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Brega, E.; Salina, E.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: The NORMA code is intended to solve multigroup diffusion problems or two-group diffusion-depletion problems in three dimensions. In order to simulate the fuel depletion, the core lifetime is divided into suitable burnup steps in which the nuclear parameters, required by the spatial solution of the diffusion equation in steady-state critical condition, are computed as functions of local burnup, coolant density and burnup-weighted coolant density (or spectral ratio) and are corrected for dilute Boron, Sm and Xe poisoning and Doppler effects with the further provision for an automatic dilute Boron adjustment to achieve criticality. Moreover, at the beginning of each burnup step, general directions for fuel management are provided. 2 - METHODS: The burnup model is implicit and the overall solution scheme at each burnup step is iterative. The multigroup diffusion equation is approximated by a coarse-mesh fourth-order polynomial method. More precisely, the numerical approximation is represented by coarse-mesh finite-difference equations corrected by discontinuity factors internally computed so as to match the accuracy of a fourth-order polynomial approach. Then, the finite-difference equations are solved for the nodal neutron fluxes by the usual inner/outer scheme. The equivalent homogenised nuclear parameters including the discontinuity factors correcting for homogenisation errors, according to Henry's generalised equivalence theory, are updated by trilinear interpolation in input libraries of reference values. The thermal-hydraulic model (an upgraded version of the COBRA-3C code called COBRA-EN) is based on three partial differential equations that describe the conservation of mass, energy and momentum for the water liquid/vapor mixture and the interaction of the two-phase coolant with the system structures. Optionally, a fourth equation can be added which tracks the vapor mass separately and which, along with the correlations for vapor generation and slip ratio, replaces the subcooled quality and quality/void fraction correlations, needed by the homogeneous model. In each coolant channel, the one-dimensional (z) fluid dynamics equations in the vertical direction as well as the one-dimensional (r) equation in the horizontal direction that models the heat transfer in solid structures are approximated by finite differences. The resulting equations of hydrodynamic phenomena form a system of coupled nonlinear equations that are solved by the upflow scheme (allowed only when no reverse flow is predicted) or by a Newton-Raphson iteration procedure (needed when the vapor mass continuity equation is added). The heat-transfer equations in the solid structures are treated implicitly. Moreover, a full boiling curve is provided, comprising the basic heat-transfer regimes, each represented by a set of optional correlations for the heat-transfer coefficient between a solid surface and the coolant bulk. 3 - Restrictions on the complexity of the problem: The data-dependent arrays are contained in the named Common block BLANK whose standard length of 4x10(6) bytes can be changed by modifying a PARAMETER statement in an include file (see the Installation Directions)

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27 Oct 1999; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1388.html; Country of input: International Atomic Energy Agency (IAEA); 1 ref.

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APPROXIMATIONS, BORON, BURNUP, COMPUTER PROGRAM DOCUMENTATION, CONTINUITY EQUATIONS, COOLANTS, CRITICALITY, DIFFUSION, DIFFUSION EQUATIONS, FUEL MANAGEMENT, FUEL RODS, GROUP THEORY, HEAT TRANSFER, INTERPOLATION, ITERATIVE METHODS, N CODES, NEUTRON FLUX, NEUTRONS, NONLINEAR PROBLEMS, ONE-DIMENSIONAL CALCULATIONS, POLYNOMIALS, POWER DISTRIBUTION, REACTIVITY, REACTOR CORES, REACTOR SAFETY, SLIP RATIO, STEADY-STATE CONDITIONS, THERMAL HYDRAULICS, THERMODYNAMICS, THREE-DIMENSIONAL CALCULATIONS, TWO-PHASE FLOW, VAPORS, VOID FRACTION, WATER COOLED REACTORS, WEBSITES

BARYONS, CALCULATION METHODS, COMPUTER CODES, DIFFERENTIAL EQUATIONS, DIMENSIONLESS NUMBERS, DOCUMENT TYPES, ELEMENTARY PARTICLES, ELEMENTS, ENERGY TRANSFER, EQUATIONS, FERMIONS, FLUID FLOW, FLUID MECHANICS, FLUIDS, FUEL ELEMENTS, FUNCTIONS, GASES, HADRONS, HYDRAULICS, MANAGEMENT, MATHEMATICAL SOLUTIONS, MATHEMATICS, MECHANICS, NUCLEAR MATERIALS MANAGEMENT, NUCLEONS, NUMERICAL SOLUTION, PARTIAL DIFFERENTIAL EQUATIONS, RADIATION FLUX, REACTOR COMPONENTS, REACTORS, SAFETY, SEMIMETALS

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Kloosterman, Jan Leen

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description of program or function: Contains experimental benchmarks which can be used for the validation of burnup code systems and accompanied data libraries. Although the benchmarks presented here are thoroughly described in literature, it is in many cases not straightforward to retrieve unambiguously the correct input data and corresponding results from the benchmark Descriptions. Furthermore, results which can easily be measured, are sometimes difficult to calculate because of conversions to be made. Therefore, emphasis has been put to clarify the input of the benchmarks and to present the benchmark results in such a way that they can easily be calculated and compared. For more thorough Descriptions of the benchmarks themselves, the literature referred to here should be consulted. This benchmark book is divided in 11 chapters/files containing the following in text and tabular form: chapter 1: Introduction; chapter 2: Burnup Credit Criticality Benchmark Phase 1-B; chapter 3: Yankee-Rowe Core V Fuel Inventory Study; chapter 4: H.B. Robinson Unit 2 Fuel Inventory Study; chapter 5: Turkey Point Unit 3 Fuel Inventory Study; chapter 6: Turkey Point Unit 3 Afterheat Power Study; chapter 7: Dickens Benchmark on Fission Product Energy Release of U-235; chapter 8: Dickens Benchmark on Fission Product Energy Release of Pu-239; chapter 9: Yarnell Benchmark on Decay Heat Measurements of U-233; chapter 10: Yarnell Benchmark on Decay Heat Measurements of U-235; chapter 11: Yarnell Benchmark on Decay Heat Measurements of Pu-239

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11 Jun 1999; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1606.html; Country of input: International Atomic Energy Agency (IAEA); 11 refs.

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ACTINIDE NUCLEI, ALPHA DECAY RADIOISOTOPES, COMPUTER CODES, DOCUMENT TYPES, ENERGY, EVEN-ODD NUCLEI, HEAVY ION DECAY RADIOISOTOPES, HEAVY NUCLEI, INTERNAL CONVERSION RADIOISOTOPES, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, MANAGEMENT, MATERIALS, MINUTES LIVING RADIOISOTOPES, NEON 24 DECAY RADIOISOTOPES, NUCLEAR MATERIALS MANAGEMENT, NUCLEI, PLUTONIUM ISOTOPES, RADIOACTIVE MATERIALS, RADIOISOTOPES, SPONTANEOUS FISSION RADIOISOTOPES, TESTING, URANIUM ISOTOPES, YEARS LIVING RADIOISOTOPES

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Durmayaz, Ahmet

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: A computer program for the fast computation of the thermodynamic and transport properties of heavy water (D

_{2}O) at saturation, in subcooled liquid and superheated vapor states. Specific volume (or density), specific enthalpy, specific entropy, constant-pressure specific heat and temperature at saturation are calculated by a number of piecewise continuous approximation functions of (and their derivatives are calculated with respect to) pressure whereas pressure at saturation is calculated by a piecewise continuous approximation function of temperature for heavy water. Density in subcooled liquid state, specific volume in super-heated vapor state, specific enthalpy, specific entropy and constant-pressure specific heat in both of these states are calculated by some piecewise continuous approximation functions of pressure and temperature for heavy water. The correlations used in the calculation of these thermodynamic properties of heavy water were derived by fitting some appropriate curves to the data given in the steam tables by Hill et al (1981). The whole set of correlations and the approximation method used in their derivation are presented by Durmayaz (1997). Dynamic viscosity and thermal conductivity for heavy water are calculated as functions of temperature and density with the correlations given by Hill et al (1981), by Matsunaga and Nagashima (1983) and by Kestin et al (1984). Surface tension for heavy water is calculated as a function of temperature with the correlation given by Crabtree and Siman-Tov (1993). 2 - Methods: A group of pressure-enthalpy (P-h) pairs can be given in an input data file or assigned in the main program without knowing the state in which fluid takes place. In this case, first, the enthalpies at saturation corresponding to the given pressure are computed. Second, the state is determined by comparing the given enthalpy to the saturation enthalpies. Then, the properties are computed. Program D_{2}O can also be linked as a subroutine of another main program. In this case, any thermodynamic property at saturation can be computed directly as a function of saturation pressure as well as saturation pressure is determined directly as a function of saturation temperature. Similarly, if a P-h pair is known in subcooled liquid or superheated vapor state, any thermodynamic or transport property can be computed directly. If a pressure-temperature (P-T) pair is known in subcooled liquid or superheated vapor state, enthalpy can also be computed directly. 3 - Restrictions on the complexity of the problem: The range of use and the maximum error for each correlation of the thermodynamic properties of heavy water are presented by Durmayaz (1997), and are also given in the statements in the beginning of each function subprogram for each correlationPrimary Subject

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30 Oct 2000; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1535.html; Country of input: International Atomic Energy Agency (IAEA); 6 refs.

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[en] 1 - Description of program or function: Format: DTF format and the structure is adopted from the MACKLIB-IV library. Number of groups: group library of reaction cross sections, gas production, kerma and DPA. Materials: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-Nat, Al-27, Si-28, P-31, S-Nat, Cl-Nat, Ar-36, Ar-38, Ar-40, K-Nat, Ca-Nat, Ti-Nat, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Zr-Nat, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Ag-107, Ag-109, Cd-Nat, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-12,2 Sn-124, Ba-130, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Hf-174, Hf-176, Hf-177, Hf- 178, Hf-179, Hf-180, Ta-181, W-Nat, W-182, W-183, W-184, W-186, Re-185, Re-187, Pb-Nat, Bi-209. Temperatures: T=293.6 K. Origin: GEFF-2 and GEPDL. RFL-2 is a group library of reaction cross sections, gas production, kerma and DPA based upon GEFF-2 and GEPDL - which are included in the package ZZ-GEFF-2-GENDF - and upon DECNET - included in the ZZ-DECNET-GENDF package (see below the description of these libraries). RFL-2 has been derived from them by the GENTORFL code (GENdf To RFL). Its primary use is to complete the neutron transport libraries in ANISN or FIDO format with data normally not present in the traditional files. It includes all GEFF-2 materials at T=293.6 K and σ0 = infinity; as qualifying point it gives 'delayed' kerma and 'delayed' gamma-ray production matrices, i.e. the energy release and the photons, respectively, generated by the decay of radioactive nuclei produced in the primary reactions; decay events that occur within 10000 seconds from the primary reaction are taken into account. The library includes many isotopes, since for each natural element included in GEFF-2 the decay of all component isotopes have been traced out. The library is in DTF format and the structure is adopted from the MACKLIB-IV library. The list of the included materials is given in one separate file. The position of the reactions in the RFL-2 matrix are given in the ZZRFL2.POS file. Source libraries GEFF-2: This P5 library for fusion neutronics calculations is based upon the 175 neutron, 42 photon VITAMIN-J group structure with the standard weighting function: Maxwellian (at the temperature to which the material is referenced) + 1/E + fission spectrum + 1/E + fusion peak + 1/E. It includes 93 materials - almost all from EFF-2 basic data; but Ag-107, Ag-109, natural Cd, the 6 Hf isotopes and the 4 W isotopes have been taken from JEF-2.2 - at 3 temperatures and 6 dilution values; 10 thermal groups are provided below 3 eV. Neutron cross sections and diffusion matrices, photon and gas production, kerma and DPA are given. GEPDL: GEPDL is a 42 group photon interaction P8 library based upon EPDL-90 to be used in conjunction with GEFF-2 for preparing neutron + photon coupled libraries by means of codes such as MATXSR - to produce data for TRANSX - or SMILER - an AMPX-77 module to produce AMPX Master Libraries from GENDF structured data. DECNET: DECNET is a library for fusion damage computations of 175 neutron +42 photon VITAMIN-J energy group with the standard weighting function: Maxwellian (at the temperature to which the material is referenced) + 1/E + fission spectrum + 1/E + fusion peak + 1/E; it includes neutron kerma and gamma-ray production data from radioactive nuclei at 3 temperatures with the same materials of ZZ-GEFF-2-GENDF (see below) from 1-H-1 to Bi-209, mostly taken from EFF-2 with some nuclides from JEF-2.2 - Ag-107, Ag-109, Cd, the 6 Hf isotopes and the 4 W isotopes; however the list of the materials disagrees with that of GEFF-2 in that all elemental nuclides have been split into the components isotopes to follow the respective decay chain; and not all materials of GEFF-2 produce nuclei which disintegrate within the assumed decay time of 10000 seconds delay from the primary events which produced the radioactive nucleus. The format of the library is GENDF 2 - Method of solution: NJOY has been used to process both GEFF-2 and GEPDL; since the processing activities lasted over a couple of years, less important materials were processed with the 91.13 version of the code and the 91.38 has been used to process all the remaining materials, including the whole GEPDL. Some very important modifications have been introduced in the code as far as the kerma computation is concerned, in order to solve problems due to physical inconsistencies of the data and non-standard formats of some materials, e.g. the lump reaction MT 10 - neutron diffusion in the continuum energy region - was not properly treated by the kerma module HEATR. The DECNET library has been produced by the DECKER code which has been developed for this purpose

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25 Apr 1997; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1545.html; Country of input: International Atomic Energy Agency (IAEA); 11 refs.

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ATOMIC DISPLACEMENTS, BARIUM 130, BISMUTH 209, CALCULATION METHODS, COMPUTER PROGRAM DOCUMENTATION, DILUTION, FISSION SPECTRA, GAMMA RADIATION, GROUP CONSTANTS, HAFNIUM 174, HYDROGEN 1, IRON 54, KERMA, MANGANESE 55, MOLYBDENUM 95, NEUTRON DIFFUSION EQUATION, NEUTRON TRANSPORT, NUCLEAR DAMAGE, NUCLEAR DATA COLLECTIONS, PHOTONS, RHENIUM 185, RHENIUM 187, SILVER 107, TANTALUM 181, TIN 115, TIN 117, TUNGSTEN 183, WEBSITES, WEIGHTING FUNCTIONS, Z CODES

ALKALINE EARTH ISOTOPES, ALPHA DECAY RADIOISOTOPES, BARIUM ISOTOPES, BETA DECAY RADIOISOTOPES, BETA-MINUS DECAY RADIOISOTOPES, BISMUTH ISOTOPES, BOSONS, COMPUTER CODES, CROSS SECTIONS, DAYS LIVING RADIOISOTOPES, DIFFERENTIAL EQUATIONS, DIFFUSION EQUATIONS, DOCUMENT TYPES, ELECTROMAGNETIC RADIATION, ELEMENTARY PARTICLES, EQUATIONS, EVEN-EVEN NUCLEI, EVEN-ODD NUCLEI, FUNCTIONS, HAFNIUM ISOTOPES, HEAVY NUCLEI, HYDROGEN ISOTOPES, INTERMEDIATE MASS NUCLEI, INTERNAL CONVERSION RADIOISOTOPES, IONIZING RADIATIONS, IRON ISOTOPES, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, LIGHT NUCLEI, MANGANESE ISOTOPES, MASSLESS PARTICLES, MOLYBDENUM ISOTOPES, NEUTRAL-PARTICLE TRANSPORT, NUCLEI, ODD-EVEN NUCLEI, PARTIAL DIFFERENTIAL EQUATIONS, PHYSICAL RADIATION EFFECTS, RADIATION EFFECTS, RADIATION TRANSPORT, RADIATIONS, RADIOISOTOPES, RHENIUM ISOTOPES, SECONDS LIVING RADIOISOTOPES, SILVER ISOTOPES, SPECTRA, STABLE ISOTOPES, TANTALUM ISOTOPES, TIN ISOTOPES, TUNGSTEN ISOTOPES, YEARS LIVING RADIOISOTOPES

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Turnbull, J.A.

Studsvik Scandpower AB, SE-611 82 Nykoeping (Sweden); Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Studsvik Scandpower AB, SE-611 82 Nykoeping (Sweden); Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description: The INTER-RAMP programme of work was designed to meet the following main objectives: - To establish the failure-safe operating limits of 20 standard-type, un-pressurized BWR fuel rods on overpower ramping at the burn-up levels of 10 and 20 MWd/kgU. The over-power ramping was to be performed at a fast ramp rate of about 4 kW/m/min with the preceding base irradiation performed with an idealised patterned cyclic variation in power levels designed to represent realistic variations in operating boiling water power reactors. - To study the influence of three main design parameters on fuel rod performance under power ramping, i.e. clad heat treatment (recrystallised anneal 'RX', vs cold work plus stress relief anneal 'SR'), pellet/clad diametral gap size, and fuel density. - To study the failure mechanism and associated phenomena. - To furnish data suitable for fuel modelling work. In order to meet these objectives, much effort was expended to record the individual fuel rod power history in great detail and also to determine the fuel rod changes, based on a pre-irradiation characterisation on a pellet by pellet basis, nondestructive examinations at stages throughout the irradiation history and selective detailed destructive examination of all rods following power ramping

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17 Nov 1997; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1555.html; Country of input: International Atomic Energy Agency (IAEA); 1 ref.

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Turnbull, J.A.

AEA Technology, Materials and Chemistry Group, 220.29 Harwell, Didcot, Oxfordshire OX11 0RA (United Kingdom); Health and Safety Executive, Bootle, Merseyside L20 3LZ (United Kingdom); Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AEA Technology, Materials and Chemistry Group, 220.29 Harwell, Didcot, Oxfordshire OX11 0RA (United Kingdom); Health and Safety Executive, Bootle, Merseyside L20 3LZ (United Kingdom); Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description: The series of experiments performed by AEA Technology at Harwell and released by the Health and Safety Executive investigates the relative importance of the effective gas atom diffusion coefficient, the capacity of grain boundary bubbles and the fuel grain size under conditions more nearly representative of a fault transient rather than of normal operation. Small samples of UO

_{2}fuel were obtained from CAGR fuel pins and annealed in a pre-defined high temperature, 1500 to 1900 C, for periods of between 2 and 40 hours. The release rate of 85-Kr was monitored continuously throughout each test. The rate at which the final temperature was attained varied from 0.1 to 8 degrees C/s in order to determine whether or not the behaviour of intergranular bubbles was sensitive to changes in temperature ramp rate within this range. The fuel used for these tests had a burn-up of ∼17 MWd/kgU and had two mean linear intercept grain sizes, 6 and 18 micronsPrimary Subject

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15 Mar 1999; [html]; Available on-line: http://www.nea.fr/abs/html/nea-1594.html; Country of input: International Atomic Energy Agency (IAEA); 4 refs.

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ACTINIDE COMPOUNDS, BETA DECAY RADIOISOTOPES, BETA-MINUS DECAY RADIOISOTOPES, CHALCOGENIDES, COMPUTER CODES, DOCUMENT TYPES, EVEN-ODD NUCLEI, FUEL ELEMENTS, HEAT TREATMENTS, HOURS LIVING RADIOISOTOPES, INTERMEDIATE MASS NUCLEI, ISOMERIC TRANSITION ISOTOPES, ISOTOPES, KRYPTON ISOTOPES, MICROSECONDS LIVING RADIOISOTOPES, MICROSTRUCTURE, NUCLEI, OXIDES, OXYGEN COMPOUNDS, PELLETS, RADIOISOTOPES, REACTOR COMPONENTS, SIZE, URANIUM COMPOUNDS, URANIUM OXIDES, YEARS LIVING RADIOISOTOPES

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Shapiro, A.B.; Vitez, G.A.

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] Description of program or function: TOPAZ2D is a two- dimensional implicit finite element computer codes for heat transfer analysis. TOPAZ2D can also be used to solve electrostatic and magnetostatic problems. The program solves for the steady-state or transient temperature or electrostatic and magnetostatic potential field on two-dimensional planar or axisymmetric geometries. Material properties may be temperature- or potential-dependent and either isotropic or orthotropic. A variety of time- and temperature- dependent boundary conditions can be specified including temperature, flux, convection, and radiation. By implementing the user subroutine feature, users can model chemical reaction kinetics and allow for any type of functional representation of boundary conditions and internal heat generation. The program can solve problems of diffuse and specular band radiation in an enclosure coupled with conduction in the material surrounding the enclosure. Additional features include thermal contact resistance across an interface, bulk fluids, phase change, and energy balances

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6 Nov 1996; [html]; Available on-line: http://www.nea.fr/abs/html/nesc9801.html; Country of input: International Atomic Energy Agency (IAEA); 4 refs.

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AXIAL SYMMETRY, BOUNDARY CONDITIONS, CHEMICAL REACTION KINETICS, CHEMICAL REACTIONS, COMPUTER PROGRAM DOCUMENTATION, CONVECTION, ENERGY BALANCE, FINITE ELEMENT METHOD, GEOMETRY, SOLIDS, STEADY-STATE CONDITIONS, T CODES, TEMPERATURE DEPENDENCE, THERMAL CONDUCTION, THERMAL STRESSES, TWO-DIMENSIONAL CALCULATIONS, WEBSITES

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Radiation Shielding Information Center, Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States); Fluor Daniel Northwest, Richland, Washington (United States); University of Washington, Seattle, Washington (United States); Organisation for Economic Co-Operation and Development, Nuclear Energy Agency - OECD/NEA, Le Seine Saint-Germain, 12 boulevard des Iles, F-92130 Issy-les-Moulineaux (France)

AbstractAbstract

[en] 1 - Description of program or function: ATHENA

_{2}D was written to simulate a hypothetical water reflood of a highly-damaged light water reactor (such as the Three-Mile-Island Unit-2 after meltdown, with a packed debris bed near the center of the core), but with insufficiently-borated reflood water. A recriticality transient may result because of the potentially more reactive debris bed. ATHENA-2D solves the transient multigroup neutron diffusion equations in (r,z) geometry. Executing in parallel with the transient neutronics, is a single-phase computational fluid dynamics (CFD) model, driven by multichannel thermal hydraulics based on detailed pin models. Numerous PV-Wave procedure files are included on the distribution media, useful for those who already have PV-Wave from Visual Numerics. These procedures are documented in the 'README' files included on the distribution CD. Some reactor lattice computer code such as WIMS-E, CCC-576/WIMSD4, or CCC-656/WIMSD5B is required for the creation of macroscopic cross section libraries, given pin-cell geometries. WIMS-E is a commercial product available from AEA Technologies, England, WIMS is not included on the ATHENA_{2}D distribution CD. Several auxiliary routines are included in the package. TFMAX: Utility that searches through ATHENA_{2}D binary output to find the maximum fuel temperature over space and time. POST_{V}EL: Utility that searches through ATHENA_{2}D binary output to find maximum scalar and flow field values (over space) and outputs normalization factors as a function of time. These results are used to correctly scale animations. CONVT: If executing ATHENA_{2}D on a PC under Windows, this utility converts one form of binary output (directly from ATHENA_{2}D) to another, which is readable by PV-Wave for Windows (PV-Wave is data animation and visualization software from Visual Numerics, Inc.) CALC_{M}TX: Post-processing utility for calculating the model coefficients for the calculation matrix. 2 - Methods: Both the neutronics and CFD equations are solved by successive-over relaxation (SOR) iteration. Chebychev extrapolation is available for the neutronics equations, although this option has not been thoroughly tested. The CFD equations use the semi-implicit method for pressure-linked equations (SIMPLE) iteration to achieve self-consistent solutions for mass, momentum and energy. Asymptotic acceleration of the pressure-correction equation (part of the SIMPLE iteration) is available. The one-dimensional thermal hydraulic fuel pin models employ fast tri-diagonal matrix inversion for a single time step. 3 - Restrictions on the complexity of the problem: ATHENA_{2}D assumes moderator boiling occurs before fuel remelt, as there is no mechanism to handle fuel remelt and relocation. However, even the most severe transients simulated during the course of the PhD work showed that this is reasonable. In a thermal neutron spectrum, time constants are longer than those in a fast spectrum. Given typical fuel piece dimensions, moderator boiling always occurred before the peak fuel temperature reached the melting point. Another limitation is the single (liquid) phase CFD model. Boiling is treated, but only insofar as to calculate local void fractions that feed back to the neutronics equations through local cross section interpolation based on reduced moderator density. The reactivity effects of voids being explicitly transported away from their point of origin is not treated. However, these effects are believed to be small as the introduction of voids tends to be a primary shutdown mechanism for these severe transients. Improved two-phase modeling would only affect the details of the shutdown phase of the transient, not the total energy release. Other Limitations: Reactivity feedback effects arising from any potential fluidized bed motion of fuel particles in the debris bed is not treated. No fuel motion is modeled. Blackbody (radiative) heat transfer is not modeled. Radiolytic gas bubble formation and its effect on reactivity is not modeled. There is no treatment of gamma or neutron thermalization heating in the moderator. There is no treatment of high-temperature, fuel-clad-water chemical interaction which is a potential hydrogen gas source term. There is no turbulence energy dissipation in the CFD model. The current library of heat transfer correlations is limited and could be improvedPrimary Subject

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17 Nov 1999; [html]; Available on-line: http://www.nea.fr/abs/html/psr-0431.html; Country of input: International Atomic Energy Agency (IAEA); 8 refs.

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A CODES, ASYMPTOTIC SOLUTIONS, COMPUTER PROGRAM DOCUMENTATION, COMPUTERIZED SIMULATION, CROSS SECTIONS, ENERGY LOSSES, EXTRAPOLATION, FLUIDIZED BEDS, FUEL PARTICLES, FUEL PINS, HEAT TRANSFER, MELTDOWN, MELTING POINTS, MULTIGROUP THEORY, NEUTRON DIFFUSION EQUATION, NUCLEAR DATA COLLECTIONS, ONE-DIMENSIONAL CALCULATIONS, RADIOLYSIS, REACTIVITY, REACTIVITY COEFFICIENTS, REACTOR LATTICES, REACTOR SAFETY, SHUTDOWN, THERMAL HYDRAULICS, THERMAL NEUTRONS, THERMALIZATION, THREE MILE ISLAND-2 REACTOR, TIME DEPENDENCE, VOID FRACTION, WEBSITES

ACCIDENTS, BARYONS, CHEMICAL RADIATION EFFECTS, CHEMICAL REACTIONS, COMPUTER CODES, DECOMPOSITION, DIFFERENTIAL EQUATIONS, DIFFUSION EQUATIONS, DOCUMENT TYPES, ELEMENTARY PARTICLES, ENERGY TRANSFER, ENRICHED URANIUM REACTORS, EQUATIONS, FERMIONS, FLUID MECHANICS, FUEL ELEMENTS, HADRONS, HYDRAULICS, LOSSES, MATHEMATICAL SOLUTIONS, MECHANICS, NEUTRON TRANSPORT THEORY, NEUTRONS, NUCLEONS, NUMERICAL SOLUTION, PARTIAL DIFFERENTIAL EQUATIONS, PHYSICAL PROPERTIES, POWER REACTORS, PWR TYPE REACTORS, RADIATION EFFECTS, REACTOR ACCIDENTS, REACTOR COMPONENTS, REACTORS, SAFETY, SIMULATION, SLOWING-DOWN, THERMAL REACTORS, THERMODYNAMIC PROPERTIES, TRANSITION TEMPERATURE, TRANSPORT THEORY, WATER COOLED REACTORS, WATER MODERATED REACTORS

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AbstractAbstract

[en] Verifying the treatment value(Monitor Unit) unnecessarily involves too many simple and repetitive calculation processes, that is, individual computation process using the data(PDD value, Scp Factor, SSD Factor, Tray Factor) on the data book. We intend to minimize the time required to check the Monitor Unit through computerized calculation. Using (multiplication), /(division), +(sum), if function, among others, which are present in the Excell program, MS office program, the Monitor Unit was obtainable through A/P value, Scp Factor and PDD value, Wedge Factor. From the verification of the computations of Monitor Unit for 60 patients previously treated, we were able to obtain an error rate of ±0.028 MU. Computerized calculation of the Monitor Unit could save the burden of Technologist.

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Journal of the Korean Society for Radiotherapeutic Technology; ISSN 1598-8449; ; v. 11(1); p. 28-32

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[en] The existence of a real integral solution of the generalised linear diffusion equation, in accordance with the original Feynman integral solution, is shown to be valid for affine potentials. The findings provide a new insight into the nature of the functional integral solution of the diffusion and the Schroedinger equation. (author)

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Singapore Journal of Physics; ISSN 0217-4251; ; v. 12(1); p. 95-101

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