Site response calculations for nuclear power plants
Wight, L.H.
1975-01-01
Six typical sites consisting of three soil profiles with average shear wave velocities of 800, 1800, and 5000 ft/sec as well as two soil depths of 200 and 400 ft were considered. Seismic input to these sites was a synthetic accelerogram applied at the surface and corresponding to a statistically representative response spectrum. The response of each of these six sites to this input was calculated with the SHAKE program. The results of these calculations are presented
Sensor response time calculation with no stationary signals from a Nuclear Power Plant
Vela, O.; Vallejo, I.
1998-01-01
Protection systems in a Nuclear Power Plant have to response in a specific time fixed by design requirements. This time includes the event detection (sensor delay) and the actuation time system. This time is obtained in refuel simulating the physics event, which trigger the protection system, with an electric signal and measuring the protection system actuation time. Nowadays sensor delay is calculated with noise analysis techniques. The signals are measured in Control Room during the normal operation of the Plant, decreasing both the cost in time and personal radioactive exposure. The noise analysis techniques require stationary signals but normally the data collected are mixed with process signals that are no stationary. This work shows the signals processing to avoid no-stationary components using conventional filters and new wavelets analysis. (Author) 2 refs
Global nuclear-structure calculations
Moeller, P.; Nix, J.R.
1990-01-01
The revival of interest in nuclear ground-state octupole deformations that occurred in the 1980's was stimulated by observations in 1980 of particularly large deviations between calculated and experimental masses in the Ra region, in a global calculation of nuclear ground-state masses. By minimizing the total potential energy with respect to octupole shape degrees of freedom in addition to ε 2 and ε 4 used originally, a vastly improved agreement between calculated and experimental masses was obtained. To study the global behavior and interrelationships between other nuclear properties, we calculate nuclear ground-state masses, spins, pairing gaps and Β-decay and half-lives and compare the results to experimental qualities. The calculations are based on the macroscopic-microscopic approach, with the microscopic contributions calculated in a folded-Yukawa single-particle potential
Abdou, M.A.; Gohar, Y.; Wright, R.Q.
1978-07-01
MACK-IV calculates nuclear response functions important to the neutronics analysis of nuclear and fusion systems. A central part of the code deals with the calculation of the nuclear response function for nuclear heating more commonly known as the kerma factor. Pointwise and multigroup neutron kerma factors, individual reactions, helium, hydrogen, and tritium production response functions are calculated from any basic nuclear data library in ENDF/B format. The program processes all reactions in the energy range of 0 to 20 MeV for fissionable and nonfissionable materials. The program also calculates the gamma production cross sections and the gamma production energy matrix. A built-in computational capability permits the code to calculate the cross sections in the resolved and unresolved resonance regions from resonance parameters in ENDF/B with an option for Doppler broadening. All energy pointwise and multigroup data calculated by the code can be punched, printed and/or written on tape files. Multigroup response functions (e.g., kerma factors, reaction cross sections, gas production, atomic displacements, etc.) can be outputted in the format of MACK-ACTIVITY-Table suitable for direct use with current neutron (and photon) transport codes
Reactor calculations and nuclear information
Lang, D.W.
1977-12-01
The relationship of sets of nuclear parameters and the macroscopic reactor quantities that can be calculated from them is examined. The framework of the study is similar to that of Usachev and Bobkov. The analysis is generalised and some properties required by common sense are demonstrated. The form of calculation permits revision of the parameter set. It is argued that any discrepancy between a calculation and measurement of a macroscopic quantity is more useful when applied directly to prediction of other macroscopic quantities than to revision of the parameter set. The mathematical technique outlined is seen to describe common engineering practice. (Author)
Techniques of nuclear structure calculations
Dyson, R.D.
1967-04-01
The quasiparticle method for identical particles interacting through pairing forces has been extended by others for use with systems of neutrons and protons. The method is to project isospin from separately considered neutron and proton quasiparticle wavefunctions. This is discussed in detail, and it seems that the projection may not be important. Therefore unprojected quasiparticle wavefunctions are tried with some success as a basis of states in which to diagonalize a realistic nuclear Hamiltonian. Brief unrelated calculations on nuclei of mass 19 and the SU(3) classification of states in the p-f shell are also presented. (author)
Nuclear data library in design calculation
Hirano, Go; Kosaka, Shinya
2006-01-01
In core design calculation, nuclear data takes part as multi group cross section library during the assembly calculation, which is the first stage of a core design calculation. This report summarizes the multi group cross section libraries used in assembly calculations and also presents the methods adopted for resonance and assembly calculation. (author)
Broyden's method in nuclear structure calculations
Baran, Andrzej; Bulgac, Aurel; Forbes, Michael McNeil; Hagen, Gaute; Nazarewicz, Witold; Schunck, Nicolas; Stoitsov, Mario V.
2008-01-01
Broyden's method, widely used in quantum chemistry electronic-structure calculations for the numerical solution of nonlinear equations in many variables, is applied in the context of the nuclear many-body problem. Examples include the unitary gas problem, the nuclear density functional theory with Skyrme functionals, and the nuclear coupled-cluster theory. The stability of the method, its ease of use, and its rapid convergence rates make Broyden's method a tool of choice for large-scale nuclear structure calculations
Nuclear winter - a calculative experiment
Aleksandrov, V.B.; Stenchikov, G.L.
1985-01-01
Using a hydrodynamic model of the Earth climate the climatic consequences following carbon dioxide concentration augmentation in the Earth atmosphere, effects of aerosol contamination and solar constant variation due to the use of nuclear weapon are studied. Results of studying the sensitivity of average annual climatic regime of the atmosphere and ocean general circulation to a sudde extremely strong, long-term change in optical properties of the air in the short-wave portion of the spectrum are discussed. These changes could be caused by contamination of the atmosphere with dust during a nuclear conflict and soot resulting from fires. It is shown, that after nuclear war according to practically any scenario, people who would survive the first blow will find themselves in conditions of a severe cold, darkness, absence of water, food and fuel under the effect of a powerful radiation, contaminants, diseases and under extreme pycological stress
Nuclear calculation of the thorium reactor
Hirakawa, Naohiro
1998-01-01
Even if for a reactor using thorium (and 233-U), its nuclear design calculation procedure is similar to the case using conventional 235-U, 238-U and plutonium. As nuclear composition varies with time on operation of nuclear reactor, calculation of its mean cross section should be conducted in details. At that time, one-group cross section obtained by integration over a whole of energy range is used for small member group. And, as the nuclear data for a base of its calculation is already prepared by JENDL3.2 and nuclear data library derived from it, the nuclear calculation of a nuclear reactor using thorium has no problem. From such a veiwpoint, IAEA has organized a coordinated research program of 'Potential of Th-based Fuel Cycles to Constrain Pu and to reduce Long-term Waste Toxicities' since 1996. All nations entering this program were regulated so as to institute by selecting a nuclear fuel cycle thinking better by each nation and to examine what cycle is expected by comparing their results. For a promise to conduct such neutral comparison, a comparison of bench mark calculations aiming at PWR was conducted to protect that the obtained results became different because of different calculation method and cross section adopted by each nation. Therefore, it was promoted by entrance of China, Germany, India, Israel, Japan, Korea, Russia and USA. The SWAT system developed by Tohoku University is used for its calculation code, by using which calculated results on the bench mark calculation at the fist and second stages and the nuclear reactor were reported. (G.K.)
Subcritical calculation of the nuclear material warehouse
Garcia M, T.; Mazon R, R.
2009-01-01
In this work the subcritical calculation of the nuclear material warehouse of the Reactor TRIGA Mark III labyrinth in the Mexico Nuclear Center is presented. During the adaptation of the nuclear warehouse (vault I), the fuel was temporarily changed to the warehouse (vault II) and it was also carried out the subcritical calculation for this temporary arrangement. The code used for the calculation of the effective multiplication factor, it was the Monte Carlo N-Particle Extended code known as MCNPX, developed by the National Laboratory of Los Alamos, for the particles transport. (Author)
Responsability of nuclear industry
Cadiz Deleito, J.C.
1985-01-01
Since the beginning of nuclear industry, civil responsibility with damages to the public health and properties was a critical problem, because the special conditions of this industry (nuclear accident, damages could be very high but probability of these events is very low). Legal precepts, universally accepted, in the first 60 years for all countries interested in nuclear energy are being revised, then 20 years of experience. The civil responsibility limited is being questioned and indemnities updated. (author)
Three dimensional diffusion calculations of nuclear reactors
Caspo, N.
1981-07-01
This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)
Nuclear friction calculated from nucleon currents
Pi, M.; Vinas, X.; Barranco, M.; La Rana, G.; Leray, S.; Lucas, R.; Ngo, C.; Tomasi, E.
1984-01-01
Nuclear friction can be connected to the number of nucleons exchanged between two interacting nuclei. The proximity scaling allows to reduce this problem to a calculation of the nucleon current between two semi infinite slabs of nuclear matter facing each other. In this paper we review the approximations and the results concerning this problem with a special emphasis on the physical ideas. Applications of nucleons currents to Fermi jets and to the calculation of a part of the imaginary potential are also discussed
TINTE. Nuclear calculation theory description report
Gerwin, H.; Scherer, W.; Lauer, A. [Forschungszentrum Juelich GmbH (DE). Institut fuer Energieforschung (IEF), Sicherheitsforschung und Reaktortechnik (IEF-6); Clifford, I. [Pebble Bed Modular Reactor (Pty) Ltd. (South Africa)
2010-01-15
The Time Dependent Neutronics and Temperatures (TINTE) code system deals with the nuclear and the thermal transient behaviour of the primary circuit of the High-temperature Gas-cooled Reactor (HTGR), taking into consideration the mutual feedback effects in twodimensional axisymmetric geometry. This document contains a complete description of the theoretical basis of the TINTE nuclear calculation, including the equations solved, solution methods and the nuclear data used in the solution. (orig.)
Křístková, Anežka; Malkin, Vladimir G. [Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dúbravská cesta 9, SK-84536 Bratislava (Slovakia); Komorovsky, Stanislav; Repisky, Michal [Centre for Theoretical and Computational Chemistry, University of Tromsø - The Arctic University of Norway, N-9037 Tromsø (Norway); Malkina, Olga L., E-mail: olga.malkin@savba.sk [Institute of Inorganic Chemistry, Slovak Academy of Sciences, Dúbravská cesta 9, SK-84536 Bratislava (Slovakia); Department of Inorganic Chemistry, Comenius University, Bratislava (Slovakia)
2015-03-21
In this work, we report on the development and implementation of a new scheme for efficient calculation of indirect nuclear spin-spin couplings in the framework of four-component matrix Dirac-Kohn-Sham approach termed matrix Dirac-Kohn-Sham restricted magnetic balance resolution of identity for J and K, which takes advantage of the previous restricted magnetic balance formalism and the density fitting approach for the rapid evaluation of density functional theory exchange-correlation response kernels. The new approach is aimed to speedup the bottleneck in the solution of the coupled perturbed equations: evaluation of the matrix elements of the kernel of the exchange-correlation potential. The performance of the new scheme has been tested on a representative set of indirect nuclear spin-spin couplings. The obtained results have been compared with the corresponding results of the reference method with traditional evaluation of the exchange-correlation kernel, i.e., without employing the fitted electron densities. Overall good agreement between both methods was observed, though the new approach tends to give values by about 4%-5% higher than the reference method. On the average, the solution of the coupled perturbed equations with the new scheme is about 8.5 times faster compared to the reference method.
Křístková, Anežka; Malkin, Vladimir G.; Komorovsky, Stanislav; Repisky, Michal; Malkina, Olga L.
2015-01-01
In this work, we report on the development and implementation of a new scheme for efficient calculation of indirect nuclear spin-spin couplings in the framework of four-component matrix Dirac-Kohn-Sham approach termed matrix Dirac-Kohn-Sham restricted magnetic balance resolution of identity for J and K, which takes advantage of the previous restricted magnetic balance formalism and the density fitting approach for the rapid evaluation of density functional theory exchange-correlation response kernels. The new approach is aimed to speedup the bottleneck in the solution of the coupled perturbed equations: evaluation of the matrix elements of the kernel of the exchange-correlation potential. The performance of the new scheme has been tested on a representative set of indirect nuclear spin-spin couplings. The obtained results have been compared with the corresponding results of the reference method with traditional evaluation of the exchange-correlation kernel, i.e., without employing the fitted electron densities. Overall good agreement between both methods was observed, though the new approach tends to give values by about 4%-5% higher than the reference method. On the average, the solution of the coupled perturbed equations with the new scheme is about 8.5 times faster compared to the reference method
Nuclear Data Processing for Reactor Physics Calculation
Suwoto; Zuhair; Pandiangan, Tumpal
2003-01-01
Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV
Concurrent algorithms for nuclear shell model calculations
Mackenzie, L.M.; Macleod, A.M.; Berry, D.J.; Whitehead, R.R.
1988-01-01
The calculation of nuclear properties has proved very successful for light nuclei, but is limited by the power of the present generation of computers. Starting with an analysis of current techniques, this paper discusses how these can be modified to map parallelism inherent in the mathematics onto appropriate parallel machines. A prototype dedicated multiprocessor for nuclear structure calculations, designed and constructed by the authors, is described and evaluated. The approach adopted is discussed in the context of a number of generically similar algorithms. (orig.)
Calculating the new global nuclear terrorism threat
2001-01-01
Experts from around the world are meeting at the IAEA on 29 October to 2 November at an international symposium on nuclear safeguards, verification, and security. A special session on 2 November focuses on the issue of combating nuclear terrorism. Although terrorists have never used a nuclear weapon, reports that some terrorist groups, particularly al-Qaeda, have attempted to acquire nuclear material is a cause of great concern. According to the IAEA, since 1993, there have been 175 cases of trafficking in nuclear material and 201 cases of trafficking in other radioactive sources (medical, industrial). However, only 18 of these cases have actually involved small amounts of highly enriched uranium or plutonium, the material needed to produce a nuclear bomb. IAEA experts judge the quantities involved to be insufficient to construct a nuclear explosive device. The IAEA experts have evaluated the risks for nuclear terrorism in these three categories: Nuclear facilities; Nuclear Material; Radioactive Sources. The IAEA is proposing a number of new initiatives, including strengthening border monitoring, helping States search for and dispose of orphan sources and strengthening the capabilities of the IAEA Emergency Response Centre to react to radiological emergencies following a terrorist attack. In the short term, the IAEA estimates that at least $30-$50 million annually will be needed to strengthen and expand its programs to meet this terrorist threat
Nuclear structure calculations for astrophysical applications
Moeller, P.; Kratz, K.L.
1992-01-01
Here we present calculated results on such diverse properties as nuclear energy levels, ground-state masses and shapes, β-decay properties and fission-barrier heights. Our approach to these calculations is to use a unified theoretical framework within which the above properties can all be studied. The results are obtained in the macroscopic-microscopic approach in which a microscopic nuclear-structure single-particle model with extensions is combined with a macroscopic model, such as the liquid drop model. In this model the total potential energy of the nucleus may be calculated as a function of shape. The maxima and minima in this function correspond to such features as the ground state, fission saddle points and shape-isomeric states. Various transition rate matrix elements are determined from wave-functions calculated in the single-particle model with pairing and other relevant residual interactions taken into account
Calculation models for a nuclear reactor
Tashanii, Ahmed Ali
2010-01-01
Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)
Hartree-Fock calculations of nuclear masses
Quentin, P.
1976-01-01
Hartree-Fock calculations pertaining to the determination of nuclear binding energies throughout the whole chart of nuclides are reviewed. Such an approach is compared with other methods. Main techniques in use are shortly presented. Advantages and drawbacks of these calculations are also discussed with a special emphasis on the extrapolation towards nuclei far from the stability valley. Finally, a discussion of some selected results from light to superheavy nuclei, is given [fr
Microscopic calculations of nuclear structure and nuclear correlations
Wiringa, R.B.
1992-01-01
A major goal in nuclear physics is to understand how nuclear structure comes about from the underlying interactions between nucleons. This requires modelling nuclei as collections of strongly interacting particles. Using realistic nucleon-nucleon potentials, supplemented with consistent three-nucleon potentials and two-body electroweak current operators, variational Monte Carlo methods are used to calculate nuclear ground-state properties, such as the binding energy, electromagnetic form factors, and momentum distributions. Other properties such as excited states and low-energy reactions are also calculable with these methods
Methodology of shielding calculation for nuclear reactors
Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo
1982-01-01
A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt
Fluidization calculation on nuclear fuel kernel coating
Sukarsono; Wardaya; Indra-Suryawan
1996-01-01
The fluidization of nuclear fuel kernel coating was calculated. The bottom of the reactor was in the from of cone on top of the cone there was a cylinder, the diameter of the cylinder for fluidization was 2 cm and at the upper part of the cylinder was 3 cm. Fluidization took place in the cone and the first cylinder. The maximum and the minimum velocity of the gas of varied kernel diameter, the porosity and bed height of varied stream gas velocity were calculated. The calculation was done by basic program
Parallel computational in nuclear group constant calculation
Su'ud, Zaki; Rustandi, Yaddi K.; Kurniadi, Rizal
2002-01-01
In this paper parallel computational method in nuclear group constant calculation using collision probability method will be discuss. The main focus is on the calculation of collision matrix which need large amount of computational time. The geometry treated here is concentric cylinder. The calculation of collision probability matrix is carried out using semi analytic method using Beckley Naylor Function. To accelerate computation speed some computer parallel used to solve the problem. We used LINUX based parallelization using PVM software with C or fortran language. While in windows based we used socket programming using DELPHI or C builder. The calculation results shows the important of optimal weight for each processor in case there area many type of processor speed
Calculation of Monte Carlo importance functions for use in nuclear-well logging calculations
Soran, P.D.; McKeon, D.C.; Booth, T.E.
1989-07-01
Importance sampling is essential to the timely solution of Monte Carlo nuclear-logging computer simulations. Achieving minimum variance (maximum precision) of a response in minimum computation time is one criteria for the choice of an importance function. Various methods for calculating importance functions will be presented, new methods investigated, and comparisons with porosity and density tools will be shown. 5 refs., 1 tab
Nuclear Research Center IRT reactor dynamics calculation
Aleman Fernandez, J.R.
1990-01-01
The main features of the code DIRT, for dynamical calculations are described in the paper. With the results obtained by the program, an analysis of the dynamic behaviour of the Research Reactor IRT of the Nuclear Research Center (CIN) is performed. Different transitories were considered such as variation of the system reactivity, coolant inlet temperature variation and also variations of the coolant velocity through the reactor core. 3 refs
Reactivity calculation with reduction of the nuclear power fluctuations
Suescun Diaz, Daniel; Senra Martinez, Aquilino
2009-01-01
A new formulation is presented in this paper for the calculation of reactivity, which is simpler than the formulation that uses the Laplace and Z transforms. A treatment is also made to reduce the intensity of the noise found in the nuclear power signal used in the calculation of reactivity. Two classes of different filters are used for that. This treatment is based on the fact that the reactivity can be written by using the compose Simpson's rule resulting in a sum of two convolution terms with response to the impulse that is characteristic of a linear system. The linear part is calculated by using the filter named finite impulse response filter (FIR). The non-linear part is calculated using the filter exponentially adjusted by the least squares method, which does not cause attenuation in the reactivity calculation.
Reactivity calculation with reduction of the nuclear power fluctuations
Suescun Diaz, Daniel [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914 RJ (Brazil)], E-mail: dsuescun@hotmail.com; Senra Martinez, Aquilino [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914 RJ (Brazil)
2009-05-15
A new formulation is presented in this paper for the calculation of reactivity, which is simpler than the formulation that uses the Laplace and Z transforms. A treatment is also made to reduce the intensity of the noise found in the nuclear power signal used in the calculation of reactivity. Two classes of different filters are used for that. This treatment is based on the fact that the reactivity can be written by using the compose Simpson's rule resulting in a sum of two convolution terms with response to the impulse that is characteristic of a linear system. The linear part is calculated by using the filter named finite impulse response filter (FIR). The non-linear part is calculated using the filter exponentially adjusted by the least squares method, which does not cause attenuation in the reactivity calculation.
Final disposal room structural response calculations
Stone, C.M.
1997-08-01
Finite element calculations have been performed to determine the structural response of waste-filled disposal rooms at the WIPP for a period of 10,000 years after emplacement of the waste. The calculations were performed to generate the porosity surface data for the final set of compliance calculations. The most recent reference data for the stratigraphy, waste characterization, gas generation potential, and nonlinear material response have been brought together for this final set of calculations
Bertsch, G.F.
1983-01-01
These lectures present the theory of the nuclear response in the Random Phase Approximation (RPA). In the first lecture, various relations are derived between densities and currents which give rise to the well-known sum rules. Then RPA is derived via the time-dependent Hartree theory. The various formulations of RPA are shown: the configuration space representation, the coordinate space representation, the Landau theory of infinite systems and the RPA for separable interactions constrained by consistency. The remarkable success of RPA in describing the collective density oscillations of closed shell nuclei is illustrated with a few examples. In the final lecture, the σtau response is discussed with the application of simple theoretical considerations to the empirical data. Finally, we point out several problems which remain in the response theory. (author)
Parquet theory in nuclear structure calculations
Bergli, Elise
2010-01-01
The thesis concerns a numerical implementation of the Parquet summation of diagrams within Green's functions theory applied to calculations of nuclear systems. The main motivation has been to investigate whether it is possible to develop this approach to a level comparable in accuracy and reliability to other ab initio nuclear structure methods. The Green's functions approach is theoretically well-established in many-body theory, but to our knowledge, no actual application to nuclear systems has been previously published. It has a number of desirable properties, foremost the gently scaling with system size compared to direct diagonalization and the closeness to experimentally accessible quantities. The main drawback is the numerical instabilities due to the pole structure of the one-particle propagator, leading to convergence difficulties. This issue is one of the main focal points of the work presented in this thesis, and strategies to improve the convergence properties are described and investigated. We have applied the method both to a simple model which can be solved by exact diagonalization and to the more realistic 4 He system. The results shows that our implementation is close to the exact solution in the simple model as long as the interaction strengths are small. As the number of particles increases, convergence is increasingly hard to obtain. In the 4 He case, we obtain results in the vicinity of the results from comparable approaches. The numerical in-stabilities in the current implementation still prevents the desired accuracy and stability necessary to achieve the current benchmark standards. (Author)
Low-energy calculations for nuclear photodisintegration
Deflorian S.
2016-01-01
Full Text Available In the Standard Solar Model a central role in the nucleosynthesis is played by reactions of the kind XZ1A11+XZ2A22→YZ1+Z2A1+A2+γ${}_{{Z_1}}^{{A_1}}{X_1} + {}_{{Z_2}}^{{A_2}}{X_2} \\to {}_{{Z_1} + {Z_2}}^{{A_1} + {A_2}}Y + \\gamma $, which enter the proton-proton chains. These reactions can also be studied through the inverse photodisintegration reaction. One option is to use the Lorentz Integral Transform approach, which transforms the continuum problem into a bound state-like one. A way to check the reliability of such methods is a direct calculation, for example using the Kohn Variational Principle to obtain the scattering wave function and then directly calculate the response function of the reaction.
MCNP capabilities for nuclear well logging calculations
Forster, R.A.; Little, R.C.; Briesmeister, J.F.; Hendricks, J.S.
1990-01-01
The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. This paper discusses how the general-purpose continuous-energy Monte Carlo code MCNP (Monte Carlo neutron photon), part of the LARTCS, provides a computational predictive capability for many applications of interest to the nuclear well logging community. The generalized three-dimensional geometry of MCNP is well suited for borehole-tool models. SABRINA, another component of the LARTCS, is a graphics code that can be used to interactively create a complex MCNP geometry. Users can define many source and tally characteristics with standard MCNP features. The time-dependent capability of the code is essential when modeling pulsed sources. Problems with neutrons, photons, and electrons as either single particle or coupled particles can be calculated with MCNP. The physics of neutron and photon transport and interactions is modeled in detail using the latest available cross-section data
LLNL nuclear data libraries used for fusion calculations
Howerton, R.J.
1984-01-01
The Physical Data Group of the Computational Physics Division of the Lawrence Livermore National Laboratory has as its principal responsibility the development and maintenance of those data that are related to nuclear reaction processes and are needed for Laboratory programs. Among these are the Magnetic Fusion Energy and the Inertial Confinement Fusion programs. To this end, we have developed and maintain a collection of data files or libraries. These include: files of experimental data of neutron induced reactions; an annotated bibliography of literature related to charged particle induced reactions with light nuclei; and four main libraries of evaluated data. We also maintain files of calculational constants developed from the evaluated libraries for use by Laboratory computer codes. The data used for fusion calculations are usually these calculational constants, but since they are derived by prescribed manipulation of evaluated data this discussion will describe the evaluated libraries
Qualification of γ-heating calculation in nuclear reactors
Ravaux, Simon
2013-01-01
During the last few years, the γ-heating issue has gained in stature, mainly for the safety of the 3. generation reactors in which a stainless steel reflector is inserted. The purpose of this work is the qualification of the needed tools for calculation of the γ-heating in the nuclear reactors. In a nuclear reactor, all the photons are directly or indirectly produced by the neutron-matter interactions. Thus, the first phase of this work is a critical analysis of the photon production data in the standard nuclear data library. New evaluations have been proposed to the next version of the JEFF library after that some omissions have been found. They have partly been accepted for JEFF-3.2. Two particle-transport codes are currently developed in the CEA: the deterministic code APOLLO2 and the Monte Carlo code TRIPOLI4. The second part of this work is the qualification of both these codes by interpreting an integral experiment called PERLE. The experimental set-up is made by a LWR pin assembly surrounded by a stainless steel reflector in which the γ-heating is measured by Thermo-luminescent Detector (TLD). A calculation scheme has been proposed for both APOLLO2 and TRIPOLI4 in order to calculate the TLD's responses. Comparisons between calculations and measurements have shown that TRIPOLI4 gives a satisfactory estimation of the γ-heating in the reflector. These discrepancies are within the experimental 1 σ uncertainty. Before the qualification, APOLLO2 has been previously validated against TRIPOLI4 reference calculation. This validation gives an estimation of the bias due to the deterministic approximations of the transport equation resolution. The qualification has shown that the discrepancies between APOLLO2 predictions and TLD's measurements are in the same range as experimental uncertainties. (author) [fr
Nuclear data preparation and discrete ordinates calculation
Carmignani, B.
1980-01-01
These lectures deal with the use of the GAM-GATHER and GAM-THERMOS chains for the calculation of lattice cross sections and within use of the discrete ordinates one dimensional ANISN code for the calculation of criticality and flux distribution of the cell and of the whole reactor. As an example the codes are applied to the calculation of a PWR. Results of different approximations are compared. (author)
Calculation of nuclear radius using alpha decay
Castro, R.B. de.
1988-01-01
Using a Quantum Theory approach for the Alpha-Decay process, a formula is deduced for determination of the nuclear radius of the s-state, that is, a nuclear model with a spherical shell. The hypothesis that it is possible to individualize the alpha particle and the daughter nucleus at the moment of the alpha particle emission is considered. In considered in these conditions, the treatment of a two body problem considered as point particles, repelling each other by Coulomb's Law. Using the new values of the fundamental physical constants, experimentally determinated, by substitution of their numerical values in the proposed, new values of nuclear radii are obtained. These values are compared with those found in the literature. (author) [pt
Chiral nucleon-nucleon forces in nuclear structure calculations
Coraggio L.
2016-01-01
Full Text Available Realistic nuclear potentials, derived within chiral perturbation theory, are a major breakthrough in modern nuclear structure theory, since they provide a direct link between nuclear physics and its underlying theory, namely the QCD. As a matter of fact, chiral potentials are tailored on the low-energy regime of nuclear structure physics, and chiral perturbation theory provides on the same footing two-nucleon forces as well as many-body ones. This feature fits well with modern advances in ab-initio methods and realistic shell-model. Here, we will review recent nuclear structure calculations, based on realistic chiral potentials, for both finite nuclei and infinite nuclear matter.
Nuclear excitations in the nuclear response theory
Nguyen Van Giai.
1983-01-01
The Random Phase Approximation is used to calculate the response of a nucleus to an external field. The method allows a full treatment of continuum effects. Effective interactions to be used are discussed, and the consistency between the mean field and the residual interaction is stressed. Some applications to excitation properties of spherical nuclei are shown
The MCEF code for nuclear evaporation and fission calculations
Deppman, A.; Pina, S.R. de; Likhachev, V.P.; Mesa, J.; Arruda-Neto, J.D.T.; Rodriguez, O.; Goncalves, M.
2001-11-01
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)
Effective operators in nuclear-structure calculations
Barrett, Bruce R
2005-01-01
A brief review of the history of the use of many-body perturbation theory to determine effective operators for shell-model calculations, i.e., for calculations in truncated model spaces, is given, starting with the ground-breaking work of Arima and Horie for electromagnetic moments. The problems encountered in utilizing this approach are discussed. New methods based on unitary-transformation approaches are introduced and analyzed. The old problems persist, but the new methods allow us to obtain a better insight into the nature of the physics involved in these processes
MACK/MACKLIB system for nuclear response functions
Abdou, M.A.; Gohar, Y.
1978-01-01
The MACK computer program calculates energy pointwise and multigroup nuclear response functions from basic nuclear data in ENDF/B format. The new version of the program, MACK-IV, incorporates major developments and improvements aimed at maximizing the utilization of available nuclear data and ensuring energy conservation in nuclear heating calculations. A new library, MACKLIB-IV, of nuclear response functions was generated in the CTR energy group structure of 171 neutron groups and 36 gamma groups. The library was prepared using MACK-IV, and ENDF/B-IV, and is suitable for fusion, fusion--fission hybrids, and fission applications. 3 figures, 4 tables
MACK/MACKLIB system for nuclear response functions
Abdou, M.A.; Gohar, Y.M.
1978-01-01
The MACK computer program calculates energy pointwise and multigroup nuclear response functions from basic nuclear data in ENDF/B format. The new version of the program MACK-IV, incorporates major developments and improvements aimed at maximizing the utilization of available nuclear data and ensuring energy conservation in nuclear heating calculations. A new library, MACKLIB-IV, of nuclear response functions was generated in the CTR energy group structure of 171 neutron groups and 36 gamma groups. The library was prepared using MACK-IV and ENDF/B-IV and is suitable for fusion, fusion-fission hydrids, and fission applications
Computations of nuclear response functions with MACK-IV
Abdou, M.A.; Gohar, Y.
1978-01-01
The MACK computer program calculates energy pointwise and multigroup nuclear response functions from basic nuclear data in ENDF/B format. The new version of the program, MACK-IV, incorporates major developments and improvements aimed at maximizing the utilization of available nuclear data and ensuring energy conservation in nuclear heating calculations. A new library, MACKLIB-IV, of nuclear response functions was generated in the CTR energy group structure of 171 neutron groups and 36 gamma groups. The library was prepared using MACK-IV and ENDF/B-IV and is suitable for fusion, fusion-fission hybrids, and fission applications
Computations of nuclear response functions with MACK-IV
Abdou, M A; Gohar, Y
1978-01-01
The MACK computer program calculates energy pointwise and multigroup nuclear response functions from basic nuclear data in ENDF/B format. The new version of the program, MACK-IV, incorporates major developments and improvements aimed at maximizing the utilization of available nuclear data and ensuring energy conservation in nuclear heating calculations. A new library, MACKLIB-IV, of nuclear response functions was generated in the CTR energy group structure of 171 neutron groups and 36 gamma groups. The library was prepared using MACK-IV and ENDF/B-IV and is suitable for fusion, fusion-fission hybrids, and fission applications.
MACK/MACKLIB system for nuclear response functions
Abdou, M.A.; Gohar, Y.M.
1978-03-15
The MACK computer program calculates energy pointwise and multigroup nuclear response functions from basic nuclear data in ENDF/B format. The new version of the program MACK-IV, incorporates major developments and improvements aimed at maximizing the utilization of available nuclear data and ensuring energy conservation in nuclear heating calculations. A new library, MACKLIB-IV, of nuclear response functions was generated in the CTR energy group structure of 171 neutron groups and 36 gamma groups. The library was prepared using MACK-IV and ENDF/B-IV and is suitable for fusion, fusion-fission hydrids, and fission applications.
Calculation of ex-core detector responses
Wouters, R. de; Haedens, M. [Tractebel Engineering, Brussels (Belgium); Baenst, H. de [Electrabel, Brussels (Belgium)
2005-07-01
The purpose of this work carried out by Tractebel Engineering, is to develop and validate a method for predicting the ex-core detector responses in the NPPs operated by Electrabel. Practical applications are: prediction of ex-core calibration coefficients for startup power ascension, replacement of xenon transients by theoretical predictions, and analysis of a Rod Drop Accident. The neutron diffusion program PANTHER calculates node-integrated fission sources which are combined with nodal importance representing the contribution of a neutron born in that node to the ex-core response. These importance are computed with the Monte Carlo program MCBEND in adjoint mode, with a model of the whole core at full power. Other core conditions are treated using sensitivities of the ex-core responses to water densities, computed with forward Monte Carlo. The Scaling Factors (SF), or ratios of the measured currents to the calculated response, have been established on a total of 550 in-core flux maps taken in four NPPs. The method has been applied to 15 startup transients, using the average SF obtained from previous cycles, and to 28 xenon transients, using the SF obtained from the in-core map immediately preceding the transient. The values of power (P) and axial offset (AOi) reconstructed with the theoretical calibration agree well with the measured values. The ex-core responses calculated during a rod drop transient have been successfully compared with available measurements, and with theoretical data obtained by alternative methods. In conclusion, the method is adequate for the practical applications previously listed. (authors)
Nuclear fuel cycle cost and cost calculation
Schmiedel, P.; Schricker, W.
1975-01-01
Four different methods of calculating the cost of the fuel cycle are explained, starting from the individual cost components with their specific input data. The results (for LWRs) are presented in tabular form and in the form of diagrams. (RB) [de
Calculation of nuclear parameters for some heavy isotopes
Corcuera, R.P.; Pinheiro, A.M.B.S.
1981-01-01
Some integrals are calculated using different weighting functions, the basic data come from two different nuclear data libraries, ENDF/B IV and ENDL/78. Significant discrepancies are found when are or the other lirary are used. (author) [pt
Semiclassical theory for the nuclear response function
Stroth, U.
1986-01-01
In the first part of this thesis it was demonstrated how on a semiclassical base a RPA theory is developed and applied to electron scattering. It was shown in which fields of nuclear physics this semiclassical theory can be applied and how it is to be understood. In this connection we dedicated an extensive discussion to the Fermi gas model. From the free response function we calculated the RPA response with a finite-range residual interaction which we completely antisymmetrize. In the second part of this thesis we studied with our theory (e,e') data for the separated response functions. (orig./HSI) [de
Methods for tornado frequency calculation of nuclear power plant
Liu Haibin; Li Lin
2012-01-01
In order to take probabilistic safety assessment of nuclear power plant tornado attack event, a method to calculate tornado frequency of nuclear power plant is introduced based on HAD 101/10 and NUREG/CR-4839 references. This method can consider history tornado frequency of the plant area, construction dimension, intensity various along with tornado path and area distribution and so on and calculate the frequency of different scale tornado. (authors)
Scoping calculations of power sources for nuclear electric propulsion
Difilippo, F.C.
1994-05-01
This technical memorandum describes models and calculational procedures to fully characterize the nuclear island of power sources for nuclear electric propulsion. Two computer codes were written: one for the gas-cooled NERVA derivative reactor and the other for liquid metal-cooled fuel pin reactors. These codes are going to be interfaced by NASA with the balance of plant in order to making scoping calculations for mission analysis
Thermodynamics of Rh nuclear spins calculated by exact diagonalization
Lefmann, K.; Ipsen, J.; Rasmussen, F.B.
2000-01-01
We have employed the method of exact diagonalization to obtain the full-energy spectrum of a cluster of 16 Rh nuclear spins, having dipolar and RK interactions between first and second nearest neighbours only. We have used this to calculate the nuclear spin entropy, and our results at both positi...
Amelina, M.E.; Arsent'ev, S.V.; Molchanov, A.S.
2009-01-01
The article considers the method of calculation of rates for insurance of civil responsibility for nuclear damage during transportation of nuclear materials, which can minimize the insurer's costs for this type of insurance in situation when there is no statistics available and it is not possible to calculate the insurance rate by the traditional means using the probability theory
Criticality calculation of the nuclear material warehouse of the ININ
Garcia, T.; Angeles, A.; Flores C, J.
2013-10-01
In this work the conditions of nuclear safety were determined as much in normal conditions as in the accident event of the nuclear fuel warehouse of the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ). The warehouse contains standard fuel elements Leu - 8.5/20, a control rod with follower of standard fuel type Leu - 8.5/20, fuel elements Leu - 30/20, and the reactor fuel Sur-100. To check the subcritical state of the warehouse the effective multiplication factor (keff) was calculated. The keff calculation was carried out with the code MCNPX. (Author)
Formation for the calculation of reactivity without nuclear power history
Suescun Diaz, Daniel; Senra Martinez, Aquilino; Carvalho Da Silva, Fernando
2007-01-01
This paper presents a new method for the solution of the inverse point kinetics equation. This method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. With the imposition of conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has very special characteristics, amongst which the possibility of using longer sampling period, and the possibility of restarting the calculation, after its interruption, allowing the calculation of reactivity in a non-continuous way. Beside that, the reactivity can be obtained independent of the nuclear power memory. (author)
Validation of calculational methods for nuclear criticality safety - approved 1975
Anon.
1977-01-01
The American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, N16.1-1975, states in 4.2.5: In the absence of directly applicable experimental measurements, the limits may be derived from calculations made by a method shown to be valid by comparison with experimental data, provided sufficient allowances are made for uncertainties in the data and in the calculations. There are many methods of calculation which vary widely in basis and form. Each has its place in the broad spectrum of problems encountered in the nuclear criticality safety field; however, the general procedure to be followed in establishing validity is common to all. The standard states the requirements for establishing the validity and area(s) of applicability of any calculational method used in assessing nuclear criticality safety
Calculation of fission gases internal pressure in nuclear fuel rods
Vasconcelos Santana, M. de.
1981-12-01
Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt
Calculation of integrated biological response in brachytherapy
Dale, Roger G.; Coles, Ian P.; Deehan, Charles; O'Donoghue, Joseph A.
1997-01-01
Purpose: To present analytical methods for calculating or estimating the integrated biological response in brachytherapy applications, and which allow for the presence of dose gradients. Methods and Materials: The approach uses linear-quadratic (LQ) formulations to identify an equivalent biologically effective dose (BED eq ) which, if applied to a specified tissue volume, would produce the same biological effect as that achieved by a given brachytherapy application. For simple geometrical cases, BED multiplying factors have been derived which allow the equivalent BED for tumors to be estimated from a single BED value calculated at a dose reference point. For more complex brachytherapy applications a voxel-by-voxel determination of the equivalent BED will be more accurate. Equations are derived which when incorporated into brachytherapy software would facilitate such a process. Results: At both high and low dose rates, the BEDs calculated at the dose reference point are shown to be lower than the true values by an amount which depends primarily on the magnitude of the prescribed dose; the BED multiplying factors are higher for smaller prescribed doses. The multiplying factors are less dependent on the assumed radiobiological parameters. In most clinical applications involving multiple sources, particularly those in multiplanar arrays, the multiplying factors are likely to be smaller than those derived here for single sources. The overall suggestion is that the radiobiological consequences of dose gradients in well-designed brachytherapy treatments, although important, may be less significant than is sometimes supposed. The modeling exercise also demonstrates that the integrated biological effect associated with fractionated high-dose-rate (FHDR) brachytherapy will usually be different from that for an 'equivalent' continuous low-dose-rate (CLDR) regime. For practical FHDR regimes involving relatively small numbers of fractions, the integrated biological effect to
The giant resonances in hot nuclei. Linear response calculations
Braghin, F.L.; Vautherin, D.; Abada, A.
1995-01-01
The isovector response function of hot nuclear matter is calculated using various effective Skyrme interactions. For Skyrme forces with a small effective mass the strength distribution is found to be nearly independent of temperature, and shows little collective effects. In contrast effective forces with an effective mass close to unity produce at zero temperature sizeable collective effects which disappear at temperatures of a few MeV. The relevance of these results for the saturation of the multiplicity of photons emitted by the giant dipole resonance in hot nuclei observed in recent experiments beyond T = 3 MeV is discussed. (authors). 12 refs., 3 figs
Lattice QCD Calculations in Nuclear Physics towards the Exascale
Joo, Balint
2017-01-01
The combination of algorithmic advances and new highly parallel computing architectures are enabling lattice QCD calculations to tackle ever more complex problems in nuclear physics. In this talk I will review some computational challenges that are encountered in large scale cold nuclear physics campaigns such as those in hadron spectroscopy calculations. I will discuss progress in addressing these with algorithmic improvements such as multi-grid solvers and software for recent hardware architectures such as GPUs and Intel Xeon Phi, Knights Landing. Finally, I will highlight some current topics for research and development as we head towards the Exascale era This material is funded by the U.S. Department of Energy, Office Of Science, Offices of Nuclear Physics, High Energy Physics and Advanced Scientific Computing Research, as well as the Office of Nuclear Physics under contract DE-AC05-06OR23177.
Nuclear data sets for reactor design calculations - approved 1975
Anon.
1978-01-01
This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types
American National Standard: nuclear data sets for reactor design calculations
1983-01-01
This standard identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include criteria for acceptance of evaluated nuclear data sets, criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types
American National Standard nuclear data sets for reactor design calculations
Anon.
1975-01-01
A standard is presented which identifies and describes the specifications for developing, preparing, and documenting nuclear data sets to be used in reactor design calculations. The specifications include (a) criteria for acceptance of evaluated nuclear data sets, (b) criteria for processing evaluated data and preparation of processed continuous data and averaged data sets, and (c) identification of specific evaluated, processed continuous, and averaged data sets which meet these criteria for specific reactor types
The cost of nuclear electricity: economic values and political calculations
Stauffer, T.
1985-01-01
The subject is covered in sections: introduction (monetary inflation; US-style rate-base formula; cost escalation); electricity generation costs (rate-base calculation formula; regulatory versus economic costs; inflationary case; cost-of-service rates versus inflation; first year electricity costs); rate shock (A. comparison with oil; B. nuclear case; C. comparison with coal/nuclear system; vintaged electricity costs versus growth and inflation); conclusions. (U.K.)
Multiscale fluctuations in nuclear response
Lacroix, D.; Chomaz, Ph.
1999-01-01
The nuclear collective response is investigated in the framework of a doorway picture in which the spreading width of the collective emotion is described as a coupling to more and more complex configurations. It is shown that this coupling induces fluctuations of the observed strength. In the case of a hierarchy of overlapping decay channels, Ericson fluctuations are observed at different scales. Methods for extracting these scales and the related lifetimes are discussed. Finally, it is shown that the coupling of different states at one level of complexity to some common decay channels at the next level, may produce interference-like patterns in the nuclear response. This quantum effect leads to anew type of fluctuations with a typical width related to the level spacing. (author)
Multiscale fluctuations in nuclear response
Lacroix, D.; Chomaz, Ph
1999-01-01
The nuclear collective response is investigated in the framework of a doorway picture in which the spreading width of the collective emotion is described as a coupling to more and more complex configurations. It is shown that this coupling induces fluctuations of the observed strength. In the case of a hierarchy of overlapping decay channels, Ericson fluctuations are observed at different scales. Methods for extracting these scales and the related lifetimes are discussed. Finally, it is shown that the coupling of different states at one level of complexity to some common decay channels at the next level, may produce interference-like patterns in the nuclear response. This quantum effect leads to anew type of fluctuations with a typical width related to the level spacing. (author) 25 refs.
Evaluated nuclear data file libraries use in nuclear-physical calculations
Gritsaj, O.O.; Kalach, N.Yi.; Kal'chenko, O.Yi.; Kolotij, V.V.; Vlasov, M.F.
1994-01-01
The necessity of nuclear updated usage is founded for neutron experiment modeling calculations, for preparation of suitable data for reactor calculations and for other applications that account of detail energetic structure of cross section is required. The scheme of system to coordinate the work to collect and to prepare evaluated nuclear data on an international scale is presented. Main updated and recommended nuclear data libraries and associated computer programs are reviewed. Total neutron cross sections for 28 energetic groups calculated on the base of natural mixture iron isotopes evaluated nuclear data file (BROND-2, 1991) have been compared with BNAB-78 data. (author). 7 refs., 1 tab., 4 figs
Naito, Yoshitaka; Ihara, Hitoshi; Katakura, Jun-ichi; Hara, Toshiharu.
1986-08-01
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)
Nuclear data for actinide production and depletion calculations
Benjamin, R.W.
1978-01-01
The status of nuclear cross section data required for actinide depletion calculations in thermal reactors is summarized, and recommendations are made for future work. The primary fertile and fissile nuclides ( 232 Th, 233 U, 235 U, 238 U, and 239 Pu) are not reviewed. Nuclear data for the transactinium mass region are, with few exceptions, reasonably complete and adequate for current thermal-reactor depletion calculations. There is a real need, however, for well-documented reactor production studies to use as benchmarks for data testing. 3 figures, 6 tables
EMPIRE-II statistical model code for nuclear reaction calculations
Herman, M [International Atomic Energy Agency, Vienna (Austria)
2001-12-15
EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)
Society response to nuclear energy
Santamaria, N. C.
2007-01-01
Energy demand in the world is growing increasingly, among other factors due to economic development. Every way of producing electricity has got their own drawbacks and has implicit environmental impact. Among all the energy sources, nuclear energy is the most polemic because of the way it is presented by the mass media. This aspect provokes controversy to occidental societies which reject this kind of energy with arguments normally based on a wrong and insufficient knowledge of the matter. The antinuclear discourse, promoted late in the seventies, has gone deeply into the collective social unconscious and has undermined public acceptance of nuclear energy due to the fact, deeply exploited by antinuclear groups, of linking nuclear energy with the atomic bombing of Hiroshima and Nagasaki. In this sense, it is important to mention that in Japan there was a profound resentment and opposition to nuclear energy, because the memory of the nuclear bombings was permanently alive. However when the Japanese government told its people that this energy was necessary to boost their industrial development, Japanese citizens in an unprecedented attitude of patriotism overcame their most antagonist feelings, in order to contribute to the industrial development of their country. The result was that most of them voted in favour. Presently Japan gets 30% of its energy by means of 56 nuclear power plants and 1 more is under construction. Antinuclear groups took as their best emblem the accident of Chernobyl to justify their opposition to the nuclear power plants. The manipulation of this accident has been one of the most shameful in the nuclear history. It is widely known among the experts that the reactor used in Chernobyl was a type of military plutonium converter with a positive temperature reactivity coefficient, which made very dangerous its functioning. Any nuclear regulatory commission in democratic and responsible countries would have never authorized the use of this reactor
Nuclear data and multigroup methods in fast reactor calculations
Gur, Y.
1975-03-01
The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)
Therapeutic Applications of Monte Carlo Calculations in Nuclear Medicine
Sgouros, George
2003-01-01
This book examines the applications of Monte Carlo (MC) calculations in therapeutic nuclear medicine, from basic principles to computer implementations of software packages and their applications in radiation dosimetry and treatment planning. It is written for nuclear medicine physicists and physicians as well as radiation oncologists, and can serve as a supplementary text for medical imaging, radiation dosimetry and nuclear engineering graduate courses in science, medical and engineering faculties. With chapters is written by recognised authorities in that particular field, the book covers the entire range of MC applications in therapeutic medical and health physics, from its use in imaging prior to therapy to dose distribution modelling targeted radiotherapy. The contributions discuss the fundamental concepts of radiation dosimetry, radiobiological aspects of targeted radionuclide therapy and the various components and steps required for implementing a dose calculation and treatment planning methodology in ...
Development of nuclear models for higher energy calculations
Bozoian, M.; Siciliano, E.R.; Smith, R.D.
1988-01-01
Two nuclear models for higher energy calculations have been developed in the regions of high and low energy transfer, respectively. In the former, a relativistic hybrid-type preequilibrium model is compared with data ranging from 60 to 800 MeV. Also, the GNASH exciton preequilibrium-model code with higher energy improvements is compared with data at 200 and 318 MeV. In the region of low energy transfer, nucleon-nucleus scattering is predominately a direct reaction involving quasi-elastic collisions with one or more target nucleons. We discuss various aspects of quasi-elastic scattering which are important in understanding features of cross sections and spin observables. These include (1) contributions from multi-step processes; (2) damping of the continuum response from 2p-2h excitations; (3) the ''optimal'' choice of frame in which to evaluate the nucleon-nucleon amplitudes; and (4) the effect of optical and spin-orbit distortions, which are included in a model based on the RPA the DWIA and the eikonal approximation. 33 refs., 15 figs
theory and calculation of the design of nuclear reactor
Refaat, R.A.
1994-01-01
For the sake of formation of a complete general code for nuclear power reactor design, this thesis deals with a great part of this code. the code links the solution of the neutron integral transport equation by the multigroup treatment (76 energy groups) for the calculation of the reactor cell parameters by the fuel management program that solves the neutron diffusion equation inside a large number of nuclear fuel assemblies. the lattice cell code is modified to accommodate the calculation of lattice cell parameters for more than one enrichment ( one after the other). it is also modified to calculate the burn up parameters using unequal time steps. these two modifications are complicated but necessary for the link between the cell program and fuel management program. the comparison between the results of the fitted cross sections and that given by the cell calculations shows the necessity of using the cell code cross sections. this is also necessary for the sake of generality for any type of reactors. the comparison for the fuel management calculation depending on fitted data and that depending on cell calculation data insures the necessity for using the cell data i.e. insures the necessity of linking the cell calculation program by the fuel management program
Uncertainty quantification in lattice QCD calculations for nuclear physics
Beane, Silas R. [Univ. of Washington, Seattle, WA (United States); Detmold, William [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Orginos, Kostas [College of William and Mary, Williamsburg, VA (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Savage, Martin J. [Institute for Nuclear Theory, Seattle, WA (United States)
2015-02-05
The numerical technique of Lattice QCD holds the promise of connecting the nuclear forces, nuclei, the spectrum and structure of hadrons, and the properties of matter under extreme conditions with the underlying theory of the strong interactions, quantum chromodynamics. A distinguishing, and thus far unique, feature of this formulation is that all of the associated uncertainties, both statistical and systematic can, in principle, be systematically reduced to any desired precision with sufficient computational and human resources. As a result, we review the sources of uncertainty inherent in Lattice QCD calculations for nuclear physics, and discuss how each is quantified in current efforts.
Dynamical calculations of nuclear fission and heavy-ion reactions
Nix, J.R.; Sierk, A.J.
1984-01-01
With the goal of determining the magnitude and mechanism of nuclear dissipation from comparisons of predictions with experimental data, we describe recent calculations in a unified macroscopic-microscopic approach to large-amplitude collective nuclear motion such as occurs in fission and heavy-ion reactions. We describe the time dependence of the distribution function in phase space of collective coordinates and momenta by a generalized Fokker-Planck equation. The nuclear potential energy of deformation is calculated as the sum of repulsive Coulomb and centrifugal energies and an attractive Yukawa-plus-exponential potential, the inertia tensor is calculated for a superposition of rigid-body rotation and incompressible, nearly irrotational flow by use of the Werner-Wheeler method, and the dissipation ensor that describes the conversion of collective energy into single-particle excitation energy is calculated for two prototype mechanisms that represent opposite extremes of large and small dissipation. We solve the generalized Hamilton equations of motion for the first moments of the distribution function to obtain the mean translational fission-fragment kinetic energy and mass of a third fragment that sometimes forms between the two end fragments, as well as dynamical thresholds, capture cross sections, and ternary events in heavy-ion reactions. 33 references
Benchmark calculation of nuclear design code for HCLWR
Suzuki, Katsuo; Saji, Etsuro; Gakuhari, Kazuhiko; Akie, Hiroshi; Takano, Hideki; Ishiguro, Yukio.
1986-01-01
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)
COMRAD96, Nuclear Fuel Burnup and Depletion Calculation System
Suyama, K.; Masukawa, F.; Ido, M.; Enomoto, M.; Takyu, S.; Hara, T.
2002-01-01
1 - Description of program or function: Burn-up calculation of nuclear fuel. 2 - Methods: Matrix exponential method, Bateman Equation. 3 - Restrictions on the complexity of the problem: a) One-grouped cross section library should be prepared for the fuel system to be analyzed using UNITBURN. However, UNITBURN is not available now for UNIX systems. b) Gamma ray spectrometry calculation will fail using the attached piflib routine. This problem has already been rectified in the internal version. 4 - Typical running time: Two minutes for standard burn-up calculation on Sun ULTRA 30. 5 - Unusual features - a) Selection of Matrix exponential method, or Bateman Equation. b) JDDL, a detailed decay chain data based on ENSDF. 6 - Related or auxiliary programs: UNITBURN: Burnup calculation code unit cell system
Nuclear law: organization and responsibilities
Ha Vinh Phuong.
1986-01-01
The paper emphasizes the importance of a special legislation insuring the governmental control of nuclear applications and other related activities. This legislation must establish the authority in charge for the development of peaceful applications of nuclear energy and the specialized body legally competent to insure an independent control of nuclear activities, it must define the principles and the conditions for licensing nuclear activities insuring the physical protection of nuclear materials and installations and must establish the specific rules for nuclear liability in the case of a nuclear accident. A list of IAEA publications related to the safety of nuclear power plants is included
Calculation of heat generation due to nuclear radiation in nuclear reactors
Torres, L.M.R.; Gomes, I.C.; Maiorino, J.R.
1986-01-01
The study is performed for caculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN code, that solves the one-dimensional transport equation using the discrete ordinate method, to include nuclear heating calculations. Tests of the implemented modifications were performed in problems of nuclear heating due to radiation energy deposition in a fusion reactor. (Author) [pt
The role of desk calculators in nuclear data evaluation
Motta, M.
1980-01-01
The performances of the modern Desk Calculators are more and more increasing. Consequently, the best feature for the definition of a Desk Calculator seems to be prices and volume occupation. The interactive operating mode and the low installation and maintaining costs make the use of these computing machines very likely and economical. A list of tasks which are profitably performed in nuclear data preparation are presented here. A lot of practical applications are given through the formulas and the list of codes written in BASIC language, which is commonly adopted for these small computers. (author)
SIMCRI: a simple computer code for calculating nuclear criticality parameters
Nakamaru, Shou-ichi; Sugawara, Nobuhiko; Naito, Yoshitaka; Katakura, Jun-ichi; Okuno, Hiroshi.
1986-03-01
This is a user's manual for a simple criticality calculation code SIMCRI. The code has been developed to facilitate criticality calculation on a single unit of nuclear fuel. SIMCRI makes an extensive survey with a little computing time. Cross section library MGCL for SIMCRI is the same one for the Monte Carlo criticality code KENOIV; it is, therefore, easy to compare the results of the two codes. SIMCRI solves eigenvalue problems and fixed source problems based on the one space point B 1 equation. The results include infinite and effective multiplication factor, critical buckling, migration area, diffusion coefficient and so on. SIMCRI is comprised in the criticality safety evaluation code system JACS. (author)
Calculation of the viscosity of nuclear waste glass systems
Shah, R.; Behrman, E.C.; Oksoy, D.
1990-01-01
Viscosity is one of the most important processing parameters and one of the most difficult to calculate theoretically, particularly for multicomponent systems like nuclear waste glasses. Here, the authors propose a semi-empirical approach based on the Fulcher equation, involving identification of key variables, for which coefficients are then determined by regression analysis. Results are presented for two glass systems, and compared to results of previous workers and to experiment. The authors also sketch a first-order statistical mechanical perturbation theory calculation for the effects on viscosity of a change in composition of the melt
Cluster monte carlo method for nuclear criticality safety calculation
Pei Lucheng
1984-01-01
One of the most important applications of the Monte Carlo method is the calculation of the nuclear criticality safety. The fair source game problem was presented at almost the same time as the Monte Carlo method was applied to calculating the nuclear criticality safety. The source iteration cost may be reduced as much as possible or no need for any source iteration. This kind of problems all belongs to the fair source game prolems, among which, the optimal source game is without any source iteration. Although the single neutron Monte Carlo method solved the problem without the source iteration, there is still quite an apparent shortcoming in it, that is, it solves the problem without the source iteration only in the asymptotic sense. In this work, a new Monte Carlo method called the cluster Monte Carlo method is given to solve the problem further
Axial power distribution calculation using a neural network in the nuclear reactor core
Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1997-12-31
This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)
Axial power distribution calculation using a neural network in the nuclear reactor core
Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)
1998-12-31
This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)
Hauser*5, a computer code to calculate nuclear cross sections
Mann, F.M.
1979-07-01
HAUSER*5 is a computer code that uses the statistical (Hauser-Feshbach) model, the pre-equilibrium model, and a statistical model of direct reactions to predict nuclear cross sections. The code is unrestricted as to particle type, includes fission and capture, makes width-fluctuation corrections, and performs three-body calculations - all in minimum computer time. Transmission coefficients can be generated internally or supplied externally. This report describes equations used, necessary input, and resulting output. 2 figures, 4 tables
Covariance matrices for nuclear cross sections derived from nuclear model calculations
Smith, D. L.
2005-01-01
The growing need for covariance information to accompany the evaluated cross section data libraries utilized in contemporary nuclear applications is spurring the development of new methods to provide this information. Many of the current general purpose libraries of evaluated nuclear data used in applications are derived either almost entirely from nuclear model calculations or from nuclear model calculations benchmarked by available experimental data. Consequently, a consistent method for generating covariance information under these circumstances is required. This report discusses a new approach to producing covariance matrices for cross sections calculated using nuclear models. The present method involves establishing uncertainty information for the underlying parameters of nuclear models used in the calculations and then propagating these uncertainties through to the derived cross sections and related nuclear quantities by means of a Monte Carlo technique rather than the more conventional matrix error propagation approach used in some alternative methods. The formalism to be used in such analyses is discussed in this report along with various issues and caveats that need to be considered in order to proceed with a practical implementation of the methodology
Semi-classical calculation of the spin-isospin response functions
Chanfray, G.
1987-03-01
We present a semi-classical calculation of the nuclear response functions beyond the Thomas-Fermi approximation. We apply our formalism to the spin-isospin responses and show that the surface peaked h/2π corrections considerably decrease the ratio longitudinal/transverse as obtained through hadronic probes
Minaret, a deterministic neutron transport solver for nuclear core calculations
Moller, J-Y.; Lautard, J-J.
2011-01-01
We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)
Minaret, a deterministic neutron transport solver for nuclear core calculations
Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)
2011-07-01
We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)
Advanced nuclear data for radiation-damage calculations
MacFarlane, R.E.; Foster, D.G. Jr.
1983-01-01
Accurate calculations of atomic displacement damage in materials exposed to neutrons require detailed spectra for primary recoil nuclei. Such data are not available from direct experimental measurements. Moreover, they cannot always be computed accurately starting from evaluated nuclear data libraries such as ENDF/B-V that were developed primarily for neutron transport applications, because these libraries lack detailed energy-and-angle distributions for outgoing charged particles. Fortunately, a new generation of nuclear model codes is now available that can be used to fill in the missing spectra. One example is the preequilibrium statistical-model code GNASH. For heating and damage applications, a supplementary code called RECOIL has been developed. RECOIL uses detailed reaction data from GNASH, together with angular distributions based on Kalbach-Mann systematics to compute the energy and angle distributions of recoil nuclei. The energy-angle distributions for recoil nuclei and outgoing particles are written out in the new ENDF/B File 6 format. The result is a complete set of nuclear data that can be used to calculate displacement-energy production, heat production, gas production, transmutation, and activation. Sample results for iron are given and compared to the results of conventional damage models such as those used in NJOY
Historical trend of nuclear matter calculation and its recent developments
Kohno, Michio
2006-01-01
He guide line to understand nuclear properties on the basis of nuclear force was started in the 1950's by the Brueckner theory. The theory established the fundamental framework to formulate the picture to consider both the two nucleon and tensor correlations as well as Pauli effect inside the nuclei. In the 1960's the theory was developed to obtain ground state energy on the perturbation many-body theory. The growth and refinement of the Brueckner theory in the 1970's and after are overviewed and the computer code developments in the 1980's are mentioned. Concerning the many-body correlation problem Italian group has calculated up to three-body correlations in the Brueckner theory. At present, effective interaction nuclear theory is coming into a new level and actively studied by the introduction of low momentum interaction based on the renormalization group theory, by full application of the coupled cluster method, by the application of Skyrme Hartree-Fock method in wide range and by the reconsideration of the energy density functional method in relation to the relativistic mean field method. Owing to the recent remarkable progress of computers, calculations which were impossible to be executed in old days are now done rather easily. (S. Funahashi)
The status of nuclear data for transmutation calculations
Wilson, W.B.; England, T.R.; MacFarlane, R.E.; Muir, D.W.; Young, P.G.
1995-01-01
At this point, the accurate description of transmutation products in a radiation environment is more a nuclear data problem than a code development effort. We have used versions of the CINDER code for over three decades to describe the transmutation of nuclear reactor fuels in radiation environments. The need for the accurate description of reactor neutron-absorption, decay-power, and decay-spectra properties have driven many AEC, ERDA, and DOE supported nuclear data development efforts in this period. The level of cross-section, decay, and fission-yield data has evolved from rudimentary to a comprehensive ENDF/B-VI library permitting great precision in reactor calculations. The precision of the data supporting reactor simulations provides a sturdy foundation for the data base required for the wide range of transmutation problems currently studied. However, such reactor problems are typically limited to neutron energies below 10 MeV or so; reaction and decay data are required for actinides of, say, 90 ≤ Z ≤ 96 neutron-rich fission products of 22 ≤ Z ≤ 72. The expansion into reactor structural materials and fusion systems extends these ranges in energy and Z somewhat. The library of nuclear data, constantly growing in breadth and quality with international cooperation, is now described in the following table
Ab Initio Calculations Of Nuclear Reactions And Exotic Nuclei
Quaglioni, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2014-05-05
Our ultimate goal is to develop a fundamental theory and efficient computational tools to describe dynamic processes between nuclei and to use such tools toward supporting several DOE milestones by: 1) performing predictive calculations of difficult-to-measure landmark reactions for nuclear astrophysics, such as those driving the neutrino signature of our sun; 2) improving our understanding of the structure of nuclei near the neutron drip line, which will be the focus of the DOE’s Facility for Rare Isotope Beams (FRIB) being constructed at Michigan State University; but also 3) helping to reveal the true nature of the nuclear force. Furthermore, these theoretical developments will support plasma diagnostic efforts at facilities dedicated to the development of terrestrial fusion energy.
Calculation simulation of equivalent irradiation swelling for dispersion nuclear fuel
Cai Wei; Zhao Yunmei; Gong Xin; Ding Shurong; Huo Yongzhong
2015-01-01
The dispersion nuclear fuel was regarded as a kind of special particle composites. Assuming that the fuel particles are periodically distributed in the dispersion nuclear fuel meat, the finite element model to calculate its equivalent irradiation swelling was developed with the method of computational micro-mechanics. Considering irradiation swelling in the fuel particles and the irradiation hardening effect in the metal matrix, the stress update algorithms were established respectively for the fuel particles and metal matrix. The corresponding user subroutines were programmed, and the finite element simulation of equivalent irradiation swelling for the fuel meat was performed in Abaqus. The effects of the particle size and volume fraction on the equivalent irradiation swelling were investigated, and the fitting formula of equivalent irradiation swelling was obtained. The results indicate that the main factors to influence equivalent irradiation swelling of the fuel meat are the irradiation swelling and volume fraction of fuel particles. (authors)
Microscopic nuclear structure calculations with modern meson-exchange potentials
Hjort-Jensen, M.; Osnes, E.; Muether, H.; Schmid, K.W.; Kuo, T.T.S.
1990-07-01
The report presents the results of microscopic nuclear shell-model calculations using three different nucleon-nucleon potentials. These are the phenomenological Reid-Soft-Core potential and the meson-exchange potentials of the Paris and the Bonn groups. It is found that the Bonn potential yields sd-shell matrix elements which are more attractive than those obtained with the Reid or the Paris potentials. The harmonic-oscillator matrix elements of the Bonn potential are also in better agreement with the empirically derived matrix elements of Wildenthal. The implications are discussed. 27 refs., 4 figs., 1 tab
Nuclear calculation methods for light water moderated reactors
Hicks, D.
1961-02-01
This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)
Design and structural calculation of nuclear power plant mechanical components
Amaral, J.A.R. do
1986-01-01
The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt
Calculation of dynamic hydraulic forces in nuclear plant piping systems
Choi, D.K.
1982-01-01
A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)
Nuclear model calculations on cyclotron production of {sup 51}Cr
Kakavand, Tayeb [Imam Khomeini International Univ., Qazvin (Iran, Islamic Republic of). Dept. of Physics; Aboudzadeh, Mohammadreza [Nuclear Science and Technology Research Institute/AEOI, Karaj (Iran, Islamic Republic of). Agricultural, Medical and Industrial Research School; Farahani, Zahra; Eslami, Mohammad [Zanjan Univ. (Iran, Islamic Republic of). Dept. of Physics
2015-12-15
{sup 51}Cr (T{sub 1/2} = 27.7 d), which decays via electron capture (100 %) with 320 keV gamma emission (9.8 %), is a radionuclide with still a large application in biological studies. In this work, ALICE/ASH and TALYS nuclear model codes along with some adjustments are used to calculate the excitation functions for proton, deuteron, α-particle and neutron induced on various targets leading to the production of {sup 51}Cr radioisotope. The production yields of {sup 51}Cr from various reactions are determined using the excitation function calculations and stopping power data. The results are compared with corresponding experimental data and discussed from point of view of feasibility.
Subcriticality calculation in nuclear reactors with external neutron sources
Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br
2007-07-01
The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)
Subcriticality calculation in nuclear reactors with external neutron sources
Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da
2007-01-01
The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)
Calculations of hydrogen detonations in nuclear containments by the random choice method
Delichatsios, M.A.; Genadry, M.B.
1983-01-01
Computer codes were developed for the prediction of pressure histories at different points of a nuclear containment wall due to postulated internal hydrogen detonations. These pressure histories are required to assess the structural response of a nuclear containment to hydrogen detonations. The compressible flow equations including detonation, which was treated as a sharp fluid discontinuity, were solved by the random choice method which reproduces maximum pressures and discontinuities sharply. The computer codes were validated by calculating pressure profiles and maximum wall pressures for plane and spherical geometries and comparing the results with exact analytic solutions. The two-dimensional axisymmetric program was used to calculate wall pressure histories in an actual nuclear containment. The numerical results for wall pressures are presented in a dimensionless form, which allows their use for different combinations of hydrogen concentration, and initial conditions. (orig.)
Variance and covariance calculations for nuclear materials accounting using ''MAVARIC''
Nasseri, K.K.
1987-07-01
Determination of the detection sensitivity of a materials accounting system to the loss of special nuclear material (SNM) requires (1) obtaining a relation for the variance of the materials balance by propagation of the instrument errors for the measured quantities that appear in the materials balance equation and (2) substituting measured values and their error standard deviations into this relation and calculating the variance of the materials balance. MAVARIC (Materials Accounting VARIance Calculations) is a custom spreadsheet, designed using the second release of Lotus 1-2-3, that significantly reduces the effort required to make the necessary variance (and covariance) calculations needed to determine the detection sensitivity of a materials accounting system. Predefined macros within the spreadsheet allow the user to carry out long, tedious procedures with only a few keystrokes. MAVARIC requires that the user enter the following data into one of four data tables, depending on the type of the term in the materials balance equation; the SNM concentration, the bulk mass (or solution volume), the measurement error standard deviations, and the number of measurements made during an accounting period. The user can also specify if there are correlations between transfer terms. Based on these data entries, MAVARIC can calculate the variance of the materials balance and the square root of this variance, from which the detection sensitivity of the accounting system can be determined
Variance and covariance calculations for nuclear materials accounting using 'MAVARIC'
Nasseri, K.K.
1987-01-01
Determination of the detection sensitivity of a materials accounting system to the loss of special nuclear material (SNM) requires (1) obtaining a relation for the variance of the materials balance by propagation of the instrument errors for the measured quantities that appear in the materials balance equation and (2) substituting measured values and their error standard deviations into this relation and calculating the variance of the materials balance. MAVARIC (Materials Accounting VARIance Calculations) is a custom spreadsheet, designed using the second release of Lotus 1-2-3, that significantly reduces the effort required to make the necessary variance (and covariance) calculations needed to determine the detection sensitivity of a materials accounting system. Predefined macros within the spreadsheet allow the user to carry out long, tedious procedures with only a few keystrokes. MAVARIC requires that the user enter the following data into one of four data tables, depending on the type of the term in the materials balance equation; the SNM concentration, the bulk mass (or solution volume), the measurement error standard deviations, and the number of measurements made during an accounting period. The user can also specify if there are correlations between transfer terms. Based on these data entries, MAVARIC can calculate the variance of the materials balance and the square root of this variance, from which the detection sensitivity of the accounting system can be determined
Impact of nuclear data uncertainty on safety calculations for spent nuclear fuel geological disposal
Herrero J.J.
2017-01-01
Full Text Available In the design of a spent nuclear fuel disposal system, one necessary condition is to show that the configuration remains subcritical at time of emplacement but also during long periods covering up to 1,000,000 years. In the context of criticality safety applying burn-up credit, k-eff eigenvalue calculations are affected by nuclear data uncertainty mainly in the burnup calculations simulating reactor operation and in the criticality calculation for the disposal canister loaded with the spent fuel assemblies. The impact of nuclear data uncertainty should be included in the k-eff value estimation to enforce safety. Estimations of the uncertainty in the discharge compositions from the CASMO5 burn-up calculation phase are employed in the final MCNP6 criticality computations for the intact canister configuration; in between, SERPENT2 is employed to get the spent fuel composition along the decay periods. In this paper, nuclear data uncertainty was propagated by Monte Carlo sampling in the burn-up, decay and criticality calculation phases and representative values for fuel operated in a Swiss PWR plant will be presented as an estimation of its impact.
Nuclear data for the calculation of thermal reactor reactivity coefficients
1989-01-01
On its 15th meeting in Vienna, 16-20 June 1986, the International Nuclear Data Committee (INDC) considered it important to review the accuracy with which changes in thermal reactor reactivity resulting from changes in temperature and coolant density can be predicted. It was noted that reactor physicists in several countries had to adjust the thermal neutron cross-section data base in order to reproduce measured reactivity coefficients. Consequently, it appeared to be essential to examine the consistency of the integral and differential cross-section data and to make all the information available which has a bearing on reactivity coefficient prediction. Following the recommendation of the INDC, the Nuclear Data Section of the International Atomic Energy Agency, therefore, convened the Advisory Group Meeting on Nuclear Data for the Calculation of Thermal Reaction Reactivity Coefficients, in Vienna, Austria, 7-10 Dec. 1987. The Conclusions and Recommendations of the meeting together with the papers presented, are submitted in the present document. A separate abstract was prepared for each of these 12 papers. Refs, figs and tabs
Nuclear performance calculations for the ELMO Bumpy Torus Reactor (EBTR) reference design
Santoro, R.T.; Barnes, J.M.
1977-12-01
The nuclear performance of the ELMO Bumpy Torus Reactor reference design has been calculated using the one-dimensional discrete ordinates code ANISN and the latest available ENDF/B-IV transport cross-section data and nuclear response functions. The calculated results include estimates of the spatial and integral heating rate with emphasis on the recovery of fusion neutron energy in the blanket assembly and minimization of the energy deposition rates in the cryogenic magnet coil assemblies. The tritium breeding ratio in the natural lithium-laden blanket was calculated to be 1.29 tritium nuclei per incident neutron. The radiation damage in the reactor structural material and in the magnet assembly is also given
Propagation of nuclear data uncertainties in fuel cycle calculations using Monte-Carlo technique
Diez, C.J.; Cabellos, O.; Martinez, J.S.
2011-01-01
Nowadays, the knowledge of uncertainty propagation in depletion calculations is a critical issue because of the safety and economical performance of fuel cycles. Response magnitudes such as decay heat, radiotoxicity and isotopic inventory and their uncertainties should be known to handle spent fuel in present fuel cycles (e.g. high burnup fuel programme) and furthermore in new fuel cycles designs (e.g. fast breeder reactors and ADS). To deal with this task, there are different error propagation techniques, deterministic (adjoint/forward sensitivity analysis) and stochastic (Monte-Carlo technique) to evaluate the error in response magnitudes due to nuclear data uncertainties. In our previous works, cross-section uncertainties were propagated using a Monte-Carlo technique to calculate the uncertainty of response magnitudes such as decay heat and neutron emission. Also, the propagation of decay data, fission yield and cross-section uncertainties was performed, but only isotopic composition was the response magnitude calculated. Following the previous technique, the nuclear data uncertainties are taken into account and propagated to response magnitudes, decay heat and radiotoxicity. These uncertainties are assessed during cooling time. To evaluate this Monte-Carlo technique, two different applications are performed. First, a fission pulse decay heat calculation is carried out to check the Monte-Carlo technique, using decay data and fission yields uncertainties. Then, the results, experimental data and reference calculation (JEFF Report20), are compared. Second, we assess the impact of basic nuclear data (activation cross-section, decay data and fission yields) uncertainties on relevant fuel cycle parameters (decay heat and radiotoxicity) for a conceptual design of a modular European Facility for Industrial Transmutation (EFIT) fuel cycle. After identifying which time steps have higher uncertainties, an assessment of which uncertainties have more relevance is performed
Calculation of Direct photon production in nuclear collisions
Cepila, J
2012-01-01
Prompt photons produced in a hard reaction are not expected to be accompanied by any final state interaction, either energy loss or absorption and one should not expect any nuclear effects at high pT . However, data from the PHENIX experiment indicates large-pT suppression in d+Au and central Au+Au collisions that cannot be accompanied by coherent phenomena. We propose a mechanism based on the energy sharing problem at large pT near the kinematic limit that is induced by multiple initial state interactions and that improves the agreement of calculations with PHENIX data. We calculate inclusive direct photon production cross sections in p+p collisions at RHIC and LHC energies using the color dipole approach without any additional parameter. Our predictions are in good agreement with the available data. Within the same framework, we calculate direct photon production rates in d+A and A+A collisions at RHIC energy. We also provide predictions for the same process in p+A collisions at LHC energy. Since the kinema...
Moment methods with effective nuclear Hamiltonians; calculations of radial moments
Belehrad, R.H.
1981-02-01
A truncated orthogonal polynomial expansion is used to evaluate the expectation value of the radial moments of the one-body density of nuclei. The expansion contains the configuration moments, , , and 2 >, where R/sup (k)/ is the operator for the k-th power of the radial coordinate r, and H is the effective nuclear Hamiltonian which is the sum of the relative kinetic energy operator and the Bruckner G matrix. Configuration moments are calculated using trace reduction formulae where the proton and neutron orbitals are treated separately in order to find expectation values of good total isospin. The operator averages are taken over many-body shell model states in the harmonic oscillator basis where all particles are active and single-particle orbitals through six major shells are included. The radial moment expectation values are calculated for the nuclei 16 O, 40 Ca, and 58 Ni and find that is usually the largest term in the expansion giving a large model space dependence to the results. For each of the 3 nuclei, a model space is found which gives the desired rms radius and then we find that the other 5 lowest moments compare favorably with other theoretical predictions. Finally, we use a method of Gordon (5) to employ the lowest 6 radial moment expectation values in the calculation of elastic electron scattering from these nuclei. For low to moderate momentum transfer, the results compare favorably with the experimental data
Key, S.W.
1985-01-01
The results of two calculations related to the impact response of spent nuclear fuel shipping casks are compared to the benchmark results reported in a recent study by the Japan Society of Mechanical Engineers Subcommittee on Structural Analysis of Nuclear Shipping Casks. Two idealized impacts are considered. The first calculation utilizes a right circular cylinder of lead subjected to a 9.0 m free fall onto a rigid target, while the second calculation utilizes a stainless steel clad cylinder of lead subjected to the same impact conditions. For the first problem, four calculations from graphical results presented in the original study have been singled out for comparison with HONDO III. The results from DYNA3D, STEALTH, PISCES, and ABAQUS are reproduced. In the second problem, the results from four separate computer programs in the original study, ABAQUS, ANSYS, MARC, and PISCES, are used and compared with HONDO III. The current version of HONDO III contains a fully automated implementation of the explicit-explicit partitioning procedure for the central difference method time integration which results in a reduction of computational effort by a factor in excess of 5. The results reported here further support the conclusion of the original study that the explicit time integration schemes with automated time incrementation are effective and efficient techniques for computing the transient dynamic response of nuclear fuel shipping casks subject to impact loading. (orig.)
Cost calculations for decommissioning and dismantling of nuclear research facilities
Andersson, I.; Backe, S.; Cato, A.; Lindskog, S.; Efraimsson, H.; Iversen, Klaus; Salmenhaara, S.; Sjoeblom, R.
2008-07-01
Today, it is recommended that planning of decommission should form an integral part of the activities over the life cycle of a nuclear facility (planning, building and operation), but it was only in the nineteen seventies that the waste issue really surface. Actually, the IAEA guidelines on decommissioning have been issued as recently as over the last ten years, and international advice on finance of decommissioning is even younger. No general international guideline on cost calculations exists at present. This implies that cost calculations cannot be performed with any accuracy or credibility without a relatively detailed consideration of the radiological prerequisites. Consequently, any cost estimates based mainly on the particulars of the building structures and installations are likely to be gross underestimations. The present study has come about on initiative by the Swedish Nuclear Power Inspectorate (SKI) and is based on a common need in Denmark, Finland, Norway and Sweden. The content of the report may be briefly summarised as follows. The background covers design and operation prerequisites as well as an overview of the various nuclear research facilities in the four participating countries: Denmark, Finland, Norway and Sweden. The purpose of the work has been to identify, compile and exchange information on facilities and on methodologies for cost calculation with the aim of achieving an 80 % level of confidence. The scope has been as follows: 1) to establish a Nordic network 2) to compile dedicated guidance documents on radiological surveying, technical planning and financial risk identification and assessment 3) to compile and describe techniques for precise cost calculations at early stages 4) to compile plant and other relevant data A separate section is devoted in the report to good practice for the specific purpose of early but precise cost calculations for research facilities, and a separate section is devoted to techniques for assessment of cost
Cost calculations for decommissioning and dismantling of nuclear research facilities
Andersson, I. (Studsvik Nuclear AB (Sweden)); Backe, S. (Institute for Energy Technology (Norway)); Cato, A.; Lindskog, S. (Swedish Nuclear Power Inspectorate (Sweden)); Efraimsson, H. (Swedish Radiation Protection Authority (Sweden)); Iversen, Klaus (Danish Decommissioning (Denmark)); Salmenhaara, S. (VTT Technical Research Centre of Finland (Finland)); Sjoeblom, R. (Tekedo AB, (Sweden))
2008-07-15
Today, it is recommended that planning of decommission should form an integral part of the activities over the life cycle of a nuclear facility (planning, building and operation), but it was only in the nineteen seventies that the waste issue really surface. Actually, the IAEA guidelines on decommissioning have been issued as recently as over the last ten years, and international advice on finance of decommissioning is even younger. No general international guideline on cost calculations exists at present. This implies that cost calculations cannot be performed with any accuracy or credibility without a relatively detailed consideration of the radiological prerequisites. Consequently, any cost estimates based mainly on the particulars of the building structures and installations are likely to be gross underestimations. The present study has come about on initiative by the Swedish Nuclear Power Inspectorate (SKI) and is based on a common need in Denmark, Finland, Norway and Sweden. The content of the report may be briefly summarised as follows. The background covers design and operation prerequisites as well as an overview of the various nuclear research facilities in the four participating countries: Denmark, Finland, Norway and Sweden. The purpose of the work has been to identify, compile and exchange information on facilities and on methodologies for cost calculation with the aim of achieving an 80 % level of confidence. The scope has been as follows: 1) to establish a Nordic network 2) to compile dedicated guidance documents on radiological surveying, technical planning and financial risk identification and assessment 3) to compile and describe techniques for precise cost calculations at early stages 4) to compile plant and other relevant data A separate section is devoted in the report to good practice for the specific purpose of early but precise cost calculations for research facilities, and a separate section is devoted to techniques for assessment of cost
A relativistic point coupling model for nuclear structure calculations
Buervenich, T.; Maruhn, J.A.; Madland, D.G.; Reinhard, P.G.
2002-01-01
A relativistic point coupling model is discussed focusing on a variety of aspects. In addition to the coupling using various bilinear Dirac invariants, derivative terms are also included to simulate finite-range effects. The formalism is presented for nuclear structure calculations of ground state properties of nuclei in the Hartree and Hartree-Fock approximations. Different fitting strategies for the determination of the parameters have been applied and the quality of the fit obtainable in this model is discussed. The model is then compared more generally to other mean-field approaches both formally and in the context of applications to ground-state properties of known and superheavy nuclei. Perspectives for further extensions such as an exact treatment of the exchange terms using a higher-order Fierz transformation are discussed briefly. (author)
Progress in theoretical calculation of transactinium isotope nuclear data
Salvy, J.
1984-05-01
Considerable progress has been made in effective use of nuclear theory for evaluation purposes. During the past few years, a number of basic improvements have developed in nuclear models commonly used for data evaluation. Actinide data evaluation can also use such improvements, but in the actinide region a further complication arises from the presence of fission competition. Nevertheless, systematic prescriptions for calculating even predicting neutron cross sections within an extended actinide region are available. Many efforts in several laboratorie are currently devoted to improving nuclear codes to be used for evaluation purposes. However at the present time numerous basic parameters associated with the neutron-induced fission process as well as neutron and gamma-ray competition have to be predetermined as input. Systematic studies of the behaviour of these parameters have been initiated with the aim of finding general trends hopefully useful for extrapolation in cases where direct information is lacking. Such trends can emerge from suitable examination of a large number of coherent experimental data, coherent theoretical results, or a combination these. This seems at the present time to be the most promising means for improving the actinide data evaluation. The aim of this paper is only to review briefly some of the main improvements either achieved or under way. The concern will be theoretical aspects useful for evaluating actinide data in the restricted incident neutron energy range from 10 KeV to 20 MeV. It is intended to focus on examples of systematics and on some improvements expected from microscopic methods under development
Unmanned Mobile Monitoring for Nuclear Emergency Response
Choi, YoungSoo; Park, JongWon; Kim, TaeWon; Jeong, KyungMin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
Severe accidents at nuclear power plant have led to significant consequences to the people, the environment or the facility. Therefore, the appropriate response is required for the mitigation of the accidents. In the past, most of responses were performed by human beings, but it was dangerous and risky. In this paper, we proposed unmanned mobile system for the monitoring of nuclear accident in order to response effectively. For the integrity of reactor cooling and containment building, reactor cooling pipe and hydrogen distribution monitoring with unmanned ground vehicle was designed. And, for the safety of workers, radiation distribution monitoring with unmanned aerial vehicle was designed. Unmanned mobile monitoring system was proposed to respond nuclear accidents effectively. Concept of reinforcing the integrity of RCS and containment building, and radiation distribution monitoring were described. RCS flow measuring, hydrogen distribution measuring and radiation monitoring deployed at unmanned vehicle were proposed. These systems could be a method for the preparedness of effective response of nuclear accidents.
Calculation of nuclear spin-spin coupling constants using frozen density embedding
Götz, Andreas W., E-mail: agoetz@sdsc.edu [San Diego Supercomputer Center, University of California San Diego, 9500 Gilman Dr MC 0505, La Jolla, California 92093-0505 (United States); Autschbach, Jochen [Department of Chemistry, University at Buffalo, State University of New York, Buffalo, New York 14260-3000 (United States); Visscher, Lucas, E-mail: visscher@chem.vu.nl [Amsterdam Center for Multiscale Modeling (ACMM), VU University Amsterdam, Theoretical Chemistry, De Boelelaan 1083, 1081 HV Amsterdam (Netherlands)
2014-03-14
We present a method for a subsystem-based calculation of indirect nuclear spin-spin coupling tensors within the framework of current-spin-density-functional theory. Our approach is based on the frozen-density embedding scheme within density-functional theory and extends a previously reported subsystem-based approach for the calculation of nuclear magnetic resonance shielding tensors to magnetic fields which couple not only to orbital but also spin degrees of freedom. This leads to a formulation in which the electron density, the induced paramagnetic current, and the induced spin-magnetization density are calculated separately for the individual subsystems. This is particularly useful for the inclusion of environmental effects in the calculation of nuclear spin-spin coupling constants. Neglecting the induced paramagnetic current and spin-magnetization density in the environment due to the magnetic moments of the coupled nuclei leads to a very efficient method in which the computationally expensive response calculation has to be performed only for the subsystem of interest. We show that this approach leads to very good results for the calculation of solvent-induced shifts of nuclear spin-spin coupling constants in hydrogen-bonded systems. Also for systems with stronger interactions, frozen-density embedding performs remarkably well, given the approximate nature of currently available functionals for the non-additive kinetic energy. As an example we show results for methylmercury halides which exhibit an exceptionally large shift of the one-bond coupling constants between {sup 199}Hg and {sup 13}C upon coordination of dimethylsulfoxide solvent molecules.
Responsible stewardship of nuclear materials
Hannum, W.H.
1994-01-01
The ability to tap the massive energy potential of nuclear fission was first developed as a weapon to end a terrible world war. Nuclear fission is also a virtually inexhaustible energy resource, and is the only energy supply in certain areas in Russia, Kazakhstan and elsewhere. The potential link between civilian and military applications has been and continues to be a source of concern. With the end of the Cold War, this issue has taken a dramatic turn. The U.S. and Russia have agreed to reduce their nuclear weapons stockpiles by as much as two-thirds. This will make some 100 tonnes of separated plutonium and 500 tonnes of highly enriched uranium available, in a form that is obviously directly usable for weapons. The total world inventory of plutonium is now around 1000 tonnes and is increasing at 60-70 tonnes per year. There is even more highly enriched uranium. Fortunately the correct answer to what to do with excess weapons material is also the most attractive. It should be used and reused as fuel for fast reactors. Material in use (particularly nuclear material) is very easy to monitor and control, and is quite unattractive for diversion. Active management of fissile materials not only makes a major contribution to economic stability and well-being, but also simplifies accountability, inspection and other safeguards processes; provides a revenue stream to pay for the necessary safeguards; and, most importantly, limits the prospective world inventory of plutonium to only that which is used and useful
Development of a power-period calculation unit for nuclear reactor Control
Martin, J.
1966-10-01
The apparatus studied is a digital calculating assembly which makes it possible to prepare and to present numerically the period and power of a nuclear reactor during operation, from start-up to nominal power. The pulses from a fission chamber are analyzed continuously, using real time. A small number of elements is required because of the systematic use of a calculation technique comprising the determination of a base 2 logarithm by a linear approximation. The accuracy obtained for the period is of the order of 14%; the response time of the order of the calculated period value. An approximate value of the power (30%) is given at each calculation cycle together with the power thresholds required for the control. (author) [fr
Validation of iron nuclear data for the neutron calculation of nuclear reactors
Vaglio-Gaudard, C.
2010-01-01
The GEN-III and GEN-IV reactors will be equipped with heavy reflectors. However, the existing integral validation of the iron nuclear data in the latest JEFF3 European library in the frame of the neutron calculation of the heavy reflector is very partial: some results exist concerning fast reactors but there is no result corresponding to the LWR heavy reflector. No clear trend on the JEFF3 iron cross sections was brought into evidence up to now for fission reactor calculations. Iron nuclear data were completely re-evaluated in the JEFF3 library. Despite the fact that iron is widely used in the nuclear industry, large uncertainties are still associated with its nuclear data, particularly its inelastic cross section which is very important in the neutron slowing down. A validation of 56 Fe nuclear data was performed on the basis of the analysis of integral experiments. Two major critical experiments, the PERLE experiment and the Gas Benchmark, were interpreted with 3D reference Monte-Carlo calculations and the JEFF3.1.1 library. The PERLE experiment was recently performed in the EOLE zero-power facility (CEA Cadarache). This experiment is dedicated to heavy reflector physics in GEN-III light water reactors. It was especially conceived for the validation of iron nuclear data. The Gas Benchmark is representative of a Gas Fast Reactor with a stainless steel reflector (with no fertile blanket) in the MASURCA facility (CEA Cadarache). Radial traverses of reaction rates were measured to characterize flux attenuation at various energies in the reflector. The results of the analysis of both experiments show good agreement between the calculations and the measurements, which is confirmed by the analysis of complementary experiments (ZR-6M, MISTRAL4, CIRANO-ZONA2B). A process of re-estimating the 56 Fe nuclear data was implemented on the basis of feedback from these two experiments and the RDN code. This code relies on a non-linear regression method using an iterative
A nuclear engineer's ethical responsibility to society
Kemeny, L.G.
1989-01-01
Chernobyl notwithstanding, this paper seeks to illustrate why, on numerous fronts, nuclear technology provides the safest, cleanest and most effective method of base-load power generation. In particular it seeks to demonstrate that, despite the strident rhetoric and media exposure given to the anti-nuclear lobby, the technology is fundamental to the quality of life and the equitable sharing of energy by the year 2000. Therefore, the safety and technological superiority of the nuclear fuel cycle together with its high technology peripheral benefits both societal and fiscal are viewed as an ever increasing challenge and motivation which constitutes a major part of the nuclear engineer's ethical responsibility to society
Calculational framework for safety analyses of non-reactor nuclear facilities
Coleman, J.R.
1994-01-01
A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks
A new approach to calculating spatial impulse responses
Jensen, Jørgen Arendt
1997-01-01
Using linear acoustics the emitted and scattered ultrasound field can be found by using spatial impulse responses as developed by Tupholme (1969) and Stepanishen (1971). The impulse response is calculated by the Rayleigh integral by summing the spherical waves emitted from all of the aperture...
Calculation methods of reactivity using derivatives of nuclear power and Filter fir
Diaz, Daniel Suescun
2007-01-01
This work presents two new methods for the solution of the inverse point kinetics equation. The first method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. Applying some conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has special characteristics, amongst which the possibility of using different sampling periods, and the possibility of restarting the calculation, after its interruption associated it with a possible equipment malfunction, allowing the calculation of reactivity in a non-continuous way. Apart from this reactivity can be obtained with or without dependency on the nuclear power memory. The second method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. The reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. In this method it can be pointed out that the linear part is equivalent to a filter named Finite Impulse Response (Fir). The Fir filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive way. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way. The proposed methods were validated using signals with random noise and showing the relationship between the reactivity difference and the degree of the random noise. (author)
Dose-Response Calculator for ArcGIS
Hanser, Steven E.; Aldridge, Cameron L.; Leu, Matthias; Nielsen, Scott E.
2011-01-01
The Dose-Response Calculator for ArcGIS is a tool that extends the Environmental Systems Research Institute (ESRI) ArcGIS 10 Desktop application to aid with the visualization of relationships between two raster GIS datasets. A dose-response curve is a line graph commonly used in medical research to examine the effects of different dosage rates of a drug or chemical (for example, carcinogen) on an outcome of interest (for example, cell mutations) (Russell and others, 1982). Dose-response curves have recently been used in ecological studies to examine the influence of an explanatory dose variable (for example, percentage of habitat cover, distance to disturbance) on a predicted response (for example, survival, probability of occurrence, abundance) (Aldridge and others, 2008). These dose curves have been created by calculating the predicted response value from a statistical model at different levels of the explanatory dose variable while holding values of other explanatory variables constant. Curves (plots) developed using the Dose-Response Calculator overcome the need to hold variables constant by using values extracted from the predicted response surface of a spatially explicit statistical model fit in a GIS, which include the variation of all explanatory variables, to visualize the univariate response to the dose variable. Application of the Dose-Response Calculator can be extended beyond the assessment of statistical model predictions and may be used to visualize the relationship between any two raster GIS datasets (see example in tool instructions). This tool generates tabular data for use in further exploration of dose-response relationships and a graph of the dose-response curve.
Licensee responsibility for nuclear power plant safety
Schneider, Horst
2010-01-01
Simple sentences easy to grasp are desirable in regulations and bans. However, in a legal system, their meaning must be unambiguous. Article 6, Paragraph 1 of the EURATOM Directive on a community framework for the nuclear safety of nuclear facilities of June 2009 states that 'responsibility for the nuclear safety of a nuclear facility is incumbent primarily on the licensee.' The draft 'Safety Criteria for Nuclear Power Plants, Revision D, April 2009' of the German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) (A Module 1, 'Safety Criteria for Nuclear Power Plants: Basic Safety Criteria' / '0 Principles' Paragraph 2) reads: 'Responsibility for ensuring safety rests with the licensee. He shall give priority to compliance with the safety goal over the achievement of other operational objectives.' In addition, the existing rules and regulations, whose rank is equivalent to that of international regulations, assign priority to the safety goal to be pursued by the licensee over all other objectives of the company. The operator's responsibility for nuclear safety can be required and achieved only on the basis of permits granted, which must meet legal requirements. The operator's proximity to plant operation is the reason for his 'primary responsibility.' Consequently, verbatim incorporation of Article 6, Paragraph 1 of the EURATOM Directive would only be a superscript added to existing obligations of the operator - inclusive of a safety culture designed as an incentive to further 'the spirit of safety-related actions' - without any new legal contents and consequences. In the reasons of the regulation, this would have to be clarified in addition to the cryptic wording of 'responsibility.. primarily,' at the same time expressing that operators and authorities work together in a spirit of openness and trust. (orig.)
Duluc, Matthieu; Bardelay, Aurélie; Celik, Cihangir; Heinrichs, Dave; Hopper, Calvin; Jones, Richard; Kim, Soon; Miller, Thomas; Troisne, Marc; Wilson, Chris
2017-09-01
AWE (UK), IRSN (France), LLNL (USA) and ORNL (USA) began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron and gamma), useful for emergency response teams and public authorities. This paper will present, firstly the motivation and the long term objectives for this update, then the overview of the initial configurations for updated calculations and preliminary results obtained with modern 3D codes.
Duluc Matthieu
2017-01-01
Full Text Available AWE (UK, IRSN (France, LLNL (USA and ORNL (USA began a long term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. This document, published almost 20 years ago, gives order of magnitude estimates of key parameters, such as number of fissions and doses (neutron and gamma, useful for emergency response teams and public authorities. This paper will present, firstly the motivation and the long term objectives for this update, then the overview of the initial configurations for updated calculations and preliminary results obtained with modern 3D codes.
Calculation of nuclear excitation in an electron transition
Pisk, K. (Institut Rudjer Boskovic, Zagreb (Yugoslavia)); Kaliman, Z. (Rijeka Univ. (Yugoslavia). Faculty of Pedagogics); Logan, B.A. (Ottawa Univ., ON (Canada). Ottawa-Carleton Centre for Physics)
1989-11-06
We have made a theoretical investigation of nuclear excitation during an electron transition (NEET). Our approach allows us to express the NEET probabilities in terms of the excited nuclear level width, the energy difference between the nuclear and electron transition, the Coulomb interaction between the initial electron states, and the electron level width. A comparison is made with the available experimental results. (orig.).
Nuclear data requirements for fission reactor neutronics calculations
Finck, P.
1998-01-01
The paper discusses current European nuclear data measurement and evaluation requirements for fission reactor technology applications and problems involved in meeting the requirements. Reference is made to the NEA High Priority Nuclear Data Request List and to the production of the new JEFF-3 library of evaluated nuclear data. There are requirements for both differential (or basic) nuclear data measurements and for different types of integral measurement critical facility measurements and isotopic sample irradiation measurements. Cross-section adjustment procedures are being used to take into account the simpler types of integral measurement, and to define accuracy needs for evaluated nuclear data
Responsibilities of the nuclear-weapon states
Wang Jun
1994-01-01
The responsibilities of Nuclear Weapon States are presented by a straightforward analysis together with the ways in which they could fulfill them. The complete undertaking of all the commitments by the Nuclear Weapon States may take a long time. However they do not have a single excuse to neglect such a historic opportunity to do their best to provide a genuinely secure world environment for the international community, of which they too are members
A High Performance Block Eigensolver for Nuclear Configuration Interaction Calculations
Aktulga, Hasan Metin; Afibuzzaman, Md.; Williams, Samuel; Buluc, Aydin; Shao, Meiyue
2017-01-01
As on-node parallelism increases and the performance gap between the processor and the memory system widens, achieving high performance in large-scale scientific applications requires an architecture-aware design of algorithms and solvers. We focus on the eigenvalue problem arising in nuclear Configuration Interaction (CI) calculations, where a few extreme eigenpairs of a sparse symmetric matrix are needed. Here, we consider a block iterative eigensolver whose main computational kernels are the multiplication of a sparse matrix with multiple vectors (SpMM), and tall-skinny matrix operations. We then present techniques to significantly improve the SpMM and the transpose operation SpMM T by using the compressed sparse blocks (CSB) format. We achieve 3-4× speedup on the requisite operations over good implementations with the commonly used compressed sparse row (CSR) format. We develop a performance model that allows us to correctly estimate the performance of our SpMM kernel implementations, and we identify cache bandwidth as a potential performance bottleneck beyond DRAM. We also analyze and optimize the performance of LOBPCG kernels (inner product and linear combinations on multiple vectors) and show up to 15× speedup over using high performance BLAS libraries for these operations. The resulting high performance LOBPCG solver achieves 1.4× to 1.8× speedup over the existing Lanczos solver on a series of CI computations on high-end multicore architectures (Intel Xeons). We also analyze the performance of our techniques on an Intel Xeon Phi Knights Corner (KNC) processor.
Nuclear criticality safety calculational analysis for small-diameter containers
LeTellier, M.S.; Smallwood, D.J.; Henkel, J.A.
1995-11-01
This report documents calculations performed to establish a technical basis for the nuclear criticality safety of favorable geometry containers, sometimes referred to as 5-inch containers, in use at the Portsmouth Gaseous Diffusion Plant. A list of containers currently used in the plant is shown in Table 1.0-1. These containers are currently used throughout the plant with no mass limits. The use of containers with geometries or material types other than those addressed in this evaluation must be bounded by this analysis or have an additional analysis performed. The following five basic container geometries were modeled and bound all container geometries in Table 1.0-1: (1) 4.32-inch-diameter by 50-inch-high polyethylene bottle; (2) 5.0-inch-diameter by 24-inch-high polyethylene bottle; (3) 5.25-inch-diameter by 24-inch-high steel can (open-quotes F-canclose quotes); (4) 5.25-inch-diameter by 15-inch-high steel can (open-quotes Z-canclose quotes); and (5) 5.0-inch-diameter by 9-inch-high polybottle (open-quotes CO-4close quotes). Each container type is evaluated using five basic reflection and interaction models that include single containers and multiple containers in normal and in credible abnormal conditions. The uranium materials evaluated are UO 2 F 2 +H 2 O and UF 4 +oil materials at 100% and 10% enrichments and U 3 O 8 , and H 2 O at 100% enrichment. The design basis safe criticality limit for the Portsmouth facility is k eff + 2σ < 0.95. The KENO study results may be used as the basis for evaluating general use of these containers in the plant
Calculated energy response of lithium fluoride finger-tip dosimeters
Johns, T.F.
1965-07-01
Calculations have been made of the energy response of the lithium fluoride thermoluminescent dosimeters being used at A.E.E. Winfrith for the measurement of radiation doses to the finger-tips of people handling radio-active materials. It is shown that the energy response is likely to be materially affected if the sachet in which the powder is held contains elements with atomic numbers much higher than 9 (e.g. if the sachet is made from polyvinyl chloride). (author)
Seeliger, D.
1993-01-01
This contribution contains a brief presentation and comparison of the different Statistical Multistep Approaches, presently available for practical nuclear data calculations. (author). 46 refs, 5 figs
Calculation of the well depth parameter to the nuclear potential
Kim, Y.U.; Kim, Y.J.
1984-01-01
Well depth parameter S or range correction factor S-1 is computed for several nuclear potentials such as square, Gaussian, exponential and Yukawa wells. A simple central force is assumed for nuclear potential between nucleons. We adopted only two parameters for potentials and attempted to clarify the fundamental nature of the nuclear forces that bind a proton and a neutron into a deuteron. Results thus obtained were used for an estimate of first order correction to simple square well model. (Author)
Nuclear responses in INTOR plasma stabilization elements
Gohar, Y.; Gilligan, J.; Jung, J.; Mattas, R.F.; Miley, G.H.; Wiffen, F.W.; Yang, S.
1985-01-01
Nuclear responses in the plasma stabilization elements were studied in a parametric fashion as a part of the transient electromagnetics critical issue C of ETR/INTOR activity. The main responses are neutron fluence and radiation dose in the insulator material, induced resistivity and atomic displacement in the conductor material, nuclear heating and life analysis for the elements. Copper and aluminum conductors with either MgAl 2 O 4 or MgO insulating material were investigated. Radiation damage and life analysis for these elements were also discussed
Dielectric response of periodic systems from quantum Monte Carlo calculations.
Umari, P; Willamson, A J; Galli, Giulia; Marzari, Nicola
2005-11-11
We present a novel approach that allows us to calculate the dielectric response of periodic systems in the quantum Monte Carlo formalism. We employ a many-body generalization for the electric-enthalpy functional, where the coupling with the field is expressed via the Berry-phase formulation for the macroscopic polarization. A self-consistent local Hamiltonian then determines the ground-state wave function, allowing for accurate diffusion quantum Monte Carlo calculations where the polarization's fixed point is estimated from the average on an iterative sequence, sampled via forward walking. This approach has been validated for the case of an isolated hydrogen atom and then applied to a periodic system, to calculate the dielectric susceptibility of molecular-hydrogen chains. The results found are in excellent agreement with the best estimates obtained from the extrapolation of quantum-chemistry calculations.
Nuclear structure effects on calculated fast neutron reaction cross sections
Avrigeanu, V.
1992-01-01
The importance of accurate low-lying level schemes for reaction cross section calculation and need for microscopically calculated levels are proved with reference to fast neutron induced reactions in the A = 50 atomic mass range. The uses of the discrete levels both for normalization of phenomenological level density approaches and within Hauser-Feshbach calculations are discussed in this respect. (Author)
Ibrahim, Ahmad M.; Polunovskiy, Eduard; Loughlin, Michael J.; Grove, Robert E.; Sawan, Mohamed E.
2016-01-01
turns that are closer to the plasma source, it is our recommendation to adopt the utilization of ADVANTG in similar calculations to increase the reliability of MCNP calculations of detailed distributions of nuclear responses.
Ibrahim, Ahmad M., E-mail: ibrahimam@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Polunovskiy, Eduard; Loughlin, Michael J. [ITER Organization, Route de Vinon Sur Verdon, 13067 St. Paul Lez Durance (France); Grove, Robert E. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Sawan, Mohamed E. [University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706 (United States)
2016-11-01
turns that are closer to the plasma source, it is our recommendation to adopt the utilization of ADVANTG in similar calculations to increase the reliability of MCNP calculations of detailed distributions of nuclear responses.
Variational Monte Carlo calculations of nuclear ground states
Wiringa, R.B.
1990-01-01
A major goal in nuclear physics is to understand how nuclear structure comes about from the underlying interactions between nucleons. This requires modelling nuclei as collections of strongly interacting nucleons. We start with realistic nucleon-nucleon potentials, supplemented with consistent three-nucleon potentials and two-body electroweak current operators, and try to predict nuclear ground properties, such as the binding energy, density and momentum distributions, and electromagnetic form factors. We also seek to predict other properties of nuclei such as excited states and low-energy reactions. 21 refs., 14 figs., 5 tabs
Du Yanjun; Liu Qingcheng; Liu Hongzhang; Qin Guoxiu
2009-01-01
In order to find the feasibility of calculating mine radiation dose based on γ field theory, this paper calculates the γ radiation dose of a mine by means of γ field theory based calculation method. The results show that the calculated radiation dose is of small error and can be used to monitor mine environment of nuclear radiation. (authors)
Key Response Planning Factors for the Aftermath of Nuclear Terrorism
Buddemeier, B R; Dillon, M B
2009-01-21
Despite hundreds of above-ground nuclear tests and data gathered from Hiroshima and Nagasaki, the effects of a ground-level, low-yield nuclear detonation in a modern urban environment are still the subject of considerable scientific debate. Extensive review of nuclear weapon effects studies and discussions with nuclear weapon effects experts from various federal agencies, national laboratories, and technical organizations have identified key issues and bounded some of the unknowns required to support response planning for a low-yield, ground-level nuclear detonation in a modern U.S. city. This study, which is focused primarily upon the hazards posed by radioactive fallout, used detailed fallout predictions from the advanced suite of three-dimensional (3-D) meteorology and plume/fallout models developed at Lawrence Livermore National Laboratory (LLNL), including extensive global Key Response Planning Factors for the Aftermath of Nuclear Terrorism geographical and real-time meteorological databases to support model calculations. This 3-D modeling system provides detailed simulations that account for complex meteorology and terrain effects. The results of initial modeling and analysis were presented to federal, state, and local working groups to obtain critical, broad-based review and feedback on strategy and messaging. This effort involved a diverse set of communities, including New York City, National Capitol Regions, Charlotte, Houston, Portland, and Los Angeles. The largest potential for reducing casualties during the post-detonation response phase comes from reducing exposure to fallout radiation. This can be accomplished through early, adequate sheltering followed by informed, delayed evacuation.B The response challenges to a nuclear detonation must be solved through multiple approaches of public education, planning, and rapid response actions. Because the successful response will require extensive coordination of a large number of organizations, supplemented by
Calculating the Responses of Self-Powered Radiation Detectors.
Thornton, D. A.
Available from UMI in association with The British Library. The aim of this research is to review and develop the theoretical understanding of the responses of Self -Powered Radiation Detectors (SPDs) in Pressurized Water Reactors (PWRs). Two very different models are considered. A simple analytic model of the responses of SPDs to neutrons and gamma radiation is presented. It is a development of the work of several previous authors and has been incorporated into a computer program (called GENSPD), the predictions of which have been compared with experimental and theoretical results reported in the literature. Generally, the comparisons show reasonable consistency; where there is poor agreement explanations have been sought and presented. Two major limitations of analytic models have been identified; neglect of current generation in insulators and over-simplified electron transport treatments. Both of these are developed in the current work. A second model based on the Explicit Representation of Radiation Sources and Transport (ERRST) is presented and evaluated for several SPDs in a PWR at beginning of life. The model incorporates simulation of the production and subsequent transport of neutrons, gamma rays and electrons, both internal and external to the detector. Neutron fluxes and fuel power ratings have been evaluated with core physics calculations. Neutron interaction rates in assembly and detector materials have been evaluated in lattice calculations employing deterministic transport and diffusion methods. The transport of the reactor gamma radiation has been calculated with Monte Carlo, adjusted diffusion and point-kernel methods. The electron flux associated with the reactor gamma field as well as the internal charge deposition effects of the transport of photons and electrons have been calculated with coupled Monte Carlo calculations of photon and electron transport. The predicted response of a SPD is evaluated as the sum of contributions from individual
The Nuclear Review: the Institution of Nuclear Engineers' response to the Review of Nuclear Power
Anon.
1994-01-01
The United Kingdom Government's Nuclear Review currently underway, addresses whether and in what form nuclear power should continue to be part of the country's power generation capability. This article sets out the response of the Institution of Nuclear Engineers to the Nuclear Review. This pro-nuclear group emphasises the benefits to be gained from diversity of generation in the energy supply industry. The environmentally benign nature of nuclear power is emphasised, in terms of gaseous emissions. The industry's excellent safety record also argues in favour of nuclear power. Finally, as power demand increases globally, a health U.K. nuclear industry could generate British wealth through power exports and via the construction industry. The Institution's view on radioactive waste management is also set out. (UK)
Accelerating Full Configuration Interaction Calculations for Nuclear Structure
Yang, Chao; Sternberg, Philip; Maris, Pieter; Ng, Esmond; Sosonkina, Masha; Le, Hung Viet; Vary, James; Yang, Chao
2008-01-01
One of the emerging computational approaches in nuclear physics is the full configuration interaction (FCI) method for solving the many-body nuclear Hamiltonian in a sufficiently large single-particle basis space to obtain exact answers - either directly or by extrapolation. The lowest eigenvalues and corresponding eigenvectors for very large, sparse and unstructured nuclear Hamiltonian matrices are obtained and used to evaluate additional experimental quantities. These matrices pose a significant challenge to the design and implementation of efficient and scalable algorithms for obtaining solutions on massively parallel computer systems. In this paper, we describe the computational strategies employed in a state-of-the-art FCI code MFDn (Many Fermion Dynamics - nuclear) as well as techniques we recently developed to enhance the computational efficiency of MFDn. We will demonstrate the current capability of MFDn and report the latest performance improvement we have achieved. We will also outline our future research directions
He Liu; Xiao Bo; Song Yumeng
2014-01-01
Non-nuclear steam run up compared with nuclear steam run up, can verify the design, manufacture, installation quality of the unit, at the same time shorten the follow-up duration of the entire group ready to start debugging time. In this paper, starting from the first law of thermodynamics, Analyzed Heat balance Calculation and Feasibility analysis for Initial startup of Fuqing nuclear Turbine unit with Non-nuclear steam, By the above calculation, to the system requirements and device status on the basis of technical specifications, confirmed the feasibility of Non-nuclear steam running up in theory. After the implementation of the Non-nuclear turn of Fuqing unit, confirmed the results fit with the actual process. In summary, the Initial startup of Fuqing turbine unit with Non-nuclear steam is feasible. (authors)
Variational Calculation for the Equation of State of Hot Asymmetric Nuclear Matter
Togashi, Hajime; Kanzawa, Hiroaki; Takano, Masatoshi
2010-01-01
We calculate the equation of state (EOS) of asymmetric nuclear matter at finite temperatures with the cluster variational method based on the realistic nuclear Hamiltonian composed of the AV18 and UIX nuclear potentials. The free energy is calculated with an extension of the variational method proposed by Schmidt and Pandharipande. The obtained thermodynamic quantities such as entropy, internal energy, pressure and chemical potential derived from the free energy are reasonable. It is also found that the present variational calculation is self-consistent. These thermodynamic quantities are essential ingredients in our project for constructing a new nuclear EOS applicable to supernova simulations.
Nuclear emergency preparedness and response in Germany
Miska, H.
2009-01-01
Off-site nuclear emergency response in Germany is divided into disaster response under the responsibility of the Laender and measures for precautionary radiation protection pursuant to the Precautionary Radiation Protection Act under the lead of federal authorities. Early countermeasures at the regional level require a different management than long-term and comprehensive actions of precautionary radiation protection. As situations may arise in which measures of both approaches overlap with regard to place and time, it is essential to make thorough preparations in order to avoid problems with implementation. (orig.)
Calculation of reactivity using a finite impulse response filter
Suescun Diaz, Daniel [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil); Senra Martinez, Aquilino [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil)], E-mail: aquilino@lmp.ufrj.br; Carvalho Da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, CEP 21941-914, RJ (Brazil)
2008-03-15
A new formulation is presented in this paper to solve the inverse kinetics equation. This method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. Reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. This new method of reactivity calculation has very special features, amongst which it can be pointed out that the linear part is characterized by a filter named finite impulse response (FIR). The FIR filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive form. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way.
Jeong, Kwan Seong; Lee, Dong Gyu; Jung, Chong Hun; Lee, Kune Woo
2006-01-01
The estimated decommissioning cost of nuclear research reactor is calculated by applying a unit cost factor-based engineering cost calculation method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning cost of nuclear research reactor is composed of labor cost, equipment and materials cost. Labor cost of decommissioning costs in decommissioning works are calculated on the basis of working time consumed in decommissioning objects. In this paper, the unit cost factors and work difficulty factors which are needed to calculate the labor cost in estimating decommissioning cost of nuclear research reactor are derived and figured out.
Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong
2014-01-01
The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)
Calculation of nuclear electromagnetic pulse propagation along the earth's surface
Liang Rui; Zheng Yi; Song Lijun; Zhang Xueqin; Lip Peng
2010-01-01
It calculates the LF/VLF wave of NEMP propagation along the earth's surface. The earth-wave and the sky-wave are taken into account in the calculation. With the distance increase, the earth wave attenuates fast than the sky wave, and the time difference between the earth wave and the sky wave is reduced. (authors)
Shielding calculations for ships carrying irradiated nuclear fuel
Burstall, R.F.; Dean, M.H.
1983-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the UK and France, for reprocessing. About twenty transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose rate greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large, and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both gamma radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board the ships, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
Shielding calculations for ships carrying irradiated nuclear fuel
Dean, M.H.
1985-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
Model calculations of nuclear data for biologically-important elements
Chadwick, M.B.; Blann, M.; Reffo, G.; Young, P.G.
1994-05-01
We describe calculations of neutron-induced reactions on carbon and oxygen for incident energies up to 70 MeV, the relevant clinical energy in radiation neutron therapy. Our calculations using the FKK-GNASH, GNASH, and ALICE codes are compared with experimental measurements, and their usefulness for modeling reactions on biologically-important elements is assessed
Evaluation of covariance in theoretical calculation of nuclear data
Kikuchi, Yasuyuki
1981-01-01
Covariances of the cross sections are discussed on the statistical model calculations. Two categories of covariance are discussed: One is caused by the model approximation and the other by the errors in the model parameters. As an example, the covariances are calculated for 100 Ru. (author)
Protection against nuclear terrorism: the IAEA response
Dodd, B.
2002-01-01
Full text: As a result of the events of 11 September 2001, the International Atomic Energy Agency (IAEA) identified possible threats from acts of nuclear terrorism. A report to the Board of Governors in November 2001 summarized the IAEA's ongoing work in areas relevant to the prevention and mitigation of the consequences of such acts and outlined proposals for a number of new and/or enhanced activities. Four main threats were addressed: theft of a nuclear weapon; acquisition of nuclear material; acquisition of other radioactive material; and violent acts against nuclear facilities. These proposals have been further refined and the new plan was approved in principle at the March 2002 board meeting. In the beginning, implementation will be dependent on member state contributions to a voluntary fund. Proposed new or enhanced activities are grouped into eight areas: I. Physical protection of nuclear material and nuclear facilities; II. Detection of malicious activities involving nuclear and other radioactive materials; III. State systems for nuclear material accountancy and control; IV. Security of radioactive material other than nuclear material; V. Assessment of safety/security related vulnerability of nuclear facilities; VI. Response to malicious acts, or threats thereof; VII. Adherence to and implementation of international agreements, guidelines and recommendations; VIII. Nuclear security co-ordination and information management. After an overview, this paper focuses on activity area IV, which deals with the radiological terrorism issues involving radioactive sources. A strategy for evaluation of the IAEA's role is presented, covering an analysis of the likely threats and possible scenarios. This leads to an assessment of the most desirable sources from a terrorist's viewpoint. The strategy then examines how terrorists might acquire such sources and attempts to determine the best ways to prevent their acquisition. Further activities are proposed to prevent the use
Calculations to support design of a nuclear material tracking system
Carter, L.L.; Eggers, R.F.; Williams, T.L.
1991-01-01
The Westinghouse Hanford Company is developing a nuclear material tracking system called NTRAK for the US Department of Energy at the Savannah River site. The NTRAK system is designed to determine the position and approximate magnitude of packages of special nuclear material (SNM) moving through a nuclear plant. The NTRAK accomplishes this by using special assemblies of detectors called modules to measure the gamma radiation emitted by the SNM. After measurement, raw data are processed to determine the direction to and position of the gamma-ray source. In order for the NTRAK method of SNM tracking to work, the gamma-ray signal at the detector modules must be at least four standard deviations above background. This paper addresses the use of the Monte Carlo computer code for neutron and photon transport (MCNP) to (a) predict the radiation emitted by plutonium oxide sources and (b) predict the counting rate of NaI detectors measuring those sources
Criticality safety calculations for the nuclear waste disposal canisters
Anttila, M.
1996-12-01
The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)
Response surfaces and sensitivity analyses for an environmental model of dose calculations
Iooss, Bertrand [CEA Cadarache, DEN/DER/SESI/LCFR, 13108 Saint Paul lez Durance, Cedex (France)]. E-mail: bertrand.iooss@cea.fr; Van Dorpe, Francois [CEA Cadarache, DEN/DTN/SMTM/LMTE, 13108 Saint Paul lez Durance, Cedex (France); Devictor, Nicolas [CEA Cadarache, DEN/DER/SESI/LCFR, 13108 Saint Paul lez Durance, Cedex (France)
2006-10-15
A parametric sensitivity analysis is carried out on GASCON, a radiological impact software describing the radionuclides transfer to the man following a chronic gas release of a nuclear facility. An effective dose received by age group can thus be calculated according to a specific radionuclide and to the duration of the release. In this study, we are concerned by 18 output variables, each depending of approximately 50 uncertain input parameters. First, the generation of 1000 Monte-Carlo simulations allows us to calculate correlation coefficients between input parameters and output variables, which give a first overview of important factors. Response surfaces are then constructed in polynomial form, and used to predict system responses at reduced computation time cost; this response surface will be very useful for global sensitivity analysis where thousands of runs are required. Using the response surfaces, we calculate the total sensitivity indices of Sobol by the Monte-Carlo method. We demonstrate the application of this method to one site of study and to one reference group near the nuclear research Center of Cadarache (France), for two radionuclides: iodine 129 and uranium 238. It is thus shown that the most influential parameters are all related to the food chain of the goat's milk, in decreasing order of importance: dose coefficient 'effective ingestion', goat's milk ration of the individuals of the reference group, grass ration of the goat, dry deposition velocity and transfer factor to the goat's milk.
Shindo, R.; Yamashita, K.; Murata, I.
1991-01-01
The nuclear design code system for the HTTR consists of one dimensional cell burnup computer code, developed in JAERI and the TWOTRAN-2 transport code. In order to satisfy related design criteria, uncertainty of the calculation was investigated by comparing the calculated and experimental results. The experiments were performed with a graphite moderated critical assembly. It was confirmed that discrepancies between calculations and experiments were small enough to be allowed in the nuclear design of HTTR. 8 refs, 6 figs
Ainsworth, T.L.
1983-01-01
The Δ(1232) plays an important role in determining the properties of nuclear and neutron matter. The effects of the Δ resonance are incorporated explicitly by using a coupled channel formalism. A method for constraining a lowest order variational calculation, appropriate when nucleon internal degrees of freedom are made explicity, is presented. Different N-N potentials were calculated and fit to phase shift data and deuteron properties. The potentials were constructed to test the relative importance of the Δ resonance on nuclear properties. The symmetry energy and incompressibility of nuclear matter are generally reproduced by this calculation. Neutron matter results lead to appealing neutron star models. Fermi liquid parameters for 3 He are calculated with a model that includes both direct and induced terms. A convenient form of the direct interaction is obtained in terms of the parameters. The form of the direct interaction ensures that the forward scattering sum rule (Pauli principle) is obeyed. The parameters are adjusted to fit the experimentally determined F 0 /sup s/, F 0 /sup a/, and F 1 /sup s/ Landau parameters. Higher order Landau parameters are calculated by the self-consistent solution of the equations; comparison to experiment is good. The model also leads to a preferred value for the effective mass of 3 He. Of the three parameters only one shows any dependence on pressure. An exact sum rule is derived relating this parameter to a specific summation of Landau parameters
Use of nuclear reaction models in cross section calculations
Grimes, S.M.
1975-03-01
The design of fusion reactors will require information about a large number of neutron cross sections in the MeV region. Because of the obvious experimental difficulties, it is probable that not all of the cross sections of interest will be measured. Current direct and pre-equilibrium models can be used to calculate non-statistical contributions to neutron cross sections from information available from charged particle reaction studies; these are added to the calculated statistical contribution. Estimates of the reliability of such calculations can be derived from comparisons with the available data. (3 tables, 12 figures) (U.S.)
Three-dimensional calculation of inhomogeneous nuclear matter
Okamoto, Minoru; Maruyama, Toshiki; Yabana, Kazuhiro; Tatsumi, Toshitaka
2012-01-01
We numerically explore the pasta structures and properties of low-density symmetric nuclear matter without any assumption on the geometry. We observe conventional pasta structures, while a mixture of the pasta appears as a meta-stable state at some transient densities. We also analyze the lattice structure of droplets.
Three-dimensional calculation of inhomogeneous nuclear matter
Okamoto, Minoru; Maruyama, Toshiki; Yabana, Kazuhiro; Tatsumi, Toshitaka [Graduate School of Pure and Applied Science, University of Tsukuba (Japan); Advanced Science Research Center, Japan Atomic Energy Agency (Japan); Graduate School of Pure and Applied Science, University of Tsukuba (Japan); Department of Physics, Kyoto University (Japan)
2012-11-12
We numerically explore the pasta structures and properties of low-density symmetric nuclear matter without any assumption on the geometry. We observe conventional pasta structures, while a mixture of the pasta appears as a meta-stable state at some transient densities. We also analyze the lattice structure of droplets.
Tropospheric response to a nuclear exchange
Penner, J.E.
1983-10-01
The immediate effects of a full-scale nuclear war would be large and severe. The survivors of such a war would have to endure possible changes in the chemical structure of the atmosphere. These changes may come about as a result of changes caused by the nuclear explosions themselves (direct effects) or as a result of changes caused by fires that may start after the explosions (indirect effects). This paper focuses on the expected global-scale changes in the chemical structure of the atmosphere from both direct and indirect effects after a full-scale nuclear exchange. The immediate effects of a nuclear explosion include the creation of a hot mass of air or fireball which rises in the atmosphere to a level that depends on the yield of the explosion. Because the fireball is hot, it is able to dissociate atmospheric nitrogen, N 2 . As the fireball cools, nitrogen atoms recombine with oxygen to form nitrogen oxides, NO and NO 2 . In addition, dust and recondensed gases are swept up through the stem of the fireball and deposited at the same level to which the fireball rises. This paper focuses on the response of atmospheric ozone to a nuclear war
Shang Yanlong; Cai Qi; Chen Lisheng; Zhang Yangwei
2012-01-01
In this paper, the combined method of response surface and importance sampling was applied for calculation of parameter failure probability of the thermodynamic system. The mathematics model was present for the parameter failure of physics process in the thermodynamic system, by which the combination arithmetic model of response surface and importance sampling was established, then the performance degradation model of the components and the simulation process of parameter failure in the physics process of thermodynamic system were also present. The parameter failure probability of the purification water system in nuclear reactor was obtained by the combination method. The results show that the combination method is an effective method for the calculation of the parameter failure probability of the thermodynamic system with high dimensionality and non-linear characteristics, because of the satisfactory precision with less computing time than the direct sampling method and the drawbacks of response surface method. (authors)
Recent developments in nuclear reaction theories and calculations
Gardner, D.G.
1980-01-01
A brief review is given of some recent developments in the fields of optical model potentials; level densities; and statistical model, precompound, and direct reaction codes and calculations. Significant developments have occurred in all of these fields since the 1977 Conference on Neutron Cross Sections, which will greatly enhance the ability to calculate high-energy neutron-induced reaction cross sections in the next few years. 11 figures, 3 tables
Progress on calculations of nuclear data at Tsinghua University
Chen Zhenpeng
1995-01-01
The calculated cross sections of direct inelastic scattering neutron from Ni, the research on using parameters of SOM in calculation of CCOM (coupled-channel optical model) and the reduced R-matrix analysis of n+ 16 O between 6.2 and 10.5 MeV are described. The total cross section and (n,α) cross section of n + 16 O are shown out. (2 tabs., 1 fig.)
Frequency Calculation For Loss Coolant Accident In The Nuclear Reactor
Sony, DT
1996-01-01
LOCA as initiating event is engineering judgement, because it is rare condition. So, to determine LOCA frequency used be probability and statistic method. By probability and statistic method was estimated from size, weld, age, learning curve and quality, etc. it has been calculated for LOCA frequency in the simplified piping system model, especially estimates from size and weld factors. From calculation, LOCA frequency is 9,82.10 - 6/year
The nuclear response in extended RPA theories
Wambach, J.
1991-01-01
Linear response theories for small-amplitude nuclear motion are discussed. A review is given of the Random-Phase-Approximation and its successes and shortcomings are pointed out. Systematic improvements are presented which include two-particle two-hole excitations and account for dissipative effects due to binary collisions of nucleons in the mean field. These improved theories describe inelastic scattering experiments satisfactorily over a wide kinematical range. (author)
Nuclear-magnetic-resonance quantum calculations of the Jones polynomial
Marx, Raimund; Spoerl, Andreas; Pomplun, Nikolas; Schulte-Herbrueggen, Thomas; Glaser, Steffen J.; Fahmy, Amr; Kauffman, Louis; Lomonaco, Samuel; Myers, John M.
2010-01-01
The repertoire of problems theoretically solvable by a quantum computer recently expanded to include the approximate evaluation of knot invariants, specifically the Jones polynomial. The experimental implementation of this evaluation, however, involves many known experimental challenges. Here we present experimental results for a small-scale approximate evaluation of the Jones polynomial by nuclear magnetic resonance (NMR); in addition, we show how to escape from the limitations of NMR approaches that employ pseudopure states. Specifically, we use two spin-1/2 nuclei of natural abundance chloroform and apply a sequence of unitary transforms representing the trefoil knot, the figure-eight knot, and the Borromean rings. After measuring the nuclear spin state of the molecule in each case, we are able to estimate the value of the Jones polynomial for each of the knots.
Robot-borne fault tolerant calculators for nuclear use
Giraud, A.; Robiolle, M.
1995-01-01
The use of robots has become a necessity in civil nuclear industry. Electronic systems of such robots must tolerate cumulative ionizing radiation dose effects. Today's objective is to reach a 3 kGy dose resistance. Difficulties and costs involved during on-site maintenance imply to warrant at least one functioning mode in the case of system failure. To improve the behaviour of robot-borne systems, the CEA Department for Nuclear Engineering Studies (DEIN) has developed a method for the selection of industrial electronic components and has built computer architectures which allows to break free from some cumulative dose sensitive parameters. This paper presents the MICADO and CADMOS architectures developed at the DEIN. (J.S.). 15 refs., 5 figs
Calculation of nuclear moment of inertia with proper treatment of pairing interaction
Tazaki, S.; Ando, Y.; Hasegawa, M.
1997-01-01
An attempt to calculate nuclear moments of inertia treating the pairing interaction exactly is reported. As usual, hamiltonian is composed of the Nilsson's singleparticle energies and the pairing interaction, but the eigenstates and the eigenvalues are calculated exactly in a realistic, sufficiently large model space. The method of calculating the moment of inertia is presented. (author)
Nuclear power history calculation for subcritical systems using Euler-MacLaurin formula
Henrice Junior, Edson; Goncalves, Alessandro da Cruz
2013-01-01
This paper presents an efficient method for calculating the reactivity using inverse point kinetic equation for subcritical systems by applying the Euler-MacLaurin summation formula to calculate the nuclear power history. In accordance with the accuracy of the numerical results, this method does not require a large number of points for calculation, providing accurate results with low computational cost. (author)
Merger of Nuclear Data with Criticality Safety Calculations
Derrien, H.; Larson, N.M.; Leal, L.C.
1999-09-20
In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes � measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations � which in the past have been treated independently.
Merger of Nuclear Data with Criticality Safety Calculations
Derrien, H.; Larson, N.M.; Leal, L.C.
1999-01-01
In this paper we report on current activities related to the merger of differential/integral data (especially in the resolved-resonance region) with nuclear criticality safety computations. Techniques are outlined for closer coupling of many processes measurement, data reduction, differential-data analysis, integral-data analysis, generating multigroup cross sections, data-testing, criticality computations which in the past have been treated independently
The social responsibility of nuclear energy
Mizuo, Junichi
2008-01-01
Interest in the concept of Social Responsibility (SR) has increased recently. Continuing advances in the pace of innovative science and information technology development, growing competition in the world's markets, economic globalization and harsh criticism from communities have all drawn attention to the behavior of different organizations. As a way of drawing global attention to the fulfillment of SR, the goal of coexistence assumes an increasingly significant role from the standpoint of sustainable development of organizations and society. This implies that SR involves two responsibilities: the primary responsibility being an obligation to society, and the secondary responsibility being a positive contribution to society. Seen from the same perspective, Nuclear Energy (NE) is expected to make a positive contribution to the advancement of society and to encourage a safety culture that prevents serious accidents while also encouraging the sound development of organizations and society. ('Society' includes the environment and the economy, with the same sense as a 'triple bottom line'.) Considering the Social Responsibility of Nuclear Energy (NSR) from these points-of-view, NE should coexist with multiple stakeholders. The purpose of this paper is to clarify the relationship between NE and society, to define a framework for problem-solving, and finally to suggest changes in NSR as a whole. (author)
Three-dimensional Monte Carlo calculation of some nuclear parameters
Günay, Mehtap; Şeker, Gökmen
2017-09-01
In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99-95% Li20Sn80 + 1-5% RG-Pu, 99-95% Li20Sn80 + 1-5% RG-PuF4, and 99-95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion-fission hybrid reactor system. Beryllium (Be) zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.
Three-dimensional Monte Carlo calculation of some nuclear parameters
Günay Mehtap
2017-01-01
Full Text Available In this study, a fusion-fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20Sn80 + 1-5% RG-Pu, 99–95% Li20Sn80 + 1-5% RG-PuF4, and 99–95% Li20Sn80 + 1-5% RG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be zone with the width of 3 cm was used for the neutron multiplication between the liquid first wall and blanket. This study analyzes the nuclear parameters such as tritium breeding ratio (TBR, energy multiplication factor (M, heat deposition rate, fission reaction rate in liquid first wall, blanket and shield zones and investigates effects of reactor grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.
ADL-3. Nuclear data library for activation and transmutation calculations
Grudzevich, O.T.; Zelenetskij, A.V.; Ignatyuk, A.V.; Pashchenko, A.B.
1995-07-01
It is shown that the use of simplified approaches to calculate threshold neutron reaction cross-sections is not acceptable for the generation of cross-section libraries. Although rigorous models are complex and involve laborious calculations, they provide the only reliable means for evaluating cross-sections when no experimental data are available. A brief description is given of the new version of the library ADL-3 generated by the authors. It contains 18,200 excitation functions of reactions induced by neutrons of up to 20 MeV. The threshold reaction cross-sections have been calculated in the Hauser-Feshbach-Moldauer formalism with allowance for the contribution of non-equilibrium processes. The cross-sections obtained have been tested by comparison with experimental data and evaluations from other libraries. (author)
Hicks, H.G.
1981-11-01
This report presents calculated gamma radiation exposure rates and ground deposition of related radionuclides resulting from three types of event that deposited detectable radioactivity outside the Nevada Test Site complex, namely, underground nuclear detonations, tests of nuclear rocket engines and tests of nuclear ramjet engines
Garcia, T.; Angeles, A.; Flores C, J., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)
2013-10-15
In this work the conditions of nuclear safety were determined as much in normal conditions as in the accident event of the nuclear fuel warehouse of the reactor TRIGA Mark III of the Instituto Nacional de Investigaciones Nucleares (ININ). The warehouse contains standard fuel elements Leu - 8.5/20, a control rod with follower of standard fuel type Leu - 8.5/20, fuel elements Leu - 30/20, and the reactor fuel Sur-100. To check the subcritical state of the warehouse the effective multiplication factor (keff) was calculated. The keff calculation was carried out with the code MCNPX. (Author)
Proportional counter response calculations for gallium solar neutrino detectors
Kouzes, R.T.; Reynolds, D.
1989-01-01
Gallium bases solar neutrino detectors are sensitive to the primary pp reaction in the sun. Two experiments using gallium, SAGE in the Soviet Union and GALLEX in Europe, are under construction and will produce data by 1989. The radioactive /sup 71/Ge produced by neutrinos interacting with the gallium detector material, is chemically extracted and counted in miniature proportional counters. A number of calculations have been carried out to simulate the response of these counters to the decay of /sup 71/Ge and to background events
Use of the Local Variation Methods for Nuclear Design Calculations
Zhukov, A.I.
2006-01-01
A new problem-solving method for steady-state equations, which describe neutron diffusion, is presented. The method bases on a variation principal for steady-state diffusion equations and direct search the minimum of a corresponding functional. Benchmark problem calculation for power of fuel assemblies show ∼ 2% relative accuracy
Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium
Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed
2013-01-01
The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed
Calculation of Lightning Transient Responses on Wind Turbine Towers
Xiaoqing Zhang
2013-01-01
Full Text Available An efficient method is proposed in this paper for calculating lightning transient responses on wind turbine towers. In the proposed method, the actual tower body is simplified as a multiconductor grid in the shape of cylinder. A set of formulas are given for evaluating the circuit parameters of the branches in the multiconductor grid. On the basis of the circuit parameters, the multiconductor grid is further converted into an equivalent circuit. The circuit equation is built in frequency-domain to take into account the effect of the frequency-dependent characteristic of the resistances and inductances on lightning transients. The lightning transient responses can be obtained by using the discrete Fourier transform with exponential sampling to take the inverse transform of the frequency-domain solution of the circuit equation. A numerical example has been given for examining the applicability of the proposed method.
Okuducu, S.; Sarac, H.; Akti, N. N.; Boeluekdemir, M. H.; Tel, E.
2010-01-01
In this study the nuclear energy level density based on nuclear collective excitation mechanism has been identified in terms of the low-lying collective level bands at near the neutron binding energy. Nuclear level density parameters of some light deformed medical radionuclides used widely in medical applications have been calculated by using different collective excitation modes of observed nuclear spectra. The calculated parameters have been used successfully in estimation of the neutron-capture cross section basic data for the production of new medical radionuclides. The investigated radionuclides have been considered in the region of mass number 40< A< 100. The method used in the present work assumes equidistance spacing of the collective coupled state bands of the interest radionuclides. The present calculated results have been compared with the compiled values from the literatures for s-wave neutron resonance data.
Ab Initio Nuclear Structure and Reaction Calculations for Rare Isotopes
Draayer, Jerry P. [Louisiana State Univ., Baton Rouge, LA (United States)
2014-09-28
We have developed a novel ab initio symmetry-adapted no-core shell model (SA-NCSM), which has opened the intermediate-mass region for ab initio investigations, thereby providing an opportunity for first-principle symmetry-guided applications to nuclear structure and reactions for nuclear isotopes from the lightest p-shell systems to intermediate-mass nuclei. This includes short-lived proton-rich nuclei on the path of X-ray burst nucleosynthesis and rare neutron-rich isotopes to be produced by the Facility for Rare Isotope Beams (FRIB). We have provided ab initio descriptions of high accuracy for low-lying (including collectivity-driven) states of isotopes of Li, He, Be, C, O, Ne, Mg, Al, and Si, and studied related strong- and weak-interaction driven reactions that are important, in astrophysics, for further understanding stellar evolution, X-ray bursts and triggering of s, p, and rp processes, and in applied physics, for electron and neutrino-nucleus scattering experiments as well as for fusion ignition at the National Ignition Facility (NIF).
Nuclear matter calculations with a pseudoscalar-pseudovector chiral model
Niembro, R.; Marcos, S.; Bernardos, P. [University of Cantabria, Faculty of Sciences, Department of Modern Physics, 39005 Santander (Spain); Fomenko, V.N. [St Petersburg University for Railway Engineering, Department of Mathematics, 197341 St Petersburg (Russian Federation); Savushkin, L.N. [St Petersburg University for Telecomunications, Department of Physics, 191065 St Petersburg (Russian Federation); Lopez-Quelle, M. [University of Cantabria, Faculty of Sciences, Department of Applied Physics, 39005 Santander, Spain (Spain)
1998-10-01
A mixed pseudoscalar-pseudovector {pi}N coupling relativistic Lagrangian is obtained from a pure pseudoscalar chiral one, by transforming the nucleon field according to a generalized Weinberg transformation, which depends on a mixing parameter. The interaction is generated by the {sigma}, {omega} and {pi} meson exchanges. Within the Hartree-Fock context, pion polarization effects, including the {delta} isobar, are considered in the random phase approximation in nuclear matter. These effects are interpreted, in a non-relativistic framework, as a modification of the range and intensity of a Yukawa-type potential by means of a simple function which takes into account the nucleon-hole and {delta}-hole excitations. Results show stability of relativistic nuclear matter against pion condensation. Compression modulus is diminished by the combined effects of the nucleon and {delta} polarization towards the usually accepted experimental values. The {pi}N interaction strength used in this paper is less than the conventional one to ensure the viability of the model. The fitting parameters of the model are the scalar meson mass m{sub {sigma}} and the {omega}-N coupling constant g{sub {omega}}. (author)
Ab Initio Nuclear Structure and Reaction Calculations for Rare Isotopes
Draayer, Jerry P.
2014-01-01
We have developed a novel ab initio symmetry-adapted no-core shell model (SA-NCSM), which has opened the intermediate-mass region for ab initio investigations, thereby providing an opportunity for first-principle symmetry-guided applications to nuclear structure and reactions for nuclear isotopes from the lightest p-shell systems to intermediate-mass nuclei. This includes short-lived proton-rich nuclei on the path of X-ray burst nucleosynthesis and rare neutron-rich isotopes to be produced by the Facility for Rare Isotope Beams (FRIB). We have provided ab initio descriptions of high accuracy for low-lying (including collectivity-driven) states of isotopes of Li, He, Be, C, O, Ne, Mg, Al, and Si, and studied related strong- and weak-interaction driven reactions that are important, in astrophysics, for further understanding stellar evolution, X-ray bursts and triggering of s, p, and rp processes, and in applied physics, for electron and neutrino-nucleus scattering experiments as well as for fusion ignition at the National Ignition Facility (NIF).
Therapeutic Applications of Monte Carlo Calculations in Nuclear Medicine
Coulot, J
2003-01-01
Monte Carlo techniques are involved in many applications in medical physics, and the field of nuclear medicine has seen a great development in the past ten years due to their wider use. Thus, it is of great interest to look at the state of the art in this domain, when improving computer performances allow one to obtain improved results in a dramatically reduced time. The goal of this book is to make, in 15 chapters, an exhaustive review of the use of Monte Carlo techniques in nuclear medicine, also giving key features which are not necessary directly related to the Monte Carlo method, but mandatory for its practical application. As the book deals with therapeutic' nuclear medicine, it focuses on internal dosimetry. After a general introduction on Monte Carlo techniques and their applications in nuclear medicine (dosimetry, imaging and radiation protection), the authors give an overview of internal dosimetry methods (formalism, mathematical phantoms, quantities of interest). Then, some of the more widely used Monte Carlo codes are described, as well as some treatment planning softwares. Some original techniques are also mentioned, such as dosimetry for boron neutron capture synovectomy. It is generally well written, clearly presented, and very well documented. Each chapter gives an overview of each subject, and it is up to the reader to investigate it further using the extensive bibliography provided. Each topic is discussed from a practical point of view, which is of great help for non-experienced readers. For instance, the chapter about mathematical aspects of Monte Carlo particle transport is very clear and helps one to apprehend the philosophy of the method, which is often a difficulty with a more theoretical approach. Each chapter is put in the general (clinical) context, and this allows the reader to keep in mind the intrinsic limitation of each technique involved in dosimetry (for instance activity quantitation). Nevertheless, there are some minor remarks to
Moeller, M. P.; Urbanik, II, T.; Desrosiers, A. E.
1982-03-01
This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuatlon tlmes for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies.
Moeller, M.P.; Desrosiers, A.E.; Urbanik, T. II
1982-03-01
This paper describes the methodology and application of the computer model CLEAR (Calculates Logical Evacuation And Response) which estimates the time required for a specific population density and distribution to evacuate an area using a specific transportation network. The CLEAR model simulates vehicle departure and movement on a transportation network according to the conditions and consequences of traffic flow. These include handling vehicles at intersecting road segments, calculating the velocity of travel on a road segment as a function of its vehicle density, and accounting for the delay of vehicles in traffic queues. The program also models the distribution of times required by individuals to prepare for an evacuation. In order to test its accuracy, the CLEAR model was used to estimate evacuation times for the emergency planning zone surrounding the Beaver Valley Nuclear Power Plant. The Beaver Valley site was selected because evacuation time estimates had previously been prepared by the licensee, Duquesne Light, as well as by the Federal Emergency Management Agency and the Pennsylvania Emergency Management Agency. A lack of documentation prevented a detailed comparison of the estimates based on the CLEAR model and those obtained by Duquesne Light. However, the CLEAR model results compared favorably with the estimates prepared by the other two agencies. (author)
Nuclear structure calculations in the dynamic-interaction propagator approach
Engelbrecht, C.A.; Hahne, F.J.W.; Heiss, W.D.
1978-01-01
The dynamic-interaction propagator approach provides a natural method for the handling of energy-dependent effective two-body interactions induced by collective excitations of a many-body system. In this work this technique is applied to the calculation of energy spectra and two-particle strengths in mass-18 nuclei. The energy dependence is induced by the dynamic exchange of the lowest 3 - octupole phonon in O 16 , which is described within a normal static particle-hole RPA. This leads to poles in the two-body self-energy, which can be calculated if other fermion lines are restricted to particle states. The two-body interaction parameters are chosen to provide the correct phonon energy and reasonable negative-parity mass-17 and positive-parity mass-18 spectra. The fermion lines must be dressed consistently with the same exchange phonon to avoid redundant solutions or ghosts. The negative-parity states are then calculated in a parameter-free way which gives good agreement with the observed spectra [af
Inventory charge calculations in the nuclear fuel cycle
Salmon, R.
1975-09-01
Simplified methods are presented for the calculation of inventory charges or carrying charges on fuel, which represent the indirect component of the fuel cycle cost. These methods permit rapid calculation of the changes in fuel cycle cost caused by changes in the amount or timing of fuel cycle expenditures. The methods are developed by applying the discounted cash flow procedure to a single batch of fuel. In typical cases, this would be a batch representing equilibrium or steady-state reactor operation. The cost equations used are the same as those used in the computer code REFCO, described in ORNL-4695, which was based on the discounted cash flow procedure with continuous discounting. Equivalent procedures using the fixed charge rate concept also are developed. This is done in such a way that consistency with the discounted cash flow procedure is maintained. The fixed charge rate used here is defined in terms of tax rates and the interest rates on debt and equity capital. An effective inventory time is also defined. This is a function of the lead or lag time, the interest rates on capital, and the exposure time of the batch. Tabulated values of the effective inventory time and other useful functions, such as the ratio of indirect to direct cost, are included. Cost calculations using these tables agree with those produced by REFCO, the accuracy being within 0.001 mill/kWhr in the cases studied. (U.S.)
Hydraulic simulation of the systems of a nuclear power plant for charges calculation in piping
Masriera, N.
1990-01-01
This work presents a general description of the methodology used by the ENACE S.A. Fluids Working Group for hydraulics simulation of a nuclear power plant system for the calculation charges in piping. (Author) [es
Sensitivity of nuclear power plant structural response to aircraft impact
Buchhardt, F.; Magiera, G.; Matthees, W.; Weber, M.
1984-01-01
In this paper a sensitivity study for aircraft impact is performed concerning the excitation of internal components, with particular regard to nonlinear structural material behaviour in the impact area. The nonlinear material values are varied within the bandwidth of suitable material strength, depending on local stiffness pre-calculations. The analyses are then performed on a globally discretized three-dimensional finite element model of a nuclear power plant, using a relatively fine mesh. For specified nodal points results are evaluated by comparing their response spectra. (Author) [pt
ANL calculational methodologies for determining spent nuclear fuel source term
McKnight, R. D.
2000-01-01
Over the last decade Argonne National Laboratory has developed reactor depletion methods and models to determine radionuclide inventories of irradiated EBR-II fuels. Predicted masses based on these calculational methodologies have been validated using available data from destructive measurements--first from measurements of lead EBR-II experimental test assemblies and later using data obtained from processing irradiated EBR-II fuel assemblies in the Fuel Conditioning Facility. Details of these generic methodologies are described herein. Validation results demonstrate these methods meet the FCF operations and material control and accountancy requirements
Dirac-Fock atomic electronic structure calculations using different nuclear charge distributions
Visscher, L; Dyall, KG
1997-01-01
Numerical Hartree-Fock calculations based on the Dirac-Coulomb Hamiltonian for the first 109 elements of the periodic table are presented. The results give the total electronic energy, as a function of the nuclear model that is used, for four different models of the nuclear charge distribution. The
Benchmarking nuclear models for Gamow–Teller response
Litvinova, E.; Brown, B.A.; Fang, D.-L.; Marketin, T.; Zegers, R.G.T.
2014-01-01
A comparative study of the nuclear Gamow–Teller response (GTR) within conceptually different state-of-the-art approaches is presented. Three nuclear microscopic models are considered: (i) the recently developed charge-exchange relativistic time blocking approximation (RTBA) based on the covariant density functional theory, (ii) the shell model (SM) with an extended “jj77” model space and (iii) the non-relativistic quasiparticle random-phase approximation (QRPA) with a Brueckner G-matrix effective interaction. We study the physics cases where two or all three of these models can be applied. The Gamow–Teller response functions are calculated for 208 Pb, 132 Sn and 78 Ni within both RTBA and QRPA. The strengths obtained for 208 Pb are compared to data that enable a firm model benchmarking. For the nucleus 132 Sn, also SM calculations are performed within the model space truncated at the level of a particle–hole (ph) coupled to vibration configurations. This allows a consistent comparison to the RTBA where ph⊗phonon coupling is responsible for the spreading width and considerable quenching of the GTR. Differences between the models and perspectives of their future developments are discussed.
Benchmarking nuclear models for Gamow–Teller response
Litvinova, E., E-mail: elena.litvinova@wmich.edu [Department of Physics, Western Michigan University, Kalamazoo, MI 49008-5252 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Brown, B.A. [Department of Physics and Astronomy, Michigan State University, East Lansing, MI 48824-1321 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Fang, D.-L. [National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Joint Institute for Nuclear Astrophysics, Michigan State University, East Lansing, MI 48824-1321 (United States); Marketin, T. [Physics Department, Faculty of Science, University of Zagreb (Croatia); Zegers, R.G.T. [Department of Physics and Astronomy, Michigan State University, East Lansing, MI 48824-1321 (United States); National Superconducting Cyclotron Laboratory, Michigan State University, East Lansing, MI 48824-1321 (United States); Joint Institute for Nuclear Astrophysics, Michigan State University, East Lansing, MI 48824-1321 (United States)
2014-03-07
A comparative study of the nuclear Gamow–Teller response (GTR) within conceptually different state-of-the-art approaches is presented. Three nuclear microscopic models are considered: (i) the recently developed charge-exchange relativistic time blocking approximation (RTBA) based on the covariant density functional theory, (ii) the shell model (SM) with an extended “jj77” model space and (iii) the non-relativistic quasiparticle random-phase approximation (QRPA) with a Brueckner G-matrix effective interaction. We study the physics cases where two or all three of these models can be applied. The Gamow–Teller response functions are calculated for {sup 208}Pb, {sup 132}Sn and {sup 78}Ni within both RTBA and QRPA. The strengths obtained for {sup 208}Pb are compared to data that enable a firm model benchmarking. For the nucleus {sup 132}Sn, also SM calculations are performed within the model space truncated at the level of a particle–hole (ph) coupled to vibration configurations. This allows a consistent comparison to the RTBA where ph⊗phonon coupling is responsible for the spreading width and considerable quenching of the GTR. Differences between the models and perspectives of their future developments are discussed.
Task 4 Improvised Nuclear Device Response Curves
Alai, Maureen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Neuscamman, Stephanie [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2016-05-31
LLNL performed fallout and nuclear blast modeling for the 60 cities using the NARAC modeling system and predominant weather patterns determined in a previous Task 4 effort. LLNL performed model simulations and analyses to identify and provide response curves (expressed as two-dimensional contours) for radioactive fallout deposition, transport, population, and blast overpressure as a function of yield, weather, location and time. These contours can then be further combined and correlated with infrastructure and population databases to estimate city specific effects on KPFs such as impacted infrastructure and casualty rates.
Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code
Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu
2014-01-01
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)
Koo, Bon Seung; Lee, Kyung Hoon; Song, Jae Seung; Park, Sang Yoon
2013-01-01
In this paper, the basic nuclear characteristics of major emitter materials were surveyed. In addition, preliminary calculations of Cobalt-Vanadium fixed incore detector were performed using the Monte Carlo code. Calculational results were cross-checked by KARMA. KARMA is a two-dimensional multigroup transport theory code developed by the KAERI and approved by Korean regularity agency to be employed as a nuclear design tool for a Korean commercial pressurizer water reactor. The nuclear characteristics of the major emitter materials were surveyed, and preliminary calculations of the hybrid fixed incore detector were performed with the MCNP code. The eigenvalue and pin-by-pin fission power distributions were calculated and showed good agreement with the KARMA calculation results. As future work, gamma power distributions as well as several types of XS of the emitter, insulator, and collector regions for a Co-V ICI assembly will be evaluated and compared
Electromagnetically induced nuclear beta decay calculated by a Green's function method
Reiss, H.R.
1984-01-01
The transition probability for enhancement of forbidden nuclear beta decay by an applied plane-wave electromagnetic field is calculated in a nonrelativistic spinless approximation by a Green's function method. The calculation involves a stationary-phase approximation. The stationary phase points in the presence of an intense field are located in very different positions than they are in the field-free case. In order-of-magnitude terms, the results are completely consistent with an earlier, much more complete wave-function calculation which includes spin and relativistic effects. Both the present Green's function calculation and the earlier wave function calculation give electromagnetic contributions in first-forbidden nuclear beta decay matrix elements which are of order (R 0 /lambda-dash-bar/sub C/) 2 with respect to allowed decays, where R 0 is the nuclear radius and lambda-dash-bar/sub C/ is the electron Compton wavelength
Some Qualitative Requirements for Testing of Nuclear Emergency Response Robots
Eom, Heungseop; Cho, Jai Wan; Choi, Youngsoo; Jeong, Kyungmin
2014-01-01
Korea Atomic Energy Research Institute (KAERI) is carrying out the project 'Development of Core Technology for Remote Response in Nuclear Emergency Situation', and as a part of the project, we are studying the reliability and performance requirements of nuclear emergency response robots. In this paper, we described some qualitative requirements for testing of nuclear emergency response robots which are different to general emergency response robots. We briefly introduced test requirements of general emergency response robots and described some qualitative aspects of test requirements for nuclear emergency response robots. When considering an immature field-robot technology and variety of nuclear emergency situations, it seems hard to establish quantitative test requirements of these robots at this time. However, based on studies of nuclear severe accidents and the experience of Fukushima NPP accident, we can expect some test requirements including quantitative ones for nuclear emergency response robots
Calculations of nuclear energies using the energy density formalism
Pu, W.W.T.
1975-01-01
The energy density formalism (EDF) is used to investigate two problems. In this formalism the energy of the nucleus is expressed as a functional of its density. The nucleus energy is obtained by minimizing the functional with respect to the density. The first problem has to do with the stability of nuclei having shapes of different degrees of central depression (bubble shapes). It is shown that the bubble shapes are energetically favorable only for unrealistically large nuclei. Particularly, the super heavy nucleus that has been suggested (Z = 114, N = 184) prefers a shape with constant central density. These results are in good agreement with earlier calculations using the liquid drop model. The second problem concerns an anomaly detected experimentally in the isotope shift of mercury. The isotope shifts among a long chain of mercury isotopes show a sudden change as the neutron number is reduced. In particular, the experimental result suggests that the effective size of the charge distributions of 183 Hg and 185 Hg are as large as that of 196 Hg. Such sudden changes in other nuclei have been attributed to a sudden onset of permanent quadruple deformation. In the case of mercury there is no experimental evidence for deformed shapes. It was, therefore, suggested that the proton distribution might develop a central depression in the lighter isotopes. The EDF is used to investigate the mercury isotope shift anomaly following the aforementioned suggestion. Specifically, nucleon densities with different degrees of central depression are generated. Energies corresponding to these densities are obtained. To allow for shell effects, nucleon densities are obtained from single-particle wave functions. Calculations are made for a few mercury isotopes, especially for 184 Hg. The results are that in all cases the energy is lower for densities corresponding to a solid spherical shape
Ruud, Kenneth; Helgaker, Trygve; Kobayashi, Rika
1994-01-01
to corresponding individual gauges for localized orbitals (IGLO) results. The London results show better basis set convergence than IGLO, especially for heavier atoms. It is shown that the choice of active space is crucial for determination of accurate nuclear shielding constants.......Nuclear shielding calculations are presented for multiconfigurational self-consistent field wave functions using London atomic orbitals (gauge invariant atomic orbitals). Calculations of nuclear shieldings for eight molecules (H2O, H2S, CH4, N2, CO, HF, F2, and SO2) are presented and compared...
Political pressure on nuclear - responsibility or business?
Petrech, Rastislav; Holy, Robert
2001-01-01
Environmental Impact Assessment study prior to bringing the plant in a commercial operation - will change nothing. I believe that the outcomes will reinforce the arguments in favour of construction and commissioning of new sources of clean, safe, and reliable generation of electricity. Is it just diverting the public's attention from other problems or an attempt to gain a political capital? Is the political pressure on nuclear a real responsibility or a sheer business? Most politicians have no idea about the nuclear power, and if they do, it is only very misty one. It is necessary to invite them to visit plants, discuss with them and give them clear arguments. A clear statement of an energy policy and relying on the safe and economic advantageous nuclear power can also attract a large portion of electors in the future. We should try so that energy policy in post-communist countries got into election programmes of each political party, since the energy mix policy cannot be tailored only for one election period of a government, but for a time of 20 to 30 years in advance
Critical seismic response of nuclear reactors
Drenick, R.F.; Wang, P.C.; Yun, C.B.; Philappacopoulos, A.J.
1980-01-01
This paper deals with the problem of how to assess the seismic resistance of important structures, particularly of nuclear reactor structures. Most of the design procedures presently in use or under investigation are based on design response spectra obtained by statistical evaluation of past ground excitations assuming certain probability distributions (e.g., normal or lognormal) of response peaks. Artificial time histories generated from these spectra are also used. However, it is not clear whether these approaches lead to designs that can be relied upon on the confidence levels that are presumably desired for structures such as nuclear reactors whose integrity during an earthquake is of considerable importance. The trouble lies with the fact that the analysis of structural integrity seems highly sensitive to the assumptions regarding the nature of those probability distributions: small variations, especially in the tails of the distributions, can induce large changes in the desired results. This greatly weakens the reliance that can be placed in many assessments of earthquake resistance. In this paper, a new method is developed which has the potential of avoiding those weaknesses. It is more specifically based on assumptions that seem well supported by seismological observations but side-steps others, especially those regarding probability distribution of ground motions, which are more conjectural. (orig./RW)
Monte Carlo calculation of the nuclear temperature coefficient in fast reactors
Matthes, W.
1974-04-15
A Monte Carlo program for the calculation of the nuclear temperature coefficient for fast reactors is described. The special difficulties for this problem are the energy and space dependence of the cross sections and the calculation of differential eifects. These difficulties are discussed in detail and the way for their solution chosen in this program is described. (auth)
Improvements to the nuclear model code GNASH for cross section calculations at higher energies
Young, P.G.; Chadwick, M.B.
1994-01-01
The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV
Nuclear decay data for dosimetry calculation. Revised data of ICRP Publication 38
Endo, Akira; Yamaguchi, Yasuhiro
2005-02-01
New nuclear decay data used for dose calculation have been compiled for 1034 radionuclides, which are significant in medical, environmental and occupational exposures. The decay data were assembled from decay data sets of the Evaluated Nuclear Structure Data File (ENSDF), the latest version as of 2003. Basic nuclear properties in the ENSDF that are particularly important for calculating energies and intensities of radiations were examined and updated by referring to UNBASE2003/AME2003, the database for nuclear and decay properties of nuclides. In addition, modification of incomplete ENSDF was done for their format errors, level schemes, normalization records, and so on. The energies and intensities of emitted radiations by the nuclear decay and the subsequent atomic process were computed from the ENSDF using the computer code EDISTR04. EDISTR04 is an enhanced version of EDISTR used for assembling ICRP Publication 38 (ICRP38), and incorporates updates of atomic data and computation methods for calculating atomic radiations and spontaneous fission radiations. Quality assurance of the compiled data has been made by comparisons with various experimental data and decay databases prepared from different computer codes and data libraries. A package of the data files, called DECDC2 (Nuclear DECay Data for Dosimetry Calculation, Version 2), will succeed ICRP38 that has been used extensively in dose calculation and will be utilized in various fields. (author)
Koning, A.
2008-01-01
Full text: Masses: Adopted Goriely HFB masses in TALYS as theoretical default instead of Moeller. Audi-Wapstra, Moeller and HFB masses tested formally with TALYS. Levels. Adopted latest discrete level update (2006) by Belgya (as sent by Capote) in TALYS. Tested with TALYS. Resonances. Adopted RIPL-2 D0 collection in TALYS. Tested by TALYS. Optical model. Coordinated Optical model segment for RIPL-3. Adopted Soukhovitskii CC potential as default for actinides. Covariances: Confirmed OMP parameter uncertainties from last meeting. Level density. Produced consistent set of level density parameters for CTM, BFM, GSM and HFM. Local models (per nucleus) and global models (systematics). With and without effective collective enhancement. Included and tested with TALYS Gamma-ray strength. Adopted Goriely HFB strength function tables as option (not default) in TALYS. Both formally tested and validated with TALYS. Fission. Adopted Sin-Capote WKB approximation in TALYS as option for fission calculations. Formally tested. RIPL-2/3 validation. Very extensive formal tests and validation procedures with TALYS. MONKEY code for random input files (has found RIPL errors in the past). Automatic comparison with all available EXFOR cross section data (for level density study). Started work on global parameter uncertainties (for covariances). SALTY nuclear data library (final version under construction): - 60 MeV n,g,p,d,t,h,a activation files for 1200 nuclides - 200 MeV n,g,p,d,t,h,a transport files for 250 nuclides RIPL is automatically being used by all TALYS users (and TALYS-related publications). TALYS-1.0 release in December 2007 (delay because of level densities). (author)
Financing responsibility for nuclear waste disposal
2004-01-01
The basic premise for financing arrangements for the disposal of nuclear waste is that the nuclear industry - not the taxpayer - must bear the costs. Present regulations, however, are imperfect in this regard. The Inquiry therefore proposes extending the financial liability of the nuclear industry and introducing new fee-setting arrangements. It is proposed that a new law be enacted to regulate these changes. The present financing system is regulated in the 'Financing Act' 1. Under this Act, the licensed owner and operator of a nuclear reactor is required to pay an annual fee and provide guarantees to the State. Four companies are reactor owners. These companies are wholly or partly owned by other companies according to various arrangements. Each reactor owner is responsible for its own dismantling costs and for its share of allocated common costs of disposal and related measures. If there is insufficient money in the funds, the nuclear industry will still be liable. The basic premise of the Inquiry is that the financing system should be designed so as to minimise the risk that the State (and taxpayers) will need to step in and pay. Although the nuclear industry is intended to have full liability for payment, in practice it does not. This is because the formal full liability for payment in the nuclear industry rests with the reactor companies and not where the industry's long-term ability to pay is to be found. Essentially, the present arrangements mean that: - Companies that cannot be expected to have any long-term ability to pay have unlimited liability, and - Companies that can be expected to have an ability to pay have very limited liability. The Inquiry therefore proposes that ability to pay and liability are brought into line by a formal assumption by owning companies of the sort of liability for payment that now rests solely with the reactor companies. This means that the owning company in each group that is best suited to bear the liability for payment
Crabol, B.; Manesse, D.; Robeau, D.
1989-07-01
The available calculation tools of the Crisis Technical Center (CTC), for the analysis and evaluation of radiation effects from a nuclear accident, are presented. The CTC calculation unit depends on local means, and on the National Meteorology system, in order to collect the data needed for the atmospheric waste diffusion evaluation. For the radiation dose calculations, plotters and software allowing the analysis of all waste Kinetics and all the meteorological conditions are available. The work developed by CTC calculation unit enables an easy application of the calculation tools as well as the results obtention. Images from data bases are provided to complete the obtained results [fr
The nuclear heating calculation scheme for material testing in the future Jules Horowitz Reactor
Huot, N.; Aggery, A.; Blanchet, D.; Courcelle, A.; Czernecki, S.; Di-Salvo, J.; Doederlein, C.; Serviere, H.; Willermoz, G.
2004-01-01
An innovative nuclear heating calculation scheme for materials testing carried out in in the future Jules Horowitz reactor (JHR) is described. A heterogeneous gamma source calculation is first performed at assembly level using the deterministic code APOLLO2. This is followed by a Monte Carlo gamma transport calculation in the whole core using the TRIPOLI4 code. The calculated gamma sources at the assembly level are applied in the whole core simulation using a weighting based on power distribution obtained from the neutronic core calculation. (authors)
Radionuclide composition in nuclear fuel waste. Calculations performed by ORIGEN2
Lyckman, C.
1996-01-01
The report accounts for results from calculations on the content of radionuclides in nuclear fuel waste. It also accounts for the results from calculations on the neutron flow from spent fuel, which is very important during transports. The calculations have been performed using the ORIGEN2 software. The results have been compared to other results from earlier versions of ORIGEN and some differences have been discovered. This is due to the updating of the software. 7 refs, 10 figs, 15 tabs
Effective calculation algorithm for nuclear chains of arbitrary length and branching
Chirkov, V.A.; Mishanin, B.V.
1994-01-01
An effective algorithm for calculation of the isotope concentration in the spent nuclear fuel when it is kept in storage, is presented. Using the superposition principle and representing the transfer function in a rather compact form it becomes possible achieve high calculation speed and a moderate computer code size. The algorithm is applied for the calculation of activity, energy release and toxicity of heavy nuclides and products of their decay when the fuel is kept in storage. (authors). 1 ref., 4 tabs
Emergency response and nuclear risk governance. Nuclear safety at nuclear power plant accidents
Kuhlen, Johannes
2014-01-01
The present study entitled ''Emergency Response and Nuclear Risk Governance: nuclear safety at nuclear power plant accidents'' deals with issues of the protection of the population and the environment against hazardous radiation (the hazards of nuclear energy) and the harmful effects of radioactivity during nuclear power plant accidents. The aim of this study is to contribute to both the identification and remediation of shortcomings and deficits in the management of severe nuclear accidents like those that occurred at Chernobyl in 1986 and at Fukushima in 2011 as well as to the improvement and harmonization of plans and measures taken on an international level in nuclear emergency management. This thesis is divided into a theoretical part and an empirical part. The theoretical part focuses on embedding the subject in a specifically global governance concept, which includes, as far as Nuclear Risk Governance is concerned, the global governance of nuclear risks. Due to their characteristic features the following governance concepts can be assigned to these risks: Nuclear Safety Governance is related to safety, Nuclear Security Governance to security and NonProliferation Governance to safeguards. The subject of investigation of the present study is as a special case of the Nuclear Safety Governance, the Nuclear Emergency governance, which refers to off-site emergency response. The global impact of nuclear accidents and the concepts of security, safety culture and residual risk are contemplated in this context. The findings (accident sequences, their consequences and implications) from the analyses of two reactor accidents prior to Fukushima (Three Mile Iceland in 1979, Chernobyl in 1986) are examined from a historical analytical perspective and the state of the Nuclear Emergency governance and international cooperation aimed at improving nuclear safety after Chernobyl is portrayed by discussing, among other topics, examples of &apos
Particle-hole calculation of the longitudinal response function of 12C
Dellafiore, A.; Lenz, F.; Brieva, F.A.
1985-01-01
The longitudinal response function of 12 C in the range of momentum transfers 200 MeV/c< or =q< or =550 MeV/c is calculated in the Tamm-Dancoff approximation. The particle-hole Green's function is evaluated by means of a doorway-state expansion. This method allows us to take into account finite-range residual interactions in the continuum, including exchange processes. At low momentum transfers, calculations agree qualitatively with the data. The data cannot be reproduced at momentum transfers around 450 MeV/c. This discrepancy can be accounted for neither by uncertainties in the residual interaction, nor by more complicated processes in the nuclear final states
Marguet, S.D.
1993-01-01
Owing to the prevalence in France of nuclear generated electricity, the french utility, EDF focuses much research on fuel cycle strategy. In this context, analysis of scenarios combining problems related to planning and economics, but also reactor physics, necessitate a relatively thorough understanding of fuel response to irradiation. The main purpose of the fuel strategy program codes is to predict mass balance modifications with time for the main actinides involved in the cycle, including the minor actinides associated with the current back end fuel cycle key issues. Considering the large number of calculations performed by a strategy code in an iterative process covering a range of about a hundred years, it was important to develop basic computation modules for both the ''reactor'' and ''fabrication'' items. These had to be high speed routines, but on an accuracy level compatible with the strategy code efficiency. At the end of 1992, the EDF Research and Development Division (EDF/DER) developed a very simple, extremely fast method of calculating transuranian isotope masses. This approach, which resulted in the STRAPONTIN software, considerably increased the scope of the EDF/DER fuel strategy code TIRELIRE without undue impairment of machine time requirements for a scenario. (author). 2 figs., 2 tabs., 3 refs
Nuclear spin response studies in inelastic polarized proton scattering
Jones, K.W.
1988-01-01
Spin-flip probabilities S/sub nn/ have been measured for inelastic proton scattering at incident proton energies around 300 MeV from a number of nuclei. At low excitation energies S/sub nn/ is below the free value. For excitation energies above about 30 MeV for momentum transfers between about 0.35 fm/sup /minus/1/ and 0.65 fm/sup / minus/1/ S/sub nn/ exceeds free values significantly. These results suggest that the relative ΔS = 1(ΔS = 0 + ΔS = 1) nuclear spin response approaches about 90% in the region of the enhancement. Comparison of the data with slab response calculations are presented. Decomposition of the measured cross sections into σ(ΔS = 0) and σ(ΔS = 1) permit extraction of nonspin-flip and spin-flip dipole and quadrupole strengths. 29 refs., 11 figs
Planning and implementing nuclear emergency response facilities
Williams, D.H.
1983-01-01
After Three Mile Island, Arkansas Nuclear One produced a planning document called TMI-2 Response Program. Phase I of the program defined action plans in nine areas: safety assessment, training, organization, public information, communication, security, fiscal-governmental, technical and logistical support. Under safety assessment, the staff was made even better prepared to handle radioactive material. Under training, on site simulators for each unit at ANO were installed. The other seven topics interface closely with each other. An emergency control center is diagrammed. A habitable technical support system was created. A media center, with a large media area, and an auditorium, was built. Electric door strike systems increased security. Phone networks independently run via microwave were installed. Until Three Mile Island, logistical problems were guesswork. That incident afforded an opportunity to better identify and prepare for these problems
CO2 impulse response curves for GWP calculations
Jain, A.K.; Wuebbles, D.J.
1993-01-01
The primary purpose of Global Warming Potential (GWP) is to compare the effectiveness of emission strategies for various greenhouse gases to those for CO 2 , GWPs are quite sensitive to the amount of CO 2 . Unlike all other gases emitted in the atmosphere, CO 2 does not have a chemical or photochemical sink within the atmosphere. Removal of CO 2 is therefore dependent on exchanges with other carbon reservoirs, namely, ocean and terrestrial biosphere. The climatic-induced changes in ocean circulation or marine biological productivity could significantly alter the atmospheric CO 2 lifetime. Moreover, continuing forest destruction, nutrient limitations or temperature induced increases of respiration could also dramatically change the lifetime of CO 2 in the atmosphere. Determination of the current CO 2 sinks, and how these sinks are likely to change with increasing CO 2 emissions, is crucial to the calculations of GWPs. It is interesting to note that the impulse response function is sensitive to the initial state of the ocean-atmosphere system into which CO 2 is emitted. This is due to the fact that in our model the CO 2 flux from the atmosphere to the mixed layer is a nonlinear function of ocean surface total carbon
Extension of responsibilities of the State Office for Nuclear Safety
Hrehor, M.
1995-01-01
The responsibilities of the State Office for Nuclear Safety have been extended by Act No. 85/1995 to cover protection against ionizing radiation. The following responsibilities of the State Office for Nuclear Safety are defined by the Act: a) state surveillance over nuclear safety of nuclear facilities, and over radioactive waste and spent fuel management; b) state surveillance over nuclear materials, their record-keeping and accountancy; c) state surveillance over the safeguarding of nuclear facilities and nuclear materials; d) state surveillance over selected materials, facilities and technologies used in the nuclear field, as well as dual-purpose materials and facilities; e) state surveillance over protection against ionizing radiation; f) coordination of the performance of the Radiation Monitoring Network over the Czech Republic and responsibility for international exchange of data on the radiological situation. The Act is reproduced in full, and the organizational structure of the Office is shown in a chart. (J.B.)
Response function of spin-isospin nuclear excitations
Salvetti, A.R.
1986-01-01
The selected aspects of spin-isospir nuclear excitations are studied. The spreading width of M/ states in even Ca isotopes for the purpose of trying to understand the missing strenght specially in 44 Ca, was estimated. The doorway calculation, was used, considering the level of complexity next to the independent particle M/ state. Using a nuclear matter context, the system response function to a spin-isospin probe and verify how the response function behaves for free fermions and in the ring approximation was studied. Higher correlations to polarization propagation such as the induced interaction and self-energy corrections was introduced. The dopping of colletive effects by such collisions terms was verified. It was investigate how to estimate the short range term of the effective interaction in the spin-isospin channel and the possibility of detecting a difference between these short range terms in the longitudinal and the transverse channel, for understanding the absence of pior condensation precursor states and negative results in a recent attempt to detect differences between longitudinal and transverse response functions one naively expects theoretically. (author) [pt
Diaz, Daniel Suescun
2007-07-01
This work presents two new methods for the solution of the inverse point kinetics equation. The first method is based on the integration by parts of the integral of the inverse point kinetics equation, which results in a power series in terms of the nuclear power in time dependence. Applying some conditions to the nuclear power, the reactivity is represented as first and second derivatives of this nuclear power. This new calculation method for reactivity has special characteristics, amongst which the possibility of using different sampling periods, and the possibility of restarting the calculation, after its interruption associated it with a possible equipment malfunction, allowing the calculation of reactivity in a non-continuous way. Apart from this reactivity can be obtained with or without dependency on the nuclear power memory. The second method is based on the Laplace transform of the point kinetics equations, resulting in an expression equivalent to the inverse kinetics equation as a function of the power history. The reactivity can be written in terms of the summation of convolution with response to impulse, characteristic of a linear system. For its digital form the Z-transform is used, which is the discrete version of the Laplace transform. In this method it can be pointed out that the linear part is equivalent to a filter named Finite Impulse Response (Fir). The Fir filter will always be, stable and non-varying in time, and, apart from this, it can be implemented in the non-recursive way. This type of implementation does not require feedback, allowing the calculation of reactivity in a continuous way. The proposed methods were validated using signals with random noise and showing the relationship between the reactivity difference and the degree of the random noise. (author)
Model calculations of the influence of population distribution on the siting of nuclear power plants
Nielsen, F.; Walmod-Larsen, O.
1984-02-01
This report was prepared for a working group established in April 1981 by the Danish Environmental Protection Agency with the task of investigating siting problems of nuclear power stations in Denmark. The purpose of the working group was to study the influence of the population density around a site on nuclear power safety. The importance of emergency planning should be studied as well. In this model study two specific accident sequences were simulated on a 1000 MWe nuclear power plant. The plant was assumed to be placed in the center of two different model population distributions. The concequences for the two population distributions from the two accidents were calculated for the most frequent weather conditions. Doses to individuals were calculated for the bone marrow, lungs, gastrointestinal tract, thyroidea and for the whole body. The collective whole body doses were also calculated for the two populations considered. (author)
Zhu Zhenghe; Luo Deli; Feng Kaiming
2013-01-01
The present work is to calculate the magnetic thermodynamically functions, i.e. energy, the intensity of magnetization, enthalpy, entropy and Gibbs function for nuclear magnetic moments of T, D and neutron n at 2 T and 1, 50, 100 and 150 K from partition functions. It is shown that magnetic saturation of thermonuclear plasma does not easily occur for nuclear magneton is only of 10 -3 of Bohr magneton. The work done by magnetic field is considerable. (authors)
Marzo, M.A.S.
1986-01-01
The INSPECT software package was developed in the Pacific Northwest Laboratory for statistical calculations in nuclear material accountability. The programs apply the inspection and evaluation methodology described in Part of the Safeguards Technical Manual. In this paper the implementation of INSPECT at the Safeguards Division of CNEN, and the main characteristics of INSPECT are described. The potential applications of INSPECT to the nuclear material accountability is presented. (Author) [pt
Bettes, R.S.
1984-01-01
The paper discusses the real time performance calculations for the turbine cycle and reactor and steam generators of a nuclear power plant. Program accepts plant measurements and calculates performance and efficiency of each part of the cycle: reactor and steam generators, turbines, feedwater heaters, condenser, circulating water system, feed pump turbines, cooling towers. Presently, the calculations involve: 500 inputs, 2400 separate calculations, 500 steam properties subroutine calls, 200 support function accesses, 1500 output valves. The program operates in a real time system at regular intervals
Problems in calculating reactor model (primary circuit) for nuclear power plant diagnostics
Markov, P.
1986-01-01
Some results are presented of the calculation of eigen-vibrations of the system of WWER-440 nuclear reactor vessels in a vacuum and in a liquid. Computer code BOSOR 4 has been written for calculating forced vibrations of shells with axial symmetry and of a simplified system of reactor vessels. A description is given of this code, which is based on the so-called energy method of finite differences. Briefly discussed is the feasibility of applying the results of the latest computation techniques in the diagnostics of the major components of a nuclear reactor. (Z.M.)
Benefits of Parallel I/O in Ab Initio Nuclear Physics Calculations
Laghave, Nikhil; Sosonkina, Masha; Maris, Pieter; Vary, James P.
2009-01-01
Many modern scientific applications rely on highly parallel calculations, which scale to 10's of thousands processors. However, most applications do not concentrate on parallelizing input/output operations. In particular, sequential I/O has been identified as a bottleneck for the highly scalable MFDn (Many Fermion Dynamics for nuclear structure) code performing ab initio nuclear structure calculations. In this paper, we develop interfaces and parallel I/O procedures to use a well-known parallel I/O library in MFDn. As a result, we gain efficient input/output of large datasets along with their portability and ease of use in the downstream processing.
Distribution and Parameter's Calculations of Television Cameras Inside a Nuclear Facility
El-kafas, A.A.
2009-01-01
In this work, a distribution of television cameras and parameter's calculation inside and outside a nuclear facility is presented. Each of exterior and interior camera systems will be described and explained. The work shows the overall closed circuit television system. Fixed and moving cameras with various lens format and different angles of view are used. The calculations of width of images sensitive area and Lens focal length for the cameras will be introduced. The work shows the camera locations and distributions inside and outside the nuclear facility. The technical specifications and parameters for cameras selection are tabulated
Suescun-Diaz, Daniel [Surcolombiana Univ., Neiva (Colombia). Groupo de Fisica Teorica; Narvaez-Paredes, Mauricio [Javeriana Univ., Cali (Colombia). Groupo de Matematica y Estadistica Aplicada Pontificia; Lozano-Parada, Jamie H. [Univ. del Valle, Cali (Colombia). Dept. de Ingenieria
2016-03-15
In this paper, the generalisation of the 4th-order Adams-Bashforth-Moulton predictor-corrector method is proposed to numerically solve the point kinetic equations of the nuclear reactivity calculations without using the nuclear power history. Due to the nature of the point kinetic equations, different predictor modifiers are used in order improve the precision of the approximations obtained. The results obtained with the prediction formulas and generalised corrections improve the precision when compared with previous methods and are valid for various forms of nuclear power and different time steps.
Molina, G.
1985-01-01
Utilization of nuclear energy to produce or generate electricity is a growing practice in the world, since it represent an economic and safe option to replace fossil fuels. During operation of nuclear power plants, radioactive materials are produced. A small fraction of these material are released to environment in the form of liquid or gaseous effluents. Estimation of radiation doses causing by effluents release has three purposes. During design phase of a nuclear station it is useful to adapt the wastes treatment systems to acceptable limits. During licensing phase, the regulator organism verifies the design of nuclear station effectuating estimation of doses. Finally, during operation of a nuclear station, before every unload of radioactive effluents, radiation doses should be evaluate in order to fulfill technical specifications, which limit the release of radioactive materials to environment. 1. To perform calculations of individual doses due to liquid radioactive effluents unload in units 1 and 2 of Laguna Verde nuclear power plant (In licensing phase). 2. To perform a parametric study of the effect of unload recirculation over individual dose, since recirculation has two principal effects: thermodynamical effects in nuclear station and radioactivity concentration, the last can affect the fullfilment of dose limits. 3. To perform the calculation of collective doses causes by unloads of liquid effluents within a radius of 80 Kms. of nuclear station caused by unload of liquid radioactive effluents during normal operation of nuclear power plant and does not include doses caused during accident conditions. In Mexico the organism in charge of regulation of peaceful uses of nuclear energy is Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) and for Laguna Verde licensing, the regulations of country who manufactured the reactor was adopted, it is to say United States of America. In Appendix 'C' units used along this work are explained. Unless another
Bosq, J.Ch.
1998-01-01
This thesis concerns the definition and the validation of the ERANOS neutronic calculation system for steel reflected fast reactors. The calculation system uses JEF2.2 evaluated nuclear data, the ECCO cell code and the BISTRO and VARIANT transport codes. After a description of the physical phenomena induced by the existence of the these sub-critical media, an inventory of the past studies related to steel reflectors is reported. A calculational scheme taking into account the important physical phenomena (strong neutronic slowing-down, presence of broad resonances of the structural materials and spatial variation of the spectrum in the reflector) is defined. This method is validated with the TRIPOLI4 reference Monte-Carlo code. The use of this upgraded calculation method for the analysis of the part of the CIRANO experimental program devoted to the study of steel reflected configurations leads to discrepancies between the calculated and measured values. These remaining discrepancies obtained for the reactivity and the fission rate traverses are due to inaccurate nuclear data for the structural materials. The adjustment of these nuclear data in order to reduce these discrepancies id demonstrated. The additional uncertainty associated to the integral parameters of interest for a nuclear reactor (reactivity and power distribution) induced by the replacement of a fertile blanket by a steel reflector is determined for the Superphenix reactor and is proved to be small. (author)
Kwan-Seong Jeong; Dong-Gyu Lee; Chong-Hun Jung; Kune-Woo Lee
2007-01-01
Available in abstract form only. Full text of publication follows: The uncertainties of decommissioning costs increase high due to several conditions. Decommissioning cost estimation depends on the complexity of nuclear installations, its site-specific physical and radiological inventories. Therefore, the decommissioning costs of nuclear research facilities must be estimated in accordance with the detailed sub-tasks and resources by the tasks of decommissioning activities. By selecting the classified activities and resources, costs are calculated by the items and then the total costs of all decommissioning activities are reshuffled to match with its usage and objectives. And the decommissioning cost of nuclear research facilities is calculated by applying a unit cost factor method on which classification of decommissioning works fitted with the features and specifications of decommissioning objects and establishment of composition factors are based. Decommissioning costs of nuclear research facilities are composed of labor cost, equipment and materials cost. Of these three categorical costs, the calculation of labor costs are very important because decommissioning activities mainly depend on labor force. Labor costs in decommissioning activities are calculated on the basis of working time consumed in decommissioning objects and works. The working times are figured out of unit cost factors and work difficulty factors. Finally, labor costs are figured out by using these factors as parameters of calculation. The accuracy of decommissioning cost estimation results is much higher compared to the real decommissioning works. (authors)
MOSRA-SRAC. Lattice calculation module of the modular code system for nuclear reactor analyses MOSRA
Okumura, Keisuke
2015-10-01
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.)
Status and developing of nuclear emergency response techniques in China
Jiangang, Zhang; Bing, Zhao; Rongyao, Tang; Xiaoxiao, Xu
2008-01-01
Full text: Nuclear Emergency preparedness and response in China is consistent with international basic principle of nuclear safety and emergency response. Nuclear emergency response techniques in China developed with nuclear power from 1980s. The status of nuclear emergency techniques in China are: 1) China have plentiful experiences and abilities in the fields of nuclear facility emergency planning and preparedness, nuclear accident consequence assessment, emergency monitoring, and emergency advisory; 2) Emergency assistance ability in China has a foundation, however it cannot satisfy national requirement; 3) Emergency planning and preparedness is not based on hazard assessment; 4) Remote monitoring and robot techniques in not adaptable to the requirements of nuclear emergency response; 5) A consistent emergency assessment system is lack in China. In this paper, it is analyzed what is the developing focal points of nuclear emergency response techniques in China, and it is proposed that the main points are: a) To develop the research of emergency preparedness on the base of hazard analysis; b) To improve remote monitoring and robot ability during nuclear emergency; c) To develop the response technique research with anti-terrorism. (author)
A new calculation formula of the nuclear cross-section of therapeutic protons
Waldemar Ulmer
2014-03-01
Full Text Available Purpose: We have previously developed for nuclear cross-sections of therapeutic protons a calculation model, which is founded on the collective model as well as a quantum mechanical many particle problem to derive the S matrix and transition probabilities. In this communication, we show that the resonances can be derived by shifted Gaussian functions, whereas the unspecific nuclear interaction compounds can be represented by an error function, which also provides the asymptotic behavior. Method: The energy shifts can be interpreted in terms of necessary domains of energy to excite typical nuclear processes. Thus the necessary formulas referring to previous calculations of nuclear cross-sections will be represented. The mass number AN determines the strong interaction range, i.e. RStrong = 1.2·10-13·AN1/3cm. The threshold energy ETh of the energy barrier is determined by the condition Estrong = ECoulomb. Results and Conclusion: A linear combination of Gaussians, which contain additional energy shifts, and an error function incorporate a possible representation of Fermi-Dirac statistics, which is applied here to nuclear excitations and reaction with release of secondary particles. The new calculation formula provides a better understanding of different types of resonances occurring in nuclear interactions with protons. The present study is mainly a continuation of published papers.1-3--------------------------------Cite this article as: Ulmer W. A new calculation formula of the nuclear cross-section of therapeutic protons. Int J Cancer Ther Oncol 2014; 2(2:020211. DOI: 10.14319/ijcto.0202.11
Self-consistent green function calculations for isospin asymmetric nuclear matter
Mansour, Hesham; Gad, Khalaf; Hassaneen, Khaled S.A.
2010-01-01
The one-body potentials for protons and neutrons are obtained from the self-consistent Green-function calculations of asymmetric nuclear matter, in particular their dependence on the degree of proton/neutron asymmetry. Results of the binding energy per nucleon as a function of the density and asymmetry parameter are presented for the self-consistent Green function approach using the CD-Bonn potential. For the sake of comparison, the same calculations are performed using the Brueckner-Hartree-Fock approximation. The contribution of the hole-hole terms leads to a repulsive contribution to the energy per nucleon which increases with the nuclear density. The incompressibility for asymmetric nuclear matter has been also investigated in the framework of the self-consistent Green-function approach using the CD-Bonn potential. The behavior of the incompressibility is studied for different values of the nuclear density and the neutron excess parameter. The nuclear symmetry potential at fixed nuclear density is also calculated and its value decreases with increasing the nucleon energy. In particular, the nuclear symmetry potential at saturation density changes from positive to negative values at nucleon kinetic energy of about 200 MeV. For the sake of comparison, the same calculations are performed using the Brueckner-Hartree-Fock approximation. The proton/neutron effective mass splitting in neutron-rich matter has been studied. The predicted isospin splitting of the proton/neutron effective mass splitting in neutron-rich matter is such that m n * ≥ m p * . (author)
FLATT - a computer programme for calculating flow and temperature transients in nuclear fuels
Venkat Raj, V.; Koranne, S.M.
1976-01-01
FLATT is a computer code written in Fortran language for BESM-6 computer. The code calculates the flow transients in the coolant circuit of a nuclear reactor, caused by pump failure, and the consequent temperature transients in the fuel, clad, and the coolant. In addition any desired flow transient can be fed into the programme and the resulting temperature transients can be calculated. A case study is also presented. (author)
"Cloud" functions and templates of engineering calculations for nuclear power plants
Ochkov, V. F.; Orlov, K. A.; Ko, Chzho Ko
2014-10-01
The article deals with an important problem of setting up computer-aided design calculations of various circuit configurations and power equipment carried out using the templates and standard computer programs available in the Internet. Information about the developed Internet-based technology for carrying out such calculations using the templates accessible in the Mathcad Prime software package is given. The technology is considered taking as an example the solution of two problems relating to the field of nuclear power engineering.
Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges
B. McLeod
2002-01-01
This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M and O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M and O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report
The development of nuclear power and emergency response
Pan Ziqiang
2007-01-01
Nuclear power is a safe, clean energy, which has been evidenced by the history of nuclear power development. Nuclear power is associated with very low risk but not equal to zero. Accident emergency response and preparedness is a final barrier necessary to reduce potential risks that may arise from nuclear power plants, which must be enhanced. In the course of accident emergency response and preparedness, it is highly necessary to draw domestic and foreign experiences and lessons. Lastly, the paper presents the discussions of some issues which merit attention with respect to emergency response and preparedness in China. (authors)
Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations
Ware Tim
2017-01-01
Full Text Available The ANSWERS Software Service is developing a number of techniques to better understand and quantify uncertainty on calculations of the neutron multiplication factor, k-effective, in nuclear fuel and other systems containing fissile material. The uncertainty on the calculated k-effective arises from a number of sources, including nuclear data uncertainties, manufacturing tolerances, modelling approximations and, for Monte Carlo simulation, stochastic uncertainty. For determining the uncertainties due to nuclear data, a set of application libraries have been generated for use with the MONK10 Monte Carlo and the WIMS10 deterministic criticality and reactor physics codes. This paper overviews the generation of these nuclear data libraries by Latin hypercube sampling of JEFF-3.1.2 evaluated data based upon a library of covariance data taken from JEFF, ENDF/B, JENDL and TENDL evaluations. Criticality calculations have been performed with MONK10 and WIMS10 using these sampled libraries for a number of benchmark models of fissile systems. Results are presented which show the uncertainty on k-effective for these systems arising from the uncertainty on the input nuclear data.
Experience at Los Alamos with use of the optical model for applied nuclear data calculations
Young, P.G.
1994-01-01
While many nuclear models are important in calculations of nuclear data, the optical model usually provides the basic underpinning of analyses directed at data for applications. An overview is given here of experience in the Nuclear Theory and Applications Group at Los Alamos National Laboratory in the use of the optical model for calculations of nuclear cross section data for applied purposes. We consider the direct utilization of total, elastic, and reaction cross sections for neutrons, protons, deuterons, tritons, 3 He and alpha particles in files of evaluated nuclear data covering the energy range of 0 to 200 MeV, as well as transmission coefficients for reaction theory calculations and neutron and proton wave functions direct-reaction and Feshbach-Kerman-Koonin analyses. Optical model codes such as SCAT and ECIS and the reaction theory codes COMNUC, GNASH FKK-GNASH, and DWUCK have primarily been used in our analyses. A summary of optical model parameterizations from past analyses at Los Alamos will be given, including detailed tabulations of the parameters for a selection of nuclei
Experience at Los Alamos with use of the optical model for applied nuclear data calculations
Young, P.G.
1998-01-01
While many nuclear models are important in calculations of nuclear data, the optical model usually provides the basic underpinning of analyses directed at data for applications. An overview is given here of experience in the Nuclear Theory and Applications Group at Los Alamos National Laboratory in the use of the optical model for calculations of nuclear cross section data for applied purposes. We consider the direct utilization of total, elastic, and reaction cross sections for neutrons, protons, deuterons, tritons, 3 He and alpha particles in files of evaluated nuclear data covering the energy range of 0 to 200 MeV, as well as transmission coefficients for reaction theory calculations and neutron and proton wave functions in direct-reaction and Feshbach-Kerman-Koonin analyses. Optical model codes such as SCAT and ECIS and the reaction theory codes COMNUC, GNASH, FKK-GNASH, and DWUCK have primarily been used in our analyses. A summary of optical model parameterizations from past analyses at Los Alamos will be given, including detailed tabulations of the parameters for a selection of nuclei. (author)
Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations
Ware, Tim; Dobson, Geoff; Hanlon, David; Hiles, Richard; Mason, Robert; Perry, Ray
2017-09-01
The ANSWERS Software Service is developing a number of techniques to better understand and quantify uncertainty on calculations of the neutron multiplication factor, k-effective, in nuclear fuel and other systems containing fissile material. The uncertainty on the calculated k-effective arises from a number of sources, including nuclear data uncertainties, manufacturing tolerances, modelling approximations and, for Monte Carlo simulation, stochastic uncertainty. For determining the uncertainties due to nuclear data, a set of application libraries have been generated for use with the MONK10 Monte Carlo and the WIMS10 deterministic criticality and reactor physics codes. This paper overviews the generation of these nuclear data libraries by Latin hypercube sampling of JEFF-3.1.2 evaluated data based upon a library of covariance data taken from JEFF, ENDF/B, JENDL and TENDL evaluations. Criticality calculations have been performed with MONK10 and WIMS10 using these sampled libraries for a number of benchmark models of fissile systems. Results are presented which show the uncertainty on k-effective for these systems arising from the uncertainty on the input nuclear data.
Kellö, Vladimir [Department of Physical Chemistry, Comenius University, SK-842 15 Bratislava (Slovakia)
2015-01-22
Highly correlated scalar relativistic calculations of electric field gradients at nuclei in diatomic molecules in combination with accurate nuclear quadrupole coupling constants obtained from microwave spectroscopy are used for determination of nuclear quadrupole moments.
Computer Security Incident Response Planning at Nuclear Facilities
2016-06-01
The purpose of this publication is to assist Member States in developing comprehensive contingency plans for computer security incidents with the potential to impact nuclear security and/or nuclear safety. It provides an outline and recommendations for establishing a computer security incident response capability as part of a computer security programme, and considers the roles and responsibilities of the system owner, operator, competent authority, and national technical authority in responding to a computer security incident with possible nuclear security repercussions
Walsh, M.L.; Agnew, D.A.; Donnelly, K.E.
1984-01-01
The response of the Ontario Hydro Thermoluminescence Dosimetry System to beta radiation in nuclear power station environments was evaluated. Synthetic beta spectra were constructed, based on activity samples from heat transport systems and fuelling machine contamination smears at nuclear power stations. Using these spectra and dosemeter energy response functions, an overall response factor for the skin dosemeter relative to skin dose at 7 mg.cm -2 was calculated. This calculation was done assuming three specific geometries: (1) an infinite uniformly contaminated plane source at a distance of 33 cm (50 mg.cm -2 total shielding) from the receptor; (2) an infinite cloud surrounding the receptor; (3) a point source at 33 cm. Based on these calculations, a conservative response factor of 0.7 has been chosen. This provides an equation for skin dose assignment, i.e. Skin Dose = 1.4 x Skin Dosemeter Reading when the skin dosemeter is directly calibrated in mGy(gamma). (author)
Response of nuclear emulsions to ionizing radiations
Katz, R.; Pinkerton, F.E.
1975-01-01
Heavy ion tracks in Ilford K-2 emulsion are simulated with a computer program which makes use of the delta-ray theory of track structure, and the special assumption that the response of this emulsion to gamma-rays is 8-or-more hit. The Ilford K-series of nuclear emulsions is produced from a parent stock called K.0 emulsion, sensitized to become K.1 to K.5, and desensitized to become K-1 to K-3. Our simulations demonstrate that the emulsions K.5 through K.0 to K-1 are 1-or-more hit detectors, while K-2 is an 8-or-more hit detector. We have no data for K-3 emulsion. It would appear that emulsions of intermediate hittedness might be produced by an intermediate desensitization, to mimic or match the RBE-LET variations of biological cells, perhaps to produce a ''rem-dosimeter''. In the K-2 emulsion no developable gains are produced by stopping H, He, and Li ions. The emulsion has ''threshold-like'' properties, resembling etchable track detectors. It should prove useful in the measurement of high LET dose in a strong low LET background, as for pions or neutrons. Since it can be expected to accumulate and repair ''sub-lethal damage'', to display the ion-kill and gamma-kill inactivation modes, the grain-count and track width regimes, it may serve to model biological effects. (auth)
Calculational tools for the evaluation of nuclear cross-section and spectra data
Gardner, M.A.
1985-01-01
A technique based on discrete energy levels rather than energy level densities is presented for nuclear reaction calculations. The validity of the technique is demonstrated via theoretical and experimental agreement for cross sections, isomer-ratios and gamma-ray strength functions. 50 refs., 7 figs
NUCORE - A system for nuclear structure calculations with cluster-core models
Heras, C.A.; Abecasis, S.M.
1982-01-01
Calculation of nuclear energy levels and their electromagnetic properties, modelling the nucleus as a cluster of a few particles and/or holes interacting with a core which in turn is modelled as a quadrupole vibrator (cluster-phonon model). The members of the cluster interact via quadrupole-quadrupole and pairing forces. (orig.)
Kitahara, Yoshihisa; Kishimoto, Yoichiro; Narita, Osamu; Shinohara, Kunihiko
1979-01-01
Several Calculation methods for relative concentration (X/Q) and relative cloud-gamma dose (D/Q) of the radioactive materials released from nuclear facilities by posturated accident are presented. The procedure has been formulated as a Computer program PANDA and the usage is explained. (author)
Single-particle basis and translational invariance in microscopic nuclear calculations
Ehfros, V.D.
1977-01-01
The approach to the few-body problem is considered which allows to use the simple single-particle basis without violation of the translation invariance. A method is proposed to solve the nuclear reaction problems in the single-particle basis. The method satisfies the Pauli principle and the translation invariance. Calculation of the matrix elements of operators is treated
Methodology and conclusions of activation calculations of WWER-440 type nuclear power plants
Babcsány, Boglárka, E-mail: boglarka.babcsany@reak.bme.hu; Czifrus, Szabolcs; Fehér, Sándor
2015-04-01
Highlights: • Activation calculation of two WWER-440 type nuclear power plants. • Detailed description of the applied activation calculation methodology. • Graphical results for total activity and waste index categorization. • General conclusions for activation applicable in the case of PWR reactors. - Abstract: Activation calculations for two nuclear power plants of WWER-440 type have been performed by the authors in order to assist the decommissioning planning by assessing the radioactive inventory present at the time of and at different times after the final shutdown. According to related international literature and studies performed earlier by the authors, considering the activity more than 99% of this inventory is concentrated in the materials directly surrounding the reactor core, where the predominant evolution of radionuclides is generated by neutron induced nuclear reactions. In order to obtain the highest possible accuracy in modelling, three-dimensional Monte Carlo neutron transport calculations were performed. Besides the methods and models applied to these analyses, the paper also summarizes the results that can be generally applied to such nuclear power plant types. At the time of shutdown, the total activity of the stainless steel components is about 6 × 10{sup 16} Bq and 1.3 × 10{sup 17} Bq for the two NPPs considered. The biological shielding concrete constitutes approximately 7 × 10{sup 13} Bq and 1.1 × 10{sup 14} Bq.
Maucec, M
2005-01-01
Monte Carlo simulations for nuclear logging applications are considered to be highly demanding transport problems. In this paper, the implementation of weight-window variance reduction schemes in a 'manual' fashion to improve the efficiency of calculations for a neutron logging tool is presented.
SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS
D. A. Thomas
1996-01-01
The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report
Calculation of the neutron activation parameters from recently evaluated nuclear data
Lopez Aldama, Daniel; Diaz Martinez, Nereida C.
1999-01-01
Neutron Activation Analysis (NAA) requires the values for nuclear data such as the 2200 m/s cross section so, the resonance integral I0, the parameter Q0 and the well-known Westcott factors. The availability of recently evaluated nuclear data libraries as the ENDF/B-VI Rev. 5, JEF 2.2, CENDL-2.1 and JENDL-3.2, makes possible to derive the above quantities from the basic nuclear data. It could be very helpful for those NAA parameters, which are unknown or difficult to measure accurately. The procedure to compute the NAA parameters includes the processing of the evaluated nuclear data and the calculation of each parameter directly from its definition. The evaluated nuclear data libraries ENDF/B-VI Rev. 5 and JENDL 3.2 were selected as the main sources of basic nuclear data. The ENDF pre-processing codes were used for processing the source evaluated data and a modified version of the INTER code was applied to calculate the required NAA integrals. The NAA parameters were computed for more than 30 important isotopes. The obtained results were compared with experimental values whenever possible
Calculation Of Recycle And Open Cycle Nuclear Fuel Cost Using Lagistase Method
Djoko Birmano, Moch
2002-01-01
. To be presented the calculation of recycle and open cycle nuclear fuel cost for LWR type that have net power of 600 MWe. This calculation using LEGECOST method developed by IAEA which have characteristics,where i.e. money is stated in constant money (no inflation),discount rate is equalized with interest rate and not consider tax and depreciation.As a conclusion is that open cycle nuclear fuel cost more advantage because it is cheaper than recycle nuclear fuel cost. This is caused that at present, reprocessing process disadvantage because it has not found yet more efficient and cheaper method, besides price of fresh uranium is still cheap. In future, the cost of recycle nuclear fuel cycle will be more competitive toward the cost of open nuclear fuel cycle if is found technology of reprocessing process that more advance, efficient and cheap. Increase of Pu use for reactor fuel especially MOX type will rise Pu price that finally will decrease the cost of recycle nuclear fuel cycle
Impact of nuclear data on sodium-cooled fast reactor calculations
Aures, A.; Bostelmann, F.; Zwermann, W.; Velkov, K.
2016-01-01
Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors. (authors)
Calculation of LUEC using HEEP Software for Nuclear Hydrogen Production Plant
Kim, Jongho; Lee, Kiyoung; Kim, Minhwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2015-05-15
To achieve the hydrogen economy, it is very important to produce a massive amount of hydrogen in a clean, safe and efficient way. Nuclear production of hydrogen would allow massive production of hydrogen at economic prices while avoiding environments pollution by reducing the release of carbon dioxide. A Very High Temperature Reactor (VHTR) is considered as an efficient reactor to couple with the thermo-chemical Sulfur Iodine (SI) cycle to achieve the hydrogen economy. HEEP(Hydrogen Economy Evaluation Program) is one of the software tools developed by IAEA to evaluate the economy of the nuclear hydrogen production system by estimating unit hydrogen production cost. In this paper, the LUHC (Levelized Unit Hydrogen Cost) is calculated by using HEEP for nuclear hydrogen production plant, which consists of 4 modules of 600 MWth VHTR coupled with SI process. The levelized unit hydrogen production cost(LUHC) was calculated by the HEEP software.
The risk of major nuclear accident: calculation and perception of probabilities
Leveque, Francois
2013-01-01
Whereas before the Fukushima accident, already eight major accidents occurred in nuclear power plants, a number which is higher than that expected by experts and rather close to that corresponding of people perception of risk, the author discusses how to understand these differences and reconcile observations, objective probability of accidents and subjective assessment of risks, why experts have been over-optimistic, whether public opinion is irrational regarding nuclear risk, and how to measure risk and its perception. Thus, he addresses and discusses the following issues: risk calculation (cost, calculated frequency of major accident, bias between the number of observed accidents and model predictions), perceived probabilities and aversion for disasters (perception biases of probability, perception biases unfavourable to nuclear), the Bayes contribution and its application (Bayes-Laplace law, statistics, choice of an a priori probability, prediction of the next event, probability of a core fusion tomorrow)
Improving the calculated core stability by the core nuclear design optimization
Partanen, P.
1995-01-01
Three different equilibrium core loadings for TVO II reactor have been generated in order to improve the core stability properties at uprated power level. The reactor thermal power is assumed to be uprated from 2160 MW th to 2500 MW th , which moves the operating point after a rapid pump rundown where the core stability has been calculated from 1340 MW th and 3200 kg/s to 1675 MW th and 4000 kg/s. The core has been refuelled with ABB Atom Svea-100 -fuel, which has 3,64% w/o U-235 average enrichment in the highly enriched zone. PHOENIX lattice code has been used to provide the homogenized nuclear constants. POLCA4 static core simulator has been used for core loadings and cycle simulations and RAMONA-3B program for simulating the dynamic response to the disturbance for which the stability behaviour has been evaluated. The core decay ratio has been successfully reduced from 0,83 to 0,55 mainly by reducing the power peaking factors. (orig.) (7 figs., 1 tab.)
Measured and calculated absorptance of tracks of fast heavy ions in Ilford G5 nuclear emulsion
Jensen, M.; Mathiesen, O.
1975-04-01
A modified form of the delta-ray theory of track formation developed by Katz and coworkers has been used to calculate the absorptance of tracks of fast heavy ions in Ilford G5 nuclear emulsion. The theoretical data have been compared with results of different photometrical investigations reported in the literature. In most cases the theoretical predictions are found to be in good agreement with experiments. This suggests that the theory can be used in the planning and execution of future experiments, i.e. to optimize the geometry of the photometer and to obtain an absolute charge calibration of the detector. It is shown that the basic photometrical properties of an emulsion stack can be described by a single quantity which can be determined from measurements. Knowing this quantity it is possible to predict the response of the emulsion stack for different types of photometers. The practical limits of the use of the modified theory at high and low levels of energy dose are discussed. (Auth.)
Measured and calculated absorptance of tracks of fast heavy ions in Ilford G5 nuclear emulsion
Jensen, M.; Mathiesen, O.
1976-01-01
A modified form of the delta-ray theory of track formation developed by Katz and coworkers has been used to calculate the absorptance of tracks of fast heavy ions in Ilford G5 nuclear emulsion. The theoretical data have been compared with results of different photometrical investigations reported in the literature. In most cases the theoretical predictions are found to be in good agreement with experiments. This suggests that the theory can be used in the planning and execution of future experiments, i.e. to optimize the geometry of the photometer and to obtain an absolute charge calibration of the detector. It is shown that the basic photometrical properties of an emulsion stack can be described by a single quantity which can be determined from measurements. Knowing this quantity it is possible to predict the response of the detector system, for different types of photometers. The practical limits of the use of the modified theory at high and low levels of energy dose are discussed. (Auth.)
Evaluation of WIMS-D/4 nuclear data library used on TRIGA reactor calculation
Chen Wei; Xie Zhongsheng; Jiang Xinbiao; Chen Da
1997-01-01
The 69 groups constants of H in ZrH, 166 Er and 167 Er generated by NJOY and GASKET codes are inserted into WIMS nuclear data library WIMS-CNDC and WIMS-NINT libraries used on RTIGA reactor calculation are obtained. In order to check WIMS-CNDC and WIMS-NINT libraries, the scattering cross-section is compared with that in WIMS-IJS library. The group constant, K ∞ and temperature coefficient are calculated by using WIMS-CNDC, WIMS-NINT and WIMS-IJS. The results show the both libraries are suitable for calculation of TRIGA reactor
Nuclear model calculations below 200 MeV and evaluation prospects
Koning, A.J.; Bersillon, O.; Delaroche, J.P.
1994-08-01
A computational method is outlined for the quantum-mechanical prediction of the whole double-differential energy spectrum. Cross sections as calculated with the code system MINGUS are presented for (n,xn) and (p,xn) reactions on 208 Pb and 209 Bi. Our approach involves a dispersive optical model, comprehensive discrete state calculations, renormalized particle-hole state densities, a combined MSD/MSC model for pre-equilibrium reactions and compound nucleus calculations. The relation with the evaluation of nuclear data files is discussed. (orig.)
RPA spin-isospin nuclear response in the deep inelastic region
Alberico, W.M.; Molinari, A.; De Pace, A.; Johnson, M.B.; Ericson, M.
1985-11-01
The spin-isospin volume responses of a finite nucleus are evaluated in the RPA frame, utilizing a harmonic oscillator basis. Particular emphasis is given to the mixing between the longitudinal and transverse couplings, which arise at the nuclear surface. We show that it reduces somewhat the contrast between the two spin responses. We compare the calculated transverse response with the experimental one extracted from deep inelastic electron scattering
The bad debt of nuclear responsibility
Haverkamp, J.
2010-01-01
Nuclear regulation plays an important role in keeping the visible and invisible threats of radioactivity at bay. This article argues that the largest gains in nuclear safety can be made if nuclear activities with economically, socially and environmentally viable alternatives (such as nuclear energy and nuclear weapons) are phased out - similarly to what has been done for toxic substances. Unless this takes place, the legacy of radioactive waste for future generations will only increase. In this respect, the article highlights the importance of accountability, independence and transparency of nuclear regulators to contain the risks stemming from radioactivity. In the interest of safety, other uses of nuclear technology should, as a rule, apply the best available technology and the best regulatory practice. As is made clear in the article, the EURATOM Nuclear Safety Directive, adopted in 2009, hardly addresses any of these issues. The article ends on a set of demands from Greenpeace to meet the challenges posed by nuclear safety, concluding that there is still a long way to go before the sector can be considered even reasonably safe. (author)
Introducing PCTRAN as an evaluation tool for nuclear power plant emergency responses
Cheng, Yi-Hsiang; Shih, Chunkuan; Chiang, Show-Chyuan; Weng, Tung-Li
2012-01-01
Highlights: ► PCTRAN is integrated with an atmospheric dispersion algorithm. ► The improved PCTRAN acts as an accident/incident simulator and a data exchange system. ► The software helps the responsible organizations decide the rescue and protective actions. ► The evaluation results show the nuclear power plant accident and its off-site dose consequences. ► The software can be used for nuclear power plant emergency responses. - Abstract: Protecting the public from radiation exposure is important if a nuclear power plant (NPP) accident occurs. Deciding appropriate protective actions in a timely and effective manner can be fulfilled by using an effective accident evaluation tool. In our earlier work, we have integrated PCTRAN (Personal Computer Transient Analyzer) with the off-site dose calculation model. In this study, we introduce PCTRAN as an evaluation tool for nuclear power plant emergency responses. If abnormal conditions in the plant are monitored or observed, the plant staffs can distinguish accident/incident initiation events. Thus, the responsible personnel can immediately operate PCTRAN and set up those accident/incident initiation events to simulate the nuclear power plant transient or accident in conjunction with off-site dose distributions. The evaluation results consequently help the responsible organizations decide the rescue and protective actions. In this study, we explain and demonstrate the capabilities of PCTRAN for nuclear emergency responses, through applying it to simulate the postulated nuclear power plant accident scenarios.
Koponen, B.L.; Hampel, V.E.
1982-01-01
This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41
Lima, Fernando R.A.; Lira, Carlos A.B.O.; Gandini, Augusto
1995-01-01
During the last two decades perturbative methods became an efficient tool to perform sensitivity analysis in nuclear reactor safety problems. In this paper, a comparative study taking into account perturbation formalisms (Diferential and Matricial Mthods and generalized Perturbation Theory - GPT) is considered. Then a few number of applications are described to analyze the sensitivity of some functions relavant to thermal hydraulics designs or safety analysis of nuclear reactor cores and steam generators. The behaviours of the nuclear reactor cores and steam generators are simulated, respectively, by the COBRA-IV-I and GEVAP codes. Results of sensitivity calculations have shown a good agreement when compared to those obtained directly by using the mentioned codes. So, a significative computational time safe can be obtained with perturbative methods performing sensitivity analysis in nuclear power plants. (author). 25 refs., 5 tabs
Impact of the measurement data on the CORD-2 nuclear design calculations of the NPP Krsko
Kromar, M.; Kurincic, B.
2004-01-01
The CORD-2 package was developed at Jozef Stefan Institute and has been validated for the nuclear design calculations of PWR cores. It has been used for the independent verification of the NPP Krsko nuclear design for the last 6 cycles of operation. The accuracy of the package is very good fulfilling all criteria usually imposed on the design prediction of the reactor nuclear parameters. To obtain as robust package as possible and to eliminate potential systematic errors of the package, it was decided to rely on measured core power distributions. In core power measurements, which are performed each month of reactor operation, are used to obtain fuel assemblies burnup histories. Consequently, burnup distributions obtained from the power measurements of all previous cycles are taken as a starting point at the beginning of the considered cycle. Since a lot of experience has been gained with the package, it was decided to evaluate the impact of measurement data on the accuracy of the calculations. Burnup calculations of all 19 cycles of the NPP Krsko are repeated, building simultaneously the calculated library of burnup histories for all fuel assemblies. The basic reactor parameters such as HZP critical boron concentration, isothermal temperature coefficient, control rod worth and cycle length are compared to the results obtained with CORD-2 standard sequence of calculation and direct measurements.(author)
Experimental study and nuclear model calculations of {sup 3}He-induced nuclear reactions on zinc
Al-Abyad, M.; Mohamed, Gehan Y. [Nuclear Research Centre, Atomic Energy Authority, Physics Department (Cyclotron Facility), Cairo (Egypt); Ditroi, F.; Takacs, S.; Tarkanyi, F. [Hungarian Academy of Sciences (ATOMKI), Institute for Nuclear Research, Debrecen (Hungary)
2017-05-15
Excitation functions of {sup 3}He-induced nuclear reactions on natural zinc were measured using the standard stacked-foil technique and high-resolution gamma-ray spectrometry. From their threshold energies up to 27 MeV, the cross-sections for {sup nat}Zn ({sup 3}He,xn) {sup 69}Ge, {sup nat}Zn({sup 3}He,xnp) {sup 66,67,68}Ga, and {sup nat}Zn({sup 3}He,x){sup 62,65}Zn reactions were measured. The nuclear model codes TALYS-1.6, EMPIRE-3.2 and ALICE-IPPE were used to describe the formation of these products. The present data were compared with the theoretical results and with the available experimental data. Integral yields for some important radioisotopes were determined. (orig.)
Pham Ngoc Khoi; Nguyen Kim Dung
2016-03-01
Recognizing the significant value and necessity of publishing the scientific document of experimental and calculational works on the Dalat Nuclear Research Reactor (DNRR) physics and engineering for research, operation, training activities as well as for international scientific exchange, Vietnam Atomic Energy Agency (VAEA) and Vietnam Atomic Energy Institute have completed editing to publish the “Experimental and Calculational Works on Characteristics of THE DALAT NUCLEAR RESEARCH REACTOR” which consists of 26 typical papers representing the most important experimental and calculational results of the DNRR physics and engineering obtained during 30 years of operation and exploitation with the contribution of Vietnamese and former USSR’s experts, especially scientists and engineers working at the Reactor Center of the NRI
Summation of Parquet diagrams as an ab initio method in nuclear structure calculations
Bergli, Elise; Hjorth-Jensen, Morten
2011-01-01
Research highlights: → We present a Green's function based approach for doing ab initio nuclear structure calculations. → In particular the sum the subset of so-called Parquet diagrams. → Applying the theory to a simple but realistic model, results in good agreement with other ab initio methods. → This opens up for ab initio calculations for medium-heavy nuclei. - Abstract: In this work we discuss the summation of the Parquet class of diagrams within Green's function theory as a possible framework for ab initio nuclear structure calculations. The theory is presented and some numerical details are discussed, in particular the approximations employed. We apply the Parquet method to a simple model, and compare our results with those from an exact solution. The main conclusion is that even at the level of approximation presented here, the results shows good agreement with other comparable ab initio approaches.
Calculational model for condensation of water vapor during an underground nuclear detonation
Knox, R.J.
1975-01-01
An empirally derived mathematical model was developed to calculate the pressure and temperature history during condensation of water vapor in an underground-nuclear-explosion cavity. The condensation process is non-isothermal. Use has been made of the Clapeyron-Clausius equation as a basis for development of the model. Analytic fits to the vapor pressure and the latent heat of vaporization for saturated-water vapor, together with an estimated value for the heat-transfer coefficient, have been used to describe the phenomena. The calculated pressure-history during condensation has been determined to be exponential, with a time constant somewhat less than that observed during the cooling of the superheated steam from the explosion. The behavior of the calculated condensation-pressure compares well with the observed-pressure record (until just prior to cavity collapse) for a particular nuclear-detonation event for which data is available
Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel
L. Angers
2001-01-01
The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k eff ) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR
Calculation of fuel burn-up and fuel reloading for the Dalat Nuclear Research Reactor
Lan, Nguyen Phuoc; Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Binh, Do Quang [Nuclear Research Inst., Da Lat (Viet Nam)
1994-10-01
Calculation of fuel burnup and fuel reloading for the Dalat Nuclear Research Reactor was carried out by using a new programme named HEXA-BURNUP, realized in a PC. The programme is used to calculate the following parameters of the Dalat reactor: a/Critical configurations of the core loaded with 69, 72, 74, 86, 88, 89 and 92 fuel elements. The effective multiplication coefficients equal 1 within the error ranges of less than 0.38%. b/ The thermal neutron flux distribution in the reactor. The calculated results agree with the experimental data measured at 11 typical positions. c/The average fuel burn-up for the period from Feb. 1984 to Sep. 1992. The difference between calculation and experiment is only about 1.9%. 10 fuel reloading versions are calculated, from which an optimal version is proposed. (author). 9 refs., 4 figs., 5 tabs.
Suescun D, D. [Universidad Surcolombiana, Av. Pastrana Borrero - Carrera 1, Neiva, Huila (Colombia); Rojas A, O., E-mail: daniel.suescun@usco.edu.co [Universidad Popular Autonoma del Estado de Puebla, Av. 9 Pte 1908, Barrio de Santiago, 72410 Puebla (Mexico)
2017-09-15
The measurement of reactivity is a function of time and its calculation results from the variation in nuclear power from the inverse equation of punctual kinetics. This equation is a differential integral, where the term of the integral conserves the historical power and the differential part is directly related to the period of the reactor. In practice, in a nuclear plant, sensors are required to record the signals. For example, the movements of the control rods that cause the fluctuations of nuclear power over time commonly generate signals with noise, an event that makes difficult to estimate the reactivity. Thus is necessary and very useful to build digital reactivity meters in real time, since allows a reactor to be operated with greater security. The calculation of the reactivity is carried out using punctual kinetics, especially the concentration of delayed neutron precursors. In this work we present a new way to reduce the fluctuations in the calculation of the reactivity, for the high precision we propose the generalization of the predictor and corrector of the Adams-Bashforth-Moulton (ABM) method of order 4 to solve numerically the equations of the point kinetics for the calculation of the reactivity, without using the power history, due to the nature of the equations of the punctual kinetics, the modifiers of the different predictors are used to increase the accuracy in the approximation obtained accompanied by the filter known as Savitzky-Golay (Sg), allow to reduce the fluctuations of reactivity. It is known that the Sg filter softens and does not attenuate the nuclear power regardless of its shape, guarantees to reduce noise levels up to σ = 0.01, with a calculation time step of σ = 0.01, s. This formulation uses a polynomial approximation of Gram, with a degree d = 2, to calculate the convolution coefficients by means of an analytical formula that is implemented computationally and avoids problems of bad conditioning, caused by the inversion of a
Majer, P.
1990-01-01
The fundamentals are outlined of the discounted value flows method, which is used in industrial countries for calculating the specific electricity production costs. Actual calculations were performed for the first two units of the Temelin nuclear power plant. All costs associated with the construction, operation and decommissioning of this nuclear power plant were taken into account. With a high degree of certainty, the specific production costs of the Temelin nuclear power plant will lie within the range of 0.32 to 0.36 CSK/kWh. Nearly all results of the sensitivity analysis performed for the possible changes in the input values fall within this range. An increase in the interest rate to above 8% is an exception; this, however, can be regarded as rather improbable on a long-term basis. Sensitivity analysis gave evidence that the results of the electricity production cost calculations for the Temelin nuclear power plant can be considered sufficiently stable. (Z.M.). 7 figs., 2 tabs., 14 refs
Pandey, Anil Kumar; Sharma, Sanjay Kumar; Sharma, Punit; Gupta, Priyanka; Kumar, Rakesh
2013-01-01
It is important to ensure that as low as reasonably achievable (ALARA) concept during the radiopharmaceutical (RPH) dose administration in pediatric patients. Several methods have been suggested over the years for the calculation of individualized RPH dose, sometimes requiring complex calculations and large variability exists for administered dose in children. The aim of the present study was to develop a software application that can calculate and store RPH dose along with patient record. We reviewed the literature to select the dose formula and used Microsoft Access (a software package) to develop this application. We used the Microsoft Excel to verify the accurate execution of the dose formula. The manual and computer time using this program required for calculating the RPH dose were compared. The developed application calculates RPH dose for pediatric patients based on European Association of Nuclear Medicine dose card, weight based, body surface area based, Clark, Solomon Fried, Young and Webster's formula. It is password protected to prevent the accidental damage and stores the complete record of patients that can be exported to Excel sheet for further analysis. It reduces the burden of calculation and saves considerable time i.e., 2 min computer time as compared with 102 min (manual calculation with the calculator for all seven formulas for 25 patients). The software detailed above appears to be an easy and useful method for calculation of pediatric RPH dose in routine clinical practice. This software application will help in helping the user to routinely applied ALARA principle while pediatric dose administration. (author)
Russian Nuclear Power Response to the Fukushima Accident
Asmolov, Vladimir G.
2011-01-01
Conclusions: 1. Nuclear safety is not based only on regulators. The prime responsibility for nuclear safety rests with operating organizations which have the necessary experience and knowledge. 2. Improvement in safety can be reached through better sharing of operation experience and improvements in technology. The IAEA is to increase interactions with utilities and nuclear industry. 3. The IAEA is to declare clearly the recognition of the role of operating organizations and nuclear industry in safe, efficacious and sustainable nuclear power development and to strengthen cooperation with them
Development of a nuclear spallation simulation code and calculations of primary spallation products
Nishida, Takahiko; Nakahara, Yasuaki; Tsutsui, Tsuneo
1986-08-01
In order to make evaluations of computational models for the nuclear spallation reaction from a nuclear physics point of view, a simulation code NUCLEUS has been developed by modifying and combining the Monte Carlo codes NMTC/JAERI and NMTA/JAERI for calculating only the nuclear spallation reaction (intranuclear cascade + evaporation and/or fast fission) between a nucleus and a projectile without taking into consideration of internuclear transport. New several plotting routines have been provided for the rapid process of much more event data, obtained by using the ARGUS plotting system. The results obtained by our code can be directly compared with the experimental results using by thin foil experiments in which internuclear multiple collisions have little effects, and will serve to upgrade the calculational methods and the values of nuclear parameters currently used in the calculations. Some discussions are done about the preliminary computational results obtained by using NUCLEUS. The mass distribution and charge dispersion of reaction products are examined in some detail for the nuclear spallation reaction between incident protons and target nuclei, such as U, Pb and Ag, in the energy range from 0.5 GeV to 3.0 GeV. These results show that the distribution of reaction products ceases to change its form as the proton energy increases over about 2 GeV. The same tendency is seen in the energy dependence of the number of primary particles emitted from a nucleus. After spallation reactions, a variety of nuclei, especially many neutron deficient nuclides with nuclear charges nearly equal to ones of a target nucleus, are produced. Due to their short lifetime most of them will change to stable nuclides in due time. Finally, some important issues are discussed to improve the present simulation method. (author)
A prototype nuclear emergency response decision making expert system
Chang, C.; Shih, C.; Hong, M.; Yu, W.; Su, M.; Wang, S.
1990-01-01
A prototype of emergency response expert system developed for nuclear power plants, has been fulfilled by Institute of Nuclear Energy Research. Key elements that have been implemented for emergency response include radioactive material dispersion assessment, dynamic transportation evacuation assessment, and meteorological parametric forecasting. A network system consists of five 80386 Personal Computers (PCs) has been installed to perform the system functions above. A further project is still continuing to achieve a more complicated and fanciful computer aid integral emergency response expert system
Nuclear Energy Response in the EMF27 Study
Kim, Son H.; Wada, Kenichi; Kurosawa, Atsushi; Roberts, Matthew
2014-01-01
The nuclear energy response for mitigating global climate change across eighteen participating models of the EMF27 study is investigated. Diverse perspectives on the future role of nuclear power in the global energy system are evident in the broad range of nuclear power contributions from participating models of the study. In the Baseline scenario without climate policy, nuclear electricity generation and shares span 0 - 66 EJ/ year and 0 - 25% in 2100 for all models, with a median nuclear electricity generation of 39 EJ/year (1,389 GWe at 90% capacity factor) and median share of 9%. The role of nuclear energy increased under the climate policy scenarios. The median of nuclear energy use across all models doubled in the 450 ppm CO2e scenario with a nuclear electricity generation of 67 EJ/year (2,352 GWe at 90% capacity factor) and share of 17% in 2100. The broad range of nuclear electricity generation (11 - 214 EJ/year) and shares (2 - 38%) in 2100 of the 450 ppm CO2e scenario reflect differences in the technology choice behavior, technology assumptions and competitiveness of low carbon technologies. Greater clarification of nuclear fuel cycle issues and risk factors associated with nuclear energy use are necessary for understanding the nuclear deployment constraints imposed in models and for improving the assessment of the nuclear energy potential in addressing climate change
Preliminary results on food consumption rates for off-site dose calculation of nuclear power plants
Lee, Gab Bock; Chung, Yang Geun; Bang, Sun Young; Kang, Duk Won
2005-01-01
The Internal dose by food consumption mostly account for radiological dose of public around nuclear power plants(NPP). But, food consumption rate applied to off-site dose calculation in Korea which is the result of field investigation around Kori NPP by the KAERI in 1988. is not reflected of the latest dietary characteristics. The Ministry of Health and Welfare Affairs has investigated the food and nutrition of nations every 3 years based on the Law of National Health Improvement. To update the food consumption rates of the maximum individual, the analysis of the national food investigation results and field surveys around nuclear power plant sites have been carried out
Do nuclear engineering educators have a special responsibility
Weinberg, A.M.
1977-01-01
Each 1000 MW(e) reactor in equilibrium contains 15 x 10 9 Ci of radioactivity. To handle this material safety requires an extremely high level of expertise and commitment - in many respects, an expertise that goes beyond what is demanded of any other technology. If one grants that nuclear engineering is more demanding than other engineering because the price of failure is greater, one must ask how can we inculcate into the coming generations of nuclear engineers a full sense of the responsibility they bear in practising their profession. Clearly a first requirement is that all elements of the nuclear community -utility executives, equipment engineers, operating engineers, nuclear engineers, administrators - must recognize and accept the idea that nuclear energy is something special, and that therefore its practitioners must be special. This sense must be instilled into young nuclear engineers during their education. A special responsibility therefore devolves upon nuclear engineering educators: first, to recognize the special character of their profession, and second, to convey this sense to their students. The possibility of institutionalizing this sense of responsibility by establishing a nuclear Hippocratic Oath or special canon of ethics for nuclear engineers ought to be discussed within the nuclear community. (author)
Supplier responsibility for nuclear material quality
Stuart, P.S.; Dohna, A.E.
1976-01-01
Nuclear materials must be delivered by either the manufacturer or the distributor with objective, documented evidence that the material was manufactured, inspected, and tested by proven techniques performed by qualified personnel working to documented procedures. Measurement devices used for acceptance must be of proven accuracy. The material and all records must be identified for positive traceability as part of the quality history of the nuclear components, system, or structure in which the material was used. In conclusion, the nuclear material supplier must join the fabricator, the installer, and the user in effective implementation of the total systems approach to the application of quality assurance principles to all phases of procurement, fabrication, installation, and use of the safety-related components, systems, and structures in a nuclear power plant
Analysis of offsite dose calculation methodology for a nuclear power reactor
Moser, D.M.
1995-01-01
This technical study reviews the methodology for calculating offsite dose estimates as described in the offsite dose calculation manual (ODCM) for Pennsylvania Power and Light - Susquehanna Steam Electric Station (SSES). An evaluation of the SSES ODCM dose assessment methodology indicates that it conforms with methodology accepted by the US Nuclear Regulatory Commission (NRC). Using 1993 SSES effluent data, dose estimates are calculated according to SSES ODCM methodology and compared to the dose estimates calculated according to SSES ODCM and the computer model used to produce the reported 1993 dose estimates. The 1993 SSES dose estimates are based on the axioms of Publication 2 of the International Commission of Radiological Protection (ICRP). SSES Dose estimates based on the axioms of ICRP Publication 26 and 30 reveal the total body estimates to be the most affected
Herman, M.; Capote, R.; Sin, M.
2013-08-01
EMPIRE is a modular system of nuclear reaction codes, comprising various nuclear models, and designed for calculations over a broad range of energies and incident particles. The system can be used for theoretical investigations of nuclear reactions as well as for nuclear data evaluation work. Photons, nucleons, deuterons, tritons, helions ( 3 He), α's, and light or heavy ions can be selected as projectiles. The energy range starts just above the resonance region in the case of a neutron projectile, and extends up to few hundred MeV for heavy ion induced reactions. The code accounts for the major nuclear reaction models, such as optical model, Coupled Channels and DWBA (ECIS06 and OPTMAN), Multi-step Direct (ORION + TRISTAN), NVWY Multi-step Compound, exciton model (PCROSS), hybrid Monte Carlo simulation (DDHMS), and the full featured Hauser-Feshbach model including width fluctuations and the optical model for fission. Heavy ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters based on the RIPL-3 library covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers, and γ-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations (BARFIT, MOMFIT). The results can be converted into the ENDF-6 format using the accompanying EMPEND code. Modules of the ENDF Utility Codes and the ENDF Pre-Processing codes are applied for ENDF file verification. The package contains the full EXFOR library of experimental data in computational format C4 that are automatically retrieved during the calculations. EMPIRE contains the resonance module that retrieves data from the electronic version of the Atlas of Neutron Resonances by Mughabghab (not provided with the EMPIRE distribution), to produce resonance section and related covariances for the
Radulovic, Vladimir; Barbot, Loic; Fourmentel, Damien; Villard, Jean-Francois [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St Paul-Lez-Durance, (France); Snoj, Luka; Zerovnik, Gasper [Jozef Stefan Institute, Reactor Physics Department, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Trkov, Andrej [IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna, (Austria)
2015-07-01
Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which
The ripple electromagnetic calculation: accuracy demand and possible responses
Cocilovo, V.; Ramogida, G.; Formisano, A.; Martone, R.; Portone, A.; Roccella, M.; Roccella, R.
2006-01-01
Due to a number of causes (the finite number of toroidal field coils or the presence of concentrate blocks of magnetic materials, as the neutral beam shielding) the actual magnetic configuration in a Tokamak differs from the desired one. For example, a ripple is added to the ideal axisymmetric toroidal field, impacting the equilibrium and stability of the plasma column; as a further example the magnetic field out of plasma affects the operation of a number of critical components, included the diagnostic system and the neutral beam. Therefore the actual magnetic field has to be suitably calculated and his shape controlled within the required limits. Due to the complexity of its design, the problem is quite critical for the ITER project. In this paper the problem is discussed both from mathematical and numerical point of view. In particular, a complete formulation is proposed, taking into account both the presence of the non linear magnetic materials and the fully 3D geometry. Then the quality level requirements are discussed, included the accuracy of calculations and the spatial resolution. As a consequence, the numerical tools able to fulfil the quality needs while requiring reasonable computer burden are considered. In particular possible tools based on numerical FEM scheme are considered; in addition, in spite of the presence of non linear materials, the practical possibility to use Biot-Savart based approaches, as cross check tools, is also discussed. The paper also analyses the possible geometrical simplifications of the geometry able to make possible the actual calculation while guarantying the required accuracy. Finally the characteristics required for a correction system able to effectively counteract the magnetic field degradation are presented. Of course a number of examples will be also reported and commented. (author)
Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies
Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.
1981-01-01
The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru
PROLIB: code to create production library of nuclear data for design calculations
Wittkopf, W.A.; Tilford, J.M.; Furtney, M.
1977-02-01
The PROLIB program creates, updates, and edits the production library used in the B and W nuclear design system. The production library contains the material cross section data required to perform the thermal and epithermal spectrum calculations in the NULIF program. PROLIB collapses cross section data from the master libraries, produced by the ETOGM and THOR programs, to the desired production library group structures. The physics models that are used, the calculations that are performed in PROLIB, the input, and the output are described. Information that is required to use PROLIB along with a sample problem that illustrates the input and output formats and that provides a benchmark problem are given
Avdic, S.; Pesic, M.
1992-01-01
The ORTEC 580 Neutron Spectrometer system contains a detector unit in diode coincidence arrangement for measurement of fast neutron spectrum in the energy range from 1 MeV to 14 MeV. Numerical code HE3 for computation of semiconductor 3 He detector efficiency in a collimated neutron beam is based on analytical method in infinite diode approximation and Monte Carlo method for real spectrometer geometry. Calculations are performed in the first collision approximation in the detector active volume including evaluation of correction factors. Accuracy of relative detector efficiency calculation is improved by using neutron cross section from nuclear library ENDF/B-6. (author)
Calculation of static harmonics of a nuclear reactor using CITATION code
Belchior Junior, A.; Moreira, J.M.L.
1989-01-01
The CITATION code, which solves the multigroup diffusion equation by the finite difference method, calculates the fundamental λ-mode (harmonic) for nuclear reactors. In this work, two fission source correction methods are attempted to obtain higher λ-modes through the CITATION code. The two methods are compared, their advantages and disadvantages analysed and verified against analytical solutions. Two dimensional harmonic modes are calculated for the IEA-R1 research reactor and for the ANGRA-I power reactor. The results are shown in graphics and tables. (author) [pt
Strategies for CT tissue segmentation for Monte Carlo calculations in nuclear medicine dosimetry
Braad, P E N; Andersen, T; Hansen, Søren Baarsgaard
2016-01-01
in the ICRP/ICRU male phantom and in a patient PET/CT-scanned with 124I prior to radioiodine therapy. Results: CT number variations body CT examinations at effective CT doses ∼2 mSv. Monte Carlo calculated absorbed doses depended on both the number of media types and accurate......Purpose: CT images are used for patient specific Monte Carlo treatment planning in radionuclide therapy. The authors investigated the impact of tissue classification, CT image segmentation, and CT errors on Monte Carlo calculated absorbed dose estimates in nuclear medicine. Methods: CT errors...
Finite element method used in strength calculations of nuclear power plant pressure vessels
Hanulak, E.
1987-01-01
A software system based on the use of the finite element method in linear and nonlinear elastomechanics was developed for assessing the strength and service life of steam generators and pressurizers for WWER type nuclear power plants. The individual programs are briefly described. They are written in FORTRAN IV, some modules are in ASSEMBLER. Programs EGUSAP, NEANKO, ROSYNA are designed for the calculation of stress and deformation, programs ROSYNA, NEANKO and NTEPLO are used for the calculation of temperature fields. Programs SPOJ and STATES are used for assessing the strength and service life of screw joints and other nodes of the WWER-440 type steam generators and pressurizers. (Z.M.)
Microscopic calculations of nuclear matter collective flow in Nb(400 MeV/N) + Nb
Hoffer, J.B.; Kruse, H.; Molitoris, J.J.; Stoecker, H.
1984-01-01
The recent experimental observation of sidewards peaks in the emission pattern of fragments emitted in collisions of heavy nuclear systems has stimulated a dispute among theorists about how to interpret these data. It has been shown that the observations are in agreement with the results of macroscopic nuclear fluid dynamical calculations, but several microscopic calculations done to simulate the sidewards emission (via the intranuclear cascade (INC) approach) failed - the angular distributions obtained where always forward peaked. A many body equations of motion (EOM) approach to study heavy ion collision has been developed. The approach is analogous to the early work of Bodmer et al., and Wilets et al. Hamilton's equations of motion are solved for an ensemble of nucleons with simultaneous mutual two-body interactions between all particles. The model predicts the sidewards emission peaks for the Nb + Nb reaction
Calculation of health risks from spent-nuclear-fuel transportation accidents
Chen, S.Y.; Yuan, Y.C.
1987-01-01
Models developed to analyze potential radiological health risks from various accident scenarios during transportation of spent nuclear fuels are described. The models are designed both for detailed route-specific risk analyses and for use in conducting overall risk analyses for route selection and related decision-making activities. The radiological risks calculated include individual dose commitments, collective dose commitments, and long-term (100-year) environmental dose commitments to a population following release of radioactivity. To facilitate route-specific analysis, a state-level database was developed and incorporated into the model. Route-specific analysis is demonstrated by the calculation of radiological risks resulting from various accident scenarios, as postulated by the recent US Nuclear Regulatory Commission Modal Study, for four representative states selected from various regions of the United States. 10 refs., 3 figs., 3 tabs
Thermal-hydraulic calculation and analysis for QNPP (Qinshan Nuclear Power Plant) containment
Xie Hui; Zhou Jie; He Yingchao
1993-01-01
Three containment thermal-hydraulic codes CONTEMPT-LT/028, CONTEMPT-4/MOD3 and COMPARE are used to compute and analyse the Qinshan Nuclear Power Plant (QNPP) containment response under LOCA or MSLB conditions. An evaluation of the capability of containment of QNPP is given
Rossi, Pedro Carlos Russo
2011-01-01
This work presents a study of high energy nuclear reactions which are fundamental to dene the source term in accelerator driven systems. These nuclear reactions, also known as spallation, consist in the interaction of high energetic hadrons with nucleons in the atomic nucleus. The phenomenology of these reactions consist in two step. In the rst, the proton interacts through multiple scattering in a process called intra-nuclear cascade. It is followed by a step in which the excited nucleus, coming from the intranuclear cascade, could either, evaporates particles to achieve a moderate energy state or fission. This process is known as competition between evaporation and fission. In this work the main nuclear models, Bertini and Cugnon are reviewed, since these models are fundamental for design purposes of the source term in ADS, due to lack of evaluated nuclear data for these reactions. The implementation and validation of the calculation methods for the design of the source is carried out to implement the methodology of source design using the program MCNPX (Monte Carlo N-Particle eXtended), devoted to calculation of transport of these particles and the validation performed by an international cooperation together with a Coordinated Research Project (CRP) of the International Atomic Energy Agency and available jobs, in order to qualify the calculations on nuclear reactions and the de-excitation channels involved, providing a state of the art of design and methodology for calculating external sources of spallation for source driven systems. The CRISP, is a brazilian code for the phenomenological description of the reactions involved and the models implemented in the code were reviewed and improved to continue the qualification process. Due to failure of the main models in describing the production of light nuclides, the multifragmentation reaction model was studied. Because the discrepancies in the calculations of production of these nuclides are attributes to the
Comparison of burnup calculation results using several evaluated nuclear data files
Suyama, Kenya; Katakura, Jun-ichi; Nomura, Yasushi
2002-01-01
Burn-up calculation and comparison of the results were carried out to clarify the differences among the following latest evaluated nuclear data libraries: JENDL-3.2, ENDF/B-VI and JEF-2.2. The analyses showed that the differences seen among the current evaluated nuclear data libraries are small for evaluation of the amounts of many uranium and plutonium isotopes. However, several nuclides important for evaluation of nuclear fuel cycle as 238 Pu, 244 Cm, 149 Sm and 134 Cs showed large differences among used libraries. The chain analyses for the isotopes were conducted and the reasons for the differences were discussed. Based on the discussion, information of important cross section to obtain better agreement with the experimental results for 238 Pu, 244 Cm, 149 Sm and 134 Cs was shown. (author)
Accurate and efficient calculation of response times for groundwater flow
Carr, Elliot J.; Simpson, Matthew J.
2018-03-01
We study measures of the amount of time required for transient flow in heterogeneous porous media to effectively reach steady state, also known as the response time. Here, we develop a new approach that extends the concept of mean action time. Previous applications of the theory of mean action time to estimate the response time use the first two central moments of the probability density function associated with the transition from the initial condition, at t = 0, to the steady state condition that arises in the long time limit, as t → ∞ . This previous approach leads to a computationally convenient estimation of the response time, but the accuracy can be poor. Here, we outline a powerful extension using the first k raw moments, showing how to produce an extremely accurate estimate by making use of asymptotic properties of the cumulative distribution function. Results are validated using an existing laboratory-scale data set describing flow in a homogeneous porous medium. In addition, we demonstrate how the results also apply to flow in heterogeneous porous media. Overall, the new method is: (i) extremely accurate; and (ii) computationally inexpensive. In fact, the computational cost of the new method is orders of magnitude less than the computational effort required to study the response time by solving the transient flow equation. Furthermore, the approach provides a rigorous mathematical connection with the heuristic argument that the response time for flow in a homogeneous porous medium is proportional to L2 / D , where L is a relevant length scale, and D is the aquifer diffusivity. Here, we extend such heuristic arguments by providing a clear mathematical definition of the proportionality constant.
FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations
Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.
1980-05-01
Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described
POPCYCLE: a computer code for calculating nuclear and fossil plant levelized life-cycle power costs
Hardie, R.W.
1982-02-01
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
About the application of MCNP4 code in nuclear reactor core design calculations
Svarny, J.
2000-01-01
This paper provides short review about application of MCNP code for reactor physics calculations performed in SKODA JS. Problems of criticality safety analysis of spent fuel systems for storage and transport of spent fuel are discussed and relevant applications are presented. Application of standard Monte Carlo code for accelerator driven system for LWR waste destruction is shown and conclusions are reviewed. Specific heterogeneous effects in neutron balance of WWER nuclear cores are solved for adjusting standard design codes. (Authors)
Application of optimization numerical methods in calculation of the two-particle nuclear reactions
Titarenko, N.N.
1987-01-01
An optimization packet of PEAK-OPT applied programs intended for solution of problems of absolute minimization of functions of many variables in calculations of cross sections of binary nuclear reactions is described. The main algorithms of computerized numerical solution of systems of nonlinear equations for the least square method are presented. Principles for plotting and functioning the optimization software as well as results of its practical application are given
Toledo Piza, A.F.R. de.
1987-01-01
The Random Phase Approximation (RPA) treatment of nuclear small amplitude vibrations including particle-hole continua is handled in terms of previously developed techniques to treat single-particle resonances in a reaction theoretical framework. A hierarchy of interpretable approximations is derived and a simple working approximation is proposed which involves a numerical effort no larger than that involved in standard, discrete RPA calculations. (Author) [pt
Calculational prediction of fuel burn-up for the Dalat Nuclear Research Reactor
Nguyen Phuoc Lan; Do Quang Binh
2016-01-01
In this paper, the method of expanding operators and functions in the neutron diffusion equations as chains of time variable is used for calculation of fuel burn-up of the Dalat nuclear reactors. A computer code, named BURREF, programmed in language Fortran-77 running on IBM PC-AT, has been developed based on this method to predict the fuel burn-up of the Dalat reactor. Some results will be presented here. (author)
Brenk, H.D.; Vogt, K.J.
1977-01-01
An evaluation of the environmental impact of nuclear plants according to paragraph 45 of the Radiation Protection Directive of the Federal Republic of Germany requires the calculation of dose conversion factors indicating the correlation between the contaminated medium and individual radiation exposure. The present study is to be conceived as a contribution to discussion on this subject. For the determination of radiation exposure caused by the waste air of nuclear plants, models are being specified for computing the dose conversion factors for the external exposure pathways of β-submersion, γ-submersion and γ-radiation from contaminated ground as well as the internal exposure pathways of inhalation and ingestion, which further elaborate and improve the models previously applied, especially as far as the ingestion pathway is concerned, which distinguishes between 6 major food categories. The computer models are applied to those radionuclides which are significan for nuclear emitters, in particular nuclear light-water power stations. The results obtained for the individual exposure pathways and affected organs are specified in the form of tables. For this purpose, calculations were first of all carried out for the so-called 'reference man'. The results can be transferred to population groups with different consumption habits (e.g. vegetarians) by the application of correction factors. The models are capable of being extended with a view to covering other age groups. (orig.) [de
Legal responsibility for damage caused by nuclear accidents
Brink, J.J. van den.
1986-04-01
In this essay a treatment is given of the legal third-party risks of licencees of nuclear power plants. It is regarded to what extent the actual responsibility arrangements provide an adequate protection to the citizen against potential risks of a nuclear power plant. (Auth.)
Analytical predictions of SGEMP response and comparisons with computer calculations
de Plomb, E.P.
1976-01-01
An analytical formulation for the prediction of SGEMP surface current response is presented. Only two independent dimensionless parameters are required to predict the peak magnitude and rise time of SGEMP induced surface currents. The analysis applies to limited (high fluence) emission as well as unlimited (low fluence) emission. Cause-effect relationships for SGEMP response are treated quantitatively, and yield simple power law dependencies between several physical variables. Analytical predictions for a large matrix of SGEMP cases are compared with an array of about thirty-five computer solutions of similar SGEMP problems, which were collected from three independent research groups. The theoretical solutions generally agree with the computer solutions as well as the computer solutions agree with one another. Such comparisons typically show variations less than a ''factor of two.''
Current nuclear threats and possible responses
Lamb, Frederick K.
2005-04-01
Over the last 50 years, the United States has spent more than 100 billion developing and building a variety of systems intended to defend its territory against intercontinental-range ballistic missiles. Most of these systems never became operational and ultimately all were judged ineffective. The United States is currently spending about 10 billion per year developing technologies and systems intended to defend against missiles that might be acquired in the future by North Korea or Iran. This presentation will discuss these efforts ad whether they are likely to be more effective than those of the past. It will also discuss the proper role of anti-ballistic programs at a time when the threat of a nuclear attack on the U.S. by terrorists armed with nuclear weapons is thought to be much higher than the threat of an attack by nuclear-armed ballistic missles.
Emergency preparedness and response plan for nuclear facilities in Indonesia
Nur Rahmah Hidayati; Pande Made Udiyani
2009-01-01
All nuclear facilities in Indonesia are owned and operated by the National Nuclear Energy Agency (BATAN). The programs and activities of emergency planning and preparedness in Indonesia are based on the existing nuclear facilities, i.e. research reactors, research reactor fuel fabrication plant, radioactive waste treatment installation and radioisotopes production installation. The assessment is conducted to learn of status of emergency preparedness and response plan for nuclear facilities in Indonesia and to support the preparation of future Nuclear Power Plant. The assessment is conducted by comparing the emergency preparedness and response system in Indonesia to the system in other countries such as Japan and Republic of Korea, since the countries have many Nuclear Power Plants and other nuclear facilities. As a result, emergency preparedness response plan for existing nuclear facility in Indonesia has been implemented in many activities such as environmental monitoring program, facility monitoring equipment, and the continuous exercise of emergency preparedness and response. However, the implementation need law enforcement for imposing the responsibility of the coordinators in National Emergency Preparedness Plan. It also needs some additional technical support systems which refer to the system in Japan or Republic of Korea. The systems must be completed with some real time monitors which will support the emergency preparedness and response organization. The system should be built in NPP site before the first NPP will be operated. The system should be connected to an Off Site Emergency Center under coordination of BAPETEN as the regulatory body which has responsibility to control of nuclear energy in Indonesia. (Author)
Dolgaya, A.A.; Uzdin, A.M.; Indeykin, A.V.
1993-01-01
The investigation object is the design amplitude of accelerograms, which are used in the evaluation of seismic stability of responsible structures, first and foremost, NPS. The amplitude level is established depending on the degree of responsibility of the structure and on the prevailing period of earthquake action on the construction site. The investigation procedure is based on statistical analysis of 310 earthquakes. At the first stage of statistical data-processing we established the correlation dependence of both the mathematical expectation and root-mean-square deviation of peak acceleration of the earthquake on its prevailing period. At the second stage the most suitable law of acceleration distribution about the mean was chosen. To determine of this distribution parameters, we specified the maximum conceivable acceleration, the excess of which is not allowed. Other parameters of distribution are determined according to statistical data. At the third stage the dependencies of design amplitude on the prevailing period of seismic effect for different structures and equipment were established. The obtained data made it possible to recommend to fix the level of safe-shutdown (SSB) and operating basis earthquakes (OBE) for objects of various responsibility categories when designing NPS. (author)
Consideration of vertical seismic response spectrum in nuclear safety review
Sun Zaozhan; Huang Bingchen
2011-01-01
The basic requirements for civil nuclear installation are introduced in the article. Starting from the basic concept of seismic response spectrum, the authors analyze the site seismic response spectrum and the design seismic response spectrum that desire much consideration. By distinguishing the absolute seismic response spectrum and relative seismic response spectrum, the authors analyze the difference and relationship between the vertical seismic response spectrum and horizontal seismic response spectrum. The authors also bring forward some suggestions for determining the site vertical seismic response spectrum by considering the fact in our country. (authors)
Ondra, Frantisek; Daniska, Vladimir; Rehak, Ivan; Necas, Vladimir
2009-01-01
The aim of the article is a development of analytical methodology for evaluation of input data inaccuracies impact on calculation of cost and other output decommissioning parameters. This methodology is based on analytical model calculations using the OMEGA code and taking into account the probability of input data inaccuracies occurrence also. To achieve about mentioned aim, the article identifies possible sources of input data inaccuracies and analyzes their level of impact on output parameters. Then the methodology for calculation of input parameters inaccuracies impact is developed, based on analytical model calculation. The model calculation takes into consideration output parameters impact on cost and other decommissioning output parameters in analytical way. The methodology used in model calculations is original, more over it implements the international standardized structure (IAEA, OECD/NEA, EC) [6] of decommissioning cost for the first time. A probabilistic occurrence of input data inaccuracies is taken into consideration and implemented in the methodology developed. A correction factors matrix for evaluation of input data inaccuracies impact on decommissioning output parameters is set up. The matrix contains parameters based on model calculations using the proposed methodology. Finally the methodology for application of correction factor matrix is proposed and tested; the methodology is used for calculation of contingency in the standardized structure which reflected the level of input data inaccuracies. The cost for individual decommissioning projects for common nuclear power plants are in the range 300 - 500 mil. EUR. Contingencies are from 10% to 30%, depending on the level of detailed during preparation of decommissioning projects. A implementation about mentioned methodology in the OMEGA code improves the accuracy of contingency. Consequently it makes calculated contingency more trustworthy and makes calculated decommissioning cost closer to reality
Nuclear fear and children: the impact of parental nuclear activism, responsivity, and fear
LaGuardia, M.R.
1986-01-01
This study examines the extent to which parental nuclear fear, parental activism, and parental responsivity is associated with children's (age 10) nuclear fear. Other associated variables investigated include: nuclear denial, general anxiety and fear, and the personal characteristics of sex, socio-economic status, and academic aptitude. Findings indicate that children attend to nuclear issues when their parents attend to a significant degree. Children's hopelessness about the arms race is increased as parents' worry about nuclear war increases. Children's fear about not surviving a nuclear war increases as parents' worry about survivability decreases. Children who have more general fears also indicated that they have a high level of hopelessness, pervasive worry, and much concern about being able to survive a nuclear war. Children with a high degree of general anxiety did not indicate high degrees of nuclear fears. Children with high academic aptitude were more knowledgeable about nuclear issues and expressed more fears about the nuclear threat. Boys demonstrated more knowledge about nuclear issues than girls, and girls expressed much more frequent fear and worry about the nuclear threat than boys. Parents of lower socio-economic statues (SES) expressed more denial about the nuclear threat and were more pro-military than the higher SES parents.
Cost calculations at early stages of nuclear research facilities in the nordic countries
Iversen, Klaus; Salmenhaara, Seppo; Backe, Steinar; Cato, Anna; Lindskog, Staffan; Callander, Clas; Efraimsson, Henrik; Andersson, Inga; Sjoeblom, Rolf
2007-01-01
The Nordic countries Denmark, Norway and Sweden, and to some extent also Finland, had very large nuclear research and development programs for a few decades starting in the nineteen fifties. Today, only some of the facilities are in use. Some have been decommissioned and dismantled while others are at various stages of planning for shutdown. The perspective ranges from imminent to several decades. It eventually became realized that considerable planning for the future decommissioning is warranted and that an integral part of this planning is financial, including how financial funds should be acquired, used and allocated over time. This necessitates that accurate and reliable cost estimates be obtained at all stages. However, this is associated with fundamental difficulties and treacherous complexities, especially for the early ones. Eventually, Denmark and Norway decided not to build any nuclear power plants while Finland and Sweden did. This is reflected in the financing where the latter countries have established systems with special funds in which money is being collected now to cover the future costs for the decommissioning of the research facilities. Nonetheless, the needs for planning for the decommissioning of nuclear research facilities are very similar. However, they differ considerably from those of nuclear power reactors, especially with regard to cost calculations. It has become apparent in the course of work that summation types of cost estimation methodologies give rise to large systematic errors if applied at early stages, in which case comparison based assessments are less biased and may be more reliable. Therefore, in order to achieve the required quality of the cost calculations, it is necessary that data and experience from authentic cases be utilized in models for cost calculations. It also implies that this calculation process should include a well adopted learning process. Thus, a Nordic co-operation has been established for the exchange and
Kubo, H.; Harada, K.; Sakaeda, T.; Yamamoto, Y.
2013-01-01
On the basis of the Wilsonian renormalization group (WRG) analysis of nuclear effective field theory (NEFT) including pions, we propose a practical calculational scheme in which the short-distance part of one-pion exchange (S-OPE) is removed and represented as contact terms. The long-distance part of one-pion exchange (L-OPE) is treated as perturbation. The use of dimensional regularization (DR) for diagrams consisting only of contact interactions considerably simplifies the calculation of scattering amplitude and the renormalization group equations. NLO results for nucleon-nucleon elastic scattering in the S-waves are obtained and compared with experiments. A brief comment on NNLO calculations is given. (author)
Thorlaksen, B.
1981-05-01
Nuclear cross sections for fuel assemblies of the more recent Westinghouse designs, representing two different PWR reactor cores, are calculated as functions of average fuel temperature, moderator density, and moderator poison concentration. The cross-section functions are verified by referring to Westinghouse power-shape calculations and other analysis. Computations on the side reflector resulted in significantly higher albedo values than used previously for BWR's in similar nodal codes. This led to an investigation of the influence of the internodal coupling coefficients on the power shape. It is concluded that the calculated power shape is strongly dependent, on the choise of coupling coefficients. However, it is shown that ''the correct'' set of coupling coefficients depends mostly on the nodal configuration, and that it is fairly independent of the power condition. (author)
n + 2759Co(En≤20 MeV) nuclear data calculation and analysis
Wang Shunuan
2006-01-01
Whole set of nuclear data calculation in ENDF/B-6 format for n + 27 59 Co (E n ≤20 MeV) has been finished by using spherical optical model, coupled channel optical model, pre-equilibrium exciton model and Hauser-Fashbach equilibrium statistical model. The calculated cross sections, angular distributions, spectrum and double differential cross sections by using codes of APOM, ECIS95 and UNF are compared with all existing experimental data for n + 27 59 Co(E n ≤20 MeV) takefrom EXFOR. The calculated results are analyzed from point of view of theoretical model and model parameters used. The work is for CENDL-3. (authors)
Verification of design calculations of a PGNAA setup using nuclear track ejectors
Naqvi, A.A. E-mail: aanaqvi@kfupm.edu.sa; Fazal-ur-Rehman,; Nagadi, .M.; Maslehuddin, M.; Khateeb-ur-Rehman; Kidwai, S
2004-02-01
A rectangular moderator assembly has been designed for the PGNAA setup at ing Fahd University of Petroleum and Minerals (KFUPM). The design calculations of the rectangular moderator, which were obtained through Monte Carlo simulation, have been verified experimentally through thermal neutron field measurement using CR-39 nuclear track detectors (NTDs). These measurements were carried out at the KFUPM 350 keV accelerator using 2.8 MeV pulsed neutron beam from D(d,n) reaction. The thermal neutron yield was measured inside the sample volume of the rectangular moderator by two NTDs fixed at back and front end of the sample cavity. The good agreement between he experimental results and the results of the calculations shows useful application of NTDs in verification of design calculations of a PGNAA setup.
Cross Sections Calculations of ( d, t) Nuclear Reactions up to 50 MeV
Tel, E.; Yiğit, M.; Tanır, G.
2013-04-01
In nuclear fusion reactions two light atomic nuclei fuse together to form a heavier nucleus. Fusion power is the power generated by nuclear fusion processes. In contrast with fission power, the fusion reaction processes does not produce radioactive nuclides. The fusion will not produce CO2 or SO2. So the fusion energy will not contribute to environmental problems such as particulate pollution and excessive CO2 in the atmosphere. Fusion powered electricity generation was initially believed to be readily achievable, as fission power had been. However, the extreme requirements for continuous reactions and plasma containment led to projections being extended by several decades. In 2010, more than 60 years after the first attempts, commercial power production is still believed to be unlikely before 2050. Although there have been significant research and development studies on the inertial and magnetic fusion reactor technology, there is still a long way to go to penetrate commercial fusion reactors to the energy market. In the fusion reactor, tritium self-sufficiency must be maintained for a commercial power plant. Therefore, for self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. Working out the systematics of ( d, t) nuclear reaction cross sections is of great importance for the definition of the excitation function character for the given reaction taking place on various nuclei at different energies. Since the experimental data of charged particle induced reactions are scarce, self-consistent calculation and analyses using nuclear theoretical models are very important. In this study, ( d, t) cross sections for target nuclei 19F, 50Cr, 54Fe, 58Ni, 75As, 89Y, 90Zr, 107Ag, 127I, 197Au and 238U have been investigated up to 50 MeV deuteron energy. The excitation functions for ( d, t) reactions have been calculated by pre-equilibrium reaction mechanism. Calculation results have been also compared with the available measurements in
Soviet medical response to the Chernobyl nuclear accident
Linnemann, R.E.
1987-01-01
The nuclear accident at Chernobyl was the worst in the history of nuclear power. It tested the organized medical response to mass radiation casualties. This article reviews the Soviet response as reported at the 1986 postaccident review meeting in Vienna and as determined from interviews. The Soviets used three levels of care: rescue and first aid at the plant site; emergency treatment at regional hospitals; and definitive evaluation and treatment in Moscow. Diagnosis, triage, patient disposition, attendant exposure, and preventive actions are detailed. The United States would be well advised to organize its resources definitively to cope with future nonmilitary nuclear accidents
Nuclear emergency planning and response in the Netherlands after Chernobyl
Bergman, L.J.W.M.; Kerkhoven, I.P.
1989-01-01
After Chernobyl an extensive project on nuclear emergency planning and response was started in the Netherlands. The objective of this project was to develop a (governmental) structure to cope with accidents with radioactive materials, that can threaten the Dutch community and neighbouring countries. The project has resulted in a new organizational structure for nuclear emergency response, that differs on major points from the existing plans and procedures. In this paper an outline of the new structure is given. Emphasis is placed on accidents with nuclear power plants
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
Garcia-Herranz, Nuria; Cabellos, Oscar; Sanz, Javier; Juan, Jesus; Kuijper, Jim C.
2008-01-01
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files
Propagation of statistical and nuclear data uncertainties in Monte Carlo burn-up calculations
Garcia-Herranz, Nuria [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain)], E-mail: nuria@din.upm.es; Cabellos, Oscar [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, UPM (Spain); Sanz, Javier [Departamento de Ingenieria Energetica, Universidad Nacional de Educacion a Distancia, UNED (Spain); Juan, Jesus [Laboratorio de Estadistica, Universidad Politecnica de Madrid, UPM (Spain); Kuijper, Jim C. [NRG - Fuels, Actinides and Isotopes Group, Petten (Netherlands)
2008-04-15
Two methodologies to propagate the uncertainties on the nuclide inventory in combined Monte Carlo-spectrum and burn-up calculations are presented, based on sensitivity/uncertainty and random sampling techniques (uncertainty Monte Carlo method). Both enable the assessment of the impact of uncertainties in the nuclear data as well as uncertainties due to the statistical nature of the Monte Carlo neutron transport calculation. The methodologies are implemented in our MCNP-ACAB system, which combines the neutron transport code MCNP-4C and the inventory code ACAB. A high burn-up benchmark problem is used to test the MCNP-ACAB performance in inventory predictions, with no uncertainties. A good agreement is found with the results of other participants. This benchmark problem is also used to assess the impact of nuclear data uncertainties and statistical flux errors in high burn-up applications. A detailed calculation is performed to evaluate the effect of cross-section uncertainties in the inventory prediction, taking into account the temporal evolution of the neutron flux level and spectrum. Very large uncertainties are found at the unusually high burn-up of this exercise (800 MWd/kgHM). To compare the impact of the statistical errors in the calculated flux with respect to the cross uncertainties, a simplified problem is considered, taking a constant neutron flux level and spectrum. It is shown that, provided that the flux statistical deviations in the Monte Carlo transport calculation do not exceed a given value, the effect of the flux errors in the calculated isotopic inventory are negligible (even at very high burn-up) compared to the effect of the large cross-section uncertainties available at present in the data files.
Sunder, S.; Shoesmith, D.W.; Kolar, M.; Leneveu, D.M.
1998-01-01
Calculation of used nuclear fuel dissolution rates in a geological disposal vault requires a knowledge of the redox conditions in the vault. For redox conditions less oxidizing than those causing UO 2 oxidation to the U 3 O 7 , stage, a thermodynamically-based model is appropriate. For more oxidizing redox conditions a kinetic or an electrochemical model is needed to calculate these rates. The redox conditions in a disposal vault will be affected by the radiolysis of groundwater by the ionizing radiation associated with the fuel. Therefore, we have calculated the alpha-, beta- and gamma-dose rates in water in contact with the reference used fuel in the Canadian Nuclear Fuel Waste Management Program (CNFWMP) as a function of cooling time. Also, we have determined dissolution rates of UO 2 fuel as a function of alpha and gamma dose rates from our electrochemical measurements. These room-temperature rates are used to calculate the dissolution rates of used fuel at 100 o C, the highest temperature expected in a container in the CNFWMP, as a function of time since emplacement. It is shown that beta radiolysis of water will be the main cause of oxidation of used CANDU fuel in a failed container. The use of a kinetic or an electrochemical corrosion model, to calculate fuel dissolution rates, is required for a period of ∼1000 a following emplacement of copper containers in the geologic disposal vault envisaged in the CNFWMP. Beyond this time period a thermodynamically-based model adequately predicts the fuel dissolution rates. The results presented in this paper can be adopted to calculate used fuel dissolution rates for other used UO 2 fuels in other waste management programs. (author)
Responsibilities and capabilities of a nuclear energy programme implementing organization
2009-01-01
An appropriate infrastructure is essential for the efficient, safe, reliable and peaceful use of nuclear power. The IAEA was encouraged by its Member States to provide assistance to those considering the introduction of nuclear power. These countries face the challenge of building a national nuclear infrastructure to support a first nuclear power plant. The IAEA is responding to their needs through increased technical assistance, missions and workshops, and with new and updated technical publications in the IAEA Nuclear Energy Series. Milestones in the Development of a National Infrastructure for Nuclear Power, an IAEA Nuclear Energy Series publication (NG-G-3.1), provides detailed guidance on a holistic approach to national nuclear infrastructure development, over three phases. Nineteen issues are identified in this guide, ranging from development of a government's national position on nuclear power to planning for procurement related to the first NPP. An important element of the holistic approach is an entity that can help prepare the decision makers in a country to make a knowledgeable commitment to nuclear power, and then to coordinate infrastructure development efforts among various implementing organizations so that they arrive at the point of readiness to issue a bid tender at the same time. In the Milestones guide, this entity is called a nuclear energy programme implementing organization (NEPIO). As a growing number of Member States started to consider the nuclear power option, they asked for guidance from the IAEA on how to launch a nuclear power programme. In particular, Member States requested additional information on how to establish a NEPIO, especially in the earliest phases of a programme. This report has been prepared to provide information on the responsibilities and capabilities of a NEPIO, as well as to give an indication on how it relates to other key national organizations in the implementation of a nuclear power programme, such as the owner
More efficient response to nuclear emergencies
1979-12-01
Data provided by the local authorities in the counties in which the Oskarshamn and Barsebaeck nuclear power plants are situated is presented. The data is for planning of evaluation in the case of a reactor accident and includes population, population distribution, age distribution, institutions such as schools and hospitals, transport, both public and private and accommodation possibilities. Agricultural and domestic animal data are also provided. (J.I.W.)
Preventing nuclear terrorism: responses to terrorist grievances
Beres, L.R.
1987-01-01
The US is vulnerable to nuclear terrorism, despite the presence of physical security and other measures. Although these measures are important, they are insufficient to prevent or deter terrorism. What, then, is the answer? The author feels it lies in a hitherto neglected dimension of terrorism: its underlying political grievances. The principal grievance that potential terrorists have against the US concerns misguided elements of US foreign policy. These elements are moving the US on a seemingly inexorable collision course with terrorism and, more than likely, with nuclear terrorism. The US represents a serious threat to many people and groups who feel directly the effects of a foreign policy mired in strident anti-Sovietism: opponents of the US-NATO Euromissile deployments, populations seeking to secure their human rights from repressive regimes supported by the US, and governments seeking self-determination but embattled by insurgents backed by US arms, equipment, and advisers. In many cases, the US foreign policy stance in one country has aroused suspicion and anger within the region as a whole. The collision course need not be inevitable. The US can take a number of steps in the political arena that would greatly reduce the threat of nuclear terrorism
A calculation model for a HTR core seismic response
Buland, P.; Berriaud, C.; Cebe, E.; Livolant, M.
1975-01-01
The paper presents the experimental results obtained at Saclay on a HTGR core model and comparisons with analytical results. Two series of horizontal tests have been performed on the shaking table VESUVE: sinusoidal test and time history response. Acceleration of graphite blocks, forces on the boundaries, relative displacement of the core and PCRB model, impact velocity of the blocks on the boundaries were recorded. These tests have shown the strongly non-linear dynamic behaviour of the core. The resonant frequency of the core is dependent on the level of the excitation. These phenomena have been explained by a computer code, which is a lumped mass non-linear model. Good correlation between experimental and analytical results was obtained for impact velocities and forces on the boundaries. This comparison has shown that the damping of the core is a critical parameter for the estimation of forces and velocities. Time history displacement at the level of PCRV was reproduced on the shaking table. The analytical model was applied to this excitation and good agreement was obtained for forces and velocities. (orig./HP) [de
Uncertain hybrid model for the response calculation of an alternator
Kuczkowiak, Antoine
2014-01-01
The complex structural dynamic behavior of alternator must be well understood in order to insure their reliable and safe operation. The numerical model is however difficult to construct mainly due to the presence of a high level of uncertainty. The objective of this work is to provide decision support tools in order to assess the vibratory levels in operation before to restart the alternator. Based on info-gap theory, a first decision support tool is proposed: the objective here is to assess the robustness of the dynamical response to the uncertain modal model. Based on real data, the calibration of an info-gap model of uncertainty is also proposed in order to enhance its fidelity to reality. Then, the extended constitutive relation error is used to expand identified mode shapes which are used to assess the vibratory levels. The robust expansion process is proposed in order to obtain robust expanded mode shapes to parametric uncertainties. In presence of lack-of knowledge, the trade-off between fidelity-to-data and robustness-to-uncertainties which expresses that robustness improves as fidelity deteriorates is emphasized on an industrial structure by using both reduced order model and surrogate model techniques. (author)
Gritzay, O.; Kalchenko, O.
2010-01-01
Full text: Scientific support of NPPs has to cover several important aspects of scientific and organization activity, namely:1.Training for group of high skilled specialists to do the following work: o nuclear data generation for engineer calculations; o engineer calculations to ensure the safety operation of NPPs; o experimental-calculation support of fluence dosimetry at NPP. 2.Development of up-to-date computer base, equipped with necessary program packages for nuclear data generation and engineer calculations. 3.The updated Libraries of Evaluated Nuclear Data (ENDF), such as ENDF/B-VII (USA), JENDL-3.3 (Japan) and JEFF-3.1 (Europe), RUSFOND ( Russia) and as a result the generation of specialized nuclear data multi-group libraries for special purpose engineer calculations.To reach these purposes, the Ukrainian Nuclear Data Center (UKRNDC) was organized and developed for more, than 10 years (since 1996).The capabilities of the UKRNDC are detailed below. o Modern ENDF libraries, first of all the general purpose libraries, such as ENDF/B-7.0, -6.8, JEFF-3.1.1, JENDL-3.3, etc. These databases contain recommended, evaluated cross sections, spectra, angular distributions, fission product yields, photo-atomic and thermal scattering law data, with emphasis on neutron induced reactions.o Codes for processing these data, updated to the last versions of ENDF and other libraries. First of all these are PREPRO 2007 package (Updated March 17, 2007) and NJOY package updated to versions NJOY-158 and NJOY-253 (in 2009). These codes may give the possibilities to produce the multi-group data for needed spectrum of interacting particles (neutrons, protons, gammas) and temperatures.o Computer base of several specialized server stations, such as ESCALA- S120 (analogous to IBM -240 with RISC 6000 processor) operating under OS under OS UNIX (version AIX 5.1) and IBM PC operating under Linux Red Hat 7.2.o The set of PC computers joined in UKRNDC network, operating mainly in OS Windows
M. Gross
2004-01-01
The purpose of this scientific analysis is to define the sampled values of stochastic (random) input parameters for (1) rockfall calculations in the lithophysal and nonlithophysal zones under vibratory ground motions, and (2) structural response calculations for the drip shield and waste package under vibratory ground motions. This analysis supplies: (1) Sampled values of ground motion time history and synthetic fracture pattern for analysis of rockfall in emplacement drifts in nonlithophysal rock (Section 6.3 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (2) Sampled values of ground motion time history and rock mechanical properties category for analysis of rockfall in emplacement drifts in lithophysal rock (Section 6.4 of ''Drift Degradation Analysis'', BSC 2004 [DIRS 166107]); (3) Sampled values of ground motion time history and metal to metal and metal to rock friction coefficient for analysis of waste package and drip shield damage to vibratory motion in ''Structural Calculations of Waste Package Exposed to Vibratory Ground Motion'' (BSC 2004 [DIRS 167083]) and in ''Structural Calculations of Drip Shield Exposed to Vibratory Ground Motion'' (BSC 2003 [DIRS 163425]). The sampled values are indices representing the number of ground motion time histories, number of fracture patterns and rock mass properties categories. These indices are translated into actual values within the respective analysis and model reports or calculations. This report identifies the uncertain parameters and documents the sampled values for these parameters. The sampled values are determined by GoldSim V6.04.007 [DIRS 151202] calculations using appropriate distribution types and parameter ranges. No software development or model development was required for these calculations. The calculation of the sampled values allows parameter uncertainty to be incorporated into the rockfall and structural response calculations that support development of the seismic scenario for the
Calculation-measurement comparison for control rods reactivity in RA-3 nuclear reactor
Estryk, Guillermo; Gomez, Angel
2002-01-01
The RA-3 Nuclear Reactor of the Atomic Energy National Commission from Argentina, begun working with high enrichment fuel elements in 1967, and turned to low enrichment by 1990. During 1999 it was found out that several fuel elements had problems, so more than 50 % of them had to be removed from the core. Because of this, it was planned to go from core 93 to core 94 with special care from nuclear safety point of view. Core 94 was preceded by other five, T-1 to T-5, only as transitory ones. The care implied several nuclear parameters measurements: core reactivity excess, calibration of control rods, etc. Calculations were performed afterwards to simulate those measurements using the neutron diffusion code PUMA. The comparison shows a good agreement for more than 80% of the cases with differences lower than 10% in reactivity. The greatest differences were found in the last part of the control rods calibration and a better calculation of cell constants is planned to be done in order to improve the adjustment. (author)
Calculation of the atomic electric dipole moment of Pb2+ induced by nuclear Schiff moment
Ramachandran, S. M.; Latha, K. V. P.; Meenakshisundaram, N.
2017-07-01
We report the atomic electric dipole moment induced by the P, T violating interactions in the nuclear/sub-nuclear level, for 207Pb2+ and 207Pb, owing to the recent interest in the ferroelectric crystal PbTiO3 as one of the candidates for investigating macroscopic P, T-odd effects. In this paper, we calculate the atomic electric dipole moments of 207Pb and Pb2+, parametrized in terms of the P, T-odd coupling parameter, the nuclear Schiff moment (NSM), S, in the frame-work of the coupled-perturbed Hartree-Fock theory. We estimate the Schiff moment of Pb2+ using the experimental result of a system, which is electronically similar to the Pb2+ ion. We present the dominant contributions of the electric dipole moment (EDM) matrix elements and the important correlation effects contributing to the atomic EDM of Pb2+. Our results provide the first ever calculated EDM of the Pb2+ ion, and an estimate of its NSM from which the P, T-odd energy shift in a PbTiO3 crystal can be evaluated.
Approximate calculation method for integral of mean square value of nonstationary response
Aoki, Shigeru; Fukano, Azusa
2010-01-01
The response of the structure subjected to nonstationary random vibration such as earthquake excitation is nonstationary random vibration. Calculating method for statistical characteristics of such a response is complicated. Mean square value of the response is usually used to evaluate random response. Integral of mean square value of the response corresponds to total energy of the response. In this paper, a simplified calculation method to obtain integral of mean square value of the response is proposed. As input excitation, nonstationary white noise and nonstationary filtered white noise are used. Integrals of mean square value of the response are calculated for various values of parameters. It is found that the proposed method gives exact value of integral of mean square value of the response.
Nuclear-data uncertainty propagations in burnup calculation for the PWR assembly
Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei
2017-01-01
Highlights: • The DRAGON 5.0 and NECP-CACTI have been implemented in UNICORN. • The effects of different neutronics methods on S&U results were quantified. • Uncertainty analysis has been applied to burnup calculation of PWR assembly. • The uncertainties of eigenvalue and few-group constants have been quantified. - Abstract: In this paper, our home-developed lattice code NECP-CACTI has been implemented into our UNICORN code to perform sensitivity and uncertainty analysis for the lattice calculations. The verified multigroup cross-section perturbation model and methods of the sensitivity and uncertainty analysis are established and applied to different lattice codes in UNICORN. As DRAGON5.0 and NECP-CACTI are available for the lattice calculations in UNICORN now, the effects of different neutronics methods (including methods for the neutron-transport and resonance self-shielding calculations) on the results of sensitivity and uncertainty analysis were studied in this paper. Based on NECP-CACTI, uncertainty analysis using the statistical sampling method has been performed to the burnup calculation for the fresh-fueled TMI-1 assembly, propagating the nuclear-data uncertainties to k_∞ and two-group constants of the lattice calculation with depletions. As results shown, for different neutronics methods, it can be observed that different methods of the neutron-transport calculation introduce no differences to the results of sensitivity and uncertainty analysis, while different methods of the resonance self-shielding calculation would impact the results. With depletions of the TMI-1 assembly, for k_∞, the relative uncertainty varies between 0.45% and 0.60%; for two-group constants, the largest variation is between 0.35% and 2.56% for vΣ_f_,_2. Moreover, the most significant contributors to the uncertainty of k_∞ and two-group constants varied with depletions are determined.
Capote Noy, R.
2004-08-01
A summary is given of the First Research Coordination Meeting on Parameters for Calculation of Nuclear Reactions of Relevance to Non-Energy Nuclear Applications (Reference Input Parameter Library: Phase III), including a critical review of the RIPL-2 file. The new library should serve as input for theoretical calculations of nuclear reaction data at incident energies up to 200 MeV, as needed for energy and non-energy modern applications of nuclear data. Technical discussions and the resulting work plan of the Coordinated Research Programme are summarized, along with actions and deadlines. Participants' contributions to the RCM are also attached. (author)
Freeman, L.B. (ed.)
1978-08-01
The calculational model used in the analysis of LWBR nuclear performance is described. The model was used to analyze the as-built core and predict core nuclear performance prior to core operation. The qualification of the nuclear model using experiments and calculational standards is described. Features of the model include: an automated system of processing manufacturing data; an extensively analyzed nuclear data library; an accurate resonance integral calculation; space-energy corrections to infinite medium cross sections; an explicit three-dimensional diffusion-depletion calculation; a transport calculation for high energy neutrons; explicit accounting for fuel and moderator temperature feedback, clad diameter shrinkage, and fuel pellet growth; and an extensive testing program against experiments and a highly developed analytical standard.
Calculation code evaluating the confinement of a nuclear facility in case of fires
Laborde, J.C.; Prevost, C.; Vendel, J.
1995-01-01
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
Calculation code evaluating the confinement of a nuclear facility in case of fires
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.
Comprehensive nuclear model calculations: theory and use of the GNASH code
Young, P.G.; Arthur, E.D.; Chadwick, M.B.
1998-01-01
The theory and operation of the nuclear reaction theory computer code GNASH is described, and detailed instructions are presented for code users. The code utilizes statistical Hauser-Feshbach theory with full angular momentum conservation and includes corrections for preequilibrium effects. This version is expected to be applicable for incident particle energies between 1 keV and 150 MeV and for incident photon energies to 140 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. A number of new features compared to previous versions are described in this manual, including the following: (1) inclusion of multiple preequilibrium processes, which allows the model calculations to be performed above 50 MeV; (2) a capability to calculate photonuclear reactions; (3) a method for determining the spin distribution of residual nuclei following preequilibrium reactions; and (4) a description of how preequilibrium spectra calculated with the FKK theory can be utilized (the 'FKK-GNASH' approach). 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 93 Nb and 12-MeV neutrons incident on 238 U. 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. Results from a variety of other cases are illustrated. (author)
Guidelines for nuclear plant response to an earthquake
1989-12-01
Guidelines have been developed to assist nuclear plant personnel in the preparation of earthquake response procedures for nuclear power plants. The objectives of the earthquake response procedures are to determine (1) the immediate effects of an earthquake on the physical condition of the nuclear power plant, (2) if shutdown of the plant is appropriate based on the observed damage to the plant or because the OBE has been exceeded, and (3) the readiness of the plant to resume operation following shutdown due to an earthquake. Readiness of a nuclear power plant to restart is determined on the basis of visual inspections of nuclear plant equipment and structures, and the successful completion of surveillance tests which demonstrate that the limiting conditions for operation as defined in the plant Technical Specifications are met. The guidelines are based on information obtained from a review of earthquake response procedures from numerous US and foreign nuclear power plants, interviews with nuclear plant operations personnel, and a review of reports of damage to industrial equipment and structures in actual earthquakes. 7 refs., 4 figs., 4 tabs
STRUCTURAL CALCULATIONS FOR THE CODISPOSAL OF TRIGA SPENT NUCLEAR FUEL IN A WASTE PACKAGE
S. Mastilovic
1999-01-01
The purpose of this analysis is to determine the structural response of a TRIGA Department of Energy (DOE) spent nuclear fuel (SNF) codisposal canister placed in a 5-Defense High Level Waste (DHLW) waste package (WP) and subjected to a tipover design basis event (DBE) dynamic load; the results will be reported in terms of displacements and stress magnitudes. This activity is associated with the WP design
Report of nuclear utility industry responses to Kemeny Commission recommendations
1989-02-01
The purpose of this paper is to provide a report of nuclear utility industry progress in responding to the recommendations of the President's Commission on the Accident at Three Mile Island (The Kemeny Commission). On April 11, 1979, in response to TMI, President Carter established a Commission to conduct '.... a comprehensive study and investigation of the recent accident involving the nuclear power facility on Three Mile Island in Pennsylvania'. The Commission was chaired by Dr. John G. Kemeny, then President of Dartmouth College. (A list of all members of The Kemeny Commission is provided in Attachment to the Appendix ). The report of the commission's findings and recommendations was transmitted to the President in October 1979. During this same period, the nuclear utility industry responded to TMI by creating the Institute of Nuclear Power Operations (INPO) with a mission to promote the highest levels of safety and reliability - to promote excellence - in the operation of nuclear electric generating plants. In addition, the Nuclear Safety Analysis Center (NSAC) was established at the Electric Power Research Institute (EPRI to evaluate the accident and assist in determining the best industry response. In a White House paper (and press release) of December 7 1979, the President announced that he agreed fully with the spirit and intent of al the Kemeny Commission recommendations and requested that the industry and The Nuclear Regulatory Commission (NRC) comply with the recommendations. The President also recognized the industry initiative in establishing INPO and called for several actions involving the Institute; the President directed the Department of Energy and other government agencies to provide assistance to INPO and the industry. An overall status of the nuclear utility industry responses to Kemeny Commission recommendations in the key areas directly related to nuclear plant operations is provided below. A more detailed status of industry responses to the
Report of nuclear utility industry responses to Kemeny Commission recommendations
NONE
1989-02-15
The purpose of this paper is to provide a report of nuclear utility industry progress in responding to the recommendations of the President's Commission on the Accident at Three Mile Island (The Kemeny Commission). On April 11, 1979, in response to TMI, President Carter established a Commission to conduct '.... a comprehensive study and investigation of the recent accident involving the nuclear power facility on Three Mile Island in Pennsylvania'. The Commission was chaired by Dr. John G. Kemeny, then President of Dartmouth College. (A list of all members of The Kemeny Commission is provided in Attachment to the Appendix ). The report of the commission's findings and recommendations was transmitted to the President in October 1979. During this same period, the nuclear utility industry responded to TMI by creating the Institute of Nuclear Power Operations (INPO) with a mission to promote the highest levels of safety and reliability - to promote excellence - in the operation of nuclear electric generating plants. In addition, the Nuclear Safety Analysis Center (NSAC) was established at the Electric Power Research Institute (EPRI to evaluate the accident and assist in determining the best industry response. In a White House paper (and press release) of December 7 1979, the President announced that he agreed fully with the spirit and intent of al the Kemeny Commission recommendations and requested that the industry and The Nuclear Regulatory Commission (NRC) comply with the recommendations. The President also recognized the industry initiative in establishing INPO and called for several actions involving the Institute; the President directed the Department of Energy and other government agencies to provide assistance to INPO and the industry. An overall status of the nuclear utility industry responses to Kemeny Commission recommendations in the key areas directly related to nuclear plant operations is provided below. A more detailed status of industry responses to the
The possibility of nuclear war: Appraisal, coping and emotional response
Kanofsky, S.
1989-01-01
This study used Lazarus and Folkman's (1984) model of appraisal and coping to explore people's emotional response to the possibility of nuclear war. Sixty-seven women and 49 men participated in a questionnaire study. The sample represented a cross-section of Americans by age and ethnic group but had more education and higher occupational status scores than is typical for the greater population. Sampling limitations and the political climate at the time of questionnaire administration suggested that the present findings be interpreted cautiously. Nevertheless, results suggested the importance of appraisal, defined in this study as the estimated probability of nuclear war and beliefs that citizen efforts to reduce the likelihood of nuclear war can be effective, and coping as factors in people's nuclear threat related emotional response. Six of the study's 11 hypotheses received at least partial confirmation. One or more measures of nuclear threat-related emotional distress were positively correlated with probability estimates of nuclear war, individual and collective response efficacy beliefs, and seeking social support in regard to the nuclear threat. Negative correlations were found between measures of threat-related distress and both trust in political leaders and distancing. Statistically significant relationships contrary to the other five hypotheses were also obtained. Measures of threat-related distress were positively, rather than negatively, correlated with escape avoidance and positive reappraisal coping efforts. Appraisal, coping, and emotion variables, acting together, predicted the extent of political activism regarding the nuclear arms race. It is useful to consider attitudes toward the nuclear arms race, distinguishing between intensity and frequency of emotional distress, and between measures of trait, state, and concept-specific emotionality in understanding emotional responses
Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro
2001-01-01
Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within ±10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the 92 Mo(n, 2n) 91g Mo reaction in FENDL, and lack of activation cross section data, e.g., the 138 Ba(n, 2n) 137m Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)
Maekawa, Fujio; Wada, Masayuki; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2001-01-01
Validity of decay heat calculations for safety designs of fusion reactors was investigated by using decay heat experimental data on thirty-two fusion reactor relevant materials obtained at the 14-MeV neutron source facility of FNS in JAERI. Calculation codes developed in Japan, ACT4 and CINAC version 4, and nuclear data bases such as JENDL/Act-96, FENDL/A-2.0 and Lib90 were used for the calculation. Although several corrections in algorithms for both the calculation codes were needed, it was shown by comparing calculated results with the experimental data that most of activation cross sections and decay data were adequate. In cases of type 316 stainless steel and copper which were important for ITER, prediction accuracy of decay heat within {+-}10% was confirmed. However, it was pointed out that there were some problems in parts of data such as improper activation cross sections, e,g., the {sup 92}Mo(n, 2n){sup 91g}Mo reaction in FENDL, and lack of activation cross section data, e.g., the {sup 138}Ba(n, 2n){sup 137m}Ba reaction in JENDL. Modifications of cross section data were recommended for 19 reactions in JENDL and FENDL. It was also pointed out that X-ray and conversion electron energies should be included in decay data. (author)
Talamo, Alberto; Gohar, Y.; Rabiti, C.; Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I.
2009-01-01
One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.
Talamo, Alberto [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: atalamo@anl.gov; Gohar, Y. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Rabiti, C. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83403 (United States); Aliberti, G.; Kondev, F.; Smith, D.; Zhong, Z. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.; Serafimovich, I. [Joint Institute for Power and Nuclear Research-Sosny, National Academy of Sciences (Belarus)
2009-07-21
One of the most reliable experimental methods for measuring the subcriticality level of a nuclear fuel assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology simulating the Sjoestrand method, which allows comparing the experimental and analytical reaction rates and the obtained subcriticality levels. In this methodology, the reaction rate is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the delayed fission neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction rate is vanished. The obtained reaction rate is then superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The analytical results of this new calculation methodology have shown an excellent agreement with the experimental data available from the YALINA-Booster facility of Belarus. This methodology can be used to calculate Bell and Glasstone spatial correction factor.
Crabol, B.; Romeo, E.; Nester, K.
1992-01-01
In case of an accident in a nuclear power plant near the French-German border different schemes for dispersion calculations in both countries will currently be applied. An intercomparison of these schemes initiated from the German-French Commission for the safety of nuclear installations (DFK) revealed in some meteorological situations large differences in the resulting concentrations for radionuclides. An ad hoc working group was installed by the DFK with the mandate to analyse the reasons for the different model results and also to consider new theoretical concepts. The working group has agreed to apply a Gaussian puff model for emergency response calculations. The results of the model based on turbulence parameterization via similarity approach or spectral theory - have been compared with tracer experiments for different emission heights and atmospheric stability regimes. As a reference the old modelling approaches have been included in the study. The simulations with the similarity approach and the spectral theory show a slightly better agreement to the measured concentration data than the schemes used in the past. Instead of diffusion categories both new approaches allow a continuous characterization of the atmospheric dispersion conditions. Because the spectral approach incorporates the sampling time of the meteorological data as an adjustable parameter thereby offering the possibility to adjust the dispersion model to different emission scenarios this turbulence parameterization scheme will be foreseen as the basis for a joint French-German puff model
Chang, Jong Hwa; Lee, Jeong Yeon; Lee, Young Ouk; Sukhovitski, Efrem Sh [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-01-01
Programs SHEMMAN and OPTMAN (Version 6) have been developed for determinations of nuclear Hamiltonian parameters and for optical model calculations, respectively. The optical model calculations by OPTMAN with coupling schemes built on wave functions functions of non-axial soft-rotator are self-consistent, since the parameters of the nuclear Hamiltonian are determined by adjusting the energies of collective levels to experimental values with SHEMMAN prior to the optical model calculation. The programs have been installed at Nuclear Data Evaluation Laboratory of KAERI. This report is intended as a brief manual of these codes. 43 refs., 9 figs., 1 tabs. (Author)
Cost calculations for decommissioning and dismantling of nuclear research facilities, Phase 1
Andersson, Inga; Backe, S.; Iversen, Klaus; Lindskog, S; Salmenhaara, S.; Sjoeblom, R.
2006-11-01
Today, it is recommended that planning of decommission should form an integral part of the activities over the life cycle of a nuclear facility. However, no actual international guideline on cost calculations exists at present. Intuitively, it might be tempting to regard costs for decommissioning of a nuclear facility as similar to those of any other plant. However, the presence of radionuclide contamination may imply that the cost is one or more orders of magnitude higher as compared to a corresponding inactive situation, the actual ratio being highly dependent on the level of contamination as well as design features and use of the facility in question. Moreover, the variations in such prerequisites are much larger than for nuclear power plants. This implies that cost calculations cannot be performed with any accuracy or credibility without a relatively detailed consideration of the radiological and other prerequisites. Application of inadequate methodologies especially at early stages has often lead to large underestimations. The goals of the project and the achievements described in the report are as follows: 1) Advice on good practice with regard to: 1a) Strategy and planning; 1b) Methodology selection; 1c) Radiological surveying; 1d) Uncertainty analysis; 2) Techniques for assessment of costs: 2a) Cost structuring; 2b) Cost estimation methodologies; 3) Compilation of data for plants, state of planning, organisations, etc.; 3a) General descriptions of relevant features of the nuclear research facilities; 3b) General plant specific data; 3c) Example of the decommissioning of the R1 research reactor in Sweden; 3d) Example of the decommissioning of the DR1 research reactor in Denmark. In addition, but not described in the present report, is the establishment of a Nordic network in the area including an internet based expert system. It should be noted that the project is planned to exist for at least three years and that the present report is an interim one
Cost calculations for decommissioning and dismantling of nuclear research facilities, Phase 1
Andersson, Inga [StudsvikNuclear AB (Sweden); Backe, S. [Institute for Energy Technology (Norway); Iversen, Klaus [Danish Decommissioning (Denmark); Lindskog, S [Swedish Nuclear Power Inspectorate (Sweden); Salmenhaara, S. [VTT Technical Research Centre of Finland (Finland); Sjoeblom, R. [Tekedo AB (Sweden)
2006-11-15
Today, it is recommended that planning of decommission should form an integral part of the activities over the life cycle of a nuclear facility. However, no actual international guideline on cost calculations exists at present. Intuitively, it might be tempting to regard costs for decommissioning of a nuclear facility as similar to those of any other plant. However, the presence of radionuclide contamination may imply that the cost is one or more orders of magnitude higher as compared to a corresponding inactive situation, the actual ratio being highly dependent on the level of contamination as well as design features and use of the facility in question. Moreover, the variations in such prerequisites are much larger than for nuclear power plants. This implies that cost calculations cannot be performed with any accuracy or credibility without a relatively detailed consideration of the radiological and other prerequisites. Application of inadequate methodologies especially at early stages has often lead to large underestimations. The goals of the project and the achievements described in the report are as follows: 1) Advice on good practice with regard to: 1a) Strategy and planning; 1b) Methodology selection; 1c) Radiological surveying; 1d) Uncertainty analysis; 2) Techniques for assessment of costs: 2a) Cost structuring; 2b) Cost estimation methodologies; 3) Compilation of data for plants, state of planning, organisations, etc.; 3a) General descriptions of relevant features of the nuclear research facilities; 3b) General plant specific data; 3c) Example of the decommissioning of the R1 research reactor in Sweden; 3d) Example of the decommissioning of the DR1 research reactor in Denmark. In addition, but not described in the present report, is the establishment of a Nordic network in the area including an internet based expert system. It should be noted that the project is planned to exist for at least three years and that the present report is an interim one
Technical/institutional prerequisite for nuclear forensics response framework
Tamai, Hiroshi; Okubo, Ayako; Kimura, Yoshiki; Kokaji, Lisa; Shinohara, Nobuo; Tomikawa, Hirofumi
2016-01-01
Nuclear Forensics capability has been developed under the international collaborations. For its effective function, technical development in analysis of seized nuclear materials as well as the institutional development in comprehensive response framework are required under individual national responsibility. In order to keep the “chain of custody” in the proper operation of sample collection at the event scene, radiological analysis at the laboratory, storage of the samples, and further inspection and trial, close cooperation and information sharing between relevant organisations are essential. IAEA issues the Implementing Guide to provide the model action plan and assists individual national development. Some countries at the advancing stage of national response framework, promote the international cooperation for the technical improvement and awareness cultivation. Examples in such national developments will be introduced and prospective technical/institutional prerequisite for nuclear forensics response framework will be studied. (author)
Microscopic equation of state calculations: 1. Nuclear matter. 2. Liquid helium 3
Heyer, J.P.
1989-01-01
A new method for calculating the equation of state of extended Fermi systems is proposed and applied to nuclear matter and liquid 3 He. New techniques are developed for summing up the particle-particle (pp) and particle-hole (ph) ring diagrams to all orders in the calculation of the ground state shift ΔE 0 for many-body systems. Analytic expressions for ΔE pp P 0 , the contribution from all of the pp ring diagrams to ΔE 0 , and ΔE ph 0 , the corresponding contribution from all of the ph ring diagrams, have been obtained. It has been shown that the pp ring diagram sum may be written as an integral over frequency, involving the particle-particle Green's function. A similar integral expression is derived for the ph ring diagram sum. Two methods are developed for carrying out the frequency integrations, namely the multipole and transition amplitude methods. These methods have been tested on an exactly-solvable many-fermion model, a modified Lipkin model, and compared. The author has studied the instability of nuclear matter at both zero and finite temperature within the pp ring diagram framework. He has found using the Gogny D1 effective nucleon-nucleon interaction, complex eigenvalues of an RPA-type secular equation are obtained in a well-defined temperature-density region. When complex eigenvalues occur, the thermodynamic potential becomes complex. The possible connection between the occurrence of complex eigenvalues and liquid-gas phase separation is discussed. The pp ring diagrams are also found to lower the compression modulus of nuclear matter. Lastly, the pp ring diagram method is applied to the calculation of the ground state energy of normal and spin-polarized liquid 3 He. We have found a binding energy per particle (BE/A) of 1.45 degree K and 1.79 degree K for the normal and spin-polarized systems, respectively
Calculation of shielding and radiation doses for PET/CT nuclear medicine facility
Mollah, A.S.; Muraduzzaman, S.M.
2011-01-01
Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)
Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional
2017-11-01
Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)
NUCADA - two adaptations of the system NUCORE for nuclear structure calculations
Heras, C.A.; Abecasis, S.M.
1983-01-01
Calculation of nuclear energy levels and their electromagnetic properties (transitions only between levels of the same parity). The nucleus is modelled as a cluster of a few particles and/or holes interacting with a core which in turn is either modelled as a quadrupole-octupole vibrator (cluster-phonon model) or of unspecified nature (cluster-core model). The members of the cluster interact via quadrupole-quadrupole and pairing forces in the first case, and via a delta force in the second. (orig.)
A simplified model for calculating early offsite consequences from nuclear reactor accidents
Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.
1988-07-01
A personal computer-based model, SMART, has been developed that uses an integral approach for calculating early offsite consequences from nuclear reactor accidents. The solution procedure uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast-running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarking and detailed sensitivity/uncertainty analyses using SMART are presented. 34 refs., 21 figs., 24 tabs
US/JAERI calculational benchmarks for nuclear data and codes intercomparison. Article 8
Youssef, M.Z.; Jung, J.; Sawan, M.E.; Nakagawa, M.; Mori, T.; Kosako, K.
1986-01-01
Prior to analyzing the integral experiments performed at the FNS facility at JAERI, both US and JAERI's analysts have agreed upon four calculational benchmark problems proposed by JAERI to intercompare results based on various codes and data base used independently by both countries. To compare codes the same data base is used (ENDF/B-IV). To compare nuclear data libraries, common codes were applied. Some of the benchmarks chosen were geometrically simple and consisted of a single material to clearly identify sources of discrepancies and thus help in analysing the integral experiments
Adjustement of Dancoff factor for calculating the cell of fluidized bed nuclear reactor
Borges, V.; Sefidvash, F.
1988-01-01
A new nuclear reactor design based on the fluidized bed concept is under reserch and development. It utilized spherical fuel of slightly enriched zircaloy-clad uranium dioxide fluidized by light water under pressure since the Leopard code has been developed for light water reactor analysis, it was necessary to develop a method to determine the dimensions of the hypothetical fuel rod lattice, which are neutronically equivalent to the spherical fuel pellet lattice. This method is shown to calculate the Dancoff factor correctly. (author) [pt
Application of ultrasonic inspection data in strength calculations for nuclear power plant equipment
Ovchinnikov, A.V.; Rivkin, E.Yu.; Vasilchenko, G.S.; Zvezdin, Yu.I.
1991-01-01
Several kinds of test specimens were produced with three types of defects of defined sizes and positions in the particular localities of weld joints. Such specimens have been used for defect parameter characterization by ultrasonic testing. The principles for schematization of such defects and the formulae for the stress intensity factor calculations for elliptical and semielliptical cracks have been worked out. Methods for defining the sizes of defect which are acceptable have been designed for use for use on operational nuclear power plant equipment and take account of the mutual effects of the force, thermal and residual stresses. The method can be used in the brittle, transitional and tough material state. (author)
CPU time reduction strategies for the Lambda modes calculation of a nuclear power reactor
Vidal, V.; Garayoa, J.; Hernandez, V. [Universidad Politecnica de Valencia (Spain). Dept. de Sistemas Informaticos y Computacion; Navarro, J.; Verdu, G.; Munoz-Cobo, J.L. [Universidad Politecnica de Valencia (Spain). Dept. de Ingenieria Quimica y Nuclear; Ginestar, D. [Universidad Politecnica de Valencia (Spain). Dept. de Matematica Aplicada
1997-12-01
In this paper, we present two strategies to reduce the CPU time spent in the lambda modes calculation for a realistic nuclear power reactor.The discretization of the multigroup neutron diffusion equation has been made using a nodal collocation method, solving the associated eigenvalue problem with two different techniques: the Subspace Iteration Method and Arnoldi`s Method. CPU time reduction is based on a coarse grain parallelization approach together with a multistep algorithm to initialize adequately the solution. (author). 9 refs., 6 tabs.
Approximate techniques for calculating gamma ray dose rates in nuclear power plants
Lahti, G.P.
1986-01-01
Although today's computers have made three-dimensional discrete ordinates transport codes a virtual reality, there is still a need for approximate techniques for estimating radiation environments. This paper discusses techniques for calculating gamma ray dose rates in nuclear power plants where Compton scattering is the dominant attenuation mechanism. The buildup factor method is reviewed; its use and misuse are discussed. Several useful rules-of-thumb are developed. The paper emphasizes the need for understanding the fundamental physics and draws heavily on the old, classic references
Calculation of the spin-isospin response functions in an extended semi-classical theory
Chanfray, G.
1987-01-01
We present a semi-classical calculation of the spin isospin response-functions beyond Thomas-Fermi theory. We show that surface-peaked ℎ 2 corrections reduce the collective effects predicted by Thomas-Fermi calculations. These effects, small for a volume response, become important for surface responses probed by hadrons. This yields a considerable improvement of the agreement with the (p, p') Los Alamos data
Improvements in the nuclear accident response system in Brazil
Estrada, J.J.S.; Azevedo, E.M.; Knofel, T.M.J.; Recio, J.C.A.; Alves, R.N.
1998-01-01
The National Commission on Nuclear Energy has been making outstanding effort to improve its nuclear and radiological accident response systems since the tragic accident in Goiania. Most of this effort is related to nuclear area although the radiological accident has been also considered. This paper describes the improvements in the CNEN response system structure, discusses several topics involving those related to emergency planning and preparedness, and points out some deficiencies that need to be corrected also. The situation during the Goiania accident was more disadvantageous than nowadays, so it is believed that none of the actual deficiencies are sufficient to guess that the population and the environment will not be protected in case of a nuclear or radiological accident
Nuclear response functions at large energy and momentum transfer
Bertozzi, W.; Moniz, E.J.; Lourie, R.W.
1991-01-01
Quasifree nucleon processes are expected to dominate the nuclear electromagnetic response function for large energy and momentum transfers, i.e., for energy transfers large compared with nuclear single particle energies and momentum transfers large compared with typical nuclear momenta. Despite the evident success of the quasifree picture in providing the basic frame work for discussing and understanding the large energy, large momentum nuclear response, the limits of this picture have also become quite clear. In this article a selected set of inclusive and coincidence data are presented in order to define the limits of the quasifree picture more quantitatively. Specific dynamical mechanisms thought to be important in going beyond the quasifree picture are discussed as well. 75 refs, 37 figs
Prevention and preparedness for response to nuclear and radiological threats
Pradeepkumar, K.S.
2016-01-01
Challenges from smuggled or illegally transported radioactive sources with malevolent intention of causing potential threats to the society are much higher to those potential radiological emergencies from misplaced, orphan or lost radioactive sources. Large number of radioactive sources world over is transported for its application in various fields. The emergency preparedness and response system is less developed for potential radiological emergencies caused by them compared to those at nuclear facilities which are kept in readiness to respond to any kind of emergency. After the terrorist attack on WTC of 2001, there is significant concern world over about the malicious use of nuclear and other radioactive material. This calls for prevention of stealing/smuggling of radioactive materials and improving the emergency response system. Use of Radiological Dispersal Device (RDD) and Improvised Nuclear Device (IND) are considered as possible radiological and nuclear threats, can lead to large area contamination in addition to the injuries caused by blast and thermal effects. (author)
Linear cascade calculations of matrix due to neutron-induced nuclear reactions
Avila, Ricardo E
2000-01-01
A method is developed to calculate the total number of displacements created by energetic particles resulting from neutron-induced nuclear reactions. The method is specifically conceived to calculate the damage in lithium ceramics by the 6L i(n, α)T reaction. The damage created by any particle is related to that caused by atoms from the matrix recoiling after collision with the primary particle. An integral equation for that self-damage is solved by interactions, using the magic stopping powers of Ziegler, Biersack and Littmark. A projectile-substrate dependent Kinchin-Pease model is proposed, giving and analytic approximation to the total damage as a function of the initial particle energy (au)
Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant
Lerner, Ana M.; Madariaga, Marcelo R.
1999-01-01
In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)
Nuclear criticality safety calculations for a K-25 site vacuum cleaner
Shor, J.T.; Haire, M.J.
1997-02-01
A modified Nilfisk model GSJ dry vacuum cleaner is used throughout the K-25 Site to collect dry forms of highly enriched uranium (HEU). When vacuuming, solids are collected in a cyclone-type separator vacuum cleaner body. Calculations were done with the SCALE (KENO V.a) computer code to establish conditions at which a nuclear criticality event might occur if the vacuum cleaner was filled with fissile solution. Conditions evaluated included full (12-in. water) reflection and nominal (1-in. water) reflection, and full (100%) and 20% 235 U enrichment. Validation analyses of SCALE/KENO and the SCALE 27-group cross sections for nuclear criticality safety applications indicate that a calculated k eff + 2σ eff + 2σ ≥ 0.9605 is considered unsafe and may be critical. Critical conditions were calculated to be 70 g U/L for 100% 235 U and full 12-in. water reflection. This corresponds to a minimum critical mass of approximately 1,400 g 235 U for the approximate 20.0-L volume of the vacuum cleaner. The actual volume of the vacuum cleaner is smaller than the modeled volume because some internal materials of construction were assumed to be fissile solution. The model was an overestimate, for conservatism, of fissile solution occupancy. At nominal reflection conditions, the critical concentration in a vacuum cleaner full of UO 2 F 2 solution was calculated to be 100 g 235 U/L, or 2,000 g mass of 100% 235 U. At 20% 235 U for the 20.0-L volume of the vacuum cleaner. At 15% 235 U enrichment and full reflection, critical conditions were not reached at any possible concentration of uranium as a uranyl fluoride solution. At 17.5% 235 U enrichment, criticality was reached at approximately 1,300 g U/L which is beyond saturation at 25 C
Meteorological considerations in emergency response capability at nuclear power plant
Fairobent, J.E.
1985-01-01
Meteorological considerations in emergency response at nuclear power plants are discussed through examination of current regulations and guidance documents, including discussion of the rationale for current regulatory requirements related to meteorological information for emergency response. Areas discussed include: major meteorological features important to emergency response; onsite meteorological measurements programs, including redundant and backup measurements; access to offsite sources of meteorological information; consideration of real-time and forecast conditions and atmospheric dispersion modeling
Verifying a nuclear weapon`s response to radiation environments
Dean, F.F.; Barrett, W.H.
1998-05-01
The process described in the paper is being applied as part of the design verification of a replacement component designed for a nuclear weapon currently in the active stockpile. This process is an adaptation of the process successfully used in nuclear weapon development programs. The verification process concentrates on evaluating system response to radiation environments, verifying system performance during and after exposure to radiation environments, and assessing system survivability.
Exercises for radiological and nuclear emergency response. Planing - performance - evaluation
Bayer, A.; Faleschini, J.; Goelling, K.; Stapel, R.; Strobl, C.
2010-01-01
The report of the study group emergency response seminar covers the following topics: (A) purpose of exercises and exercise culture: fundamentals and appliances for planning, performance and evaluation; (B) exercises in nuclear facilities; (C) exercises of national authorities and aid organizations on nuclear scenarios; exercises of national authorities and aid organizations on other radiological scenarios; (D) exercises in industrial plants, universities, medical facilities and medical services, and research institutes; (E) transnational exercises, international exercises; (F): exercises on public information.
How the Nuclear Applications Laboratories Help in Strengthening Emergency Response
2014-01-01
Safety is one of the most important considerations when engaging in highly advanced scientific and technological activities. In this respect, utilizing the potential of nuclear technology for peaceful purposes also involves risks, and nuclear techniques themselves can be useful in strengthening emergency response measures related to the use of nuclear technology. In the case of a nuclear incident, the rapid measurement and subsequent monitoring of radiation levels are top priorities as they help to determine the degree of risk faced by emergency responders and the general public. Instruments for the remote measurement of radioactivity are particularly important when there are potential health risks associated with entering areas with elevated radiation levels. The Nuclear Science and Instrumentation Laboratory (NSIL) — one of the eight laboratories of the Department of Nuclear Sciences and Applications (NA) in Seibersdorf, Austria — focuses on developing a variety of specialized analytical and diagnostic instruments and methods, and transferring knowledge to IAEA Member States. These include instruments capable of carrying out remote measurements. This emergency response work carried out by the NA laboratories supports health and safety in Member States and supports the IAEA’s mandate to promote the safe and peaceful use of nuclear energy
The risk of a major nuclear accident: calculation and perception of probabilities
Leveque, Francois
2013-07-01
The accident at Fukushima Daiichi, Japan, occurred on 11 March 2011. This nuclear disaster, the third on such a scale, left a lasting mark in the minds of hundreds of millions of people. Much as Three Mile Island or Chernobyl, yet another place will be permanently associated with a nuclear power plant which went out of control. Fukushima Daiichi revived the issue of the hazards of civil nuclear power, stirring up all the associated passion and emotion. The whole of this paper is devoted to the risk of a major nuclear accident. By this we mean a failure initiating core meltdown, a situation in which the fuel rods melt and mix with the metal in their cladding. Such accidents are classified as at least level 5 on the International Nuclear Event Scale. The Three Mile Island accident, which occurred in 1979 in the United States, reached this level of severity. The explosion of reactor 4 at the Chernobyl plant in Ukraine in 1986 and the recent accident in Japan were classified as class 7, the highest grade on this logarithmic scale. The main difference between the top two levels and level 5 relates to a significant or major release of radioactive material to the environment. In the event of a level-5 accident, damage is restricted to the inside of the plant, whereas, in the case of level-7 accidents, huge areas of land, above or below the surface, and/or sea may be contaminated. Before the meltdown of reactors 1, 2 and 3 at Fukushima Daiichi, eight major accidents affecting nuclear power plants had occurred worldwide. This is a high figure compared with the one calculated by the experts. Observations in the field do not appear to fit the results of the probabilistic models of nuclear accidents produced since the 1970's. Oddly enough the number of major accidents is closer to the risk as perceived by the general public. In general we tend to overestimate any risk relating to rare, fearsome accidents. What are we to make of this divergence? How are we to reconcile
Initial Human Response to Nuclear Radiation
1982-04-01
symptomatic response to radiation. In the second phase, the models will be used to infer performance effects. DNA staff members Cyrus Knowles and David ...P. Setty ATTN: K. Schwartz ATTN: J. NcGahan Kamn Tempo System Planning Corp ATTN: R. Miller ATTN: J. JonesATTN: G. Perks Kamen Tempo AiT: S. Shrier
Legal responsibility in case of a nuclear accident
Nabhane, M. F.
1988-01-01
Numerous laws have been elaborated in order to determine the legal responsibility in case of a nuclear accident. These laws were made necessary because of intervention of the factor 'error' in the nuclear accident. The legal definition of 'error' assumes that it results from non-respect or negligence of established norms on the part of the persons who manipulate the instruments of radioactive production. Nuclear research should not be undertaken in a country without the formal engagement of the central authorities to take the necessary dispositions to ensure the security and safety of the populations and their possessions. The world community should not admit a scientific activity in the nuclear field in the absence of guarantees for the safety and the security of man. The state that permits the production of nuclear energy is legally responsible for any failure that might result in radioactive spills. Considering the possibility of error and the dangers attached to the manipulation of radioactive material, the legislators have elaborated a series of laws, which take into consideration two principles: a)The inalienable right of man to life as conceived in the monotheistic religions and proclaimed by positive law; and b)The responsibility of the state for the safety and security of its citizens. Of course, error is human; but if man may make an error of judgement in ordinary normal life, he does not have the right to make the least miscalculation when this might lead to a nuclear disaster. (author)
Rosatom's Crisis Response Centre within the national nuclear safety system
Smirnov, S.N.; Komarovskij, A.V.; Moskalev, V.A.
2011-01-01
The Rosatom Corporation includes a number of subsidiaries associated with nuclear energy use as well as with the military, scientific, technological, nuclear and radiation safety management aspects. The Rosatom Corporation has a well-established and efficient industry-wide system of emergency prevention and response, whose purpose is to ensure safe functioning of the nuclear industry, protection of personnel, the public and nature from potential dangers; it is also a functional subsystem of the unified national system of emergency prevention and response. Overall management of the system is performed by Director General of the Rosatom Corporation, overall methodological management - by the Department of Licensing, Nuclear and Radiation Safety; everyday management of the emergency prevention and response system, round-the-clock monitoring and informational support - by the Rosatom Crisis and Response Centre (CRC). CRC acts as the national focal point for warning and communication in Russia, which provides continuous round-the-clock preparedness to cooperate with the IAEA's Incident and Emergency Centre using the formats of the ENATOM international emergency response system, similar national crisis response centres abroad [ru
Calculation Of Radon Gas Inhaled By A Front End Worker Of The Nuclear Fuel Cycles
Soedardjo
1996-01-01
The calculation of Radon gas inhaled by workers in a front end nuclear fuel cycle has been studied. The cycle of front end nuclear fuel is underground uranium exploration on Remaja tunnel West Kalimantan. The activities of mining consider of drilling, blasting, transporting mineral and tunnel supporting were chosen, It is assumed that in one month, for a worker has four assignments namely in tunnel I, tunnel II, tunnel III and tunnel IV, The activities in the mine are divided into some categories, namely 12 hours of drilling, 2 hours of after blasting, 9 hours of mineral transportation and 8 hours of tunnel support construction. The result of calculation shows that the average Radon gas concentration on each particular location is still less than maximum permissible concentration 300 pCi/l. The prediction of maximum dose inhaled by one uranium miner during a year is 2.3 x 10 2 μCi which is less than BATAN regulation 7.3 x 10 2 μCi
Calculated effects of backscattering on skin dosimetry for nuclear fuel fragments
Aydarous, A. Sh
2008-01-01
The size of hot particles contained in nuclear fallout ranges from 10 nm to 20 μm for the worldwide weapons fallout. Hot particles from nuclear power reactors can be significantly bigger (100 μm to several millimetres). Electron backscattering from such particles is a prominent secondary effect in beta dosimetry for radiological protection purposes, such as skin dosimetry. In this study, the effect of electron backscattering due to hot particles contamination on skin dose is investigated. These include parameters such as detector area, source radius, source energy, scattering material and source density. The Monte-Carlo Neutron Particle code (MCNP4C) was used to calculate the depth dose distribution for 10 different beta sources and various materials. The backscattering dose factors (BSDF) were then calculated. A significant dependence is shown for the BSDF magnitude upon detector area, source radius and scatterers. It is clearly shown that the BSDF increases with increasing detector area. For high Z scatterers, the BSDF can reach as high as 40 and 100% for sources with radii 0.1 and 0.0001 cm, respectively. The variation of BSDF with source radius, source energy and source density is discussed. (authors)
Evaluated Nuclear Data Library for Transport Calculations at Energies up to 150 MeV
Korovin, Yu.A.; Konobeyev, A.Yu.; Pilnov, G.B.; Stankovskiy, A.Yu.
2005-01-01
A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The first version of neutron sub-library has been completed and described in the present paper. The library contains nuclear data for transport, heating, and shielding applications for 242 nuclides ranging in atomic number from 8 to 82 in the energy region of primary neutrons from 10-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (Revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The evaluation of emitted particle energy and angular distributions at the energies above 20 MeV was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross sections, elastic cross sections, and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3m or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF/B-VI format using the MF=3/MT=5 and MF=6/MT=5 representations
Computerization of off-site dose calculations at two nuclear power plants
Lei, W.; Robertson, C.E.; Moore, G.T.; Rawls, B.E.; Sipp, J.R.
1988-01-01
The Brunswick Nuclear Project (BNP) consists of two boiling water reactors designed to generate a total net output of 1642 MWe. Unit 2 achieved commercial production in 1975, and Unit 1 began commercial operation in 1977. The Harris nuclear Project (HNP) is an 860 MWe pressurized water reactor that entered commercial operation in May of 1987. both plant sites are operated by Carolina Power and Light Company (CP and L). During January 1984, BNP replaced its older effluent technical specifications (part of the plant's original license) with the newer generation of Radiological Effluent Technical Specifications (RETS) mandated by the U.S. Nuclear Regulatory Commission. The RETS for HNP were integrated directly into this initial technical specifications. The initial version of the ODCM for BNP was drafted by a vendor and then extensively rewritten by the plant staff. The manual for HNP was drafted by CP and L corporate staff. At the outset, it was realized how impractical it would be to attempt to manually perform all of the dose calculations and keep the necessary records. The alternative-computerization-required extensive in-plant and corporate efforts to identify the computing resources (hardware) needed, create the software for ODCM implementation, and test, verify, validate, and document the software. This paper discusses these efforts
Miyasaka, Yasuhiko
1979-01-01
The fuel used for the research reactors of Japan Atomic Energy Research Institute (JAERI) is presently highly enriched uranium of 93%. However, the U.S. government (the supplier of fuel) is claiming to utilize low or medium enriched uranium from the viewpoint of resistivity to nuclear proliferation, and the availability of highly enriched uranium is becoming hard owing to the required procedure. This report is described on the results of nuclear calculation which is the basis of fuel design in the countermeasures to the reduction of enrichment. The basic conception in the reduction of enrichment is three-fold: to lower the latent potential of nuclear proliferation as far as possible, to hold the present reactor performance as far as possible, and to limit the reduction in the range which is not accompanied by the modification of reactor core construction and cooling system. This time, the increase of the density and thickness of fuel plates and the effect of enrichment change to 45% on reactivity and neutron flux were investigated. The fuel of UAl sub(x) - Al system was assumed, which was produced by powder metallurgical method. The results of investigations on JRR-2 and JMTR reactors revealed that 45% enriched fuel does not affect the performances much. However, deterioration of the performances is not neglegible if further reduction is needed. In future, the influence of the burn-up effect of fuel on the life of reactor cores must be investigated. (Wakatsuki, Y.)
Nuclear power and global warming: a first cost-benefit calculation
Hope, C.
1994-01-01
This paper investigates the costs and benefits of a modest nuclear power programme in the European Union to combat the threat of global warming. The nuclear programme is found to bring a double benefit. The first and more obvious benefit is that the economic impacts of global warming are reduced. The second benefit is counter-intuitive; most people would expect it to be a cost. It comes from the stimulus to the economy from the construction of the nuclear plant, which, with the recycling of carbon tax revenues, offsets its construction and operating costs, and may even cause consumers' expenditure to rise. Calculations in this paper show that over the period to 2100 the mean net present value of the first benefit is 6 billion European Currency Units (ECU; 1 ECU is about Dollars 1), while the second benefit has a mean net present value of 159 billion ECU. However both benefits, particularly the second, are still very uncertain, to the extent that even their sign is not yet definitely established. (author)
MONTE CARLO CALCULATION OF THE ENERGY RESPONSE OF THE NARF HURST-TYPE FAST- NEUTRON DOSIMETER
De Vries, T. W.
1963-06-15
The response function for the fast-neutron dosimeter was calculated by the Monte Carlo technique (Code K-52) and compared with a calculation based on the Bragg-Gray principle. The energy deposition spectra so obtained show that the response spectra become softer with increased incident neutron energy ahove 3 Mev. The K-52 calculated total res nu onse is more nearly constant with energy than the BraggGray response. The former increases 70 percent from 1 Mev to 14 Mev while the latter increases 135 percent over this energy range. (auth)
International responsibility of using nuclear energy for peaceful purposes
Ouenat, N.
2008-01-01
Although the stability of the idea of international responsibility in public international law, the international jurisprudence has not settled on a definition. The concept of international responsibility is no longer limited to the legal effects or consequences under international law to violate its provisions. The states recognized that the customary principles governing the international responsibility in public international law does not take into account the specificities of nuclear dangers, this sought to conclude a number of international conventions include a special system of nuclear liability not based on the wrongful act, but on the principle of keeping things, and it requires the existence of an international regime for nuclear liability in order to establish measures and procedures to achieve the implementation of the provisions for compensation unhindered by national legal systems. There is no doubt that the use of nuclear energy in time of peace falls within the scope of internationally prohibited acts. Atomic activities undertaken by the State within its borders for peaceful purposes are considered legitimate activities as long as they have taken necessary measures to avoid damage to neighboring countries. States has tended to conclude international agreements under which disputes that may result from the use of nuclear energy can be solved. The existing international legal framework on Civil Liability for Nuclear Damage consists of three major interrelated agreements: Paris Convention on civil liability in the field of nuclear energy, Vienna Convention on Civil Liability for civil damages and the Brussels Convention on Civil Liability in the Field of Maritime Carriage of Nuclear Materials.
Screening calculations for radioactive waste releases from non-nuclear facilities
Xu, Shulan; Soederman, Ann-Louis
2009-02-01
A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 μSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 μSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment plant
Optimized shielding calculation to the transport of 131I employed in nuclear medicine
Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.
1996-01-01
The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs
Mochamad Nasrullah; Sudi Arianto
2005-01-01
Nuclear power plant as one alternative power plant for Indonesia is expected to attract interest of investors to invest in electricity sector. Calculation of investment cost and electricity tariff is a nearly necessary Information needed by investors. Spread sheet calculations on construction cost including Interest During Construction and escalation as well as financial viability are implemented. Result of the study show that overnight cost before escalation is US $ 2.682.865.200,- and after IDC and escalation it becomes US $ 3.795.712.088 or 1.807,5 US$/k We. Levelized Tariff is at around 4,57 cents/kWh. Levelized Tariff is 3,5 cents/kWh not feasible to the project of because all financial parameter show negative value. The project is financially feasible if calculated levelized tariff within arrange of 4,0 cents/kWh-5,5 cents/kWh. The most profitable tariff for investor is within arrange of 4,87 cents/kWh - 5,11 cents/kWh. (author)
Younker, J.L.; Wilson, W.E.; Sinnock, S.
1986-01-01
In support of the US Department of Energy Nevada Nuclear Waste Storage Investigations Project, ground-water travel times were calculated for flow paths in both the saturated and unsaturated zones at Yucca Mountain, a potential site for a high-level radioactive waste repository in southern Nevada. The calculations were made through a combined effort by Science Applications International Corporation, Sandia National Laboratories, and the US Geological Survey. Travel times in the unsaturated zone were estimated by dividing the flow path length by the ground-water velocity, where velocities were obtained by dividing the vertical flux by the effective porosity of the rock types along assumed vertical flow paths. Saturated zone velocities were obtained by dividing the product of the bulk hydraulic conductivity and hydraulic gradient by the effective porosity. Total travel time over an EPA-established 5-km flow path was then calculated to be the sum of the travel times in the two parts of the flow path. Estimates of ground water fluxes and travel times are critical for evaluating the favorability of the Yucca Mountain site because they provide the basis for estimating the potential for radionuclides to reach the accessible environment within certain time limits
Comparison between different computational schemes for variational calculations in nuclear structure
Puddu, G.
2009-01-01
We compare several iteration methods for angular-momentum- and parity-projected Hartree-Fock calculations. We used the Anderson update, the modified Broyden method, newly introduced in nuclear-structure calculations, and variants of the Broyden-Fletcher-Goldhaber-Shanno methods (BFGS). We performed ground-state calculations for 18 C and 6 Li using the two-body Hamiltonian obtained from the CDBonn-2000 potential via the Lee-Suzuki renormalization method. We found that BFGS methods are superior to both the Anderson update and to the modified Broyden method. In the case of 6 Li we found that the Anderson update and modified Broyden method do not converge to the angular-momentum- and parity-projected Hartree-Fock minimum. The reason is traced back to the lack of a mechanism that guarantees a decrease of the energy from one iteration to the next and to the fact that these methods guarantee a stationary solution rather than a minimum of the energy. (orig.)
Calculation of radiation dose rates from a spent nuclear fuel shipping cask
Chen, S.Y.; Yuan, Y.C.
1988-01-01
Radiation doses from a spent nuclear fuel cask are usually from various phases of operations during handling, shipping, and storage of the casks. Assessment of such doses requires knowledge of external radiation dose rates at various locations surrounding a cask. Under current practices, dose rates from gamma photons are usually estimated by means of point- or line-source approaches incorporating the conventional buildup factors. Although such simplified approaches may at times be easy to use, their accuracy has not been verified. For example, those simplified methods have not taken into account influencing factors such as the geometry of the cask and the presence of the ground surface, and the effects of these factors on the calculated dose rates are largely unknown. Moreover, similar empirical equations for buildup factors currently do not exist for neutrons. The objective of this study is to use a more accurate approach in calculating radiation dose rates for both neutrons and gamma photons from a spent fuel cask. The calculation utilizes the more sophisticated transport method and takes into account the geometry of the cask and the presence of the ground surface. The results of a detailed study of dose rates in the near field (within 20 meters) are presented and, for easy application, the cask centerline dose rates are fitted into empirical equations at cask centerline distances up to 2000 meters from the surface of the cask
Water hammer calculation and analysis in main feedwater system of PWR nuclear power plants
Wang Xin; Han Weishi
2010-01-01
The main feedwater system of a nuclear power plant is an important part in ensuring the cooling of the steam generator. Moreover, it is the main pipe section where water hammers frequently occur. Studying the regular patterns of water hammers to the main feedwater system is significant to the stable operation of the system. The paper focuses on the study of water hammers through Flowmaster's transient calculating function to establish a mathematical model with boundary conditions such as a feedwater pump, control valves, etc.; calculation of the water hammers pressure when feedwater pumps and control valves shut down; exporting the instantaneous change in solution of pressure. Combined with engineering practical examples, the conclusions verify the viability of calculating the water hammers pressure through Flowmaster's transient function, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively, changing the intervals of closing signals to feedwater pumps and control valves to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (authors)
Sato, Hajime
2011-01-01
Severe accidents at nuclear plants can result in long-standing and large-scale disasters encompassing wide areas. The public may have special concerns regarding these plants and radiation-related health risks. It has therefore been argued that risk communications efforts, along with rigid safety management of nuclear plants, are imperative to prevent such accidents, mitigate their impacts, and alleviate public concerns. This article introduces a set of laws, acts, codes, and guidelines concerning nuclear safety in Japan. In addition, the preparedness and mitigation plans and programs for dealing with nuclear accidents and possible disasters are also discussed. Furthermore, the ongoing accidents at the Fukushima nuclear power plants following the Great East Japan Earthquake in 2011, and the government response to them are presented. A set of points regarding the management and communications of power plant accidents are discussed. (author)
Response of high Tc superconducting Josephson junction to nuclear radiation
Ding Honglin; Zhang Wanchang; Zhang Xiufeng
1992-10-01
The development of nuclear radiation detectors and research on high T c superconducting nuclear radiation detectors are introduced. The emphases are the principle of using thin-film and thick-film Josephson junctions (bridge junction) based on high T c YBCO superconductors to detect nuclear radiation, the fabrication of thin film and thick-film Josephson junction, and response of junction to low energy gamma-rays of 59.5 keV emitted from 241 Am and beta-rays of 546 keV. The results show that a detector for measuring nuclear radiation spectrum made of high T c superconducting thin-film or thick-film, especially, thick-film Josephson junction, certainly can be developed
Calculation of the process of vacuum drying of a metal-concrete container with spent nuclear fuel
Karyakin, Yu. E.; Lavrent'ev, S. A.; Pavlyukevich, N. V.; Pletnev, A. A.; Fedorovich, E. D.
2012-01-01
An algorithm and results of calculation of the process of vacuum drying of a metal-concrete container intended for long-term "dry" storage of spent nuclear fuel are presented. A calculated substantiation of the initial amount of moisture in the container is given.
2013-01-01
Following the March 2011 accident at the Fukushima Daiichi nuclear power plant, all NEA member countries took early action to ensure and confirm the continued safety of their nuclear power plants and the protection of the public. After these preliminary safety reviews, all countries with nuclear facilities carried out comprehensive safety reviews, often referred to as 'stress tests', which reassessed safety margins of nuclear facilities with a primary focus on challenges related to conditions experienced at the Fukushima Daiichi nuclear power plant, for example extreme external events and the loss of safety functions, or capabilities to cope with severe accidents. As appropriate, improvements are being made to safety and emergency response systems to ensure that nuclear power plants are capable of withstanding events that lead to loss of electrical power and/or cooling capability. In the weeks following the accident, the NEA immediately began establishing expert groups in the nuclear safety and radiological protection areas, as well as contributing to information exchange with the Japanese authorities and other international organisations. It promptly provided a forum for high-level decision makers and regulators within the G8-G20 frameworks. The NEA actions taken at the international level in response to the accident have been carried out primarily by the three NEA standing technical committees concerned with nuclear and radiation safety issues - the Committee on Nuclear Regulatory Activities (CNRA), the Committee on the Safety of Nuclear Installations (CSNI) and the Committee on Radiation Protection and Public Health (CRPPH) - under the leadership of the CNRA. More than two years following the accident, the NEA continues to assist the Japanese authorities in dealing with their nuclear safety and recovery efforts as well as to facilitate international co-operation on nuclear safety and radiological protection matters. It is strongly supporting the establishment of
Nuclear response beyond mean field theory
Brand, M.G.E.; Allaart, K.; Dickhoff, W.H.
1990-01-01
An extension of the RPA equations is derived, with emphasis on the relation between the single-particle Green function and the polarization propagator. Including second order self-energy contributions the resulting particle-hole interaction includes the coupling to two-particle-two-hole (2p2h) states and the resulting response satisfies relevant conservation laws. This aspect of the theory is shown to be essential to obtain reliable and meaningful results for excitation strengths and to avoid ghost solutions. This method is applied to electromagnetic and charge exchange excitations in 48 Ca up to 100 MeV. A G-matrix interaction based on meson exchange is used which takes care of short-range correlations. The results compare favourably with measured excitation strengths and electromagnetic form factors both at low energy as well as in the giant resonance region. Remaining discrepancies point in the direction of further strength reduction due to short-range correlations as well as a possible stronger coupling to 2p2h states at low energy. (orig.)
Responsibility, fairness and credibility - ethical dilemmas concerning nuclear waste
Nilsson, Annika
1999-01-01
This book is written with the aim to stimulate discussion and studies on ethical aspects of nuclear waste management. It does not penetrate the technical details of the different stages of waste handling, but concentrates on questions such as our responsibility towards future generations etc
Two-body tensor interactions in the nuclear matter response function
Besprosvany, J.
1997-01-01
The inclusive scattering response of nuclear matter is studied in the regime of large momentum transfer q, and around the quasielastic peak. We review interaction corrections to free propagation as embodied in the impulse approximation. Calculations of the two-body and many-body corrections within an eikonal approach are presented. These use an approximated two-body density matrix which takes account of spin and isospin degrees of freedom. Both calculations give similar and sizable corrections at q = 550 MeV and reproduce data extrapolated from finite nuclei; this indicates the relevance of two-body tensor contributions in this regime. (Author)
Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers
Cho, Byungoh.
1990-01-01
Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function
Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y.
2006-01-01
The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)
Dose calculation for accident situations at WWR-S type spent nuclear fuel repository
Margeanu, S.; Florescu, G.
2006-01-01
Full text: The Spent Nuclear Fuel Repository at IFIN-HH Bucharest (SNFR IFIN-HH) consists in four pools, repository hall, radiological monitoring system, ventilation system and auxiliary systems. At the moment the remaining activity in the repository is about 3500 Ci. Despite of the small activity, for emergency preparedness purposes, several accident scenarios, with a non zero probability of occurrence during the repository lifetime, have been postulated. Evaluations of radiological consequences to personnel, general public and environment, for each accident scenario have been performed. The radioactive inventory was evaluated with ORIGEN code from SCALE computer code system and radiological consequences were evaluated with COSYMA computer code. Assumptions for the source term determination, meteorological conditions and release, are presented. The calculated values of doses and risk are also presented. The impact of these accident scenarios on population and environment is also discussed. (authors)
Garcia, A.E.; Parkansky, D.G.
1993-01-01
In the Embalse nuclear power plant (CNE), the Regional Overpower Protection System acting on the Shutdown Systems number 1 and number 2 protects the reactor against overpowers in the reactor field for a localized peaking or a power increase in the reactor as a whole. This report summarizes the results of the critical channel power calculation for the time average powers configuration for the 380 reactor field channels. The final purpose of this work is to analyze and eventually modify the detector set points. Other reactor configurations are being analyzed. The report also presents a sensitivity analysis in order to evaluate potential sources of error and uncertainties which could affect the ROP performance. (author)
Garg, S.B.
1991-01-01
A detailed investigation is carried out to determine the effect of different level density prescriptions on the computed neutron nuclear data of Ni-58 in the energy range 5-25 MeV. Calculations are performed in the framework of the multistep Hauser-Feshbach statistical theory including the Kalbach exciton model and Brink-Axel giant dipole resonance model for radiative capture. Level density prescriptions considered in this investigation are based on the original Gilbert-Cameron, improved Gilbert-Cameron, backshifted Fermi-gas and the Ignatyuk, et al. approaches. The effect of these prescriptions is discussed, with special reference to (n,p), (n,2n), (n,alpha) and total particle-production cross sections. (author). 17 refs, 8 figs
Sher, R. [Rudolph Sher Associates, Stanford, CA (United States); Li, J. [Polestar Applied Technology, Inc., Los Altos, CA (United States)
1995-02-01
NAUAHYGROS is a computer code to calculate the behavior of fission product and other aerosol particles in the containment of a nuclear reactor following a severe accident. It is an extension of the German code NAUA, which has been in widespread use for many years. Early versions of NAUA treated various aerosol phenomena in dry atmospheres, including aerosol agglomeration, diffusion (plateout), and settling processes. Later versions added treatments of steam condensation on particles in saturated or supersaturated containment atmospheres. The importance of these condensation effects on aerosol removal rates was demonstrated in large scale simulated containment tests. The additional features incorporated in NAUAHYGROS include principally a treatment of steam condensation on hygroscopic aerosols, which can grow as a result of steam condensation even in superheated atmospheres, and improved modelling of steam condensation on the walls of the containment. The code has been validated against the LACE experiments.
Numerical calculation of the tensor of diffusion in the nuclear reactor cells by Monte-Carlo method
Gorodkov, S.S.; Kalugin, M.A.
2009-01-01
New algorithm based on the sequential application of the RMS path method has been proposed for the diffusion constants calculation. The offered algorithm conforms to the diffusion constants calculation in arbitrary segments of nuclear reactors without detail description of geometry, dependence of cross-sections from energy or neutron scattering anisotropy by kernel medium. The proposed algorithm is used for the diffusion constants calculation in uranium-graphite reactor sells
Lyckman, C
1996-01-01
The report accounts for results from calculations on the content of radionuclides in nuclear fuel waste. It also accounts for the results from calculations on the neutron flow from spent fuel, which is very important during transports. The calculations have been performed using the ORIGEN2 software. The results have been compared to other results from earlier versions of ORIGEN and some differences have been discovered. This is due to the updating of the software. 7 refs, 10 figs, 15 tabs.
ZZ NUCDECAYCALC, Nuclear Decay Data for Radiation Dosimetry Calculation for ICRP
2002-01-01
1 - Description or function: The Dosimetry Research Group (DRG) of the Health Sciences Research Division at ORNL has for several years maintained data bases of nuclear decay data for use in dosimetric calculations. The data on mean and unique energy plus intensity have been previously published, in abridged form, in Publication 38 of the International Commission on Radiological Protection (ICRP 1983). This data base was designed to address the needs in medical, environmental, and occupational radiation protection. DLC-172/NUCDECAY is required by the CCC-620/SEECAL program to calculate age-dependent specific effective energies. 2 - Methods: The unabridged data used in preparing ICRP Publication 38 are distributed in electronic form in this package. The collection consists of data on the energies and intensities of radiations emitted by the 825 radionuclides reported, although abridged, in ICRP Publication 38 plus an additional 13 radionuclides evaluated during preparation of a monograph for the Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine. Each collection is contained in an ASCII file (INDEXR.DAT) which is a sorted list of the radionuclides containing the decay chain information. The utility code DecayCalc extracts the decay data from the library for radionuclide(s) specified by the user. It computes the activities of radionuclides present after decay and ingrowth over a user-specified time period from 1 minute to 50 years. Decay data for any decay chain may be displayed and printed either in tabular form or graphically. DecayCalc, in a slightly modified version, will be a part of CCC-553/Rascal v3. DecayCalc is a Windows application that runs under Microsoft Windows 95 or 98, or Microsoft Windows NT 4.0 or later. The Compac Fortran 77 compiler was used to compile the code. The full source for DecayCalc is not provided but will be distributed when Rascal V3 is released
Input/Output of ab-initio nuclear structure calculations for improved performance and portability
Laghave, Nikhil
2010-01-01
Many modern scientific applications rely on highly computation intensive calculations. However, most applications do not concentrate as much on the role that input/output operations can play for improved performance and portability. Parallelizing input/output operations of large files can significantly improve the performance of parallel applications where sequential I/O is a bottleneck. A proper choice of I/O library also offers a scope for making input/output operations portable across different architectures. Thus, use of parallel I/O libraries for organizing I/O of large data files offers great scope in improving performance and portability of applications. In particular, sequential I/O has been identified as a bottleneck for the highly scalable MFDn (Many Fermion Dynamics for nuclear structure) code performing ab-initio nuclear structure calculations. We develop interfaces and parallel I/O procedures to use a well-known parallel I/O library in MFDn. As a result, we gain efficient I/O of large datasets along with their portability and ease of use in the down-stream processing. Even situations where the amount of data to be written is not huge, proper use of input/output operations can boost the performance of scientific applications. Application checkpointing offers enormous performance improvement and flexibility by doing a negligible amount of I/O to disk. Checkpointing saves and resumes application state in such a manner that in most cases the application is unaware that there has been an interruption to its execution. This helps in saving large amount of work that has been previously done and continue application execution. This small amount of I/O provides substantial time saving by offering restart/resume capability to applications. The need for checkpointing in optimization code NEWUOA has been identified and checkpoint/restart capability has been implemented in NEWUOA by using simple file I/O.
Medeiros, Marcos P.C.; Rebello, Wilson F.; Andrade, Edson R.; Silva, Ademir X.
2015-01-01
Nuclear explosions are usually described in terms of its total yield and associated shock wave, thermal radiation and nuclear radiation effects. The nuclear radiation produced in such events has several components, consisting mainly of alpha and beta particles, neutrinos, X-rays, neutrons and gamma rays. For practical purposes, the radiation from a nuclear explosion is divided into i nitial nuclear radiation , referring to what is issued within one minute after the detonation, and 'residual nuclear radiation' covering everything else. The initial nuclear radiation can also be split between 'instantaneous or 'prompt' radiation, which involves neutrons and gamma rays from fission and from interactions between neutrons and nuclei of surrounding materials, and 'delayed' radiation, comprising emissions from the decay of fission products and from interactions of neutrons with nuclei of the air. This work aims at presenting isodose curves calculations at ground level by Monte Carlo simulation, allowing risk assessment and consequences modeling in radiation protection context. The isodose curves are related to neutrons produced by the prompt nuclear radiation from a hypothetical nuclear explosion with a total yield of 20 KT. Neutron fluency and emission spectrum were based on data available in the literature. Doses were calculated in the form of ambient dose equivalent due to neutrons H*(10) n - . (author)
Medeiros, Marcos P.C.; Rebello, Wilson F.; Andrade, Edson R., E-mail: rebello@ime.eb.br, E-mail: daltongirao@yahoo.com.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Secao de Engenharia Nuclear; Silva, Ademir X., E-mail: ademir@nuclear.ufrj.br [Corrdenacao dos Programas de Pos-Graduacao em Egenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2015-07-01
Nuclear explosions are usually described in terms of its total yield and associated shock wave, thermal radiation and nuclear radiation effects. The nuclear radiation produced in such events has several components, consisting mainly of alpha and beta particles, neutrinos, X-rays, neutrons and gamma rays. For practical purposes, the radiation from a nuclear explosion is divided into {sup i}nitial nuclear radiation{sup ,} referring to what is issued within one minute after the detonation, and 'residual nuclear radiation' covering everything else. The initial nuclear radiation can also be split between 'instantaneous or 'prompt' radiation, which involves neutrons and gamma rays from fission and from interactions between neutrons and nuclei of surrounding materials, and 'delayed' radiation, comprising emissions from the decay of fission products and from interactions of neutrons with nuclei of the air. This work aims at presenting isodose curves calculations at ground level by Monte Carlo simulation, allowing risk assessment and consequences modeling in radiation protection context. The isodose curves are related to neutrons produced by the prompt nuclear radiation from a hypothetical nuclear explosion with a total yield of 20 KT. Neutron fluency and emission spectrum were based on data available in the literature. Doses were calculated in the form of ambient dose equivalent due to neutrons H*(10){sub n}{sup -}. (author)
Calculation of financial compensation due of municipalities hosting nuclear waste deposit
Silva, Renata A. da; Simoes, Francisco Fernando L.; Martins, Vivian B.
2011-01-01
The present work evaluates the math from monthly financial transfers to municipalities with technical viability for building of initial or intermediate repository for storing of radioactivity nuclear waste: gloves, sneakers, mask, resins and filters came from thermonuclear facilities. Several aspects have been considered as the geological factors of the site as presence of capable faults, groundwater vulnerability, infiltration of seawater. Also, it was take into account socioeconomic factors: population density, costs for construction, maintenance and operation of repository; size and activity of waste; among others. Hereafter, we have presented the key features of low and average activity repository and high activity repository even as initial, intermediate and final repository and the possible environment impact. The methodology for calculation of financial compensation of municipalities was established by CNEN will be applied for a specific assumed municipality. The analysis of financial compensation due to the specific nuclear waste deposit and the possible guidelines for the use of that compensation by the municipality will be analyzed. In addiction, it will be compared the model for compensation used for nuclear wastes with other plants receiving permanent wastes from cemeteries and sanitary landfills, where the land should not be allowed for the human activities the same as: crops, livestock and buildings. Also, comparison with royalties and indemnities were paid by facilities of energy production as hydroelectric dams as well as petroleum and gas exploration plants. The destination of financial compensation transfer to the municipality is in charge of the city administration. The compensation could be applied of investments in education and culture, health, sanitation works, improvement of public transport, environment, among others. It will be discussed the cost-benefit relation for the assumed municipality. (author)
Calculation of financial compensation due of municipalities hosting nuclear waste deposit
Silva, Renata A. da, E-mail: renata.amaral@ufrj.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Simoes, Francisco Fernando L.; Martins, Vivian B., E-mail: flamego@ien.gov.b [Instituto de Engenharia Nuclear (LIMA/IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. Impactos Ambientais
2011-07-01
The present work evaluates the math from monthly financial transfers to municipalities with technical viability for building of initial or intermediate repository for storing of radioactivity nuclear waste: gloves, sneakers, mask, resins and filters came from thermonuclear facilities. Several aspects have been considered as the geological factors of the site as presence of capable faults, groundwater vulnerability, infiltration of seawater. Also, it was take into account socioeconomic factors: population density, costs for construction, maintenance and operation of repository; size and activity of waste; among others. Hereafter, we have presented the key features of low and average activity repository and high activity repository even as initial, intermediate and final repository and the possible environment impact. The methodology for calculation of financial compensation of municipalities was established by CNEN will be applied for a specific assumed municipality. The analysis of financial compensation due to the specific nuclear waste deposit and the possible guidelines for the use of that compensation by the municipality will be analyzed. In addiction, it will be compared the model for compensation used for nuclear wastes with other plants receiving permanent wastes from cemeteries and sanitary landfills, where the land should not be allowed for the human activities the same as: crops, livestock and buildings. Also, comparison with royalties and indemnities were paid by facilities of energy production as hydroelectric dams as well as petroleum and gas exploration plants. The destination of financial compensation transfer to the municipality is in charge of the city administration. The compensation could be applied of investments in education and culture, health, sanitation works, improvement of public transport, environment, among others. It will be discussed the cost-benefit relation for the assumed municipality. (author)
Model calculating annual mean atmospheric dispersion factor for coastal site of nuclear power plant
无
2001-01-01
This paper describes an atmospheric dispersion field experiment performed on the coastal site of nuclear power plant in the east part of China during 1995 to 1996. The three-dimension joint frequency are obtained by hourly observation of wind and temperature on a 100m high tower; the frequency of the “event day of land and sea breezes” are given by observation of surface wind and land and sea breezes; the diffusion parameters are got from measurements of turbulent and wind tunnel simulation test.A new model calculating the annual mean atmospheric dispersion factor for coastal site of nuclear power plant is developed and established.This model considers not only the effect from mixing release and mixed layer but also the effect from the internal boundary layer and variation of diffusion parameters due to the distance from coast.The comparison between results obtained by the new model and current model shows that the ratio of annual mean atmospheric dispersion factor gained by the new model and the current one is about 2.0.
Shao, Meiyue; Aktulga, H. Metin; Yang, Chao; Ng, Esmond G.; Maris, Pieter; Vary, James P.
2018-01-01
We describe a number of recently developed techniques for improving the performance of large-scale nuclear configuration interaction calculations on high performance parallel computers. We show the benefit of using a preconditioned block iterative method to replace the Lanczos algorithm that has traditionally been used to perform this type of computation. The rapid convergence of the block iterative method is achieved by a proper choice of starting guesses of the eigenvectors and the construction of an effective preconditioner. These acceleration techniques take advantage of special structure of the nuclear configuration interaction problem which we discuss in detail. The use of a block method also allows us to improve the concurrency of the computation, and take advantage of the memory hierarchy of modern microprocessors to increase the arithmetic intensity of the computation relative to data movement. We also discuss the implementation details that are critical to achieving high performance on massively parallel multi-core supercomputers, and demonstrate that the new block iterative solver is two to three times faster than the Lanczos based algorithm for problems of moderate sizes on a Cray XC30 system.
Liu Yang; Cai Qi; Lin Xiaoling
2011-01-01
Based on the operation characteristics of the nuclear power unit, the radioactive inventory of activated parts was calculated by ORIGEN2, and the effects of bum-up, operation mode and power change on the radioactive inventory for activated parts were analyzed. The results indicated that the radioactive inventory grew with the increasing of burn-up, and when the actual operation time was longer than the effective operation time, the increasing rate of nuclide activity approximated the burn-up increasing; Radioactive inventory of activated parts was influenced directly by the operation modes of the nuclear power unit, and under same reactor load, operation power and bum-up, the radioactive inventory for non-continuous operation mode is less than that for the continuous operation mode. Effects of operation modes on radioactive inventory reversed with half life of nuclides. Under same bum-up and longer operation time, the effect of operation power change on the radioactive inventory is not obvious, (authors)
Nuclear vorticity and the low-energy nuclear response. Towards the neutron drip line
Papakonstantinou, P.; Athens Univ.; Wambach, J.; Ponomarev, V.Y.; Mavrommatis, E.
2004-01-01
The transition density and current provide valuable insight into the nature of nuclear vibrations. Nuclear vorticity is a quantity related to the transverse transition current. In this work, we study the evolution of the strength distribution, related to density fluctuations, and the vorticity strength distribution, as the neutron drip line is approached. Our results on the isoscalar, natural-parity multipole response of Ni isotopes, obtained by using a self-consistent Skyrme-Hartree-Fock+continuum RPA model, indicate that, close to the drip line, the low-energy response is dominated by L > 1 vortical transitions. (orig.)
Silva, M
2006-01-01
The Atucha I Nuclear Power Plant (CNA-I) has enough room to store its spent fuel (SF) in damp in its two pool houses until the middle of 2015.Before that date there is the need to have an interim dry storage system for spent fuel that would make possible to empty at least one of the pools, whether to keep the plant operating if its useful life is extended, or to be able to empty the reactor core in case of decommissioning.Nucleolectrica Argentina S.A. (NA-SA) and the Comision Nacional de Energia Atomica (CNEA), due to their joint responsibility in the management of the SF, have proposed interim dry storage systems.These systems have to be evaluated in order to choose one of them by the end of 2006.In this work the Monte Carlo code MCNP was used to make the criticality and shielding calculations corresponding to the model proposed by CNEA.This model suggests the store of sealed containers with 36 or 37 SF in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.The results of the criticality calculations indicates that the solutions of SF proposed have widely fulfilled the requirements of subcriticality, even in supposed extreme accidental situations.Regarding the transference cask, the SF dose rate estimations allow us to make a feedback for the design aiming to the geometry and shielding improvements.Regarding the store modules, thicknesses ranges of concrete walls are suggested in order to fulfill the dose requirements stated by the Autoridad Regulatoria Nuclear Argentina [es
The responsibility of nuclear suppliers in the international market
Thiessen, C.W.; Cotton, J.D.; Tait, B.I.
1978-01-01
As a result of escalating oil prices, the developing countries are more and more turning to nuclear energy. As these countries become a more significant part of the international market, unique issues must be faced by the utility and country concerned and the nuclear supplier. A fundamental prerequisite of a successful supplier/customer relationship is a good relationship between their two governments. The supplier can fulfil an effective liaison function to ensure efficiency of communications in this area. In the preplanning phase the supplier can play a major role in supplying information relative to siting, unit size, system planning and plant economics. Once the nuclear choice is made, the pros and cons of purchasing plants on a turnkey versus non-turnkey basis must be weighed. The customer must balance his available resources and experience against his desires for acquisition of overall plant and systems knowledge. During the planning and procurement phase the subject of training, maintenance and service facilities can be an effective part of a comprehensive nuclear programme. In fact, local training and service centres could logically be the first step of a long-term localization effort. With respect to localization the supplier should be capable and experienced in developing this capability and willing to make a long-term commitment in this area. Through a clearer understanding of the responsibilities of the nuclear supplier and a clear delineation of his role, developing nations can more adequately assess the basis upon which to move forward on a nuclear programme. (author)
Preparedness for and response to a radiological or nuclear incident
Norman Coleman, C.
2014-01-01
Public health and medical planning for a nuclear or radiological incident requires a complex, multi-faceted systematic approach involving federal, state and local governments, private sector organizations, academia, industry, international partners and individual experts and volunteers. The approach developed by the U.S. Department of Health and Human Services in collaboration with other U.S. Departments is the result of efforts from government and non-government experts that connect the available capabilities, resources, guidance tools, underlying concepts and science into the Nuclear Incident Medical Enterprise (NlME). It is a systems approach that can be used to support planning for, response to, and recovery from the effects of a nuclear incident. Experience is gained in exercises specific to radiation but also from other mass casualty incidents as there are many principles and components in common. Resilience and the ability to mitigate the consequences of a nuclear incident are enhanced by effective planning, preparation and training, timely response, clear communication, and continuous improvements based on new science, technology, experience and ideas. Recognizing that preparation for a radiological or nuclear incident will be a lower priority for healthcare workers and responders due to other demands, the Radiation Emergency Medical Management website has been developed with the National Library of Medicine. This includes tools for education and training, just-in-time medical management and triage among others. Most of the components of NIME are published in the peer review medical and disaster medicine literature to help ensure high quality and accessibility. While NIME is a continuous work-in-progress, the current status of the public health and medical preparedness and response for a nuclear incident is presented. (author)
Fukushima nuclear accident and the social responsibility of science
Yoshioka, Hitoshi
2011-01-01
Five months had passed since Fukushima Daiichi Nuclear Power Plant (NPP) accident occurred but still there was no knowing when the accident ended. Released radioactivity seemed to be greater than one million terra Bq and if there occurred an explosive rupture of containment vessel due to the failure of containment vent or occurrence of steam explosion, the amount of released radioactivity might amount to be at least equivalent to or surpass that of Chernobyl NPP accident. There existed still a risk that overheating and meltdown of nuclear fuels might reoccur with loss of cooling due to a possible giant aftershock. This article described total views on significant disaster that the accident brought about on many neighboring residents or wide range of people. After a general discussion about what was social responsibility of scientists, social responsibility of scientists for Fukushima Daiichi NPP accident was discussed. Responsibility of omission was also argued. (T. Tanaka)
Evaluating nuclear power plant crew performance during emergency response drills
Rabin, D.
1999-01-01
The Atomic Energy Control Board (AECB) is responsible for the regulation of the health, safety and environmental consequences of nuclear activities in Canada. Recently, the Human Factors Specialists of the AECB have become involved in the assessment of emergency preparedness and emergency response at nuclear facilities. One key contribution to existing AECB methodology is the introduction of Behaviourally Anchored Rating Scales (BARS) to measure crew interaction skills during emergency response drills. This report presents results of an on-going pilot study to determine if the BARS provide a reliable and valid means of rating the key dimensions of communications, openness, task coordination and adaptability under simulated emergency circumstances. To date, the objectivity of the BARS is supported by good inter-rater reliability while the validity of the BARS is supported by the agreement between ratings of crew interaction and qualitative and quantitative observations of crew performance. (author)
Calculation of Excore Detector Responses upon Control Rods Movement in PGSFR
Ha, Pham Nhu Viet; Lee, Min Jae; Kang, Chang Moo; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-05-15
The Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) safety design concept, which aims at achieving IAEA's safety objectives and GIF's safety goals for Generation-IV reactor systems, is mainly focused on the defense in depth for accident detection, prevention, control, mitigation and termination. In practice, excore neutron detectors are widely used to determine the spatial power distribution and power level in a nuclear reactor core. Based on the excore detector signals, the reactor control and protection systems infer the corresponding core power and then provide appropriate actions for safe and reliable reactor operation. To this end, robust reactor power monitoring, control and core protection systems are indispensable to prevent accidents and reduce its detrimental effect should one occur. To design such power monitoring and control systems, numerical investigation of excore neutron detector responses upon various changes in the core power level/distribution and reactor conditions is required in advance. In this study, numerical analysis of excore neutron detector responses (DRs) upon control rods (CRs) movement in PGSFR was carried out. The objective is to examine the sensitivity of excore neutron detectors to the core power change induced by moving CRs and thereby recommend appropriate locations to locate excore neutron detectors for the designing process of the PGSFR power monitoring systems. Section 2 describes the PGSFR core model and calculation method as well as the numerical results for the excore detector spatial weighting functions, core power changes and detector responses upon various scenarios of moving CRs in PGSFR. The top detector is conservatively safe because it overestimated the core power level. However, the lower and bottom detectors still functioned well in this case because they exhibited a minor underestimation of core power of less than ∼0.5%. As a secondary CR was dropped into the core, the lower detector was
Calculating the sensitivity of wind turbine loads to wind inputs using response surfaces
Rinker, Jennifer M.
2016-01-01
at a low computational cost. Sobol sensitivity indices (SIs) can then be calculated with relative ease using the calibrated response surface. The proposed methodology is demonstrated by calculating the total sensitivity of the maximum blade root bending moment of the WindPACT 5 MW reference model to four......This paper presents a methodology to calculate wind turbine load sensitivities to turbulence parameters through the use of response surfaces. A response surface is a high-dimensional polynomial surface that can be calibrated to any set of input/output data and then used to generate synthetic data...... turbulence input parameters: a reference mean wind speed, a reference turbulence intensity, the Kaimal length scale, and a novel parameter reflecting the nonstationarity present in the inflow turbulence. The input/output data used to calibrate the response surface were generated for a previous project...
Monte Carlo code Serpent calculation of the parameters of the stationary nuclear fission wave
V. M. Khotyayintsev
2017-12-01
Full Text Available n this work, propagation of the stationary nuclear fission wave was simulated for series of fixed power values using Monte Carlo code Serpent. The wave moved in the axial direction in 5 m long cylindrical core of fast reactor with pure 238U raw fuel. Stationary wave mode arises some period later after the wave ignition and lasts sufficiently long to determine kef with high enough accuracy. The velocity characteristic of the reactor was determined as the dependence of the wave velocity on the neutron multiplication factor. As we have recently shown within a one-group diffusion description, the velocity characteristic is two-valued due to the effect of concentration mechanisms, while thermal feedback affects it only quantitatively. The shape and parameters of the velocity characteristic critically affect feasibility of the reactor design since stationary wave solutions of the lower branch are unstable and do not correspond to any real waves in self-regulated reactor, like CANDLE. In this work calculations were performed without taking into account thermal feedback. They confirm that theoretical dependence correctly describes the shape of the velocity characteristic calculated using the results of the Serpent modeling.
On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies
Mikhin, V.I.; Zhukov, A.V.
1985-01-01
One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements
Benchmarking of MCNP for calculating dose rates at an interim storage facility for nuclear waste.
Heuel-Fabianek, Burkhard; Hille, Ralf
2005-01-01
During the operation of research facilities at Research Centre Jülich, Germany, nuclear waste is stored in drums and other vessels in an interim storage building on-site, which has a concrete shielding at the side walls. Owing to the lack of a well-defined source, measured gamma spectra were unfolded to determine the photon flux on the surface of the containers. The dose rate simulation, including the effects of skyshine, using the Monte Carlo transport code MCNP is compared with the measured dosimetric data at some locations in the vicinity of the interim storage building. The MCNP data for direct radiation confirm the data calculated using a point-kernel method. However, a comparison of the modelled dose rates for direct radiation and skyshine with the measured data demonstrate the need for a more precise definition of the source. Both the measured and the modelled dose rates verified the fact that the legal limits (<1 mSv a(-1)) are met in the area outside the perimeter fence of the storage building to which members of the public have access. Using container surface data (gamma spectra) to define the source may be a useful tool for practical calculations and additionally for benchmarking of computer codes if the discussed critical aspects with respect to the source can be addressed adequately.
Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.
1995-01-01
During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences
Jancauskas, J.R.
1994-01-01
The purpose of degrading voltage relays (DVRs) is to ensure that adequate voltage is available to operate all Class 1E loads at all voltage distribution levels. Should voltage drop below the setpoint of the DVRs, the Class 1E power system is disconnected from its supply and resequenced onto the diesel generators in order to restore system voltages to acceptable levels. These relays represent one of the two levels of voltage protection required for the onsite power system. Determining the proper setpoint for degraded voltage relays in nuclear generating stations is a complex task which requires a complete understanding of the Class 1E power distribution system. Despite the importance of degraded voltage relay setpoint calculations, most of the available references only give clues on how not to set these relays rather than provide guidance on how to determine the appropriate setpoint. This paper presents an approach for performing these calculations which attempts to ensure that all of the relevant design issues are addressed
Evaluation of response factors for seismic probabilistic safety assessment of nuclear power plants
Ebisawa, K.; Abe, K.; Muramatsu, K.; Itoh, M.; Kohno, K.; Tanaka, T.
1994-01-01
This paper presents a method for evaluating 'response factors' of components in nuclear power plants for use in a seismic probabilistic safety assessment (PSA). The response factor here is a measure of conservatism included in response calculations in seismic design analysis of components and is defined as a ratio of conservative design resonse to actual response. This method has the following characteristic features: (1) The components are classified into several groups based on the differences in their location and in the vibration models used in design response analyses; (2) the response factors are decomposed into subfactors corresponding to the stages of the seismic response analyses in the design practices; (3) the response factors for components are calculated as products of subfactors; (4) the subfactors are expressed either as a single value or as a function of parameters that influence the response of components. This paper describes the outline of this method and results from an application to a sample problem in which response factors were quantified for examples of components selected from the groups. (orig.)
Seismic response analysis for a deeply embedded nuclear power plant
Chen, W.W.H.; Chatterjee, M.; Day, S.M.
1979-01-01
One of the important aspect of the aseimic design of nuclear power plants is the evaluation of the seismic soil-structure interaction effect due to design earthquakes. The soil-structure interaction effect can initiate rocking and result in different soil motions compared to the free field motions, thus significantly affecting the structural response. Two methods are generally used to solve the seismic soil-structure interaction problems: the direct finite element method (FLUSH) and the substructure or impedance approach. This paper presents the results of the horizontal seismic soil-structure interaction analysis using the impedance aproach and the direct finite element method for a deeply embedded nuclear power plant. (orig.)
Limits on the experimental simulation of nuclear fuel rod response
Hagar, R.C.
1980-01-01
The steady-state and transient effects of intrinxic geometric and material property differences between typical nuclear fuel pins and electric fuel pin simulators (FPSs) are identified. The effectiveness of varying the transient power supplied to the FPS in reducing the differences between the transient responses of nuclear fuel pins and FPSs is investigated. This effectiveness is shown to be limited by the heat capacity of the FPS, the allowed range of the power program, and different FPS power requirements at different positions on a full-length FPS
Calculating the sensitivity of wind turbine loads to wind inputs using response surfaces
Rinker, Jennifer M.
2016-01-01
This paper presents a methodology to calculate wind turbine load sensitivities to turbulence parameters through the use of response surfaces. A response surface is a highdimensional polynomial surface that can be calibrated to any set of input/output data and then used to generate synthetic data at a low computational cost. Sobol sensitivity indices (SIs) can then be calculated with relative ease using the calibrated response surface. The proposed methodology is demonstrated by calculating the total sensitivity of the maximum blade root bending moment of the WindPACT 5 MW reference model to four turbulence input parameters: a reference mean wind speed, a reference turbulence intensity, the Kaimal length scale, and a novel parameter reflecting the nonstationarity present in the inflow turbulence. The input/output data used to calibrate the response surface were generated for a previous project. The fit of the calibrated response surface is evaluated in terms of error between the model and the training data and in terms of the convergence. The Sobol SIs are calculated using the calibrated response surface, and the convergence is examined. The Sobol SIs reveal that, of the four turbulence parameters examined in this paper, the variance caused by the Kaimal length scale and nonstationarity parameter are negligible. Thus, the findings in this paper represent the first systematic evidence that stochastic wind turbine load response statistics can be modeled purely by mean wind wind speed and turbulence intensity. (paper)
Screening calculations for radioactive waste releases from non-nuclear facilities
Shulan Xu; Soederman, Ann-Louis
2009-02-15
A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 muSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 muSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment
In-medium no-core shell model for ab initio nuclear structure calculations
Gebrerufael, Eskendr
2017-01-01
In this work, we merge two successful ab initio nuclear-structure methods, the no-core shell model (NCSM) and the multi-reference in-medium similarity renormalization group (IM-SRG), to define a novel many-body approach for the comprehensive description of ground and excited states of closed- and open-shell medium-mass nuclei. Building on the key advantages of the two methods - the decoupling of excitations at the many-body level in the IM-SRG, and the exact diagonalization in the NCSM applicable up to medium-light nuclei - their combination enables fully converged no-core calculations for an unprecedented range of nuclei and observables at moderate computational cost. The efficiency and rapid model-space convergence of the new approach make it ideally suited for ab initio studies of ground and low-lying excited states of nuclei up to the medium-mass regime. Interactions constructed within the framework of chiral effective field theory provide an excellent opportunity to describe properties of nuclei from first principles, i.e., rooted in quantum chromodynamics, they overcome the lack of predictive power of phenomenological potentials. The hard core of these interactions causes strong short-range correlations, which we soften by using the similarity-renormalization-group transformation that accelerates the model-space convergence of many-body calculations. Three-nucleon effects, which are mandatory for the correct description of bulk properties of nuclei, are included in our calculations by using the normal-ordered two-body approximation, which has been shown to be sufficient to capture the main effects of the three-nucleon interaction. Using these interactions, we analyze energies of ground and excited states in the carbon and oxygen isotopic chains, where conventional NCSM calculations are still feasible and provide an important benchmark. Furthermore, we study the Hoyle state in 12 C - a three-alpha cluster state that cannot be converged in standard NCSM
Calculation of Upper Subcritical Limits for Nuclear Criticality in a Repository
J.W. Pegram
1998-01-01
The purpose of this document is to present the methodology to be used for development of the Subcritical Limit (SL) for post closure conditions for the Yucca Mountain repository. The SL is a value based on a set of benchmark criticality multiplier, k eff results that are outputs of the MCNP calculation method. This SL accounts for calculational biases and associated uncertainties resulting from the use of MCNP as the method of assessing k eff . The context for an SL estimate include the range of applicability (based on the set of MCNP results) and the type of SL required for the application at hand. This document will include illustrative calculations for each of three approaches. The data sets used for the example calculations are identified in Section 5.1. These represent three waste categories, and SLs for each of these sets of experiments will be computed in this document. Future MCNP data sets will be analyzed using the methods discussed here. The treatment of the biases evaluated on sets of k eff results via MCNP is statistical in nature. This document does not address additional non-statistical contributions to the bias margin, acknowledging that regulatory requirements may impose additional administrative penalties. Potentially, there are other biases or margins that should be accounted for when assessing criticality (k eff ). Only aspects of the bias as determined using the stated assumptions and benchmark critical data sets will be included in the methods and sample calculations in this document. The set of benchmark experiments used in the validation of the computational system should be representative of the composition, configuration, and nuclear characteristics for the application at hand. In this work, a range of critical experiments will be the basis of establishing the SL for three categories of waste types that will be in the repository. The ultimate purpose of this document is to present methods that will effectively characterize the MCNP
Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations
Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.
1999-01-01
Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k eff of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data
Canning, Andrew
2013-03-01
Inorganic scintillation phosphors (scintillators) are extensively employed as radiation detector materials in many fields of applied and fundamental research such as medical imaging, high energy physics, astrophysics, oil exploration and nuclear materials detection for homeland security and other applications. The ideal scintillator for gamma ray detection must have exceptional performance in terms of stopping power, luminosity, proportionality, speed, and cost. Recently, trivalent lanthanide dopants such as Ce and Eu have received greater attention for fast and bright scintillators as the optical 5d to 4f transition is relatively fast. However, crystal growth and production costs remain challenging for these new materials so there is still a need for new higher performing scintillators that meet the needs of the different application areas. First principles calculations can provide a useful insight into the chemical and electronic properties of such materials and hence can aid in the search for better new scintillators. In the past there has been little first-principles work done on scintillator materials in part because it means modeling f electrons in lanthanides as well as complex excited state and scattering processes. In this talk I will give an overview of the scintillation process and show how first-principles calculations can be applied to such systems to gain a better understanding of the physics involved. I will also present work on a high-throughput first principles approach to select new scintillator materials for fabrication as well as present more detailed calculations to study trapping process etc. that can limit their brightness. This work in collaboration with experimental groups has lead to the discovery of some new bright scintillators. Work supported by the U.S. Department of Homeland Security and carried out under U.S. Department of Energy Contract no. DE-AC02-05CH11231 at Lawrence Berkeley National Laboratory.
Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium
Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R.; Sajo B, L.
2015-10-01
This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of 252 Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of 252 Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ( 7 LiF - BeF 2 - ThF 4 - UF 4 ) as fuel. For design the k eff , neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of 233 U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k eff of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)
Two-phase flow and thermal response from nuclear excursions in tuff
Rath, J.S.; Sanchez, L.C.; Taylor, L.L.
1998-05-01
Thermal hydrology calculations were performed to predict the geologic thermal and saturation response of a far-field nuclear criticality. The thermal hydrology (THX) calculations used an experimental version of a transient multi-phase fluid and energy simulator, BRAGFLO T. A total of 45 THX calculations were completed using various combinations of initial saturation S 0 , input heat generation zone (HGZ) radii r 0 , input energies E 0 , and input space power density functions (SPDFs). The thermal hydrology calculations were performed as a part the nuclear dynamics consequence analysis (NDCA) study for potential criticality consequences associated with disposal of high-level waste (HLW) and spent nuclear fuel (SNF) in an underground geologic repository. In the NDCA study it was identified that total fission energy E 0 , integrated from the power-time history, has an expected range of 10 17 --10 20 total fissions per excursion. This range of values is comparable to those reported for aqueous criticality accidents that had occurred in processing plants. The THX results show (using the conservative temperature recycle times) that a criticality frequency between 3 and 30 criticalities/yr is possible. Probability frequencies (generated by probabilistic risk analysis and the THX model) for these consequences indicate that any additional fissions are minor contributions to the biological hazards caused by the disposed fissile materials
Effects of different SSI parameters on the floor response spectra of a nuclear reactor building
Kabir, A.F.; Bolourchi, S.; Maryak, M.E.
1991-01-01
The effects of several critical soil-structure interaction (SSI) parameters on the floor response spectra (FRS) of a typical nuclear reactor building have been examined. These parameters are computation of soil impedance functions using different approaches, scattering effects (reductions in ground motion due to embedment and rigidity of building foundation) and strain dependency of soil dynamic properties. This paper reports that the significant conclusions of the study, which are applicable to a deeply embedded very rigid nuclear reactor building, are as follows: FRS generated without considering scattering effects are highly conservative; differences between FRS, generated considering strain-dependency of soil dynamic properties, and those generated suing low-strain values, are not significant; and the lumped-parameter approach of SSI calculations, which only uses a single value of soil shear modulus in impedance calculations, may not be able to properly compute the soil impedances for a soil deposit with irregularly varying properties with depth
Nuclear energy and the responsibilities of the Atomic Energy Board
De Villiers, J.W.L.
1980-01-01
The paper discusses nuclear energy and the responsibilities of the previous Atomic Energy Board, (now the Atomic Energy Corporation) of South Africa in this respect. The paper starts by giving a brief introduction to the Atomic Energy Board, its organization and its functions. Research is undertaken in various fields such as the exploitation of nuclear fuels, radiobiology, radioisotopes, etc. Certain activities of the Board was also more directly related to Koeberg. The paper covers four of these areas, namely the early studies of the feasibility of introducing nuclear power in South Africa; the services involving the Board's special expertise in certain areas which Escom makes use of; the regulatory function and the preparation for handling and disposal of radioactive waste
Response trees and expert systems for nuclear reactor operations
Nelson, W.R.
1984-02-01
The United States Nuclear Regulatory Commission is sponsoring a project performed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) to evaluate different display concepts for use in nuclear reactor control rooms. Included in this project is the evaluation of the response tree computer based decision aid and its associated displays. This report serves as an overview of the response tree methodology and how it has been implemented as a computer based decision aid utilizing color graphic displays. A qualitative assessment of the applicability of the response tree aid in the reactor control room is also made. Experience gained in evaluating the response tree aid is generalized to address a larger category of computer aids, those known as knowledge based expert systems. General characteristics of expert systems are discussed, as well as examples of their application in other domains. A survey of ongoing work on expert systems in the nuclear industry is presented, and an assessment of their potential applicability is made. Finally, recommendations for the design and evaluation of computer based decision aids are presented
Oblozinsky, P.
1997-09-01
The report contains the summary of the third and the last Research Co-ordination Meeting on ''Development of Reference Input Parameter Library for Nuclear Model Calculations of Nuclear Data (Phase I: Starter File)'', held at the ICTP, Trieste, Italy, from 26 to 29 May 1997. Details are given on the status of the Handbook and the Starter File - two major results of the project. (author)
Bedrosian, B.; Barbela, M.; Drenick, R.F.; Tsirk, A.
1980-10-01
The critical excitation method provides a new, alternative approach to methods presently used for seismic analysis of nuclear power plant structures. The critical excitation method offers the advantages that: (1) it side-steps the assumptions regarding the probability distribution of ground motions, and (2) it does not require an artificial, and to some extent arbitrarily generated, time history of ground motion, both features to which structural integrity analyses are sensitive. Potential utility of the critical excitation method is studied from the user's viewpoint. The method is reviewed and compared with the response spectrum method used in current practice, utilizing the reactor buildings of a PWR and a BWR plant in case studies. Two types of constraints on critical excitation were considered in the study. In one case, only an intensity limit was used. In the other case, imposition of an intensity limit together with limits on the maximum acceleration and/or velocity for the critical excitation is considered
Sakenas, C.A.; McKenna, T.J.; Perkins, K.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.
1987-02-01
This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. US Nuclear Regulatory Commission Response is the fifth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes NRC response modes, organizations, and official positions; roles of other federal agencies are also described briefly. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text
The calculated neutron response of a sphere with the multi-counters
Li Taosheng; Yang Lianzhen; Li Dongyu
2004-01-01
Based on the difference of the neutron distribution in the moderator, three position sensitive proportional counters which are perpendicular to each other are inserted into the moderator. The energy responses with six spherical moderators and six incidence directions have been calculated by MCNP4A code. The calculated results for two divided region methods in the radial of the spherical moderator have been analyzed and compared. (authors)
Dynamical response of the nuclear 'pasta' in neutron star crusts
Horowitz, C.J.; Perez-Garcia, M.A.; Berry, D.K.; Piekarewicz, J.
2005-01-01
The nuclear pasta - a novel state of matter having nucleons arranged in a variety of complex shapes - is expected to be found in the crust of neutron stars and in core-collapse supernovae at subnuclear densities of about 10 14 g/cm 3 . Owing to frustration, a phenomenon that emerges from the competition between short-range nuclear attraction and long-range Coulomb repulsion, the nuclear pasta displays a preponderance of unique low-energy excitations. These excitations could have a strong impact on many transport properties, such as neutrino propagation through stellar environments. The excitation spectrum of the nuclear pasta is computed via a molecular-dynamics simulation involving up to 100,000 nucleons. The dynamic response of the pasta displays a classical plasma oscillation in the 1- to 2-MeV region. In addition, substantial strength is found at low energies. Yet this low-energy strength is missing from a simple ion model containing a single-representative heavy nucleus. The low-energy strength observed in the dynamic response of the pasta is likely to be a density wave involving the internal degrees of freedom of the clusters
Kim, Kyu Tae; Kim, Oh Hwan
1999-01-01
A simplified statistical methodology is developed in order to both reduce over-conservatism of deterministic methodologies employed for PWR fuel rod internal pressure (RIP) calculation and simplify the complicated calculation procedure of the widely used statistical methodology which employs the response surface method and Monte Carlo simulation. The simplified statistical methodology employs the system moment method with a deterministic statistical methodology employs the system moment method with a deterministic approach in determining the maximum variance of RIP. The maximum RIP variance is determined with the square sum of each maximum value of a mean RIP value times a RIP sensitivity factor for all input variables considered. This approach makes this simplified statistical methodology much more efficient in the routine reload core design analysis since it eliminates the numerous calculations required for the power history-dependent RIP variance determination. This simplified statistical methodology is shown to be more conservative in generating RIP distribution than the widely used statistical methodology. Comparison of the significances of each input variable to RIP indicates that fission gas release model is the most significant input variable. (author). 11 refs., 6 figs., 2 tabs
Hutton, T.; Sublet, J.C.; Morgan, L.; Leadbeater, T.W.
2015-01-01
Highlights: • We quantify the effect of processing nuclear data from ENDF to ACE format. • We consider the differences between fission and fusion angular distributions. • C-nat(n,el) at 2.0 MeV has a 0.6% deviation between original and processed data. • Fe-56(n,el) at 14.1 MeV has a 11.0% deviation between original and processed data. • Processed data do not accurately depict ENDF distributions for fusion energies. - Abstract: Nuclear data form the basis of the radiation transport codes used to design and simulate the behaviour of nuclear facilities, such as the ITER and DEMO fusion reactors. Typically these data and codes are biased towards fission and high-energy physics applications yet are still applied to fusion problems. With increasing interest in fusion applications, the lack of fusion specific codes and relevant data libraries is becoming increasingly apparent. Industry standard radiation transport codes require pre-processing of the evaluated data libraries prior to use in simulation. Historically these methods focus on speed of simulation at the cost of accurate data representation. For legacy applications this has not been a major concern, but current fusion needs differ significantly. Pre-processing reconstructs the differential and double differential interaction cross sections with a coarse binned structure, or more recently as a tabulated cumulative distribution function. This work looks at the validity of applying these processing methods to data used in fusion specific calculations in comparison to fission. The relative effects of applying this pre-processing mechanism, to both fission and fusion relevant reaction channels are demonstrated, and as such the poor representation of these distributions for the fusion energy regime. For the nat C(n,el) reaction at 2.0 MeV, the binned differential cross section deviates from the original data by 0.6% on average. For the 56 Fe(n,el) reaction at 14.1 MeV, the deviation increases to 11.0%. We
Massicano, Felipe
2010-01-01
The nuclear medicine area has an increasing slope in the therapy of diseases, particularly in the treatment of radiosensitive tumors. Due to the high dose levels in radionuclide therapy, it is very important the accurate quantify of the dose distribution to avoid deleterious effects on healthy tissues. In Brazil, the internal dosimetry system used is the MIRD (Medical Internal Radiation Dose) based on a reference model that does not have adequate patient data to obtain a dose accurate assessment in therapy. However, in recent years, internal radionuclide dosimetry evaluates the spatial dose distribution base ad on information obtained from CT and SPECT or PET images together with the using of Monte Carlo codes. Those systems are called patient-specific dosimetry systems. In the Nuclear Engineering Center at IPEN, this methodology is in development. When the CT images are inserted into the Monte Carlo code MCNP5 through of use of a interface software called SCMS the dosimetry can be accomplished using patient-specific data, resulting in a more accurate energy deposition in organs of interest. This work aim to contribute with the development of part of that patient-specific dosimetry for therapy. To achieve this goal we have proposed three specific objectives: (1) Development of a software to convert images from Computed Tomography (CT) in the tissue parameters (ρ, ω(ι)); (2) Development of a software to perform attenuation correction in nuclear medicine tomographic images (SPECT or PET) and to provide the map of relative activity and (3) Provide data to the SCMS code by these two software. The software developed for the rst specific objective was the Image Converter Computed Tomography (ICCT), which obtained a good accuracy to determine the density and the tissue composition; the elements that had high variation were carbon and oxygen. Fortunately, this variation for the energy range used in radionuclide therapy is not detrimental to the dose distribution. A
Linear response calculation using the canonical-basis TDHFB with a schematic pairing functional
Ebata, Shuichiro; Nakatsukasa, Takashi; Yabana, Kazuhiro
2011-01-01
A canonical-basis formulation of the time-dependent Hartree-Fock-Bogoliubov (TDHFB) theory is obtained with an approximation that the pair potential is assumed to be diagonal in the time-dependent canonical basis. The canonical-basis formulation significantly reduces the computational cost. We apply the method to linear-response calculations for even-even nuclei. E1 strength distributions for proton-rich Mg isotopes are systematically calculated. The calculation suggests strong Landau damping of giant dipole resonance for drip-line nuclei.
Morrone, A.
1979-01-01
A particular type of seismic analysis performed on the Nuclear Island Buildings (NIB) complex of a nuclear power plant and the methods developed to combine torsional, rotational and translational responses are described. The NIB complex analyzed consists of various buildings supported on a common foundation mat and tied together from the underground foundation to the roof levels. Three independent building mathematical models were used for the three components of the earthquake with a lumped-mass method utilizing direct integration of the coupled equations of motion. The input ground acceleration time histories were based on three 20 s long statistically independent records whose normalized response spectra enveloped those of Regulatory Guide 1.60. A linear stochastic model was used to generate these records which simulated strong motion earthquakes. Due to site characteristics, the soil material properties were calculated considering different ranges of soil moduli below and above the foundation. (orig.)
Morrone, A.; Sigal, G.B.
1979-01-01
A particular type of seismic analysis performed on the Nuclear Island Buildings (NIB) complex of a nuclear power plant and the methods developed to combine torsional, rotational and translational responses are described. The NIB complex analyzed consists of various buildings supported on a common foundation mat and tied together from the underground foundation to the roof levels. Three independent building mathematical models were used for the three components of the earthquake with a lumped-mass method utilizing direct integration of the coupled equations of motion. The input ground acceleration time histories were based on three 20 s long statistically independent records whose mormalized response spectra enveloped those of Regulatory Guide 1.60. A linear stochastic model was used to generate these records which simulated strong motion earthquakes. Due to site characteristics, the soil material properties were calculated considering different ranges of soil moduli below and above the foundation
Japanese Nuclear Accident and U.S. Response
Douet, Randy
2011-01-01
U.S. Government response to the Fukushima accident: • Multi-agency task force (Nuclear Regulatory Commission, Department of Energy, Department of Defense) supporting Japan recovery efforts; • President Obama directed the NRC to perform a comprehensive review of U.S. reactors; • NRC established agency task force to develop lessons learned from Fukushima Daiichi accident to provide short-term and long-term analysis of the events
Calculation and applications of the frequency dependent neutron detector response functions
Van Dam, H.; Van Hagen, T.H.J.J. der; Hoogenboom, J.E.; Keijzer, J.
1994-01-01
The theoretical basis is presented for the evaluation of the frequency dependent function that enables to calculate the response of a neutron detector to parametric fluctuations ('noise') or oscillations in reactor core. This function describes the 'field view' of a detector and can be calculated with a static transport code under certain conditions which are discussed. Two applications are presented: the response of an ex-core detector to void fraction fluctuations in a BWR and of both in and ex-core detectors to a rotating neutron absorber near or inside a research reactor core. (authors). 7 refs., 4 figs
10 CFR 1.46 - Office of Nuclear Security and Incident Response.
2010-01-01
... 10 Energy 1 2010-01-01 2010-01-01 false Office of Nuclear Security and Incident Response. 1.46... Headquarters Program Offices § 1.46 Office of Nuclear Security and Incident Response. The Office of Nuclear... evaluation and assessment of technical issues involving security at nuclear facilities, and is the agency...
Ohoka, Yasunori; Tatsumi, Masahiro; Sugimura, Naoki; Tabuchi, Masato
2011-01-01
In 2010, the latest version of the Japanese Evaluated Nuclear Data Library (JENDL-4.0) has been released by JAEA. JENDL-4.0 is major update from JENDL- 3.3, and confirmed to give good accuracy by integral test for fission reactor systems such as fast neutron system and thermal neutron system. In this study, we evaluated the reactivity impact due to difference between ENDF/B-VII.0 and JENDL-4.0 for PWR fuel assembly burnup calculation using AEGIS code which has been developed by Nuclear Engineering, Ltd. in cooperation with Nuclear Fuel Industries, Ltd. and Nagoya University
2005-04-01
The aim of this publication is to serve as a practical resource for planning the medical response to a nuclear or radiological emergency. It fulfils in part functions assigned to the IAEA under Article 5.a(ii) of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (Assistance Convention), namely, to collect and disseminate to States Parties and Member States information concerning methodologies, techniques and available results of research relating to such emergencies. Effective medical response is a necessary component of the overall response to nuclear or radiological (radiation) emergencies. In general, the medical response may represent a difficult challenge for the authorities due to the complexity of the situation, often requiring specialized expertise, and special organizational arrangements and materials. To be effective, adequate planning and preparedness are needed. This manual, if implemented, should help to contribute to coherent international response. The manual provides the practical tools and generic procedures for use by emergency medical personnel during an emergency situation. It also provides guidance to be used at the stage of preparedness for development of medical response capabilities. The manual also addresses mass casualty emergencies resulting from malicious acts involving radioactive material. This part was supported by the Nuclear Security Fund. The manual was developed based on a number of assumptions about national and local capabilities. Therefore, it must be reviewed and revised as part of the planning process to match the potential accidents, threats, local conditions and other unique characteristics of the facility where it may be used
Proskuryakov, K. N.
2017-11-01
Created new scientific direction: “Diagnosis, prognosis and prevention of vibration - acoustic resonances in the nuclear power plant (NPP) equipment. The possibility of using methods for calculating and analyzing electric oscillation systems in the study of the properties of acoustic systems with a two-phase medium is proved, based on the similarity of the differential equations describing the state of these systems. Is shown that the developed methods can be used to predict and prevent the occurrence of vibration - acoustic resonances in the NPP equipment. Is shown that the volume of pressurizer at NPPs with VVER and PWR as well as boiling water reactor that exploded at Japan’s NPP Fukushima Daiichi is a Helmholtz resonator, which contain water and steam volumes and able many times increases the impact on them of outside periodic oscillations. Paper presents most important results published long before the severe accidents at NPPs Three Mile Island (TMI), Chernobyl and Fukushima Daiichi that could be used for the prediction of a severe accident scenario, identification of measuring data and process control in order to minimize the damage. Worked out results also could be useful in another industrial technologies based on applications of single and two-phase flows.
Hickman, R.; Reitter, T.
1985-01-01
The purpose of our investigation was to determine if the rapid progression of fire to flashover conditions in a furnished room, observed in a 1953 nuclear weapons test at the Nevada Test Site (the Encore Event), might be typical behavior rather than an aberration. If flashover under such conditions is indeed likely, this phenomenon is worth pursuing in view of the increased threat to buildings and human life from possible large-scale fires. We placed special emphasis on fires that occurred in modern rooms, i.e., ones furnished with upholstery and drapery materials made from synthetic polymers. Examination of photochemical processes showed them to be an unlikely explanation, either in Encore or in the future. Our calculation of rapid radiant-heating behavior of a few materials demonstrated that fabrics and fabric-covered foams would exceed their autoignition temperature when exposed to a 25-cal/cm 2 fluence from a 1-Mt air burst weapon. Because synthetic polymers have higher heating values and release heat faster during combustion than do the cellulosics used in the Encore experiment, early flashover should not be unexpected in contemporary households. However, the far-field thermal fluence required would be higher because of the absorption of thermal energy by windows and window coverings. Because of the complexity of the problem, carefully planned, full-scale experiments will be needed to finally answer the question. 39 refs., 9 figs., 8 tabs
Calculations of dosimetric parameter and REM meter response for BE(d, n) source
Chen Changmao
1988-01-01
Based on the recent data about neutron spectra, the average energy, effictive energy and conversion coefficient of fluence to dose equivalent are calculated for some Be (α, n) neutron sources which have differene types and structures. The responses of 2202D and 0075 REM meter for thses spectral neutrons are also estimated. The results indicate that the relationship between average energy and conversion coefficient or REM meter responses can be described by simple functions
Calculating the cost of research and Development in nuclear and radiation safety
Matsulevich, N.Je.; Nosovs'ka, A.A.
2010-01-01
Methodological support assessing the cost of research and development in the area of nuclear and radiation safety regulation is considered. Basic methodological recommendations for determining labor expenditures for research and development in nuclear and radiation safety are provided.
NFAP calculation of the response of a 1/6 scale reinforced concrete containment model
Costantino, C.J.; Pepper, S.; Reich, M.
1989-01-01
The details associated with the NFAP calculation of the pressure response of the 1/6th scale model containment structure are discussed in this paper. Comparisons are presented of some of the primary items of interest with those determined from the experiment. It was found from this comparison that the hoop response of the containment wall was adequately predicted by the NFAP finite element calculation, including the response in the high pressure, high strain range at which cracking of the concrete and yielding of the hoop reinforcement occurred. In the vertical or meridional direction, it was found that the model was significantly softer than predicted by the finite element calculation; that is, the vertical strains in the test were three to four times larger than computed in the NFAP calculation. These differences were noted even at low strain levels at which the concrete would not be expected to be cracked under tensile loadings. Simplified calculations for the containment indicate that the vertical stiffness of the wall is similar to that which would be determined by assuming the concrete fully cracked. Thus, the experiment indicates an anomalous behavior in the vertical direction
NFAP calculation of pressure response of 1/6th scale model containment structure
Costantino, C.J.; Pepper, S.; Reich, M.
1988-01-01
The details associated with the NFAP calculation of the pressure response of the 1/6th scale model containment structure are discussed in this paper. Comparisons are presented of some of the primary items of interest with those determined from the experiment. It was found from this comparison that the hoop response of the containment wall was adequately predicted by the NFAP finite element calculation, including the response in the high pressure, high strain range at which cracking of the concrete and yielding of the hoop reinforcement occurred. In the vertical or meridional direction, it was found that the model was significantly softer than predicted by the finite element calculation; that is, the vertical strains in the test were three to four times larger than computed in the NFAP calculation. These differences were noted even at low strain levels at which the concrete would not be expected to be cracked under tensile loadings. Simplified calculations for the containment indicate that the vertical stiffness of the wall is similar to that which would be determined by assuming the concrete fully cracked. Thus, the experiment indicates an anomalous behavior in the vertical direction
The nuclear spin response to intermediate energy protons and deuterons at low momentum transfer
Baker, F.T.; Djalali, C.; Glashausser, C.; Lenske, H.; Love, W.G.; Tomasi-Gustafsson, E.; Wambach, J.
1997-01-01
Measurements of polarization transfer in the inelastic scattering of intermediate energy protons and deuterons have yielded a wealth of data on the spin response of nuclei. This work complements the well-known studies of Gamow-Teller strength in charge-exchange reactions. The emphasis here is on a consistent determination of the S=1, T=0 response, practical only with deuterons, and on the proper separation of S=0 and S=1 strength in proton spectra for appropriate comparison with sum rules. We concentrate on two nuclei, 40 Ca and 12 C, at momentum transfers below about 1 fm -1 and on excitations up to about 50 MeV. The continuum second random phase approximation provides the primary theoretical tool for calculating and interpreting the response in terms of properties of the nucleon-nucleon force inside the nuclear medium. The reaction mechanism is described by the DWIA, applied here to continuum proton scattering almost as rigorously as it is usually applied to low energy excitations. A new DWIA formalism for the description of spin observables in deuteron scattering is used. Comparison of the proton and deuteron data with each other and with RPA/DWIA calculations yields interesting insights into the current state of understanding of collectivity and the nuclear spin response. (orig.)
Fiorito, L.; Piedra, D.; Cabellos, O.; Diez, C.J.
2015-01-01
Highlights: • We performed burnup calculations of PWR and BWR benchmarks using ALEPH and SCALE. • We propagated nuclear data uncertainty and correlations using different procedures and code. • Decay data uncertainties have negligible impact on nuclide densities. • Uncorrelated fission yields play a major role on the uncertainties of fission products. • Fission yields impact is strongly reduced by the introduction of correlations. - Abstract: Two fuel assemblies, one belonging to the Takahama-3 PWR and the other to the Fukushima-Daini-2 BWR, were modelled and the fuel irradiation was simulated with the TRITON module of SCALE 6.2 and with the ALEPH-2 code. Our results were compared to the experimental measurements of four samples: SF95-4 and SF96-4 were taken from the Takahama-3 reactor, while samples SF98-6 and SF99-6 belonged to the Fukushima-Daini-2. Then, we propagated the uncertainties coming from the nuclear data to the isotopic inventory of sample SF95-4. We used the ALEPH-2 adjoint procedure to propagate the decay constant uncertainties. The impact was inappreciable. The cross-section covariance information was propagated with the SAMPLER module of the beta3 version of SCALE 6.2. This contribution mostly affected the uncertainties of the actinides. Finally, the uncertainties of the fission yields were propagated both through ALEPH-2 and TRITON with a Monte Carlo sampling approach and appeared to have the largest impact on the uncertainties of the fission products. However, the lack of fission yield correlations results is a serious overestimation of the response uncertainties
Sensitivity and uncertainty analysis of nuclear responses in the EU HCLL TBM of ITER
Leichtle, Dieter; Fischer, Ulrich; Perel, Reuven L.; Serikov, Arkady
2011-01-01
Within the EU Fusion Technology Programme dedicated theoretical and experimental efforts are conducted to provide reliable nuclear data and computational tools for design analyses for fusion devices like ITER including qualified uncertainty estimates. In this respect, the present paper reports on sensitivity and uncertainty analyses for the EU HCLL Test Blanket Module (TBM) of ITER. Neutron flux spectra and tritium production rates have been calculated using MCNP with a modified version of the ITER Alite torus sector model with integrated TBMs. Sensitivities of such parameters to nuclear cross sections of isotopes contained in the TBM as well as in the ITER device have been calculated using the Monte Carlo code MCSEN. Uncertainties could be obtained by using existing covariance data of the important nuclear cross section files, mainly from ENDF/B-VI, SCALE6.0, but also from recent JEFF/EFF evaluations. Like in the HCLL mock-up experiment two positions at front and back of the TBM have been selected. In both cases the calculated uncertainties of the responses (tritium production rate, neutron flux) are in the range of 2-4%.
Gruel, A.
2011-01-01
Reactivity measurements by the oscillation technique, as those performed in the Minerve reactor, enable to access various neutronic parameters on materials, fuels or specific isotopes. Usually, expected reactivity effects are small, about ten pcm at maximum. Then, the modeling of these experiments should be very precise, to obtain reliable feedback on the pointed parameters. Especially, calculation biases should be precisely identified, quantified and reduced to get precise information on nuclear data. The goal of this thesis is to develop a reference calculation scheme, with well quantified uncertainties, for in-pile oscillation experiments. In this work are presented several small reactivity calculation methods, based on deterministic and/or stochastic calculation codes. Those method are compared thanks to a numerical benchmark, against a reference calculation. Three applications of these methods are presented here: a purely deterministic calculation with exact perturbation theory formalism is used for the experimental validation of fission product cross sections, in the frame of reactivity loss studies for irradiated fuel; an hybrid method, based on a stochastic calculation and the exact perturbation theory is used for the readjustment of nuclear data, here 241 Am; and a third method, based on a perturbative Monte Carlo calculation, is used in a conception study. (author) [fr
Shih, C.C.; Thuillier, R.H.
1984-01-01
The U.S. Nuclear Regulatory Commission requires that nuclear power plants provide for rapid assessment and response in the event of a radiological emergency. At the Diablo Canyon Nuclear Power Plant, Pacific Gas and Electric Company uses a system of linked central minicomputer, satellite desktop computers and microprocessors to provide decision makers with timely and pertinent information in emergency situations. The system provides for data acquisition and microprocessing at meteorological and radiological monitoring sites. Current estimates or projections of offsite dose commitment are made in real-time by a dispersion/dose calculation model. Computerized dissemination of data and calculational results to decision makers at the government and utility levels is also available. The basic system in use is a commercially available Emergency Assessment and Response System (EARS). This generic system has been modified in-house to meet requirements specific to emergency situations at the plant. Distinctive features of the modification program includes: a highly professional man-machine interaction; consideration of site-specific factors; simulation of environmental radiology for development of drill scenarios; and concise, pertinent reports as input to decision making
Calculation of Multisphere Neutron Spectrometer Response Functions in Energy Range up to 20 MeV
Martinkovic, J
2005-01-01
Multisphere neutron spectrometer is a basic instrument of neutron measurements in the scattered radiation field at charged-particles accelerators for radiation protection and dosimetry purposes. The precise calculation of the spectrometer response functions is a necessary condition of the propriety of neutron spectra unfolding. The results of the response functions calculation for the JINR spectrometer with LiI(Eu) detector (a set of 6 homogeneous and 1 heterogeneous moderators, "bare" detector within cadmium cover and without it) at two geometries of the spectrometer irradiation - in uniform monodirectional and uniform isotropic neutron fields - are given. The calculation was carried out by the code MCNP in the neutron energy range 10$^{-8}$-20 MeV.
Numerical calculation models of the elastoplastic response of a structure under seismic action
Edjtemai, Nima.
1982-06-01
Two digital calculation models developed in this work have made it possible to analyze the exact dynamic behaviour of ductile structures with one or several degrees of liberty, during earthquakes. With the first model, response spectra were built in the linear and non-linear fields for different absorption and ductility values and two types of seismic accelerograms. The comparative study of these spectra made it possible to check the validity of certain hypotheses suggested for the construction of elastoplastic spectra from corresponding linear spectra. A simplified method of non-linear seismic calculation based on the modal analysis and the spectra of elastoplastic response was then applied to structures with a varying number of degrees of liberty. The results obtained in this manner were compared with those provided by an exact calculation provided by the second digital model developed by us [fr
Parish, T.A.
1995-03-02
This project involved performing the nuclear design and safety analyses needed to modify the license issued by the Nuclear Regulatory Commission to allow operation of the Texas A& M University Nuclear Science Center Reactor (NSCR) with a core containing low enrichment uranium (LEU) fuel. The specific type of LEU fuel to be considered was the TRIGA 20-20 fuel produced by General Atomic. Computer codes for the neutronic analyses were provided by Argonne National Laboratory (ANL) and the assistance of William Woodruff of ANL in helping the NSCR staff to learn the proper use of the codes is gratefully acknowledged. The codes applied in the LEU analyses were WIMSd4/m, DIF3D, NCTRIGA and PARET. These codes allowed full three dimensional, temperature and burnup dependent calculations modelling the NSCR core to be performed for the first time. In addition, temperature coefficients of reactivity and pulsing calculations were carried out in-house, whereas in the past this modelling had been performed at General Atomic. In order to benchmark the newly acquired codes, modelling of the current NSCR core with highly enriched uranium fuel was also carried out. Calculated results were compared to both earlier licensing calculations and experimental data and the new methods were found to achieve excellent agreement with both. Therefore, even if an LEU core is never loaded at the NSCR, this project has resulted in a significant improvement in the nuclear safety analysis capabilities established and maintained at the NSCR.
Honda, K.K.
1976-01-01
As part of the structural response research program being conducted for ERDA, the response behavior of high-rise buildings in Las Vegas, Nevada, due to ground motion caused by underground nuclear explosions (UNEs) at the Nevada Test Site (NTS) has been measured for the past 12 years. Results obtained include variation in dynamic response properties as a function of amplitude of motion, influence of nonstructural partitions in the building response, and comparison of calculated and measured response. These data for three reinforced concrete high-rise buildings, all designed as moment-resisting space frames are presented
Komorovsky, Stanislav; Repisky, Michal; Malkin, Elena; Demissie, Taye B; Ruud, Kenneth
2015-08-11
We present an implementation of the nuclear spin-rotation (SR) constants based on the relativistic four-component Dirac-Coulomb Hamiltonian. This formalism has been implemented in the framework of the Hartree-Fock and Kohn-Sham theory, allowing assessment of both pure and hybrid exchange-correlation functionals. In the density-functional theory (DFT) implementation of the response equations, a noncollinear generalized gradient approximation (GGA) has been used. The present approach enforces a restricted kinetic balance condition for the small-component basis at the integral level, leading to very efficient calculations of the property. We apply the methodology to study relativistic effects on the spin-rotation constants by performing calculations on XHn (n = 1-4) for all elements X in the p-block of the periodic table and comparing the effects of relativity on the nuclear SR tensors to that observed for the nuclear magnetic shielding tensors. Correlation effects as described by the density-functional theory are shown to be significant for the spin-rotation constants, whereas the differences between the use of GGA and hybrid density functionals are much smaller. Our calculated relativistic spin-rotation constants at the DFT level of theory are only in fair agreement with available experimental data. It is shown that the scaling of the relativistic effects for the spin-rotation constants (varying between Z(3.8) and Z(4.5)) is as strong as for the chemical shieldings but with a much smaller prefactor.
Emergency Preparedness and Response at Nuclear Power Plants in Pakistan
Khan, L. A.; Qamar, M. A.; Liaquat, M.R., E-mail: samasl@yahoo.com [Pakistan Atomic Energy Commission, Islamabad (Pakistan)
2014-10-15
Emergency preparedness and response arrangements at Nuclear Power Plants (NPPs) in Pakistan have been reevaluated in the light of Fukushima Daiichi accident. Appropriate measures have been taken to strengthen and effectively implement the on-site and off-site emergency plans. Verification of these plans is conducted through regulatory review and by witnessing periodic emergency drills and exercises conducted by the NPPs in the fulfilment of the regulatory requirements. Emergency Planning Zones (EPZs) have been revised at NPPs. A multi discipline reserve force has been formed for assistance during severe accidents. Nuclear Emergency Management System (NEMS) has been established at the national level in order to make necessary arrangements for responding to nuclear and radiological emergencies. Training programs for first responders and medical professionals have been launched. Emergencies coordination centres have been established at national and corporate levels. Public awareness program has been initiated to ensure that the surrounding population is provided with appropriate information on emergency planning and response. To share national and international operational experience, Pakistan has arranged various workshops and developed a strong link with International Atomic Energy Agency (IAEA). (author)