitioning the directional half-sphere into a number of θ polar levels, and each θ level into a number of φ azimuthal intervals. In two dimensional calculations the discretization of the azimuthal level is dominant. In the last issue of the HELIOS code (version 1.10) the maximum value of azimuthal discretization is increased from 4 to 12. This gives the possibility to calculate large (core-wide or near core-wide) systems with appropriate accuracy, which extends the applicability of the HELIOS program. This paper presents the experience gained from HELIOS calculations of large systems having several WWER-440 assemblies. The examined parameter is the core-wide power distribution, which was inadequately calculated by former versions of HELIOS. The application of high azimuthal discretization gives substantial improvement in accuracy, compared to reference solutions calculated by MCNP Monte-Carlo code. Although HELIOS is designed to calculate assembly-wide systems, it is now applicable to core-wide systems. Using its possibilities, the area of application is extended to calculate reference solutions for core-wide programs or to examine spectral changes of few-group cross sections due to burnup in real situations. Some potential areas of application are presented in the paper, together with the limitations of those applications. (Author)
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In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)
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Nuclear technology development pointed out the need for a new assessment of the fuel cycle back-end. Treatment and disposal of radioactive wastes arising from nuclear fuel reprocessing is known as one of the problems not yet satisfactorily solved, together with separation process of uranium and plutonium from fission products in highly irradiated fuels. Aim of this work is to present an improvement of the computer code for solvent extraction process calculation previously designed by the authors. The modeling of the extraction system has been modified by introducing a new method for calculating the distribution coefficients. The new correlations were based on deriving empirical functions for not only the apparent equilibrium constants, but also the solvation number. The mathematical model derived for calculating separation performance has been then tested for up to ten components and twelve theoretical stages with minor modifications to the convergence criteria. Suitable correlations for the calculation of the distribution coefficients of Uranium, Plutonium, Nitric Acid and fission products were constructed and used to successfully simulate several experimental conditions. (Author)
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EFFI calculates the electromagnetic field and vector potential in coil systems of arbitrary geometry. The coils are made from circular arc and/or straight segments of rectangular cross-section conductor. EFFI can also calculate magnetic flux lines, magnetic force, and inductance. The methods used for the calculations are based on a combination of analytical and numerical integration of the Biot--Savart law for a volume distribution of current. These methods yield accurate field values inside and outside the conductor. All input to EFFI is format-free and is checked for validity before any calculations are done. Any errors detected during the check produce a diagnostic that lists the error, the code's objection to it, and the number of the offending data card. EFFI produces output in both printed and graphical form. Each page of output is labeled with the title of the problem, the time, the computer and date of the run, and the version number and compilation date for EFFI. In addition, each column of numbers on each page is appropriately labeled. Examples from the coil design for the Mirror Fusion Test Facility (MFTF) and a divertor design for a Tokamak reactor are used for illustration
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This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
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The calculations performed for the Almaraz Unit 2 PWR using the code packages of the Atomic Energy Corporation of South Africa Ltd. are summarized. These calculations were done as part of the IAEA Coordinated Research Programme on In-Core Fuel Management Code Package Validation for LWRs. A brief description of the one-dimensional cross section generation package as well as of the Level II (scoping type) global core calculational package which was used is given. Detailed results are presented in several appendices. 29 figs., 20 tabs., 10 refs
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This work presents the Physics of Source Driven Systems (ADS). It shows some statics and Kinetics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)
KENO-IV code benchmark calculation, (6)
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A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multigroup constants library MGCL. The present report describes the results of a benchmark test using criticality experiments about Plutonium fuel in various shape. In all, 33 cases of experiments have been calculated for Pu(NO3)4 aqueous solution, Pu metal or PuO2-polystyrene compact in various shape (sphere, cylinder, rectangular parallelepiped). The effective multiplication factors calculated for the 33 cases distribute widely between 0.955 and 1.045 due to wide range of system variables. (author)
Calculation code of the fission products activity
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The document describes the two codes for the calculation of the fission products activity. The ''Pepin le bref'' code gives the exact value of the beta and gamma activities of completely known fission products. The code ''Plus Pepin'' introduces the beta and gamma activities whose properties are partially known. (A.L.B.)
International Nuclear Information System (INIS)
Full text: The implementation in the reaction code system EMPIRE-2.19 of an advanced formalism for fission cross-section calculation has been completed. The formalism is based on the optical model for fission and can be applied for nuclei exhibiting double- or triple-humped barrier starting from sub-barrier excitation energies. The optical model for fission, initially developed to describe the resonant structure of the fission cross section at sub-barrier excitation energies due to the vibrational states in the second well of a double-humped fission barrier, was extended to light actinides by including the relations for the transmission coefficients through a complex triple-humped fission barrier. The real part of the fission barrier is parameterised as a function of the nucleus deformation by five smoothly joined parabolas. The imaginary potential is introduced only in the deformation range corresponding to the second well because the tertiary well is supposed to be shallow enough to neglect the damping of class III vibrational states. The transition states are assumed to be rotational states built on vibrational or non-collective band-heads. As the excitation energy increases, the shell effect, which causes the splitting of the outer barrier, diminishes and the outer humps lump into a single one. Therefore, in the present formalism, triple-humped barriers are associated only to the discrete transition states; the contribution of continuum to the fission coefficients is calculated considering a double-humped barrier. The parameters of the second single barrier equivalent with the outer humps are being determined from the condition of equal transmission coefficients. The saddle-point transition states in continuum are described by level densities (BCS below the critical energy and a modified version of Fermi Gas above) accounting for collective enhancements specific to the nuclear shape asymmetry at each saddle point . The neutron cross sections of 232Th in the
KENO-IV code benchmark calculation, (4)
International Nuclear Information System (INIS)
A series of benchmark tests has been undertaken in JAERI in order to examine the capability of JAERI's criticality safety evaluation system consisting of the Monte Carlo calculation code KENO-IV and the newly developed multi-group constants library MGCL. The present paper describes the results of a test using criticality experiments about slab-cylinder system of uranium nitrate solution. In all, 128 cases of experiments have been calculated for the slab-cylinder configuration with and without plexiglass reflector, having the various critical parameters such as the number of cylinders and height of the uranium nitrate solution. It is shown among several important results that the code and library gives a fairly good multiplication factor, that is, k sub(eff) -- 1.0 for heavily reflected cases, whereas k sub(eff) -- 0.91 for the unreflected ones. This suggests the necessity of more advanced treatment of the criticality calculation for the system where neutrons can easily leak out during slowing down process. (author)
The VADMAP code to calculate the SAF of photon
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A computer code VADMAP has been developed to calculate the Specific Absorbed Fraction, SAF, of photon. The development of the code is aimed at efficient and systematic preparation of the SAF data files for several different human phantoms in a suitable form as a direct input data file to DOSimetric DAta Calculation system, DOSDAC, which is being developed at Japan Atomic Energy Research Institute, JAERI. This document describes the methodology used in the code, the code structure, user's information including the way of implementing the code on FACOM/M-380, and the performance through calculation and preparation of the SAF data file. In order to show the performance of the code, a set of the SAF values for an adult human phantom was calculated and was organized to prepare the SAF file. Comparing the calculated SAF values with those tabulated in ORNL-5000, the quality of the code was examined. (author)
Coupling CFD code with system code and neutron kinetic code
Energy Technology Data Exchange (ETDEWEB)
Vyskocil, Ladislav, E-mail: Ladislav.Vyskocil@ujv.cz; Macek, Jiri
2014-11-15
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent.
Coupling CFD code with system code and neutron kinetic code
International Nuclear Information System (INIS)
Highlights: • Coupling interface between CFD code Fluent and system code Athlet was created. • Athlet code is internally coupled with neutron kinetic code Dyn3D. • Explicit coupling of overlapped computational domains was used. • A coupled system of Athlet/Dyn3D+Fluent codes was successfully tested on a real case. - Abstract: The aim of this work was to develop the coupling interface between CFD code Fluent and system code Athlet internally coupled with neutron kinetic code Dyn3D. The coupling interface is intended for simulation of complex transients such as Main Steam Line Break scenarios, which cannot be modeled separately first by system and neutron kinetic code and then by CFD code, because of the feedback between the codes. In the first part of this article, the coupling method is described. Explicit coupling of overlapped computational domains is used in this work. The second part of the article presents a demonstration simulation performed by the coupled system of Athlet/Dyn3D and Fluent. The “Opening a Steam Dump to the Atmosphere” test carried out at the Temelin NPP (VVER-1000) was simulated by the coupled system. In this simulation, the primary and secondary circuits were modeled by Athlet, mixing in downcomer and lower plenum was simulated by Fluent and heat generation in the core was calculated by Dyn3D. The results of the simulation with Athlet/Dyn3D+Fluent were compared with the experimental data and the results from a calculation performed with Athlet/Dyn3D without Fluent
Two-dimensional sensitivity calculation code: SENSETWO
International Nuclear Information System (INIS)
A SENSETWO code for the calculation of cross section sensitivities with a two-dimensional model has been developed, on the basis of first order perturbation theory. It uses forward neutron and/or gamma-ray fluxes and adjoint fluxes obtained by two-dimensional discrete ordinates code TWOTRAN-II. The data and informations of cross sections, geometry, nuclide density, response functions, etc. are transmitted to SENSETWO by the dump magnetic tape made in TWOTRAN calculations. The required input for SENSETWO calculations is thus very simple. The SENSETWO yields as printed output the cross section sensitivities for each coarse mesh zone and for each energy group, as well as the plotted output of sensitivity profiles specified by the input. A special feature of the code is that it also calculates the reaction rate with the response function used as the adjoint source in TWOTRAN adjoint calculation and the calculated forward flux from the TWOTRAN forward calculation. (author)
Calculation code revised MIXSET for Purex process
International Nuclear Information System (INIS)
Revised MIXSET is a FORTRAN IV calculation code developed to simulate steady and transient behaviors of the Purex extraction process and calculate the optimum operating condition of the process. Revised MIXSET includes all the functions of MIXSET code as shown below. a) Maximum chemical system of eight components can be handled with or without mutual dependence of the distribution of components. b) The flowrate and concentration of feed can be renewed successively at any state, transient or steady, for searching optimum operating conditions. c) Optimum inputs of feed concentrations and flowrates can be calculated to satisfy both of specification and recovery rate of a product. d) Radioactive decay reactions can be handled on each component. Besides these functions, the following chemical reactions concerned in Purex process are newly-included in Revised MIXSET code and the quantitative changes of components such as H+, U(IV), U(VI), Pu(III), Pu(IV), NH2OH, N2H4 can be simulated. 1st Gr. (i) reduction of Pu(IV); U4+ + 2Pu4+ + 2H2O → UO22+ + 2Pu3+ + 4H+. (ii) oxidation of Pu(III); 2Pu3+ + 3H+ + NO3- → 2Pu4+ + HNO2 + H2O. (iii) oxidation of U(IV); U4+ + NO3- + H2O → UO22+ + H+ + HNO2 2U4+ + O2 + 2H2O → 2UO22+ + 4H+. (iv) decomposition of HNO2; HNO2 + N2H5+ → HN3 + 2H2O + H+. (author)
International Nuclear Information System (INIS)
The lack of suitable benchmark problems makes it difficult to test sensitivity codes with a covariance library. A benchmark problem has therefore been defined for one- and two-dimensional sensitivity and uncertainity analysis codes and code systems. The problem, representative of a fusion reactor blanket, has a simple, three-zone /tau/-z geometry containing a D-T fusion neutron source distributed in a central void region surrounded by a thick 6LiH annulus. The response of interest is the 6Li tritium production per source neutron, T6. The calculation has been performed with SENSIBL using other codes from the AARE code system as a test of both SENSIBL and the linked, modular system. The caluclation was performed using the code system in the standard manner with a covariance data library in the COVFILS-2 format but modified to contain specifically tailored covariance data for H and 6Li (Path A). The calculation was also performed by a second method which uses specially perturbed H and Li cross sections (Path B). This method bypasses SENSIBL and allows a hand calculation of the benchmark T6 uncertainties. The results of Path A and Path B were total uncertainties in T6 of 0.21% and 0.19%, respectively. The closeness of the results for this challenging test gives confidence that SENSIBL and the AARE system will perform well for realistic sensitivity and uncertainty analyses
CRACKEL: a computer code for CFR fuel management calculations
International Nuclear Information System (INIS)
The CRACKLE computer code is designed to perform rapid fuel management surveys of CFR systems. The code calculates overall features such as reactivity, power distributions and breeding gain, and also calculates for each sub-assembly plutonium content and power output. A number of alternative options are built into the code, in order to permit different fuel management strategies to be calculated, and to perform more detailed calculations when necessary. A brief description is given of the methods of calculation, and the input facilities of CRACKLE, with examples. (author)
Spectroscopic calculation code ASPECT and its application
International Nuclear Information System (INIS)
The Code ASPECT is available for calculations of electronic levels of atoms and ions by the intermediate coupling scheme. This scheme is characterized by the simultaneous diagonalization of Hamiltonians for electronic repulsion, spin orbit interaction and crystal field effect. ASPECT performs the sorting of microstates involved in the electronic configuration in problem, calculation of matrix elements of these Hamiltonians, and diagonalization of the summed matrix. As input data, the calculation needs only parameter values of Slater integrals. ASPECT is also applied to calculate transition probabilities between the electronic levels obtained by this code. ASPECT is particularly focused on complex configurations containing f-electrons as met in Lanthanides and Actinides, which are not easily treated by an algebraic method. For convenience of users, Slater integral values for configurations fn of Lanthanides and Actinides are installed in the code so that users may select merely the atomic number. This document is composed of three parts. The first part (Chapter 1-3) describes quantum mechanical principles to calculate matrix elements of each unperturbed Hamiltonian and transition probabilities. The second part (Chapter 4) explains the structure of the code, and the last part (Chapter 5) serves as the manual for applications of this code, in which some samples are included. The third part (Chapter 6) is added as supplement for users who will improve this code. (author)
ASME Code Calculations for the CC Cryostat
Energy Technology Data Exchange (ETDEWEB)
Luther, R.D.; /Fermilab
1987-11-04
This engineering note contains the ASHE Code calculations for the CC Cryostat prepared by the manufacturer, Richmond-Lox Equipment Company. Most of these were taken from calculations initially prepared by Fermilab personne1and pub1ished in Eng. Note 68.
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.
International Nuclear Information System (INIS)
This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and parameter survey using developed analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. Evaporation calculations using cathode processor calculation code with distillation process, which was developed in 2000, were evaluated. By selecting proper input data (time step, mesh size etc.), the results showed that the present code agreed well for the evaporation rate of cadmium., and the capability of the distillation process design and simulation with the code has been confirmed. Parameter surveys using developed analytical model were performed for the purpose of reflection of cooling system design of the pyrochemical process cell. 4 cases of cooling flow patterns were surveyed at the normal and low flow rate conditions. From the result of parameter surveys, it was shown that the cooling pattern with direct cooling for heating facilities in the lower cell and balk cooling for upper cell is desirable. (author)
International Nuclear Information System (INIS)
Highlights: → The specifications required for the analyses of the destructive assay data taken from irradiated fuel in Ohi-1 and Ohi-2 PWRs were documented in this paper. → These data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. → These destructive assay data are suitable for the benchmarking of the burnup calculation code systems. - Abstract: The isotopic composition of spent nuclear fuels is vital data for studies on the nuclear fuel cycle and reactor physics. The Japan Atomic Energy Agency (JAEA) has been active in obtaining such data for pressurized water reactor (PWR) and boiling water reactor (BWR) fuels, and some data has already been published. These data have been registered with the international Spent Fuel Isotopic Composition Database (SFCOMPO) and widely used as international benchmarks for burnup calculation codes and libraries. In this paper, Assay Data of Spent Nuclear Fuel from two fuel assemblies irradiated in the Ohi-1 and Ohi-2 PWRs in Japan are shown. The destructive assay data from Ohi-2 have already been published. However, these data were not suitable for the benchmarking of calculation codes and libraries because several important specifications and data were not included. This paper summarizes the details of destructive assay data and specifications required for analyses of isotopic composition from Ohi-1 and Ohi-2. For precise burnup analyses, the burnup values of destructive assay samples were re-evaluated in this study. These destructive assay data were analyzed using the SWAT2.1 code, and the calculation results showed good agreement with experimental results. This indicates that the quality of destructive assay data from Ohi-1 and Ohi-2 PWRs is high, and that these destructive assay data are suitable for the benchmarking of burnup calculation code systems.
Calculation code MIXSET for Purex process
International Nuclear Information System (INIS)
MIXSET is a FORTRAN IV calculation code for Purex process that simulate the dynamic behavior of solvent extraction processes in mixer-settlers. Two options permit terminating dynamic phase by time or by achieving steady state. These options also permit continuing calculation successively using new inputs from a arbitrary phase. A third option permits artificial rapid close to steady state and a fourth option permits searching optimum input to satisfy both of specification and recovery rate of product. MIXSET handles maximum chemical system of eight components with or without mutual dependence of the distribution of the components. The chemical system in MIXSET includes chemical reactions and/or decaying reaction. Distribution data can be supplied by third-power polynominal equations or tables, and kinetic data by tables or given constants. The fluctuation of the interfacial level height in settler is converted into the flow rate changes of organic and aqueous stream to follow dynamic behavior of extraction process in detail. MIXSET can be applied to flowsheet study, start up and/or shut down procedure study and real time process management in countercurrent solvent extraction processes. (auth.)
Development of transient neutron transport calculation code
International Nuclear Information System (INIS)
A transient neutron transport code for time-dependent analyses of neutronics systems, named DOT4-T, has been developed. The code is based on the Discrete Ordinates code DOT4.2, which solves the steady-state neutron transport equation in two dimensions. For the discretization of time variable, a direct method, the fully implicit and unconditionally stable time integration scheme, has been employed. The resulting code has been tested using several one-dimensional and two-dimensional benchmark problems, and the results obtained with DOT4-T shows very satisfactory agreement with the benchmark problem results. (authors)
TEA: A Code Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2016-07-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. and Eriksson. It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature–pressure pairs. We tested the code against the method of Burrows & Sharp, the free thermochemical equilibrium code Chemical Equilibrium with Applications (CEA), and the example given by Burrows & Sharp. Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is available under a reproducible-research, open-source license via https://github.com/dzesmin/TEA.
International Nuclear Information System (INIS)
The possible return to power during the strong cooling down following a steam line break in the pressurized water reactors is a potential hazard. Among others, this transient can be characterized by the recriticality temperature which strongly depends on several three dimensional effects existing in the reactor core. This parameter can be evaluated by a best estimate neutronic code but an appropriate conservatism must be assured. Due to the nature of this transient the recriticality temperature is practically non-measurable and its uncertainty is hard to predict. Even so, some Russian exceptional measurements and appropriate core data referred in the literature can be found. In this paper our common practice is outlined, the measurements are presented and a simulation made by the Hungarian VVER-440 specific code system called KARATE-440 is described on the basis of the available plant data.
MCOR - Monte Carlo depletion code for reference LWR calculations
International Nuclear Information System (INIS)
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
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
International Nuclear Information System (INIS)
Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO2-1, BAPL-UO2-2 and BAPL-UO2-3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations of
Energy Technology Data Exchange (ETDEWEB)
Blazy-Aubignac, L
2007-09-15
The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)
Shielding calculational system for plutonium
International Nuclear Information System (INIS)
A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)
Energy Technology Data Exchange (ETDEWEB)
Mazurier, J
1999-05-28
This thesis has been performed in the framework of national reference setting-up for absorbed dose in water and high energy photon beam provided with the SATURNE-43 medical accelerator of the BNM-LPRI (acronym for National Bureau of Metrology and Primary standard laboratory of ionising radiation). The aim of this work has been to develop and validate different user codes, based on PENELOPE Monte Carlo code system, to determine the photon beam characteristics and calculate the correction factors of reference dosimeters such as Fricke dosimeters and graphite calorimeter. In the first step, the developed user codes have permitted the influence study of different components constituting the irradiation head. Variance reduction techniques have been used to reduce the calculation time. The phase space has been calculated for 6, 12 and 25 MV at the output surface level of the accelerator head, then used for calculating energy spectra and dose distributions in the reference water phantom. Results obtained have been compared with experimental measurements. The second step has been devoted to develop an user code allowing calculation correction factors associated with both BNM-LPRI's graphite and Fricke dosimeters thanks to a correlated sampling method starting with energy spectra obtained in the first step. Then the calculated correction factors have been compared with experimental and calculated results obtained with the Monte Carlo EGS4 code system. The good agreement, between experimental and calculated results, leads to validate simulations performed with the PENELOPE code system. (author)
Sample calculations on fuel rod behaviour during a LOCA with the code system SSYST-MOD 1
International Nuclear Information System (INIS)
The present paper shows results generated with SSYST, a program system developed for the analysis of the LWR fuel rod behaviour during a LOCA. A blowdown experiment in an out-of-pile test facility is analysed. The aim of the calculations is to demonstrate the influence of the various separate models, each describing a particular phenomenon such as rod internal pressure or rod mechanics on the behaviour of a hot rod by switching on these models sequentially. Calculations showed that the models presently included in the SSYST-system are able to describe the thermal and mechanical rod behaviour qualitatively in a correct way and that they may well be used to analyse the rod behaviour in a LWR during a LOCA. (Auth.)
International Nuclear Information System (INIS)
Corrosive conditions in the BWR primary cooling system are usually expressed by the corrosion index, ECP. In order to determine ECP at any location in the primary cooling system, ECP should be evaluated by computer simulation codes consisting of water radiolysis models to determine the concentrations of corrosive radiolytic species and mixed potential models to determine ECP based on corrosive species. Measures for mitigation of SCC crack growth rate by decreasing ECP are authorized by the JSME Standards; however, measures for mitigation of ECP by hydrogen addition have not been authorized yet. In the paper, standard procedures to authorize the computer simulation codes based on the verification and validation (V and V) method are proposed. The numerical justification of every code applied as a standard code should be verified and its accuracy and applicability for plant analysis should be validated. Benchmark analysis for verification procedures is proposed while a comparison of the calculated results with the measured ones for the evaluated plant is also proposed for the validation procedures. It is strongly recommended that the results of V and V evaluation of the codes that might be applied for evaluation of corrosive conditions in operating power plants are published in a peer-reviewed journal before their application. (author)
International Nuclear Information System (INIS)
The newly developed, advanced, light water reactor (LWR) simulation code, RELAP5, is used to analyze the response of Semiscale Mod-3 Test S-07-6. The objective of Test S-07-6 was to provide reference data to evaluate LWR integral blowdown, refill, and reflood behavior during a 200% cold leg break with emergency core coolant (ECC) injected into the intact loop cold leg. The calculated test results using RELAP5 illustrate many of the nonequilibrium and nonhomogeneous aspects of the ECC injection which are not directly observable in the test data. These results also demonstrate the capability of the RELAP5 code and compare well with the test data fo break flow, pressure, temperature, and density throughout the Semiscale Mod-3 system. The periodic depletion and replenishment of ECC water in the downcomer shown in the test data is also shown in the calculation
Exposure calculation code module for reactor core analysis: BURNER
International Nuclear Information System (INIS)
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules
RTP: Radionuclides inventories calculation using ORIGEN Code
International Nuclear Information System (INIS)
ORIGEN is a widely used computer code for calculating the buildup, decay, and processing of radioactive materials. The ORIGEN code was created by famous and reputable nuclear institution in United States, Oak Ridge National Laboratory (ORNL). For a nuclear reactor, either it is a nuclear power reactor or nuclear research reactor, the radionuclide inventories data is important. This data is acquired by performing source term modelling. A fresh nuclear fuel could not cause any harm to human. However, used nuclear fuel could pose danger threat to human. The fission products particularly long-lived radionuclides for example H-3, Co-60, Cs-137 that are generated inside the fuel yield a significance amount of radioactivity. Therefore, there is no doubt that for a facility having a nuclear reactor, it is vital to anticipate the amount of fission products inside the fuel together with the radioactivity that it may emit. Sufficient information on the radionuclide inventories allows the facility to provide adequate shielding protection and ensure safe transportation of nuclear fuel, when it is needed. This paper briefly describes application of ORIGEN code to calculate the radionuclides inventories of TRIGA-PUSPATI REACTOR (RTP) fuel. (author)
Data calculation program for RELAP 5 code
Energy Technology Data Exchange (ETDEWEB)
Silvestre, Larissa J.B.; Sabundjian, Gaiane, E-mail: larissajbs@usp.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2015-07-01
As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)
Data calculation program for RELAP 5 code
International Nuclear Information System (INIS)
As the criteria and requirements for a nuclear power plant are extremely rigid, computer programs for simulation and safety analysis are required for certifying and licensing a plant. Based on this scenario, some sophisticated computational tools have been used such as the Reactor Excursion and Leak Analysis Program (RELAP5), which is the most used code for the thermo-hydraulic analysis of accidents and transients in nuclear reactors. A major difficulty in the simulation using RELAP5 code is the amount of information required for the simulation of thermal-hydraulic accidents or transients. The preparation of the input data leads to a very large number of mathematical operations for calculating the geometry of the components. Therefore, a mathematical friendly preprocessor was developed in order to perform these calculations and prepare RELAP5 input data. The Visual Basic for Application (VBA) combined with Microsoft EXCEL demonstrated to be an efficient tool to perform a number of tasks in the development of the program. Due to the absence of necessary information about some RELAP5 components, this work aims to make improvements to the Mathematic Preprocessor for RELAP5 code (PREREL5). For the new version of the preprocessor, new screens of some components that were not programmed in the original version were designed; moreover, screens of pre-existing components were redesigned to improve the program. In addition, an English version was provided for the new version of the PREREL5. The new design of PREREL5 contributes for saving time and minimizing mistakes made by users of the RELAP5 code. The final version of this preprocessor will be applied to Angra 2. (author)
International Nuclear Information System (INIS)
The project encompasses the following project tasks and problems: (1) Studies relating to complete failure of the main heat transfer system; (2) Pebble flow; (3) Development of computer codes for detailed calculation of hypothetical accidents; (a) the THERMIX/RZKRIT temperature buildup code (covering a.o. a variation to include exothermal heat sources); (b) the REACT/THERMIX corrosion code (variation taking into account extremely severe air ingress into the primary loop); (c) the GRECO corrosion code (variation for treating extremely severe water ingress into the primary loop); (d) the KIND transients code (for treating extremely fast transients during reactivity incidents. (4) Limiting devices for safety-relevant quantities. (5) Analyses relating to hypothetical accidents. (a) hypothetical air ingress; (b) effects on the fuel particles induced by fast transients. The problems of the various tasks are defined in detail and the main results obtained are explained. The contributions reporting the various project tasks and activities have been prepared for separate retrieval from the database. (orig./HP)
Electrical Conductivity Calculations from the Purgatorio Code
Energy Technology Data Exchange (ETDEWEB)
Hansen, S B; Isaacs, W A; Sterne, P A; Wilson, B G; Sonnad, V; Young, D A
2006-01-09
The Purgatorio code [Wilson et al., JQSRT 99, 658-679 (2006)] is a new implementation of the Inferno model describing a spherically symmetric average atom embedded in a uniform plasma. Bound and continuum electrons are treated using a fully relativistic quantum mechanical description, giving the electron-thermal contribution to the equation of state (EOS). The free-electron density of states can also be used to calculate scattering cross sections for electron transport. Using the extended Ziman formulation, electrical conductivities are then obtained by convolving these transport cross sections with externally-imposed ion-ion structure factors.
Energy Technology Data Exchange (ETDEWEB)
Guerrero Araque, J. E.; Mendez Villafane, R.
2013-07-01
This work intends to heavily describe simulation steps used in code MCNPX for calculation for Neutron response of a BSS with passive or active detector. Has it been calculated with MCNPX the matrix response of a system of Bonner spheres, with passive or active detector, which described in detail the steps to be followed by the code are part of the solution. (Author)
Coupled code calculation of rod withdrawal at power accident
Energy Technology Data Exchange (ETDEWEB)
Grgić, Davor, E-mail: davor.grgic@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Benčik, Vesna, E-mail: vesna.bencik@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Šadek, Siniša, E-mail: sinisa.sadek@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia)
2013-08-15
Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.
Coupled code calculation of rod withdrawal at power accident
International Nuclear Information System (INIS)
Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced
User's guide to the SEPHIS computer code for calculating the Thorex solvent extraction system
Energy Technology Data Exchange (ETDEWEB)
Watson, S.B.; Rainey, R.H.
1979-05-01
The SEPHIS computer program was developed to simulate the countercurrent solvent extraction process. The code has now been adapted to model the Acid Thorex flow sheet. This report represents a practical user's guide to SEPHIS - Thorex containing a program description, user information, program listing, and sample input and output.
ESCADRE and ICARE code systems
International Nuclear Information System (INIS)
The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab
FORTRAN code-evaluation system
Capps, J. D.; Kleir, R.
1977-01-01
Automated code evaluation system can be used to detect coding errors and unsound coding practices in any ANSI FORTRAN IV source code before they can cause execution-time malfunctions. System concentrates on acceptable FORTRAN code features which are likely to produce undesirable results.
International Nuclear Information System (INIS)
The KDF9/EGDON programme FADDEEV has been written to investigate a technique for the calculation of the matrix of frequency responses G(jw) describing the response of the output vector y from the multivariable differential/algebraic system S to the drive of the system input vector u. S: Ex = Ax + Bu, y = Cx, G(jw) = C(jw E - A )-1 B. The programme uses an algorithm due to Faddeev and has been written with emphasis upon: (a) simplicity of programme structure and computational technique which should enable a user to find his way through the programme fairly easily, and hence facilitate its manipulation as a subroutine in a larger code; (b) rapid computational ability, particularly in systems with fairly large number of inputs and outputs and requiring the evaluation of the frequency responses at a large number of frequencies. Transport or time delays must be converted by the user to Pade or Bode approximations prior to input. Conditions under which the algorithm fails to give accurate results are identified, and methods for increasing the accuracy of the calculations are discussed. The conditions for accurate results using FADDEEV indicate that its application is specialized. (author)
A code to calculate multigroup constants for fast neutron reactor
International Nuclear Information System (INIS)
KQCS-2 code is a new improved version of KQCS code, which was designed to calculate multigroup constants for fast neutron reactor. The changes and improvements on KQCS are described in this paper. (author)
International Nuclear Information System (INIS)
In order to evaluate core characteristics of fast reactors, a computer code system ARCADIAN-FBR has been developed by utilizing the existing analysis codes and the latest nuclear data library JENDL-3.3. The validity of ARCADIAN-FBR was verified by using the experimental data obtained in the MONJU core physics tests. The results of analyses are in good agreement with the experimental data and the applicability of ARCADIAN-FBR for fast reactor core analysis is confirmed. Using ARCADIAN-FBR, the sodium void reactivity worth, which is an important parameter in the safety analysis of fast reactors, was analyzed for MONJU core. 241Pu in the core fuel is transmuted to 241Am due to disintegrations. Therefore, the effect of 241Am accumulation on the sodium void reactivity worth was evaluated for MONJU core. As a result of calculation, it was confirmed that the accumulation of 241Am significantly influences on the sodium void reactivity worth and hence on the safety analysis of sodium-cooled fast reactors. (author)
International Nuclear Information System (INIS)
This work is devoted to the qualification of the reactor codes in comparison with experiments. The first part deals with the interpretation by CEA's calculational tools of critical experiments in uranium-plutonium-light water lattices. In the second part, a thorough analysis of the start-up experiments of a pressurized light water power reactor is presented. The calculation was made with both FRAMATOME and CEA's tools, which required the development of an interface code
Accumulative Landings System Code Tables
National Oceanic and Atmospheric Administration, Department of Commerce — Code Tables Used In Landings System. These tables assign meanings to the codes that appear in the data tables. Code tables exist for species, gear, state, county,...
International Nuclear Information System (INIS)
Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNECFD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements
MOx benchmark calculations by deterministic and Monte Carlo codes
International Nuclear Information System (INIS)
Highlights: ► MOx based depletion calculation. ► Methodology to create continuous energy pseudo cross section for lump of minor fission products. ► Mass inventory comparison between deterministic and Monte Carlo codes. ► Higher deviation was found for several isotopes. - Abstract: A depletion calculation benchmark devoted to MOx fuel is an ongoing objective of the OECD/NEA WPRS following the study of depletion calculation concerning UOx fuels. The objective of the proposed benchmark is to compare existing depletion calculations obtained with various codes and data libraries applied to fuel and back-end cycle configurations. In the present work the deterministic code NEWT/ORIGEN-S of the SCALE6 codes package and the Monte Carlo based code MONTEBURNS2.0 were used to calculate the masses of inventory isotopes. The methodology to apply the MONTEBURNS2.0 to this benchmark is also presented. Then the results from both code were compared.
Rotamak equilibrium calculations using the PEST code
International Nuclear Information System (INIS)
This report describes the use of the equilibrium part of the Princeton equilibrium and stability code PEST to model rotamak equilibria with an applied toroidal magnetic field. An overview of the code is provided, together with a list of required input data. The simulation of a range of equilibria measured in the ANSTO rotamak shows that the rotamak approximately satisfies magnetohydrodynamic equilibrium. Of particular interest is the presence of large diamagnetic poloidal current about the magnetic axis which produces a peak in the plasma pressure on the magnetic axis. For a low toroidal field, however, poloidal current of opposite direction is simultaneously driven on flux surfaces distant from the magnetic axis, producing paramagnetism
Presentation of the NABE calculation code
International Nuclear Information System (INIS)
The purpose of the NABE code is to follow up the physical and chemical parameters of a concrete cell when large amounts of sodium are released by accident in the reactor building. The code contains several modules: behaviour of the concrete, heating-up, water and CO2 release, reactions between liquid and vapor sodium with the concrete breakdown products, combustion of sodium and hydrogen (water), power released by the fission products in the liquid, the gas and the concrete, behaviour of the atmosphere of the cell (pressure, temperature, leaks), sodium vaporization and boiling, and condensation on the cold walls
CONDOR: neutronic code for fuel elements calculation with rods
International Nuclear Information System (INIS)
CONDOR neutronic code is used for the calculation of fuel elements formed by fuel rods. The method employed to obtain the neutronic flux is that of collision probabilities in a multigroup scheme on two-dimensional geometry. This code utilizes new calculation algorithms and normalization of such collision probabilities. Burn-up calculations can be made before the alternative of applying variational methods for response flux calculations or those corresponding to collision normalization. (Author)
Gao, Wen
2015-01-01
This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
International Nuclear Information System (INIS)
In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data
Xenon poisoning calculation code for miniature neutron source reactor (MNSR)
Institute of Scientific and Technical Information of China (English)
无
2001-01-01
In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.
CONSUL code package application for LMFR core calculations
International Nuclear Information System (INIS)
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Harrington, Joseph; Bowman, M. Oliver
2015-01-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method o...
Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
Energy Technology Data Exchange (ETDEWEB)
Aleshin, Sergey S.; Gorodkov, Sergey S.; Shcherenko, Anna I. [Nuclear Research Centre ' ' Kurchatov Institute' ' , Moscow (Russian Federation)
2015-09-15
A burn-up calculation of VVER's cores by Monte-Carlo code is complex process and requires large computational costs. This fact makes Monte-Carlo codes usage complicated for project and operating calculations. Previously prepared isotopic compositions are proposed to use for the Monte-Carlo code (MCU) calculations of different states of VVER's core with burnt fuel. Isotopic compositions are proposed to calculate by an approximation method. The approximation method is based on usage of a spectral functionality and reference isotopic compositions, that are calculated by engineering codes (TVS-M, PERMAK-A). The multiplication factors and power distributions of FA and VVER with infinite height are calculated in this work by the Monte-Carlo code MCU using earlier prepared isotopic compositions. The MCU calculation data were compared with the data which were obtained by engineering codes.
Calculation code PULCO for Purex process in pulsed column
International Nuclear Information System (INIS)
The calculation code PULCO, which can simulate the Purex process using a pulsed column as an extractor, has been developed. The PULCO is based on the fundamental concept of mass transfer that the mass transfer within a pulsed column occurs through the interface of liquid drops and continuous phase fluid, and is the calculation code different from conventional ones, by which various phenomena such as the generation of liquid drops, their rising and falling, and the unification of liquid drops actually occurring in a pulsed column are exactly reflected and can be correctly simulated. In the PULCO, the actually measured values of the fundamental quantities representing the extraction behavior of liquid drops in a pulsed column are incorporated, such as the mass transfer coefficient of each component, the diameter and velocity of liquid drops in a pulsed column, the holdup of dispersed phase, and axial turbulent flow diffusion coefficient. The verification of the results calculated with the PULCO was carried out by installing a pulsed column of 50 mm inside diameter and 2 m length with 40 plate stage in a glove box for unirradiated uranium-plutonium mixed system. The results of the calculation and test were in good agreement, and the validity of the PULCO was confirmed. (Kako, I.)
Electronic transport calculations in the ONETEP code: Implementation and applications
Bell, Robert A.; Dubois, Simon M.-M.; Payne, Michael C.; Mostofi, Arash A.
2015-08-01
We present an approach for computing Landauer-Büttiker ballistic electronic transport for multi-lead devices containing thousands of atoms. The method is implemented in the ONETEP linear-scaling density-functional theory code and uses matrix elements calculated from first-principles. Using a compact yet accurate basis of atom-centred non-orthogonal generalised Wannier functions that are optimised in situ to their unique local chemical environment, the transmission and related properties of very large systems can be calculated efficiently and accurately. Other key features include the ability to simulate devices with an arbitrary number of leads, to compute eigenchannel decompositions, and to run on highly parallel computer architectures. We demonstrate the scale of the calculations made possible by our approach by applying it to the study of electronic transport between aligned carbon nanotubes, with system sizes up to 2360 atoms for the underlying density-functional theory calculation. As a consequence of our efficient implementation, computing electronic transport from first principles in systems containing thousands of atoms should be considered routine, even on relatively modest computational resources.
International Nuclear Information System (INIS)
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)
2007-09-15
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.
Development and validation of a nodal code for core calculation
International Nuclear Information System (INIS)
The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency
Energy Technology Data Exchange (ETDEWEB)
Jan, S.; Laedermann, J.P.; Bochud, F.; Ferragut, A.; Bordy, J.M.; Parisi, L.L.; Abou-Khalil, R.; Longeot, M.; Kitsos, S.; Groetz, J.E.; Villagrasa, C.; Daures, J.; Martin, E.; Henriet, J.; Tsilanizara, A.; Farah, J.; Uyttenhove, W.; Perrot, Y.; De Carlan, L.; Vivier, A.; Kodeli, I.; Sayah, R.; Hadid, L.; Courageot, E.; Fritsch, P.; Davesne, E.; Michel, X.
2010-07-01
This document gathers the slides of the available presentations given during these conference days. Twenty seven presentations are assembled in the document and deal with: 1 - GATE: calculation code for medical imaging, radiotherapy and dosimetry (S. Jan); 2 - estimation of conversion factors for the measurement of the ambient dose equivalent rate by in-situ spectroscopy (J.P. Laedermann); 3 - geometry specific calibration factors for nuclear medicine activity meters (F. Bochud); 4 - Monte Carlo simulation of a rare gases measurement system - calculation and validation, ASGA/VGM system (A. Ferragut); 5 - design of a realistic radiation field for the calibration of the dosemeters used in interventional radiology/cardiology (medical personnel dosimetry) (J.M. Bordy); 6 - determination of the position and height of the KALINA facility chimney at CEA Cadarache (L.L. Parisi); 7 - MERCURAD{sup TM} - 3D simulation software for dose rates calculation (R. Abou-Khalil); 8 - PANTHERE - 3D software for gamma dose rates simulation of complex nuclear facilities (M. Longeot); 9 - radioprotection, from the design to the exploitation of radioactive materials transportation containers (S. Kitsos); 10 - post-simulation processing of MCNPX responses in neutron spectroscopy (J.E. Groetz); 11 - last developments of the Geant4 Monte Carlo code for trace amounts simulation in liquid water at the molecular scale (C. Villagrasa); 12 - Calculation of H{sub p}(3)/K{sub air} conversion coefficients using PENELOPE Monte-Carlo code and comparison with MCNP calculation results (J. Daures); 13 - artificial neural networks, a new alternative to Monte Carlo calculations for radiotherapy (E. Martin); 14 - use of case-based reasoning for the reconstruction and handling of voxelized fantoms (J. Henriet); 15 - resolution of the radioactive decay inverse problem for dose calculation in radioprotection (A. Tsilanizara); 16 - use of NURBS-type fantoms for the study of the morphological factors influencing
Burnup calculations using serpent code in accelerator driven thorium reactors
International Nuclear Information System (INIS)
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 232Th and mixed 233U 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.)
Calculations of angular momentum coupling coefficients on a computer code
International Nuclear Information System (INIS)
In this study, Clebsch-Gordan coefficients, 3j symbols, Racah coefficients, Wigner's 6j and 9j symbols were calculated by a prepared computer code of COEFF. The computer program COEFF is described which calculates angular momentum coupling coefficients and expresses them as quotient of two integers multiplied by the square root of the quotient of two integers. The program includes subroutines to encode an integer into its prime factors, to decode of prime factors back into an integer , to perform basic arithmetic operations on prime-coded numbers, as well as subroutines which calculate the coupling coefficients themselves. The computer code COEFF had been prepared to run on a VAX. In this study we rearranged the code to run on PC and tested it successfully. The obtained values in this study, were compared with the values of other computer programmes. A pretty good agreement is obtained between our prepared computer code and other computer programmes
TEA: A Code for Calculating Thermochemical Equilibrium Abundances
Blecic, Jasmina; Bowman, M Oliver
2015-01-01
We present an open-source Thermochemical Equilibrium Abundances (TEA) code that calculates the abundances of gaseous molecular species. The code is based on the methodology of White et al. (1958) and Eriksson (1971). It applies Gibbs free-energy minimization using an iterative, Lagrangian optimization scheme. Given elemental abundances, TEA calculates molecular abundances for a particular temperature and pressure or a list of temperature-pressure pairs. We tested the code against the method of Burrows & Sharp (1999), the free thermochemical equilibrium code CEA (Chemical Equilibrium with Applications), and the example given by White et al. (1958). Using their thermodynamic data, TEA reproduces their final abundances, but with higher precision. We also applied the TEA abundance calculations to models of several hot-Jupiter exoplanets, producing expected results. TEA is written in Python in a modular format. There is a start guide, a user manual, and a code document in addition to this theory paper. TEA is ...
Energy Technology Data Exchange (ETDEWEB)
Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu
2000-07-01
The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)
Calculation codes in radiation protection, radiation physics and dosimetry
International Nuclear Information System (INIS)
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
Revisiting the calculation of effective free distance of turbo codes
Chatzigeorgiou, I; Wassell, I. J.
2008-01-01
The expression for the minimum Hamming weight of the output of a constituent convolutional encoder, when its input is a weight-2 sequence is revisited. The new expression particularly facilitates the calculation of the effective free distance of recently proposed schemes, namely non-systematic turbo codes and pseudo-randomly punctured turbo codes.
Burnup calculation methodology in the serpent 2 Monte Carlo code
International Nuclear Information System (INIS)
This paper presents two topics related to the burnup calculation capabilities in the Serpent 2 Monte Carlo code: advanced time-integration methods and improved memory management, accomplished by the use of different optimization modes. The development of the introduced methods is an important part of re-writing the Serpent source code, carried out for the purpose of extending the burnup calculation capabilities from 2D assembly-level calculations to large 3D reactor-scale problems. The progress is demonstrated by repeating a PWR test case, originally carried out in 2009 for the validation of the newly-implemented burnup calculation routines in Serpent 1. (authors)
Description of the CAREM Reactor Neutronic Calculation Codes
International Nuclear Information System (INIS)
In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included
Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation
International Nuclear Information System (INIS)
Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)
Validation of BGCore System for Burnup Calculations
International Nuclear Information System (INIS)
BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. BGCore interfaces the Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating fuel composition during irradiation and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. The objective of this study is validation of the BGCore system against well established and verified, state of the art computer codes for thermal and fast spectrum lattices
Energy Technology Data Exchange (ETDEWEB)
NONE
2003-07-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
International Nuclear Information System (INIS)
A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)
CONSUL code package application for LMFR core calculations
Energy Technology Data Exchange (ETDEWEB)
Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)
2008-07-01
CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)
Verification test calculations for the Source Term Code Package
International Nuclear Information System (INIS)
The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs
Validation of the Monteburns code for criticality calculation of TRIGA reactors
Energy Technology Data Exchange (ETDEWEB)
Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)
2002-07-01
Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)
Hot zero power reactor calculations using the Insilico code
Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.
2016-06-01
In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.
Calculation of doppler coefficient of reactivity by WIMS code
International Nuclear Information System (INIS)
The Doppler coefficient of reactivity is an important factor in prediction of several transients in light water reactors. Some of the past studies raised the question about the 10% uncertainty that traditionally was taken in calculations of Doppler coefficient by LWR lattice code. In order to bridge the gap of lack of accurate benchmark problem to evaluate the accuracy of Doppler effect, Mosteller et al. proposed a computational benchmark problem of Doppler coefficient to evaluate the accuracy and consistency of LWR lattice physics code. In this paper we present the results obtained from WIMS-D4 lattice code and compare it with those obtained by CELL-2 lattice code part of the EPRI-PRESS reactor physics package. The results obtained from the Monte Carlo code MCNP-3A served as reference for both cases, and was taken from ref 1. (authors). 4 refs., 2 figs., 1 tab
The MCEF code for nuclear evaporation and fission calculations
International Nuclear Information System (INIS)
We present an object oriented algorithm, written in the Java programming language, which performs a Monte Carlo calculation of the evaporation-fission process taking place inside an excited nucleus. We show that this nuclear physics problem is very suited for the object oriented programming by constructing two simple objects: one that handles all nuclear properties and another that takes care of the nuclear reaction. The MCEF code was used to calculate important results for nuclear reactions, and here we show examples of possible uses for this code. (author)
Full-core pin-power calculations using Monte Carlo codes
International Nuclear Information System (INIS)
Pin wise calculations of core power distribution have been performed for a criticality mock up installation that models a WWER-1000 reactor. Two Monte Carlo codes have been applied for solving of this problem: the MCNP4B code and the KENO-VI code from the SCALE 4.4 system. The codes use different kinds of neutron cross section data: pointwise continuous-energy ENDF/B-VI data and multigroup ENDF/B-V data. Comparisons of calculated results show that the MCNP4B and KENO-VI results are in good agreement. (authors)
International Nuclear Information System (INIS)
Aiming at validation of depletion codes against WWER-440 spent fuel data some calculations of isotope composition of WWER-440 spent fuel assembly have been carried out by the NESSEL-NUKO code system. The initial data and data for the comparisons were taken from the ISTS burn up credit project data, recently published in the ISTC report 'Radiochemical Assays of Irradiated WWER-440 Fuel for Use in Spent Fuel Burnup Credit Activities. The specific work scope included the destructive assay (DA) of spent fuel assembly rod segments with an - -38.5 MWd/KgU burn up from a single WWER-440 fuel assembly from the Novovorenezh reactor in Russia (Authors)
Data processing code system for foil experiments
International Nuclear Information System (INIS)
A code system has been developed for an efficient measurement of reaction rates in foil irradiation experiments. The code system consists of four codes, namely of, (i) setting up experimental parameters and collecting γ-ray spectrum data, (ii) analysing γ-ray spectrum, (iii) calculating reaction rate distributions, and (iv) furnishing utility programs. This code system provides a useful tool of data processing of irradiated foil to obtain the γ-ray spectrum and the reaction rate distribution. These procedures can be executed automatically. The routine for processing foil counting data covers the following functions : the data smoothing, the peak searching by means of the first and second derivative methods, and the determination of the photo peak area and its error with use of a functional fitted by a non-linear least squares method. The code for reaction rate calculation has the following functions : the determination of decay constants of each isotope by using decay data of foil counting and the calculation of reaction rates after correcting irradiation time and weight of a foil. These codes are written by FORTRAN-77 for mini-computer PDP-11/44 (DEC), of which the maximum program memory size is limited to 32k bytes. (author)
International Nuclear Information System (INIS)
Recently several code systems have been developed and utilized at Japan Atomic Energy Research Institute. It is not easy to construct and maintain a code system, but the fact is not well known in the researchers. For this reason, it will be very useful to publish informations about design concepts, characteristics, necessary computer features and amounts of invested manpower for the developments of some recent JAERI code systems. In this report, a general view of required manpower on unification of nuclear codes is discussed and four code systems, i.e., SRAC for thermal reactor analysis, TRITON for Tokamak MHD analysis, SPEEDI for emergent environmental dose prediction and RADHEAT for radiation shielding analysis are presented. They are described from aspects of (1) purpose and schedule of development, (2) outline of system, (3) results of benchmark tests, (4) utilized computer features, (5) invested manpower, and (6) desirable computer features. Finally common aspects of four code systems from viewpoint of necessary computer hardwares and softwares are discussed for future development of code systems. (author)
SLAROM: a code for cell homogenization calculation of fast reactor
International Nuclear Information System (INIS)
A revised version of the SLAROM code has been developed. The main function of SLAROM is to perform the cell homogenization calculation of a fast power reactor and a fast critical assembly. The code uses the JFS2 or JFS3 type cross section set as a multi-group cross section library. The region dependent effective cross sections are calculated by taking account of the heterogeneity effect of resonance shielding for heavy nuclides. The integral transport equations are solved by using the collision probability method. SLAROM installs collision probability calculation routines for various geometries encountered in a fast reactor analysis. The effective multiplication factor (ksub(eff)) calculation or buckling search mode is available. The cell homogenized cross sections are obtained by weighting with the fine structure flux and volumes. The calculation of anisotropic diffusion coefficient is based on the Benoist's definition with use of the directional collision probability. The averaged macroscopic and microscopic cross sections are saved on the Partitioned Data Set file with a unified format. In addition to the cell calculation, another module is equipped to solve one dimensional diffusion equations in normal and adjoint modes. The fluxes obtained by this module can be used to collapse the fine group cross sections into the broad group structure. The perturbation calculation is also available. This report describes the calculational method adopted in the SLAROM code, input data and job control statements instructions, structure of the code, file requirement and sample input and output data. Since the input data are punched in a free format, users will be easy to prepare them. The description of auxiliary programs is given in Appendix for a help of the data handling on the PDS file. (author)
The best estimate codes applied to VVER calculations validation methodology
International Nuclear Information System (INIS)
The best estimate thermal hydraulic codes for PWRs and BWRs accident analysis were elaborated in 80-th years. The main goal of the best estimate codes (BEC) calculations was to obtain the real picture of the reactor facilities parameters changes during transient and accident regimes. This codes were based on 5--6 equations mathematical models taking into account separate flows of water and vapor. Validation of the best estimate thermal hydraulic codes applied VVER calculations is the main objective of the article. The concept of CSNI test matrices application to the codes validation is presented. The ways of the matrices improvement for this purpose is outlined. Taking into account deficiency of the operating integral test facilities a special attention is focused on the original separate effects experimental data obtained in Ukraine and which was not included to the CSNI matrices. To facilitate application of the obtained data to the codes validation the numerical experiment method was elaborated. The example of the method application is described
Blind Recognition Algorithm of Turbo Codes for Communication Intelligence Systems
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Ali Naseri
2011-11-01
Full Text Available Turbo codes are widely used in land and space radio communication systems, and because of complexity of structure, are custom in military communication systems. In electronic warfare, COMINT systems make attempt to recognize codes by blind ways. In this Paper, the algorithm is proposed for blind recognition of turbo code parameters like code kind, code-word length, code rate, length of interleaver and delay blocks number of convolution code. The algorithm calculations volume is0.5L3+1.25L, therefore it is suitable for real time systems.
International Nuclear Information System (INIS)
Relief valves are used in industrial plants for instance as safeguarding of the system pressure in case of pump failures. Pump failures and automatic changeovers to redundant aggregates cause a flow reversal which induces the stop valve closure. This process can cause water hammer effects in the piping system. The backflow velocity defines the maximum load in the piping system. The presented approach taking into account this effect of medium displacement in the RELAP calculations yields significant differences to the former results. Validation using experimental data is required.
Calculation of SPERT Reactor benchmarks using 3D diffusion code DIREN
International Nuclear Information System (INIS)
The three dimensional diffusion code DIREN was developed at Institute for Nuclear Research (INR) Pitesti for reactor physics calculations for natural uranium and advanced CANDU reactors. Cell codes used are WIMS (from NEA library) and DRAGON (available in open source system). The latter is used also for super cell modeling of reactor control devices. These codes and the auxiliary programs were linked together in a calculation system. In order to apply WIMS-DRAGON-DIREN system to LWR, first the reactor SPERT benchmarks problems were calculated. The core including the control rods was modeled in three dimensional geometry. Following the calculations of the critical height (Hcrit), three dimensional power and flux distributions were obtained. The standard procedure used for CANDU reactor calculations (incremental cross sections for reactivity devices) underestimated the worth of control rods. A simple procedure to obtain the internal boundary conditions was developed using the super cell code DRAGON. Also the DIREN 3D diffusion code was modified to apply inner boundary conditions at control rods assigned volumes. Applying the inner boundary conditions yielded results closer to the measured values (e.g. the measured Hcrit was 49.53 cm as compared to 53.15 cm, the calculated one on 7 groups for nominal temperature). The reactivity coefficients for temperature and density required in transient's simulations were also calculated. The sample test problem T83 (hot stand-by, fast transient) was simulated using the RELAP code. (authors)
Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART
Hursin, Mathieu
2010-01-01
The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, bu...
Verification of the COCAGNE core code using cluster depletion calculations
International Nuclear Information System (INIS)
EDF/R and D is developing a new calculation scheme based on the transport- Simplified Pn (SPn) approach. The lattice code used is the deterministic code APOLLO2, developed at CEA with the support of EDF and AREVA-NP. The core code is the code COCAGNE, developed at EDF R and D. The latter can take advantage of a microscopic depletion solver which improves the treatment of spectral history effects. This solver can resort to a specific correction based on the use of the Pu239 concentration as a spectral indicator. In order to evaluate the improvements brought by this Pu239 correction model, one uses (3x3 assemblies) cluster depletion calculations as test-cases. UOX and UOX/MOX clusters are both considered. As a reference, APOLLO2 depletion calculations of these clusters, using a critical boron (CB) search scheme at each calculation step, are performed. This choice of methodology (using CB search instead of a fixed average CB) enables to highlight historical spectral effects related to the boron concentration. This methodology is also more consistent with the depletion calculation of real cores. Pin by pin COCAGNE calculations are performed and compared with the APOLLO2 results. The analysis of the results obtained shows that the boron concentration computed by COCAGNE gets more consistent with APOLLO2 when the Pu239 corrector is used, especially for UOX/MOX clusters. As for pin power distribution, the use of the Pu239 model also enables to reduce slightly the gap between APOLLO2 and COCAGNE. This work will be extended to clusters with gadolinium-poisoned fuel assemblies and reflector regions. (author)
A new neutronics analysis code system for fast reactors and validation
International Nuclear Information System (INIS)
A new neutronics analysis code system has been developed for detailed analysis of fast reactor cores. The code system is composed of a calculation code of effective cross sections, an assembly calculation code based on the method of characteristics, and a full core transport/diffusion calculation code. The validity of the code system is investigated by applying it to the prototype fast reactor Monju, and by comparing the calculation results with measured ones. (author)
Physical Model and Calculation Code for Fuel Coolant Interactions
International Nuclear Information System (INIS)
The base of the physical model of the FCI in discussion are the shock tube experiments performed with UO2 and Na. The experiments demonstrated that the process of 'vapour explosion' in constrained configurations consists of successive cycles. According to the experimental results, the interaction model described here divides each cycle into three phases: Fuel-Coolant-Contact (Phase A), Ejection and reentry of the coolant (Phase B), Impact and Fragmentation (Phase C). The results of some calculations, performed for different values of the system pressure resp. of the coolant bulk temperature, are compiled in plots displaying the series of successive ejection events of an interaction; the plots represent the ejection height and the corresponding pressure in the vapour volume versus time. The pressure peaks mark the impact pressure pulse resp. the vapour pressure immediately after contact, which-ever is greater; the acoustic pressure peak of Phase A has not been plotted. The third and the following cycles reveal the oscillating reentry behaviour. A physical model has been proposed to describe fuel coolant interactions in shock-tube geometry. The corresponding code is presently in an advanced state of development. A principal feature, of the code is the consistent application of the Fourier-equation throughout the whole interaction process; this reveals some peculiarities which do not become evident otherwise. The two dimensional representation of the coolant flow provides the basis for a fragmentation mechanism, thus relieving from the necessity to introduce the fragmentation as an input parameter. These features may indicate a step towards a more realistic comprehension of the subject
WIPP Benchmark calculations with the large strain SPECTROM codes
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This report provides calculational results from the updated Lagrangian structural finite-element programs SPECTROM-32 and SPECTROM-333 for the purpose of qualifying these codes to perform analyses of structural situations in the Waste Isolation Pilot Plant (WIPP). Results are presented for the Second WIPP Benchmark (Benchmark II) Problems and for a simplified heated room problem used in a parallel design calculation study. The Benchmark II problems consist of an isothermal room problem and a heated room problem. The stratigraphy involves 27 distinct geologic layers including ten clay seams of which four are modeled as frictionless sliding interfaces. The analyses of the Benchmark II problems consider a 10-year simulation period. The evaluation of nine structural codes used in the Benchmark II problems shows that inclusion of finite-strain effects is not as significant as observed for the simplified heated room problem, and a variety of finite-strain and small-strain formulations produced similar results. The simplified heated room problem provides stratigraphic complexity equivalent to the Benchmark II problems but neglects sliding along the clay seams. The simplified heated problem does, however, provide a calculational check case where the small strain-formulation produced room closures about 20 percent greater than those obtained using finite-strain formulations. A discussion is given of each of the solved problems, and the computational results are compared with available published results. In general, the results of the two SPECTROM large strain codes compare favorably with results from other codes used to solve the problems
Decay heat calculation an international nuclear code comparison
International Nuclear Information System (INIS)
The results of an international code comparison on decay heat are presented and discussed. Participants from more than ten laboratories calculated, using the same input data, decay heat for thirteen cooling times between 1 and 1013 sec. Two irradiation cases were proposed: fission pulse and 3x107 seconds of irradiation of 235U fuel. The results are analysed and compared. This inter-comparison shows that, if the same input data are given, most of the codes give very similar results for the decay heat and consequently also for the fission product contribution
International Nuclear Information System (INIS)
The development of a high-accuracy reactor benchmark analysis capability is described. This capability has been incorporated into a revised and extended version of the lattice analysis program HAMMER. Previous analyses using the HAMMER program required the introduction of correction factors obtained from more rigorous treatments of various effects such as resonance capture and neutron leakage. The present version of the program will remove the ambiguities associated with the introduction of such correction factors by optionally performing the more rigorous calculations internally or by automating the correctional procedure
Verification calculations for the WWER version of the TRANSURANUS code
International Nuclear Information System (INIS)
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
SPECTRAL AMPLITUDE CODING OCDMA SYSTEMS USING ENHANCED DOUBLE WEIGHT CODE
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F.N. HASOON
2006-12-01
Full Text Available A new code structure for spectral amplitude coding optical code division multiple access systems based on double weight (DW code families is proposed. The DW has a fixed weight of two. Enhanced double-weight (EDW code is another variation of a DW code family that can has a variable weight greater than one. The EDW code possesses ideal cross-correlation properties and exists for every natural number n. A much better performance can be provided by using the EDW code compared to the existing code such as Hadamard and Modified Frequency-Hopping (MFH codes. It has been observed that theoretical analysis and simulation for EDW is much better performance compared to Hadamard and Modified Frequency-Hopping (MFH codes.
Progress on burnup calculation methods coupling Monte Carlo and depletion codes
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Leszczynski, Francisco [Comision Nacional de Energia Atomica, San Carlos de Bariloche, RN (Argentina). Centro Atomico Bariloche]. E-mail: lesinki@cab.cnea.gob.ar
2005-07-01
Several methods of burnup calculations coupling Monte Carlo and depletion codes that were investigated and applied for the author last years are described. here. Some benchmark results and future possibilities are analyzed also. The methods are: depletion calculations at cell level with WIMS or other cell codes, and use of the resulting concentrations of fission products, poisons and actinides on Monte Carlo calculation for fixed burnup distributions obtained from diffusion codes; same as the first but using a method o coupling Monte Carlo (MCNP) and a depletion code (ORIGEN) at a cell level for obtaining the concentrations of nuclides, to be used on full reactor calculation with Monte Carlo code; and full calculation of the system with Monte Carlo and depletion codes, on several steps. All these methods were used for different problems for research reactors and some comparisons with experimental results of regular lattices were performed. On this work, a resume of all these works is presented and discussion of advantages and problems found are included. Also, a brief description of the methods adopted and MCQ system for coupling MCNP and ORIGEN codes is included. (author)
TEMP: a computer code to calculate fuel pin temperatures during a transient
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The computer code TEMP calculates fuel pin temperatures during a transient. It was developed to accommodate temperature calculations in any system of axi-symmetric concentric cylinders. When used to calculate fuel pin temperatures, the code will handle a fuel pin as simple as a solid cylinder or as complex as a central void surrounded by fuel that is broken into three regions by two circumferential cracks. Any fuel situation between these two extremes can be analyzed along with additional cladding, heat sink, coolant or capsule regions surrounding the fuel. The one-region version of the code accurately calculates the solution to two problems having closed-form solutions. The code uses an implicit method, an explicit method and a Crank-Nicolson (implicit-explicit) method
Fuel rod under power oscillations; calculations with the ENIGMA code
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Power oscillations in a BWR may result from a series of events starting from a re-circulation pump trip or can be initiated during start-up at low-flow conditions by other perturbations. Whole core and regional oscillations have been observed. Severe consequences may be anticipated if the instability diverges and the reactor protection system fails (no scram) in all phases of the incident (ATWS). Power peaks higher than ten times of the pre-transient power level have been speculated to appear. Low-magnitude oscillations have been observed at the TVO plant, Olkiluoto 1987, and at the Lasalle-2 plant, 1988, and in other BWRs world-wide. Typically, a boiling water reactor has an unstable operational point at low flow and high power conditions. The physical phenomenon behind the instability is density wave oscillations leading to boiling boundary oscillations and void fraction fluctuations across the heated channel. These in turn, make the fission power vary. The typical frequency of the oscillations seems to be of the order of 0.5 Hz, and thus the power peak for a fuel rod is considerably wider than a RIA-pulse, for instance. Large oscillations can result in elevated fuel temperatures, accelerated fission gas release and additional internal loads on the cladding. These effects may be more severe for a high burnup rod with a large fission gas inventory and a closed gap. Therefore, an experiment has been proposed to be conducted at Halden reactor for simulating the fuel rod response under power oscillations. As there is lack of knowledge also on the relevant boundary conditions, pre-calculations with various input options have been performed and are further suggested. Calculations with FRAPTRAN code have shown the importance of the cladding-coolant heat transfer to the fuel temperature. The applicability of the ENIGMA code to this kind of transients was confirmed. To support the planning of the proposed Halden test, estimates on fuel and cladding temperatures as well as
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Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)
1988-04-01
A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.
International Nuclear Information System (INIS)
A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs
International Nuclear Information System (INIS)
The aim of this report is the comparison between two cell code systems which base on two different physical approaches. To this end wet and dry PROTEUS buffer cell reactivities are computed using the transport codes ONEDANT and SURCU together with self-shielded cross section ENDF/B-IV based libraries obtained using NMATXS module and postprocessor code TRANSX-CTR. TRANSX-CTR performs self-shielding of resonance cross sections using Bondarenko method. The same calculations are then repeated using new NJOY module MICROR and two region spectrum code MICROX-2, to perform more accurate pointwise resonance shielding solving slowing down equations. It is shown that the dry cell is not sensitive to the resonance cross sections, because the spectrum is too hard and energy of most neutrons is higher than resonance energies of fissionable isotopes. However the lattice is sensitive to the fission spectrum used. Also the computations pertaining to the water moderated cell show sensitivity to different weighting functions used to calculate fission spectra from fission matrices, because the spectrum is still too hard. But now large differencies in the reactivity of the cell arise between the shielding factor method and the computations based on the pointwise self-shielding. This is because the shielding factor method even if sophisticated (Bondarenko model) is not representative in lower resonance range where resonances are not narrow. (author)
Introduction to reactor lattice calculations by the WIMSD code
International Nuclear Information System (INIS)
The present report is based on lectures delivered at the Workshop on Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety hold in International Centre of Theoretical Physics, Trieste, in March 1998. The main goal of the set of lectures was to give the basis of reactor physics calculations for participants working on nuclear data.The last lectures, devoted to WIMS including the WIMSD code users. Following this general line the material is divided into three parts: The first part includes a short description of neutron transport phenomena limited to those definitions that are necessary to understand the approach to practical solution of the problem given in the second part on reactor lattice transport calculations. The detailed discussion of the neutron cross sections has been skipped as this subject has been treated in detail by other lectures. In the third part those versions of the well-known WIMSD code which are distributed by NEA Data Bank are described. The general structure of the code is given supplied in a more detailed description of aspects being the most common points of misunderstanding for the code users. (author)
Analytical stress tensor and pressure calculations with the CRYSTAL code
Doll, Klaus
2010-01-01
Abstract The calculation of the stress tensor and related properties and its implementation in the CRYSTAL code are described. The stress tensor is obtained from the earlier implemented analytical gradients with respect to the cell parameters. Subsequently, the pressure and enthalpy is computed, and a test concerning the pressure driven phase transition in KI is used as an illustration. Finally, the possibility of applying external pressure is implemented. The ...
SCRAM calculations with the KIKO3D code
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Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code. The spatial effects near to the position of ionisation chambers were studied. As was expected, reactivities calculated from the flux curves of different nodes using inverse point kinetics are in a wide range. The dynamic and static reactivities in case of asymmetric SCRAM differ considerably as a result of the slow flux shape redistribution. The effect of source neutrons from spontaneous fission and the node-wise delayed neutron fraction on the results is also presented. (Authors)
User's manual of the MKENO-DAR code system
International Nuclear Information System (INIS)
The computer code manual of MKENO-DAR which is a direct angular representation Monte Carlo code for criticality safety analysis is already issued as JAERI-M report, however, complex pre-stage data handlings and calculations by auxiliary programs are required before the execution of MKENO-DAR. The MKENO-DAR CODE SYSTEM widely spans a whole code system including MKENO-DAR and other pre-stage auxiliary programs. This report discusses the systematic treatment of the MKENO-DAR CODE SYSTEM and shows the simplified calculation technique from the user side of view. (author)
Wake field calculations with three-dimensional BCI code
International Nuclear Information System (INIS)
The new MAFIA code T3 is introduced, belonging to a family of fully 3-D codes for computer-aided design of RF cavities developed by the MAFIA collaboration at DESY, KFA Juelich and LANL. T3 is a 3-D extension of TBCI, solving Maxwell's equations in the time domain using the FIT ansatz, allowing the use of structures of arbitrary shape and dielectric material insertions, and integrating the wake potential at arbitrary positions. An open boundary condition was implemented to simulate infinite beam pipes. An IBM 3081 with 5 Mbytes of main storage can handle problems up to 80,000 mesh points while a window option enables the treatment of very long structures using up to 1,000,000 mesh points. Together with its postprocessors W3COR and W3OUT and the MAFIA mesh generator, this code was used to calculate wake potentials for several beam pipe components (i.e. vacuum chambers, vacuum chamber junctions etc.) which required fully 3-D calculations. Comparison of T3 results with TBCI calculations in the case of a cylindrically symmetric structure (pillbox) showed the agreement within a few percent
Energy Technology Data Exchange (ETDEWEB)
Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx
2007-07-01
The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)
SRAC95; general purpose neutronics code system
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Okumura, Keisuke; Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio
1996-03-01
SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author).
SRAC95; general purpose neutronics code system
International Nuclear Information System (INIS)
SRAC is a general purpose neutronics code system applicable to core analyses of various types of reactors. Since the publication of JAERI-1302 for the revised SRAC in 1986, a number of additions and modifications have been made for nuclear data libraries and programs. Thus, the new version SRAC95 has been completed. The system consists of six kinds of nuclear data libraries(ENDF/B-IV, -V, -VI, JENDL-2, -3.1, -3.2), five modular codes integrated into SRAC95; collision probability calculation module (PIJ) for 16 types of lattice geometries, Sn transport calculation modules(ANISN, TWOTRAN), diffusion calculation modules(TUD, CITATION) and two optional codes for fuel assembly and core burn-up calculations(newly developed ASMBURN, revised COREBN). In this version, many new functions and data are implemented to support nuclear design studies of advanced reactors, especially for burn-up calculations. SRAC95 is available not only on conventional IBM-compatible computers but also on scalar or vector computers with the UNIX operating system. This report is the SRAC95 users manual which contains general description, contents of revisions, input data requirements, detail information on usage, sample input data and list of available libraries. (author)
The GSCAN code: From GNASH reaction calculations to ENDF nuclear data files
International Nuclear Information System (INIS)
The GSCAN code, which is enhanced version of the GNXS code included in the GNASH code package presented at the IAEA Trieste Workshops, utilizes an output made by the Hauser-Feshbach and preequilibrium GNASH reaction model code. Its main purposes are: (1) To convert the calculated cross sections into ENDF-6 format; (2) To calculate the emission spectra of A ≥ 5 secondary particles (recoils) and represent them in ENDF-6 format; and (3) To display all the exclusive reaction channels that contributed to a given inclusive emission channel (production cross section). This code has been widely used at Los Alamos in the production of the high-energy data files that extend up to 150 MeV for incident neutrons and protons, for enhanced radiation transport simulations of accelerator-driven systems. (author)
Comparison of code calculations with experiments on containment response during LOCA conditions
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A series of experiments were performed on a one-tenth scale model of PHWR containment, incorporating pressure suppression system. The pressure-temperature transients in the model containment observed during simulated LOCA (Loss of Coolant) blowdown conditions were compared against calculated results form computer code PACSR, for purposes of verification of the code. Comparison of results indicated that calculated values of peak pressure in various compartment were significantly higher than observed ones. This disagreement was attributed mainly to modelling for energy absorption from containment atmosphere to structural surfaces, this effect being particularly important in a scaled down model. Good agreement between calculation and experiment was obtained after heat transfer correlation for energy absorption on surfaces were modified in the code. The study demonstrates the conservatism of the results from the code. (author). 6 refs., 1 tab., 9 figs
HERMET: cell neutronic calculation code for MTR (materials testing reactors) fuels
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The HERMET neutronic calculation code was developed for resolution of systems, at a cell calculation level in one-dimensional plain geometry (MTR), preserving its heterogeneous character with or without reflecting boundary conditions and reducing the cost as regards time and machine-memory. This code also includes the burn-up calculation which may be performed with the critical spectra B0, B1 or the one improved by leakages corresponding to the buckling given by the user. The burn-up scheme may be carried out by a transport equation with intermediate stages without flux reevaluation or by a predictor-corrector scheme. (Author)
Comparison of computer code calculations with FEBA test data
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The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.)
International Nuclear Information System (INIS)
A survey is given of the behavior of the reactor core during an accident associated with main steam header break. This problem is a benchmark exercise for the verification of thermohydraulic system codes connected with three-dimensional hexagonal reactor dynamics codes. The accident is initiated by a symmetrical break of the main steam header at the end of the 1st fuelling cyclein the hot shutdown condition with the control rod groups fully inserted except for group 4. The calculations were performed by using the externally coupled codes ATHLET Mod. 1.1 Cycle C and DYN3DH1.1/M3. The two codes are described at their external coupling is outlined. The standard WWER-440/213 input deck of the ATHLET code was used. The tresults of steady-state calculations as well as burnup calculations are given. The analysis of the transient process is aimed at the time behavior of safety related parameters in selected components of the primary and secondary circuits as well as at the global parameters of the reactor core and behavior in selected fuel assemblies
Refuelling design and core calculations at NPP Paks: codes and methods
International Nuclear Information System (INIS)
This article gives a brief review of the computer codes used in the fuel management practice at NPP Paks. The code package consist of the HELIOS neutron and gamma transport code for preparation of few-group cross section library, the CERBER code to determine the optimal core loading patterns and the C-PORCA code for detailed reactor physical analysis of different reactor states. The last two programs have been developed at the NPP Paks. HELIOS gives sturdy basis for our neutron physical calculation, CERBER and C-PORCA programs have been enhanced in great extent for last years. Methods and models have become more detailed and accurate as regards the calculated parameters and space resolution. Introduction of a more advanced data handling algorithm arbitrary move of fuel assemblies can be followed either in the reactor core or storage pool. The new interactive WINDOWS applications allow easier and more reliable use of codes. All these computer code developments made possible to handle and calculate new kind of fuels as profiled Russian and BNFL fuel with burnable poison or to support the reliable reuse of fuel assemblies stored in the storage pool. To extend thermo-hydraulic capability, with KFKI contribution the COBRA code will also be coupled to the system (Authors)
Calculation of reactor pressure vessel fluence using TORT code
International Nuclear Information System (INIS)
TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Unit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) for all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library, BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The maximum fast neutron fluence calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effective full power days is 1.784x1018n/cm2. The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree
System Design Description for the TMAD Code
International Nuclear Information System (INIS)
This document serves as the System Design Description (SDD) for the TMAD Code System, which includes the TMAD code and the LIBMAKR code. The SDD provides a detailed description of the theory behind the code, and the implementation of that theory. It is essential for anyone who is attempting to review or modify the code or who otherwise needs to understand the internal workings of the code. In addition, this document includes, in Appendix A, the System Requirements Specification for the TMAD System
Tokamak plasma power balance calculation code (TPC code) outline and operation manual
International Nuclear Information System (INIS)
This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)
Model Calculation of Fission Product Yields Data using GEF Code
International Nuclear Information System (INIS)
Fission yields data are classified with spontaneous fission data and neutron induced fission data. The fission product yields data at several energy points for the limited actinides are included in nuclear data libraries such as ENDF/B, JEFF and JENDL because production of those is based mainly on experimental results and it is very difficult to conduct experiments for all actinides and continuous energies. Therefore, in order to obtain fission yields data without experimental data, a theoretical fission model should be introduced to produce the yields data. GEneral Fission model (GEF) is developed to predict the properties for fissioning systems that have not been measured and that are not accessible to experiment. In this study, the fission yields data generated from GEF code are compared with the measured data and the recently available nuclear data libraries. The GEF code is very powerful tool to generate fission yields without measurements. Also, it can produce the distribution of fission product yields for continuous neutron energy while measured data are given only at several energies. The fission yields data of 235U have been tentatively generated with GEF code in this work. Comparing GEF results with measurements and recently released evaluated fission yields data, it is confirmed that GEF code can successfully predict the fission yields data. With its sophisticated model, GEF code is playing a significant role in nuclear industry
Code system for fast reactor neutronics analysis
International Nuclear Information System (INIS)
A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)
KARATE - a code for VVER-440 core calculation
Energy Technology Data Exchange (ETDEWEB)
Gado, J.; Hegedus, Cs.J.; Hegyi, Gy.; Kereszturi, A.; Makai, M.; Maraczi, Cs.; Telbisz, M.
1994-12-31
A modular calculation system has been elaborated at the KFKI Atomic Energy Research Institute for VVER-440 cores. The purpose of KARATE is the calculation of neutron physical and thermal-hydraulic processes in the core at normal, startup, and slow transient conditions. KARATE is under validation and verification (V&V) against mathematical, experimental, and operational data.
Adjoint Monte Carlo techniques and codes for organ dose calculations
International Nuclear Information System (INIS)
Adjoint Monte Carlo simulations can be effectively used for the estimation of doses in small targets when the sources are extended in large volumes or surfaces. The main features of two computer codes for calculating doses at free points or in organs of an anthropomorphic phantom are described. In the first program (REBEL-3) natural gamma-emitting sources are contained in the walls of a dwelling room; in the second one (POKER-CAMP) the user can specify arbitrary gamma sources with different spatial distributions in the environment: in (or on the surface of) the ground and in the air. 3 figures
Calculations of WWER cells and assemblies by WIMS-7B code
International Nuclear Information System (INIS)
A study of the nuclear data libraries of the WIMS-7B code have been performed in calculations of computational benchmark problems. The benchmarks cover pin cell, single fuel assembly with several different fuel types, moderator densities. Fuel depletion is performed to a burnup of 60 MWd/kgNM in the WWER-1000 pin cell. The results of the analysis of the benchmark with different code systems have been compared and indicated good agreement among the different methods and data. (Authors)
Computer code for shielding calculations of x-rays rooms
International Nuclear Information System (INIS)
The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)
The PHREEQE Geochemical equilibrium code data base and calculations
International Nuclear Information System (INIS)
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)
Directory of Open Access Journals (Sweden)
Asghar Mesbahi
2015-09-01
Full Text Available Introduction Radiotherapy with small fields is used widely in newly developed techniques. Additionally, dose calculation accuracy of treatment planning systems in small fields plays a crucial role in treatment outcome. In the present study, dose calculation accuracy of two commercial treatment planning systems was evaluated against Monte Carlo method. Materials and Methods Siemens Once or linear accelerator was simulated, using MCNPX Monte Carlo code, according to manufacturer’s instructions. Three analytical algorithms for dose calculation including full scatter convolution (FSC in TiGRT, along with convolution and superposition in XiO system were evaluated for a small solid liver tumor. This solid tumor with a diameter of 1.8 cm was evaluated in a thorax phantom, and calculations were performed for different field sizes (1×1, 2×2, 3×3 and4×4 cm2. The results obtained in these treatment planning systems were compared with calculations by MC method (regarded as the most reliable method. Results For FSC and convolution algorithm, comparison with MC calculations indicated dose overestimations of up to 120%and 25% inside the lung and tumor, respectively in 1×1 cm2field size, using an 18 MV photon beam. Regarding superposition, a close agreement was seen with MC simulation in all studied field sizes. Conclusion The obtained results showed that FSC and convolution algorithm significantly overestimated doses of the lung and solid tumor; therefore, significant errors could arise in treatment plans of lung region, thus affecting the treatment outcomes. Therefore, use of MC-based methods and super position is recommended for lung treatments, using small fields and beamlets.
Improvements of the subgroup resonance calculation code SUGAR
International Nuclear Information System (INIS)
Highlights: • Improvement of subgroup resonance code SUGAR is undertaken to increase efficiency. • Resonant nuclides are grouped for complex isotope compositions. • An in-house matrix MOC solver is employed to replace the AutoMOC solver. • These techniques speed 5–32 without any loss of accuracy or geometric flexibility. - Abstract: Due to its accuracy and geometric flexibility, the subgroup method is becoming a more and more attractive resonance calculation approach dedicated to obtaining resonance group macroscopic cross sections from multi-group libraries. In order to increase the efficiency of our subgroup code SUGAR, this paper contributes to the development of the code from four aspects. Firstly, subgroup parameters were proved to be problem-independent and the number of subgroups can be chosen automatically. This motivated us to produce a new multi-group library. Secondly, what subgroup method really needs is the relative subgroup flux within each multi-group instead of the relative multi-group flux between different groups. Thus, it is unnecessary to iteratively calculate in the whole energy range if the connections between different energy ranges can be approximated by a simple method. Thirdly, for problems with complex isotope compositions, resonant nuclides could be grouped according to their resonance characteristics. By this grouping, computational effort could be significantly reduced since nominal resonant nuclides turn out to be these nuclide groups rather than the actual nuclides. Finally, considering that most of the computational effort is spent on solving the subgroup neutron transport equation, an in-house matrix MOC solver is employed to replace the AutoMOC solver. In this way, the higher speed of the matrix MOC solver can be fully utilized by our subgroup code. To verify these theories and to prove the improvements, a series of benchmark problems were solved. It is demonstrated by these numerical results that these techniques can
Development of the Joyo MK-II core bowing reactivity calculation code
International Nuclear Information System (INIS)
The study on the passive safety test by using the Experimental Fast Reactor Joyo has been performed to demonstrate the inherent safety of fast breeder reactors. In this study, emphasis was placed on the improvement on the accuracy of the feedback reactivity analysis. As a bowing reactivity might play a significant roll in ATWS analysis because of its effectively short time constant and relatively large magnitude, an emphasis was placed upon the evaluation of the analysis precision of bowing reactivity. Taking into account of the refueling and irradiation history of the individual core component, the core bowing behavior in Joyo has been analyzed by using the MK-II core management code system MAGI, the interface code TETRAS which interpolate neutron flux and coolant temperature at the position of wrapper tube, and the core bowing calculation code BEACON. Calculation accuracy of above mentioned system was evaluated through the comparison of calculated and measured permanent distortion of subassemblies. In 1996, core bowing reactivity was calculated by AURORA code using the above calculated bowing behavior of individual core component as input. But because an approximate two dimensional material reactivity worth map was utilized in AURORA, it was made clear that some amount of error caused by extrapolation could not be neglected. Therefore calculation code ARCHCOM (Analysis of Reactivity Change due to Core Mechanics) which utilize three dimensional material reactivity worth map as input was developed for the Joyo MK-II core bowing reactivity calculation. This code reduces above mentioned extrapolation error that used to be occurred at isolated core component, such as control rod or irradiation rig and at the interface region between fuel and reflector which had sharp bowing reactivity worth gradient. (author)
International Nuclear Information System (INIS)
The paper aims to present the main physical principles for selection of design characteristics of the fast reactor control rods (CR) system. The brief analysis of problems of CR physical calculations is given. Four components are described for the correction to the control rod worth calculated by the routine method based on the few - group three - dimensional diffusion code (TRIGEX) in hexagonal geometry. Principle considerations are given for the choice of the original task discretization methods implemented in this code to minimize the total error. Brief information is given about methods and codes used for the evaluation of error components of control rod worths calculated in a standard way. The results of experimental and calculational investigations of control rod physical characteristics are presented. These results were obtained at BFS critical assemblies simulating LMFBR cores. The investigations have been carried out for different types of core configurations. The experimental and calculated values are given on the distortion of power distribution due to the control rod insertion in the core. (author). 51 refs, 9 figs, 5 tabs
Calculation code evaluating the confinement of a nuclear facility in case of fires
Energy Technology Data Exchange (ETDEWEB)
Laborde, J.C.; Prevost, C.; Vendel, J. [and others
1995-02-01
Accident events involving fire are quite frequent and could have a severe effect on the safety of nuclear facilities. As confinement must be maintained, the ventilation and filtration systems have to be designed to limit radioactive release to the environment. To determine and analyse the consequences of a fire on the contamination confinement, IPSN, COGEMA and SGN are participating in development of a calculation code based on introduction, in the SIMEVENT ventilation code, of various models associated to fire risk and mass transfer in the ventilation networks. This calculation code results from the coupling of the SIMEVENT code with several models describing the temperature in a room resulting of a fire, the temperatures along the ventilation ducts, the contamination transfers through out the ventilation equipments (ducts, dampers, valves, air cleaning systems) and the High Efficiency Particulate Air (HEPA) filters clogging. The paper proposed presents the current level of progress in development of this calculation code. It describes, in particular, the empirical model used for the clogging of HEPA filters by the aerosols derived from the combustion of standard materials used in the nuclear industry. It describes, also, the specific models used to take into account the mass transfers and resulting from the basic mechanisms of aerosols physics. In addition, an assessment of this code is given using the example of a simple laboratory installation.
Code accuracy evaluation of ISP 35 calculations based on NUPEC M-7-1 test
International Nuclear Information System (INIS)
Quantitative evaluation of code uncertainties is a necessary step in the code assessment process, above all if best-estimate codes are utilised for licensing purposes. Aiming at quantifying the code accuracy, an integral methodology based on the Fast Fourier Transform (FFT) has been developed at the University of Pisa (DCMN) and has been already applied to several calculations related to primary system test analyses. This paper deals with the first application of the FFT based methodology to containment code calculations based on a hydrogen mixing and distribution test performed in the NUPEC (Nuclear Power Engineering Corporation) facility. It is referred to pre-test and post-test calculations submitted for the International Standard Problem (ISP) n. 35. This is a blind exercise, simulating the effects of steam injection and spray behaviour on gas distribution and mixing. The result of the application of this methodology to nineteen selected variables calculated by ten participants are here summarized, and the comparison (where possible) of the accuracy evaluated for the pre-test and for the post-test calculations of a same user is also presented. (author)
Burnup calculations of TR-2 Research Reactor with Monteburns Monte Carlo Code
International Nuclear Information System (INIS)
Full text: In this study, some neutronic calculations of first and second core cycles of 5 MW pool type TR-2 Research Reactor have been performed using Multi-Step Monte Carlo Burnup Code System MONTEBURNS and the results were compared with the values of experiments and other codes. Time dependent keff distribution and burnup ratios belong to first and second core cycles of TR-2 Research Reactor were compared and quite good consistence in the results were observed. After modeling the first and second core cycles of TR-2 with MCNP5 Monte Carlo code, MCNP5 used in MONTEBURNS code has been parallelized in 8 HP ProLiant BL680C G5 systems with 4 quad-core Intel Xeon E7330 CPU, utilizing the MPI parallel protocol and simulations were performed on the 128 cores Linux parallel computing machine system. The computation time was reduced by parallelization of MONTEBURNS which uses MCNP in many steps. (authors)
Module type plant system dynamics analysis code (MSG-COPD). Code manual
International Nuclear Information System (INIS)
MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)
Development of calculation code of fission products specific activity in primary coolant
International Nuclear Information System (INIS)
Based on an assumption of that there is a design basis fuel defect level from reactor startup, calculation method of fission products specific activities in primary coolant is studied. Time-dependent nuclide activities in defect fuel are calculated by ORIGEN code, and nuclide releases from the defect fuel are considered. After processed by interface codes, data are used by PCFPA code which is used to calculate nuclide activities in the coolant. PCFPA solves differential equations by unit of decay chain, and totally considers decay's contribution to nuclide activities, and considers different system design between secondary and third generation plants such as AP1000. The method could provide the maximum of specific activity during plant operation and their results are consistent with data in AP1000 DCD(Rev.16). The method could be applicable to shielding design in secondary and third generation plants such as AP1000. (authors)
TRANS-I: A fast calculating computer code for the calculation of reactivity transients
International Nuclear Information System (INIS)
In literature is shown that the adiabatic and the quasistatic approximation to space time neutron kinetics are generally fast and conservative methods for calculating reactivity transients. Nevertheless if a feedback reactivity is considered these methods predict too high values of peak flux, energy production and temperature. It is demonstrated, that the deficiency of adiabatic and quasistatic method can be removed, if the mean fuel temperature is multiplied by a weighting factor to get a corrected temperature for calculating Doppler-feedback. The code TRANS-I including this modification is presented. (author)
Process of cross section generation for radiation shielding calculations, using the NJOY code
International Nuclear Information System (INIS)
The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author)
International Nuclear Information System (INIS)
A benchmark problem was proposed to reproduce an experiment for target membrane structure cooling of Accelerator Driven System at the 10th meeting of IWGAR (International Working Group of Advanced Nuclear Reactors Thermal Hydraulic) by the Fluid Phenomena in Energy Exchanges Section of IAHR (International Association of Hydraulic Engineering and Research). The benchmark calculation has been carried out with AQUA and FLUENT codes to estimate the code validity for liquid metal thermal-hydraulics application. As a result of comparison between numerical analyses and experiment, it is concluded as follows: Inlet flow rate at the distributing grid much affects a coolant temperature and temperature pulsation near the membrane. The coolant temperature decreases and the pulsation decays rapidly as the flow rate toward the membrane center increases. On downstream of the distributing grid, numerical results agree with experimental data except that numerical analysis tends to overestimate the coolant temperature pulsation. Numerical results show that the decrease of coolant temperature and the dissipation of pulsation tend to be underestimated when the flow rate toward the membrane center increases. In FLUENT code, the dissipation of coolant temperature is underestimated more than in AQUA code because FLUENT code tends to overestimate the flow rate toward the membrane center. But the same tendency of the dissipation behavior is shown in AQUA code. A turbulent model is less influenced on the coolant behavior in this benchmark analysis. Because Prandtl (Pr) number of liquid metal is low and the turbulent flow is not developed sufficiently in the conditions of the experiment. (author)
Performance of independent dose calculation in helical tomotherapy: implementation of the MCSIM code
International Nuclear Information System (INIS)
Currently, a software-based second check dose calculation for helical tomotherapy (HT) is not available. The goal of this study is to evaluate the dose calculation accuracy of the in-house software using EGS4 /MCSIM Monte Carlo environment against the treatment planning system calculations. In-house software was used to convert HT treatment plan information into a non-helical format. The MCSIM dose calculation code was evaluated by comparing point dose calculations and dose profiles against those from the HT treatment plan. Fifteen patients, representing five treatment sites, were used in this comparison. Point dose calculations between the HT treatment planning system and the EGS4 /MCSIM Monte Carlo environment had percent difference values below 5 % for the majority of this study. Vertical and horizontal planar profiles also had percent difference values below 5 % for the majority of this study. Down sampling was seen to improve speed without much loss of accuracy. EGS4 /MCSIM Monte Carlo environment showed good agreement with point dose measurements, compared to the HT treatment plans. Vertical and horizontal profiles also showed good agreement. Significant time saving may be obtained by down-sampling beam projections. The dose calculation accuracy of the in-house software using the MCSIM code against the treatment planning system calculations was evaluated. By comparing point doses and dose profiles, the EGS4 /MCSIM Monte Carlo environment was seen to provide an accurate independent dose calculation.
Benchmark calculation of CANDU end shielding system
Energy Technology Data Exchange (ETDEWEB)
Roh, Gyuhong; Choi, Hangbok [KAERI, Taejon (Korea, Republic of)
1998-05-01
A shielding analysis was performed for the end shield of CANDU 6 reactor. The one-dimensional discrete ordinate code ANISN with a 38-group neutron-gamma library, extracted from DLC-37D library, was used to estimate the dose rate for the natural uranium CANDU reactor. For comparison, MCNP-4B calculation was performed for the same system using continuous, discrete and multi-group libraries. The comparison has shown that the total dose rate of the ANISN calculation agrees well with that of the MCNP calculation. However, the individual dose rate (neutron and gamma) has shown opposite trends between ANISN and MCNP estimates, which may require a consistent library generation for both codes.
Kumawat, Soma; Ravi Kumar, M.
2016-07-01
Double Weight (DW) code family is one of the coding schemes proposed for Spectral Amplitude Coding-Optical Code Division Multiple Access (SAC-OCDMA) systems. Modified Double Weight (MDW) code for even weights and Enhanced Double Weight (EDW) code for odd weights are two algorithms extending the use of DW code for SAC-OCDMA systems. The above mentioned codes use mapping technique to provide codes for higher number of users. A new generalized algorithm to construct EDW and MDW like codes without mapping for any weight greater than 2 is proposed. A single code construction algorithm gives same length increment, Bit Error Rate (BER) calculation and other properties for all weights greater than 2. Algorithm first constructs a generalized basic matrix which is repeated in a different way to produce the codes for all users (different from mapping). The generalized code is analysed for BER using balanced detection and direct detection techniques.
International Nuclear Information System (INIS)
This paper provides for various cases of 235U enrichment or Pu isotopic vectors, and different reflectors, new minimum critical values of uranyl nitrate and plutonium nitrate solutions (H+=0) obtained by the standard IRSN calculation route and the new isopiestic density laws. Comparisons are also made with other more accurate routes showing that the standard one's results are most often conservative and usable for criticality safety assessments. (author)
Fragmentation calculation by intranuclear-cascade-evaporation code
Energy Technology Data Exchange (ETDEWEB)
Shigyo, Nobuhiro; Iga, Kiminori; Ishibashi, Kenji [Kyushu Univ., Fukuoka (Japan)
1997-03-01
High Energy Transport Code (HETC) based on the intranuclear-cascade-evaporation model is modified for calculating the fragmentation cross section. For the intranuclear-cascade process, nucleon-nucleon cross sections are used for collision computation; effective in-medium-corrected cross sections are adopted instead of the original free-nucleon collision. The exciton model is adopted for improvement of backward nucleon-emission cross section for low-energy nucleon-incident events. The fragmentation reaction is incorporated into the original HETC as a subroutine set by the use of the systematics of the reaction. The modified HETC (HETC-3STEP/FRG) reproduces experimental fragment yields to a reasonable degree. (author)
Computer access security code system
Collins, Earl R., Jr. (Inventor)
1990-01-01
A security code system for controlling access to computer and computer-controlled entry situations comprises a plurality of subsets of alpha-numeric characters disposed in random order in matrices of at least two dimensions forming theoretical rectangles, cubes, etc., such that when access is desired, at least one pair of previously unused character subsets not found in the same row or column of the matrix is chosen at random and transmitted by the computer. The proper response to gain access is transmittal of subsets which complete the rectangle, and/or a parallelepiped whose opposite corners were defined by first groups of code. Once used, subsets are not used again to absolutely defeat unauthorized access by eavesdropping, and the like.
Development of neutral transport lattice code DENT-2D and benchmark calculation
International Nuclear Information System (INIS)
We developed new transport lattice code called DENT-2D (Deterministic Neutral Particle Transport Code in 2-D imensional Space)primarily to generate few- group constants for the reactor physics analysis diffusion codes. This code is designed to be coupled with KAERI reactor analysis nodal code, MASTER [1] ,to complete the design system package. CASMO-3 and HELIOS have been used in generating the few- group constant for MASTER. Currently DENT-2D includes only neutron particle transport calculation in 2-dimensional Cartesian geometry. The characteristics method is adopted for the spatial discretization, which is advantageous for the treatment of the complicated geometry structure and the highly anisotropic scattering. The subgroup method is used for the resonance treatment. B1 approximation has been used to obtain the criticality spectrum considering the leakage effect in the real core situation. The exponential matrix method has been used for the depletion calculation. The results of benchmark calculations show that the prediction capability of DENT-2D is comparable to the other lattice codes such as HELIOS and CASMO-3
Development of PC version code system for radiation dose estimation
International Nuclear Information System (INIS)
Since a direct access from a radiation work site to a main frame computer is usually difficult, evaluation of radiation sources and/or doses are often carried out by handcalculations with less accuracy. So considering a recent remarkable progress of PC (Personal computer), we have developed an interactive code system of PC version to calculate dose equivalent values with high accuracy. It consists of a radiation source calculation code ORIGEN-2 and point kernel shielding calculation codes, QAD-CGGP2 and G33-GP2. With the present system, you can easily obtain dose equivalent values at any detector point starting from radiation source estimation. Validity of these codes have been verified individually on a main frame computer through various benchmark calculations. Thus we verified the present PC version system by comparing the PC calculations with those using a main frame computer. Excellent agreement was obtained between them. (author)
International Nuclear Information System (INIS)
A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)
Recent developments in the Los Alamos radiation transport code system
Energy Technology Data Exchange (ETDEWEB)
Forster, R.A.; Parsons, K. [Los Alamos National Lab., NM (United States)
1997-06-01
A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results.
Codes complex for quick transport 3D neutron calculations of WWER
International Nuclear Information System (INIS)
Complex Surface Values System that capable to carry out the all stages of neutron physical stationary calculations for reactors WWER and PWR is presented. This complex based oneself on using the Surface Values Methods. The approach consists in maximum possible account of specific features in reactor calculations. There is a certain amount of specific features in reactor problems due attention to which is crucial for attaining the main goal-organizing of reactor code.s complexes those provide economical and safe reactors' operating. It is important for the estimation of a code quality to fix of the methodical component but it is necessary to have in view of scale of other ones of result uncertainties. We mention in passing it afterwards. Complex Surface Values System present the approach of building computational reactor models that account for the specifics of reactor problems outlined above. This approach is called the methods of surface values. It utilizes method of surface pseudo-sources for calculating cells within active cores and method of surface harmonics for calculating the whole core or certain sub-assemblies that contain several cells. The part of total complex - the complex Surface Values Lattices - is destined for cells, assembly and sub-assembly calculations of thermal water reactors. The complex Surface Values Lattices use micro constant libraries prepared in the format WIMS. There are libraries in such format based on different initial files of valued data (ENDBF, JEFF, UKNDL, JENDL) on site IAEA. We used these libraries for comparing and choused the one similar to recommended IAEA library. The code Surface Values Core is contained in the total complex Surface Values System besides the complex Surface Values Lattice. This code provide the total scale calculations of reactor's cores. The equations of the Surface Harmonics Method are suitable well for inside reactor core. Other approximations are necessary for neutron description inside reflector
Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor
International Nuclear Information System (INIS)
The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' with the 140 MWt irradiation core using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation transient of a large-scale fast breeder reactor. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present test was selected and simulated using the NETFLOW++ code. Major plant parameters are simulated with good agreement such a similar accuracy as the Mimir-N2 exclusive code for 'Joyo'. As a result, it is concluded that the NETFLOW++ is applicable to the natural circulation analysis of sodium-cooled fast reactors with the similar scale of the prototype reactor 'Monju'. (author)
Relative efficiency calculation of a HPGe detector using MCNPX code
International Nuclear Information System (INIS)
High-purity germanium detectors (HPGe) are mandatory tools for spectrometry because of their excellent energy resolution. The efficiency of such detectors, quoted in the list of specifications by the manufacturer, frequently refers to the relative full-energy peak efficiency, related to the absolute full-energy peak efficiency of a 7.6 cm x 7.6 cm (diameter x height) NaI(Tl) crystal, based on the 1.33 MeV peak of a 60Co source positioned 25 cm from the detector. In this study, we used MCNPX code to simulate a HPGe detector (Canberra GC3020), from Real-Time Neutrongraphy Laboratory of UFRJ, to survey the spectrum of a 60Co source located 25 cm from the detector in order to calculate and confirm the efficiency declared by the manufacturer. Agreement between experimental and simulated data was achieved. The model under development will be used for calculating and comparison purposes with the detector calibration curve from software Genie2000™, also serving as a reference for future studies. (author)
Computer codes for the calculation of vibrations in machines and structures
International Nuclear Information System (INIS)
After an introductory paper on the typical requirements to be met by vibration calculations, the first two sections of the conference papers present universal as well as specific finite-element codes tailored to solve individual problems. The calculation of dynamic processes increasingly now in addition to the finite elements applies the method of multi-component systems which takes into account rigid bodies or partial structures and linking and joining elements. This method, too, is explained referring to universal computer codes and to special versions. In mechanical engineering, rotary vibrations are a major problem, and under this topic, conference papers exclusively deal with codes that also take into account special effects such as electromechanical coupling, non-linearities in clutches, etc. (orig./HP)
HELIAS module development for systems codes
International Nuclear Information System (INIS)
In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs
HELIAS module development for systems codes
Energy Technology Data Exchange (ETDEWEB)
Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.
2015-02-15
In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.
A FACSIMILE code for calculating void swelling and creep, with vacancy loops present: version VS4
International Nuclear Information System (INIS)
This FACSIMILE code calculates void swelling and creep of irradiated materials, taking into account the effects of cavities, interstitial loops, vacancy loops, dislocation network and either grain boundaries or foil surfaces. The creep calculations are based on SIPA theory (stress induced preferred absorption), with no preferred nucleation. Either interactive or non-interactive options are available for the sink strength equations, but rate limitation is not incorporated. FACSIMILE is a computer program for solving simultaneous differential equations, and this VS4 code is one of a series of codes for calculating void swelling using increasingly complex theories. Other reports describing the VS1 and VS2 codes explain their use under control of the TSO system of the Harwell IBM 3033 computer, and explain the basic organization of the codes as required for use by FACSIMILE. The creep theory assumes that the material is under a constant uniaxial tensile stress during the irradiation. Three directions are considered for network parameters relative to the direction of the stress, and two directions for interstitial and vacancy loops. To give a full picture of these various contributions to the total creep, a large set of output parameter values are printed for each demanded dose value via a FORTRAN subroutine. (author)
International Nuclear Information System (INIS)
Two methods of calculating criticality are available in the 3D generalised geometry Monte Carlo particle transport code SPARTAN (Bending and Heffer, 1975). The first is a matrix technique in which the multiplication constant and source distribution of the system under study are calculated from estimates of fission probabilities and the second a method in which the multiplication constant is inferred from estimates of changes in neutron population over a number of neutron generations. Modifications are described which have been made to the way in which these methods are used in SPARTAN in order to improve the efficiency of criticality calculations. (author)
LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code
International Nuclear Information System (INIS)
Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs
Quasiparticle GW calculations within the GPAW electronic structure code
DEFF Research Database (Denmark)
Hüser, Falco
The GPAW electronic structure code, developed at the physics department at the Technical University of Denmark, is used today by researchers all over the world to model the structural, electronic, optical and chemical properties of materials. They address fundamental questions in material science...... properties are to a large extent governed by the physics on the atomic scale, that means pure quantum mechanics. For many decades, Density Functional Theory has been the computational method of choice, since it provides a fairly easy and yet accurate way of determining electronic structures and related...... respect to the system one wants to investigate by choosing a certain functional or by tuning parameters. A succesful alternative is the so-called GW approximation. It is mathematically precise and gives a physically well-founded description of the complicated electron interactions in terms of screening...
International Nuclear Information System (INIS)
The Electron-Gamma Shower (EGS) code system is a general purpose package for the Monte Carlo simulation of the coupled transport of electrons and photons in an arbitrary geometry for particles with energies above a few keV up to several hundred GeV (depending on the atomic numbers of the target materials). This report introduces a new, enhanced version called EGS5. In addition to explaining and documenting the various enhancements and changes to the previous version (EGS4), this document includes several introductory and advanced tutorials on the use of EGS5, and also contains the EGS5 User Manual. Our intention has been to make this document wholly self-contained so that the user need not refer to the original EGS4 manual (SLAC-265) in order to use the code. To this end, we have taken the liberty of incorporating into Chapter 2 of this report those portions of Chapter 2 of SLAC-265 which describe physics models of EGS4 retained by EGS5, thereby documenting all the physics contained in EGS5. (author)
DeCART code verifications by numerical benchmark calculations of HTTR
International Nuclear Information System (INIS)
DeCART code verifications have been performed through the numerical benchmark calculations of HTTR. The reference calculations have been carried out using the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART code gives less negative MTC and RTC than the McCARD code does and thus the DeCART code underestimates the multiplication factors at states with high moderator and reflector temperatures. However, the DeCART code predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of the DeCART code increases with burnup. While the DeCART code error in a 2-dimensional core depletion calculation decreases with burnup up to around 500 FPD. (author)
Expansion of the CHR bone code system
International Nuclear Information System (INIS)
This report describes the coding system used in the Center for Human Radiobiology (CHR) to identify individual bones and portions of bones of a complete skeletal system. It includes illustrations of various bones and bone segments with their respective code numbers. Codes are also presented for bone groups and for nonbone materials
GOBLIN computer code. Comparison between calculations and TLTA small break test
International Nuclear Information System (INIS)
GOBLIN calcuations have been performed for two simulation tests of the boiling water reactor (BWR) small break loss-of-coolant accidents (LOCAs) which were conducted in the two loop test apparatus (TLTA). The first test investigated the small break with nondegraded emergency core coolant (ECC) systems and the second test studied the same small break but with degraded ECC systems in which the high pressure core spray (HPCS) was assumed unavailable. Very good agreement between test data and calculations is achieved. The second test is the most challenging from code comparison point of view and the code prediction of the complicated mass distribution pattern which changes with time is very satisfactory. In the first test and to some extent late in the second test multidimensional subchannel effects are evident in the core bundle region. These are not and cannot be reproduced by the code since the bundle model of GOBLIN is strictly one-dimensional. (Author)
International Nuclear Information System (INIS)
The method to calculate the response function of spherical BF3 proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF3 proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within ±10%. (author)
A guide to the AUS modular neutronics code system
International Nuclear Information System (INIS)
A general description is given of the AUS modular neutronics code system, which may be used for calculations of a very wide range of fission reactors, fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations, one-dimensional transport calculations, and one, two, and three-dimensional diffusion calculations, burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system
Recent progress with large-scale ab initio calculations: the CONQUEST code
Bowler, D. R.; Choudhury, R.; Gillan, M. J.; Miyazaki, T.
While the success of density functional theory (DFT) has led to its use in a wide variety of fields such as physics, chemistry, materials science and biochemistry, it has long been recognised that conventional methods are very inefficient for large complex systems, because the memory requirements scale as N 2 and the cpu requirements as N 3 (where N is the number of atoms). The principles necessary to develop methods with linear scaling of the cpu and memory requirements with system size (O(N ) methods) have been established for more than ten years, but only recently have practical codes showing this scaling for DFT started to appear. We report recent progress in the development of the Conquest code, which performs O(N ) DFT calculations on parallel computers, and has a demonstrated ability to handle systems of over 10000 atoms. The code can be run at different levels of precision, ranging from empirical tight-binding, through ab initio tight-binding, to full ab initio , and techniques for calculating ionic forces in a consistent way at all levels of precision will be presented. Illustrations are given of practical Conquest calculations in the strained Ge/Si(001) system.
Energy Technology Data Exchange (ETDEWEB)
Antunes, Alberi
2008-07-01
This work presents the Physics of Source Driven Systems (ADS). It shows some statics and K{sub i}netics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)
Tandem Mirror Reactor Systems Code (Version I)
International Nuclear Information System (INIS)
A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost
CALOR89: The code system for calorimeter analysis and design
International Nuclear Information System (INIS)
As part of a strong experimental high energy physics program, a substantial effort must be involved in calculational analysis of the detector system. This calculational capability must be fundamentally sound and based on previous interchange between theoretical calculations and experimental test programs. The CALOR89[1-6] system for analyzing calorimeters offers a solid approach for investigating all facets of detector systems and has been used in many calculational studies. CALOR89 is one of two major code systems recommended for analysis of SSC detector systems
14 CFR 234.8 - Calculation of on-time performance codes.
2010-01-01
... (AVIATION PROCEEDINGS) ECONOMIC REGULATIONS AIRLINE SERVICE QUALITY PERFORMANCE REPORTS § 234.8 Calculation of on-time performance codes. (a) Each reporting carrier shall calculate an on-time performance code... 14 Aeronautics and Space 4 2010-01-01 2010-01-01 false Calculation of on-time performance...
International Nuclear Information System (INIS)
The post test calculations of three runs of the experiment UPTF-TRAM A2 using the thermohydraulic system code ATHLET are presented. This separate-effects-test was designed to investigate the formation of flow regimes in a hot leg of a PWR under different two-phase flow conditions. The formation of a stratified flow has influence on the distribution of the coolant mass in the primary system. Using an input model with the same nodalization for the represented components as employed for the analyses of integral transients in PWR-facilities the following results and knowledges were gained. The process of mixture level increase in the reactor vessel just after the start of the steam injection into the core simulator could only be simulated with differences to the experiment. The displacement of the water inventory in the upper plenum and the initial water carry over into the hot legs were strongly overpredicted. Under all specified mass flow conditions the formation of a stratified flow in the horizontal part of the hot leg has been calculated in correspondence to the experiment. Except for the observed change of the flow regime the spatial water distribution in the hot leg depending on the boundary conditions could be simulated qualitatively. The ATHLET code always predicts a flow regime with strong two-phase momentum transfer, which could not be observed in all test phases. (orig.)
Calculations of neutron penetration through graphite medium with Monte Carlo code MCNP
International Nuclear Information System (INIS)
Experiments for fast neutron penetration through graphite are analysed with the continuous energy Monte Carlo code MCNP. Reaction rates and energy spectra obtained with the MCNP are compared with measured values and calculated ones with McBEND code. And validity of penetration calculation with the MCNP is comfirmed. In addition, it is revealed that the MCNP code using Weight-Window method is well applicable to calculations of neutron penetration through graphite up to 70 cm in depth. (author)
Stepšys, A.; Mickevicius, S.; Germanas, D.; Kalinauskas, R. K.
2014-11-01
This new version of the HOTB program for calculation of the three and four particle harmonic oscillator transformation brackets provides some enhancements and corrections to the earlier version (Germanas et al., 2010) [1]. In particular, new version allows calculations of harmonic oscillator transformation brackets be performed in parallel using MPI parallel communication standard. Moreover, higher precision of intermediate calculations using GNU Quadruple Precision and arbitrary precision library FMLib [2] is done. A package of Fortran code is presented. Calculation time of large matrices can be significantly reduced using effective parallel code. Use of Higher Precision methods in intermediate calculations increases the stability of algorithms and extends the validity of used algorithms for larger input values. Catalogue identifier: AEFQ_v4_0 Program summary URL: http://cpc.cs.qub.ac.uk/summaries/AEFQ_v4_0.html Program obtainable from: CPC Program Library, Queen’s University of Belfast, N. Ireland Licensing provisions: GNU General Public License, version 3 Number of lines in programs, including test data, etc.: 1711 Number of bytes in distributed programs, including test data, etc.: 11667 Distribution format: tar.gz Program language used: FORTRAN 90 with MPI extensions for parallelism Computer: Any computer with FORTRAN 90 compiler Operating system: Windows, Linux, FreeBSD, True64 Unix Has the code been vectorized of parallelized?: Yes, parallelism using MPI extensions. Number of CPUs used: up to 999 RAM(per CPU core): Depending on allocated binomial and trinomial matrices and use of precision; at least 500 MB Catalogue identifier of previous version: AEFQ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 181, Issue 2, (2010) 420-425 Does the new version supersede the previous version? Yes Nature of problem: Calculation of matrices of three-particle harmonic oscillator brackets (3HOB) and four-particle harmonic oscillator brackets (4HOB) in a more
Code system to compute radiation dose in human phantoms
International Nuclear Information System (INIS)
Monte Carlo photon transport code and a code using Monte Carlo integration of a point kernel have been revised to incorporate human phantom models for an adult female, juveniles of various ages, and a pregnant female at the end of the first trimester of pregnancy, in addition to the adult male used earlier. An analysis code has been developed for deriving recommended values of specific absorbed fractions of photon energy. The computer code system and calculational method are described, emphasizing recent improvements in methods
Computer codes used in the calculation of high-temperature thermodynamic properties of sodium
International Nuclear Information System (INIS)
Three computer codes - SODIPROP, NAVAPOR, and NASUPER - were written in order to calculate a self-consistent set of thermodynamic properties for saturated, subcooled, and superheated sodium. These calculations incorporate new critical parameters (temperature, pressure, and density) and recently derived single equations for enthalpy and vapor pressure. The following thermodynamic properties have been calculated in these codes: enthalpy, heat capacity, entropy, vapor pressure, heat of vaporization, density, volumetric thermal expansion coefficient, compressibility, and thermal pressure coefficient. In the code SODIPROP, these properties are calculated for saturated and subcooled liquid sodium. Thermodynamic properties of saturated sodium vapor are calculated in the code NAVAPOR. The code NASUPER calculates thermodynamic properties for super-heated sodium vapor only for low (< 1644 K) temperatures. No calculations were made for the supercritical region
SINFAC - SYSTEMS IMPROVED NUMERICAL FLUIDS ANALYSIS CODE
Costello, F. A.
1994-01-01
. On the first pass, the user finds that the calculated outlet conditions of the last component do not match the estimated inlet conditions of the first. The user then modifies the estimated inlet conditions of the first component in an attempt to match the calculated values. The user estimated values are called State Variables. The differences between the user estimated values and calculated values are called the Error Variables. The procedure systematically changes the State Variables until all of the Error Variables are less than the user-specified iteration limits. The solution procedure is referred to as SCX. It consists of two phases, the Systems phase and the Controller phase. The X is to imply experimental. SCX computes each next set of State Variables in two phases. In the first phase, SCX fixes the controller positions and modifies the other State Variables by the Newton-Raphson method. This first phase is the Systems phase. Once the Newton-Raphson method has solved the problem for the fixed controller positions, SCX next calculates new controller positions based on Newton's method while treating each sensor-controller pair independently but allowing all to change in one iteration. This phase is the Controller phase. SINFAC is available by license for a period of ten (10) years to approved licensees. The licenced program product includes the source code for the additional routines to SINDA, the SINDA object code, command procedures, sample data and supporting documentation. Additional documentation may be purchased at the price below. SINFAC was created for use on a DEC VAX under VMS. Source code is written in FORTRAN 77, requires 180k of memory, and should be fully transportable. The program was developed in 1988.
Code package of the physics calculation and fuel management of uranium hydride zirconium reactor
International Nuclear Information System (INIS)
The code package of uranium hydride zirconium reactor physics calculation is established. considering the thermalization of H in ZrH, the nuclear data of H in ZrH in WIMS library pattern are provided and WIMS-N2 library is obtained. The cell parameters are calculated using WIMS-D/4 code and SIMS-N2 library. The diffusion calculation is performed using CITATION code and SIXTUS-2 code. The in-core fuel management code XPR-ICFM is obtained on the basis of SIXTUS-2 code. To check the accuracy and reliability of the code package, the critical keff and the value of the control rod of abroad TRIGA, the pulsed reactor in china are calculated. The results are satisfied
Development of tokamak reactor systems analysis code 'TORSAC'
International Nuclear Information System (INIS)
This report describes Tokamak Reactor Systems Analysis Code ''TORSAC'' which has been developed in order to assess the impact of the design choises on reactor systems and to improve tokamak designs in wide parameter range. This computer code has following functions. (1) Systematic sensitivity analysis for a set of given design parameters, (2) Cost calculation of a new reactor concept designed automatically as a result of systematic sensitivity analysis. (author)
Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment
International Nuclear Information System (INIS)
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter und the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 times the hydraulic pipe diameter. (orig.)
Validation of the ATHLET-code 2.1A by calculation of the ECTHOR experiment
International Nuclear Information System (INIS)
Before a numerical code (e.g. ATHLET) is used for simulation of physical phenomena being new or unknown for the code and/or the user, the user ensures the applicability of the code and his own experience of handling with it by means of a so-called validation. Parametric studies with the code are executed for that matter and the results have to be compared with verified experimental data. Corresponding reference values are available in terms of so-called single-effect-tests (e.g. ECTHOR). In this work the system-code ATHLET Mod. 2.1 Cycle A is validated by post test calculation of the ECTHOR experiment due to the above named aspects. With the ECTHOR-tests the clearing of a water-filled model of a loop seal by means of an air-stream was investigated including momentum exchange at the phase interface under adiabatic and atmospheric conditions. The post test calculations show that the analytical results meet the experimental data within the reproducibility of the experiments. Further findings of the parametric studies are: - The experimental results obtained with the system water-air (ECTHOR) can be assigned to a water-steam-system, if the densities of the phases are equal in both cases. - The initial water level in the loop seal has no influence on the results as long as the gas mass flow is increased moderately. - The loop seal is appropriately nodalized if the mean length of the control volumes accords approx. 1.5 tim es the hydraulic pipe diameter. (orig.)
NUFACE: An interface code for the calculation of nuclear responses
International Nuclear Information System (INIS)
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs
NUFACE: An interface code for the calculation of nuclear responses
Energy Technology Data Exchange (ETDEWEB)
Henderson, D.L. (Oak Ridge National Lab., TN (USA)); Gomes, I.C. (Tennessee Univ., Knoxville, TN (USA))
1990-01-01
The NUFACE interface code computes nuclear responses for use in the nuclear analysis of a given tokamak reactor design. The NUFACE code operates on the neutron and gamma fluxes provided by the one-dimensional neutral-particle transport code ONEDANT. Zonewise and zone-boundary responses are computed to obtain both zone-integrated values and maximum surface values. Information on each material mixture within a zone and on each element or isotope constituent of each material is computed. This feature allows for a detailed analysis of the reactor whereby one can easily identify the fractional contribution to the response of interest from each material and each element or isotope. 4 refs., 4 figs., 3 tabs.
International Nuclear Information System (INIS)
This report presents a detailed analysis of the results obtained by five estimate calculations concerning the safety of the first test of the Sodium-water reaction experimental programme, FLASH. This programme is carried out the frame of the LMFBR R and D European Agreement. The calculations were executed by the PLEXUS code of the CASTEM system
Accuracy evaluation of pin exposure calculations in current LWR core design codes
International Nuclear Information System (INIS)
The deregulated utility environment and better utilization of fuel assemblies in nuclear power plants has allowed designers to burn fuel assemblies to maximum allowable exposures. Any uncertainties, associated with the technical approach and numerical methods used to perform pin exposure calculations may cause either peak power exposure to exceed the Nuclear Regulatory Commission (NRC) exposure limit or lead to excessive conservatism and thus inefficient fuel utilization. In this work, a Monte Carlo based coupled depletion code (MCNP5/ORIGEN-S) is utilized to provide reference solutions in order to assess the accuracy of pin power and pin exposure reconstruction methods in the current commercial and licensed three-dimensional (3D) nodal Light Water Reactor (LWR) core design codes. The developed at Pennsylvania State University (PSU) MCNP5/ORIGEN-S coupled depletion code system was validated using measured data from the PSU TRIGA research reactor critical experiments. A number of test cases (modeling benchmarks) representative of LWRs were developed starting from the least complex model towards more complicated and more realistic models. The accuracy evaluation of the pin reconstruction methods was performed by using the CASMO-4 and SIMULATE-3 codes as the representative of current commercial LWR core design systems. Two-dimensional (2D) transport calculations with the TRITON module from the SCALE5 package were employed to produce the spectrum averaged cross-section libraries as a function of burnup for ORIGEN-S calculations. The burnup dependent cross-section libraries are specifically generated for each lattice configuration type. For the MCNP5 calculations continuous cross-section libraries for different isotopes at hot operating temperatures are generated and subsequently utilized. Realistic lattice configurations of the GE13 BWR fuel assemblies (unrodded and rodded) depleted under operating conditions were studied in this research because of their heterogeneous
ERINNI an optical model Fortran IV code for the calculation of multiple cascading particle emissions
International Nuclear Information System (INIS)
ERINNI is an extension of the CERBERO code to the calculations of compound nucleus decay cascades. Up to three successive decays are considered. Optical model transmission coefficients are internally calculated. (authors0
SIMULATE-3K linkage with reactor systems codes
International Nuclear Information System (INIS)
SIMULATE-3K is Studsvik Scandpower's best-estimate three-dimensional core kinetics code. SIMULATE-3K has been coupled to several best-estimate reactor systems codes including, RELAP5-3D, RELAP5-3.3, TRACE V5.0, and RETRAN-3D. The coupled codes can be applied to existing reactors and to advanced reactor designs. The S3K linkage to each of the systems codes is a direct, explicit coupling of the two codes on a synchronous time-step basis. The coupling provides an execution method for the S3K three-dimensional neutronic model using the Nuclear Steam Supply System (NSSS) boundary conditions calculated by the systems code. Also, it allows the S3K calculated total core power and core power distributions to drive the system model core. Detailed calculations from the component codes result in a methodology for analyzing limiting transients such as steam line breaks, rod drops/ejections, and ATWS scenarios. These transient events require detailed three- dimensional core data and information about the behavior of NSSS components. A coupled analysis of these transients is important because the core behavior is closely tied to the NSSS system. For example, to capture the timing and characteristics of the important thermal-hydraulic phenomena and/or operations events, such as valve closures, safety injection, or control system interactions, requires a detailed plant model. The Peach Bottom 2 turbine trip transient is used to assess the accuracy of the coupled code calculations. Comparisons of the important plant parameters to results from RELAP5-3D, RELAP5-3.3, and TRACE V5.0 calculations are shown and discussed. The MSLB benchmark is also used to demonstrate the capabilities of the coupled code systems. Comparisons of the calculated reactor power to the reference data are shown can discussed. The comparisons demonstrate the applicability of S3K, either standalone or coupled with a system analysis code, to properly model system response during accident scenarios. (author)
Computer code for double beta decay QRPA based calculations
International Nuclear Information System (INIS)
The computer code developed by our group some years ago for the evaluation of nuclear matrix elements, within the QRPA and PQRPA nuclear structure models, involved in neutrino-nucleus reactions, muon capture and β± processes, is extended to include also the nuclear double beta decay
A New, Efficient Stellar Evolution Code for Calculating Complete Evolutionary Tracks
Kovetz, Attay; Prialnik, Dina
2008-01-01
We present a new stellar evolution code and a set of results, demonstrating its capability at calculating full evolutionary tracks for a wide range of masses and metallicities. The code is fast and efficient, and is capable of following through all evolutionary phases, without interruption or human intervention. It is meant to be used also in the context of modeling the evolution of dense stellar systems, for performing live calculations for both normal star models and merger-products. The code is based on a fully implicit, adaptive-grid numerical scheme that solves simultaneously for structure, mesh and chemical composition. Full details are given for the treatment of convection, equation of state, opacity, nuclear reactions and mass loss. Results of evolutionary calculations are shown for a solar model that matches the characteristics of the present sun to an accuracy of better than 1%; a $1 \\Msun$ model for a wide range of metallicities; a series of models of stellar populations I and II, for the mass rang...
MUXS: a code to generate multigroup cross sections for sputtering calculations
International Nuclear Information System (INIS)
This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
International Nuclear Information System (INIS)
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements (10B, 3He, 6Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate
International Nuclear Information System (INIS)
Since 1974, Thermodata has been working on developing an Integrated Information System in Inorganic Chemistry. A major effort was carried on the thermochemical data assessment of both pure substances and multicomponent solution phases. The available data bases are connected to powerful calculation codes (GEMINI = Gibbs Energy Minimizer), which allow to determine the thermodynamical equilibrium state in multicomponent systems. The high interest of such an approach is illustrated by recent applications in as various fields as semi-conductors, chemical vapor deposition, hard alloys and nuclear safety. (author). 26 refs., 6 figs
Development of the multistep compound process calculation code
Energy Technology Data Exchange (ETDEWEB)
Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan)
1998-03-01
A program `cmc` has been developed to calculate the multistep compound (MSC) process by Feshback-Kerman-Koonin. A radial overlap integral in the transition matrix element is calculated microscopically, and comparisons are made for neutron induced {sup 93}Nb reactions. Strengths of the two-body interaction V{sub 0} are estimated from the total MSC cross sections. (author)
Development and verification of Monte Carlo burnup calculation system
International Nuclear Information System (INIS)
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.)
Calculating system reliability with SRFYDO
Energy Technology Data Exchange (ETDEWEB)
Morzinski, Jerome [Los Alamos National Laboratory; Anderson - Cook, Christine M [Los Alamos National Laboratory; Klamann, Richard M [Los Alamos National Laboratory
2010-01-01
SRFYDO is a process for estimating reliability of complex systems. Using information from all applicable sources, including full-system (flight) data, component test data, and expert (engineering) judgment, SRFYDO produces reliability estimates and predictions. It is appropriate for series systems with possibly several versions of the system which share some common components. It models reliability as a function of age and up to 2 other lifecycle (usage) covariates. Initial output from its Exploratory Data Analysis mode consists of plots and numerical summaries so that the user can check data entry and model assumptions, and help determine a final form for the system model. The System Reliability mode runs a complete reliability calculation using Bayesian methodology. This mode produces results that estimate reliability at the component, sub-system, and system level. The results include estimates of uncertainty, and can predict reliability at some not-too-distant time in the future. This paper presents an overview of the underlying statistical model for the analysis, discusses model assumptions, and demonstrates usage of SRFYDO.
The JAERI code system for evaluation of BWR ECCS performance
International Nuclear Information System (INIS)
Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)
Code-B-1 for stress/strain calculation for TRISO fuel particle (Contract research)
International Nuclear Information System (INIS)
We have developed Code-B-1 for the prediction of the failure probabilities of the coated fuel particles for the high temperature gas-cooled reactors (HTGRs) under operation by modification of an existing code. A finite element method (FEM) is employed for the stress calculation part and Code-B-1 can treat the plastic deformation of the coating layer of the coated fuel particles which the existing code cannot treat. (author)
The 'hot rod methodology' for intermediate break LOCA calculations with the CATHARE2 code
International Nuclear Information System (INIS)
The nuclear fuel that will be used for next decades in French PWR may be quite different from current ones, including technological evolutions of cladding alloys, new types of assemblies, etc. Beside these structural evolutions, calculation methodologies are also changing with the use of Best-estimate multi-physics multi-scale coupled calculations, and complex physical models development at 3D local scale. In the frame of its LOCA R and D program, the French 'Institut de Radioprotection et de Surete Nucleaire' is developing its own calculation methodology to figure out the maximum cladding temperature and the maximum oxidation rate reached by the hottest rod in the core. This HOT ROD methodology has been worked out for Intermediate Break LOCA calculations using the 'Best-Estimate' CATHARE2 system code. This calculation is based on a first modeling called 1D SYSTEM calculation which simulates the thermohydraulical behavior of the whole reactor. During this computation, key parameters are stored at core ends in order to be used as boundary conditions for a second calculation called 1D CHAINED calculation nodelizing in one dimension the hottest assembly of the core. Based on experimental evidences, the hot assembly has been considered to have no influence on the thermohydraulical behavior of the mean core. Hence its hydraulical environment has been modified at each calculation's timestep in order to fit with the SYSTEM mean one. This paper details the methodology's validation phase on PERICLES 2D BOIL-UP tests, which have quite the same swollen-level evolution's rate as in the boil up phase of an intermediate break LOCA. (author)
International Nuclear Information System (INIS)
The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems
A new assembly-level Monte Carlo neutron transport code for reactor physics calculations
International Nuclear Information System (INIS)
This paper presents a new assembly-level Monte Carlo neutron transport code, specifically intended for diffusion code group-constant generation and other reactor physics calculations. The code is being developed at the Technical Research Centre of Finland (VTT), under the working title 'Probabilistic Scattering Game', or PSG. The PSG code uses a method known as Woodcock tracking to simulate neutron histories. The advantages of the method include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. The main drawback is the inability to calculate reaction rates in optically thin volumes. This narrows the field of application to calculations involving parameters integrated over large volumes. The main features of the PSG code and the Woodcock tracking method are introduced. The code is applied in three example cases, involving infinite lattices of two-dimensional LWR fuel assemblies. Comparison calculations are carried out using MCNP4C and CASMO-4E. The results reveal that the code performs quite well in the calculation cases of this study, especially when compared to MCNP. The PSG code is still under extensive development and there are both flaws in the simulation of the interaction physics and programming errors in the source code. The results presented here, however, seem very encouraging, especially considering the early development stage of the code. (author)
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
Development of the next generation reactor analysis code system, MARBLE
International Nuclear Information System (INIS)
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)
Energy Technology Data Exchange (ETDEWEB)
Holly R. Trellue
1998-12-01
Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.
Energy Technology Data Exchange (ETDEWEB)
Hirayama, Hideo; Namito, Yoshihito; /KEK, Tsukuba; Bielajew, Alex F.; Wilderman, Scott J.; U., Michigan; Nelson, Walter R.; /SLAC
2005-12-20
In the nineteen years since EGS4 was released, it has been used in a wide variety of applications, particularly in medical physics, radiation measurement studies, and industrial development. Every new user and every new application bring new challenges for Monte Carlo code designers, and code refinements and bug fixes eventually result in a code that becomes difficult to maintain. Several of the code modifications represented significant advances in electron and photon transport physics, and required a more substantial invocation than code patching. Moreover, the arcane MORTRAN3[48] computer language of EGS4, was highest on the complaint list of the users of EGS4. The size of the EGS4 user base is difficult to measure, as there never existed a formal user registration process. However, some idea of the numbers may be gleaned from the number of EGS4 manuals that were produced and distributed at SLAC: almost three thousand. Consequently, the EGS5 project was undertaken. It was decided to employ the FORTRAN 77 compiler, yet include as much as possible, the structural beauty and power of MORTRAN3. This report consists of four chapters and several appendices. Chapter 1 is an introduction to EGS5 and to this report in general. We suggest that you read it. Chapter 2 is a major update of similar chapters in the old EGS4 report[126] (SLAC-265) and the old EGS3 report[61] (SLAC-210), in which all the details of the old physics (i.e., models which were carried over from EGS4) and the new physics are gathered together. The descriptions of the new physics are extensive, and not for the faint of heart. Detailed knowledge of the contents of Chapter 2 is not essential in order to use EGS, but sophisticated users should be aware of its contents. In particular, details of the restrictions on the range of applicability of EGS are dispersed throughout the chapter. First-time users of EGS should skip Chapter 2 and come back to it later if necessary. With the release of the EGS4 version
International Nuclear Information System (INIS)
MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (keff, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors)
Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)
International Nuclear Information System (INIS)
The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab
SENVAR: a code for handling chemical uncertainties in solubility calculations
International Nuclear Information System (INIS)
In the planning for a repository for spent nuclear fuel it is important to know the solubility of some important solid phases in order to, for example, predict migration of radionuclides from the repository. The method presented in the present paper investigates the effect of uncertainties in thermodynamical data, i.e. stability and solubility constants, for the calculated solubility of a solid phase. The adopted approach is simple Monte Carlo sampling. The investigation is mainly made in three steps. First a preliminary sensitivity analysis where the important parameters are determined. This is done by holding each of the parameters at a fixed value for a given number of solubility calculations. During this time all other parameters are varied according to a pre-set random matrix. The variance for each stationary parameter is then calculated and the one with the smallest variance is deemed the most important one and so on. The parameters that are deemed important are then transferred into the uncertainty analysis. There, each parameter may be given a separate interval for the uncertainty and then a couple of thousand solubility calculations are made where the values of the parameters are varied according to the Monte Carlo method. The results from these calculations are used to estimate the effect of the uncertainties in a plot showing the density function of the solubility and some statistical estimators. The solubility calculations are also used to give data to a stepwise regression program which estimates the importance of each parameter entered into the uncertainty analysis. The regression error is also shown in order to make it easy to determine which values may be correct or not
The use of the codes from MCU family for calculations of WWER type reactors
International Nuclear Information System (INIS)
The MCU-RFFI/A and MCU-REA codes developed within the framework of the long term MCU project are widely used for calculations of neutron physic characteristics of WWER type reactors. Complete descriptions of the codes are available in both Russian and English. The codes are verified and validated by means of the comparison of calculated results with experimental data and mathematical benchmarks. The codes are licensed by Russian Nuclear and Criticality Safety Regulatory Body (Gosatomnadzor RF) (Code Passports: N 61 of 17.10.1966 and N 115 of 02.03.2000 accordingly)). The report gives examples of WWER reactor physic tasks important for practice solved using the codes from the MCU family. Some calculational results are given too. (Authors)
A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core
Energy Technology Data Exchange (ETDEWEB)
Chang, Won Pyo; Ha, Ki Suk; Lee, Yong Bum
2009-09-15
A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has
A Comparative Study of the Code Calculations for Local Flow Blockages in the KALIMER-150 Core
International Nuclear Information System (INIS)
A sub-channel blockage may be caused by ingression of damaged fuel debris or foreign obstacles into a core fuel subassembly for a liquid metal reactor(LMR) due to its geometrical compactness of the core design. Local coolant temperature could rise during the incident and it might eventually lead to the degradation of the fuel rods. An analysis computer code is obviously needed not only to assure the safe design of the core, but also to design an effective monitoring system to prevent it from propagating to a serious consequence. The code, therefore, must be capable of representing the thermal-hydraulic phenomena anticipated during the incident reasonably enough to be used to evaluate fuel rod intactness. Most of the technically leading countries for LMR have developed and are using the codes for sub-channel blockage analyses, Korea couldn't afford the resources sponsoring the develop of such a code past years. It was realized later that such an analysis code would ultimately be a prerequisite in the future licensing process of KALIMER as its conceptual design was being elaborated. Since those advanced countries had been reluctant to transfer such the codes, MATRA-LMR/FB had to be developed independently in Korea. It is a revised version of the existing MATRA-LMR code which was aimed for the core sub-channel analysis of LMRs. Some of its models have been improved so appropriately to be able to analyze the sub-channel blockages. Nevertheless, an experiment relevant to the sub-channel blockages had never been conducted for investigating the phenomenon in Korea. Further more, very few experimental data are available on published papers or reports world wide. Under this circumstance, a study has been made as an effort to evaluate the prediction capability of the MATRA-LMR/FB code by comparing the calculation results of the SABRE code, which had already been applied to EFR design. In result, a discrepancy has been observed in a case, but an overall agreement has been
SAMDIST: A computer code for calculating statistical distributions for R-matrix resonance parameters
Energy Technology Data Exchange (ETDEWEB)
Leal, L.C.; Larson, N.M.
1995-09-01
The SAMDIST computer code has been developed to calculate distribution of resonance parameters of the Reich-Moore R-matrix type. The program assumes the parameters are in the format compatible with that of the multilevel R-matrix code SAMMY. SAMDIST calculates the energy-level spacing distribution, the resonance width distribution, and the long-range correlation of the energy levels. Results of these calculations are presented in both graphic and tabular forms.
Blockage calculation of LMFBR core subassembly with subchannel code-SOBOS
International Nuclear Information System (INIS)
Sodium-boiling is a very important subject to be considered in the Liquid Metal Fast Breeder Reactor (LMFBR) design. Blockage is one of the most important causes of sodium boiling. The author shows the calculation results with subchannel code 'SOBOS' with the advanced subchannel model. And the results of calculations match that of experiments very well, indicating that the subchannel code could be used to calculate the blockage boiling
Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code
Litaize, Olivier; Archier, Pascal; Becker, Bjorn; Schillebeeckx, Peter; Kopecky, Stefan
2013-03-01
This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR). In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.
Validation of capture yield calculations in the Resolved Resonance Energy Range with CONRAD code
Directory of Open Access Journals (Sweden)
Schillebeeckx Peter
2013-03-01
Full Text Available This paper deals with the validation of the multiple scattering corrections developed in the CONRAD code for the capture yield calculations in the Resolved Resonance energy Range (RRR. In order to calculate the capture yields, analytic and stochastic calculation schemes implemented in CONRAD are described and compared with the analysis code SAMMY/SAMSMC. The results are in excellent agreement for a variety of samples. We concentrate the discussion here on 238U, 197Au and 55Mn.
Numerical identification of bacteria with a hand-held calculator as an alternative to code books.
Schindler, J; Schindler, Z
1982-01-01
The Hewlett-Packard HP 41C hand-held calculator can be used for the numerical identification of bacteria. The dimensions of the identification matrix are limited to about 30 by 22; however, many groups of clinically important bacteria can be numerically identified by this method. Hand-held calculators can be used as an alternative to code books. At present, these calculators and additional tests can help solve identification problems in profiles not contained in code books.
Benchmark calculation of nuclear design code for HCLWR
International Nuclear Information System (INIS)
In the calculation of the lattice cell for High Conversion Light Water Reactors, big differences of nuclear design parameters appear between the results obtained by various methods and nuclear data libraries. The validity of the calculation can be verified by the critical experiment. The benchmark calculation is also efficient for the estimation of the validity in wide range of lattice parameters and burnup. As we do not have many measured data. The benchmark calculations were done by JAERI and MAPI, using SRAC and WIMS-E respectively. The problem covered the wide range of lattice parameters, i.e., from tight lattice to the current PWR lattice. The comparison was made on the effective multiplication factor, conversion ratio, and reaction rate of each nuclide, including burnup and void effects. The difference of the result is largest at the tightest lattice. But even at that lattice, the difference of the effective multiplication factor is only 1.4 %. The main cause of the difference is the neutron absorption rate U-238 in resonance energy region. The difference of other nuclear design parameters and their cause were also grasped. (author)
A CFD validation methodology for containment code calculations of hydrogen mixing and recombination
International Nuclear Information System (INIS)
In the frame of ANSALDO activities on containment hydrogen accident events, a simulation procedure was developed to qualify and verify the calculations performed by simplified containment computer codes through the use of a full 3-D Navier Stokes solver. The methodology aims to reduce the computational time usually associated with a general purpose CFD code complete simulation of the containment transient, limiting on the other hand the loss of accuracy typical of the use of a simplified Containment Code. This goal has been fulfilled by the development of a calculation procedure organised in several different steps able to verify the calculated transient parameters by GOTHIC3.4 (the simplified code) with specific calculations performed with CFX4.2 (the CFD code). The paper describes the main milestones of the methodology development and summarizes main results, findings, as well the possible direction of use of the performed work. (authors)
Development and validation of Monte-Carlo burnup calculation code MCNTRANS
International Nuclear Information System (INIS)
A new nuclear fuel burnup calculation code MCNTRANS based on MCNP was introduced in this paper. The neutronics calculation parameter was extracted from the MCNP5 reaction rate tally result, while a graph theory algorithm was implemented to track the burnup chain and the analytic solution of the Bateman equation was given. At the same time, the detailed physical process was considered to improve the accuracy and serviceability of this code, and prediction-correction method was used to allow a large burnup step. The OECD/NEA and JAERI pin cell benchmark problems were used to validate the code MCNTRANS while a reference result was given by other code. It can be concluded that the calculation results of MCNTRANS are generally consistent with the experimental result and that of the other burnup codes, and part of the actinides and fission products calculation result show better accuracy. (authors)
Design, experiments and Relap5 code calculations for the perseo facility
International Nuclear Information System (INIS)
Research on innovative safety systems for light water reactors addressed to heat removal by in-pool immersed heat exchangers, led to design, build-up and test the PERSEO facility at SIET laboratories. The research started with the CEA-ENEA proposal of improving the GE-SBWR isolation condenser system, by moving the triggering valve from the high pressure primary side of the reactor to the low pressure pool side. A new configuration of the system was defined with the heat exchanger contained in a small pool, connected at bottom and top to a large water reservoir pool, the triggering valve being located on the pool bottom connecting pipe. ENEA funded the whole activity that included the definition and build-up of a new heat exchanger pool, on the basis of the already existing PANTHERS IC-PCC facility, at SIET laboratories, and the new plant requirements. The heat exchanger connections to the pressure vessel were maintained. An experimental campaign was executed at full scale and full thermal-hydraulic conditions for investigating the behaviour and performance of the plant in steady and unsteady conditions. The Relap5 code was utilised during all phases of the research: for the heat exchanger pool dimension definition and from pre-test and post-test analyses. The Cathare code was applied too from pre-test and post-test analyses. This paper deals with the experimental and calculated results limited to the Relap5 code
Nuclear modules of ITER tokamak systems code
International Nuclear Information System (INIS)
Nuclear modules were developed to model various reactor components in the ITER systems code. These modules include first wall, tritium breeding blanket (or shield), bulk shield, reactor vault, impurity control, and tritium system. The function of these modules is to define the performance parameters for each component as a function of the reactor operating conditions. Several design options and cost algorithms are included for each component. The first wall, blanket and shield modules calculate the beryllium zone thickness, the disruptions results, the nuclear responses in different components including the toroidal field coils. Tungsten shield/water coolant/steel structure and steel shield/water coolant are the shield options for the inboard and outboard sections of the reactor. Lithium nitrate dissolved in the water coolant with a variable beryllium zone thickness in the outboard section of the reactor provides the tritium breeding capability. The reactor vault module defines the thickness of the reactor wall and the roof based on the dose equivalent during operation including skyshine contribution. The impurity control module provides the design parameters for the divertor including plate design, heat load, erosion rate, tritium permeation through the plate material to the coolant, plasma contamination by sputtered impurities, and plate lifetime. Several materials: Be, C, V, Mo, and W can be used for the divertor plate to cover a range of plasma edge temperatures. The tritium module calculates tritium and deuterium flow rates for the reactor plant. The tritium inventory in the fuelers, neutral beams, vacuum pumps, impurity control, first wall, and blanket is calculated. Tritium requirements are provided for different operating conditions. The nuclear models are summarized in this paper including the different design options and key analyses of each module
Calculation capability of NETFLOW++ code for natural circulation in sodium cooled fast reactor
International Nuclear Information System (INIS)
The present paper describes the simulation of the natural circulation in the secondary heat transport system (HTS) after an intentional plant trip of the experimental fast reactor 'Joyo' at 140MWt power using the plant dynamics analysis code NETFLOW++. This code is an integrated network code to calculate the nuclear steam supply system (NSSS) and the balance of the plant (BOP), i.e., turbine/feedwater system developed by the author. Up to now, the code has been validated using transient data of the experimental sodium facility PLANDTL, experimental fast reactor 'Joyo' and the prototype fast breeder reactor 'Monju'. These validations are steps to evaluate the natural circulation of a large-scale fast breeder reactor because it is important that the code can simulate not only a large plant but also a small plant or apparatus in order to have versatility. Therefore, the former validation results are introduced to show the degree of agreement. In order to consolidate the applicability of the code to the evaluation of the natural circulation, the present simulation was selected. Natural circulation tests were conducted at the experimental fast reactor 'Joyo' for the two reactor cores. The latest natural circulation test was conducted from the rated power of the Mark-II irradiation core at 100 MWt, and the NETFLOW++ code was validated using the measured data. At this occasion, the importance of the inter-subassembly heat transfer was recognized by the author in order to predict the core outlet temperature although the whole plant behavior was little effected. The 'Monju' reactor conducted some natural circulation tests in the primary and secondary heat transport systems utilizing the heat of primary pump operation. The NETFLOW++ code was validated using these data. Although these were not the complete natural circulation expected in the power plant,these results were important because basic factors relating the natural circulation such as the buoyancy force,the static
International Nuclear Information System (INIS)
The authors report calculations performed using the MNCP and PENELOPE codes to determine the Hp(3)/K air conversion coefficient which allows the Hp(3) dose equivalent to be determined from the measured value of the kerma in the air. They report the definition of the phantom, a 20 cm diameter and 20 cm high cylinder which is considered as representative of a head. Calculations are performed for an energy range corresponding to interventional radiology or cardiology (20 keV-110 keV). Results obtained with both codes are compared
ETF system code: composition and applications
International Nuclear Information System (INIS)
A computer code has been developed for application to ETF tokamak system and conceptual design studies. The code determines cost, performance, configuration, and technology requirements as a function of tokamak parameters. The ETF code is structured in a modular fashion in order to allow independent modeling of each major tokamak component. The primary benefit of modularization is that it allows updating of a component module, such as the TF coil module, without disturbing the remainder of the system code as long as the input/output to the modules remains unchanged. The modules may be run independently to perform specific design studies, such as determining the effect of allowable strain on TF coil structural requirements, or the modules may be executed together as a system to determine global effects, such as defining the impact of aspect ratio on the entire tokamak system
Radiative Transfer Code: Application to the calculation of PAR
Indian Academy of Sciences (India)
D Emmanuel; D Phillippe; C Malik
2000-12-01
The production of carbon in the ocean, the so-called primary production, depends on various physico- biological parameters: the biomass and nutrient amounts in oceans, the salinity and temperature of the water and the light available in the water column. We focus on the visible spectrum of the solar radiation defined as the Photosynthetically Active Radiation (PAR). We developed a model (Chami et al. 1997) to simulate the behavior of the solar beam in the atmosphere and the ocean. We first describe the theoretical basis of the code and the method we used to solve the radiative transfer equation (RTE): the successive orders of scattering (SO). The second part deals with a sensitivity study of the PAR just above and below the sea surface for various atmospheric conditions. In a cloudy sky, we computed a ratio between vector fluxes just above the sea surface and spherical fluxes just beneath the sea surface. When the optical thickness of the cloud increases this ratio remains constant and around 1.29. This parameter is convenient to convert vector flux at the sea surface as retrieved from satellite to PAR. Subsequently, we show how solar radiation as vector flux rather than PAR leads to an underestimate of the primary production up to 40% for extreme cases.
Development of leak rate calculation model and code in piping
International Nuclear Information System (INIS)
Background: With the development of fracture mechanics, Leak-Before-Break (LBB) is widely used in nuclear power plant piping design. Purpose: In order to support the application of LBB, leak rate through crack need to be calculated. Methods: In this text, an analytical flow model is developed based on homogeneous non-equilibrium model, a computer program is also developed based on this analytical model. Results: Comparison between the results from the above program and test results shows that deviations of calc. results to test results are within ±50%. Conclusions: Conclusions can be got that this analytical model and computer program meet the requirement of engineering application. (authors)
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Baratta, A.J.
1997-07-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.
Verification of DeCART/CAPP code system for VHTR by HTTR core analysis
International Nuclear Information System (INIS)
The DeCART/CAPP code system has been developed and verified against the numerical benchmark calculations for an HTTR. The reference calculations have been carried out by the Monte Carlo McCARD code in which a double heterogeneity model was used. Verification results show that the DeCART/CAPP code system gives less negative MTC and RTC than the McCARD code, and thus the DeCART code overestimates the multiplication factors at states with a high moderator and reflector temperature. However, the DeCART/CAPP code system predicts more negative FTC than McCARD code does. In the depletion calculation for the HTTR single cell and single block, the error of DeCART/CAPP code system increases with the burnup (authors)
Code of hybrid calculation for the study of heat exchangers
International Nuclear Information System (INIS)
A series integration method has been chosen, machine time being similar to space, the computer storing the solutions passed to allow for an approximation at the finished differences of the timed derivatives. Space distributions of the functions of state (enthalpy and temperature) of the two fluids are calculated by analogical integration in the direction of the circulation of the fluids. This is made possible by the disconnection of the two fluids thanks to a method of prediction-correction for the equation of the wall. The great rapidity of the analogical integrations and the absence of iterations make the calculation in actual time possible (assuming that 200 points are taken for the storage of the solutions passed, the time required for a full calculation is 0.4 second). It is, therefore, possible to integrate this simulation in an actual (for example Nuclear Station) or simulated loop. The exchanger may or may not be with counter-current, with or without change of phase. Thermodynamic tables with two variables are introduced into the computer, interpolation is effected in analogical and this allows, inter alia, enthalpy, temperature, pressure and volumic mass to be perfectly in phase one with the other. The work bears most particularly on the accuracy and stability of simulation of the timed derivatives which are obtained by difference between an analogical function and the same function stored in the computer at the preceding step. Particular emphasis is made to bear on the operation of the model, a monitor program controlling the various phases of utilization in function of the type of exchanger, experiments to be made; a program of control of the results enables the operator to obtain the results either immediately in printed form, curve tracer or oscilloscope, put to scale, or in deferred a more elaborate processing of the results. The same program allows for the simulation of any exchanger (different fluid, variable geometry, etc.) or any number of
Code Formal Verification of Operation System
Directory of Open Access Journals (Sweden)
Yu Zhang
2010-12-01
Full Text Available with the increasing pressure on non-function attributes (security, safety and reliability requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operation system kernel in C code level. We present a case study to the verification of real-world C systems code derived from an implementation of μC/OS – II in the end.
Power distribution and fuel depletion calculation for a PWR, using LEOPARD and CITATION codes
International Nuclear Information System (INIS)
By modifying LEOPARD a new program, LEOCIT, has been developed in which additional subroutines prepare cross-section libraries in 1, 2 or 4 energy groups and subsequently record these on disc or tape in a format appropriate for direct input to the CITATION code. Use of LEOCIT in conjunction with CITATION is demonstrated by simulating the first depletion cycle of Angra Unit 1. In these calculations two energy groups are used in 1/4, X - Y geometry to give the soluble boron curve, the fuel depletion and the point to point power distribution in Angra 1. Finally relevant results obtained here are compared with those published by Westinghouse, CNEN and Furnas and recommendations are made to improve the system of neutronic calculation developed in this work. (Author)
FOOD: an interactive code to calculate internal radiation doses from contaminated food products
International Nuclear Information System (INIS)
An interactive code, FOOD, has been written in BASIC for the UNIVAC 1108 to facilitate calculation of internal radiation doses to man from radionuclides in food products. In the dose model, vegetation may be contaminated by either air or irrigation water containing radionuclides. The model considers two mechanisms for radionuclide contamination of vegetation: direct deposition on leaves and uptake from soil through the root system. The user may select up to 14 food categories with corresponding consumption rates, growing periods and either irrigation rates or atmospheric deposition rates. These foods include various kinds of produce, grains and animal products. At present, doses may be calculated for the skin, total body and five internal organs from 190 radionuclides. Dose summaries can be displayed at the local terminal. Further details on percent contribution to dose by nuclide and by food type are available from an auxiliary high-speed printer. This output also includes estimated radionuclide concentrations in soil, plants and animal products
The PASC-3 code system and the UNIPASC environment
International Nuclear Information System (INIS)
A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and its associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified, Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab
Method of tallying adjoint fluence and calculating kinetics parameters in Monte Carlo codes
International Nuclear Information System (INIS)
A method of using iterated fission probability to estimate the adjoint fluence during particles simulation, and using it as the weighting function to calculate kinetics parameters βeff and A in Monte Carlo codes, was introduced in this paper. Implements of this method in continuous energy Monte Carlo code MCNP and multi-group Monte Carlo code MCMG are both elaborated. Verification results show that, with regardless additional computing cost, using this method, the adjoint fluence accounted by MCMG matches well with the result computed by ANISN, and the kinetics parameters calculated by MCNP agree very well with benchmarks. This method is proved to be reliable, and the function of calculating kinetics parameters in Monte Carlo codes is carried out effectively, which could be the basement for Monte Carlo codes' utility in the analysis of nuclear reactors' transient behavior. (authors)
Study on the GEANT4 code applications to the dose calculation using imaging data
Lee, JeongOk; Kim, JhinKee; Kwon, HyeongCheol; Kim, JungSoo; Kim, BuGil; Jeong, DongHyeok
2015-01-01
The use of GEANT4 code has increased in the medical field. There are various studies to calculate the patient dose distributions with the GEANT4 code using the imaging data. In present study, the Monte Carlo simulations based on the DICOM data were performed to calculate absorbed dose in the patient's body. Various visualization tools were equipped in the GEANT4 code to display the detector construction, however there are limitations to display the DICOM images. In addition, it is difficult to display the dose distributions on the imaging data of the patient. Recently, gMocren code, volume visualization tool for GEANT4 simulation, has been developed and used in volume visualization of image files. In this study, the imaging data based absorbed dose distributions in patient were performed by using the gMocren code. The dosimetric evaluations with TLD and film dosimetry methods were carried out to verify the calculation results.
Study on GEANT4 code applications to dose calculation using imaging data
Lee, Jeong Ok; Kang, Jeong Ku; Kim, Jhin Kee; Kwon, Hyeong Cheol; Kim, Jung Soo; Kim, Bu Gil; Jeong, Dong Hyeok
2015-07-01
The use of the GEANT4 code has increased in the medical field. Various studies have calculated the patient dose distributions by users the GEANT4 code with imaging data. In present study, Monte Carlo simulations based on DICOM data were performed to calculate the dose absorb in the patient's body. Various visualization tools are installed in the GEANT4 code to display the detector construction; however, the display of DICOM images is limited. In addition, to displaying the dose distributions on the imaging data of the patient is difficult. Recently, the gMocren code, a volume visualization tool for GEANT4 simulation, was developed and has been used in volume visualization of image files. In this study, the imaging based on the dose distributions absorbed in the patients was performed by using the gMocren code. Dosimetric evaluations with were carried out by using thermo luminescent dosimeter and film dosimetry to verify the calculated results.
International Nuclear Information System (INIS)
A short review is given for models using in thermohydraulic code HYDRA-IBRAE/LM and the results of verification of calculational code on the problems of liquid metal coolant flow and heat transfer. It is shown that developed version of code HYDRA-IBRAE/LM simulates one-phase flow of lead, sodium and lead-bismuth coolants with high accuracy and the processes of sodium boiling with good one. The results of applied calculations of once-through steam generators are considered. It is pointed out that code HYDRA-IBRAE/LM represents correctly physics of the processes and phenomena taking place in the steam generator. The results of cross-verification calculations of lead steam generator by codes HYDRA-IBRAE/LM and TRIANA-4 show satisfactory agreement of results on temperatures of coolants and materials of channel walls in Field tube
International Nuclear Information System (INIS)
In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)
International Nuclear Information System (INIS)
The double-heterogeneity characterising pebble-bed high temperature reactors (HTRs) makes Monte Carlo based calculation tools the most suitable for detailed core analyses. These codes can be successfully used to predict the isotopic evolution during irradiation of the fuel of this kind of cores. At the moment, there are many computational systems based on MCNP that are available for performing depletion calculation. All these systems use MCNP to supply problem dependent fluxes and/or microscopic cross sections to the depletion module. This latter then calculates the isotopic evolution of the fuel resolving Bateman's equations. In this paper, a comparative analysis of three different MCNP-based depletion codes is performed: Montburns2.0, MCNPX2.6.0 and BGCore. Monteburns code can be considered as the reference code for HTR calculations, since it has been already verified during HTR-N and HTR-N1 EU project. All calculations have been performed on a reference model representing an infinite lattice of thorium-plutonium fuelled pebbles. The evolution of k-inf as a function of burnup has been compared, as well as the inventory of the important actinides. The k-inf comparison among the codes shows a good agreement during the entire burnup history with the maximum difference lower than 1%. The actinide inventory prediction agrees well. However significant discrepancy in Am and Cm concentrations calculated by MCNPX as compared to those of Monteburns and BGCore has been observed. This is mainly due to different Am-241 (n,γ) branching ratio utilized by the codes. The important advantage of BGCore is its significantly lower execution time required to perform considered depletion calculations. While providing reasonably accurate results BGCore runs depletion problem about two times faster than Monteburns and two to five times faster than MCNPX.
Preliminary study of coupling CFD code FLUENT and system code RELAP5
International Nuclear Information System (INIS)
Highlights: • System code RELAP5/MOD3.1 is coupled with CFD code FLUENT through DLL and UDF. • Transient water flow in a simple straight tube is tested using the coupled tool. • Simulation of Edwards’ pipe blowdown experiment using the coupled tool is conducted. • Coupled analysis of a more comprehensive thermal–hydraulic system is performed. - Abstract: The present paper discusses a coupling strategy of the 3D (three-dimensional) computational fluid dynamics (CFD) code ANSYS-FLUENT with the best estimate 1D (one-dimensional) thermal–hydraulic system code RELAP5/MOD3.1. Preliminarily, by using DLL (Dynamic Link Library) technology and FLUENT UDF (User Defined Functions), an explicit coupling method expected to be able to support the analysis of multi-purpose thermal–hydraulic phenomena in nuclear reactor systems has been developed. Calculations for two test cases using the coupled FLUENT/RELAP5 code have been carried out to test and demonstrate the coupling capability: (i) the first one consisting of single-phase water transient flow in a square straight tube with well controlled mass flow rates; (ii) the second one illustrating the process of single-phase water flow in a system including two closed loops and one vessel, on which loss of loop water flow due to pump trip and increase of loop water temperature are studied. Both reasonable 1D systematic behaviors and 3D distribution information are naturally obtained for the test cases. Besides, a study of a highly transient experiment problem, i.e. Edwards–O’Brien pipe blowdown problem, has been performed by using the coupled FLUENT/RELAP5 code. The results are compared with standalone RELAP5 calculation and available experimental data, which shows the coupled FLUENT/RELAP5 code’s acceptable potential for the capability of analyzing either simple single-phase or complex two-phase flow problem
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
International Nuclear Information System (INIS)
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs
Chromaticity calculations and code comparisons for x-ray lithography source XLS and SXLS rings
Energy Technology Data Exchange (ETDEWEB)
Parsa, Z.
1988-06-16
This note presents the chromaticity calculations and code comparison results for the (x-ray lithography source) XLS (Chasman Green, XUV Cosy lattice) and (2 magnet 4T) SXLS lattices, with the standard beam optic codes, including programs SYNCH88.5, MAD6, PATRICIA88.4, PATPET88.2, DIMAD, BETA, and MARYLIE. This analysis is a part of our ongoing accelerator physics code studies. 4 figs., 10 tabs.
SFR whole core burnup calculations with TRIPOLI-4 Monte Carlo code
International Nuclear Information System (INIS)
Under the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration benchmark was recently established on the neutronic analysis of four Sodium-cooled Fast Reactor (SFR) concepts of the Generation- IV nuclear energy systems. As the whole core Monte Carlo depletion calculation is one of the essential challenges of current reactor physics studies, the continuous-energy TRIPOLI-4 Monte Carlo transport code was firstly used in this study to perform whole core 3D neutronic calculations for these four SFR cores. Two medium size (1000 MWt) and two large size (3600 MWt) SFR of GEN-IV systems were analyzed. The medium size SFR concepts are from the Advanced Burner Reactors (ABR). The large size SFR concepts are from the self-breeding reactors. The TRIPOLI-4 depletion calculations were made with MOX and metallic U-Pu-Zr fuels for the ABR cores and with MOX and Carbide (U,Pu)C fuels for the self-breeding cores. The whole core reactor physics parameters calculations were performed for the BOEC and EOEC (Beginning and End of Equilibrium Cycle) conditions. This paper summarizes the TRIPOLI-4 calculation results of Keff, βeff, sodium void worth, Doppler constant, control rod worth, and core power distributions for the BOEC and EOEC conditions. The one-cycle depletion calculation results of the core inventory of U and TRU (Pu, Am, Cm, and Np) are also analyzed, after 328.5 days depletion irradiation for the ABR cores, 410 days for the large MOX core, and 500 days for the large carbide core. (author)
Core design calculations of IRIS reactor using modified CORD-2 code package
International Nuclear Information System (INIS)
Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)
Validation of the VTT's reactor physics code system
International Nuclear Information System (INIS)
At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)
Fission gas activities in the fuel-to-clad gap calculated with the code FUROM
International Nuclear Information System (INIS)
The fuel behaviour code FUROM (FUel ROd Model) has been in use and under improvement for several years at the Hungarian Academy of Sciences KFKI Atomic Energy Research Institute. Several new features are added to it each year. In the present paper an extended fission gas release model is introduced. This model is suitable for the calculation of the release of not only stable but also radioactive isotopes. Code calculations are compared to international results. (authors)
Energy Technology Data Exchange (ETDEWEB)
Perfetti, Christopher M [ORNL; Martin, William R [University of Michigan; Rearden, Bradley T [ORNL; Williams, Mark L [ORNL
2012-01-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the SHIFT Monte Carlo code within the Scale code package. The methods were used for several simple test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods.
Application of Monte Carlo code EGS4 to calculate gamma exposure buildup factors
International Nuclear Information System (INIS)
Exposure buildup factors up to 40 mean free paths ranging from 0.015 MeV to 15 MeV photon energy were calculated by using the Monte Carlo simulation code EGS4 for ordinary concrete. The calculation involves PHOTX cross section library, a point isotropic source, infinite uniform medium model and a particle splitting method and considers the Bremsstrahlung, fluorescent effect, correlative (Rayleigh) scatter. The results were compared with the relevant data. Results show that the data of the buildup factors calculated by the Monte Carlo code EGS4 was reliable. The Monte Carlo method can be used widely to calculate gamma-ray exposure buildup factors. (authors)
International Nuclear Information System (INIS)
The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested
Evaluation of Monte Carlo Codes Regarding the Calculated Detector Response Function in NDP Method
Energy Technology Data Exchange (ETDEWEB)
Tuan, Hoang Sy Minh; Sun, Gwang Min; Park, Byung Gun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The basis of the NDP is the irradiation of a sample with a thermal or cold neutron beam and the subsequent release of charged particles due to neutron-induced exoergic charged particle reactions. Neutrons interact with the nuclei of elements and release mono-energetic charged particles, e.g. alpha particles or protons, and recoil atoms. Depth profile of the analyzed element can be obtained by making a linear transformation of the measured energy spectrum by using the stopping power of the sample material. A few micrometer of the material can be analyzed nondestructively, and on the order of 10nm depth resolution can be obtained depending on the material type with NDP method. In the NDP method, the one first steps of the analytical process is a channel-energy calibration. This calibration is normally made with the experimental measurement of NIST Standard Reference Material sample (SRM-93a). In this study, some Monte Carlo (MC) codes were tried to calculate the Si detector response function when this detector accounted the energy charges particles emitting from an analytical sample. In addition, these MC codes were also tried to calculate the depth distributions of some light elements ({sup 10}B, {sup 3}He, {sup 6}Li, etc.) in SRM-93a and SRM-2137 samples. These calculated profiles were compared with the experimental profiles and SIMS profiles. In this study, some popular MC neutron transport codes are tried and tested to calculate the detector response function in the NDP method. The simulations were modeled based on the real CN-NDP system which is a part of Cold Neutron Activation Station (CONAS) at HANARO (KAERI). The MC simulations are very successful at predicting the alpha peaks in the measured energy spectrum. The net area difference between the measured and predicted alpha peaks are less than 1%. A possible explanation might be bad cross section data set usage in the MC codes for the transport of low energetic lithium atoms inside the silicon substrate.
Decay Power Calculation for Safety Analysis of Innovative Reactor Systems
International Nuclear Information System (INIS)
In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)
Using deterministic codes to accelerate continuous energy Monte-Carlo standards calculations
International Nuclear Information System (INIS)
Deterministic codes are usually used for critical parameters or one dimension geometry calculations. Advantages of the use of deterministic codes are speed of the calculation and the absence of standard deviation on the keff results. Nevertheless, the deterministic results are affected by several intrinsic uncertainties as energetic condensation or self-shielding. So the way to proceed at CEA expert criticality group (CEA/SERMA/CP2C) is to always check the main results (minimum critical or maximal permissible values and un-moderated values) with a punctual Monte Carlo calculation. These last years, in particular cases (pure actinide fissile media, exotic reflectors), large discrepancies have been observed between the keff calculated by the CRISTAL V1 route reference (continuous energy Monte Carlo code TRIPOLI-4) and the keff target (by the standard route APOLLO2-Sn). The problematic for these cases was how to transpose the keff discrepancies observed between standard and reference routes to the dimensions (mass, thickness...) or how to reduce the keff discrepancies using optimized options of the deterministic code. One solution to transpose discrepancies is to iterate on dimensions using a punctual Monte Carlo code to achieve the desired keff eigenvalue. But, the amount of time for obtaining a good standard deviation and also the desired keff eigenvalue inside the Monte Carlo calculation uncertainty can quickly increase. The principle of the method presented in this paper is that the discrepancy between deterministic code and Monte-Carlo code, calculated at the same dimension, is low variable with the dimension. Therefore, correcting the keff eigenvalue on which the deterministic code converge with the discrepancy observed, leads to a dimension nearer to the true dimension (i.e. the dimension where Monte-Carlo code keff calculation is close to the keff eigenvalue). If the keff eigenvalue is outside the Monte Carlo uncertainty, the discrepancy is recalculated and
A general dimensional neutron diffusion calculation code: ADC
International Nuclear Information System (INIS)
A FORTRAN computer program ADC is developed for the FACOM 230-75 computer to be capable of solving eigenvalue problems of neutron diffusion equation in one, two and three spatial dimensions. The available coordinate systems are orthogonal (X), (X,Y), (X,Y,Z) and cylindrical (R,Z), (R,THETA), (R,THETA,Z). The outer boundary condition for the neutron flux can be chosen to be symmetric, zero flux or log-derivative condition. The present program can be used also for obtaining the adjoint flux. (author)
International Nuclear Information System (INIS)
ZOCO V is a computer code which can calculate the time- and space- dependent pressure distribution in containments of water-cooled nuclear power reactors (both full pressure containments and pressure suppression systems) following a loss-of-coolant accident, caused by the rupture of a main coolant or steam pipe
Systems Improved Numerical Fluids Analysis Code
Costello, F. A.
1990-01-01
Systems Improved Numerical Fluids Analysis Code, SINFAC, consists of additional routines added to April, 1983, version of SINDA. Additional routines provide for mathematical modeling of active heat-transfer loops. Simulates steady-state and pseudo-transient operations of 16 different components of heat-transfer loops, including radiators, evaporators, condensers, mechanical pumps, reservoirs, and many types of valves and fittings. Program contains property-analysis routine used to compute thermodynamic properties of 20 different refrigerants. Source code written in FORTRAN 77.
Directory of Open Access Journals (Sweden)
Diego Ferraro
2011-01-01
Full Text Available Monte Carlo neutron transport codes are usually used to perform criticality calculations and to solve shielding problems due to their capability to model complex systems without major approximations. However, these codes demand high computational resources. The improvement in computer capabilities leads to several new applications of Monte Carlo neutron transport codes. An interesting one is to use this method to perform cell-level fuel assembly calculations in order to obtain few group constants to be used on core calculations. In the present work the VTT recently developed Serpent v.1.1.7 cell-oriented neutronic calculation code is used to perform cell calculations of a theoretical BWR lattice benchmark with burnable poisons, and the main results are compared to reported ones and with calculations performed with Condor v.2.61, the INVAP's neutronic collision probability cell code.
Code Formal Verification of Operation System
Yu Zhang; Yunwei Dong; Huo Hong; Fan Zhang
2010-01-01
with the increasing pressure on non-function attributes (security, safety and reliability) requirements of an operation system, high–confidence operation system is becoming more important. Formal verification is the only known way to guarantee that a system is free of programming errors. We research on formal verification of operation system kernel in system code level and take theorem proving and model checking as the main technical methods to resolve the key techniques of verifying operatio...
Calculation of the CAREM reactor with the HUEMUL-PUMA-THERMIT chain of codes
International Nuclear Information System (INIS)
The purpose of the work was the evaluation of the the CAREM 25 reactor core, using a chain of codes (HUEMUL-PUMA-THERMIT) different to the one used in the original design (CONDOR-CITVAP-THERMIT). First, we performed a partial validation of the our codes in lattices similar to CAREM and reproduced a benchmark for simulation of gadolinium burnup. The results were considered satisfactory for this stage of the project. Then, we calculated the core along the normal operating equilibrium cycle and in hot and cold shut-down conditions. The main outcome of our evaluation confirms the general behaviour of the reference calculations except in one important point referring to the cold shut down. In this condition, the failure of one single rod of bank number 13 of the shut down system, leaves the core in a supercritical state at the beginning of the cycle and this anomaly persists during almost a third of the overall cycle. A new design of the core is proposed with minor modifications of the reference one, without introducing new types of fuel elements and keeping the same fuel management scheme. This new core fulfills all the design requirements. (author)
Calculation of equilibria at elevated temperatures using the MINTEQ geochemical code
Energy Technology Data Exchange (ETDEWEB)
Smith, R.W.
1988-12-01
Coefficients and equations for calculating mineral hydrolysis constants, solubility products and formation constants for 60 minerals and 57 aqueous species in the 13 component thermodynamic system K/sub 2/O-Na/sub 2/O-CaO-MgO-FeO-Al/sub 2/O/sub 3/-SiO/sub 2/-CO/sub 2/-H/sub 2/O-HF-HCl-H/sub 2/S-H/sub 2/SO/sub 4/ are presented in a format suitable for inclusion in the MINTEQ computer code. The temperature functions presented for minerals are based on the MINTEQ data base at 25/degree/C and the integration of analytical heat capacity power functions. This approach ensures that the temperature functions join smoothly with the low-temperature data base. A new subroutine, DEBYE, was added to MINTEQ that is used to calculate the theoretical Debye-Hueckel parameters A and B as a function of temperature. In addition, this subroutine also calculates a universal value of the extended Debye-Hueckel parameter, b/sub i/, as a function of temperature. The coefficients and equations provide the capability to use MINTEQ to more accurately calculate water/rock equilibrium for temperatures of up to 250/degree/C, and in dilute, low-sulfate, near neutral groundwaters to 300/degree/C. 52 refs., 1 fig., 6 tabs.
Boltzmann-Fokker-Planck calculations using standard discrete-ordinates codes
International Nuclear Information System (INIS)
The Boltzmann-Fokker-Planck (BFP) equation can be used to describe both neutral and charged-particle transport. Over the past several years, the author and several collaborators have developed methods for representing Fokker-Planck operators with standard multigroup-Legendre cross-section data. When these data are input to a standard S/sub n/ code such as ONETRAN, the code actually solves the Boltzmann-Fokker-Planck equation rather than the Boltzmann equation. This is achieved wihout any modification to the S/sub n/ codes. Because BFP calculations can be more demanding from a numerical viewpoint than standard neutronics calculations, we have found it useful to implement new quadrature methods ad convergence acceleration methods in the standard discrete-ordinates code, ONETRAN. We discuss our BFP cross-section representation techniques, our improved quadrature and acceleration techniques, and present results from BFP coupled electron-photon transport calculations performed with ONETRAN. 19 refs., 7 figs
Code Betal to calculation Alpha/Beta activities in environmental samples
International Nuclear Information System (INIS)
A codes, BETAL, was developed, written in FORTRAN IV, to automatize calculations and presentations of the result of the total alpha-beta activities measurements in environmental samples. This code performs the necessary calculations for transformation the activities measured in total counts, to pCi/1., bearing in mind the efficiency of the detector used and the other necessary parameters. Further more, it appraise the standard deviation of the result, and calculus the Lower limit of detection for each measurement. This code is written in iterative way by screen-operator dialogue, and asking the necessary data to perform the calculation of the activity in each case by a screen label. The code could be executed through any screen and keyboard terminal, (whose computer accepts Fortran IV) with a printer connected to the said computer. (Author) 5 refs
Principles of the reactor code system RHEIN
International Nuclear Information System (INIS)
A description is given of the principles of the reactor code system RHEIN which is applied in connection with a BESM6-type computer. In transfering data between the components of the system external storage is used. The programme passage is controlled by the input data. (author)
Modular ORIGEN-S for multi-physics code systems
International Nuclear Information System (INIS)
The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including
Burn up calculations for ETRR 1 and ETRR 2 reactors with wims and origen codes
International Nuclear Information System (INIS)
For ETRR -1 and ETRR - 2 research reactor, the 235 U depletion is determined with wims and origen codes the two calculated results show good agreement with each other. The buildup of different fission products (important from both the safety and protection point of view) is also calculated. The radioactivity and decay heat of the spent fuel is determined up to 30 years
Burnup calculation capability in the PSG2 / Serpent Monte Carlo reactor physics code
International Nuclear Information System (INIS)
The PSG continuous-energy Monte Carlo reactor physics code has been developed at VTT Technical Research Centre of Finland since 2004. The code is mainly intended for group constant generation for coupled reactor simulator calculations and other tasks traditionally handled using deterministic lattices physics codes. The name was recently changed from acronym PSG to 'Serpent', and the capabilities have been extended by implementing built-in burnup calculation routines that enable the code to be used for fuel cycle studies and the modelling of irradiated fuels. This paper presents the methodology used for burnup calculation. Serpent has two fundamentally different options for solving the Bateman depletion equations: 1) the Transmutation Trajectory Analysis method (TTA), based on the analytical solution of linearized depletion chains and 2) the Chebyshev Rational Approximation Method (CRAM), an advanced matrix exponential solution developed at VTT. The first validation results are compared to deterministic CASMO-4E calculations. It is also shown that the overall running time in Monte Carlo burnup calculation can be significantly reduced using specialized calculation techniques, and that the continuous-energy Monte Carlo method is becoming a viable alternative to deterministic assembly burnup codes. (authors)
Calculation Of Extraction Optics For Ion System With Plazma Emitter
Frolov, B A
2004-01-01
The 2-D code for simulating of ion optics system of positive ion extraction from a plasma source is described. Example calculation of 100 kV optics for the extraction ion IHEP gun is presented. The trajectories of particles and emittance plots are resulted. The aberrations influ-ence strongly on ion optics for considered geometry.
Manual of Nucost 1.0 - code for calculation of nuclear power generation costs
International Nuclear Information System (INIS)
Nucost is a computer code developed at CDTN to perform cost calculation of electric power generated in PWR nuclear power plants, based on present worth cost method. The Nucost version 1.0 performs calculations of nuclear fuel cost cycle by cycle during the time life of the power plant. That calculation is performed with enough details permitting optimization and minimization. The code is also a tool to aid reload projects and economic operation of PWR reactors. This manual presents a description of Nucost version 1.0, instruction to enter data preparation and description of the Nucost output. (M.I.)
Using the ORIGEN-2 computer code for near core activation calculations
International Nuclear Information System (INIS)
The ORIGEN2 computer code is a useful tool for calculating radionuclide inventories resulting from irradiation of materials in a reactor. It is widely used to calculate activation products in irradiated metals that form the structural portion of fuel assemblies. The code is straightforward for materials within the active fuel region of a reactor core, which are subject to core average conditions. For materials outside the active core, ORIGEN2 cannot be used directly. However, ORIGEN2 can be used with the appropriate methodology to calculate the activation of materials in near core locations. This paper presents the background and a methodology for estimating radionuclide inventories in activated metals in near core locations
Validation of the ROVER-F code for ROP trip probability calculations
International Nuclear Information System (INIS)
An important task in the operation of CANDU reactors is the prevention of fuel damage as a result of fuel dryout that can occur when the fuel sheath temperature exceeds the temperature at which the coolant can efficiently remove heat. The power at which fuel dryout is expected to occur is called the critical channel power and is a function of flux shape and the fuel channel thermalhydraulics. In CANDU reactors, protection against overpowers large enough to cause dryout is provided by two regional overpower protection (ROP) systems of in-core flux detectors, arrayed through the core, each organized into three safety (or logic) channels. Each of the two independent ROP systems is associated with one of the two independent shutdown systems (SDS-1 and SDS-2). The detectors in one ROP system (associated with SDS-1) are placed in vertical penetrations, whereas the other system (associated with SDS-2) uses detectors in horizontal penetrations in the core. Each ROP system is capable of initiating the shutdown of the reactor by actuating the corresponding shutdown system. Each ROP system must be so designed that in each safety channel at l east one detector will reach its setpoint before there is damaging overpower in any fuel channel. The trip of a single detector in a safety channel will trip that channel, and the trip of two of the three safety channels in an ROP system will trip that ROP system. ROVER-F is a FORTRAN program which calculates the trip probability and the setpoint adjustment required to attain the target trip probability, for a given set of flux shapes. This calculation is performed with the assumption that the most effective safety channel is unavailable and therefore the remaining two safety channels must both trip. The calculation of trip probability itself is non-iterative, but once the trip probability of the specified system has been calculated, a convergence iteration using a binomial search is used to determine the adjustment to the trip setpoints
Energy Technology Data Exchange (ETDEWEB)
Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M
2006-07-01
The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.
A computer code for calculating a γ-external dose from a randomly distributed radioactive cloud
International Nuclear Information System (INIS)
A computer code ( CIDE ) has been developed to calculate a γ-external dose from a randomly distributed radioactive cloud. Atmospheric dispersion of radioactive materials accidentally released from a nuclear reactor needs to be estimated considering time-dependent meteorological data and terrain heights. Particle-in-Cell model is useful for that purpose, but it is not easy to calculate the dose from the randomly distributed concentration by numerical integration. In this study the mean concentration in a cell evaluated by PIC model was assumed to be uniformly distributed over that cell, which was integrated as a constant concentration by a point kernel method. The dose was obtained by summing the attributable cell doses. When the concentration of plume had a Gaussian distribution, the results of CIDE code well agreed with those of GAMPLE, which was the code for calculating the dose from the Gaussian distribution. The choice of cell sizes affecting the accuracy of the calculated results was discussed. (author)
International Nuclear Information System (INIS)
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
CORRECTED CALCULATION OF HORIZONTAL GATING SYSTEMS
Directory of Open Access Journals (Sweden)
I. A. Zayatz
2015-05-01
Full Text Available In the course of fulfillment of work the specified calculations of horizontal gating systems for various parts produced in dispensable molds were carried out. The results of work showed that the weight removal value in gating systems fluctuates in big intervals and the specified calculation of horizontal gating systems enables to calculate precisely their weight that allows to calculate quantity of metal in metal charge.
New beam-tracking simulation code using bulk-to-point calculation technique for space charge fields
Mizuno, A.
2016-02-01
A new two-dimensional beam-tracking simulation code for electron injectors using a bulk-to-point calculation technique for space charge fields was developed. The calculated space charge fields are produced not by a point charge but by a hollow cylinder that has a volume. Each tracked electron is a point charge. This bulk-to-point calculation technique for space charge fields is based on that used in the multiple beam envelope equations, which were developed by the author. The multiple beam envelope equations are a set of differential equations for investigating the beam dynamics of electron injectors and can be used to calculate bunched beam dynamics with high accuracy. However, there is one limitation. The bunched beam is assumed to be an ensemble of several segmentation pieces in both the transverse and longitudinal directions. In this bunch model, each longitudinal segmentation slice in a bunch must not warp; consequently, the accuracy of the calculated emittance is reduced in the case of a highly charged beam for calculations of a typical rf gun injector system. This limitation is related to the calculation model of longitudinal space charge fields. In the newly developed beam-tracking simulation code, the space charge field calculation scheme is upgraded and the limitation has been overcome. Therefore, the applicable range is extended while maintaining the high accuracy of emittance calculations. Simultaneously, the calculation time is markedly shortened because the emittance dependence on the segmentation number is extremely weak. In this paper, several examples of beam dynamics that cannot be calculated accurately using the multiple beam envelope equations are demonstrated using the new beam-tracking simulation code. The accuracy of the calculated emittance is also discussed.
Kinetic parameters evaluation of PWRs using static cell and core calculation codes
International Nuclear Information System (INIS)
Highlights: ► In this study, we have calculated effective delayed neutron fraction and prompt neutron lifetime in PWRs. ► New software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. ► This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. ► The ratios ((βeff)i)/((βeff)core) , which are the important input data for the reactivity accident analysis, are also calculated. - Abstract: In this paper, evaluation of the kinetic parameters (effective delayed neutron fraction and prompt neutron lifetime) in PWRs, using static cell and core calculation codes, is reported. A new software has been developed to link the WIMS, BORGES and CITATION codes in Visual C computer programming language. Using the WIMS cell calculation code, multigroup microscopic cross-sections and number densities of different materials can be generated in a binary file. By the use of BORGES code, these binary-form cross-sections and number densities are converted to a format readable by the CITATION core calculation code, by which the kinetic parameters can be finally obtained. This software is used for calculation of the kinetic parameters in a typical VVER-1000 and NOK Beznau reactor. The ratios ((βeff)i)/((βeff)core) , which are the important input data for the reactivity accident analysis, are also calculated. Benchmarking of the results against the final safety analysis report (FSAR) of the aforementioned reactors shows very good agreements with these published documents.
Improved decoding for a concatenated coding system
DEFF Research Database (Denmark)
Paaske, Erik
1990-01-01
The concatenated coding system recommended by CCSDS (Consultative Committee for Space Data Systems) uses an outer (255,233) Reed-Solomon (RS) code based on 8-b symbols, followed by the block interleaver and an inner rate 1/2 convolutional code with memory 6. Viterbi decoding is assumed. Two new...... decoding procedures based on repeated decoding trials and exchange of information between the two decoders and the deinterleaver are proposed. In the first one, where the improvement is 0.3-0.4 dB, only the RS decoder performs repeated trials. In the second one, where the improvement is 0.5-0.6 dB, both...... decoders perform repeated decoding trials and decoding information is exchanged between them...
CRONOS: A modular computational system for neutronic core calculations
International Nuclear Information System (INIS)
The CRONOS code has been designed to provide all the computational means needed for Pressurized Water Reactor calculations, including design, fuel management, follow up and accidents. CRONOS allows steady state, kinetic and transient multigroup calculations of power distribution taking into account the thermal-hydraulic feedback effects. All this can be done without any limitation on any parameter (energy groups, meshes...). The code solves either the diffusion equation or the even parity transport equation with isotropic scattering and sources. Different geometries are available such as 1, 2 or 3 dimensions cartesian geometries, 2 or 3D hexagonal geometries and cylindrical geometries. The numerical method is based on the finite difference or finite element methods. CRONOS 2 has been written with the constant will of optimizing its portability. Presently, it is running on very different computers such as IBM 3090, CRAY 1, CRAY 2, SUN 4, MIPS RS2030 or IBM RS6000. A special data structure is used in order to improve vectorization. CRONOS is based on a modular structure that allows a great flexibility of use. It is implemented in the SAPHYR system which includes assembly calculation code (APOLLO), and thermal-hydraulic core calculation code (FLICA IV). A special object oriented language, named GIBIANE, and a common tool library have been developed to chain the various computation modules of those codes. (author). 11 refs, 1 fig., 5 tabs
Reliability calculations for complex systems
Lenz, Malte; Rhodin, Johan
2011-01-01
Functionality for efficient computation of properties of system lifetimes was developed, based on the Mathematica framework. The model of these systems consists of a system structure and the components independent lifetime distributions. The components are assumed to be non-repairable. In this work a very general implementation was created, allowing a large number of lifetime distributions from Mathematica for all the component distributions. All system structures with a monotone increasing s...
International Nuclear Information System (INIS)
A numerical analysis of some neutronic parameters calculated by LEOPARD computer code compared with the literature data are presented. A computer code (LEOCIT) that is a modified version of LEOPARD, was developed, with subroutines that prepare cross sections libraries for 1,2 or 4 energy groups, writing them on tape or on disk, in special format aiming to be diretly used by citation computer codes. Finally, a simulation of the first cycle of Angra I burnup, is done, by CITATION, modelling 1/4 of the core in XY geometry, calculation, the soluble boron curve and the pin to pin power distribution, for two energy group. The more relevant results are compared with those supplied by Westinghouse, CNEN and FURNAS, and some recommendations aiming to perfect the developed system, are done. (E.G)
Nexus: A modular workflow management system for quantum simulation codes
Krogel, Jaron T.
2016-01-01
The management of simulation workflows represents a significant task for the individual computational researcher. Automation of the required tasks involved in simulation work can decrease the overall time to solution and reduce sources of human error. A new simulation workflow management system, Nexus, is presented to address these issues. Nexus is capable of automated job management on workstations and resources at several major supercomputing centers. Its modular design allows many quantum simulation codes to be supported within the same framework. Current support includes quantum Monte Carlo calculations with QMCPACK, density functional theory calculations with Quantum Espresso or VASP, and quantum chemical calculations with GAMESS. Users can compose workflows through a transparent, text-based interface, resembling the input file of a typical simulation code. A usage example is provided to illustrate the process.
A Photovoltaic System Payback Calculator
Energy Technology Data Exchange (ETDEWEB)
Riley, Daniel M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Fleming, Jeffrey E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gallegos, Gerald R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2016-06-01
The Roof Asset Management Program (RAMP) is a DOE NNSA initiative to manage roof repairs and replacement at NNSA facilities. In some cases, installation of a photovoltaic system on new roofs may be possible and desired for financial reasons and to meet federal renewable energy goals. One method to quantify the financial benefits of PV systems is the payback period, or the length of time required for a PV system to generate energy value equivalent to the system's cost. Sandia Laboratories created a simple spreadsheet-based solar energy valuation tool for use by RAMP personnel to quickly evaluate the estimated payback period of prospective or installed photovoltaic systems.
Development of BERMUDA: a radiation transport code system, 1
International Nuclear Information System (INIS)
A radiation transport code system BERMUDA has been developed for one-, two- and three-dimensional geometries. The time-independent transport equation is numerically solved using a direct integration method in a multigroup model, to obtain spatial, angular and energy distributions of neutron, gamma rays or adjoint neutron flux. As to group constants, a library with an any structure of energy groups is capable to be produced from a data base JSSTDL, or by a processing code PROF-GROUCH-G/B, selecting objective nuclear data through a retrieval system EDFSRS. Validity of the present code system has been tested by analyzing the shielding benchmark experiments. The test has shown that accurate results are obtainable with this system especially in deep penetration calculation. Described are the devised calculation method and the results of validity tests. Input data specification, job control languages and output data are also described as a user's manual for the following four neutron transport codes: BERMUDA-1DN : sphere, slab(S20), BERMUDA-2DN : cylinder (S8), BERMUDA-2DN-S16 : cylinder (S16), and BERMUDA-3DN : rectangular parallelpiped (S8). (J.P.N.)
Methods, algorithms and computer codes for calculation of electron-impact excitation parameters
Bogdanovich, P; Stonys, D
2015-01-01
We describe the computer codes, developed at Vilnius University, for the calculation of electron-impact excitation cross sections, collision strengths, and excitation rates in the plane-wave Born approximation. These codes utilize the multireference atomic wavefunctions which are also adopted to calculate radiative transition parameters of complex many-electron ions. This leads to consistent data sets suitable in plasma modelling codes. Two versions of electron scattering codes are considered in the present work, both of them employing configuration interaction method for inclusion of correlation effects and Breit-Pauli approximation to account for relativistic effects. These versions differ only by one-electron radial orbitals, where the first one employs the non-relativistic numerical radial orbitals, while another version uses the quasirelativistic radial orbitals. The accuracy of produced results is assessed by comparing radiative transition and electron-impact excitation data for neutral hydrogen, helium...
International Nuclear Information System (INIS)
A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)
Energy Technology Data Exchange (ETDEWEB)
Vitruk, S.G.; Korsun, A.S. [Moscow Engineering Physics Institute (Russian Federation); Ushakov, P.A. [Institute of Physics and Power Engineering, Obninsk (R)] [and others
1995-09-01
The multilevel mathematical model of neutron thermal hydrodynamic processes in a passive safety core without assemblies duct walls and appropriate computer code SKETCH, consisted of thermal hydrodynamic module THEHYCO-3DT and neutron one, are described. A new effective discretization technique for energy, momentum and mass conservation equations is applied in hexagonal - z geometry. The model adequacy and applicability are presented. The results of the calculations show that the model and the computer code could be used in conceptual design of advanced reactors.
A FORTRAN computer code for calculating flows in multiple-blade-element cascades
Mcfarland, E. R.
1985-01-01
A solution technique has been developed for solving the multiple-blade-element, surface-of-revolution, blade-to-blade flow problem in turbomachinery. The calculation solves approximate flow equations which include the effects of compressibility, radius change, blade-row rotation, and variable stream sheet thickness. An integral equation solution (i.e., panel method) is used to solve the equations. A description of the computer code and computer code input is given in this report.
Off-site dose calculation computer code based on ICRP-60(II) - liquid radioactive effluents -
International Nuclear Information System (INIS)
The development of computer code for calculating off-site doses(K-DOSE60) was based on ICRP-60 and the dose calculationi equations of Reg. Guide 1.109. In this paper, the methodology to compute dose for liquid effluents was described. To examine reliability of the K-DOSE60 code the results obtained from K-DOSE60 were compared with analytic solutions. For liquid effluents. The results by K-DOSE60 are in agreement with analytic solution
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
International Nuclear Information System (INIS)
POPCYCLE, a computer code designed to calculate levelized life-cycle power costs for nuclear and fossil electrical generating plants is described. Included are (1) derivations of the equations and a discussion of the methodology used by POPCYCLE, (2) a description of the input required by the code, (3) a listing of the input for a sample case, and (4) the output for a sample case
International Nuclear Information System (INIS)
The CITHAN computer code was developed at IPEN (Instituto de Pesquisas Energeticas e Nucleares) to link the HAMMER computer code with a fuel depletion routine and to provide neutron cross sections to be read with the appropriate format of the CITATION code. The problem arised due to the efforts to addapt the new version denomined HAMMER-TECHION with the routine refered. The HAMMER-TECHION computer code was elaborated by Haifa Institute, Israel within a project with EPRI. This version is at CNEN to be used in multigroup constant generation for neutron diffusion calculation in the scope of the new methodology to be adopted by CNEN. The theoretical formulation of CITHAM computer code, tests and modificatins are described. (Author)
LOLA-SYSTEM, JEN-UPM PWR Fuel Management System Burnup Code System
International Nuclear Information System (INIS)
1 - Description of program or function: The LOLA-SYSTEM is a part of the JEN-UPM code package for PWR fuel management, scope or design calculations. It is a code package for core burnup calculations using nodal theory based on a FLARE type code. The LOLA-SYSTEM includes four modules: the first one (MELON-3) generates the constants of the K-inf and M2 correlations to be input into SIMULA-3. It needs the K-inf and M2 fuel assembly values at different conditions of moderator temperature, Boron concentration, burnup, etc., which are provided by MARIA fuel assembly calculations. The main module (SIMULA-3) is the core burnup calculation code in three dimensions and one group of energy. It normally uses a geometrical representation of one node per fuel assembly or per quarter of fuel assembly. It has included a thermal hydraulic feedback on flow and voids and criticality searches on boron concentration and control rods insertion. The CONCON code makes the calculation of the albedo, transport factors, K-inf and M2 correction factors to be input into SIMULA-3. The calculation is made in the XY transversal plane. The CONAXI code is similar to CONCON, but in the axial direction. 2 - Method of solution: MELON-3 makes a mean squares fit of K-inf and M2 values at different conditions in order to determine the constants of the feedback correlations. SIMULA-3 uses a modified one-group nodal theory, with a new transport kernel that provides the same node interface leakages as a fine mesh diffusion calculation. CONCON and CONAXI determine the transport and correction factors, as well as the albedo, to be input into SIMULA-3. They are determined by a method of leakages equivalent to the detailed diffusion calculation of CARMEN or VENTURE; these factors also include the heterogeneity effects inside the node. 3 - Restrictions on the complexity of the problem: Number of axial nodes less than or equal 34. Number of material types less than or equal 30. Number of fuel assembly types less
ANIGAM: a computer code for the automatic calculation of nuclear group data
International Nuclear Information System (INIS)
The computer code ANIGAM consists mainly of the well-known programmes GAM-I and ANISN as well as of a subroutine which reads the THERMOS cross section library and prepares it for ANISN. ANIGAM has been written for the automatic calculation of microscopic and macroscopic cross sections of light water reactor fuel assemblies. In a single computer run both were calculated, the cross sections representative for fuel assemblies in reactor core calculations and the cross sections of each cell type of a fuel assembly. The calculated data were delivered to EXTERMINATOR and CITATION for following diffusion or burn up calculations by an auxiliary programme. This report contains a detailed description of the computer codes and methods used in ANIGAM, a description of the subroutines, of the OVERLAY structure and an input and output description. (oririg.)
International Nuclear Information System (INIS)
The objective of the study is to compare the thermal neutron fluxes at specimen positions of neutron radiography facility calculated by MCNP4C code with the measurement. A model for calculation was developed using details of the reactor core configuration no. 14 and neutron radiography facility installed at the existing research reactor, TRR-1/M1 reactor. Assuming all fresh fuel elements and all control rod out condition, the thermal neutron fluxes at various specimen positions were calculated using MCNP4C code. The calculation are verified by the measurement using foil activation method. Generally, the calculated neutron fluxes are overestimated by 16-20% which is reasonably good agreement and acceptable for the complex system. The discrepancy is expected to the assumption of using fresh fuel elements, all control rod out condition, and also lacks of information in develop a more accurate model for calculation. This study shows the possibility of using the MCNP4C code to verify the thermal neutron fluxes at specimen position and shielding design of the new neutron radiography facility at the new Thai research reactor
An Efficient Group Key Management Using Code for Key Calculation for Simultaneous Join/Leave: CKCS
Directory of Open Access Journals (Sweden)
Melisa Hajyvahabzadeh
2012-08-01
Full Text Available This paper presents an efficient group key management protocol, CKCS (Code for Key Calculation in Simultaneous join/leave for simultaneous join/leave in secure multicast. This protocol is based on logical key hierarchy. In this protocol, when new members join the group simultaneously, server sends only thegroup key for those new members. Then, current members and new members calculate the necessary keys by node codes and one-way hash function. A node code is a random number which is assigned to each key to help users calculate the necessary keys. Again, at leave, the server just sends the new group key to remaining members. The results show that CKCS reduces computational and communication overhead, and also message size in simultaneous join/leave.
Thermodynamic Calculations for Systems Biocatalysis
DEFF Research Database (Denmark)
Abu, Rohana; Gundersen, Maria T.; Woodley, John M.
2015-01-01
‘Systems Biocatalysis’ is a term describing multi-enzyme processes in vitro for the synthesis of chemical products. Unlike in-vivo systems, such an artificial metabolism can be controlled in a highly efficient way in order to achieve a sufficiently favourable conversion for a given target product...
The FLUFF code for calculating finned surface heat transfer -description and user's guide
International Nuclear Information System (INIS)
FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)
Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide
International Nuclear Information System (INIS)
The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems
CRX: a transport theory code for cell and assembly calculations based on characteristic method
International Nuclear Information System (INIS)
A transport theory code CRX based on characteristic method with a general geometric tracking routine for rectangular and hexagonal geometrical problems is developed and tested for heterogeneous cell and assembly calculations. Since the characteristic method treats explicitly (analytically) the streaming portion of the transport equation, CRX treats strong absorbers well and has no practical limitations placed on the geometry of the problem. To test the code, it was applied to three benchmark problems which consist of complex meshes and compared with other codes. (author)
Bar-code automated waste tracking system
International Nuclear Information System (INIS)
The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ''stop-and-go'' operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste
Bar-code automated waste tracking system
Energy Technology Data Exchange (ETDEWEB)
Hull, T.E.
1994-10-01
The Bar-Code Automated Waste Tracking System was designed to be a site-Specific program with a general purpose application for transportability to other facilities. The system is user-friendly, totally automated, and incorporates the use of a drive-up window that is close to the areas dealing in container preparation, delivery, pickup, and disposal. The system features ``stop-and-go`` operation rather than a long, tedious, error-prone manual entry. The system is designed for automation but allows operators to concentrate on proper handling of waste while maintaining manual entry of data as a backup. A large wall plaque filled with bar-code labels is used to input specific details about any movement of waste.
VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment
International Nuclear Information System (INIS)
Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)
Calculation of the Fast Flux Test Facility fuel pin tests with the WIMS-E and MCNP codes
International Nuclear Information System (INIS)
The Fuel Assembly Area (FAA) at the Fast Flux Test Facility site on the Hanford Site at Richland, Washington currently is being prepared to fabricate mixed oxide fuel (U, Pu) for the FFTF. Calculational tools are required to perform criticality safety analyses for various process locations and to establish safe limits for fissile material handling at the FAA. These codes require validation against experimental data appropriate for the compositions that will be handled. Critical array experiments performed by Bierman provide such data for mixed oxide fuel in the range Pu/(U+Pu) = 22 wt %, and with Pu-240 contents equal to 12 wt %. Both the Monte Carlo Neutron Photon (MCNP) and the Winfrith Improved Multigroup Scheme (WIMS-E) computer codes were used to calculate the neutron multiplication factor for explicit models of the various critical arrays. The W-CACTUS modules within the WIMS-E code system was used to calculate k∞ for the explicit array configuration, as well as few-group cross sections that were then used in a three-dimensional diffusion theory code for the calculation of keff for the finite array. 10 refs., 15 figs., 7 tabs
Channel coding and modulation based on chaotic systems
Kozic, Slobodan; Hasler, Martin
2007-01-01
In this thesis, a new class of codes on graphs based on chaotic dynamical systems are proposed. In particular, trellis coded modulation and iteratively decodable codes on graphs are studied. The codes are designed by controlling symbolic dynamics of chaotic systems and using linear convolutional codes. The relation between symbolic dynamics of chaotic systems and trellis aspects to minimum distance properties of coded modulations is explained. Our arguments are supported by computer simulatio...
Network Coding for Distributed Storage Systems
Dimakis, Alexandros G; Wu, Yunnan; Wainwright, Martin J; Ramchandran, Kannan
2008-01-01
Distributed storage systems provide reliable access to data through redundancy spread over individually unreliable nodes. Application scenarios include data centers, peer-to-peer storage systems, and storage in wireless networks. Storing data using an erasure code, in fragments spread across nodes, requires less redundancy than simple replication for the same level of reliability. However, since fragments must be periodically replaced as nodes fail, a key question is how to generate encoded fragments in a distributed way while transferring as little data as possible across the network. For an erasure coded system, a common practice to repair from a node failure is for a new node to download subsets of data stored at a number of surviving nodes, reconstruct a lost coded block using the downloaded data, and store it at the new node. We show that this procedure is sub-optimal. We introduce the notion of regenerating codes, which allow a new node to download \\emph{functions} of the stored data from the surviving ...
Network Coded Multicast over Multibeam Satellite Systems
R. Alegre-Godoy; Vazquez-Castro, M. A.
2015-01-01
We propose a multicast scheme for multibeam satellite systems exploiting both the multiuser and spatial diversity inherent in this type of systems while taking into account realistic physical distributions of User Terminals (UTs) over the coverage. Our proposed scheme makes use of the well-known Adaptive Coding and Modulation (ACM) feature in Digital Video Broadcasting over Satellite, 2nd Generation (DVB-S2) and Extension (DVB-S2X) standards but also incorporates a set of innovative features....
Abnormality transient analysis of Monju using a plant system code
International Nuclear Information System (INIS)
The objectives of the present study are to analyze plant transients caused by small abnormalities and to find plant parameters by which operators can recognize these small abnormalities. In order to evaluate the plant transient during an abnormal situation in the water system using the plant system code NETFLOW++, the turbine and feedwater systems should be analyzed with good precision. The code is validated using the measured data at Monju. Several abnormalities in the water system are candidates of the present study, e.g., feedwater control valve degradation, feedwater pump degradation, heat transfer degradation due to fouling on heat transfer tubes of the evaporator, loss-of-feedwater-heating, etc. All major components in the tertiary system are included in the calculation model such as the steam generators, the high-pressure turbine, the deaerator, the feedwater pump, the feedwater heaters, the feedwater control valves, the steam control valve, extraction lines and drainpipes. (author)
Energy Technology Data Exchange (ETDEWEB)
Perfetti, C.; Martin, W. [Univ. of Michigan, Dept. of Nuclear Engineering and Radiological Sciences, 2355 Bonisteel Boulevard, Ann Arbor, MI 48109-2104 (United States); Rearden, B.; Williams, M. [Oak Ridge National Laboratory, Reactor and Nuclear Systems Div., Bldg. 5700, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
2012-07-01
Three methods for calculating continuous-energy eigenvalue sensitivity coefficients were developed and implemented into the Shift Monte Carlo code within the SCALE code package. The methods were used for two small-scale test problems and were evaluated in terms of speed, accuracy, efficiency, and memory requirements. A promising new method for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was developed and produced accurate sensitivity coefficients with figures of merit that were several orders of magnitude larger than those from existing methods. (authors)
RAP-4A Computer code for thermohydraulic calculation of liquid metal cooled fuel clusters
International Nuclear Information System (INIS)
RAP-4A is a programme for calculating the fuel clusters thermal-hydraulic parameters in a fast liquid metal-cooled reactor. The code gives the possibility to calculate steady state axial distribution temperature, enthalpy, pressure drop and mass velocity . A monodimensional mathematical model along the cluster allowing the study of the single and two phase flow is used by taking into account the mixing between adjacent subchannels. Physical and mathematical models, general features and an example are presented. RAP-4A code is written in FORTRAN-IV language on IBM 370/135 computer
International Nuclear Information System (INIS)
Four calculational benchmarks have been selected to compare various nuclear data libraries based on both ENDF/B-IV and V, and to compare results from various transport codes. Discrepancies up to 20% in tritium production from 7Li were found and have been attributed mainly to differences in current ENDF/B-IV and V evaluations, while approx.4% is attributed to differences in the group structure of the libraries used. Results from MCNP and VIP Monte Carlo codes are in good agreement, but MORSE calculations show good agreement only for high threshold reactions
Energy Technology Data Exchange (ETDEWEB)
Eslinger, Paul W.; Aaberg, Rosanne L.; Lopresti, Charles A.; Miley, Terri B.; Nichols, William E.; Strenge, Dennis L.
2004-09-14
This document contains detailed user instructions for a suite of utility codes developed for Rev. 1 of the Systems Assessment Capability. The suite of computer codes for Rev. 1 of Systems Assessment Capability performs many functions.
ITP.FOR: A code to calculate thermal transients in High Level Waste Tanks
International Nuclear Information System (INIS)
A variety of processing operations for high level radioactive waste occur in the High Level Waste Tanks in the H-Area of the Savannah River Site. Thermal design constraints exist on these processes, principally to limit the amount of corrosion inhibitor which must be added to protect the tank and cooling coil materials. The required amount of corrosion inhibitor, which must subsequently be removed prior to trapping the waste in borosilicate glass, increases exponentially with temperature over a fairly narrow range (some tens of degrees Celsius). For this reason, there is a need to model the thermal-hydraulic processes occurring in the waste tanks. A FORTRAN computer code, called ITP.FOR, was written to provide a simple but reasonably accurate analysis tool for plant operation design. The code was specifically written to model Tank 48, in which the In-Tank Precipitation (ITP) process of precipitating radioactive cesium will be initiated. Although the ITP.FOR code was written as personal-use software for scoping design calculations for Tank 48, the current intent is to extend the code's applicability to other H-Area waste tanks, and to certify the code in accordance with the NRTSC Quality Assurance requirements for critical-use software (1Q-34, 1991). Since the code's capabilities have generated some interest to date, the present report is presented as interim documentation of the code's mathematical models. This documentation will eventually be supplanted by the formal documentation of the expanded and benchmarked code
A Students Attendance System Using QR Code
Directory of Open Access Journals (Sweden)
Fadi Masalha
2014-01-01
Full Text Available Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the code in order to confirm their attendance. The paper explains the high level implementation details of the proposed system. It also discusses how the system verifies student identity to eliminate false registrations.
International Nuclear Information System (INIS)
After the stopping of nuclear reactors, the decay heat power and the residual activity of the irradiated fuel don't decrease immediately to zero. The knowledge of the residual activity and the decay heat power of irradiated fuel is needed in the different areas of the nuclear reactor industry. The core physicist, the nuclear plant operator, the spent fuel conveyor and the safety authorities, all of them need data on residual activity and decay heat power. These safety parameters must be calculated, using the safety calculation codes as FISPIN, ORIGEN and APOLLO-PEPIN system which require costly and powerful computers and long running time, before transport. For this reason, the PEPABAC calculation code (module of FAKIR code) is developed, it takes into account the contribution of Fission Products, Actinides, U239 + Np 239 and delayed neutrons. PEPABAC is easy to use and provides good results of UOX and MOX fuel (from PWR) calculations. For the calculations, it uses simple formula and tabulated data libraries (which explain its short running time). The input data is reduced, it contains only the parameters from the power histogram and some informations relative to the fuel assemblies as the type of reactor and fuel assemblies, etc... It is possible, in the future, to extend the application fields to fuels from BWR. (author)
Performance Analysis of Turbo Coded OFDM System
Directory of Open Access Journals (Sweden)
Jyoti Chand
2014-05-01
Full Text Available Orthogonal Frequency Division Multiplexing (OFDM has become a popular modulation method in high speed wireless communication system. By partitioning a wideband fading channel into a flat narrowband channels, OFDM is able to mitigate the detrimental effects of multipath fading using a simple one- tap equalizer. There is a growing need to quickly transmit information wirelessly and accurately. Engineers have already combine techniques such as OFDM suitable for high data rate transmission with forward error correction (FEC methods over wireless channels. In this thesis, we enhance the system throughput of a working OFDM system by adding turbo codes. The smart use of coding and power allocation in OFDM will be useful to the desired performance at higher data rates. Simulation is to be done over Rayleigh and additive white Gaussian noise (AWGN channels. Here we also compare the two different generator polynomials. This project increases the system throughput at the same time maintaining system performance. The performance is improved by convolution coding [1].
Some benchmark calculations for VVER-1000 assemblies by WIMS-7B code
International Nuclear Information System (INIS)
Our aim in this report is to compare of calculation results, obtained with the use of different libraries, which are in the variant of the WIMS7B code. We had the three libraries: the 1986 library is based on the UKNDL files, the two 1996 libraries are based on the JEF-2.2 files, the one having the 69 group approximation, the other having the 172 group approximation. We wanted also to have some acquaintance with the new option of WIMS-7B - CACTUS. The variant of WIMS-7B was placed at our disposal by the code authors for a temporal use for 9 months. It was natural to make at comparisons with analogous values of TVS-M, MCU, Apollo-2, Casmo-4, Conkemo, MCNP, HELIOS codes, where the other different libraries were used. In accordance with our aims the calculations of unprofiled and profiled assemblies of the VVER-1000 reactor have been carried out by the option CACTUS. This option provides calculations by the characteristics method. The calculation results have been compared with the K∞ values obtained by other codes in work. The conclusion from this analysis is such: the methodical parts of errors of these codes have nearly the same values. Spacing for Keff values can be explained of the library microsections differences mainly. Nevertheless, the more detailed analysis of the results obtained is required. In conclusion the calculation of a depletion of VVER-1000 cell has been carried out. The comparison of the dependency of the multiply factor from the depletion obtained by WIMS-7B with different libraries and by the TVS-M, MCU, HELIOS and WIMS-ABBN codes in work has been performed. (orig.)
International Nuclear Information System (INIS)
The current basis for conversion coefficients for calibrating individual photon dosimeters in terms of dose equivalents is found in the series of papers by Grosswent. In his calculation the collision kerma inside the phantom is determined by calculation of the energy fluence at the point of interest and the use of the mass energy absorption coefficient. This approximates the local absorbed dose. Other Monte Carlo methods can be sued to provide calculations of the conversion coefficients. Rogers has calculated fluence-to-dose equivalent conversion factors with the Electron-Gamma Shower Version 3, EGS3, Monte Carlo program and produced results similar to Grosswent's calculations. This paper will report on calculations using the Integrated TIGER Series Version 3, ITS3, code to calculate the conversion coefficients in ICRU Tissue and in PMMA. A complete description of the input parameters to the program is given and comparison to previous results is included
BCH codes for large IC random-access memory systems
Lin, S.; Costello, D. J., Jr.
1983-01-01
In this report some shortened BCH codes for possible applications to large IC random-access memory systems are presented. These codes are given by their parity-check matrices. Encoding and decoding of these codes are discussed.
International Nuclear Information System (INIS)
This report describes a development of a wind field calculation code and an atmospheric dispersion and dose calculation code which can be used for real-time prediction in an emergency. Models used in the computer codes are a mass-consistent model for wind field and a particle diffusion model for atmospheric dispersion. In order to attain quick response even when the codes are used in a small-scale computer, high-speed iteration method (MILUCR) and kernel density method are applied to the wind field model and the atmospheric and dose calculation model, respectively. In this report, numerical models, computational codes, related files and calculation examples are shown. (author)
International Nuclear Information System (INIS)
TEMPUL is one dimensional computer code for calculating radial fuel temperature distribution in a fuel immediately after the pulse. Implementation of TEMPUL code was performed to calculate of radial temperature distribution on TRIGA fuel element. The gap between fuel element and cladding is treated to be in contact (without gap), gap is filled with air and gap is filled with helium gas, respectively. Equilateral triangular arrangement coolant channel is assumed. The calculated results on calculation of radial temperature distribution in TRIGA fuel element immediately after the pulse occur relatively high ascending tendency in zirconium rod (radius 0.3175 cm) and fuel element-cladding interface (radius 1.82245 cm) at the first second after pulse with no gap and gap filled with helium gas treatment. Rising of cladding and interface between cladding and coolant average temperature reach up to 500 oC drastically occur in the first second after the pulse. (author)
PEGASUS: a preequilibrium and multi-step evaporation code for neutron cross section calculation
International Nuclear Information System (INIS)
The computer code PEGASUS was developed to calculate neutron-induced reaction cross sections on the basis of the closed form exciton model preequilibrium theory and the multi-step evaporation theory. The cross sections and emitted particle spectra are calculated for the compound elastic scattering, (n,γ), (n,n'), (n,p), (n,α), (n,d), (n,t), (n,3He), (n,2n), (n,n'p), (n,n'α), (n,n'd), (n,n't), (n,2p) and (n,3n) reactions. The double differential cross sections of emitted particles are also calculated. The calculated results are written on a magnetic disk in the ENDF format. Parameter files and/or systematics formulas are provided for level densities, mass excess, radiation widths and inverse cross sections so that the input data to the code are made minimum. (author)
Study of magnetic island using a 3D MHD equilibrium calculation code
International Nuclear Information System (INIS)
Coupling the magnetic diagnostics and a 3D MHD equilibrium calculation code, the magnetic island is studied in the Large Helical Device (LHD) experiment. In an experiment, the collapse in the plasma core was observed in a configuration, which has large magnetic island produced by external perturbation coils. At the collapse, the temperature profile was flattened. This suggests the magnetic island evolved. The magnetic island was observed by the magnetic diagnostics. The magnetic diagnostics also suggests evolving the magnetic island. A 3D MHD equilibrium is calculated by the 3D MHD equilibrium code then signals of the magnetic diagnostics are simulated. Since the comparison of observed and calculated signals is comparable, the magnetic island in calculated equilibrium is similar to one of the experiment. (author)
Presentation of the HTR neutronics code system NEPHTIS
International Nuclear Information System (INIS)
AREVA's joint subsidiary with Siemens, Framatome ANP, has launched the ANTARES Program (AREVA New Technology based on Advanced gas-cooled Reactor for Energy Supply) for the development of an advanced commercial HTR reactor for electricity generation and process heat supply. In this context, specific development and qualification R and D programs were established with CEA to set-up NEPHTIS (Neutronics Process for HTR Innovating System), a new industrial neutronics code system for the computation of V/HTR reactors. After a brief overview of the ANTARES Program, this paper aims to present the characteristics of the HTR neutronics code system NEPHTIS and to show the validation performed against reference (Monte-Carlo) codes and versus experimental results. Nephis is a deterministic calculation scheme dedicated to the computation of prismatic block-type HTR cores. It is based on a usual two-step approach. First, the fuel element is calculated in 2D transport theory in its true heterogeneous representation (including fuel particles double-heterogeneity) using the APOLLO2 spectral code with a very fine energy meshing (172 groups). Then, this calculation provides condensed (8 groups) and homogenized cross sections to the core simulator CRONOS2, which performs 3D calculations in diffusion theory with finite-elements. Validations were performed on several geometry configurations (assembly/core; 2D/3D) for UOX fuel. The two-steps approach used in NEPHTIS is validated by comparisons with an APOLLO2 reference calculation (whole core computed in 2D transport theory) and with Monte-Carlo calculations. Specific complementary validations (e.g. control rod worth), using TRIPOLI and MCNP calculations, are presented as well in this paper. Calculations vs. experiment (C/E) comparisons have been performed on the two main existing block-type HTR reactors: HTTR and Fort Saint-Vrain (FSV). NEPHTIS results were compared with the HTTR benchmark, proposed by IAEA in 1999, on the first
Sub-channel analysis by RELAP5 system code
Energy Technology Data Exchange (ETDEWEB)
Alessandro Petruzzi; Anis Bousbia Salah [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy); Francesco D' Auria [DIMNP, Universit y of Pisa, Via Diotisalvi 2, 56126 Pisa (Italy)
2005-07-01
Full text of publication follows: Recent progress in computer technology has increased the possibilities for code calculations in predicting realistically transient scenarios in nuclear power plants. Several attempts have been engaged in order to enlarge the domain for code applications, and to allow best estimate core simulation including interaction effects between neutronics and thermal-hydraulics. In this context, Relap5/Mod3.3 system thermalhydraulic code was used as a sub-channel code for the simulation of the low-pressure boil off experiment No 5002 of Neptun test facility. The experiment constitutes one of the separate effects test (SET) in the OECD/CSNI matrix for thermalhydraulic code validation related to phase separation and vertical flow 'with or without mixture level'. The drying out of the heated elements is expect to occur at very low coolant flow rates, low pressure (about 1.1 bar) and low power level (24.6 kW). The main aim of the activity discussed in the paper is to develop a 'nodalization technology' for accurately modeling the sub-channel grade void distribution problem and in the same way to assess the degree of success in using the Relap5 system code as a sub-channel code for the analysis of local quantities during transients in nuclear reactors. All thermal-hydraulic parameters, such as the collapsed liquid level, critical heat flux time occurrence and heaters surface temperature have been predicted with reasonable accuracy. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. More accurate results have been obtained considering the surface to surface radiation heat transfer model, as well as more cross flow nodes between the test section rods. The overall analysis confirms the possibility of using the Relap5/Mod3.3 system thermal-hydraulic code as sub-channel code to predict the evolution of relevant local quantities measured during 'relevant' experiments
International Nuclear Information System (INIS)
An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis
Validation of the COBRA code for dry out power calculation in CANDU type advanced fuels
International Nuclear Information System (INIS)
Stern Laboratories perform a full scale CHF testing of the CANFLEX bundle under AECL request. This experiment is modeled with the COBRA IV HW code to verify it's capacity for the dry out power calculation . Good results were obtained: errors below 10 % with respect to all data measured and 1 % for standard operating conditions in CANDU reactors range . This calculations were repeated for the CNEA advanced fuel CARA obtaining the same performance as the CANFLEX fuel. (author)
Some questions of using coding theory and analytical calculation methods on computers
International Nuclear Information System (INIS)
Main results of investigations devoted to the application of theory and practice of correcting codes are presented. These results are used to create very fast units for the selection of events registered in multichannel detectors of nuclear particles. Using this theory and analytical computing calculations, practically new combination devices, for example, parallel encoders, have been developed. Questions concerning the creation of a new algorithm for the calculation of digital functions by computers and problems of devising universal, dynamically reprogrammable logic modules are discussed
Energy Technology Data Exchange (ETDEWEB)
Ash, J E; Marciniak, T J
1977-01-01
The Liquid Metal Fast Breeder Reactor (LMFBR) core structure consists of a matrix of hexagonal subassembly ducts. Evaluation of the safety aspects of the core structure requires that reliable computational procedures be available to predict the deformation response of the subassembly configuration to postulated local energy releases. Finite-element computer codes have been developed to calculate deflections and strains of a hexcan subassembly wrapper subjected to internal and external dynamic pressure loadings over a wide range of material-property conditions. An experimental and analytical program has been undertaken to validate and extend the codes for describing the core structural mechanics under reactor operating conditions, including, in particular, descriptions of possible subassembly-to-subassembly damage propagation. This report describes results of the first phase of the experimental program in which single hexcan sections were internally and externally hydrostatically pressurized out-of-pile at room temperature. The experimental data are compared with calculations from a two-dimensional finite-element structural-dynamics code, STRAW. Some additional comparisons were also made with calculations from a three-dimensional code, SADCAT. The correlations obtained between the computations and the hydrostatic experimental results were sufficiently good to validate the STRAW code and proceed to the next phase of the program involving the dynamic structural response.
EAI-oriented information classification code system in manufacturing enterprises
Institute of Scientific and Technical Information of China (English)
Junbiao WANG; Hu DENG; Jianjun JIANG; Binghong YANG; Bailing WANG
2008-01-01
Although the traditional information classifi-cation coding system in manufacturing enterprises (MEs) emphasizes the construction of code standards, it lacks the management of the code creation, code data transmission and so on. According to the demands of enterprise application integration (EAI) in manufacturing enter-prises, an enterprise application integration oriented information classification code system (EAIO-ICCS) is proposed. EAIO-ICCS expands the connotation of the information classification code system and assures the identity of the codes in manufacturing enterprises with unified management of codes at the view of its lifecycle.
International Nuclear Information System (INIS)
A radioactivity evaluation code system for high temperature gas-cooled reactors during normal operation was developed to study the behavior of fission products (FP) in the plants. The system consists of a code for the calculation of diffusion of FPs in fuel (FIPERX), a code for the deposition of FPs in primary cooling system (PLATO), a code for the transfer and emission of FPs in nuclear power plants (FIPPI-2), and a code for the exposure dose due to emitted FPs (FEDOSE). The FIPERX code can calculate the changes in the course of time FP of the distribution of FP concentration, the distribution of FP flow, the distribution of FP partial pressure, and the emission rate of FP into coolant. The amount of deposition of FPs and their distribution in primary cooling system can be evaluated by the PLATO code. The FIPPI-2 code can be used for the estimation of the amount of FPs in nuclear power plants and the amount of emitted FPs from the plants. The exposure dose of residents around nuclear power plants in case of the operation of the plants is calculated by the FEDOSE code. This code evaluates the dose due to the external exposure in the normal operation and in the accident, and the internal dose by the inhalation of radioactive plume and foods. Further studies of this code system by the comparison with the experimental data are considered. (Kato, T.)
Dose calculation system for remotely supporting radiotherapy
International Nuclear Information System (INIS)
The dose calculation system IMAGINE is being developed keeping in mind remotely supporting external radiation therapy using photon beams. The system is expected to provide an accurate picture of the dose distribution in a patient body, using a Monte Carlo calculation that employs precise models of the patient body and irradiation head. The dose calculation will be performed utilising super-parallel computing at the dose calculation centre, which is equipped with the ITBL computer, and the calculated results will be transferred through a network. The system is intended to support the quality assurance of current, widely carried out radiotherapy and, further, to promote the prevalence of advanced radiotherapy. Prototypes of the modules constituting the system have already been constructed and used to obtain basic data that are necessary in order to decide on the concrete design of the system. The final system will be completed in 2007. (authors)
Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations
International Nuclear Information System (INIS)
Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)
A mean field theory of coded CDMA systems
International Nuclear Information System (INIS)
We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems
A mean field theory of coded CDMA systems
Energy Technology Data Exchange (ETDEWEB)
Yano, Toru [Graduate School of Science and Technology, Keio University, Hiyoshi, Kohoku-ku, Yokohama-shi, Kanagawa 223-8522 (Japan); Tanaka, Toshiyuki [Graduate School of Informatics, Kyoto University, Yoshida Hon-machi, Sakyo-ku, Kyoto-shi, Kyoto 606-8501 (Japan); Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)], E-mail: yano@thx.appi.keio.ac.jp
2008-08-15
We present a mean field theory of code-division multiple-access (CDMA) systems with error-control coding. On the basis of the relation between the free energy and mutual information, we obtain an analytical expression of the maximum spectral efficiency of the coded CDMA system, from which a mean-field description of the coded CDMA system is provided in terms of a bank of scalar Gaussian channels whose variances in general vary at different code symbol positions. Regular low-density parity-check (LDPC)-coded CDMA systems are also discussed as an example of the coded CDMA systems.
Development of Three-dimensional Reactor Analysis Code System for Accelerator-Driven System, ADS3D
International Nuclear Information System (INIS)
To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is able to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS. (author)
KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR (V2.0) Code
International Nuclear Information System (INIS)
Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR (Version 2.0), explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006
WOLF: a computer code package for the calculation of ion beam trajectories
International Nuclear Information System (INIS)
The WOLF code solves POISSON'S equation within a user-defined problem boundary of arbitrary shape. The code is compatible with ANSI FORTRAN and uses a two-dimensional Cartesian coordinate geometry represented on a triangular lattice. The vacuum electric fields and equipotential lines are calculated for the input problem. The use may then introduce a series of emitters from which particles of different charge-to-mass ratios and initial energies can originate. These non-relativistic particles will then be traced by WOLF through the user-defined region. Effects of ion and electron space charge are included in the calculation. A subprogram PISA forms part of this code and enables optimization of various aspects of the problem. The WOLF package also allows detailed graphics analysis of the computed results to be performed
International Nuclear Information System (INIS)
As a sensitivity study the impact on the system reactivity was studied in the case that different calculational methodologies of spent fuel isotopic concentrations were used for WWER spent fuel inventory computations. The sets of isotopic concentrations obtained by calculations with different codes and libraries as a result of the CB2 international benchmark focused on WWER-440 burnup credit were used to show the spread of the calculated spent fuel system reactivity. Using the MCNP 4B code and changing the isotopics input data, the multiplication factor of an infinite array of the WWER-440 fuel pin cells was calculated. The evaluation of the results shows the sensitivity of the calculated reactivity to different calculational methodologies used for the spent fuel inventory computation. In the studied cases of the CB2 benchmark, the spread of the reference k-results relative to the mean was found less or about ±1% in spite of the fact that the data of isotopic concentrations were spread much more. (author)
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
Energy Technology Data Exchange (ETDEWEB)
Adams, C. H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.
CPS: a continuous-point-source computer code for plume dispersion and deposition calculations
Energy Technology Data Exchange (ETDEWEB)
Peterson, K.R.; Crawford, T.V.; Lawson, L.A.
1976-05-21
The continuous-point-source computer code calculates concentrations and surface deposition of radioactive and chemical pollutants at distances from 0.1 to 100 km, assuming a Gaussian plume. The basic input is atmospheric stability category and wind speed, but a number of refinements are also included.
Energy Technology Data Exchange (ETDEWEB)
Shibata, C.S.; Montes, A. [Instituto de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil); Galvao, R.M.O. [Sao Paulo Univ., SP (Brazil). Inst. de Fisica
1994-04-01
This paper describes the `FLINESH` computer code for magnetic fields calculation developed for the simulation of field configurations in plasma magnetic confinement devices. The expressions for the poloidal field and flux, the program structure and the input parameters description are presented, and also the analysis of the graphic output possibilities. (L.C.J.A.). 12 refs, 14 figs, 2 tabs.
Calculation of double differential cross sections for structural materials by PEGASUS code
International Nuclear Information System (INIS)
The neutron induced neutron and proton emission double differential cross sections were calculated with PEGASUS code for Cr, Fe and Ni and their isotopes. Results are in fair agreement with experimental data for neutron energy near 14 MeV, confirming that PEGASUS may be applied successfully to produce the double differential cross section data for JENDL. (author)
International Nuclear Information System (INIS)
The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)
A 3D coarse-mesh time dependent code for nuclear reactor kinetic calculations
International Nuclear Information System (INIS)
A course-mesh code for time-dependent multigroup neutron diffusion calculation based on a direct integration scheme for the time dependence and a low order nodal flux expansion approximation for the space variables has been implemented as a fast tool for transient analysis. (Author)
The solution of the LEU and MOX WWER-1000 calculation benchmark with the CARATE - multicell code
International Nuclear Information System (INIS)
Preparations for disposition of weapons grade plutonium in WWER-1000 reactors are in progress. Benchmark: Defined by the Kurchatov Institute (S. Bychkov, M. Kalugin, A. Lazarenko) to assess the applicability of computer codes for weapons grade MOX assembly calculations. Framework: 'Task force on reactor-based plutonium disposition' of OECD Nuclear Energy Agency. (Authors)
Calculate Some Characteristic Parameters Of VVER-1000's Fuel Assembly By MCNP4C2 Code
International Nuclear Information System (INIS)
This report presents the descriptions of parameters characteristics of the LEU and MOX Fuel Assemblies of VVER-1000 reactor, and calculation results such as infinite neutron multiplication factor kinf, two groups energies constants, neutron flux distribution by using Monte Carlo code MCNP. (author)
A Students Attendance System Using QR Code
Fadi Masalha; Nael Hirzallah
2014-01-01
Smartphones are becoming more preferred companions to users than desktops or notebooks. Knowing that smartphones are most popular with users at the age around 26, using smartphones to speed up the process of taking attendance by university instructors would save lecturing time and hence enhance the educational process. This paper proposes a system that is based on a QR code, which is being displayed for students during or at the beginning of each lecture. The students will need to scan the co...
Cooperative Regenerating Codes for Distributed Storage Systems
Shum, Kenneth W.
2011-01-01
When there are multiple node failures in a distributed storage system, regenerating the failed storage nodes individually in a one-by-one manner is suboptimal as far as repair-bandwidth minimization is concerned. If data exchange among the newcomers is enabled, we can get a better tradeoff between repair bandwidth and the storage per node. An explicit and optimal construction of cooperative regenerating code is illustrated.
International Nuclear Information System (INIS)
The objective of this study is to investigate the changes observed in the absorbed doses in mammary gland tissue when irradiated with a equipment of high dose rate known as Mammosite and introducing material resources contrary to the tissue that constitutes the mammary gland. The modeling study is performed with the code MCNPX, 2005 version, the equipment and the mammary gland and calculating the absorbed doses in tissue when introduced small volumes of air or calcium in the system. (Author)
Calculation of Plutonium content in RSG-GAS spent fuel using IAFUEL computer code
International Nuclear Information System (INIS)
It has been calculated the contain of isotopes Pu-239, Pu-240, Pu-241, and isotope Pu-242 in MTR reactor fuel types which have U-235 contain about 250 gram. The calculation was performed in three steps. The first step is to determine the library of calculation output of BOC (Beginning of Cycle). The second step is to determine the core isotope density, the weight of plutonium for one core, and one fuel isotope density. The third step is to calculate weight of plutonium in gram. All calculation is performed by IAFUEL computer code. The calculation was produced content of each Pu isotopes were Pu-239 is 6.7666 gr, Pu-240 is 1.4628 gr, Pu-241 is 0.52951 gr, and Pu-242 is 0.068952 gr
Verification of ARES transport code system with TAKEDA benchmarks
International Nuclear Information System (INIS)
Neutron transport modeling and simulation are central to many areas of nuclear technology, including reactor core analysis, radiation shielding and radiation detection. In this paper the series of TAKEDA benchmarks are modeled to verify the critical calculation capability of ARES, a discrete ordinates neutral particle transport code system. SALOME platform is coupled with ARES to provide geometry modeling and mesh generation function. The Koch–Baker–Alcouffe parallel sweep algorithm is applied to accelerate the traditional transport calculation process. The results show that the eigenvalues calculated by ARES are in excellent agreement with the reference values presented in NEACRP-L-330, with a difference less than 30 pcm except for the first case of model 3. Additionally, ARES provides accurate fluxes distribution compared to reference values, with a deviation less than 2% for region-averaged fluxes in all cases. All of these confirms the feasibility of ARES–SALOME coupling and demonstrate that ARES has a good performance in critical calculation
Verification of ARES transport code system with TAKEDA benchmarks
Zhang, Liang; Zhang, Bin; Zhang, Penghe; Chen, Mengteng; Zhao, Jingchang; Zhang, Shun; Chen, Yixue
2015-10-01
Neutron transport modeling and simulation are central to many areas of nuclear technology, including reactor core analysis, radiation shielding and radiation detection. In this paper the series of TAKEDA benchmarks are modeled to verify the critical calculation capability of ARES, a discrete ordinates neutral particle transport code system. SALOME platform is coupled with ARES to provide geometry modeling and mesh generation function. The Koch-Baker-Alcouffe parallel sweep algorithm is applied to accelerate the traditional transport calculation process. The results show that the eigenvalues calculated by ARES are in excellent agreement with the reference values presented in NEACRP-L-330, with a difference less than 30 pcm except for the first case of model 3. Additionally, ARES provides accurate fluxes distribution compared to reference values, with a deviation less than 2% for region-averaged fluxes in all cases. All of these confirms the feasibility of ARES-SALOME coupling and demonstrate that ARES has a good performance in critical calculation.
IPEN/MB-01 heavy reflector benchmark calculations using Serpent code
International Nuclear Information System (INIS)
A series of critical experiments with water-moderated square-pitched lattices with low-enriched uranium fuel rods was conducted at the IPEN/MB-01 research reactor facility, in 2005. Later, this data become some benchmarks. In one of these experiments the west face of the reactor core was covered with a set of thin SS-304 plates to simulate a heavy reflector as used in the EPR reactor (LEU-COMP-HERM-043). The plates are 3 mm thick and their width and axial length were large enough to cover one whole side of the active core of the reactor. The critical configurations were found as a function of the number of plates. Fuel rods containing UO2 with uranium enriched to 4.3% 235U were arranged in specific geometric configurations to be as close as possible to the critical state. In this work, these benchmark configurations with heavy reflectors were modeled using the Serpent Monte Carlo Code. Serpent uses a universe-based geometry model, which allows the description of practically any three-dimensional fuel or reactor configuration. Neutron transport is based on a combination of surface-to-surface ray-tracing and the Woodcock delta-tracking method. Woodcock method is many times faster than ray-tracing, so compared to MCNP code, Serpent code can bring huge gains in processing time of reactor calculations and reaction rate calculations. The results of these calculations were compared with experimental data and calculations with codes MCNP5 and SCALE6 (KENO-VI) using ENDF/B-VII.0 as cross-section input data. The codes performances are compared in terms of CPU calculation time and agreement with experimental data. Additional y, sensitivity on keff of Serpent woodcock threshold parameter was analyzed. (author)
Directory of Open Access Journals (Sweden)
Xiangbin Yu
2012-05-01
Full Text Available Differential space-time coding (DSTC technique has become a good choice when channel estimations are difficult to obtain in multiple antennas system. On the basis of analyzing the existing DSTC schemes, by introducing multi-level quadrature amplitude modulation (MQAM and matrix transform method, we develop a multi-amplitude differential space-time block coding (STBC scheme for square or non-square code matrix in MIMO systems, and give the derivation of calculation formulae of the coding advantage in detail. The developed scheme can effectively avoid the performance loss of conventional DSTC schemes based on PSK modulation (i.e. single amplitude DSTC in high spectrum efficiency. It can be applied to non-square code matrix case, and thus overcomes the shortcoming that existing DSTC schemes are only suitable for square code matrix. Compared with single amplitude DSTC schemes, our scheme has higher spectrum efficiency by carrying information not only on phases but also on amplitudes. Moreover, our scheme has linear decoding complexity, higher coding advantage, and higher code rate for more than two antennas. The simulations results show that the proposed scheme can provide lower BER than the existing single amplitude differential STBC schemes for both square and non-square code matrices.
Decay heat calculations with the CEA radioactivity data bauk and the code PEPIN
International Nuclear Information System (INIS)
The CEA radioactivity data bank, has been updated mainly from ENSDF and from some recent experimental results. This library contains the decay data for about 700 fission products (F.P.), 220 actinides and more than 1400 other nuclides. A comparison between our data and ENDF/B5 is shown for the fission products. The fission products part of this library is currently used for shielding and decay heat calculations with the PEPIN code. Calculations and spectral comparisons of the available experiments (Dickens, Lott, Yarmell ...) and other recent calculations is made for thermal fission of 235U and 239Pu using our data bank as input
International Nuclear Information System (INIS)
PEGASUS, a preequilibrium and evaporation theory code, was developed which calculates 17 neutron reaction cross sections, the particle spectra and the double differential cross sections. The code is suited to a rapid and scoping calculation. Theoretical model and the some results of calculation are presented. (author)
International Nuclear Information System (INIS)
Three computer codes MATHEW, ADPIC of LLNL and GAMPUL of JAERI for prediction of wind field, concentration and external exposure rate of airborne radioactive materials are vectorized and the results are presented. Using the continuous equation of incompressible flow as a constraint, the MATHEW calculates the three dimensional wind field by a variational method. Using the particle-in -cell method, the ADPIC calculates the advection and diffusion of radioactive materials in three dimensional wind field and terrain, and gives the concentration of the materials in each cell of the domain. The GAMPUL calculates the external exposure rate assuming Gaussian plume type distribution of concentration. The vectorized code MATHEW attained 7.8 times speedup by a vector processor FACOM230-75 APU. The ADPIC and GAMPUL are estimated to attain 1.5 and 4 times speedup respectively on CRAY-1 type vector processor. (author)
Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kliem, S.
1998-10-01
Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)
International Nuclear Information System (INIS)
This report describes a reactor-containment code, ALICE, which uses an arbitrary Lagrangian-Eulerian method to describe the coolant motion, together with a Lagrangian method to analyze the response of the containment vessel and other solid media inside a reactor containment. The finite-difference formulation used to approximate the governing equations for the motion of the coolant can be solved in either an explicit or an implicit scheme; the finite-element formulation used to approximate the governing equations for the containment vessel and other solid media can be performed only in the explicit scheme. Thus, the ALICE code can perform two types of coupling calculations for the fluid and structure (implicit-explicit and explicit-explicit). The code is generalized so that it can apply to problems either in a two-dimensional Cartesian or in a two-dimensional cylindrical-coordinate system
The GRS thermalhydraulic system code ATHLET
International Nuclear Information System (INIS)
The thermalhydraulic system code ATHLET is being developed by the Gesellschaft fuer Reaktorsicherheit (GRS) for the safety analysis of LWRs. The field of application comprises the whole spectrum of operational and abnormal transients, small and intermediate leaks as well as large breaks. An extended version (ATHLET-SA) that is being developed in cooperation with the Institut fuer Kernenergetik und Energiesysteme (IKE) at Stuttgart will allow the simulation of severe accidents with core degradation. A remarkable feature of ATHLET is the optional use of different fluiddynamic models. Depending on the type of problem to be analyzed fast running or more detailed models can be selected from essentially the same input deck. Emphasis is given to a methodical assessment procedure which provides the basis for the quantification of code uncertainties. (author). 12 refs
YNOGK: A New Public Code for Calculating Null Geodesics in the Kerr Spacetime
Yang, Xiaolin; Wang, Jiancheng
2013-07-01
Following the work of Dexter & Agol, we present a new public code for the fast calculation of null geodesics in the Kerr spacetime. Using Weierstrass's and Jacobi's elliptic functions, we express all coordinates and affine parameters as analytical and numerical functions of a parameter p, which is an integral value along the geodesic. This is the main difference between our code and previous similar ones. The advantage of this treatment is that the information about the turning points does not need to be specified in advance by the user, and many applications such as imaging, the calculation of line profiles, and the observer-emitter problem, become root-finding problems. All elliptic integrations are computed by Carlson's elliptic integral method as in Dexter & Agol, which guarantees the fast computational speed of our code. The formulae to compute the constants of motion given by Cunningham & Bardeen have been extended, which allow one to readily handle situations in which the emitter or the observer has an arbitrary distance from, and motion state with respect to, the central compact object. The validation of the code has been extensively tested through applications to toy problems from the literature. The source FORTRAN code is freely available for download on our Web site http://www1.ynao.ac.cn/~yangxl/yxl.html.
Blind Calculation of RD-14M Small Break LOCA Tests by CATHENA Code
International Nuclear Information System (INIS)
KAERI participated with the computer code CATHENA, which is used to analyze Pressurized Heavy Water Reactors (PHWRs), in an IAEA International Collaborative Standard Problem (ICSP) with the objective to benchmark and validate thermal-hydraulic computer code against qualified data for Small Break Loss of Coolant Accident (SBLOCA) scenario generated on RD-14M Test Facility. Two specific SBLOCA tests selected for this ICSP titled 'Comparison of HWR Code Predictions with SBLOCA Experimental Data', are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow down. This report presents the blind calculation results for these tests conducted by CATHENA code before the test data are distributed to participants. For B9006 test, CATHENA code simulated all the phases of the transient such as blowdown, high-pressure ECI, secondary pressure ramp, refill, switch from high pressure ECI to low pressure ECI, exponential pump ramp, and natural circulation. For B9802 test, CATHENA calculation was intended to predict temperature rise of the FES sheath due to channel boiling, and power supply trip on high FES sheath temperature (600 .deg. C) process protection trip
International Nuclear Information System (INIS)
Calculation of personnel exposure is a one of the main parameters being evaluated within the pre-decommissioning plans together with other decommissioning drivers such as costs, manpower, amounts of RAW and conventional waste and amount of discharged gaseous and liquid effluents. Alongside with manpower, the exposure is an indicator of the decommissioning process for need of staff, and quantifies impact of decommissioning on personnel from the radio hygienic point of view. At the same time it indicates suitability of individual work procedures use for decommissioning activities. For this reason it is important to estimate as precise as possible demands on personnel exposure even during preparatory decommissioning phase to quantify impact of decommissioning on personnel and eventually optimize the decommissioning process, if needed. The most appropriate way of staff exposure estimation during decommissioning preparatory phases is its calculation based on radiological and physical characteristics of equipment to be decommissioned and also quantitative and qualitative characterisation of typical decommissioning activities. On one hand, the methodology of exposure calculation should allow as much as possible realistic description and algorithmisation of exposure ways during decommissioning activities. On the other hand the calculation have to be systematic, well-arranged and clearly definable by appropriate mathematic relations. Calculation can be made by various approaches using more or less sophisticated software solutions from classic MS Excel sheets up to the complex calculation codes. In this paper, a methodology used for personnel exposure calculation and optimization implemented within the complex computer code OMEGA developed at DECOM, a.s. is described. (author)
Particle and heavy ion transport code system; PHITS
International Nuclear Information System (INIS)
Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear
Development of an effective delayed neutron fraction calculation code, BETA-K
Energy Technology Data Exchange (ETDEWEB)
Kim, Taek Kyum; Song, Hoon; Kim, Young Il; Kim, Young In; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-08-01
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method (NEM), has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction({beta}{sub eff}), neutron lifetime(l{sub eff}), fission spectrum ({chi}-bar) and fission yield data({nu}) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356 {mu}sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER. (author). 9 refs., 6 figs., 12 tabs.
Linear calculations of edge current driven kink modes with BOUT++ code
International Nuclear Information System (INIS)
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density
Development of an effective delayed neutron fraction calculation code for hexagonal core
International Nuclear Information System (INIS)
BETA-K, an effective delayed neutron fraction calculation code consistent with Nodal Expansion Method(NEM) of hexagonal geometric core, has been developed. By using relevant output files of DIF3D code, it can calculate the effective delayed neutron fraction(betaeff) and neutron lifetime(leff) for each fissionable isotope, composition of fuels and over the whole core. BETA-K code has been validated by comparing the calculated values to the measured ones of effective delayed neutron fraction in two critical experiments, BFS73-1 and BFS55-1. BFS73-1 is a metal uranium core and BFS55-1 is a metal plutonium core. The C/E values, 1.007 and 0.992 for BFS73-1 and BFS55-1 respectively, agreed well with the experimental values within the experiment errors. BETA-K code predicts 0.00709 and 0.356μ sec as the effective delayed neutron fraction and neutron life time for the uranium metallic fueled equilibrium core of 150MWe KALIMER
Linear calculations of edge current driven kink modes with BOUT++ code
Li, G. Q.; Xu, X. Q.; Snyder, P. B.; Turnbull, A. D.; Xia, T. Y.; Ma, C. H.; Xi, P. W.
2014-10-01
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
Linear calculations of edge current driven kink modes with BOUT++ code
Energy Technology Data Exchange (ETDEWEB)
Li, G. Q., E-mail: ligq@ipp.ac.cn; Xia, T. Y. [Institute of Plasma Physics, CAS, Hefei, Anhui 230031 (China); Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Xu, X. Q. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Snyder, P. B.; Turnbull, A. D. [General Atomics, San Diego, California 92186 (United States); Ma, C. H.; Xi, P. W. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); FSC, School of Physics, Peking University, Beijing 100871 (China)
2014-10-15
This work extends previous BOUT++ work to systematically study the impact of edge current density on edge localized modes, and to benchmark with the GATO and ELITE codes. Using the CORSICA code, a set of equilibria was generated with different edge current densities by keeping total current and pressure profile fixed. Based on these equilibria, the effects of the edge current density on the MHD instabilities were studied with the 3-field BOUT++ code. For the linear calculations, with increasing edge current density, the dominant modes are changed from intermediate-n and high-n ballooning modes to low-n kink modes, and the linear growth rate becomes smaller. The edge current provides stabilizing effects on ballooning modes due to the increase of local shear at the outer mid-plane with the edge current. For edge kink modes, however, the edge current does not always provide a destabilizing effect; with increasing edge current, the linear growth rate first increases, and then decreases. In benchmark calculations for BOUT++ against the linear results with the GATO and ELITE codes, the vacuum model has important effects on the edge kink mode calculations. By setting a realistic density profile and Spitzer resistivity profile in the vacuum region, the resistivity was found to have a destabilizing effect on both the kink mode and on the ballooning mode. With diamagnetic effects included, the intermediate-n and high-n ballooning modes can be totally stabilized for finite edge current density.
Aqueous Transport Code Revisions Using Geographic Information Systems
International Nuclear Information System (INIS)
STREAM II, developed at the Savannah River Site (SRS) for execution on a personal computer, is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River for emergency response management decision making. The STREAM II code consists of pre-processor, calculation, and post-processor modules. The pre-processor module provides a graphical user interface (GUI) for inputting the initial release data. The GUI passes the user specified data to the calculation module that calculates the pollutant concentrations at downstream locations and the transport times. The calculation module of the STREAM II adopts the transport module of the WASP5 code. WASP5 is a US Environmental Protection Agency water quality analysis program that simulates pollutant transport and fate through surface water using a finite difference method to solve the transport equation. The calculated downstream pollutant concentrations and travel times a re passed to the post-processor for display on the computer screen in graphical and tabular forms. To minimize the user's effort in the emergency situation, the required input parameters are limited to the time and date of release, type of release, location of release, amount and duration of release, and the calculation units. The user, however, could only select one of the seventeen predetermined locations. Hence, STREAM II could not be used for situations in which release locations differ from the seventeen predetermined locations. To eliminate this limitation, STREAM II has been revised to allow users to select the release location anywhere along the specified SRS main streams or the Savannah River by mouse-selection from a map displayed on the computer monitor. The required modifications to STREAM II using geographic information systems (GIS) software is discussed in this paper
Qualification of the HTR neutronics code system NEPHTIS
International Nuclear Information System (INIS)
AREVA's joint subsidiary with Siemens, Framatome ANP, has launched the ANTARES Program (AREVA New Technology based on Advanced gas cooled Reactor for Energy Supply) for the development of an advanced commercial HTR reactor. In this context, Framatome ANP is developing NEPHTIS in partnership with Cea and EDF. NEPHTIS is an industrial deterministic code system based on Apollo2 spectral code and Cronos2 core simulator, specially developed to handle the neutronics of prismatic block-type HTR reactors. This paper describes the NEPHTIS qualification program which includes HTTR and Fort Saint-Vrain experimental results. The preliminary qualification results for NEPHTIS are very promising. The HTTR results are well reproduced for the full core configuration (30 columns) with a discrepancy of 320 pcm. For annular configurations, intrinsically more complex to handle with a two-step deterministic approach, NEPHTIS results remain close to MCNP reference model (around 600 pcm). The isotopic content of a fuel assembly irradiated in Fort Saint-Vrain until 32 GWd/t is also well predicted by NEPHTIS and Monteburns with calculation vs. measurement discrepancies of around 3% for the main nuclides. The remaining validation domain is covered by code-to-code comparisons with Monte Carlo code systems (MCNP and Monteburns). A sample of the benchmarking between Apollo2 and Monteburns performed at assembly level on various Plutonium vectors are presented in this paper. The very good agreement between the two code systems validate the NEPHTIS depletion for UO2 and Pu fuel in HTR spectrum up to 60 GWd/t. (authors)
The spectral code Apollo2: from lattice to 2D core calculations
Energy Technology Data Exchange (ETDEWEB)
Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)
2005-07-01
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.
International Nuclear Information System (INIS)
The paper presents the computer code Mitra (Multicomponent isotope transport) which has been constructed to calculate the release of radioactive fission products from nuclear fuels under non-stationary conditions. The code is based on a new integration method fo the mass transport equation in the presence of precipitation, re-solution and radioactive decay. The starting equations and the assumed physical models are briefly described in the main part of the report. A very detailed description of the formulae used and of the Mitra subprograms are presented in extended appendices
The spectral code Apollo2: from lattice to 2D core calculations
International Nuclear Information System (INIS)
Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations
FLAME3: a three-dimensional nodal code for calculating core reactivity and power distributions
International Nuclear Information System (INIS)
The FLAME3 nodal code calculates core reactivity and three-dimensional power distributions with thermal-hydraulic feedback effects. It employs variable dimensioning, which permits the user to size his own problem subject to the total core storage of the computer. Lagrange interpolation is used for fitting variable data. This allows any input-dependent variable to be fit versus as many as three independent core variables. A transient xenon capability is included, which enhances the code's usefulness in performing maneuvering analyses. Control rod data are input by node, permitting the treatment of partial-length control rods. The various models, including the programmed equation, are described
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
Energy Technology Data Exchange (ETDEWEB)
Horhoianu, Grigore; Ionescu, Dragos Victor; Pauna, Eduard Ionut [Institute for Nuclear Research, Pitesti (Romania)
2011-05-15
The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO{sub 2} on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation. (orig.)
Comparison between CAREB code calculations and LOCA test results in the FUMEX III project
International Nuclear Information System (INIS)
The IAEA initiated a Coordinated Research Project (CRP) on improvement of computer codes used for fuel behaviour simulation under the name: FUMEX III. The Institute for Nuclear Research (INR) Pitesti participated at this CRP with ROFEM and CAREB computer codes. Recently, both codes have been improved with new models in order to extend their capabilities. The behaviour of fuel elements during high-temperature transients like LOCA is of importance to safety and licensing of power reactors. CAREB was developed for fuel transients analyses, such as LOCA and RIA. In this paper a comparison between CAREB code calculations and measured data from FIO-131 LOCA tests is presented. Several parameters were considered, including fuel sheath strains, internal element gas pressure, fuel centerline and sheath temperature, thicknesses of ZrO2 on the sheath. Fuel behavior during high-temperature transient was reasonably well modeled by CAREB code. New LOCA tests are planed to be performed in the C2-LOCA facility of the TRIGA research reactor at INR Pitesti in order to extend the experimental data base used for transient code validation. (orig.)
Comparison of calculations made with three two-dependent neutron codes TDA, MORSE and POW
International Nuclear Information System (INIS)
Three computer codes were compared to determine their usefulness in analysing a pulsed neutron experiment. The codes were a Monte Carlo code (MORSE) a diffusion kinetics code (POW) and a time dependent SN code (TDA). A series of test problems were devised to progressively model the experiment. All problems assumed a spherical system with an isotropic source. The first problem had its source in the first energy group for the first nano-second. The second problem had its source distributed in time but not distributed in energy. The third problem had a source distributed in energy and time. POW and MORSE were shown to be in good agreement, with significant differences occurring only at times when the system did not correspond with the approximations made in POW. The AAEC version of the TDA code did not handle a time-dependent source. There was also a tendency for the results beyond 50 ns to be higher than those for the other two codes for the problems having a source constant over one time interval. (author)
Design of a transport calculation system for logging sondes simulation
International Nuclear Information System (INIS)
Analysis of available resources in earth crust is performed by different techniques, one of them is neutron logging. Design of sondes that are used to make such logging is supported by laboratory experiments as well as by numerical calculations.This work presents several calculation schemes, designed to simplify the task of whom has to planify such experiments or optimize parameters of this kind of sondes.These schemes use transport calculation codes, especially DaRT, TORT and MCNP, and cross section processing modules from SCALE system.Additionally a system for DaRT and TORT data postprocessing using OpenDX is presented.It allows scalar flux spatial distribution analysis, as wells as cross section condensation and reaction rates calculation
Desch, Steven; Lorenzo, Alejandro; Ko, Byeongkwan
2016-06-01
We present a computer code we have written for general release that calculates the interior structure and mass-radius relationships of solid exoplanets up to a few Earth masses. The basic algorithm is that of Seager et al. (2007), Zeng & Sasselov (2013) and Dorn et al. (2015): the code integrates the 1-D (spherical) equation of hydrostatic equilibrium to find pressure in shells of various depths assuming a gravitational acceleration, uses the bulk modulus of the materials as inputs to an equation of state to convert pressures into density and volume in each shell, recomputes the shell thicknesses and gravitational acceleration, and iterates the solution to convergence. Unlike most existing codes, we do not impose a particular mineralogy in each shell. Instead we adopt the approach of Dorn et al. (2015), in which we impose a stoichiometry in each shell; for rocky shells and the metal core the code calls the PerpleX code (Connolly et al. 2005) to compute the mineralogy and material properties appropriate to that shell’s stoichiometry, pressure and temperature. Unique attributes of the code are as follows. The mineralogy is complete in the Fe-Mg-Si-O system, including species like FeSi and FeO in the core. We also include FeS (VII) in the core. We have also included an approximate phase diagram for water ice to account for an icy mantle. We also include the effects of adiabatic temperature profiles and a temperature jump at the core-mantle boundary. Finally, we have created a user-friendly interface allowing the code to be downloaded and used as a teaching tool. Results of the code and a demonstration of its use will be presented at the meeting.
Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code
International Nuclear Information System (INIS)
In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)
Development of Monte Carlo decay gamma-ray transport calculation system
International Nuclear Information System (INIS)
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
Radioactivity evaluation code system for high temperature gas cooled reactors
International Nuclear Information System (INIS)
A code system for the evaluation of the behavior of radioactive fission products (FP) in high temperature gas-cooled reactors (HTGR) is described. The first half of this report is devoted to the description of the conceivable behavior of FPs in the experimental very high temperature gas-cooled reactor being designed at JAERI. The transfer of FPs from the fuel to the primary coolant is considered in three steps; the release of FPs from the coated fuel particles; the diffusion of FPs within graphite sleeves; and the transfer of FPs from the sleeve surface to the coolant. As for the FP behavior within the primary coolant system, the deposition of FPs on various walls of the system is considered. As for the secondary and the thermal utilization systems, the transfer of tritium is specially considered. The calculation model for the transfer and deposition of fission products within the whole plant system is presented by a chart. The second half of this report describes the evaluation code system. The physical and mathematical models treated in each component code are presented and discussed. (Aoki, K.)
Problems of optimal data coding in hodoscopic systems
International Nuclear Information System (INIS)
An analogy system of algebraic coding theory and of hodoscopic system coding theory is considered. The connection between main parameters of coding devices and parameters of parallel coders applied in hodoscopic systems is established. The efficiency of using a proposed analogy system is illustrated on some examples of designing parallel coders with given properties
SYSMOD: user-interface for data processing, calculation codes and analysis of PWR lattices
International Nuclear Information System (INIS)
The task of the physical calculation of the reactor demand of the management of a great volume of information and inclose the stages for processing of data, calculations and analysis of their results. These stages are highly sensible to human mistakes, that's why is imprescindible that them undergo automatization, doing tracked all the process against mistake or unexpected result. The user-interface SYSMOD was developed over the platform IDE Delphi 3.0, visual language driven to events. It to consist in of the principal menu, which inclose between its options the preparation of the input data (File and Edit) to the pre-processors for the calculation codes of reactors. The output information may be showed in graphic and/or alphanumeric format (Data-Process). SYSMOD endures two applications for the management of the data base for the data during the preparation of the input for the pre-processors of the spectral calculation, so as for the organization, conservation and presentation for the obtained results. The carried out of the lattices and global codes, takes place from this application, over the platform MS-DOS (Run). SYSMOD regards the possibility for the debugging of the codes (Debugging), so as the benchmarks qualified to so effect (Benchmark). SYSMOD has been applied for the analysis of te WWER-440 of the first unity of Juragua Nuclear Power Plant. (author)
International Nuclear Information System (INIS)
The conversion coefficients, H'(d,α)/φ, for monoenergetic positrons and positron-emitting radionuclides were calculated by using the user code UCICRPM of the Monte Carlo code EGS5 to estimate the radiation dose for medical staff involved in positron emission tomography examinations. From these coefficients, the dose equivalent rates per unit activity at 0.07 and 10 mm depths in a soft tissue for a straight-line source of 2-deoxy-2-[18F]fluoro-d-glucose (18F-FDG) were calculated by using the developed user code UCF18DOSE. The dose equivalent rates per unit activity at 0.07 and 10 mm depths were measured by using a personal dosemeter (DOSE 3) under the same conditions as those considered in the calculation. The calculated dose equivalent rates per unit activity at 0.07 and 10 mm depths were 0.116 and 0.0352 pSv min-1 Bq-1, respectively, at 20 cm from the 18F-FDG injection tube. (authors)
Calculation of Gamma-ray Responses for HPGe Detectors with TRIPOLI-4 Monte Carlo Code
Lee, Yi-Kang; Garg, Ruchi
2014-06-01
The gamma-ray response calculation of HPGe (High Purity Germanium) detector is one of the most important topics of the Monte Carlo transport codes for nuclear instrumentation applications. In this study the new options of TRIPOLI-4 Monte Carlo transport code for gamma-ray spectrometry were investigated. Recent improvements include the gamma-rays modeling of the electron-position annihilation, the low energy electron transport modeling, and the low energy characteristic X-ray production. The impact of these improvements on the detector efficiency of the gamma-ray spectrometry calculations was verified. Four models of HPGe detectors and sample sources were studied. The germanium crystal, the dead layer of the crystal, the central hole, the beryllium window, and the metal housing are the essential parts in detector modeling. A point source, a disc source, and a cylindrical extended source containing a liquid radioactive solution were used to study the TRIPOLI-4 calculations for the gamma-ray energy deposition and the gamma-ray self-shielding. The calculations of full-energy-peak and total detector efficiencies for different sample-detector geometries were performed. Using TRIPOLI-4 code, different gamma-ray energies were applied in order to establish the efficiency curves of the HPGe gamma-ray detectors.
Energy Technology Data Exchange (ETDEWEB)
Hashimoto, M.; Saito, K.; Ando, H. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-05-01
The method to calculate the response function of spherical BF{sub 3} proportional counter, which is commonly used as neutron dose rate meter and neutron spectrometer with multi moderator system, is developed. As the calculation code for evaluating the response function, the existing code series NRESP, the Monte Carlo code for the calculation of response function of neutron detectors, is selected. However, the application scope of the existing NRESP is restricted, the NRESP98 is tuned as generally applicable code, with expansion of the geometrical condition, the applicable element, etc. The NRESP98 is tested with the response function of the spherical BF{sub 3} proportional counter. Including the effect of the distribution of amplification factor, the detailed evaluation of the charged particle transportation and the effect of the statistical distribution, the result of NRESP98 calculation fit the experience within {+-}10%. (author)
Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh
International Nuclear Information System (INIS)
The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)
Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations
International Nuclear Information System (INIS)
The calculator program ED (Emergency Doses) was developed from several HP-41CV calculator programs documented in the report Seven Health Physics Calculator Programs for the HP-41CV, RHO-HS-ST-5P (Rittman 1984). The program was developed to enable estimates of offsite impacts more rapidly and reliably than was possible with the software available for emergency response at that time. The ED - Revision 3, documented in this report, revises the inhalation dose model to match that of ICRP 30, and adds the simple estimates for air concentration downwind from a chemical release. In addition, the method for calculating the Pasquill dispersion parameters was revised to match the GENII code within the limitations of a hand-held calculator (e.g., plume rise and building wake effects are not included). The summary report generator for printed output, which had been present in the code from the original version, was eliminated in Revision 3 to make room for the dispersion model, the chemical release portion, and the methods of looping back to an input menu until there is no further no change. This program runs on the Hewlett-Packard programmable calculators known as the HP-41CV and the HP-41CX. The documentation for ED - Revision 3 includes a guide for users, sample problems, detailed verification tests and results, model descriptions, code description (with program listing), and independent peer review. This software is intended to be used by individuals with some training in the use of air transport models. There are some user inputs that require intelligent application of the model to the actual conditions of the accident. The results calculated using ED - Revision 3 are only correct to the extent allowed by the mathematical models. 9 refs., 36 tabs
Use of generalized curvilinear coordinate systems in electromagnetic and hybrid codes
Energy Technology Data Exchange (ETDEWEB)
Swift, D.W. [Univ. of Alaska, Fairbanks, AK (United States)
1995-07-01
The author develops a code to simulate the dynamics in the magnetosphere system. The calculation involves a single level, structured, curvilinear 2D mesh. The mesh density is varied to support regions which demand higher resolution.
Hyperspherical calculations for four-nucleon systems
International Nuclear Information System (INIS)
We develop hyperspherical calculations on the bound states of four-nucleon systems and particularly the fundamental level and the first 0+ excited states. With neglect of the Coulomb effect, we analyze the convergence of the optimal subset expansion for the binding energies calculated for central or realistic potentials. 35 refs
How accurately can we calculate thermal systems?
International Nuclear Information System (INIS)
The objective was to determine how accurately simple reactor lattice integral parameters can be determined, considering user input, differences in the methods, source data and the data processing procedures and assumptions. Three simple square lattice test cases with different fuel to moderator ratios were defined. The effect of the thermal scattering models were shown to be important and much bigger than the spread in the results. Nevertheless, differences of up to 0.4% in the K-eff calculated by continuous energy Monte Carlo codes were observed even when the same source data were used. (author)
Wall-touching kink mode calculations with the M3D code
International Nuclear Information System (INIS)
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall
Calculation of the RSG-GAS core using computer code citation-3D
International Nuclear Information System (INIS)
Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that Keff are less and greater than unity (Keffeff>1)
Wall-touching kink mode calculations with the M3D code
Energy Technology Data Exchange (ETDEWEB)
Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08542 (United States)
2015-06-15
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
VizieR Online Data Catalog: ynogkm: code for calculating time-like geodesics (Yang+, 2014)
Yang, X.-L.; Wang, J.-C.
2013-11-01
Here we present the source file for a new public code named ynogkm, aim on calculating the time-like geodesics in a Kerr-Newmann spacetime fast. In the code the four Boyer-Lindquis coordinates and proper time are expressed as functions of a parameter p semi-analytically, i.e., r(p), μ(p), φ(p), t(p), and σ(p), by using the Weiers- trass' and Jacobi's elliptic functions and integrals. All of the ellip- tic integrals are computed by Carlson's elliptic integral method, which guarantees the fast speed of the code.The source Fortran file ynogkm.f90 contains three modules: constants, rootfind, ellfunction, and blcoordinates. (3 data files).
Systemization of burnup sensitivity analysis code
International Nuclear Information System (INIS)
To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints 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 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 core '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, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user 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 functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; 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 this
System analysis of bar code laser scanner
Wang, Jianpu; Chen, Zhaofeng; Lu, Zukang
1996-10-01
This paper focuses on realizing the three important aspects of bar code scanner: generating a high quality scanning light beam, acquiring a fairly even distribution characteristic of light collection, achieving a low signal dynamic range over a large depth of field. To do this, we analyze the spatial distribution and propagation characteristics of scanning laser beam, the vignetting characteristic of optical collection system and their respective optimal design; propose a novel optical automatic gain control method to attain a constant collection over a large working depth.
International Nuclear Information System (INIS)
A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified. Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab
SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR
International Nuclear Information System (INIS)
SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)
Post test calculations of a severe accident experiment for VVER-440 reactors by the ATHLET code
Energy Technology Data Exchange (ETDEWEB)
Gyoergy, Hunor [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques (BME NTI); Trosztel, Istvan [Hungarian Academy of Sciences, Budapest (Hungary). Centre for Energy Research (MTA EK)
2013-09-15
Severe accident - if no mitigation action is taken - leads to core melt. An effective severe accident management strategy can be the external reactor pressure vessel cooling for corium localization and stabilization. For some time discussion was going on, whether the in-vessel retention can be applied for the VVER-440 type reactors. It had to be demonstrated that the available space between the reactor vessel and biological protection allows sufficient cooling to keep the melted core in the vessel, without the reactor pressure vessel losing its integrity. In order to demonstrate the feasibility of the concept an experimental facility was realized in Hungary. The facility called Cooling Effectiveness on the Reactor External Surface (CERES) is modeling the vessel external surface and the biological protection of Paks NPP. A model of the CERES facility for the ATHLET TH system code was developed. The results of the ATHLET calculation agree well with the measurements showing that the vessel cooling can be insured for a long time in a VVER-440 reactor. (orig.)
Systemization of burnup sensitivity analysis code. 2
International Nuclear Information System (INIS)
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
Shielding evaluation for e-Linac - Inter-comparison of Monte Carlo codes and analytical calculations
International Nuclear Information System (INIS)
Estimation of optimum shielding thickness is an important aspect in radiation protection as well as in assessment of cost effectiveness of any upcoming accelerator facility. Analytical calculations for shielding estimates are fast and being frequently used even though they are very approximate. Estimates by Monte Carlo codes, on the other hand is accurate, provided used in a judicious manner, but they are very time consuming and require high end computational hardware. The purpose of this work is to compare the results from various available Monte Carlo codes, such as FLUKA and EGSmc. The estimated output was also compared with the analytical techniques. For the work, an e-Linac facility of 50 MeV electron beam was used and calculations were carried out with 1 mA beam current. (author)
CHARADE: A characteristic code for calculating rate-dependent shock-wave response
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.N.; Tonks, D.L.
1991-01-01
In this report we apply spatially one-dimensional methods and simple shock-tracking techniques to the solution of rate-dependent material response under flat-plate-impact conditions. This method of solution eliminates potential confusion of material dissipation with artificial dissipative effects inherent in finite-difference codes, and thus lends itself to accurate calculation of elastic-plastic deformation, shock-to-detonation transition in solid explosives, and shock-induced structural phase transformation. Equations are presented for rate-dependent thermoelastic-plastic deformation for (100) planar shock-wave propagation in materials of cubic symmetry (or higher). Specific numerical calculations are presented for polycrystalline copper using the mechanical threshold stress model of Follansbee and Kocks with transition to dislocation drag. A listing of the CHARADE (for characteristic rate dependence) code and sample input deck are given. 26 refs., 11 figs.
Comprehensive nuclear model calculations: Introduction to the theory and use of the GNASH code
International Nuclear Information System (INIS)
A user's manual describing the theory and operation of the GNASH nuclear reaction computer code is presented. This work is based on a series of lectures describing the statistical Hauser-Feshbach plus preequilibrium version of the code with full angular momentum conservation. This version is expected to be most applicable for incident particle energies between 1 key and 50 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93Nb and 12-MeV neutrons incident on 238U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH
VVER 1000 SBO calculations with pressuriser relief valve stuck open with ASTEC computer code
International Nuclear Information System (INIS)
Highlights: ► We modelled the ASTEC input file for accident scenario (SBO) and focused analyses on the behaviour of core degradation. ► We assumed opening and stuck-open of pressurizer relief valve during performance of SBO scenario. ► ASTEC v1.3.2 has been used as a reference code for the comparison study with the new version of ASTEC code. - Abstract: The objective of this paper is to present the results obtained from performing the calculations with ASTEC computer code for the Source Term evaluation for specific severe accident transient. The calculations have been performed with the new version of ASTEC. The ASTEC V2 code version is released by the French IRSN (Institut de Radioprotection at de surete nucleaire) and Gesellschaft für Anlagen-und Reaktorsicherheit (GRS), Germany. This investigation has been performed in the framework of the SARNET2 project (under the Euratom 7th framework program) by Institute for Nuclear Research and Nuclear Energy – Bulgarian Academy of Science (INRNE-BAS).
Concatenated coding system with iterated sequential inner decoding
DEFF Research Database (Denmark)
Jensen, Ole Riis; Paaske, Erik
We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder......We describe a concatenated coding system with iterated sequential inner decoding. The system uses convolutional codes of very long constraint length and operates on iterations between an inner Fano decoder and an outer Reed-Solomon decoder...
International Nuclear Information System (INIS)
Highlights: • We present a new Monte Carlo method to perform sensitivity/perturbation calculations. • Sensitivity of keff, reaction rates, point kinetics parameters to nuclear data. • Fully continuous implicitly constrained Monte Carlo sensitivities to scattering distributions. • Implementation of the method in the continuous energy Monte Carlo code SERPENT. • Verification against ERANOS and TSUNAMI generalized perturbation theory results. - Abstract: In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the effects of nuclear data perturbation on several response functions: the effective multiplication factor, reaction rate ratios and bilinear ratios (e.g., effective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators
International Nuclear Information System (INIS)
Experiments were done on several aerosols in air atmospheres at varying temperatures and humidity conditions of interest in forming a data base for testing aerosol behavior models used as part of the process of evaluating the ''source term'' in light water reactor accidents. This paper deals with the problems of predicting the observed experimental data for suspended aerosol concentration with aerosol calculational codes. Comparisons of measured versus predicted data are provided
Efficient Calculations with Multisite Local Orbitals in a Large-Scale DFT Code CONQUEST
Nakata, A; Bowler, D. R.; Miyazaki, T.
2014-01-01
Multisite local orbitals, which are formed from linear combinations of pseudoatomic orbitals from a target atom and its neighbor atoms, have been introduced in the large-scale density functional theory calculation code CONQUEST. Multisite local orbitals correspond to local molecular orbitals so that the number of required local orbitals can be minimal. The multisite support functions are determined by using the localized filter diagonalization (LFD) method [ Phys. Rev. B 2009 , 80 , 205104 ]....
Opacity calculation for target physics using the ABAKO/RAPCAL code
Mínguez Torres, Emilio; Florido, Ricardo; Rodríguez, Rafael; Gil, J.M.; Garcia Rubiano, Jesus; Mendoza, M. A.; Suarez, D; Martel, Pablo
2010-01-01
Radiative properties of hot dense plasmas remain a subject of current interest since they play an important role in inertial confinement fusion (ICF) research, as well as in studies on stellar physics. In particular, the understanding of ICF plasmas requires emissivities and opacities for both hydro-simulations and diagnostics. Nevertheless, the accurate calculation of these properties is still an open question and continuous efforts are being made to develop new models and numerical codes th...
INTRACOIN level 1 benchmark calculations with EIR codes CONZRA, RANCH and RANCHN
International Nuclear Information System (INIS)
The authors present the results from calculations of INTRACOIN level 1, case 1 and 2 (one-dimensional advection-dispersion) benchmarks. The codes used are CONZRA and RANCH, corresponding to a semi-analytical solution of the transport equation, and RANCHN based on a fully numerical solution in the framework of the pseudo-spectral method. The influence of various boundary conditions is investigated. Excellent agreement between results from the different solution approaches is obtained. (Auth.)
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
Energy Technology Data Exchange (ETDEWEB)
Cho, Jin-Young; Kim, Kang-Seog; Lee, Chung-Chan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2006-07-01
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems.
Benchmark Calculation for the VHTR 2-D Core by Using the DeCART Code
International Nuclear Information System (INIS)
Recently, a hexagonal module has been equipped to the DeCART (Deterministic Core Analysis based on Ray Tracing) whole core code for a hexagonal core analysis. The equipment includes a ray tracing module to solve the 2-D whole-core transport problem and a multi-group CMFD module to perform an efficient transport calculation. In this paper, the capability of the DeCART hexagonal module is examined by solving VHTR core problems
Permanent boiling in rod bundles: calculations with the FLICA II B code
International Nuclear Information System (INIS)
Some calculations have been made with the FLICA II B code using a model representing rod-bundles by means of few interconnected channels. The result obtained give some idea of the similarities and differences of behavior to be expected between single channels and rod bundles during permanent forced convection sodium boiling regimes. A new phenomenon, the so-called internal flow excursion is described
Energy Technology Data Exchange (ETDEWEB)
Pahn, T.; Jonkman, J.; Rolges, R.; Robertson, A.
2012-11-01
Physically measuring the dynamic responses of wind turbine support structures enables the calculation of the applied loads using an inverse procedure. In this process, inverse means deriving the inputs/forces from the outputs/responses. This paper presents results of a numerical verification of such an inverse load calculation. For this verification, the comprehensive simulation code FAST is used. FAST accounts for the coupled dynamics of wind inflow, aerodynamics, elasticity and turbine controls. Simulations are run using a 5-MW onshore wind turbine model with a tubular tower. Both the applied loads due to the instantaneous wind field and the resulting system responses are known from the simulations. Using the system responses as inputs to the inverse calculation, the applied loads are calculated, which in this case are the rotor thrust forces. These forces are compared to the rotor thrust forces known from the FAST simulations. The results of these comparisons are presented to assess the accuracy of the inverse calculation. To study the influences of turbine controls, load cases in normal operation between cut-in and rated wind speed, near rated wind speed and between rated and cut-out wind speed are chosen. The presented study shows that the inverse load calculation is capable of computing very good estimates of the rotor thrust. The accuracy of the inverse calculation does not depend on the control activity of the wind turbine.
Calculation of effective delayed neutron fraction with modified library of Monte Carlo code
International Nuclear Information System (INIS)
Highlights: ► We propose a new Monte Carlo method to calculate the effective delayed neutron fraction by changing the library. ► We study the stability of our method. When the particles and cycles are sufficiently great, the stability is very good. ► The final result is determined to make the deviation least. ► We verify our method on several benchmarks, and the results are very good. - Abstract: A new Monte Carlo method is proposed to calculate the effective delayed neutron fraction βeff. Based on perturbation theory, βeff is calculated with modified library of Monte Carlo code. To verify the proposed method, calculations are performed on several benchmarks. The error of the method is analyzed and the way to reduce error is proposed. The results are in good agreement with the reference data
Calculation of age-dependent effective doses for external exposure using the MCNP code
Energy Technology Data Exchange (ETDEWEB)
Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)
2013-07-15
Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)
First vapor explosion calculations performed with MC3D thermal-hydraulic code
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires
1998-01-01
This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)
BETHSY 6.2TC test calculation with TRACE and RELAP5 computer code
International Nuclear Information System (INIS)
The TRACE code is still under development and it will have all capabilities of RELAP5. The purpose of the present study was therefore to assess the accuracy of the TRACE calculation of BETHSY 6.2TC test, which is 15.24 cm equivalent diameter horizontal cold leg break. For calculations the TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used. The overall results obtained with TRACE were similar to the results obtained by RELAP5/MOD3.3. The results show that the discrepancies were reasonable. (author)
International Nuclear Information System (INIS)
For safety evaluation of nuclear fuel facilities, a nuclear decay data library named JDDL and a computer code COMRAD have been developed to calculate isotopic composition of each nuclide, radiation source intensity, energy spectrum of γ-ray and neutron, and decay heat of spent fuel. JDDL has been produced mainly from the evaluated nuclear data file ENSDF to use new nuclear data. To supplement the data file for short life nuclides, the JNDC data set were also used which had been evaluated by Japan Nuclear Data Committee. Using these data, calculations became possible from short period to long period after irradiation. (author)