WorldWideScience

Sample records for external criticality calculation

  1. Calculation of the Critical Current Reduction in a Brittle Round Multifilamentary Wire due to External Forces

    NARCIS (Netherlands)

    ten Haken, Bernard; Godeke, A.; ten Kate, Herman H.J.

    1994-01-01

    A simple model is presented that can describe the electro-mechanical state of a multifilamentary wire. An elastic cylinder model is used to derive the strain state analytically. Axial and transverse forces came a position dependent critical current density in the wire. The integral critical current

  2. Consistent Set of Experiments from ICSBEP Handbook for Evaluation of Criticality Calculation Prediction of Apparatus of External Fuel Cycle with Different Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Golovko, Yury E. [FSUE ' SSC RF-IPPE' , 249033, Bondarenko Square 1, Obninsk (Russian Federation)

    2008-07-01

    Experiments with plutonium, low enriched uranium and uranium-233 from the ICSBEP1 Handbook are being considered in this paper. Among these experiments it was selected only those, which seem to be the most relevant to the evaluation of uncertainty of critical mass of mixtures of plutonium or low enriched uranium or uranium-233 with light water. All selected experiments were examined and covariance matrices of criticality uncertainties were developed along with some uncertainties were revised. Statistical analysis of these experiments was performed and some contradictions were discovered and eliminated. Evaluation of accuracy of prediction of criticality calculations was performed using the internally consistent set of experiments with plutonium, low enriched uranium and uranium-233 remained after the statistical analyses. The application objects for the evaluation of calculational prediction of criticality were water-reflected spherical systems of homogeneous aqueous mixtures of plutonium or low enriched uranium or uranium-233 of different concentrations which are simplified models of apparatus of external fuel cycle. It is shows that the procedure allows to considerably reduce uncertainty in k{sub eff} caused by the uncertainties in neutron cross-sections. Also it is shows that the results are practically independent of initial covariance matrices of nuclear data uncertainties. (authors)

  3. MCNP and OMEGA criticality calculations

    International Nuclear Information System (INIS)

    Seifert, E.

    1998-04-01

    The reliability of OMEGA criticality calculations is shown by a comparison with calculations by the validated and widely used Monte Carlo code MCNP. The criticality of 16 assemblies with uranium as fissionable is calculated with the codes MCNP (Version 4A, ENDF/B-V cross sections), MCNP (Version 4B, ENDF/B-VI cross sections), and OMEGA. Identical calculation models are used for the three codes. The results are compared mutually and with the experimental criticality of the assemblies. (orig.)

  4. IRIS core criticality calculations

    International Nuclear Information System (INIS)

    Jecmenica, R.; Trontl, K.; Pevec, D.; Grgic, D.

    2003-01-01

    Three-dimensional Monte Carlo computer code KENO-VI of CSAS26 sequence of SCALE-4.4 code system was applied for pin-by-pin calculations of the effective multiplication factor for the first cycle IRIS reactor core. The effective multiplication factors obtained by the above mentioned Monte Carlo calculations using 27-group ENDF/B-IV library and 238-group ENDF/B-V library have been compared with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. The results of Monte Carlo calculations are found to be in good agreement with the results obtained by the nodal codes. The discrepancies in effective multiplication factor are typically within 1%. (author)

  5. Criticality calculations for safety analysis

    International Nuclear Information System (INIS)

    Vellozo, S.O.

    1981-01-01

    Criticality studies in uranium nitrate and plutonium nitrate aqueous solutions were done. For uranium compound three basic computer codes are used: GAMTEC-II, DTF-IV, KENO-IV. Water was used as refletor and the results obtained with the different computer codes were analyzed and compared with the 'Handbuck zur Kriticalitat'. The cross sections and the cylindrical geometry were generated by Gamtec-II computer code. In the second compound the thickness of the recipient with plutonium nitrate are used with rectangular geometry and concret reflector. The effective multiplication constant was calculated with the Gamtec-II and Keno-IV library. The results show many differences. (E.G) [pt

  6. A Methodology Proposal to Calculate the Externalities of Liquid Biofuels

    Energy Technology Data Exchange (ETDEWEB)

    Galan, A.; Gonzalez, R.; Varela, M. [Ciemat. Madrid (Spain)

    1999-05-01

    The aim of the survey is to propose a methodology to calculate the externalities associated with the liquid bio fuels cycle. The report defines the externalities from a theoretical point of view and classifies them. The reasons to value the externalities are explained as well as the existing methods. Furthermore, an evaluation of specific environmental and non-environmental externalities is also presented. The report reviews the current situation of the transport sector, considering its environmental effects and impacts. The progress made by the ExternE and ExternE-transport projects related the externalities of transport sector is assessed. Finally, the report analyses the existence of different economic instruments to internalize the external effects of the transport sector as well as other aspects of this internalization. (Author) 58 refs.

  7. Benchmarking criticality safety calculations with subcritical experiments

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1984-06-01

    Calculation of the neutron multiplication factor at delayed criticality may be necessary for benchmarking calculations but it may not be sufficient. The use of subcritical experiments to benchmark criticality safety calculations could result in substantial savings in fuel material costs for experiments. In some cases subcritical configurations could be used to benchmark calculations where sufficient fuel to achieve delayed criticality is not available. By performing a variety of measurements with subcritical configurations, much detailed information can be obtained which can be compared directly with calculations. This paper discusses several measurements that can be performed with subcritical assemblies and presents examples that include comparisons between calculation and experiment where possible. Where not, examples from critical experiments have been used but the measurement methods could also be used for subcritical experiments

  8. Calculations in external fields in quantum chromodynamics

    International Nuclear Information System (INIS)

    Novikov, V.A.; Shifman, M.A.; Vairshtejn, A.I.; Zakharov, V.I.

    1983-01-01

    The technique of calculation of operator expansion coefficients is reviewed. The main emphasis is put on gluon operators which appear in expansion of n-point functions induced by colourless quark currents. Two convenient schemes are discussed in detail: the abstract operator method and the method based on the Fock-Schwinger gauge for the vacuum gluon field. A large number of instructive examples important from the point of view of physical applications is considered

  9. Calculation of freight externality costs for South Africa

    Directory of Open Access Journals (Sweden)

    Stefaan Swarts

    2012-11-01

    Full Text Available The purpose of this study is to quantify the marginal external costs associated with freight transport in South Africa. Six cost elements are included as externality cost items, namely, costs related to accidents, emissions, roadway land availability, policing, noise and congestion. Inputs in the calculations were a gravity-oriented freight flow model, a road transport cost model, actual transport costs for other modes, a warehousing cost survey, an inventory delay calculation and various national sources of information such as accident statistics and government budgets. Estimation techniques resulted in advances for externality cost measurement in South Africa. The quantification of the cost elements will be used to update the South African Freight Demand Model. The results show that the cost of transportation would have been 20% more if external factors were taken into account. The marginal rates of externalities can be used to develop scenarios based on alternative choices for South Africa's freight transport infrastructure configuration.

  10. Calculation of Critical Temperatures by Empirical Formulae

    Directory of Open Access Journals (Sweden)

    Trzaska J.

    2016-06-01

    Full Text Available The paper presents formulas used to calculate critical temperatures of structural steels. Equations that allow calculating temperatures Ac1, Ac3, Ms and Bs were elaborated based on the chemical composition of steel. To elaborate the equations the multiple regression method was used. Particular attention was paid to the collection of experimental data which was required to calculate regression coefficients, including preparation of data for calculation. The empirical data set included more than 500 chemical compositions of structural steel and has been prepared based on information available in literature on the subject.

  11. Influence of the external neutron sources in the criticality prediction using 1/M curve

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Valmir [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil); Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil); Martinez, Aquilino Senra [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-972 Rio de Janeiro (Brazil)]. E-mail: aquilino@lmp.ufrj.br

    2005-11-15

    The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three-dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the corresponding problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system.

  12. Influence of the external neutron sources in the criticality prediction using 1/M curve

    International Nuclear Information System (INIS)

    Pereira, Valmir; Carvalho da Silva, Fernando; Martinez, Aquilino Senra

    2005-01-01

    The influence of external neutron sources in the process to obtain the criticality condition is estimated. To reach this objective, the three-dimensional neutron diffusion equation in two groups of energy is solved, for a subcritical PWR reactor core with external neutron sources. The results are compared with the solution of the corresponding problem without external neutron sources, that is an eigenvalue problem. The method developed for this purposes it makes use of both the nodal method (for calculation of the neutron flux) and the finite differences method (for calculation of the adjoint flux). A coarse mesh finite difference method was developed for the adjoint flux calculation, which uses the output of the nodal expansion method. The results regarding the influence of the external neutron source presence for attaining criticality have shown that far from criticality it is necessary to calculate the reactivity values of the system

  13. Influence of external source location in the reactivity calculation

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra

    2011-01-01

    We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

  14. Influence of external source location in the reactivity calculation

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Silva, Fernando Carvalho da; Martinez, Aquilino Senra, E-mail: asilva@con.ufrj.b, E-mail: fernando@con.ufrj.b, E-mail: Aquilino@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2011-07-01

    We used the neutron diffusion equation with external neutron sources, in cartesian geometry and the two groups of energy, to verify the influence of external neutron source locations in the reactivity calculation. For this, a coarse mesh finite difference method was developed for the adjoint flux calculation and simplifies reactivity calculation in PWR type reactor, which uses the output of the nodal expansion method. The results were obtained for different locations on the two-dimensional plane, as well as for different types of fuel elements in the reactor core. (author)

  15. Description of Fracture Systems for External Criticality Reports

    International Nuclear Information System (INIS)

    Nicot, Jean-Philippe

    2001-01-01

    The purpose of this Analysis/Model Report (AMR) is to describe probabilistically the main features of the geometry of the fracture system in the vicinity of the repository. They will be used to determine the quantity of fissile material that could accumulate in the fractured rock underneath a waste package as it degrades. This AMR is to feed the geochemical calculations for external criticality reports. This AMR is done in accordance with the technical work plan (BSC (Bechtel SAIC Company) 2001 b). The scope of this AMR is restricted to the relevant parameters of the fracture system. The main parameters of interest are fracture aperture and fracture spacing distribution parameters. The relative orientation of the different fracture sets is also important because of its impact on criticality, but they will be set deterministically. The maximum accumulation of material depends primarily on the fracture porosity, combination of the fracture aperture, and fracture intensity. However, the fracture porosity itself is not sufficient to characterize the potential for accumulation of a fracture system. The fracture aperture is also important because it controls both the flow through the fracture and the potential plugging of the system. Other features contributing to the void space such as lithophysae are also investigated. On the other hand, no analysis of the matrix porosity is done. The parameters will be used in sensitivity analyses of geochemical calculations providing actinide accumulations and in the subsequent Monte Carlo criticality analyses

  16. DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

  17. Criticality criteria for submissions based on calculations

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1975-06-01

    Calculations used in criticality clearances are subject to errors from various sources, and allowance must be made for these errors is assessing the safety of a system. A simple set of guidelines is defined, drawing attention to each source of error, and recommendations as to its application are made. (author)

  18. The external field dependence of the BCS critical temperature

    DEFF Research Database (Denmark)

    Frank, Rupert L.; Hainzl, Christian; Seiringer, Robert

    2016-01-01

    We consider the Bardeen-Cooper-Schrieffer free energy functional for particles interacting via a two-body potential on a microscopic scale and in the presence of weak external fields varying on a macroscopic scale. We study the influence of the external fields on the critical temperature. We show...

  19. Subcriticality calculation in nuclear reactors with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. Subcriticality calculation in nuclear reactors with external neutron sources

    International Nuclear Information System (INIS)

    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)

  1. Quantization of fermions in external soliton fields and index calculation

    International Nuclear Information System (INIS)

    Grosse, H.

    1986-01-01

    We review recent results on the quantization of fermions in external fields, discuss equivalent and inequivalent representations of the canonical anticommutation relations, indicate how the requirement of implementability of gauge transformations leads to quantization conditions, determine the algebra of charges, identify the Schwinger term and remark finally how one may calculate a ground state charge. (Author)

  2. Criticality calculation method for mixer-settlers

    International Nuclear Information System (INIS)

    Gonda, Kozo; Aoyagi, Haruki; Nakano, Ko; Kamikawa, Hiroshi.

    1980-01-01

    A new criticality calculation code MACPEX has been developed to evaluate and manage the criticality of the process in the extractor of mixer-settler type. MACPEX can perform the combined calculation with the PUREX process calculation code MIXSET, to get the neutron flux and the effective multiplication constant in the mixer-settlers. MACPEX solves one-dimensional diffusion equation by the explicit difference method and the standard source-iteration technique. The characteristics of MACPEX are as follows. 1) Group constants of 4 energy groups for the 239 Pu-H 2 O solution, water, polyethylene and SUS 28 are provided. 2) The group constants of the 239 Pu-H 2 O solution are given by the functional formulae of the plutonium concentration, which is less than 50 g/l. 3) Two boundary conditions of the vacuum condition and the reflective condition are available in this code. 4) The geometrical bucklings can be calculated for a certain energy group and/or region by using the three dimentional neutron flux profiles obtained by CITATION. 5) The buckling correction search can be carried out in order to get a desired k sub(eff). (author)

  3. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  4. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  5. Criticality calculations with MCNP trademark: A primer

    International Nuclear Information System (INIS)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-01-01

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand

  6. Criticality Calculations with MCNP6 - Practical Lectures

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

    2016-11-29

    These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

  7. Criticality Calculations with MCNP6 - Practical Lectures

    International Nuclear Information System (INIS)

    Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

    2016-01-01

    These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

  8. Primer for criticality calculations with DANTSYS

    International Nuclear Information System (INIS)

    Busch, R.D.

    1996-01-01

    With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his or her facility. Typically, two types of codes are available: deterministic codes such as ANISN or DANTSYS that solve an approximate model exactly and Monte Carlo Codes such as KENO or MCNP that solve an exact model approximately. Often, the analyst feels that the deterministic codes are too simple and will not provide the necessary information, so most modeling uses Monte Carlo methods. This sometimes means that hours of effort are expended to produce results available in minutes from deterministic codes. A substantial amount of reliable information on nuclear systems can be obtained using deterministic methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico in cooperation with the Radiation Transport Group at Los Alamos National Laboratory has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. (DANTSYS is the name of a suite of codes that users more commonly know as ONEDANT, TWODANT, TWOHEX, and THREEDANT.) It assumes a college education in a technical field, but there is no assumption of familiarity with neutronics codes in general or with DANTSYS in particular. The primer is designed to teach by example, with each example illustrating two or three DANTSYS features useful in criticality analyses

  9. Parallel processing of dose calculation for external photon beam therapy

    International Nuclear Information System (INIS)

    Kunieda, Etsuo; Ando, Yutaka; Tsukamoto, Nobuhiro; Ito, Hisao; Kubo, Atsushi

    1994-01-01

    We implemented external photon beam dose calculation programs into a parallel processor system consisting of Transputers, 32-bit processors especially suitable for multi-processor configuration. Two network conformations, binary-tree and pipeline, were evaluated for rectangular and irregular field dose calculation algorithms. Although computation speed increased in proportion to the number of CPU, substantial overhead caused by inter-processor communication occurred when a smaller computation load was delivered to each processor. On the other hand, for irregular field calculation, which requires more computation capability for each calculation point, the communication overhead was still less even when more than 50 processors were involved. Real-time responses could be expected for more complex algorithms by increasing the number of processors. (author)

  10. Monte Carlo method for array criticality calculations

    International Nuclear Information System (INIS)

    Dickinson, D.; Whitesides, G.E.

    1976-01-01

    The Monte Carlo method for solving neutron transport problems consists of mathematically tracing paths of individual neutrons collision by collision until they are lost by absorption or leakage. The fate of the neutron after each collision is determined by the probability distribution functions that are formed from the neutron cross-section data. These distributions are sampled statistically to establish the successive steps in the neutron's path. The resulting data, accumulated from following a large number of batches, are analyzed to give estimates of k/sub eff/ and other collision-related quantities. The use of electronic computers to produce the simulated neutron histories, initiated at Los Alamos Scientific Laboratory, made the use of the Monte Carlo method practical for many applications. In analog Monte Carlo simulation, the calculation follows the physical events of neutron scattering, absorption, and leakage. To increase calculational efficiency, modifications such as the use of statistical weights are introduced. The Monte Carlo method permits the use of a three-dimensional geometry description and a detailed cross-section representation. Some of the problems in using the method are the selection of the spatial distribution for the initial batch, the preparation of the geometry description for complex units, and the calculation of error estimates for region-dependent quantities such as fluxes. The Monte Carlo method is especially appropriate for criticality safety calculations since it permits an accurate representation of interacting units of fissile material. Dissimilar units, units of complex shape, moderators between units, and reflected arrays may be calculated. Monte Carlo results must be correlated with relevant experimental data, and caution must be used to ensure that a representative set of neutron histories is produced

  11. Neutronic calculations for a subcritical system with external source

    International Nuclear Information System (INIS)

    Cintas, A; Lopasso, E.M; Marquez Damian, J. I

    2006-01-01

    We present a neutronic study on an A D S, systems capable of transmute minor actinides and fission products in order to reduce their radiotoxicity and mean-life.We compare neutronic parameters obtained with Scale/Tort and M C N P modelling a sub-critical system with source from a N E A Benchmark.Due to lack of nuclear data at the temperature of the system, we perform calculations at available temperature of libraries (300 K); to compensate the reactivity insertion due to the temperature change we reduce the size of the fuel zone in order to get a sub-critical system that allow u s to evaluate neutronic parameters of the system with source.We have found that the numerical results (neutron spectrum, neutron flux distributions and other neutronic parameters) are in agreement with the M C N P and with those of the benchmark participants even though the geometric models used are not exactly the same. We conclude that with the real temperature cross sections, the calculation scheme developed (Scale/Tort and M C N P) will give reliable results in A D S evaluations [es

  12. IMPORTANCE OF DELIVERY CONDITIONS IN THE EXTERNAL PRICE CALCULATION

    Directory of Open Access Journals (Sweden)

    Violeta ISAI

    2005-01-01

    Full Text Available The delivery conditions of the merchandise at export, established by the rules INCOTERMS2000, influence the external price structure. There are some conditions in which the external priceincludes only the value of the merchandise and other conditions in which, besides the value of themerchandise, the price includes also the external transport and insurance. In the case of the exportson commercial credit, when it appears the notion of external interest, this one may be included in theprice or may be invoiced separately, thus defining gross external prices and net external prices.

  13. Criticality calculation by the LTSN method

    International Nuclear Information System (INIS)

    Batistela, Claudia H.F.; Vilhena, Marco T. de; Borges, Volnei

    1997-01-01

    This work evaluates criticality parameters (multiplication factor and critical thickness) by the LTS N method in unidimensional slabs homogeneous and heterogeneous considering one-group model and isotropic scattering. The idea of the LTS N method encompasses the following steps: application of the Laplace transform into a set of discrete ordinates equations, analytical solution of the algebraic linear system for the transformed angular fluxes and their reconstruction by the Heaviside expansion technique. The novel feature of the proposed method is based upon the criticality parameters determination by solving a transcendental equation. Numerical results are reported. 12 refs., 2 tabs

  14. Identification of External Critical Success Factors in Microbial Biotechnology Firms

    Directory of Open Access Journals (Sweden)

    Alireza AZIMI

    2013-03-01

    Full Text Available Microbial biotechnology is expected to change production methods, the products themselves and the structure of the industries in the new economies. Hopefully, countries in the Middle-East, Latin America, Asia and Africa have already recognized the importance of microbial biotechnology's promise. In this sense, the importance of externalities which might affect the success or failure of these companies becomes an issue of paramount importance. In the present study, we will try to identify the main external factors which could lead in the success of microbial biotechnology firms in Iran. To do so, the research follows a qualitative research design to answer this main question. Based on our findings, critical success factors are categorized in the following categories: General Environment (GE, Political Position (PP, Economic Position (EP, and Market Position (MP.

  15. Criticality calculation of non-ordinary systems

    Energy Technology Data Exchange (ETDEWEB)

    Kalugin, A. V., E-mail: Kalugin-AV@nrcki.ru; Tebin, V. V. [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    The specific features of calculation of the effective multiplication factor using the Monte Carlo method for weakly coupled and non-asymptotic multiplying systems are discussed. Particular examples are considered and practical recommendations on detection and Monte Carlo calculation of systems typical in numerical substantiation of nuclear safety for VVER fuel management problems are given. In particular, the problems of the choice of parameters for the batch mode and the method for normalization of the neutron batch, as well as finding and interpretation of the eigenvalue spectrum for the integral fission matrix, are discussed.

  16. Criticality calculations for homogeneous mixtures of uranium and plutonium

    International Nuclear Information System (INIS)

    Spiegelberg, R. de S.H.

    1981-05-01

    Critical parameters were calculated using the one-dimensional multigroup transport theory. Calculations have been performed for water mixture of uranium metal and uranium oxides and plutonium nitrates to determine the dimensions of simple critical geometries (sphere and cylinder). The results of the calculations were plotted showing critical parameters (volume, radius or critical mass). The critical values obtained in Handbuch zur Kritikalitat were used to compare with critical parameters. A sensitivity study for the influences of mesh space size, multigroup structure and order of the S sub(n) approximation on the critical radius was carried out. The GAMTEC-II code was used to generate multigroup cross sections data. Critical radius were calculated using the one-dimensional multigroup transport code DTF-IV. (Author) [pt

  17. Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

    Energy Technology Data Exchange (ETDEWEB)

    Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-26

    The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

  18. Calculating externalities from damages in occupational health and safety

    Energy Technology Data Exchange (ETDEWEB)

    Burtraw, D; Shefftz, J

    1994-07-01

    This paper surveys the theoretical basis for the possibility that coal miner occupational health and safety damages are not adequately internalized into the production cost of mining coal and thereby impose an external cost on society.

  19. Calculating externalities from damages in occupational health and safety

    International Nuclear Information System (INIS)

    Burtraw, D.; Shefftz, J.

    1994-01-01

    This paper surveys the theoretical basis for the possibility that coal miner occupational health and safety damages are not adequately internalized into the production cost of mining coal and thereby impose an external cost on society

  20. SIMCRI: a simple computer code for calculating nuclear criticality parameters

    International Nuclear Information System (INIS)

    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)

  1. Application of MCNP in the criticality calculation for reactors

    International Nuclear Information System (INIS)

    Zhong Zhaopeng; Shi Gong; Hu Yongming

    2003-01-01

    The criticality calculation is carried out with 3-D Monte Carlo code (MCNP). The author focuses on the introduction of modelling of the core and reflector. The core description is simplified by using repetition structure function of MCNP. k eff in different control rods positions are calculated for the case of JRR3, and the results is consistent with that of the reference. This work shows that MCNP is applicable for reactor criticality calculation

  2. Calculation of the minimum critical mass of fissile nuclides

    International Nuclear Information System (INIS)

    Wright, R.Q.; Hopper, Calvin Mitchell

    2008-01-01

    The OB-1 method for the calculation of the minimum critical mass of fissile actinides in metal/water systems was described in a previous paper. A fit to the calculated minimum critical mass data using the extended criticality parameter is the basis of the revised method. The solution density (grams/liter) for the minimum critical mass is also obtained by a fit to calculated values. Input to the calculation consists of the Maxwellian averaged fission and absorption cross sections and the thermal values of nubar. The revised method gives more accurate values than the original method does for both the minimum critical mass and the solution densities. The OB-1 method has been extended to calculate the uncertainties in the minimum critical mass for 12 different fissile nuclides. The uncertainties for the fission and capture cross sections and the estimated nubar uncertainties are used to determine the uncertainties in the minimum critical mass, either in percent or grams. Results have been obtained for U-233, U-235, Pu-236, Pu-239, Pu-241, Am-242m, Cm-243, Cm-245, Cf-249, Cf-251, Cf-253, and Es-254. Eight of these 12 nuclides are included in the ANS-8.15 standard.

  3. MCNP Perturbation Capability for Monte Carlo Criticality Calculations

    International Nuclear Information System (INIS)

    Hendricks, J.S.; Carter, L.L.; McKinney, G.W.

    1999-01-01

    The differential operator perturbation capability in MCNP4B has been extended to automatically calculate perturbation estimates for the track length estimate of k eff in MCNP4B. The additional corrections required in certain cases for MCNP4B are no longer needed. Calculating the effect of small design changes on the criticality of nuclear systems with MCNP is now straightforward

  4. External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository

    International Nuclear Information System (INIS)

    McClure, J.

    2001-01-01

    This purpose of this technical report is to provide a comprehensive summary of the waste package (WP) external criticality-related risk of the Plutonium Disposition ceramic waste form, which is being developed and evaluated by the Office of Fissile Materials Disposition of the United States Department of Energy (DOE). Potential accumulation of the fissile materials, 239 Pu and 235 U, in rock formations having a favorable chemical environment for such actions, requires analysis because autocatalytic configurations, while unlikely to form, never-the-less have consequences which are undesirable and require evaluation. Secondly, the WP design has evolved necessitating a re-evaluation of the internal WP degradation scenarios that contribute to the external source terms. The scope of this study includes a summary of the revised WP degradation calculations, a summary of the accumulation mechanisms in fractures and lithophysae in the tuff beneath the WP footprint, and a summary of the criticality risk calculations from any accumulated fissile material. Accumulations of fissile material external to the WP sufficient to pose a potential criticality risk require a deposition mechanism operating over sufficient time to reach required levels. The transporting solution concentrations themselves are well below critical levels (CRWMS 2001e). The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded in High-Level Waste (HLW) glass in the standard HLW glass disposal canister. The ceramic disks would occupy approximately 12% of the HLW canister volume, while most of the remaining 88% of the volume would be occupied by HLW glass

  5. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

  6. Calculational study for criticality safety data of fissionable actinides

    International Nuclear Information System (INIS)

    Nojiri, Ichiro; Fukasaku, Yasuhiro.

    1997-01-01

    This study has been carried out to obtain basic criticality safety characteristics of minor actinides nuclides. Criticality safety data of minor actinides nuclides have been surveyed through public literatures. Critical mass of seven nuclides, Np-237, Am-241, Am-242m, Am-243, Cm-243, Cm-244 and Cm-245, have been calculated by using two code systems of criticality safety analysis, SCALE-4 and MCNP4A, under some material and reflector conditions. Some applicable cross-section libraries have been used for each code systems. Calculated data have been compared with each other and with published data. The results of this comparison shows that there is no discrepancy within the computational codes and the calculated data is strongly depend on the cross-section library. (author)

  7. Validation of calculational methods for nuclear criticality safety - approved 1975

    International Nuclear Information System (INIS)

    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

  8. Calculation of external buildings of 1300 MW PWR

    International Nuclear Information System (INIS)

    Chataignier, Jacques

    1981-01-01

    The outer containment of reactor building is composed of a cylindrical reinforced unprestressed concrete shell. Two kinds of stress have been taken into account for dimensioning the steels and checking the concrete stresses; these are: - cases of so called ''normal'' stresses that the structure is supposed to withstand without there being any effect on its operational conditions and durability, - cases of so called ''increased'' stresses which if exceeded correspond to the ruin of the structure. The strains created in the structure by the different sorts of stresses were determined by computer with the help of calculation models by finite element method to which loads are applied, broken down, as the case may be, into Fourier series when they are not axisymmetrical [fr

  9. Variational Variance Reduction for Monte Carlo Criticality Calculations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.; Larsen, Edward W.

    2001-01-01

    A new variational variance reduction (VVR) method for Monte Carlo criticality calculations was developed. This method employs (a) a variational functional that is more accurate than the standard direct functional, (b) a representation of the deterministically obtained adjoint flux that is especially accurate for optically thick problems with high scattering ratios, and (c) estimates of the forward flux obtained by Monte Carlo. The VVR method requires no nonanalog Monte Carlo biasing, but it may be used in conjunction with Monte Carlo biasing schemes. Some results are presented from a class of criticality calculations involving alternating arrays of fuel and moderator regions

  10. New model for mines and transportation tunnels external dose calculation using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Allam, Kh. A.

    2017-01-01

    In this work, a new methodology is developed based on Monte Carlo simulation for tunnels and mines external dose calculation. Tunnels external dose evaluation model of a cylindrical shape of finite thickness with an entrance and with or without exit. A photon transportation model was applied for exposure dose calculations. A new software based on Monte Carlo solution was designed and programmed using Delphi programming language. The variation of external dose due to radioactive nuclei in a mine tunnel and the corresponding experimental data lies in the range 7.3 19.9%. The variation of specific external dose rate with position in, tunnel building material density and composition were studied. The given new model has more flexible for real external dose in any cylindrical tunnel structure calculations. (authors)

  11. Iterative acceleration methods for Monte Carlo and deterministic criticality calculations

    International Nuclear Information System (INIS)

    Urbatsch, T.J.

    1995-11-01

    If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors

  12. Iterative acceleration methods for Monte Carlo and deterministic criticality calculations

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J.

    1995-11-01

    If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors.

  13. Initialization bias suppression in iterative Monte Carlo calculations: benchmarks on criticality calculation

    International Nuclear Information System (INIS)

    Richet, Y.; Jacquet, O.; Bay, X.

    2005-01-01

    The accuracy of an Iterative Monte Carlo calculation requires the convergence of the simulation output process. The present paper deals with a post processing algorithm to suppress the transient due to initialization applied on criticality calculations. It should be noticed that this initial transient suppression aims only at obtaining a stationary output series, then the convergence of the calculation needs to be guaranteed independently. The transient suppression algorithm consists in a repeated truncation of the first observations of the output process. The truncation of the first observations is performed as long as a steadiness test based on Brownian bridge theory is negative. This transient suppression method was previously tuned for a simplified model of criticality calculations, although this paper focuses on the efficiency on real criticality calculations. The efficiency test is based on four benchmarks with strong source convergence problems: 1) a checkerboard storage of fuel assemblies, 2) a pin cell array with irradiated fuel, 3) 3 one-dimensional thick slabs, and 4) an array of interacting fuel spheres. It appears that the transient suppression method needs to be more widely validated on real criticality calculations before any blind using as a post processing in criticality codes

  14. Acceleration and increased control of convergence in criticality calculations

    International Nuclear Information System (INIS)

    Jinaphanh, Alexis

    2014-01-01

    IRSN is developing a numerical simulation code called Moret to assess the nuclear criticality risk. This tool is designed to perform 3D simulations of neutron transport in a given system. It achieves this by adopting a probabilistic approach known as Monte Carlo, in which the transport of several successive generations of neutrons is calculated from an initial neutron distribution in the system under study. These generations are simulated until it is considered that convergence of the effective neutron multiplication coefficient (or K eff ) - which characterizes the gap before reaching the critical state - has been reached. Insufficient convergence can lead to underestimation of both K eff and the criticality risk. During this thesis work, A. Jinaphanh sought to improve the reliability of values by developing a new method for initializing calculations, together with a criterion used to reliably determine whether or not convergence has been reached. (author)

  15. TRIGA criticality experiment for testing burn-up calculations

    International Nuclear Information System (INIS)

    Persic, Andreja; Ravnik, Matjaz; Zagar, Tomaz

    1999-01-01

    A criticality experiment with partly burned TRIGA fuel is described. 20 wt % enriched standard TRIGA fuel elements initially containing 12 wt % U are used. Their average burn-up is 1.4 MWd. Fuel element burn-up is calculated in 2-D four group diffusion approximation using TRIGLAV code. The burn-up of several fuel elements is also measured by reactivity method. The excess reactivity of several critical and subcritical core configurations is measured. Two core configurations contain the same fuel elements in the same arrangement as were used in the fresh TRIGA fuel criticality experiment performed in 1991. The results of the experiment may be applied for testing the computer codes used for fuel burn-up calculations. (author)

  16. Presentation and qualification of criticality calculation in fuel element storage

    International Nuclear Information System (INIS)

    Ermumcu, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

    Faced with the growing size of criticality calculation requests a fast and slightly conservative method has been perfected for evaluating the effective multiplication constant of sites containing PWR type elements. This method is based on the use of the DOT 3.5 code which requires a bidimensional modelisation of the geometry of the problem and the placing into groups of the macroscopic cross sections of the various materials. This preliminary work is effected by various APOLLO calculations. This diagram is qualified by comparison with the results obtained by the Monte Carlo TRIPOLI code. Comparing the values obtained by MORET and APOLLO-DOT for the criticality of transport flask end in good agreement. For the parametric studies, a large number of calculations can be necessary, and analytical methods cost little for simple geometries. This diagram can be used for studying small transport flasks but it is particularly advantageous for storages [fr

  17. Methodology of external exposure calculation for reuse of conditional released materials from decommissioning - 59138

    International Nuclear Information System (INIS)

    Ondra, Frantisek; Vasko, Marek; Necas, Vladimir

    2012-01-01

    before expedition or temporary storage before next manufacturing activity was found as a critical activity from external exposure point of view. Significant increase of worker exposure is caused by storage of huge amount of radioactive rails in piles. Therefore it is recommended to distribute storage rails into several individual separate storage places. The optimised radionuclide concentrations in conditional released material for external exposure will be compared with radionuclide concentrations for internal exposures calculated by means of GoldSim software tool (another part of the CONRELMAT project). Based on this comparison, more conservative radionuclide concentrations will be selected. (authors)

  18. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. The spread of the results in the international calculation amounted to ± 12,000 pcm in the realistic fuel dissolver exercise n degrees 19 proposed by BNFL, and to ± 25,000 pcm in the benchmark n degrees 20 in which fissile material in solid form is surrounded by fissile material in solution. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat latter effect, permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates solicited from the participants. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism (NITAWL in the international SCALE package) to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. Improvements in the up-dated 1990 contributions, as do recent complementary reference calculations (MCNP, VIM, ultrafine slowing-down CGM calculation), confirm the need to use rigorous self-shielding methods in criticality design-oriented codes. 6 refs., 11 figs., 3 tabs

  19. Percolation with first-and-second neighbour bonds: a renormalization-group calculation of critical exponents

    International Nuclear Information System (INIS)

    Riera, R.; Oliveira, P.M.C. de; Chaves, C.M.G.F.; Queiroz, S.L.A. de.

    1980-04-01

    A real-space renormalization group approach for the bond percolation problem in a square lattice with first- and second- neighbour bonds is proposed. The respective probabilities are treated, as independent variables. Two types of cells are constructed. In one of them the lattice is considered as two interpenetrating sublattices, first-neighbour bonds playing the role of intersublattice links. This allows the calculation of both critical exponents ν and γ, without resorting to any external field. Values found for the critical indices are in good agreement with data available in the literature. The phase diagram in parameter space is also obtained in each case. (Author) [pt

  20. Evaluation and validation of criticality codes for fuel dissolver calculations

    International Nuclear Information System (INIS)

    Santamarina, A.; Smith, H.J.; Whitesides, G.E.

    1991-01-01

    During the past ten years an OECD/NEA Criticality Working Group has examined the validity of criticality safety computational methods. International calculation tools which were shown to be valid in systems for which experimental data existed were demonstrated to be inadequate when extrapolated to fuel dissolver media. A theoretical study of the main physical parameters involved in fuel dissolution calculations was performed, i.e. range of moderation, variation of pellet size and the fuel double heterogeneity effect. The APOLLO/P IC method developed to treat this latter effect permits us to supply the actual reactivity variation with pellet dissolution and to propose international reference values. The disagreement among contributors' calculations was analyzed through a neutron balance breakdown, based on three-group microscopic reaction rates. The results pointed out that fast and resonance nuclear data in criticality codes are not sufficiently reliable. Moreover the neutron balance analysis emphasized the inadequacy of the standard self-shielding formalism to account for 238 U resonance mutual self-shielding in the pellet-fissile liquor interaction. The benchmark exercise has resolved a potentially dangerous inadequacy in dissolver calculations. (author)

  1. Analysis on First Criticality Benchmark Calculation of HTR-10 Core

    International Nuclear Information System (INIS)

    Zuhair; Ferhat-Aziz; As-Natio-Lasman

    2000-01-01

    HTR-10 is a graphite-moderated and helium-gas cooled pebble bed reactor with an average helium outlet temperature of 700 o C and thermal power of 10 MW. The first criticality benchmark problem of HTR-10 in this paper includes the loading number calculation of nuclear fuel in the form of UO 2 ball with U-235 enrichment of 17% for the first criticality under the helium atmosphere and core temperature of 20 o C, and the effective multiplication factor (k eff ) calculation of full core (5 m 3 ) under the helium atmosphere and various core temperatures. The group constants of fuel mixture, moderator and reflector materials were generated with WlMS/D4 using spherical model and 4 neutron energy group. The critical core height of 150.1 cm obtained from CITATION in 2-D R-Z reactor geometry exists in the calculation range of INET China, JAERI Japan and BATAN Indonesia, and OKBM Russia. The k eff calculation result of full core at various temperatures shows that the HTR-10 has negative temperature coefficient of reactivity. (author)

  2. Linear filtering applied to Monte Carlo criticality calculations

    International Nuclear Information System (INIS)

    Morrison, G.W.; Pike, D.H.; Petrie, L.M.

    1975-01-01

    A significant improvement in the acceleration of the convergence of the eigenvalue computed by Monte Carlo techniques has been developed by applying linear filtering theory to Monte Carlo calculations for multiplying systems. A Kalman filter was applied to a KENO Monte Carlo calculation of an experimental critical system consisting of eight interacting units of fissile material. A comparison of the filter estimate and the Monte Carlo realization was made. The Kalman filter converged in five iterations to 0.9977. After 95 iterations, the average k-eff from the Monte Carlo calculation was 0.9981. This demonstrates that the Kalman filter has the potential of reducing the calculational effort of multiplying systems. Other examples and results are discussed

  3. Calculation of HTR-10 first criticality with MVP

    International Nuclear Information System (INIS)

    Xie Jiachun; Yao Lianying

    2015-01-01

    The first criticality of 10 MW pebble-bed high temperature gas-cooled reactor-test module (HTR-10) was calculated with MVP. According to the characteristics of HTR-10, the Statistical Geometry Model of MVP was employed to describe the random arrangement of coated fuel particles in the fuel pebbles and the random distribution of the fuel and dummy pebbles in the core. Compared with previous results from VSOP and MCNP, the MVP results with JENDL-3.3 library were little more different, but the results with ENDF/B-Ⅵ.8 library were very close. The relative errors were less than 0.7%, compared with the first criticality experimental results. The study shows that MVP could be used in the physics calculations for pebble bed high temperature gas-cooled reactors. (authors)

  4. Cluster monte carlo method for nuclear criticality safety calculation

    International Nuclear Information System (INIS)

    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

  5. Calculating concentration of inhaled radiolabeled particles from external gamma counting: External counting efficiency and attenuation coefficient of thorax

    International Nuclear Information System (INIS)

    Langenback, E.G.; Foster, W.M.; Bergofsky, E.H.

    1989-01-01

    We determined the overall external counting efficiency of radiolabeled particles deposited in the sheep lung. This efficiency permits the noninvasive calculation of the number of particles and microcuries from gamma-scintillation lung images of the live sheep. Additionally, we have calculated the attenuation of gamma radiation (120 keV) by the posterior chest wall and the gamma-scintillation camera collection efficiency of radiation emitted from the lung. Four methods were employed in our experiments: (1) by light microscopic counting of discrete carbonized polystyrene particles with a count median diameter (CMD) of 2.85 microns and tagged with cobalt-57, we delineated a linear relationship between the number of particles and the emitted counts per minute (cpm) detected by well scintillation counting; (2) from this conversion relationship we determined the number of particles inhaled and deposited in the lungs by scintillation counting fragments of dissected lung at autopsy; (3) we defined a linear association between the number of particles or microcuries contained in the lung and the emitted radiation as cpm detected by a gamma scintillation camera in the live sheep prior to autopsy; and (4) we compared the emitted radiation from the lungs of the live sheep to that of whole excised lungs in order to calculate the attenuation coefficient (ac) of the chest wall. The mean external counting efficiency was 4.00 X 10(4) particles/cpm (5.1 X 10(-3) microCi/cpm), the camera collection efficiency was 1 cpm/10(4) disintegrations per minute (dpm), and the ac had a mean of 0.178/cm. The external counting efficiency remained relatively constant over a range of particles and microcuries, permitting a more general use of this ratio to estimate number of particles or microcuries depositing after inhalation in a large mammalian lung if a similarly collimated gamma camera system is used

  6. Time delays between core power production and external detector response from Monte Carlo calculations

    International Nuclear Information System (INIS)

    Valentine, T.E.; Mihalczo, J.T.

    1996-01-01

    One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor

  7. Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies

  8. New calculations for critical assemblies using MCNP4B

    International Nuclear Information System (INIS)

    Adams, A.A.; Frankle, S.C.; Little, R.C.

    1997-07-01

    A suite of 41 criticality benchmarks has been modeled using MCNP trademark (version 4B). Most of the assembly specifications were obtained from the Cross Section Evaluation Working Group (CSEWG) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) compendiums of experimental benchmarks. A few assembly specifications were obtained from experimental papers. The suite contains thermal and fast assemblies, bare and reflected assemblies, and emphasizes 233 U, 235 U, 238 U, and 239 Pu. The values of k eff for each assembly in the suite were calculated using MCNP libraries derived primarily from release 2 of ENDF/B-V and release 2 of ENDF/B-VI. The results show that the new ENDF/B-VI.2 evaluations for H, O, N, B, 235 U, 238 U, and 239 Pu can have a significant impact on the values of k eff . In addition to the integral quantity k eff , several additional experimental measurements were performed and documented. These experimental measurements include central fission and reaction-rate ratios for various isotopes, and neutron leakage and flux spectra. They provide more detailed information about the accuracy of the nuclear data than can k eff . Comparison calculations were performed using both ENDF/B-V.2 and ENDF/B-VI.2-based data libraries. The purpose of this paper is to compare the results of these additional calculations with experimental data, and to use these results to assess the quality of the nuclear data

  9. External rhinoplasty: a critical analysis of 500 cases.

    Science.gov (United States)

    Foda, Hossam M T

    2003-06-01

    The study presents a comprehensive statistical analysis of a series of 500 consecutive rhinoplasties of which 380 (76 per cent) were primary and 120 (24 per cent) were secondary cases. All cases were operated upon using the external rhinoplasty technique; simultaneous septal surgery was performed in 350 (70 per cent) of the cases. Deformities of the upper two-thirds of the nose that occurred significantly more in the secondary cases included; dorsal saddling, dorsal irregularities, valve collapse, open roof and pollybeak deformities. In the lower third of the nose; secondary cases showed significantly higher incidences of depressed tip, tip over-rotation, tip asymmetry, retracted columella, and alar notching. Suturing techniques were used significantly more in primary cases, while in secondary cases grafting techniques were used significantly more. The complications encountered intra-operatively included; septal flap tears (2.8 per cent) and alar cartilage injury (1.8 per cent), while post-operative complications included; nasal trauma (one per cent), epistaxis (two per cent), infection (2.4 per cent), prolonged oedema (17 per cent), and nasal obstruction (0.8 per cent). The overall patient satisfaction rate was 95.6 per cent and the transcolumellar scar was found to be unacceptable in only 0.8 per cent of the patients.

  10. Critical Consideration of the Methods of Calculation Used in the Evaluation of the Absorbed Dose to the Skin in Cases of External Contamination; Observations Critiques sur les Methodes de Calcul Utilisees pour l'Evaluacion de la Dose Absorbee par la Peau en Cas de Contamination Externe; 041a 0420 0414 ; Observaciones Criticas Sobre los Metodos de Calculo Utilizados para Evaluar la Dosis Absorbida por la Piel en Caso de Contaminacion Externa

    Energy Technology Data Exchange (ETDEWEB)

    Casnati, E.; Breuer, F. [Gruppo di Dosimetria e Standardizzazione, CNEN, Centro di Studi Nucleari della Casaccia, Rome (Italy)

    1965-06-15

    For the calculation of the dose the Loevinger formula for beta particles and an exponential formula for electromagnetic radiations are generally used. In the present paper, the authors intend to examine whether, in the case of cutaneous contamination, the use of an exponential formula (like the formula of Rossi and Ellis) and a similar formula might be considered as satisfactory for the calculation of the beta and the gamma dose, respectively. With this aim the results obtained for this particular type of geometry by the Rossi and Ellis formula and by the Loevinger formula for beta particles have been compared; for gamma radiation the values obtained by the exponential formula and by two more elaborated formulae, suitable for the conditions of maximum photonic build-up and of null photonic build-up, were compared. The results permit the conclusion that, for the geometric conditions considered, both the dose from beta radiation and that from gamma radiation to the basal layer of the skin can be determined by the exponential type formula: the results in fact are acceptable from a health protection standpoint. (author) [French] Pour le calcul de la dose, on utilise generalement la formule de Loevinger pour les particules beta et la formule exponentielle pour le rayonnement electromagnetique. Les auteurs se proposentde determiner si, dans le cas d'une contamination cutanee, l'emploi d'une formule exponentielle (comme la formule de Rossi et Ellis) et une formule analogue peut etre considere comme satisfaisant pour le calcul de la dose beta et de la dose gamma respectivement. Ils comparent pour cela les resultats obtenus pour ce type particulier de geometrie par la formule de Rossi et Ellis et par la formule de Loevinger pour les particules beta; pour les rayons gamma, ils comparent les valeurs obtenues par la formule exponentielle et par deux formules plus elaborees qui conviennent pour les conditions de formation maximum de photons et de formation nulle de photons. Les

  11. Quantum entanglement and criticality of the antiferromagnetic Heisenberg model in an external field

    International Nuclear Information System (INIS)

    Liu Guanghua; Li Ruoyan; Tian Guangshan

    2012-01-01

    By Lanczos exact diagonalization and the infinite time-evolving block decimation (iTEBD) technique, the two-site entanglement as well as the bipartite entanglement, the ground state energy, the nearest-neighbor correlations, and the magnetization in the antiferromagnetic Heisenberg (AFH) model under an external field are investigated. With increasing external field, the small size system shows some distinct upward magnetization stairsteps, accompanied synchronously with some downward two-site entanglement stairsteps. In the thermodynamic limit, the two-site entanglement, as well as the bipartite entanglement, the ground state energy, the nearest-neighbor correlations, and the magnetization are calculated, and the critical magnetic field h c = 2.0 is determined exactly. Our numerical results show that the quantum entanglement is sensitive to the subtle changing of the ground state, and can be used to describe the magnetization and quantum phase transition. Based on the discontinuous behavior of the first-order derivative of the entanglement entropy and fidelity per site, we think that the quantum phase transition in this model should belong to the second-order category. Furthermore, in the magnon existence region (h < 2.0), a logarithmically divergent behavior of block entanglement which can be described by a free bosonic field theory is observed, and the central charge c is determined to be 1. (paper)

  12. HADOC: a computer code for calculation of external and inhalation doses from acute radionuclide releases

    International Nuclear Information System (INIS)

    Strenge, D.L.; Peloquin, R.A.

    1981-04-01

    The computer code HADOC (Hanford Acute Dose Calculations) is described and instructions for its use are presented. The code calculates external dose from air submersion and inhalation doses following acute radionuclide releases. Atmospheric dispersion is calculated using the Hanford model with options to determine maximum conditions. Building wake effects and terrain variation may also be considered. Doses are calculated using dose conversion factor supplied in a data library. Doses are reported for one and fifty year dose commitment periods for the maximum individual and the regional population (within 50 miles). The fractional contribution to dose by radionuclide and exposure mode are also printed if requested

  13. Calculating the Effect of External Shading on the Solar Heat Gain Coefficient of Windows

    Energy Technology Data Exchange (ETDEWEB)

    Kohler, Christian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Shukla, Yash [CEPT Univ., Ahmedabad (India); Rawal, Rajan [CEPT Univ., Ahmedabad (India)

    2017-08-09

    Current prescriptive building codes have limited ways to account for the effect of solar shading, such as overhangs and awnings, on window solar heat gains. We propose two new indicators, the adjusted Solar Heat Gain Coefficient (aSHGC) which accounts for external shading while calculating the SHGC of a window, and a weighted SHGC (SHGCw) which provides a seasonal SHGC weighted by solar intensity. We demonstrate a method to calculate these indices using existing tools combined with additional calculations. The method is demonstrated by calculating the effect of an awning on a clear double glazing in New Delhi.

  14. Probabilistic approach to external cloud dose calculations using onsite meteorological data

    International Nuclear Information System (INIS)

    Strenge, D.L.; Watson, E.C.; Bander, T.J.; Kennedy, W.E.

    1976-01-01

    A method is described for calculation of external total body and skin doses from accidental atmospheric releases of radionuclides based on hourly onsite meteorological data. The method involves calculation of dose values from a finite size cloud for each hourly observation for a given radionuclide inventory. These values are then used to determine the probability of occurrence of dose levels for specified release times ranging from one hour to 30 days

  15. Criticality calculations in reactor accelerator coupling experiment (Race)

    International Nuclear Information System (INIS)

    Reda, M.A.; Spaulding, R.; Hunt, A.; Harmon, J.F.; Beller, D.E.

    2005-01-01

    A Reactor Accelerator Coupling Experiment (RACE) is to be performed at the Idaho State University Idaho Accelerator Center (IAC). The electron accelerator is used to generate neutrons by inducing Bremsstrahlung photon-neutron reactions in a Tungsten- Copper target. This accelerator/target system produces a source of ∼1012 n/s, which can initiate fission reactions in the subcritical system. This coupling experiment between a 40-MeV electron accelerator and a subcritical system will allow us to predict and measure coupling efficiency, reactivity, and multiplication. In this paper, the results of the criticality and multiplication calculations, which were carried out using the Monte Carlo radiation transport code MCNPX, for different coupling design options are presented. The fuel plate arrangements and the surrounding tank dimensions have been optimized. Criticality using graphite instead of water for reflector/moderator outside of the core region has been studied. The RACE configuration at the IAC will have a criticality (k-effective) of about 0,92 and a multiplication of about 10. (authors)

  16. Automatic fission source convergence criteria for Monte Carlo criticality calculations

    International Nuclear Information System (INIS)

    Shim, Hyung Jin; Kim, Chang Hyo

    2005-01-01

    The Monte Carlo criticality calculations for the multiplication factor and the power distribution in a nuclear system require knowledge of stationary or fundamental-mode fission source distribution (FSD) in the system. Because it is a priori unknown, so-called inactive cycle Monte Carlo (MC) runs are performed to determine it. The inactive cycle MC runs should be continued until the FSD converges to the stationary FSD. Obviously, if one stops them prematurely, the MC calculation results may have biases because the followup active cycles may be run with the non-stationary FSD. Conversely, if one performs the inactive cycle MC runs more than necessary, one is apt to waste computing time because inactive cycle MC runs are used to elicit the fundamental-mode FSD only. In the absence of suitable criteria for terminating the inactive cycle MC runs, one cannot but rely on empiricism in deciding how many inactive cycles one should conduct for a given problem. Depending on the problem, this may introduce biases into Monte Carlo estimates of the parameters one tries to calculate. The purpose of this paper is to present new fission source convergence criteria designed for the automatic termination of inactive cycle MC runs

  17. Validation of a scenario-based assessment of critical thinking using an externally validated tool.

    Science.gov (United States)

    Buur, Jennifer L; Schmidt, Peggy; Smylie, Dean; Irizarry, Kris; Crocker, Carlos; Tyler, John; Barr, Margaret

    2012-01-01

    With medical education transitioning from knowledge-based curricula to competency-based curricula, critical thinking skills have emerged as a major competency. While there are validated external instruments for assessing critical thinking, many educators have created their own custom assessments of critical thinking. However, the face validity of these assessments has not been challenged. The purpose of this study was to compare results from a custom assessment of critical thinking with the results from a validated external instrument of critical thinking. Students from the College of Veterinary Medicine at Western University of Health Sciences were administered a custom assessment of critical thinking (ACT) examination and the externally validated instrument, California Critical Thinking Skills Test (CCTST), in the spring of 2011. Total scores and sub-scores from each exam were analyzed for significant correlations using Pearson correlation coefficients. Significant correlations between ACT Blooms 2 and deductive reasoning and total ACT score and deductive reasoning were demonstrated with correlation coefficients of 0.24 and 0.22, respectively. No other statistically significant correlations were found. The lack of significant correlation between the two examinations illustrates the need in medical education to externally validate internal custom assessments. Ultimately, the development and validation of custom assessments of non-knowledge-based competencies will produce higher quality medical professionals.

  18. An integral nodal variational method for multigroup criticality calculations

    International Nuclear Information System (INIS)

    Lewis, E.E.; Tsoulfanidis, N.

    2003-01-01

    An integral formulation of the variational nodal method is presented and applied to a series of benchmark critically problems. The method combines an integral transport treatment of the even-parity flux within the spatial node with an odd-parity spherical harmonics expansion of the Lagrange multipliers at the node interfaces. The response matrices that result from this formulation are compatible with those in the VARIANT code at Argonne National Laboratory. Either homogeneous or heterogeneous nodes may be employed. In general, for calculations requiring higher-order angular approximations, the integral method yields solutions with comparable accuracy while requiring substantially less CPU time and memory than the standard spherical harmonics expansion using the same spatial approximations. (author)

  19. Isopiestic density law of actinide nitrates applied to criticality calculations

    International Nuclear Information System (INIS)

    Leclaire, Nicolas; Anno, Jacques; Courtois, Gerard; Poullot, Gilles; Rouyer, Veronique

    2003-01-01

    Up to now, criticality safety experts used density laws fitted on experimental data and applied them in and outside the measurement range. Depending on the case, such an approach could be wrong for nitrate solutions. Seven components are concerned: UO 2 (NO 3 ) 2 , U(NO 3 ) 4 , Pu(NO 3 ) 4 , Pu(NO 3 ) 3 , Th(NO 3 ) 4 , Am(NO 3 ) 3 and HNO 3 . To get rid of this problem, a new methodology based on the thermodynamic concept of binary electrolytes solutions mixtures at constant water activity, so called 'isopiestic' solutions, has been developed by IRSN to calculate the nitrate solutions density. This article shortly presents the theoretical aspects of the method, its qualification using benchmarks and its implementation in IRSN graphical user interface. (author)

  20. Benchmark test of JEF-1 evaluation by calculating fast criticalities

    International Nuclear Information System (INIS)

    Pelloni, S.

    1986-06-01

    JEF-1 basic evaluation was tested by calculating fast critical experiments using the cross section discrete-ordinates transport code ONEDANT with P/sub 3/S/sub 16/ approximation. In each computation a spherical one dimensional model was used, together with a 174 neutron group VITAMIN-E structured JEF-1 based nuclear data library, generated at EIR with NJOY and TRANSX-CTR. It is found that the JEF-1 evaluation gives accurate results comparable with ENDF/B-V and that eigenvalues agree well within 10 mk whereas reaction rates deviate by up to 10% from the experiment. U-233 total and fission cross sections seem to be underestimated in the JEF-1 evaluation in the fast energy range between 0.1 and 1 MeV. This confirms previous analysis based on diffusion theory with 71 neutron groups, performed by H. Takano and E. Sartori at NEA Data Bank. (author)

  1. Prediction of prostate cancer in unscreened men: external validation of a risk calculator.

    Science.gov (United States)

    van Vugt, Heidi A; Roobol, Monique J; Kranse, Ries; Määttänen, Liisa; Finne, Patrik; Hugosson, Jonas; Bangma, Chris H; Schröder, Fritz H; Steyerberg, Ewout W

    2011-04-01

    Prediction models need external validation to assess their value beyond the setting where the model was derived from. To assess the external validity of the European Randomized study of Screening for Prostate Cancer (ERSPC) risk calculator (www.prostatecancer-riskcalculator.com) for the probability of having a positive prostate biopsy (P(posb)). The ERSPC risk calculator was based on data of the initial screening round of the ERSPC section Rotterdam and validated in 1825 and 531 men biopsied at the initial screening round in the Finnish and Swedish sections of the ERSPC respectively. P(posb) was calculated using serum prostate specific antigen (PSA), outcome of digital rectal examination (DRE), transrectal ultrasound and ultrasound assessed prostate volume. The external validity was assessed for the presence of cancer at biopsy by calibration (agreement between observed and predicted outcomes), discrimination (separation of those with and without cancer), and decision curves (for clinical usefulness). Prostate cancer was detected in 469 men (26%) of the Finnish cohort and in 124 men (23%) of the Swedish cohort. Systematic miscalibration was present in both cohorts (mean predicted probability 34% versus 26% observed, and 29% versus 23% observed, both pscreened men, but overestimated the risk of a positive biopsy. Further research is necessary to assess the performance and applicability of the ERSPC risk calculator when a clinical setting is considered rather than a screening setting. Copyright © 2010 Elsevier Ltd. All rights reserved.

  2. External dose-rate conversion factors for calculation of dose to the public

    Energy Technology Data Exchange (ETDEWEB)

    1988-07-01

    This report presents a tabulation of dose-rate conversion factors for external exposure to photons and electrons emitted by radionuclides in the environment. This report was prepared in conjunction with criteria for limiting dose equivalents to members of the public from operations of the US Department of Energy (DOE). The dose-rate conversion factors are provided for use by the DOE and its contractors in performing calculations of external dose equivalents to members of the public. The dose-rate conversion factors for external exposure to photons and electrons presented in this report are based on a methodology developed at Oak Ridge National Laboratory. However, some adjustments of the previously documented methodology have been made in obtaining the dose-rate conversion factors in this report. 42 refs., 1 fig., 4 tabs.

  3. Evaluation of Accuracy of Calculational Prediction of Criticality Based on ICSBEP Handbook Experiments

    International Nuclear Information System (INIS)

    Golovko, Yury; Rozhikhin, Yevgeniy; Tsibulya, Anatoly; Koscheev, Vladimir

    2008-01-01

    Experiments with plutonium, low enriched uranium and uranium-233 from the ICSBEP Handbook are being considered in this paper. Among these experiments it was selected only those, which seem to be the most relevant to the evaluation of uncertainty of critical mass of mixtures of plutonium or low enriched uranium or uranium-233 with light water. All selected experiments were examined and covariance matrices of criticality uncertainties were developed along with some uncertainties were revised. Statistical analysis of these experiments was performed and some contradictions were discovered and eliminated. Evaluation of accuracy of prediction of criticality calculations was performed using the internally consistent set of experiments with plutonium, low enriched uranium and uranium-233 remained after the statistical analyses. The application objects for the evaluation of calculational prediction of criticality were water-reflected spherical systems of homogeneous aqueous mixtures of plutonium or low enriched uranium or uranium-233 of different concentrations which are simplified models of apparatus of external fuel cycle. It is shows that the procedure allows to considerably reduce uncertainty in k eff caused by the uncertainties in neutron cross-sections. Also it is shows that the results are practically independent of initial covariance matrices of nuclear data uncertainties. (authors)

  4. Criticality Analysis Of TCA Critical Lattices With MNCP-4C Monte Carlo Calculation

    International Nuclear Information System (INIS)

    Zuhair

    2002-01-01

    The use of uranium-plutonium mixed oxide (MOX) fuel in electric generation light water reactor (PWR, BWR) is being planned in Japan. Therefore, the accuracy evaluations of neutronic analysis code for MOX cores have been employed by many scientists and reactor physicists. Benchmark evaluations for TCA was done using various calculation methods. The Monte Carlo become the most reliable method to predict criticality of various reactor types. In this analysis, the MCNP-4C code was chosen because various superiorities the code has. All in all, the MCNP-4C calculation for TCA core with 38 MOX critical lattice configurations gave the results with high accuracy. The JENDL-3.2 library showed significantly closer results to the ENDF/B-V. The k eff values calculated with the ENDF/B-VI library gave underestimated results. The ENDF/B-V library gave the best estimation. It can be concluded that MCNP-4C calculation, especially with ENDF/B-V and JENDL-3.2 libraries, for MOX fuel utilized NPP design in reactor core is the best choice

  5. The sensitivity of calculated doses to critical assumptions for the offsite consequences of nuclear power reactor accidents

    International Nuclear Information System (INIS)

    Moeller, M.P.; Scherpelz, R.I.; Desrosiers, A.E.

    1982-01-01

    This work analyzes the sensitivity of calculated doses to critical assumptions for offsite consequences following a PWR-2 accident at a nuclear power reactor. The calculations include three radiation dose pathways: internal dose resulting from inhalation, external doses from exposure to the plume, and external doses from exposure to contaminated ground. The critical parameters are the time period of integration for internal dose commitment and the duration of residence on contaminated ground. The data indicate the calculated offsite whole body dose will vary by as much as 600% depending upon the parameters assumed. When offsite radiation doses determine the size of emergency planning zones, this uncertainty has significant effect upon the resources allocated to emergency preparedness

  6. Calculation of external-internal flow fields for mixed-compression inlets

    Science.gov (United States)

    Chyu, W. J.; Kawamura, T.; Bencze, D. P.

    1987-01-01

    Supersonic inlet flows with mixed external-internal compressions were computed using a combined implicit-explicit (Beam-Warming-Steger/MacCormack) method for solving the three-dimensional unsteady, compressible Navier-Stokes equations in conservation form. Numerical calculations were made of various flows related to such inlet operations as the shock-wave intersections, subsonic spillage around the cowl lip, and inlet started versus unstarted conditions. Some of the computed results were compared with wind tunnel data.

  7. On the contribution of external cost calculations to energy system governance: The case of a potential large-scale nuclear accident

    International Nuclear Information System (INIS)

    Laes, Erik; Meskens, Gaston; Sluijs, Jeroen P. van der

    2011-01-01

    The contribution of nuclear power to a sustainable energy future is a contested issue. This paper presents a critical review of an attempt to objectify this debate through the calculation of the external costs of a potential large-scale nuclear accident in the ExternE project. A careful dissection of the ExternE approach resulted in a list of 30 calculation steps and assumptions, from which the 6 most contentious ones were selected through a stakeholder internet survey. The policy robustness and relevance of these key assumptions were then assessed in a workshop using the concept of a 'pedigree of knowledge'. Overall, the workshop outcomes revealed the stakeholder and expert panel's scepticism about the assumptions made: generally these were considered not very plausible, subjected to disagreement, and to a large extent inspired by contextual factors. Such criticism indicates a limited validity and useability of the calculated nuclear accident externality as a trustworthy sustainability indicator. Furthermore, it is our contention that the ExternE project could benefit greatly - in terms of gaining public trust - from employing highly visible procedures of extended peer review such as the pedigree assessment applied to our specific case of the external costs of a potential large-scale nuclear accident. - Highlights: → Six most contentious assumptions were selected through a stakeholder internet survey. → Policy robustness of these assumptions was assessed in a pedigree assessment workshop. → Assumptions were considered implausible, controversial, and inspired by contextual factors. → This indicates a limited validity and useability as a trustworthy sustainability indicator.

  8. Nuclear criticality safety calculational analysis for small-diameter containers

    International Nuclear Information System (INIS)

    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

  9. SUBDOSA: a computer program for calculating external doses from accidental atmospheric releases of radionuclides

    International Nuclear Information System (INIS)

    Strenge, D.L.; Watson, E.C.; Houston, J.R.

    1975-06-01

    A computer program, SUBDOSA, was developed for calculating external γ and β doses to individuals from the accidental release of radionuclides to the atmosphere. Characteristics of SUBDOSA are: doses from both γ and β radiation are calculated as a function of depth in tissue, summed and reported as skin, eye, gonadal, and total body dose; doses are calculated for releases within each of several release time intervals and nuclide inventories and atmospheric dispersion conditions are considered for each time interval; radioactive decay is considered during the release and/or transit using a chain decay scheme with branching to account for transitions to and from isomeric states; the dose from gamma radiation is calculated using a numerical integration technique to account for the finite size of the plume; and the program computes and lists the normalized air concentrations at ground level as a function of distance from the point of release. (auth)

  10. Calculation of age-dependent effective doses for external exposure using the MCNP code

    International Nuclear Information System (INIS)

    Hung, Tran Van

    2013-01-01

    Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)

  11. Calculation of age-dependent effective doses for external exposure using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Hung, Tran Van [Research and Development Center for Radiation Technology, ThuDuc, HoChiMinh City (VT)

    2013-07-15

    Age-dependent effective dose for external exposure to photons uniformly distributed in air were calculated. Firstly, organ doses were calculated with a series of age-specific MIRD-5 type phantoms using the Monte Carlo code MCNP. The calculations were performed for mono-energetic photon sources with source energies from 10 keV to 5 MeV and for phantoms of newborn, 1, 5, 10, and 15 years-old and adult. Then, the effective doses to the different age-phantoms from the mono-energetic photon sources were estimated based on the obtained organ doses. From the calculated results, it is shown that the effective doses depend on the body size; the effective doses in younger phantoms are higher than those in the older phantoms, especially below 100 keV. (orig.)

  12. Improved estimation of the variance in Monte Carlo criticality calculations

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard

    2008-01-01

    Results for the effective multiplication factor in a Monte Carlo criticality calculations are often obtained from averages over a number of cycles or batches after convergence of the fission source distribution to the fundamental mode. Then the standard deviation of the effective multiplication factor is also obtained from the k eff results over these cycles. As the number of cycles will be rather small, the estimate of the variance or standard deviation in k eff will not be very reliable, certainly not for the first few cycles after source convergence. In this paper the statistics for k eff are based on the generation of new fission neutron weights during each history in a cycle. It is shown that this gives much more reliable results for the standard deviation even after a small number of cycles. Also attention is paid to the variance of the variance (VoV) and the standard deviation of the standard deviation. A derivation is given how to obtain an unbiased estimate for the VoV, even for a small number of samples. (authors)

  13. Improved estimation of the variance in Monte Carlo criticality calculations

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J. Eduard [Delft University of Technology, Delft (Netherlands)

    2008-07-01

    Results for the effective multiplication factor in a Monte Carlo criticality calculations are often obtained from averages over a number of cycles or batches after convergence of the fission source distribution to the fundamental mode. Then the standard deviation of the effective multiplication factor is also obtained from the k{sub eff} results over these cycles. As the number of cycles will be rather small, the estimate of the variance or standard deviation in k{sub eff} will not be very reliable, certainly not for the first few cycles after source convergence. In this paper the statistics for k{sub eff} are based on the generation of new fission neutron weights during each history in a cycle. It is shown that this gives much more reliable results for the standard deviation even after a small number of cycles. Also attention is paid to the variance of the variance (VoV) and the standard deviation of the standard deviation. A derivation is given how to obtain an unbiased estimate for the VoV, even for a small number of samples. (authors)

  14. Critical current degradation in superconducting niobium-titanium alloys in external magnetic fields under loading

    International Nuclear Information System (INIS)

    Bojko, V.S.; Lazareva, M.B.; Starodubov, Ya.D.; Chernyj, O.V.; Gorbatenko, V.M.

    1992-01-01

    The effect of external magnetic fields on the stress at which the critical current starts to degrade (the degradation threshold σ 0 e ) under mechanical loads in superconducting Nb-Ti alloys is studied and a possible mechanism of realization of the effect observed is proposed.It is assumed that additional stresses on the transformation dislocation from the external magnetic fields are beneficial for the growth of martensite inclusions whose superconducting parameters (critical current density j k and critical temperature T k ) are lower then those in the initial material.The degradation threshold is studied experimentally in external magnetic fields H up to 7 T.The linear dependence σ 0 e (H) is observed.It is shown that external magnetic fields play an important role in the critical current degradation at the starting stages of deformation.This fact supports the assumption that the degradation of superconducting parameters under loading are due to the phenomenon of superelasticity,i.e. a reversible load-induced change in the martensite inclusions sizes rather than the reversible mechanical twinning.The results obtained are thought to be important to estimating superconducting solenoid stability in a wide range of magnetic fields

  15. The Relations among Maternal Depression, Maternal Criticism, and Adolescents' Externalizing and Internalizing Symptoms.

    Science.gov (United States)

    Frye, Alice A.; Garber, Judy

    2005-01-01

    This study examined the relations between maternal criticism and externalizing and internalizing symptoms in adolescents who varied in their risk for psychopathology. Both maternal-effects and child-effects models were examined. The sample consisted of 194 adolescents (mean age = 11.8 years) and their mothers: 146 mothers had a history of…

  16. A method for including external feed in depletion calculations with CRAM and implementation into ORIGEN

    International Nuclear Information System (INIS)

    Isotalo, A.E.; Wieselquist, W.A.

    2015-01-01

    Highlights: • A method for handling external feed in depletion calculations with CRAM. • Source term can have polynomial or exponentially decaying time-dependence. • CRAM with source term and adjoint capability implemented to ORIGEN in SCALE. • The new solver is faster and more accurate than the original solver of ORIGEN. - Abstract: A method for including external feed with polynomial time dependence in depletion calculations with the Chebyshev Rational Approximation Method (CRAM) is presented and the implementation of CRAM to the ORIGEN module of the SCALE suite is described. In addition to being able to handle time-dependent feed rates, the new solver also adds the capability to perform adjoint calculations. Results obtained with the new CRAM solver and the original depletion solver of ORIGEN are compared to high precision reference calculations, which shows the new solver to be orders of magnitude more accurate. Furthermore, in most cases, the new solver is up to several times faster due to not requiring similar substepping as the original one

  17. Neoclassical parallel flow calculation in the presence of external parallel momentum sources in Heliotron J

    Energy Technology Data Exchange (ETDEWEB)

    Nishioka, K.; Nakamura, Y. [Graduate School of Energy Science, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Nishimura, S. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki, Gifu 509-5292 (Japan); Lee, H. Y. [Korea Advanced Institute of Science and Technology, Daejeon 305-701 (Korea, Republic of); Kobayashi, S.; Mizuuchi, T.; Nagasaki, K.; Okada, H.; Minami, T.; Kado, S.; Yamamoto, S.; Ohshima, S.; Konoshima, S.; Sano, F. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-03-15

    A moment approach to calculate neoclassical transport in non-axisymmetric torus plasmas composed of multiple ion species is extended to include the external parallel momentum sources due to unbalanced tangential neutral beam injections (NBIs). The momentum sources that are included in the parallel momentum balance are calculated from the collision operators of background particles with fast ions. This method is applied for the clarification of the physical mechanism of the neoclassical parallel ion flows and the multi-ion species effect on them in Heliotron J NBI plasmas. It is found that parallel ion flow can be determined by the balance between the parallel viscosity and the external momentum source in the region where the external source is much larger than the thermodynamic force driven source in the collisional plasmas. This is because the friction between C{sup 6+} and D{sup +} prevents a large difference between C{sup 6+} and D{sup +} flow velocities in such plasmas. The C{sup 6+} flow velocities, which are measured by the charge exchange recombination spectroscopy system, are numerically evaluated with this method. It is shown that the experimentally measured C{sup 6+} impurity flow velocities do not contradict clearly with the neoclassical estimations, and the dependence of parallel flow velocities on the magnetic field ripples is consistent in both results.

  18. Calculation of conversion coefficients for radiological protection against external radiation exposures

    International Nuclear Information System (INIS)

    Zankl, M.

    2001-01-01

    Calculations are essential for radiation protection practice because organ doses and effective doses cannot be measured directly. Conversion coefficients describe the numerical relationships of protection quantities and operational quantities. The latter can be measured in practical situations using suitable dosimeters. The conversion coefficients are calculated using radiation transport codes - usually based on Monte Carlo methods - that simulate the interactions of radiation with matter in computational models of the human body. A new generation of human body models, the so-called voxel models, are constructed from image data of real persons using suitable image processing systems, consequently, they represent the human anatomy more realistically than the so-called mathematical models. The numerical effects of realistic body anatomy on the calculated conversion coefficients can amount to 70% and more for external exposures. (orig.) [de

  19. Subcriticality calculations for the FFTF reverse approach to critical experiment

    International Nuclear Information System (INIS)

    Selby, D.L.; Flanagan, G.F.

    1975-01-01

    The reverse approach to critical (RAC) experiments were performed in the ZPR-IX critical facility at Argonne National Laboratory. One of the major objectives of this project is to determine the adequacy of the low-level flux monitor (LLFM) detectors for initial loading of the Fast Flux Test Facility (FFTF). 5 references

  20. Calculating external doses from contaminated soil with the computer model SOILD

    International Nuclear Information System (INIS)

    Chen, Y.; LePoire, D.; Yu, C.

    1991-01-01

    The SOILD computer model was developed for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil. It is designed to assess external doses under various exposure scenarios that may be encountered in environmental restoration programs. The model's four major functional features address (a) dose versus source depth in soil, (b) shielding of clean cover soil, (c) area of contamination, and (d) nonuniform distribution of sources. The model can also adjust doses when there are variations in soil densities for both source and cover soils. It is supported by a data base of ∼500 radionuclides. A sample calculation was performed by SOILD to determine the effective dose equivalent for a uniform source distribution in soil. The soil density was assumed to be 1.6 g/cm 3 , and the source strength was assumed to be 1 pCi/cm 3 . The following radionuclides were studied: 60 C, 131 I, 137+D Cs, 238+D U, and 226+D Ra ('+D' denotes the parent nuclide and daughters)

  1. Neutron transport calculations of some fast critical assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Val Penalosa, J A

    1976-07-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs.

  2. Neutron transport calculations of some fast critical assemblies

    International Nuclear Information System (INIS)

    Martinez-Val Penalosa, J. A.

    1976-01-01

    To analyse the influence of the input variables of the transport codes upon the neutronic results (eigenvalues, generation times, . . . ) four Benchmark calculations have been performed. Sensitivity analysis have been applied to express these dependences in a useful way, and also to get an unavoidable experience to carry out calculations achieving the required accuracy and doing them in practical computing times. (Author) 29 refs

  3. Merger of Nuclear Data with Criticality Safety Calculations

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Feasibility study on heterogeneous method in criticality calculations

    International Nuclear Information System (INIS)

    Prati, A.

    1977-01-01

    The criticality of finite heterogeneous assemblies is analysed by the heterogeneous methods employing the Eigen-function analysis. The moderation is treated by the Fermi age theory. The system is analysed in two dimensional rectangular coordinates. The criticality and the fluxes are determined for systems with small and large number of fuel rods. The convergence and the residual error in the modal analysis are discussed. (author)

  5. Merger of Nuclear Data with Criticality Safety Calculations

    International Nuclear Information System (INIS)

    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

  6. Validation the methodology calculate critical position of control rods to the critical facility IPEN/MB-01

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Rodriguez Gual, R.

    1998-01-01

    Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used

  7. An analysis of critical heat flux on the external surface of the reactor vessel lower head

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Baek, Won Pil; Chang, Soon Heung

    1999-01-01

    CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions

  8. Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

    International Nuclear Information System (INIS)

    Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

    1991-01-01

    Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

  9. Calculating the critical temperature for Coleman-Weinberg GUTS

    International Nuclear Information System (INIS)

    Easther, R.; Moreau, W.

    1992-01-01

    We study the finite-temperature effective potential of the Higgs scalar in GUTs with Coleman-Weinberg symmetry breaking. The critical temperature is derived without employing a high-temperature approximation to the effective potential, and the limitations of such approximations are discussed. (author)

  10. Calculation of dose-rate conversion factors for external exposure to photons and electrons

    International Nuclear Information System (INIS)

    Kocher, D.C.

    1978-01-01

    Methods are presented for the calculation of dose-rate conversion factors for external exposure to photon and electron radiation from radioactive decay. A dose-rate conversion factor is defined as the dose-equivalent rate per unit radionuclide concentration. Exposure modes considered are immersion in contaminated air, immersion in contaminated water, and irradiation from a contaminated ground surface. For each radiation type and exposure mode, dose-rate conversion factors are derived for tissue-equivalent material at the body surface of an exposed individual. In addition, photon dose-rate conversion factors are estimated for 22 body organs. The calculations are based on the assumption that the exposure medium is infinite in extent and that the radionuclide concentration is uniform. The dose-rate conversion factors for immersion in contaminated air and water then follow from the requirement that all of the energy emitted in the radioactive decay is absorbed in the infinite medium. Dose-rate conversion factors for ground-surface exposure are calculated at a reference location above a smooth, infinite plane using the point-kernel integration method and known specific absorbed fractions for photons and electrons in air

  11. Dose conversion coefficients calculated using a series of adult Japanese voxel phantoms against external photon exposure

    International Nuclear Information System (INIS)

    Sato, Kaoru; Endo, Akira; Saito, Kimiaki

    2008-10-01

    This report presents a complete set of conversion coefficients of organ doses and effective doses calculated for external photon exposure using five Japanese adult voxel phantoms developed at the Japan Atomic Energy Agency (JAEA). At the JAEA, high-resolution Japanese voxel phantoms have been developed to clarify the variation of organ doses due to the anatomical characteristics of Japanese, and three male phantoms (JM, JM2 and Otoko) and two female phantoms (JF and Onago) have been constructed up to now. The conversion coefficients of organ doses and effective doses for the five voxel phantoms have been calculated for six kinds of idealized irradiation geometries from monoenergetic photons ranging from 0.01 to 10 MeV using EGS4, a Monte Carlo code for the simulation of coupled electron-photon transport. The dose conversion coefficients are given as absorbed dose and effective dose per unit air-kerma free-in-air, and are presented in tables and figures. The calculated dose conversion coefficients are compared with those of voxel phantoms based on the Caucasian and the recommended values in ICRP74 in order to discuss (1) variation of organ dose due to the body size and individual anatomy, such as position and shape of organs, and (2) effect of posture on organ doses. The present report provides valuable data to study the influence of the body characteristics of Japanese upon the organ doses and to discuss developing reference Japanese and Asian phantoms. (author)

  12. Criticality calculation of the nuclear material warehouse of the ININ

    International Nuclear Information System (INIS)

    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)

  13. Calculation notes that support accident scenario and consequence determination of a waste tank criticality

    International Nuclear Information System (INIS)

    Marusich, R.M. Westinghouse Hanford

    1996-01-01

    The purpose of this calculation note is to provide the basis for criticality consequences for the Tank Farm Safety Analysis Report (FSAR). Criticality scenario is developed and details and description of the analysis methods are provided

  14. Data for absorbed dose calculations for external sources and for emitters within the body

    International Nuclear Information System (INIS)

    Hep, J.; Valenta, V.

    1976-01-01

    Tables give data for the calculation of absorbed doses from radioactivity sources accumulated in individual body organs. The tables are arranged in such manner that the gamma energy (J) absorbed in 1 kg of target organ (19 organs and total body) are given for 18 source organs (16 different organs, total doby and surrounding air) resulting from 1 decay event, this for more than 250 radioisotopes evenly distributed in the source organ (1 J/kg=100 rad). Also given are the energies of alpha and beta radiations related to one decay. In tables having the surrounding air as the source it is assumed that the intensity of the external source is 1 decay per 1 m 3 of surrounding air which is constant in the entire half-space. The tables are only elaborated for radioisotopes with a half-life of more than 1 min. (B.S.)

  15. Cleanup techniques for Finnish urban environments and external doses from 137Cs - modelling and calculations

    International Nuclear Information System (INIS)

    Moring, M.; Markkula, M.L.

    1997-03-01

    The external doses under various radioactive deposition conditions are assessed and the efficiencies of some simple decontamination techniques (grass cutting, vacuum sweeping, hosing of paved surfaces and roofs, and felling trees) are compared in the study. The present model has been constructed for the Finnish conditions and housing areas, using 137 Cs transfer data from the Nordic and Central European studies and models. The compartment model concerns behaviour and decontamination of 137 Cs in the urban environment under summer conditions. Doses to man have been calculated for wet (light rain) and dry deposition in four typical Finnish building areas: single-family wooden houses, brick terraced-houses, blocks of flats and urban office buildings. (26 refs.)

  16. CRISTAL V1: Criticality package for burn up credit calculations

    International Nuclear Information System (INIS)

    Gomit, Jean-Michel; Cousinou, Patrick; Gantenbein, Francoise; Diop, Cheikh; Fernandez de Grado, Guy; Mijuin, Dominique; Grouiller, Jean-Paul; Marc, Andre; Toubon, Herve

    2003-01-01

    The first version of the CRISTAL package, created and validated as part of a joint project between IRSN, COGEMA and CEA, was delivered to users in November 1999. This fruitful cooperation between IRSN, COGEMA and CEA has been pursued until 2003 with the development and the validation of the package CRISTAL V1, whose main objectives are to improve the criticality safety studies including the Burn up Credit effect. (author)

  17. Use of deterministic methods in survey calculations for criticality problems

    International Nuclear Information System (INIS)

    Hutton, J.L.; Phenix, J.; Course, A.F.

    1991-01-01

    A code package using deterministic methods for solving the Boltzmann Transport equation is the WIMS suite. This has been very successful for a range of situations. In particular it has been used with great success to analyse trends in reactivity with a range of changes in state. The WIMS suite of codes have a range of methods and are very flexible in the way they can be combined. A wide variety of situations can be modelled ranging through all the current Thermal Reactor variants to storage systems and items of chemical plant. These methods have recently been enhanced by the introduction of the CACTUS method. This is based on a characteristics technique for solving the Transport equation and has the advantage that complex geometrical situations can be treated. In this paper the basis of the method is outlined and examples of its use are illustrated. In parallel with these developments the validation for out of pile situations has been extended to include experiments with relevance to criticality situations. The paper will summarise this evidence and show how these results point to a partial re-adoption of deterministic methods for some areas of criticality. The paper also presents results to illustrate the use of WIMS in criticality situations and in particular show how it can complement codes such as MONK when used for surveying the reactivity effect due to changes in geometry or materials. (Author)

  18. Monte Carlo criticality calculations accelerated by a growing neutron population

    International Nuclear Information System (INIS)

    Dufek, Jan; Tuttelberg, Kaur

    2016-01-01

    Highlights: • Efficiency is significantly improved when population size grows over cycles. • The bias in the fission source is balanced to other errors in the source. • The bias in the fission source decays over the cycle as the population grows. - Abstract: We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. As the efficiency of Monte Carlo criticality simulations is sensitive to the selected neutron population size, the method attempts to achieve the acceleration via on-the-fly control of the neutron population size. The neutron population size is gradually increased over successive criticality cycles so that the fission source bias amounts to a specific fraction of the total error in the cumulative fission source. An optimal setting then gives a reasonably small neutron population size, allowing for an efficient source iteration; at the same time the neutron population size is chosen large enough to ensure a sufficiently small source bias, such that does not limit accuracy of the simulation.

  19. A critical scoping review of external uses of comfrey (Symphytum spp.).

    Science.gov (United States)

    Frost, R; MacPherson, H; O'Meara, S

    2013-12-01

    External preparations of the herb comfrey (most commonly Symphytum officinale L.) are widely available for over-the-counter, practitioner and healthcare professional usage. Traditional practice suggests comfrey can be used to treat musculoskeletal disorders, wounds and various other conditions; however a full and critical coverage of the evidence base has not yet been undertaken. A critical scoping review was undertaken. Six bibliographic databases, 10 grey literature databases and nine trials registers were searched plus reference lists of included studies and a descriptive overview of comfrey. Randomised or non-randomised clinical trials assessing the external use of comfrey for any indication were included and methodological and reporting quality were assessed. Observational studies were included only in the assessment of adverse events. Studies were grouped and summarised according to the type of indication treated. Of 1348 identified records, 64 full texts were screened for inclusion and 26 were included in the review - 13 RCTs, 5 non-randomised controlled trials and 8 observational studies evaluating treatments for ankle distortion, back pain, abrasion wounds, venous leg ulcers and osteoarthritis. The majority of included trials had an overall unclear risk of bias due to poor quality of reporting. Few adverse events were reported. Individual clinical trials showed evidence of benefit for ankle distortion, back pain, abrasion wounds and osteoarthritis. Topical application appears to be safe but further rigorous assessment is needed. Systematic reviews focussing on particular indications may clarify the treatment effect and safety of external comfrey preparations. Copyright © 2013 Elsevier Ltd. All rights reserved.

  20. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    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.)

  1. Influence of chronic internal and acute external irradiations on the critical tissues of plants

    International Nuclear Information System (INIS)

    Kostyuk, O.P.; Ryasnenko, N.A.; Grodzins'kij, D.M.

    1998-01-01

    Peculiarities of chronic internal and acute external irradiations of the critical (as for irradiation influence) plants part, meristem, are studied. In particular, the investigation has aimed to evaluate the level of doses, accumulated by plant tissues, of the chronic internal irradiation from radiocaesium incorporated by them, and to compare its possible effect to one caused by the acute external irradiation. It is shown that the effects of both chronic and acute irradiations have similar features, and it is assumed that they have the very same mechanisms. We think that such a parameter of the plant ability to accumulate radiocaesium as the ratio of its content in a root tip and in the whole root system is a very sensible and useful criterion to estimate the irradiation influence on plants

  2. Quantum mechanical cluster calculations of critical scintillation processes

    International Nuclear Information System (INIS)

    Derenzo, Stephen E.; Klintenberg, Mattias K.; Weber, Marvin J.

    2000-01-01

    This paper describes the use of commercial quantum chemistry codes to simulate several critical scintillation processes. The crystal is modeled as a cluster of typically 50 atoms embedded in an array of typically 5,000 point charges designed to reproduce the electrostatic field of the infinite crystal. The Schrodinger equation is solved for the ground, ionized, and excited states of the system to determine the energy and electron wave function. Computational methods for the following critical processes are described: (1) the formation and diffusion of relaxed holes, (2) the formation of excitons, (3) the trapping of electrons and holes by activator atoms, (4) the excitation of activator atoms, and (5) thermal quenching. Examples include hole diffusion in CsI, the exciton in CsI, the excited state of CsI:Tl, the energy barrier for the diffusion of relaxed holes in CaF2 and PbF2, and prompt hole trapping by activator atoms in CaF2:Eu and CdS:Te leading to an ultra-fast (<50ps) scintillation rise time.

  3. Validation of KENO-based criticality calculations at Rocky Flats

    International Nuclear Information System (INIS)

    Felsher, P.D.; McKamy, J.N.; Monahan, S.P.

    1992-01-01

    In the absence of experimental data, it is necessary to rely on computer-based computational methods in evaluating the criticality condition of a nuclear system. The validity of the computer codes is established in a two-part procedure as outlined in ANSI/ANS 8.1. The first step, usually the responsibility of the code developer, involves verification that the algorithmic structure of the code is performing the intended mathematical operations correctly. The second step involves an assessment of the code's ability to realistically portray the governing physical processes in question. This is accomplished by determining the code's bias, or systematic error, through a comparison of computational results to accepted values obtained experimentally. In this paper, the authors discuss the validation process for KENO and the Hansen-Roach cross sections in use at EG and G Rocky Flats. The validation process at Rocky Flats consists of both global and local techniques. The global validation resulted in a maximum k eff limit of 0.95 for the limiting-accident scanarios of a criticality evaluation

  4. Critical evaluation of German regulatory specifications for calculating radiological exposure

    Energy Technology Data Exchange (ETDEWEB)

    Koenig, Claudia; Walther, Clemens [Hannover Univ. (Germany). Inst. of Radioecology; Smeddinck, Ulrich [Technische Univ. Braunschweig (Germany). Inst. of Law

    2015-07-01

    The assessment of radiological exposure of the public is an issue at the interface between scientific findings, juridical standard setting and political decision. The present work revisits the German regulatory specifications for calculating radiological exposure, like the already existing calculation model General Administrative Provision (AVV) for planning and monitoring nuclear facilities. We address the calculation models for the recent risk assessment regarding the final disposal of radioactive waste in Germany. To do so, a two-pronged approach is pursued. One part deals with radiological examinations of the groundwater-soil-transfer path of radionuclides into the biosphere. Processes at the so-called geosphere-biosphere-interface are examined, especially migration of I-129 in the unsaturated zone. This is necessary, since the German General Administrative Provision does not consider radionuclide transport via groundwater from an underground disposal facility yet. Especially data with regard to processes in the vadose zone are scarce. Therefore, using I-125 as a tracer, immobilization and mobilization of iodine is investigated in two reference soils from the German Federal Environment Agency. The second part of this study examines how scientific findings but also measures and activities of stakeholders and concerned parties influence juridical standard setting, which is necessary for risk management. Risk assessment, which is a scientific task, includes identification and investigation of relevant sources of radiation, possible pathways to humans, and maximum extent and duration of exposure based on dose-response functions. Risk characterization identifies probability and severity of health effects. These findings have to be communicated to authorities, who have to deal with the risk management. Risk management includes, for instance, taking into account acceptability of the risk, actions to reduce, mitigate, substitute or monitor the hazard, the setting of

  5. Quality plan for criticality safety calculations at Rocky Flats

    International Nuclear Information System (INIS)

    Pecora, D.

    1978-01-01

    The text of the plan is given, and some of the guidelines followed in writing it are discussed to aid others who may be faced with the same task. The plan is divided into four sections. The Introduction describes the general functions and purpose of the calculational program. The second section, Activities and Responsibilities, lists specific tasks and their purposes and assigns responsibility for performance. The third section references relevant documentation (e.g., ANSI standards), and the final section describes quality plans for specific functions

  6. Research on GPU acceleration for Monte Carlo criticality calculation

    International Nuclear Information System (INIS)

    Xu, Q.; Yu, G.; Wang, K.

    2013-01-01

    The Monte Carlo (MC) neutron transport method can be naturally parallelized by multi-core architectures due to the dependency between particles during the simulation. The GPU+CPU heterogeneous parallel mode has become an increasingly popular way of parallelism in the field of scientific supercomputing. Thus, this work focuses on the GPU acceleration method for the Monte Carlo criticality simulation, as well as the computational efficiency that GPUs can bring. The 'neutron transport step' is introduced to increase the GPU thread occupancy. In order to test the sensitivity of the MC code's complexity, a 1D one-group code and a 3D multi-group general purpose code are respectively transplanted to GPUs, and the acceleration effects are compared. The result of numerical experiments shows considerable acceleration effect of the 'neutron transport step' strategy. However, the performance comparison between the 1D code and the 3D code indicates the poor scalability of MC codes on GPUs. (authors)

  7. Multi-level iteration optimization for diffusive critical calculation

    International Nuclear Information System (INIS)

    Li Yunzhao; Wu Hongchun; Cao Liangzhi; Zheng Youqi

    2013-01-01

    In nuclear reactor core neutron diffusion calculation, there are usually at least three levels of iterations, namely the fission source iteration, the multi-group scattering source iteration and the within-group iteration. Unnecessary calculations occur if the inner iterations are converged extremely tight. But the convergence of the outer iteration may be affected if the inner ones are converged insufficiently tight. Thus, a common scheme suit for most of the problems was proposed in this work to automatically find the optimized settings. The basic idea is to optimize the relative error tolerance of the inner iteration based on the corresponding convergence rate of the outer iteration. Numerical results of a typical thermal neutron reactor core problem and a fast neutron reactor core problem demonstrate the effectiveness of this algorithm in the variational nodal method code NODAL with the Gauss-Seidel left preconditioned multi-group GMRES algorithm. The multi-level iteration optimization scheme reduces the number of multi-group and within-group iterations respectively by a factor of about 1-2 and 5-21. (authors)

  8. European Randomized Study of Screening for Prostate Cancer Risk Calculator: External Validation, Variability, and Clinical Significance.

    Science.gov (United States)

    Gómez-Gómez, Enrique; Carrasco-Valiente, Julia; Blanca-Pedregosa, Ana; Barco-Sánchez, Beatriz; Fernandez-Rueda, Jose Luis; Molina-Abril, Helena; Valero-Rosa, Jose; Font-Ugalde, Pilar; Requena-Tapia, Maria José

    2017-04-01

    To externally validate the European Randomized Study of Screening for Prostate Cancer (ERSPC) risk calculator (RC) and to evaluate its variability between 2 consecutive prostate-specific antigen (PSA) values. We prospectively catalogued 1021 consecutive patients before prostate biopsy for suspicion of prostate cancer (PCa). The risk of PCa and significant PCa (Gleason score ≥7) from 749 patients was calculated according to ERSPC-RC (digital rectal examination-based version 3 of 4) for 2 consecutive PSA tests per patient. The calculators' predictions were analyzed using calibration plots and the area under the receiver operating characteristic curve (area under the curve). Cohen kappa coefficient was used to compare the ability and variability. Of 749 patients, PCa was detected in 251 (33.5%) and significant PCa was detected in 133 (17.8%). Calibration plots showed an acceptable parallelism and similar discrimination ability for both PSA levels with an area under the curve of 0.69 for PCa and 0.74 for significant PCa. The ERSPC showed 226 (30.2%) unnecessary biopsies with the loss of 10 significant PCa. The variability of the RC was 16% for PCa and 20% for significant PCa, and a higher variability was associated with a reduced risk of significant PCa. We can conclude that the performance of the ERSPC-RC in the present cohort shows a high similitude between the 2 PSA levels; however, the RC variability value is associated with a decreased risk of significant PCa. The use of the ERSPC in our cohort detects a high number of unnecessary biopsies. Thus, the incorporation of ERSPC-RC could help the clinical decision to carry out a prostate biopsy. Copyright © 2016 Elsevier Inc. All rights reserved.

  9. Development of external coupling for calculation of the control rod worth in terms of burn-up for a WWER-1000 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Noori-Kalkhoran, Omid, E-mail: o_noori@yahoo.com [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Yarizadeh-Beneh, Mehdi [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of); Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Calculation of control rod worth in term of burn-up. • Calculation of differential and integral control rod worth. • Developing an external couple. • Modification of thermal-hydraulic profiles in calculations. - Abstract: One of the main problems relating to operation of a nuclear reactor is its safety and controlling system. The most widely used control systems for thermal reactors are neutron absorbent rods. In this study a code based method has been developed for calculation of integral and differential control rod worth in terms of burn-up for a WWER-1000 nuclear reactor. External coupling of WIMSD-5B, PARCS V2.7 and COBRA-EN has been used for this purpose. WIMSD-5B has been used for cell calculation and handling burn-up of the core in various days. PARCS V2.7 has been used for neutronic calculation of core and critical boron concentration search. Thermal-hydraulic calculation has been performed by COBRA-EN. An external coupling algorithm has been developed by MATLAB to couple and transfer suitable data between these codes in each step. Steady-State Power Picking Factors (PPFs) of the core and control rod worth for different control rod groups have been calculated from Beginning Of Cycle (BOC) to 289.7 Effective Full Power Days (EFPDs) in some steps. Results have been compared with the results of Bushehr Nuclear Power Plant (BNPP) Final Safety Analysis Report (FSAR). The results show a good agreement and confirm the ability of developed coupling in calculation of control rod worth in terms of burn-up.

  10. Spectral measurements in critical assemblies: MCNP specifications and calculated results

    Energy Technology Data Exchange (ETDEWEB)

    Stephanie C. Frankle; Judith F. Briesmeister

    1999-12-01

    Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k{sub eff} measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a {sup 252}Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented.

  11. Spectral measurements in critical assemblies: MCNP specifications and calculated results

    International Nuclear Information System (INIS)

    Frankle, Stephanie C.; Briesmeister, Judith F.

    1999-01-01

    Recently, a suite of 86 criticality benchmarks for the Monte Carlo N-Particle (MCNP) transport code was developed, and the results of testing the ENDF/B-V and ENDF/B-VI data (through Release 2) were published. In addition to the standard k eff measurements, other experimental measurements were performed on a number of these benchmark assemblies. In particular, the Cross Section Evaluation Working Group (CSEWG) specifications contain experimental data for neutron leakage and central-flux measurements, central-fission ratio measurements, and activation ratio measurements. Additionally, there exists another set of fission reaction-rate measurements performed at the National Institute of Standards and Technology (NIST) utilizing a 252 Cf source. This report will describe the leakage and central-flux measurements and show a comparison of experimental data to MCNP simulations performed using the ENDF/B-V and B-VI (Release 2) data libraries. Central-fission and activation reaction-rate measurements will be described, and the comparison of experimental data to MCNP simulations using available data libraries for each reaction of interest will be presented. Finally, the NIST fission reaction-rate measurements will be described. A comparison of MCNP results published previously with the current MCNP simulations will be presented for the NIST measurements, and a comparison of the current MCNP simulations to the experimental measurements will be presented

  12. Critical Values for Lawshe's Content Validity Ratio: Revisiting the Original Methods of Calculation

    Science.gov (United States)

    Ayre, Colin; Scally, Andrew John

    2014-01-01

    The content validity ratio originally proposed by Lawshe is widely used to quantify content validity and yet methods used to calculate the original critical values were never reported. Methods for original calculation of critical values are suggested along with tables of exact binomial probabilities.

  13. External radiotherapy in macular degeneration: Our technique, dosimetric calculation, and preliminary results

    International Nuclear Information System (INIS)

    Akmansu, M.; Dirican, Bahar; Oeztuerk, Berrin; Egehan, Ibrahim; Subasi, Mahmut; Or, Meral

    1998-01-01

    Purpose: This study was performed to determine the toxicity and efficacy of external-beam radiotherapy in patients with age-related subfoveal neovascularization. Methods and Materials: Between January 1996 and September 1996, 25 patients with a mean age of 70.5 (60-84) years were enrolled. All patients underwent fluorescein angiographic evaluation and documentation of their neovascular disease prior to irradiation. A total of 25 patients were treated with a total dose of 12 Gy in 6 fractions over 8 days. We used a lens-sparing technique and patients were treated with a single lateral 6-MV photon beam. To assess the risk of radiation carcinogenesis after treatment of age-related subfoveal neovascularization, we estimated the effective dose for a standard patient on the basis of tissue-weighting factors as defined by the International Commission on Radiological Protection (ICRP). The calculations were made with TLD on a male randophantom. The lens dose was found to be 0.217 Gy per fraction. Results: No significant acute morbidity was noted. Visual acuity was maintained or improved in 76% and 80% of treated patients at their 1- and 3-month follow-up examinations, respectively. On angiographic imaging, there was stabilization of subfoveal neovascular membranes in 23 patients (92%) at 3 months after irradiation. Conclusion: Our observations on these 25 patients in this study indicate that many patients will have improved or stable vision after radiotherapy treatment with low-dose irradiation

  14. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    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

  15. A calculation and measurement of the flow field in a steam condenser external to the tube nest

    International Nuclear Information System (INIS)

    Stastny, M.; Feistauer, M.

    1989-01-01

    The suggested physical and mathematical model is used to solve the flow of steam normal to the cooling tubes of condenser cross-sections in the region external to the nests. Numerical calculations are carried out by means of a multipurpose system of programmes for the finite element method and a programme for the boundary layer calculation. The results of the calculations are compared with measurements on the condenser of a 500MW steam turbine. The calculations of the flow field in a double pass condenser for the 1000MW saturated steam turbine are described. (author)

  16. Critical analysis of the condensation of water vapor at external surface of the duct

    Science.gov (United States)

    Kumar, Dileep; Memon, Rizwan Ahmed; Memon, Abdul Ghafoor; Ali, Intizar; Junejo, Awais

    2018-01-01

    In this paper, the effects of contraction of the insulation of the air duct of heating, ventilation, and air conditioning (HVAC) system is investigated. The compression of the insulation contracts it at joint, turn and other points of the duct. The energy loss and the condensation resulted from this contraction are also estimated. A mathematical model is developed to simulate the effects of this contraction on the heat gain, supply air temperature and external surface temperature of the duct. The simulation uses preliminary data obtained from an HVAC system installed in a pharmaceutical company while varying the operating conditions. The results reveal that insulation thickness should be kept greater than 30 mm and the volume flow rate of the selected air distribution system should be lower than 1.4m3/s to subside condensation on the external surface of the duct. Additionally, the optimum insulation thickness was determined by considering natural gas as an energy source and fiberglass as an insulation material. The optimum insulation thickness determined for different duct sizes varies from 28 to 45 mm, which is greater than the critical insulation thickness. Therefore, the chances of condensation on the external surface of the duct could be avoided at an optimum insulation thickness. Moreover, the effect of pressure loss coefficient of the duct fitting of air distribution system is estimated. The electricity consumption in air handling unit (AHU) decreases from 2.1 to 1.5 kW by decreasing the pressure loss coefficient from 1.5 to 0.5.

  17. External Thermal Insulation Composite Systems: Critical Parameters for Surface Hygrothermal Behaviour

    Directory of Open Access Journals (Sweden)

    Eva Barreira

    2014-01-01

    Full Text Available External Thermal Insulation Composite Systems (ETICS are often used in Europe. Despite its thermal advantages, low cost, and ease of application, this system has serious problems of biological growth causing the cladding defacement. Recent studies pointed that biological growth is due to high values of surface moisture content, which mostly results from the combined effect of exterior surface condensation, wind-driven rain, and drying process. Based on numerical simulation, this paper points the most critical parameters involved in hygrothermal behaviour of ETICS, considering the influence of thermal and hygric properties of the external rendering, the effect of the characteristics of the façade, and the consequences of the exterior and interior climate on exterior surface condensation, wind-driven rain, and drying process. The model used was previously validated by comparison with the results of an “in situ” campaign. The results of the sensitivity analyses show that relative humidity and temperature of the exterior air, atmospheric radiation, and emissivity of the exterior rendering are the parameters that most influence exterior surface condensation. Wind-driven rain depends mostly on horizontal rain, building’s height, wind velocity, and orientation. The drying capacity is influenced by short-wave absorbance, incident solar radiation, and orientation.

  18. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  19. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    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

  20. A study of physics of sub-critical multiplicative systems driven by sources and the utilization of deterministic codes in calculation of this systems

    International Nuclear Information System (INIS)

    Antunes, Alberi

    2008-01-01

    This work presents the Physics of Source Driven Systems (ADS). It shows some statics and K i netics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)

  1. A proposal for the calculation of the critical buckling of a PWR or undermoderated lattice

    International Nuclear Information System (INIS)

    Benoist, P.

    1989-01-01

    A method improving the calculation of the critical buckling of a PWR or undermorated lattice is proposed. This method takes into account the lattice heterogeneity with more detail than the existing ones; it lies on some approximations. The method requires a relatively small inplementational effort. It could be used in the calculation of fast reactors [fr

  2. Calculation and mapping of critical loads in Europe: Status report 1993

    International Nuclear Information System (INIS)

    Downing, R.J.; Hettelingh, J.P.; De Smet, P.A.M.

    1993-01-01

    The work of the RIVM Coordination Center for Effects (CCE) and National Focal Centers (NFCs) for Mapping over the past two years is summarized. The primary task of the critical loads mapping program during this period was to compute and map critical loads of sulphur in Europe. Efforts were undertaken to enhance the scientific foundations and policy relevance of the critical load program, and to foster consensus among producers and users of this information by means of three workshops. The applied calculation methods are described, as well as the resulting critical loads maps, based upon the outcomes of the workshops. Chapter 2 contains the most recent maps (May 1993) of the critical load of acidity as well as the critical load of sulphur and critical sulphur deposition, which are derived from the critical load of acidity. The chapter also contains maps of the sulphur deposition in Europe in 1980 and 1990, and the resulting exceedances. In chapter 3 the methods and equations used to derive the maps of critical loads and exceedances of acidity and sulphur are described with emphasis on the advances in the calculation methods used since the first European critical loads maps were produced in 1991. In chapter 4 the methods to be used to compute and map critical loads in the future are presented. In chapter 5 an overview of the data inputs is given, and the methods of data handling performed by the CCE to produce the current European maps of critical loads. In chapter 6 the results of an uncertainty analysis is described, which was performed on the critical loads computation methodology to assess the reliability of the computation results and the importance of the various input variables. Chapter 7 provides some conclusions and recommendations resulting from the critical load mapping activities. In Appendix 1 the reports of the can be found, with additional maps of critical loads and background variables in Appendix 2. 15 figs., 11 tabs., 156 refs

  3. Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

    Energy Technology Data Exchange (ETDEWEB)

    Maerker, R.E.; Maudlin, P.J.

    1981-02-01

    A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.

  4. Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

    International Nuclear Information System (INIS)

    Maerker, R.E.; Maudlin, P.J.

    1981-02-01

    A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment

  5. SOILD: A computer model for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil

    International Nuclear Information System (INIS)

    Chen, S.Y.; LePoire, D.; Yu, C.; Schafetz, S.; Mehta, P.

    1991-01-01

    The SOLID computer model was developed for calculating the effective dose equivalent from external exposure to distributed gamma sources in soil. It is designed to assess external doses under various exposure scenarios that may be encountered in environmental restoration programs. The models four major functional features address (1) dose versus source depth in soil, (2) shielding of clean cover soil, (3) area of contamination, and (4) nonuniform distribution of sources. The model is also capable of adjusting doses when there are variations in soil densities for both source and cover soils. The model is supported by a data base of approximately 500 radionuclides. 4 refs

  6. Criticality reference benchmark calculations for burnup credit using spent fuel isotopics

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1991-04-01

    To date, criticality analyses performed in support of the certification of spent fuel casks in the United States do not take credit for the reactivity reduction that results from burnup. By taking credit for the fuel burnup, commonly referred to as ''burnup credit,'' the fuel loading capacity of these casks can be increased. One of the difficulties in implementing burnup credit in criticality analyses is that there have been no critical experiments performed with spent fuel which can be used for computer code validation. In lieu of that, a reference problem set of fresh fuel critical experiments which model various conditions typical of light water reactor (LWR) transportation and storage casks has been identified and used in the validation of SCALE-4. This report documents the use of this same problem set to perform spent fuel criticality benchmark calculations by replacing the actual fresh fuel isotopics from the experiments with six different sets of calculated spent fuel isotopics. The SCALE-4 modules SAS2H and CSAS4 were used to perform the analyses. These calculations do not model actual critical experiments. The calculated k-effectives are not supposed to equal unity and will vary depending on the initial enrichment and burnup of the calculated spent fuel isotopics. 12 refs., 11 tabs

  7. Validation of the criticality calculation for fuel elements using the Gamtec 2 - Keno 2 and 4

    International Nuclear Information System (INIS)

    Teixeira, M.C.C.; Andrade, M.C. de

    1990-01-01

    For criticality safety in the fabrication, storage and transportation of fuel assemblies, subcriticality analysis must be done. The calculations are performed at CDTN with the GAMTEC computer code, to homogenize the fuel assembly in order to create 16 group cross-section library, and with KENO code, for determining the multiplication factor. To validate the calculational method, suitable Benchmark experiments have been done. The results show that the calculational model overestimates kef when kef+ 2 σ was considered. (author) [pt

  8. Calculated k-effectives for light water reactor typical, U + Pu nitrate solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1982-01-01

    The Department of Energy's Consolidated Fuel Reprocessing Program has as a goal the design of nuclear fuel reprocessing equipment. In order to validate computer codes used for criticality analyses in the design of such equipment, k-effectives have been calculated for several U + Pu nitrate solution critical experiments. As of January 1981, descriptions of 45 unpoisoned, U + Pu solution experiments were available in the open literature. Twelve of these experiments were performed with solutions which have physical characteristics typical of dissolved, light water reactor fuel. This paper contains a discussion of these twelve experiments, a review of the calculational procedure used to determine k-effectives, and the results of the calculations

  9. On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

    International Nuclear Information System (INIS)

    El-Osery, I.A.

    1981-01-01

    The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented

  10. Analysis and evaluation of critical experiments for validation of neutron transport calculations

    International Nuclear Information System (INIS)

    Bazzana, S.; Blaumann, H; Marquez Damian, J.I

    2009-01-01

    The calculation schemes, computational codes and nuclear data used in neutronic design require validation to obtain reliable results. In the nuclear criticality safety field this reliability also translates into a higher level of safety in procedures involving fissile material. The International Criticality Safety Benchmark Evaluation Project is an OECD/NEA activity led by the United States, in which participants from over 20 countries evaluate and publish criticality safety benchmarks. The product of this project is a set of benchmark experiment evaluations that are published annually in the International Handbook of Evaluated Criticality Safety Benchmark Experiments. With the recent participation of Argentina, this information is now available for use by the neutron calculation and criticality safety groups in Argentina. This work presents the methodology used for the evaluation of experimental data, some results obtained by the application of these methods, and some examples of the data available in the Handbook. [es

  11. Development and validation of a criticality calculation scheme based on French deterministic transport codes

    International Nuclear Information System (INIS)

    Santamarina, A.

    1991-01-01

    A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

  12. Calculations of External Irradiation from Radioactive Plume in the Early Stage of Nuclear Accident

    Czech Academy of Sciences Publication Activity Database

    Pecha, Petr; Hofman, Radek

    Vol. 50, č. 1-4 (2012), s. 420-430 ISSN 1741-5101 R&D Projects: GA MV(CZ) VG20102013018 Institutional support: RVO:67985556 Keywords : radioactivity release * photon fluence * external irradiation * data assimilation Subject RIV: AQ - Safety, Health Protection, Human - Machine http://library.utia.cas.cz/separaty/2012/AS/pecha-0385225.pdf

  13. Life cycle assessment of renewables: present issues, future outlook and implications for the calculation of external costs

    International Nuclear Information System (INIS)

    Frankl, P.

    2002-01-01

    In principle, Life Cycle Assessment (LCA) is certainly appropriate for estimating external costs of renewables, since major environmental impacts of the latter are generated in phases of the life cycle other than use. In practice however, several issues still remain. They are related to the availability and quality of Life Cycle Inventory (LCI) data, to the frit technological development of renewable energy technologies (RET), to the existence of many different applications of the latter and to a strong dependency on local conditions. Moreover, a 'static' picture of present technologies is not enough for policy indications. Therefore some kind of dynamic LCA is needed. These LCA issues are reflected in the calculation of external costs. First, the paper discusses these issues on the examples of two main technologies, namely photovoltaic (PV) and wind. Second, it discusses the results of ExternE for these two specific technologies and gives an outlook for the future. Future needs for a better use of LCA as a support tool for the calcination of external costs are identified. Finally, a new research project funded by the European Commission focused on LCI of renewables is briefly introduced and presented. (author)

  14. Literature research concerning alternative methods for validation of criticality calculation systems

    International Nuclear Information System (INIS)

    Behler, Matthias

    2016-05-01

    Beside radiochemical analysis of irradiated fuel and critical experiments, which has become a well-established basis for the validation of depletion code and criticality codes respectively, also results of oscillation experiments or the operating conditions of power reactor and research reactors can provide useful information for the validation of the above mentioned codes. Based on a literature review the potential of the utilization of oscillation experiment measurements for the validation of criticality codes is estimated. It is found that the reactivity measurements for actinides and fission products within the CERES program on the reactors DIMPLE (Winfrith, UK) and MINERVE (Cadarache, France) can give a valuable addition to the commonly used critical experiments for criticality code validation. However, there are approaches but yet no generally satisfactory solution for integrating the reactivity measurements in a quantitative bias determination for the neutron multiplication factor of typical application cases including irradiated spent fuel outside reactor cores, calculated using common criticality codes.

  15. Calculation code used in criticality analyses for the accident of JCO precipitation tank

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2000-01-01

    In order to evaluate nuclear features on criticality accident formed at the nuclear fuel processing facility in Tokai Works of the JCO, Ltd. (JCO), in Tokai-mura, Ibaraki prefecture, dynamic analyses to calculate output change after occurring the accident as well as criticality analyses to calculate reactivity added to precipitation tank, were carried out according to scenario on accident formation. For the criticality analyses, a continuous energy Monte Carlo code MCNP was used to carry out calculation of reactivity fed into the precipitation tank as correctly as possible. And, SRAC code system was used for calculation on temperature and void reactivity coefficients, effective delayed neutron ratio beta eff , and instantaneous neutron generation time required for parameters controlling transition features at criticality accident. In addition, for the dynamic analyses, because of necessity of considering on volume expansion of solution fuels used as exothermic body and radiation decomposition gas forming into solution, output behavior, numbers of nuclear fission, and so forth at initial burst portion were calculated by using TRACE and quasi-regular code, at a center of AGNES-2 promoting on its development in JAERI. Here were reported on outlines and an analysis example on calculation code using for the nuclear features evaluation. (G.K.)

  16. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    International Nuclear Information System (INIS)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility

  17. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    Energy Technology Data Exchange (ETDEWEB)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

  18. Proposal on the accelerator driven molten-salt reactor (ATW concept) benchmark calculations. (STAGE 1 - without an external neutron source)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The first stage of ATW neutronic benchmark (without an external source), based on the simple modelling of two component concept is presented. The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark is not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (author)

  19. The use of externality estimates in the calculation of adders by state PUC regulators

    International Nuclear Information System (INIS)

    Burtraw, D.; Palmer, K.; Krupnick, A.

    1994-01-01

    The primary focus of the U. S.-EC study is the development and illustration of methodologies for the estimation of marginal damages and associated externalities that result from the addition of electricity generating capacity in a specific reference environment. This paper describes how this information can be used to guide resource planning by electric utilities and State public utility commissions (PUCs). First, we discuss the 'second-best' policy environment in which PUCs must operate. We then discuss the use of 'adders' which are a policy tool that many PUCs are currently considering. Then, we introduce and estimate a formal model to calibrate these adders, based on estimates of externalities in order to promote economic efficiency in resource planning and investment decisions

  20. The use of externality estimates in the calculation of adders by state PUC regulators

    Energy Technology Data Exchange (ETDEWEB)

    Burtraw, D; Palmer, K; Krupnick, A

    1994-07-01

    The primary focus of the U. S.-EC study is the development and illustration of methodologies for the estimation of marginal damages and associated externalities that result from the addition of electricity generating capacity in a specific reference environment. This paper describes how this information can be used to guide resource planning by electric utilities and State public utility commissions (PUCs). First, we discuss the 'second-best' policy environment in which PUCs must operate. We then discuss the use of 'adders' which are a policy tool that many PUCs are currently considering. Then, we introduce and estimate a formal model to calibrate these adders, based on estimates of externalities in order to promote economic efficiency in resource planning and investment decisions.

  1. Application of the annular dispersed flow model to two-phase critical flow calculation

    International Nuclear Information System (INIS)

    Ivandaev, A.I.; Nigmatulin, B.I.

    1977-01-01

    The application of the annular dispersed flow model with an effective monodisperse core to the calculation of vapour-liquid mixture maximum rates through long pipes is discussed. An effect of the main dominant parameters such as evaporation intensity, diameter of drops picked out from the film surface and initial drop diameter at the pipe inlet on the outlet critical condition formation process has been investigated. The corresponding model constants have been determined. The calculated and experimental values of critical rates and pressure profiles along the channel have been found to be in a satisfactory agreement in the studied range of parameters. The observed non-conformity of the calculated and experimental values of critical pressures and vapour contents can be due to inadequate accuracy of the experimental techniques

  2. Experimental study and technique for calculation of critical heat fluxes in helium boiling in tubes

    International Nuclear Information System (INIS)

    Arkhipov, V.V.; Kvasnyuk, S.V.; Deev, V.I.; Andreev, V.K.

    1979-01-01

    Studied is the effect of regime parameters on critical heat loads in helium boiling in a vertical tube in the range of mass rates of 80 2 xc) and pressures of 100<=p<=200 kPa for the vapor content range corresponding to the heat exchange crisis of the first kind. The method for calculating critical heat fluxes describing experimental data with the error less than +-15% is proposed. The critical heat loads in helium boiling in tubes reduce with the growth of pressure and vapor content in the regime parameter ranges under investigation. Both positive and negative effects of the mass rate on the critical heat flux are observed. The calculation method proposed satisfactorily describes the experimental data

  3. Calculation of the fissile mass of a graphite moderated critical assembly using 93% enriched uranium

    International Nuclear Information System (INIS)

    Correa, F.; Marzo, M.A.S.; Collussi, I.; Ferreira, A.C.A.

    1976-01-01

    The critical mass of uranium has been calculated for a graphite moderated set fueled with 93% enriched uranium to be mounted on the Instituto de Energia Atomica split table Zero Power Reactor. The core composition was optimized to permit the maximum number of configurations to be studied. Analysis of three core compositions shows that 8 Kg of uranium enriched to 93% - U-235 (by weight) and 100 Kg of thorium would be sufficient for criticality experiments [pt

  4. Nuclear criticality safety calculations for a K-25 site vacuum cleaner

    International Nuclear Information System (INIS)

    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

  5. Modeling for Dose Rate Calculation of the External Exposure to Gamma Emitters in Soil

    International Nuclear Information System (INIS)

    Allam, K. A.; El-Mongy, S. A.; El-Tahawy, M. S.; Mohsen, M. A.

    2004-01-01

    Based on the model proposed and developed in Ph.D thesis of the first author of this work, the dose rate conversion factors (absorbed dose rate in air per specific activity of soil in nGy.hr - 1 per Bq.kg - 1) are calculated 1 m above the ground for photon emitters of natural radionuclides uniformly distributed in the soil. This new and simple dose rate calculation software was used for calculation of the dose rate in air 1 m above the ground. Then the results were compared with those obtained by five different groups. Although the developed model is extremely simple, the obtained results of calculations, based on this model, show excellent agreement with those obtained by the above-mentioned models specially that one adopted by UNSCEAR. (authors)

  6. Calculated K-effectives using ENDF/B-V data for U + Pu solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1981-01-01

    Effective multiplication factors for 12 critical experiments have been calculated using multigroup cross sections derived from the ENDF/B-V library. All 12 experiments contained mixed plutonium and uranium nitrate solutions. The range of hydrogen-to-fissile plutonium atom ratios spanned by these experiments was 200 to 2200. A comparison with K-effectives calculated with ENDF/B-IV data is presented

  7. Calculation of the ingestion critical dose rate for the Goiania radioactive waste repository

    International Nuclear Information System (INIS)

    Passos, E.M. dos; Martin Alves, A.S. De

    1994-01-01

    The calculation results of the critical distance for the ingestion dose rate due to a hypothetical Cs-137 release from the Abadia de Goias repository are shown. The work is based on the pathway repository-aquifer-well food chain. The calculations were based upon analytical models for the migration of radioisotopes through the aquifer and for its transfer from well water to food. (author)

  8. Calculation of criticality of the AP600 reactor with KENO V.a code

    Energy Technology Data Exchange (ETDEWEB)

    Krumbein, A; Caner, M; Shapira, M [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center

    1996-12-01

    The Westinghouse AP600 PWR has been modeled using the KENO V.a three dimensional Monte Carlo criticality program of the SCALE-PC code system. These calculations and the use of a Monte Carlo neutron transport code such as KENO will provide us with an independent check on our WIMS/CITATION calculations for the AP600 as well as for other reactors. It will also enable us to model more complicated geometries. (authors).

  9. Calculational study of benchmark critical experiments on high-enriched uranyl nitrate solution systems

    International Nuclear Information System (INIS)

    Oh, I.; Rothe, R.E.

    1978-01-01

    Criticality calculations on minimally reflected, concrete-reflected, and plastic-reflected single tanks and on arrays of cylinders reflected by concrete and plastic have been performed using the KENO-IV code with 16-group Hansen-Roach neutron cross sections. The fissile material was high-enriched (93.17% 235 U) uranyl nitrate [UO 2 (NO 3 ) 2 ] solution. Calculated results are compared with those from a benchmark critical experiments program to provide the best possible verification of the calculational technique. The calculated k/sub eff/'s underestimate the critical condition by an average of 1.28% for the minimally reflected single tanks, 1.09% for the concrete-reflected single tanks, 0.60% for the plastic-reflected single tanks, 0.75% for the concrete-reflected arrays of cylinders, and 0.51% for the plastic-reflected arrays of cylinders. More than half of the present comparisons were within 1% of the experimental values, and the worst calculational and experimental discrepancy was 2.3% in k/sub eff/ for the KENO calculations

  10. Calculation of the external dose rate in the spent fuel pool for the case to use compact racks

    International Nuclear Information System (INIS)

    Passos, E.M. dos; Alves, A.S.M.

    1988-01-01

    The possible introduction of compact racks in the spent fuel pool of the Angra 1 Nuclear Power Plant largely inreases its storage capacity, but originates an increase of the gamma radiation sources. The precise evaluation of the effects of the adoption of this option on the external gamma dose rates and also on the thickness of the concrete shielding requires the utilization of sofisticated computer codes (QAD, ANISN), which allow the calculation of the gamma dose rates through thick shielding walls. This paper describes the utilized methodology for the calculation of the modified pool shieldings, showing the obtained results for the Angra 1 NPP case. The gamma dose rate was calculated with the point Kernel model, first analytically, and later through utilization of the tridimensional multigroup QAD computer code. (author) [pt

  11. Automated calculation of myocardial external efficiency from a single 11C-acetate PET/CT scan

    DEFF Research Database (Denmark)

    Harms, Hans; Tolbod, Lars Poulsen; Hansson, Nils Henrik

    of this study was to develop and validate an automated method of calculating MEE from a single dynamic 11C-acetate PETscan. Methods: 21 subjects underwent a dynamic 27 min 11C-acetate PETscan on a Siemens Biograph TruePoint 64 PET/CTscanner. Using cluster analysis, the LV-aortic time-activity curve (TACLV......). Conclusion: Myocardial efficiencycanbe derived directly andautomatically froma single dynamic 11C-acetate PET scan. This eliminates the need for a separate CMR scan and eliminates any potential errors due to different loading conditions between CMR and PETscans.......Background: Dynamic PETwith 11C-acetate can be used to assess myocardial oxygen use which in turn is usedto calculate myocardial external efficiency (MEE), anearly marker of heart failure. MEE is defined as the ratio of total work (TW) and total energy use (TE). Calculation of TW and TE requires...

  12. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  13. Criticality calculation of the deposits for the fuel elements in RP-10 nuclear research reactor

    International Nuclear Information System (INIS)

    Aguirre, Alvaro; Bruna, Ruben

    2013-01-01

    This paper shows the results of the criticality calculation of the deposits for irradiated and non-irradiated fuel elements in the RP-10 research reactor with MCNP5 code. In all cases and for normal and incidental conditions, the effective multiplication factor (K eff ) results less than 0,90 according to the acceptance criterion. (authors).

  14. Calculation of critical concentrations of actinides in an infinite medium of silicon dioxide

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Sato, Shohei; Kawasaki, Hiromitsu

    2009-01-01

    The critical concentrations of actinides in metal-silicon-dioxide (SiO 2 ) and in metal-water (H 2 O) mixtures were calculated for 26 actinides including 233,235 U, 239,241 Pu, 242m Am, 243,245,247 Cm, and 249,251 Cf. The calculations were performed using the Monte Carlo neutron transport calculation code MCNP5 combined with the evaluated nuclear data library JENDL3.3. The results showed that the critical concentration of actinide in metal-SiO 2 mixtures was about 1/5 of that in metal-H 2 O mixtures for all the fissile nuclides investigated. The k ∞ 's of metal-SiO 2 and metal-H 2 O at one-half of the respective critical concentration of actinide, which was assumed as the subcritical concentration limit, were found to be less than 0.8 for all the actinides considered. By applying the sum-of-fractions rule to the concentrations of six nuclides in metal-SiO 2 mixtures, the subcriticality of high-level radioactive wastes was confirmed for a reported sample. The effects of different nuclear data libraries on the results of critical concentrations were found to be large for 242 Cm, 247 Cm, and 250 Cf by comparison with the results calculated with another evaluated nuclear data library, ENDF/B-VI. (author)

  15. Method of calculation of critical values of financial indicators for developing food security strategy

    Science.gov (United States)

    Aigyl Ilshatovna, Sabirova; Svetlana Fanilevna, Khasanova; Vildanovna, Nagumanova Regina

    2018-05-01

    On the basis of decision making theory (minimax and maximin approaches) the authors propose a technique with the results of calculations of the critical values of effectiveness indicators of agricultural producers in the Republic of Tatarstan for 2013-2015. There is justified necessity of monitoring the effectiveness of the state support and the direction of its improvement.

  16. Modification and validation of an analytical source model for external beam radiotherapy Monte Carlo dose calculations

    Energy Technology Data Exchange (ETDEWEB)

    Davidson, Scott E., E-mail: sedavids@utmb.edu [Radiation Oncology, The University of Texas Medical Branch, Galveston, Texas 77555 (United States); Cui, Jing [Radiation Oncology, University of Southern California, Los Angeles, California 90033 (United States); Kry, Stephen; Ibbott, Geoffrey S.; Followill, David S. [Department of Radiation Physics, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 (United States); Deasy, Joseph O. [Department of Medical Physics, Memorial Sloan Kettering Cancer Center, New York, New York 10065 (United States); Vicic, Milos [Department of Applied Physics, University of Belgrade, Belgrade 11000 (Serbia); White, R. Allen [Bioinformatics and Computational Biology, The University of Texas MD Anderson Cancer Center, Houston, Texas 77030 (United States)

    2016-08-15

    Purpose: A dose calculation tool, which combines the accuracy of the dose planning method (DPM) Monte Carlo code and the versatility of a practical analytical multisource model, which was previously reported has been improved and validated for the Varian 6 and 10 MV linear accelerators (linacs). The calculation tool can be used to calculate doses in advanced clinical application studies. One shortcoming of current clinical trials that report dose from patient plans is the lack of a standardized dose calculation methodology. Because commercial treatment planning systems (TPSs) have their own dose calculation algorithms and the clinical trial participant who uses these systems is responsible for commissioning the beam model, variation exists in the reported calculated dose distributions. Today’s modern linac is manufactured to tight specifications so that variability within a linac model is quite low. The expectation is that a single dose calculation tool for a specific linac model can be used to accurately recalculate dose from patient plans that have been submitted to the clinical trial community from any institution. The calculation tool would provide for a more meaningful outcome analysis. Methods: The analytical source model was described by a primary point source, a secondary extra-focal source, and a contaminant electron source. Off-axis energy softening and fluence effects were also included. The additions of hyperbolic functions have been incorporated into the model to correct for the changes in output and in electron contamination with field size. A multileaf collimator (MLC) model is included to facilitate phantom and patient dose calculations. An offset to the MLC leaf positions was used to correct for the rudimentary assumed primary point source. Results: Dose calculations of the depth dose and profiles for field sizes 4 × 4 to 40 × 40 cm agree with measurement within 2% of the maximum dose or 2 mm distance to agreement (DTA) for 95% of the data

  17. Use of calculations in the hygienic evaluation and prediction of external γ-irradiation levels of population

    International Nuclear Information System (INIS)

    Karpov, V.I.

    1980-01-01

    The estimation method of the annual average equivalent dose of population external irradiation calculated per 1 person and collective dose using the data of gamma radiation in buildings and inhabited points is proposed. The modified ratio (from the report of UN Scientific Committee, 1977) for the determination of external irradiation dose per 1 person is given. Dynamics and value of population exposure are determined. Particular solutions with a number of corresponding approximating functions are proposed for concrete rural and urban conditions. It is shown that with the construction of stone buildings collective dose growth considerably passes ahead of the population growth. Dynamics of the population radiation dose is forecasted upto 2000 year. It is shown that the cumulative error does not exceed +-25%

  18. Monte Carlo calculation of dose rate conversion factors for external exposure to photon emitters in soil

    CERN Document Server

    Clouvas, A; Antonopoulos-Domis, M; Silva, J

    2000-01-01

    The dose rate conversion factors D/sub CF/ (absorbed dose rate in air per unit activity per unit of soil mass, nGy h/sup -1/ per Bq kg/sup -1/) are calculated 1 m above ground for photon emitters of natural radionuclides uniformly distributed in the soil. Three Monte Carlo codes are used: 1) The MCNP code of Los Alamos; 2) The GEANT code of CERN; and 3) a Monte Carlo code developed in the Nuclear Technology Laboratory of the Aristotle University of Thessaloniki. The accuracy of the Monte Carlo results is tested by the comparison of the unscattered flux obtained by the three Monte Carlo codes with an independent straightforward calculation. All codes and particularly the MCNP calculate accurately the absorbed dose rate in air due to the unscattered radiation. For the total radiation (unscattered plus scattered) the D/sub CF/ values calculated from the three codes are in very good agreement between them. The comparison between these results and the results deduced previously by other authors indicates a good ag...

  19. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

    Carluccio, Thiago

    2011-01-01

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as k eff and k src , and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  20. Evaluation of dose equivalent rate distribution in JCO critical accident by radiation transport calculation

    CERN Document Server

    Sakamoto, Y

    2002-01-01

    In the prevention of nuclear disaster, there needs the information on the dose equivalent rate distribution inside and outside the site, and energy spectra. The three dimensional radiation transport calculation code is a useful tool for the site specific detailed analysis with the consideration of facility structures. It is important in the prediction of individual doses in the future countermeasure that the reliability of the evaluation methods of dose equivalent rate distribution and energy spectra by using of Monte Carlo radiation transport calculation code, and the factors which influence the dose equivalent rate distribution outside the site are confirmed. The reliability of radiation transport calculation code and the influence factors of dose equivalent rate distribution were examined through the analyses of critical accident at JCO's uranium processing plant occurred on September 30, 1999. The radiation transport calculations including the burn-up calculations were done by using of the structural info...

  1. Critical mass calculations for 241Am, 242mAm and 243Am

    International Nuclear Information System (INIS)

    Dias, Hemanth; Tancock, Nigel; Clayton, Angela

    2003-01-01

    Criticality mass calculations are reported for 241 Am, 242m Am and 243 Am using the MONK and MCNP computer codes with the UKNDL, JEF-2.2, ENDF/B-VI and JENDL-3.2 nuclear data libraries. Results are reported for spheres of americium metal and dioxide in bare, water reflected and steel reflected systems. Comparison of results led to the identification of a serious inconsistency in the 241 Am ENDF/B-VI DICE library used by MONK - this demonstrates the importance of using different codes to verify critical mass calculations. The 241 Am critical mass estimates obtained using UKNDL and ENDF/B-VI show good agreement with experimentally inferred data, whilst both JEF-2.2 and JENDL-3.2 produce higher estimates of critical mass. The computed critical mass estimates for 242m Am obtained using ENDF/B-VI are lower than the results produced using the other nuclear data libraries - the ENDF/B-VI fission cross-section for 242m Am is significantly higher than the other evaluations in the fast region and is not supported by recent experimental data. There is wide variation in the computed 243 Am critical mass estimates suggesting that there is still considerable uncertainty in the 243 Am nuclear data. (author)

  2. Analysis of kyoto university reactor physics critical experiments using NCNSRC calculation methodology

    International Nuclear Information System (INIS)

    Amin, E.; Hathout, A.M.; Shouman, S.

    1997-01-01

    The kyoto university reactor physics experiments on the university critical assembly is used to benchmark validate the NCNSRC calculations methodology. This methodology has two lines, diffusion and Monte Carlo. The diffusion line includes the codes WIMSD4 for cell calculations and the two dimensional diffusion code DIXY2 for core calculations. The transport line uses the MULTIKENO-Code vax Version. Analysis is performed for the criticality, and the temperature coefficients of reactivity (TCR) for the light water moderated and reflected cores, of the different cores utilized in the experiments. The results of both Eigen value and TCR approximately reproduced the experimental and theoretical Kyoto results. However, some conclusions are drawn about the adequacy of the standard wimsd4 library. This paper is an extension of the NCNSRC efforts to assess and validate computer tools and methods for both Et-R R-1 and Et-MMpr-2 research reactors. 7 figs., 1 tab

  3. OECD/NEA burnup credit criticality benchmarks phase IIIA: Criticality calculations of BWR spent fuel assemblies in storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ando, Yoshihira [Toshiba Corp., Kawasaki, Kanagawa (Japan)

    2000-09-01

    The report describes the final results of Phase IIIA Benchmarks conducted by the Burnup Credit Criticality Calculation Working Group under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA). The benchmarks are intended to confirm the predictive capability of the current computer code and data library combinations for the neutron multiplication factor (k{sub eff}) of a layer of irradiated BWR fuel assembly array model. In total 22 benchmark problems are proposed for calculations of k{sub eff}. The effects of following parameters are investigated: cooling time, inclusion/exclusion of FP nuclides and axial burnup profile, and inclusion of axial profile of void fraction or constant void fractions during burnup. Axial profiles of fractional fission rates are further requested for five cases out of the 22 problems. Twenty-one sets of results are presented, contributed by 17 institutes from 9 countries. The relative dispersion of k{sub eff} values calculated by the participants from the mean value is almost within the band of {+-}1%{delta}k/k. The deviations from the averaged calculated fission rate profiles are found to be within {+-}5% for most cases. (author)

  4. Calculation of dose conversion factors for doses in the fingernails to organ doses at external gamma irradiation in air

    International Nuclear Information System (INIS)

    Khailov, A.M.; Ivannikov, A.I.; Skvortsov, V.G.; Stepanenko, V.F.; Orlenko, S.P.; Flood, A.B.; Williams, B.B.; Swartz, H.M.

    2015-01-01

    Absorbed doses to fingernails and organs were calculated for a set of homogenous external gamma-ray irradiation geometries in air. The doses were obtained by stochastic modeling of the ionizing particle transport (Monte Carlo method) for a mathematical human phantom with arms and hands placed loosely along the sides of the body. The resulting dose conversion factors for absorbed doses in fingernails can be used to assess the dose distribution and magnitude in practical dose reconstruction problems. For purposes of estimating dose in a large population exposed to radiation in order to triage people for treatment of acute radiation syndrome, the calculated data for a range of energies having a width of from 0.05 to 3.5 MeV were used to convert absorbed doses in fingernails to corresponding doses in organs and the whole body as well as the effective dose. Doses were assessed based on assumed rates of radioactive fallout at different time periods following a nuclear explosion. - Highlights: • Elemental composition and density of nails were determined. • MIRD-type mathematical human phantom with arms and hands was created. • Organ doses and doses to nails were calculated for external photon exposure in air. • Effective dose and nail doses values are close for rotational and soil surface exposures.

  5. Nuclear criticality safety experiments, calculations, and analyses: 1958 to 1982. Volume 1. Lookup tables

    International Nuclear Information System (INIS)

    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

  6. Comparison of dose calculation algorithms for treatment planning in external photon beam therapy for clinical situations

    DEFF Research Database (Denmark)

    Knöös, Tommy; Wieslander, Elinore; Cozzi, Luca

    2006-01-01

    to the fields. A Monte Carlo calculated algorithm input data set and a benchmark set for a virtual linear accelerator have been produced which have facilitated the analysis and interpretation of the results. The more sophisticated models in the type b group exhibit changes in both absorbed dose and its...... distribution which are congruent with the simulations performed by Monte Carlo-based virtual accelerator....

  7. Modelling of an industrial environment, part 2.: External dose calculation with relevant countermeasures

    International Nuclear Information System (INIS)

    Eged, K.; Kis, Z.; Alvarez-Farizo, B.; Gil, J.; Voigt, G.

    2002-01-01

    The calculation of the collective dose and averted collective dose after applying countermeasures in an industrial environment has been divided in two parts. In the first part (Kis et al. 2002) separate Monte Carlo simulations of photon transport resulted in the air kermas per photon per unit area due to the industrial surfaces contaminated by 1 37C s at specific points using the so-called local approach. In the local approach the air kerma rates due to specific intervention elements at the evaluation locations in the whole environment are determined (Gutierrez et al. 2000). In this way the collective and averted collective dose due to the radiation from a particular intervention element (e.g. the roof of a building) can be obtained. It can, therefore, provide a ranking of the specific intervention elements based on their contribution to collective dose as well. The deposition pattern and the long-term behaviour of deposited radionuclides vary widely in natural circumstances; therefore the number of the photons emitted from the various surfaces per unit area and time can differ significantly. This means the results of the Monte Carlo simulations have to be weighted according to the number of emitted photons so that the actual radiation field can be set up. For this purpose, a dose calculation code has been developed in the framework of the TEMAS project (Gutierrez et al. 2000) which allows to calculate collective doses for different environments. This code has been applied in the present work

  8. First-principles calculation of transport property in nano-devices under an external magnetic field

    International Nuclear Information System (INIS)

    Chen Jingzhe; Zhang Jin; Han Rushan

    2008-01-01

    The mesoscopic quantum interference phenomenon (QIP) can be observed and behaves as the oscillation of conductance in nano-devices when the external magnetic field changes. Excluding the factor of impurities or defects, specific QIP is determined by the sample geometry. We have improved a first-principles method based on the matrix Green's function and the density functional theory to simulate the transport behaviour of such systems under a magnetic field. We have studied two kinds of QIP: universal conductance fluctuation (UCF) and Aharonov–Bohm effect (A–B effect). We find that the amplitude of UCF is much smaller than the previous theoretical prediction. We have discussed the origin of difference and concluded that due to the failure of ergodic hypothesis, the ensemble statistics is not applicable, and the conductance fluctuation is determined by the flux-dependent density of states (DOSs). We have also studied the relation between the UCF and the structure of sample. For a specific structure, an atomic circle, the A–B effect is observed and the origin of the oscillation is also discussed

  9. General Method for Calculating the Response and Noise Spectra of Active Fabry-Perot Semiconductor Waveguides With External Optical Injection

    DEFF Research Database (Denmark)

    Blaaberg, Søren; Mørk, Jesper

    2009-01-01

    We present a theoretical method for calculating small-signal modulation responses and noise spectra of active Fabry-Perot semiconductor waveguides with external light injection. Small-signal responses due to either a modulation of the pump current or due to an optical amplitude or phase modulatio...... amplifiers and an injection-locked laser. We also demonstrate the applicability of the method to analyze slow and fast light effects in semiconductor waveguides. Finite reflectivities of the facets are found to influence the phase changes of the injected microwave-modulated light....

  10. Gamow's calculation of the neutron star's critical mass revised

    International Nuclear Information System (INIS)

    Ludwig, Hendrik; Ruffini, Remo

    2014-01-01

    It has at times been indicated that Landau introduced neutron stars in his classic paper of 1932. This is clearly impossible because the discovery of the neutron by Chadwick was submitted more than one month after Landau's work. Therefore, and according to his calculations, what Landau really did was to study white dwarfs, and the critical mass he obtained clearly matched the value derived by Stoner and later by Chandrasekhar. The birth of the concept of a neutron star is still today unclear. Clearly, in 1934, the work of Baade and Zwicky pointed to neutron stars as originating from supernovae. Oppenheimer in 1939 is also well known to have introduced general relativity (GR) in the study of neutron stars. The aim of this note is to point out that the crucial idea for treating the neutron star has been advanced in Newtonian theory by Gamow. However, this pioneering work was plagued by mistakes. The critical mass he should have obtained was 6.9 M, not the one he declared, namely, 1.5 M. Probably, he was taken to this result by the work of Landau on white dwarfs. We revise Gamow's calculation of the critical mass regarding calculational and conceptual aspects and discuss whether it is justified to consider it the first neutron-star critical mass. We compare Gamow's approach to other early and modern approaches to the problem.

  11. International report to validate criticality safety calculations for fissile material transport

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1984-01-01

    During the past three years a Working Group established by the Organization for Economic Co-operation and Development's Nuclear Energy Agency (OECD-NEA) in Paris, France, has been studying the validity and applicability of a variety of criticality safety computer programs and their associated nuclear data for the computation of the neutron multiplication factor, k/sub eff/, for various transport packages used in the fuel cycle. The principal objective of this work has been to provide an internationally acceptable basis for the licensing authorities in a country to honor licensing approvals granted by other participating countries. Eleven countries participated in the initial study which consisted of examining criticality safety calculations for packages designed for spent light water reactor fuel transport. This paper presents a summary of this study which has been completed and reported in an OECD-NEA Report No. CSNI-71. The basic goal of this study was to outline a satisfactory validation procedure for this particular application. First, a set of actual critical experiments were chosen which contained the various material and geometric properties present in typical LWR transport containers. Secondly, calculations were made by each of the methods in order to determine how accurately each method reproduced the experimental values. This successful effort in developing a benchmark procedure for validating criticality calculations for spent LWR transport packages along with the successful intercomparison of a number of methods should provide increased confidence by licensing authorities in the use of these methods for this area of application. 4 references, 2 figures

  12. Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

    International Nuclear Information System (INIS)

    Okuno, H.

    2002-01-01

    We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

  13. Criticality calculations of various spent fuel casks - possibilities for burn up credit implementation

    International Nuclear Information System (INIS)

    Apostolov, T; Manolova, M.; Prodanova, R.

    2001-01-01

    A methodology for criticality safety analysis of spent fuel casks with possibilities for burnup credit implementation is presented. This methodology includes the world well-known and applied program systems: NESSEL-NUKO for depletion and SCALE-4.4 for criticality calculations. The abilities of this methodology to analyze storage and transportation casks with different type of spent fuel are demonstrated on the base of various tests. The depletion calculations have been carried out for the power reactors (WWER-440 and WWER-1000) and the research reactor IRT-2000 (C-36) fuel assemblies. The criticality calculation models have been developed on the basis of real fuel casks, designed by the leading international companies (for WWER-440 and WWER-1000 spent fuel assemblies), as well as for real a WWER-440 storage cask, applied at the 'Kozloduy' NPP. The results obtained show that the criticality safety criterion K eff less than 0.95 is satisfied for both: fresh and spent fuel. Besides the implementation of burnup credit allows to account for the reduced reactivity of spent fuel and to evaluate the conservatism of the fresh fuel assumption. (author)

  14. Reliable method for fission source convergence of Monte Carlo criticality calculation with Wielandt's method

    International Nuclear Information System (INIS)

    Yamamoto, Toshihiro; Miyoshi, Yoshinori

    2004-01-01

    A new algorithm of Monte Carlo criticality calculations for implementing Wielandt's method, which is one of acceleration techniques for deterministic source iteration methods, is developed, and the algorithm can be successfully implemented into MCNP code. In this algorithm, part of fission neutrons emitted during random walk processes are tracked within the current cycle, and thus a fission source distribution used in the next cycle spread more widely. Applying this method intensifies a neutron interaction effect even in a loosely-coupled array where conventional Monte Carlo criticality methods have difficulties, and a converged fission source distribution can be obtained with fewer cycles. Computing time spent for one cycle, however, increases because of tracking fission neutrons within the current cycle, which eventually results in an increase of total computing time up to convergence. In addition, statistical fluctuations of a fission source distribution in a cycle are worsened by applying Wielandt's method to Monte Carlo criticality calculations. However, since a fission source convergence is attained with fewer source iterations, a reliable determination of convergence can easily be made even in a system with a slow convergence. This acceleration method is expected to contribute to prevention of incorrect Monte Carlo criticality calculations. (author)

  15. Development and verification of an excel program for calculation of monitor units for tangential breast irradiation with external photon beams

    International Nuclear Information System (INIS)

    Woldemariyam, M.G.

    2015-07-01

    The accuracy of MU calculation performed with Prowess Panther TPS (for Co-60) and Oncentra (for 6MV and 15MV x-rays) for tangential breast irradiation was evaluated with measurements made in an anthropomorphic phantom using calibrated Gafchromic EBT2 films. Excel programme which takes in to account external body surface irregularity of an intact breast or chest wall (hence absence of full scatter condition) using Clarkson’s sector summation technique was developed. A single surface contour of the patient obtained in a transverse plane containing the MU calculation point was required for effective implementation of the programme. The outputs of the Excel programme were validated with the respective outputs from the 3D treatment planning systems. The variations between the measured point doses and their calculated counterparts by the TPSs were within the range of -4.74% to 4.52% (mean of -1.33% and SD of 2.69) for the prowess panther TPS and -4.42% to 3.14% (mean of -1.47% and SD of -3.95) for the Oncentra TPS. The observed degree of deviation may be attributed to limitations of the dose calculation algorithm within the TPSs, set up inaccuracies of the phantom during irradiation and inherent uncertainties associated with radiochromic film dosimetry. The percentage deviations between MUs calculated with the two TPSs and the Excel program were within the range of -3.45% and 3.82% (mean of 0.83% and SD of 2.25). The observed percentage deviations are within the 4% action level recommended by TG-114. This indicates that the Excel program can be confidently employed for calculation of MUs for 2D planned tangential breast irradiations or to independently verify MUs calculated with another calculation methods. (au)

  16. Calculation of Upper Subcritical Limits for Nuclear Criticality in a Repository

    International Nuclear Information System (INIS)

    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

  17. Development and external multicenter validation of Chinese Prostate Cancer Consortium prostate cancer risk calculator for initial prostate biopsy.

    Science.gov (United States)

    Chen, Rui; Xie, Liping; Xue, Wei; Ye, Zhangqun; Ma, Lulin; Gao, Xu; Ren, Shancheng; Wang, Fubo; Zhao, Lin; Xu, Chuanliang; Sun, Yinghao

    2016-09-01

    Substantial differences exist in the relationship of prostate cancer (PCa) detection rate and prostate-specific antigen (PSA) level between Western and Asian populations. Classic Western risk calculators, European Randomized Study for Screening of Prostate Cancer Risk Calculator, and Prostate Cancer Prevention Trial Risk Calculator, were shown to be not applicable in Asian populations. We aimed to develop and validate a risk calculator for predicting the probability of PCa and high-grade PCa (defined as Gleason Score sum 7 or higher) at initial prostate biopsy in Chinese men. Urology outpatients who underwent initial prostate biopsy according to the inclusion criteria were included. The multivariate logistic regression-based Chinese Prostate Cancer Consortium Risk Calculator (CPCC-RC) was constructed with cases from 2 hospitals in Shanghai. Discriminative ability, calibration and decision curve analysis were externally validated in 3 CPCC member hospitals. Of the 1,835 patients involved, PCa was identified in 338/924 (36.6%) and 294/911 (32.3%) men in the development and validation cohort, respectively. Multivariate logistic regression analyses showed that 5 predictors (age, logPSA, logPV, free PSA ratio, and digital rectal examination) were associated with PCa (Model 1) or high-grade PCa (Model 2), respectively. The area under the curve of Model 1 and Model 2 was 0.801 (95% CI: 0.771-0.831) and 0.826 (95% CI: 0.796-0.857), respectively. Both models illustrated good calibration and substantial improvement in decision curve analyses than any single predictors at all threshold probabilities. Higher predicting accuracy, better calibration, and greater clinical benefit were achieved by CPCC-RC, compared with European Randomized Study for Screening of Prostate Cancer Risk Calculator and Prostate Cancer Prevention Trial Risk Calculator in predicting PCa. CPCC-RC performed well in discrimination and calibration and decision curve analysis in external validation compared

  18. Calculated dependence of FePt damping on external field magnitude and direction

    Directory of Open Access Journals (Sweden)

    N. A. Natekar

    2017-05-01

    Full Text Available Near the Curie temperature (Tc, magnetic parameters including magnetization, anisotropy, and damping depend strongly on both temperature and length scale. This manifestation of renormalization theory is most readily seen in the case of magnetization where the magnitude of the atomic spin is largely unaffected by temperature, but the bulk magnetization vanishes at Tc. It has been previously argued that the Landau-Lifshitz-Gilbert damping parameter alpha exhibits a similar effect owing to its dependence on both atomic effects and magnon-magnon scattering, the latter having a strong length dependence. Here, we calculate, using an anisotropic exchange description of L10 FePt (Tc = 705 K, the damping (and other magnetic properties dependence on temperature for FePt at length scales around 1.0 nm as appropriate for high temperature micromagnetic simulation. While the damping reduces as the applied field along the easy direction increases, it tends to increase as the field direction is changed to in-plane. The renormalized parameters are also calculated for higher and lower Tc (770K and 630K by invoking the linear relationship between the exchange stiffness parameter and Curie temperature. This corresponds to doped and/or non-stoichiometric FePt and allows better understanding of the effects of varying anisotropy to exchange ratio.

  19. Exploring the use of a deterministic adjoint flux calculation in criticality Monte Carlo simulations

    International Nuclear Information System (INIS)

    Jinaphanh, A.; Miss, J.; Richet, Y.; Martin, N.; Hebert, A.

    2011-01-01

    The paper presents a preliminary study on the use of a deterministic adjoint flux calculation to improve source convergence issues by reducing the number of iterations needed to reach the converged distribution in criticality Monte Carlo calculations. Slow source convergence in Monte Carlo eigenvalue calculations may lead to underestimate the effective multiplication factor or reaction rates. The convergence speed depends on the initial distribution and the dominance ratio. We propose using an adjoint flux estimation to modify the transition kernel according to the Importance Sampling technique. This adjoint flux is also used as the initial guess of the first generation distribution for the Monte Carlo simulation. Calculated Variance of a local estimator of current is being checked. (author)

  20. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    Scharmer, K.

    1969-01-01

    The results of experiments in the light water cooled D 2 O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k eff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D 2 O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [fr

  1. DTADH and quantum critical phenomena caused by anisotropy and external magnetic field for spin-1/2 Heisenberg diamond chains

    International Nuclear Information System (INIS)

    Li Yanchao

    2010-01-01

    Using the transfer matrix renormalization group (TMRG) method, we study the connection between the first derivative of the thermal average of driving-term Hamiltonian (DTADH) and the trace of quantum critical behaviors at finite temperatures. Connecting with the exact diagonalization method, we give the phase diagrams and analyze the properties of each phase for both the ferromagnetic and anti-ferromagnetic frustrated J 3 anisotropy diamond chain models. The finite-temperature scaling behaviors near the critical regions are also investigated. Further, we show the critical behaviors driven by external magnetic field, analyze the formation of the 1/3 magnetic plateau and the influence of different interactions on those critical points for both the ferrimagnetic and anti-ferromagnetic distorted diamond chains.

  2. Effect of external magnetic field on critical current for the onset of virtual cathode oscillations in relativistic electron beams

    International Nuclear Information System (INIS)

    Hramov, Alexander; Koronovskii, Alexey; Morozov, Mikhail; Mushtakov, Alexander

    2008-01-01

    In this Letter we research the space charge limiting current value at which the oscillating virtual cathode is formed in the relativistic electron beam as a function of the external magnetic field guiding the beam electrons. It is shown that the space charge limiting (critical) current decreases with growth of the external magnetic field, and that there is an optimal induction value of the magnetic field at which the critical current for the onset of virtual cathode oscillations in the electron beam is minimum. For the strong external magnetic field the space charge limiting current corresponds to the analytical relation derived under the assumption that the motion of the electron beam is one-dimensional [D.J. Sullivan, J.E. Walsh, E. Coutsias, in: V.L. Granatstein, I. Alexeff (Eds.), Virtual Cathode Oscillator (Vircator) Theory, in: High Power Microwave Sources, vol. 13, Artech House Microwave Library, 1987, Chapter 13]. Such behavior is explained by the characteristic features of the dynamics of electron space charge in the longitudinal and radial directions in the drift space at the different external magnetic fields

  3. The calculation of external gamma-ray doses from airborne and deposited radionuclides in the environmental code NECTAR

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1982-02-01

    A computer program has been developed for the rapid evaluation of external gamma-ray doses from airborne and deposited radionuclide mixtures. Based on a gaussian dispersion model, the program calculates the dose at any position, including points high above ground level or upwind of the source. Meteorological frequency data for wind speed, direction, atmospheric stability and rainfall are fully taken into account. The calculational model assumes that the ground surface is perfectly flat and that gamma-ray paths are entirely in air; the possible errors caused by these and other assumptions are discussed, with suggested correction factors. The program applies various criteria to determine the best approximation or numerical integration method for each target point; execution times (on an IBM 370 machine) thus vary from less than 0.01s to about 0.3s per target point for a single weather category. The program has been incorporated in the environmental release program NECTAR. (author)

  4. Calculation of external exposure during transport and disposal of radioactive waste arisen from dismantling of steam generator

    International Nuclear Information System (INIS)

    Hornacek, M.; Necas, V.

    2014-01-01

    The dismantling of large components (reactor pressure vessel, reactor internals, steam generator) represents complex of processes involving preparation, dismantling, waste treatment and conditioning, transport and final disposal. To optimise all of these activities in accordance with the ALARA principle the prediction of the exposure of workers is an essential prerequisite. The paper deals with the calculation of external exposure of workers during transport and final disposal of heat exchange tubes of steam generator used in Slovak nuclear power plant V1 in Jaslovske Bohunice. The type of waste packages, the calculation models of truck and National Radioactive Waste Repository in Mochovce are presented. The detailed methodology of radioactive waste disposal is showed and the degree of influence of time decay (0, 5 and 10 years) on the radiological conditions during transport and disposal is studied. All of the results do not exceed the limits given in Slovak and international regulatory documents. (authors)

  5. DEEP code to calculate dose equivalents in human phantom for external photon exposure by Monte Carlo method

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro

    1991-01-01

    The present report describes a computer code DEEP which calculates the organ dose equivalents and the effective dose equivalent for external photon exposure by the Monte Carlo method. MORSE-CG, Monte Carlo radiation transport code, is incorporated into the DEEP code to simulate photon transport phenomena in and around a human body. The code treats an anthropomorphic phantom represented by mathematical formulae and user has a choice for the phantom sex: male, female and unisex. The phantom can wear personal dosimeters on it and user can specify their location and dimension. This document includes instruction and sample problem for the code as well as the general description of dose calculation, human phantom and computer code. (author)

  6. Environmental optimisation of natural gas fired engines - calculation of health externalities

    Energy Technology Data Exchange (ETDEWEB)

    Frohn, L.M.; Becker, T.; Christensen, Jesper; Hertel, O.; Silver, J.D.; Villadsen, H. (Aarhus Univ., National Environmental Research Institute, Dept. of Atmospheric Environment, Roskilde (Denmark)); Soees Hansen, M. (Aarhus Univ., National Environmental Research Institute, Dept. of Policy Analysis, Roskilde (Denmark)); Skou Andersen, M. (European Environment Agency, Copenhagen (Denmark))

    2010-07-01

    The measured emissions of WP1 of the project has been applied as input for model calculations with the EVA model system. The DEHM model which calculates the regional scale delta-concentrations has been further developed to handle the low signal to noise ratio of the delta-concentrations related to the small sources that the gas fired engines constitute. All combinations of engine settings and locations have been run as scenarios with the EVA system, however the results have been grouped into themes to investigate changes related to location as well as changes related to engine settings. New exposure-response relations have been implemented in the system related to the chemical components nitrogen dioxide, formaldehyde, ethene and propene. The choice of high-exposure location in the calculations has unfortunately turned out to be less optimal. The location at Store Valby has previously been applied in studies with the EVA system as a high-exposure site, however in previous applications, the emission sources have been large power plants with stack heights of around 150 meters. The height of the stack of the gas fired engines is only around 30 meters, and the consequence is that the emitted components reach the surface closer to the stack, thereby giving high exposure in an area located further to the southwest, where the population density is not as high as in central Copenhagen. In general the marginal health costs (in Euro pr kg) of carbon monoxide and formaldehyde emissions are very small. The emissions of formaldehyde are also small and the resulting costs for this component is therefore very small. The emission of carbon monoxide is much larger, however the small marginal cost makes the contribution to the total costs small, also for this component. The marginal health costs of nitrogen oxides and ethene emissions show little variation with engine scenario. However the general picture is that as the NO{sub x} emissions increase (either by increasing ignition

  7. Validation of the EIR LWR calculation methods for criticality assessment of storage pools

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1986-11-01

    The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system to various types of critical experiments and benchmark problems proves its good accuracy, even for heterogeneous configurations containing strong neutron absorbers such as Boral. Since the multiplication factor k eff is normally somewhat overpredicted and the spread of the results is small, this code system is validated for the calculation of storage pools, taking into account a safety margins of 1.5% on k eff . (author)

  8. Parental criticism and externalizing behavior problems in adolescents: the role of environment and genotype-environment correlation.

    Science.gov (United States)

    Narusyte, Jurgita; Neiderhiser, Jenae M; Andershed, Anna-Karin; D'Onofrio, Brian M; Reiss, David; Spotts, Erica; Ganiban, Jody; Lichtenstein, Paul

    2011-05-01

    Genetic factors are important for the association between parental negativity and child problem behavior, but it is not clear whether this is due to passive or evocative genotype-environment correlation (rGE). In this study, we applied the extended children-of-twins model to directly examine the presence of passive and evocative rGE as well as direct environmental effects in the association between parental criticism and adolescent externalizing problem behavior. The cross-sectional data come from the Twin and Offspring Study in Sweden (N = 909 pairs of adult twins) and from the Twin Study of Child and Adolescent Development (N = 915 pairs of twin children). The results revealed that maternal criticism was primarily due to evocative rGE emanating from their adolescent's externalizing behavior. On the other hand, fathers' critical remarks tended to affect adolescent problem behavior in a direct environmental way. This suggests that previously reported differences in caretaking between mothers and fathers also are reflected in differences in why parenting is associated with externalizing behavior in offspring.

  9. Parental criticism and externalizing behavior problems in adolescents– the role of environment and genotype-environment correlation

    Science.gov (United States)

    Narusyte, Jurgita; Neiderhiser, Jenae M.; Andershed, Anna-Karin; D’Onofrio, Brian M.; Reiss, David; Spotts, Erica; Ganiban, Jody; Lichtenstein, Paul

    2011-01-01

    Genetic factors are important for the association between parental negativity and child problem behavior, but it is not clear whether this is dueto passive or evocative genotype-environment correlation (rGE). In this study we applied the extended children-of-twins model to directly examine the presence of passive and evocative rGE as well as direct environmental effects in the association between parental criticism and adolescent externalizing problem behavior. The cross-sectional data come from the Twin and Offspring Study in Sweden (TOSS) (N=909 pairs of adult twins) and from the Twin study of CHild and Adolescent Development (TCHAD) (N=915 pairs of twin children). The results revealed that maternal criticism was primarily due to evocative rGE emanating from their adolescent’s externalizing behavior. On the other hand, fathers’ critical remarks tended to affect adolescent problem behavior in a direct environmental way. This suggests that previously reported differences in caretaking between mothers and fathers also are reflected in differences in why parenting is associated with externalizing behavior in offspring. PMID:21280930

  10. An Improved Computational Method for the Calculation of Mixture Liquid-Vapor Critical Points

    Science.gov (United States)

    Dimitrakopoulos, Panagiotis; Jia, Wenlong; Li, Changjun

    2014-05-01

    Knowledge of critical points is important to determine the phase behavior of a mixture. This work proposes a reliable and accurate method in order to locate the liquid-vapor critical point of a given mixture. The theoretical model is developed from the rigorous definition of critical points, based on the SRK equation of state (SRK EoS) or alternatively, on the PR EoS. In order to solve the resulting system of nonlinear equations, an improved method is introduced into an existing Newton-Raphson algorithm, which can calculate all the variables simultaneously in each iteration step. The improvements mainly focus on the derivatives of the Jacobian matrix, on the convergence criteria, and on the damping coefficient. As a result, all equations and related conditions required for the computation of the scheme are illustrated in this paper. Finally, experimental data for the critical points of 44 mixtures are adopted in order to validate the method. For the SRK EoS, average absolute errors of the predicted critical-pressure and critical-temperature values are 123.82 kPa and 3.11 K, respectively, whereas the commercial software package Calsep PVTSIM's prediction errors are 131.02 kPa and 3.24 K. For the PR EoS, the two above mentioned average absolute errors are 129.32 kPa and 2.45 K, while the PVTSIM's errors are 137.24 kPa and 2.55 K, respectively.

  11. Validation of new 240Pu cross section and covariance data via criticality calculation

    International Nuclear Information System (INIS)

    Kim, Do Heon; Gil, Choong-Sup; Kim, Hyeong Il; Lee, Young-Ouk; Leal, Luiz C.; Dunn, Michael E.

    2011-01-01

    Recent collaboration between KAERI and ORNL has completed an evaluation for 240 Pu neutron cross section with covariance data. The new 240 Pu cross section data has been validated through 28 criticality safety benchmark problems taken from the ICSBEP and/or CSEWG specifications with MCNP calculations. The calculation results based on the new evaluation have been compared with those based on recent evaluations such as ENDF/B-VII.0, JEFF-3.1.1, and JENDL-4.0. In addition, the new 240 Pu covariance data has been tested for some criticality benchmarks via the DANTSYS/SUSD3D-based nuclear data sensitivity and uncertainty analysis of k eff . The k eff uncertainty estimates by the new covariance data has been compared with those by JENDL-4.0, JENDL-3.3, and Low-Fidelity covariance data. (author)

  12. The Davidson Method as an alternative to power iterations for criticality calculations

    International Nuclear Information System (INIS)

    Subramanian, C.; Van Criekingen, S.; Heuveline, V.; Nataf, F.; Have, P.

    2011-01-01

    The Davidson method is implemented within the neutron transport core solver parafish to solve k-eigenvalue criticality transport problems. The parafish solver is based on domain decomposition, uses spherical harmonics (P_N method) for angular discretization, and nonconforming finite elements for spatial discretization. The Davidson method is compared to the traditional power iteration method in that context. Encouraging numerical results are obtained with both sequential and parallel calculations. (author)

  13. Improvement of the skeleton tables for calculation of the critical heat load

    International Nuclear Information System (INIS)

    Gotovskij, M.A.; Kvetnyj, M.A.

    2002-01-01

    Paper presents analysis of drawbacks of the skeleton tables of the critical heat flows applied in calculated heat and hydraulic codes. Paper demonstrates the necessity to take account of specific nature of mechanisms of dryout crisis, of boiling crisis at slow mass rates and the range of small underheatings up to temperature of saturation. Attention is drawn to necessity of detailed account of the natural limitations of the application field of the skeleton tables [ru

  14. Sample size calculation to externally validate scoring systems based on logistic regression models.

    Directory of Open Access Journals (Sweden)

    Antonio Palazón-Bru

    Full Text Available A sample size containing at least 100 events and 100 non-events has been suggested to validate a predictive model, regardless of the model being validated and that certain factors can influence calibration of the predictive model (discrimination, parameterization and incidence. Scoring systems based on binary logistic regression models are a specific type of predictive model.The aim of this study was to develop an algorithm to determine the sample size for validating a scoring system based on a binary logistic regression model and to apply it to a case study.The algorithm was based on bootstrap samples in which the area under the ROC curve, the observed event probabilities through smooth curves, and a measure to determine the lack of calibration (estimated calibration index were calculated. To illustrate its use for interested researchers, the algorithm was applied to a scoring system, based on a binary logistic regression model, to determine mortality in intensive care units.In the case study provided, the algorithm obtained a sample size with 69 events, which is lower than the value suggested in the literature.An algorithm is provided for finding the appropriate sample size to validate scoring systems based on binary logistic regression models. This could be applied to determine the sample size in other similar cases.

  15. Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

    International Nuclear Information System (INIS)

    Whitesides, G.H.; Stephens, M.E.

    1984-01-01

    During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

  16. Prediction calculation of HTR-10 fuel loading for the first criticality

    International Nuclear Information System (INIS)

    Jing Xingqing; Yang Yongwei; Gu Yuxiang; Shan Wenzhi

    2001-01-01

    The 10 MW high temperature gas cooled reactor (HTR-10) was built at Institute of Nuclear Energy Technology, Tsinghua University, and the first criticality was attained in Dec. 2000. The high temperature gas cooled reactor physics simulation code VSOP was used for the prediction of the fuel loading for HTR-10 first criticality. The number of fuel element and graphite element was predicted to provide reference for the first criticality experiment. The prediction calculations toke into account the factors including the double heterogeneity of the fuel element, buckling feedback for the spectrum calculation, the effect of the mixture of the graphite and the fuel element, and the correction of the diffusion coefficients near the upper cavity based on the transport theory. The effects of impurities in the fuel and the graphite element in the core and those in the reflector graphite on the reactivity of the reactor were considered in detail. The first criticality experiment showed that the predicted values and the experiment results were in good agreement with little relative error less than 1%, which means the prediction was successful

  17. Evaluation of the accuracy of group calculations for reactor criticality perturbations

    International Nuclear Information System (INIS)

    Dulin, V.A.

    1985-09-01

    For calculations of criticality perturbations it is necessary to use group constants which take into account not only the peculiarities of the intra-group flux but also those of the behaviour of the adjoint flux. A new method is proposed for obtaining bilinear-averaged constants of this type on the basis of the resonance characteristics of the importance function and the difference between the value of neutron importance at the group boundary and the group-averaged value (the bsup(+j) factor). A number of calculations are made for the ratios of reactivity coefficients in the BFS assemblies. Values have been obtained for the difference between the results of calculation with bilinear-averaged constants and those averaged conventionally (over flux). In many cases, this difference exceeds the experimental error. (author)

  18. Intact and Degraded Component Criticality Calculations of N Reactor Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    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

  19. Automation of POST Cases via External Optimizer and "Artificial p2" Calculation

    Science.gov (United States)

    Dees, Patrick D.; Zwack, Mathew R.

    2017-01-01

    optimizer functions like any other gradient-based optimizer. It has a specified variable to optimize whose value is represented as optval, a set of dependent constraints to meet with associated forms and tolerances whose value is represented as p2, and a set of independent variables known as the u-vector to modify in pursuit of optimality. Each of these quantities are calculated or manipulated at a certain phase within the trajectory. The optimizer is further constrained by the requirement that the input u-vector must result in a trajectory which proceeds through each of the prescribed events in the input file. For example, if the input u-vector causes the vehicle to crash before it can achieve the orbital parameters required for a parking orbit, then the run will fail without engaging the optimizer, and a p2 value of exactly zero is returned. This poses a problem, as this "non-connecting" region of the u-vector space is far larger than the "connecting" region which returns a non-zero value of p2 and can be worked on by the internal optimizer. Finding this connecting region and more specifically the global optimum within this region has traditionally required the use of an expert analyst.

  20. Recent R and D around the Monte-Carlo code Tripoli-4 for criticality calculation

    International Nuclear Information System (INIS)

    Hugot, F.X.; Lee, Y.K.; Malvagi, F.

    2008-01-01

    TRIPOLI-4 [1] is the fourth generation of the TRIPOLI family of Monte Carlo codes developed from the 60's by CEA. It simulates the 3D transport of neutrons, photons, electrons and positrons as well as coupled neutron-photon propagation and electron-photons cascade showers. The code addresses radiation protection and shielding problems, as well as criticality and reactor physics problems through both critical and subcritical neutronics calculations. It uses full pointwise as well as multigroup cross-sections. The code has been validated through several hundred benchmarks as well as measurement campaigns. It is used as a reference tool by CEA as well as its industrial and institutional partners, and in the NURESIM [2] European project. Section 2 reviews its main features, with emphasis on the latest developments. Section 3 presents some recent R and D for criticality calculations. Fission matrix, Eigen-values and eigenvectors computations will be exposed. Corrections on the standard deviation estimator in the case of correlations between generation steps will be detailed. Section 4 presents some preliminary results obtained by the new mesh tally feature. The last section presents the interest of using XML format output files. (authors)

  1. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  2. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  3. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  4. Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations

    Directory of Open Access Journals (Sweden)

    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.

  5. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

  6. Determining the nuclear data uncertainty on MONK10 and WIMS10 criticality calculations

    Science.gov (United States)

    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.

  7. THE DETERMINATION OF A CRITICAL VALUE FOR DYNAMIC STABILITY OF SEMICONDUCTOR LASER DIODE WITH EXTERNAL OPTICAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Remzi YILDIRIM

    1998-01-01

    Full Text Available In this study, dynamic stability analysis of semiconductor laser diodes with external optical feedback has been realized. In the analysis the frequency response of the transfer function of laser diode H jw( , the transfer m function of laser diode with external optical feedback TF jw( , and optical feedback transfer function m K jw( obtained from small signal equations has been m accomplished using Nyquist stability analysis in complex domain. The effect of optical feedback on the stability of the system has been introduced and to bring the laser diode to stable condition the working critical boundary range of dampig frequency and reflection power constant (R has been determined. In the study the reflection power has been taken as ( .

  8. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

    International Nuclear Information System (INIS)

    Aggery, A.

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  9. Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.; Funabashi, H.

    1987-01-01

    Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

  10. Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.

    1987-01-01

    Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

  11. Static analysis of material testing reactor cores:critical core calculations

    International Nuclear Information System (INIS)

    Nawaz, A. A.; Khan, R. F. H.; Ahmad, N.

    1999-01-01

    A methodology has been described to study the effect of number of fuel plates per fuel element on critical cores of Material Testing Reactors (MTR). When the number of fuel plates are varied in a fuel element by keeping the fuel loading per fuel element constant, the fuel density in the fuel plates varies. Due to this variation, the water channel width needs to be recalculated. For a given number of fuel plates, water channel width was determined by optimizing k i nfinity using a transport theory lattice code WIMS-D/4. The dimensions of fuel element and control fuel element were determined using this optimized water channel width. For the calculated dimensions, the critical cores were determined for the given number of fuel plates per fuel element by using three dimensional diffusion theory code CITATION. The optimization of water channel width gives rise to a channel width of 2.1 mm when the number of fuel plates is 23 with 290 g ''2''3''5U fuel loading which is the same as in the case of Pakistan Reactor-1 (PARR-1). Although the decrease in number of fuel element results in an increase in optimal water channel width but the thickness of standard fuel element (SFE) and control fuel element (CFE) decreases and it gives rise to compact critical and equilibrium cores. The criticality studies of PARR-1 are in good agreement with the predictions

  12. Calculation and analysis of burnup and optimum core design in accelerator driven sub-critical system

    International Nuclear Information System (INIS)

    Wang Yuwei; Yang Yongwei; Cui Pengfei

    2011-01-01

    The premise of the accelerator driven sub-critical system (ADS) in the accident is still subcritical, the biggest k eff change with burn time is less than 1.5% and the cladding material, HT9 steel, can withstand the maximum radiation damage, core fuel area is divided into fuel transmutation area and fuel multiplication area, and fuel transmutation area maintains the same fuel composition in the whole process. Through the analysis of the composition of the fuel, shape of core layout and the power distribution, etc., supposed outer and inner Pu enrichment ratio range of 1.0-1.5, then the fuel components of fuel multiplication area was adjusted. Time evolution of k eff was calculated by COUPLED2 which coupled with MCNP and ORIGEN. At the same time the power peaking factors, minoractinides transmutation rate desired to maximization and burnup were considered. A sub-critical system fitting for engineering practice was established. (authors)

  13. A novel source convergence acceleration scheme for Monte Carlo criticality calculations, part I: Theory

    International Nuclear Information System (INIS)

    Griesheimer, D. P.; Toth, B. E.

    2007-01-01

    A novel technique for accelerating the convergence rate of the iterative power method for solving eigenvalue problems is presented. Smoothed Residual Acceleration (SRA) is based on a modification to the well known fixed-parameter extrapolation method for power iterations. In SRA the residual vector is passed through a low-pass filter before the extrapolation step. Filtering limits the extrapolation to the lower order Eigenmodes, improving the stability of the method and allowing the use of larger extrapolation parameters. In simple tests SRA demonstrates superior convergence acceleration when compared with an optimal fixed-parameter extrapolation scheme. The primary advantage of SRA is that it can be easily applied to Monte Carlo criticality calculations in order to reduce the number of discard cycles required before a stationary fission source distribution is reached. A simple algorithm for applying SRA to Monte Carlo criticality problems is described. (authors)

  14. Development and External Validation of the Korean Prostate Cancer Risk Calculator for High-Grade Prostate Cancer: Comparison with Two Western Risk Calculators in an Asian Cohort.

    Science.gov (United States)

    Park, Jae Young; Yoon, Sungroh; Park, Man Sik; Choi, Hoon; Bae, Jae Hyun; Moon, Du Geon; Hong, Sung Kyu; Lee, Sang Eun; Park, Chanwang; Byun, Seok-Soo

    2017-01-01

    We developed the Korean Prostate Cancer Risk Calculator for High-Grade Prostate Cancer (KPCRC-HG) that predicts the probability of prostate cancer (PC) of Gleason score 7 or higher at the initial prostate biopsy in a Korean cohort (http://acl.snu.ac.kr/PCRC/RISC/). In addition, KPCRC-HG was validated and compared with internet-based Western risk calculators in a validation cohort. Using a logistic regression model, KPCRC-HG was developed based on the data from 602 previously unscreened Korean men who underwent initial prostate biopsies. Using 2,313 cases in a validation cohort, KPCRC-HG was compared with the European Randomized Study of Screening for PC Risk Calculator for high-grade cancer (ERSPCRC-HG) and the Prostate Cancer Prevention Trial Risk Calculator 2.0 for high-grade cancer (PCPTRC-HG). The predictive accuracy was assessed using the area under the receiver operating characteristic curve (AUC) and calibration plots. PC was detected in 172 (28.6%) men, 120 (19.9%) of whom had PC of Gleason score 7 or higher. Independent predictors included prostate-specific antigen levels, digital rectal examination findings, transrectal ultrasound findings, and prostate volume. The AUC of the KPCRC-HG (0.84) was higher than that of the PCPTRC-HG (0.79, pexternal validation. Calibration plots also revealed better performance of KPCRC-HG and ERSPCRC-HG than that of PCPTRC-HG on external validation. At a cut-off of 5% for KPCRC-HG, 253 of the 2,313 men (11%) would not have been biopsied, and 14 of the 614 PC cases with Gleason score 7 or higher (2%) would not have been diagnosed. KPCRC-HG is the first web-based high-grade prostate cancer prediction model in Korea. It had higher predictive accuracy than PCPTRC-HG in a Korean population and showed similar performance with ERSPCRC-HG in a Korean population. This prediction model could help avoid unnecessary biopsy and reduce overdiagnosis and overtreatment in clinical settings.

  15. Monte Carlo sampling on technical parameters in criticality and burn-up-calculations

    International Nuclear Information System (INIS)

    Kirsch, M.; Hannstein, V.; Kilger, R.

    2011-01-01

    The increase in computing power over the recent years allows for the introduction of Monte Carlo sampling techniques for sensitivity and uncertainty analyses in criticality safety and burn-up calculations. With these techniques it is possible to assess the influence of a variation of the input parameters within their measured or estimated uncertainties on the final value of a calculation. The probabilistic result of a statistical analysis can thus complement the traditional method of figuring out both the nominal (best estimate) and the bounding case of the neutron multiplication factor (k eff ) in criticality safety analyses, e.g. by calculating the uncertainty of k eff or tolerance limits. Furthermore, the sampling method provides a possibility to derive sensitivity information, i.e. it allows figuring out which of the uncertain input parameters contribute the most to the uncertainty of the system. The application of Monte Carlo sampling methods has become a common practice in both industry and research institutes. Within this approach, two main paths are currently under investigation: the variation of nuclear data used in a calculation and the variation of technical parameters such as manufacturing tolerances. This contribution concentrates on the latter case. The newly developed SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool) is introduced. It defines an interface to the well established GRS tool for sensitivity and uncertainty analyses SUSA, that provides the necessary statistical methods for sampling based analyses. The interfaced codes are programs that are used to simulate aspects of the nuclear fuel cycle, such as the criticality safety analysis sequence CSAS5 of the SCALE code system, developed by Oak Ridge National Laboratories, or the GRS burn-up system OREST. In the following, first the implementation of the SUnCISTT will be presented, then, results of its application in an exemplary evaluation of the neutron

  16. Critical experiment tests of bowing and expansion reactivity calculations for LMRS

    International Nuclear Information System (INIS)

    Schaefer, R.W.

    1988-01-01

    Experiments done in several LMR-type critical assemblies simulated core axial expansion, core radial expansion and bowing, coolant expansion, and control driveline expansion. For the most part new experimental techniques were developed to do these experiments. Calculations of the experiments basically used design-level methods, except when it was necessary to investigate complexities peculiar to the experiments. It was found that these feedback reactivities generally are overpredicted, but the predictions are within 30% of the experimental values. 14 refs., 2 figs., 4 tabs

  17. Investigating the minimum achievable variance in a Monte Carlo criticality calculation

    Energy Technology Data Exchange (ETDEWEB)

    Christoforou, Stavros; Eduard Hoogenboom, J. [Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands)

    2008-07-01

    The sources of variance in a Monte Carlo criticality calculation are identified and their contributions analyzed. A zero-variance configuration is initially simulated using analytically calculated adjoint functions for biasing. From there, the various sources are analyzed. It is shown that the minimum threshold comes from the fact that the fission source is approximated. In addition, the merits of a simple variance reduction method, such as implicit capture, are shown when compared to an analog simulation. Finally, it is shown that when non-exact adjoint functions are used for biasing, the variance reduction is rather insensitive to the quality of the adjoints, suggesting that the generation of the adjoints should have as low CPU cost as possible, in order to o et the CPU cost in the implementation of the biasing of a simulation. (authors)

  18. Calculational criticality analyses of 10- and 20-MW UF6 freezer/sublimer vessels

    International Nuclear Information System (INIS)

    Jordan, W.C.

    1993-02-01

    Calculational criticality analyses have been performed for 10- and 20-MW UF 6 freezer/sublimer vessels. The freezer/sublimers have been analyzed over a range of conditions that encompass normal operation and abnormal conditions. The effects of HF moderation of the UF 6 in each vessel have been considered for uranium enriched between 2 and 5 wt % 235 U. The results indicate that the nuclearly safe enrichments originally established for the operation of a 10-MW freezer/sublimer, based on a hydrogen-to-uranium moderation ratio of 0.33, are acceptable. If strict moderation control can be demonstrated for hydrogen-to-uranium moderation ratios that are less than 0.33, then the enrichment limits for the 10-MW freezer/sublimer may be increased slightly. The calculations performed also allow safe enrichment limits to be established for a 20-NM freezer/sublimer under moderation control

  19. Calculation and analysis for a series of enriched uranium bare sphere critical assemblies

    International Nuclear Information System (INIS)

    Yang Shunhai

    1994-12-01

    The imported reactor fuel assembly MARIA program system is adapted to CYBER 825 computer in China Institute of Atomic Energy, and extensively used for a series of enriched uranium bare sphere critical assemblies. The MARIA auxiliary program of resonance modification MA is designed for taking account of the effects of resonance fission and absorption on calculated results. By which, the multigroup constants in the library attached to MARIA program are revised based on the U.S. Evaluated Nuclear Data File ENDF/B-IV, the related nuclear data files are replaced. And then, the reactor geometry buckling and multiplication factor are given in output tapes. The accuracy of calculated results is comparable with those of Monte Carlo and Sn method, and the agreement with experiment result is in 1%. (5 refs., 4 figs., 3 tabs.)

  20. Validation of VHTRC calculation benchmark of critical experiment using the MCB code

    Directory of Open Access Journals (Sweden)

    Stanisz Przemysław

    2016-01-01

    Full Text Available The calculation benchmark problem Very High Temperature Reactor Critical (VHTR a pin-in-block type core critical assembly has been investigated with the Monte Carlo Burnup (MCB code in order to validate the latest version of Nuclear Data Library based on ENDF format. Executed benchmark has been made on the basis of VHTR benchmark available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments. This benchmark is useful for verifying the discrepancies in keff values between various libraries and experimental values. This allows to improve accuracy of the neutron transport calculations that may help in designing the high performance commercial VHTRs. Almost all safety parameters depend on the accuracy of neutron transport calculation results that, in turn depend on the accuracy of nuclear data libraries. Thus, evaluation of the libraries applicability to VHTR modelling is one of the important subjects. We compared the numerical experiment results with experimental measurements using two versions of available nuclear data (ENDF-B-VII.1 and JEFF-3.2 prepared for required temperatures. Calculations have been performed with the MCB code which allows to obtain very precise representation of complex VHTR geometry, including the double heterogeneity of a fuel element. In this paper, together with impact of nuclear data, we discuss also the impact of different lattice modelling inside the fuel pins. The discrepancies of keff have been successfully observed and show good agreement with each other and with the experimental data within the 1 σ range of the experimental uncertainty. Because some propagated discrepancies observed, we proposed appropriate corrections in experimental constants which can improve the reactivity coefficient dependency. Obtained results confirm the accuracy of the new Nuclear Data Libraries.

  1. Accuracy of WWR-M criticality calculations with code MCU-RFFI

    International Nuclear Information System (INIS)

    Petrov, Yu.V.; Erykalov, A.N.; Onegin, M.S.

    1999-01-01

    The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about Δρ≅±0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, α) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

  2. Accuracy of WWR-M criticality calculations with code MCU-RFFI

    Energy Technology Data Exchange (ETDEWEB)

    Petrov, Yu V [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation); Erykalov, A N; Onegin, M S [Petersburg Nuclear Physics Institute RAS, 188350 Gatchina, St. Petersburg (Russian Federation)

    1999-10-01

    The scattering and deviation of fuel element parameters by manufacturing, approximations of the reactor structure in the computer model, the partly inadequate neutron cross sections in the computer codes etc. lead to a discrepancy between the reactivity computations and data. We have compared reactivity calculations using the MCU-RRFI Monte Carlo code of critical assemblies containing WWR-M2 (36 enriched) an WWR-M5 (90%) fuel elements with benchmark experiments. The agreement was about {delta}{rho}{approx_equal}{+-}0.3%. A strong influence of the water ratio on reactivity was shown and a significant heterogeneous effect was found. We have also investigated, by full scale reactor calculations for the RETR program, the contribution to the reactivity of the main reactor structure elements: beryllium reflector, experimental channels irradiation devices inside the core, etc. Calculations show the importance of a more thorough study of the contributions of products of the (n, {alpha}) reaction in the Be reflector to the reactivity. Ways of improving the accuracy of the calculations are discussed. (author)

  3. THE ROMANIAN EXTERNAL TRADE AND THE FOREIGN DIRECT INVESTMENTS INFLOWS AFTER 2007. A CRITICAL SURVEY.

    Directory of Open Access Journals (Sweden)

    Giurgiu Adriana

    2011-12-01

    Full Text Available Focusing more on the evolution of the Romanian external trade and FDI inflows after 2007 - the year of Romania's EU integration - the paper provides insights in the trade and FDI inflows determinants in the context of the convergence process and also in relation to the financial global crisis. As the statistic data envisage, Romanian exports and imports were high over the period 2000-2010, and there were also large FDI inflows in the economy, mostly until 2008. Although the data appear to be encouraging, it seems that Romania did not succeed to maximize the benefits which should result from such a situation. Moreover, the situation after 2008 shows that Romania is in a deep recession accelerated and maintained by uninspired government policy measures with a strong negative impact on the Romanian economy. The aim of this paper is to empirically investigate Romanian export and import demand functions after the year 2007, and to compare them with those of the period 2000-2006, using quarterly data, in correlation with the FDI inflows. The main contribution of the paper is that it is an empirical analysis on Romania's FDI and external trade, providing the impact of the main determinants of export and import of Romania, by using the Engle-Granger two step method. Following Allard (2009 rather than just providing the elasticity, this method combines the elasticity with the evolution of the explanatory variables to quantify their impact during the analysed period. The analysis aim to cover all possible factors underlying the external sector performance of Romania and thus, they are complemented with country specific analysis. The empirical analysis will therefore provide some interesting insights not only in the context of the convergence process of the country with the Eurozone but also in relation to the exchange rate regime. The theoretical framework relies on the 'imperfect substitutes' model (Goldstein and Khan; 1985, and it is used in the paper

  4. The Criticality of Connecting People to Financial Results ‒ an ROI Calculation Model for Romanian FSOs

    Directory of Open Access Journals (Sweden)

    Daniela Niculescu

    2016-02-01

    Full Text Available Organisational culture and employee engagement have been the focus of recent broad-based research efforts. Adding this concern to the revealed importance of performance indicators on human capital, and that their use is getting momentum, in order to attach financial values to knowledge management assets, it becomes more and more critical to measure human capital value. Key for Romanian FSO’s managers becomes to consider that both human and financial values have a focus on adding value in every process and function in the organisation, and to perpetuate organisational profitability by the corporate culture, on the one hand, where culture is a powerful factor that helps a company to engage, on the other hand, talented people. There is a substantial concern on using ROI on Learning and Development programmes, but whilst this is still declared, Romanian FSOs do not yet have a consistent method to measure it. This study is showing the criticality of connecting people to financial results and data analysis suggests that ROI calculation has a positive impact on creating and fostering a powerful organisational culture and that employees’ awareness of ROI values within their organisation has a powerful effect on their sense of engagement. Our findings have a more practical implication for the analysed industry by shaping a formal ROI measurement mechanisms blueprint, an ROI calculation model for the Romanian FSOs, in the form of a mechanism that could be employed when considering the design of an ROI Methodology for Romanian FSOs.

  5. Three calculations of free cortisol versus measured values in the critically ill.

    Science.gov (United States)

    Molenaar, Nienke; Groeneveld, A B Johan; de Jong, Margriet F C

    2015-11-01

    To investigate the agreement between the calculated free cortisol levels according to widely applied Coolens and adjusted Södergård equations with measured levels in the critically ill. A prospective study in a mixed intensive care unit. We consecutively included 103 patients with treatment-insensitive hypotension in whom an adrenocorticotropic hormone (ACTH) test (250μg) was performed. Serum total and free cortisol (equilibrium dialysis), corticosteroid-binding globulin and albumin were assessed. Free cortisol was estimated by the Coolens method (C) and two adjusted Södergård (S1 and S2) equations. Bland Altman plots were made. The bias for absolute (t=0, 30 and 60min after ACTH injection) cortisol levels was 38, -24, 41nmol/L when the C, S1 and S2 equations were used, with 95% limits of agreement between -65-142, -182-135, and -57-139nmol/L and percentage errors of 66, 85, and 64%, respectively. Bias for delta (peak-baseline) cortisol was 14, -31 and 16nmol/L, with 95% limits of agreement between -80-108, -157-95, and -74-105nmol/L, and percentage errors of 107, 114, and 100% for C, S1 and S2 equations, respectively. Calculated free cortisol levels have too high bias and imprecision to allow for acceptable use in the critically ill. Copyright © 2015 The Canadian Society of Clinical Chemists. Published by Elsevier Inc. All rights reserved.

  6. Simulation of ferromagnetic shielding to the critical current of Bi2223/Ag tape under external fields

    Energy Technology Data Exchange (ETDEWEB)

    Gu Chen [Applied Superconductivity Research Center, Tsinghua University, Beijing 100084 (China); Alamgir, A K M [Faculty of Engineering, Yokohama National University, 75-9 Tokiwadai, Hodogaya-ku, Yokohama (Japan); Qu Timing [Applied Superconductivity Research Center, Tsinghua University, Beijing 100084 (China); Han, Z [Applied Superconductivity Research Center, Tsinghua University, Beijing 100084 (China)

    2007-03-15

    Ferromagnetic material was electroplated onto the surface of the Bi2223/Ag multi-filamentary tape and as a result changed the I{sub c}(B) characteristic of the tape correspondingly. A numerical simulation was used to investigate the influence of ferromagnetic shielding (FS) on the performance of the tape, in particular the I{sub c} behaviour under perpendicular external fields. Using finite element analysis, we are able to understand how FS alters the flux distribution within the superconductor region under any arbitrary shielding structure. The shielding width, thickness and nonlinear property of the ferromagnetic material were taken into account for the simulation. Finally, optimized shielding parameters in association with different operating fields were suggested and these values could be considered for the next run of experimental work.

  7. Simulation of ferromagnetic shielding to the critical current of Bi2223/Ag tape under external fields

    International Nuclear Information System (INIS)

    Gu Chen; Alamgir, A K M; Qu Timing; Han, Z

    2007-01-01

    Ferromagnetic material was electroplated onto the surface of the Bi2223/Ag multi-filamentary tape and as a result changed the I c (B) characteristic of the tape correspondingly. A numerical simulation was used to investigate the influence of ferromagnetic shielding (FS) on the performance of the tape, in particular the I c behaviour under perpendicular external fields. Using finite element analysis, we are able to understand how FS alters the flux distribution within the superconductor region under any arbitrary shielding structure. The shielding width, thickness and nonlinear property of the ferromagnetic material were taken into account for the simulation. Finally, optimized shielding parameters in association with different operating fields were suggested and these values could be considered for the next run of experimental work

  8. Renormalization and Central limit theorem for critical dynamical systems with weak external noise

    CERN Document Server

    Diaz-Espinosa, O

    2006-01-01

    We study of the effect of weak noise on critical one dimensional maps; that is, maps with a renormalization theory. We establish a one dimensional central limit theorem for weak noises and obtain Berry--Esseen estimates for the rate of this convergence. We analyze in detail maps at the accumulation of period doubling and critical circle maps with golden mean rotation number. Using renormalization group methods, we derive scaling relations for several features of the effective noise after long times. We use these scaling relations to show that the central limit theorem for weak noise holds in both examples. We note that, for the results presented here, it is essential that the maps have parabolic behavior. They are false for hyperbolic orbits.

  9. Critical groups vs. representative person: dose calculations due to predicted releases from USEXA

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, N.L.D., E-mail: nelson.luiz@ctmsp.mar.mil.br [Centro Tecnologico da Marinha (CTM/SP), Sao Paulo, SP (Brazil); Rochedo, E.R.R., E-mail: elainerochedo@gmail.com [Instituto de Radiprotecao e Dosimetria (lRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Mazzilli, B.P., E-mail: mazzilli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The critical group cf Centro Experimental Aramar (CEA) site was previously defined based 00 the effluents releases to the environment resulting from the facilities already operational at CEA. In this work, effective doses are calculated to members of the critical group considering the predicted potential uranium releases from the Uranium Hexafluoride Production Plant (USEXA). Basically, this work studies the behavior of the resulting doses related to the type of habit data used in the analysis and two distinct situations are considered: (a) the utilization of average values obtained from official institutions (IBGE, IEA-SP, CNEN, IAEA) and from the literature; and (b) the utilization of the 95{sup tb} percentile of the values derived from distributions fit to the obtained habit data. The first option corresponds to the way that data was used for the definition of the critical group of CEA done in former assessments, while the second one corresponds to the use of data in deterministic assessments, as recommended by ICRP to estimate doses to the so--called 'representative person' . (author)

  10. Critical groups vs. representative person: dose calculations due to predicted releases from USEXA

    International Nuclear Information System (INIS)

    Ferreira, N.L.D.; Rochedo, E.R.R.; Mazzilli, B.P.

    2013-01-01

    The critical group cf Centro Experimental Aramar (CEA) site was previously defined based 00 the effluents releases to the environment resulting from the facilities already operational at CEA. In this work, effective doses are calculated to members of the critical group considering the predicted potential uranium releases from the Uranium Hexafluoride Production Plant (USEXA). Basically, this work studies the behavior of the resulting doses related to the type of habit data used in the analysis and two distinct situations are considered: (a) the utilization of average values obtained from official institutions (IBGE, IEA-SP, CNEN, IAEA) and from the literature; and (b) the utilization of the 95 tb percentile of the values derived from distributions fit to the obtained habit data. The first option corresponds to the way that data was used for the definition of the critical group of CEA done in former assessments, while the second one corresponds to the use of data in deterministic assessments, as recommended by ICRP to estimate doses to the so--called 'representative person' . (author)

  11. Conservatism in SRS Criticality Alarm System 12 Rad Zone Calculations - How Much is Enough?

    International Nuclear Information System (INIS)

    Yates, K.R.

    2002-01-01

    Savannah River Site (SRS) uses two methods (i.e., Approximate Method and MCNP) of calculating the 12-rad zone. The reasons for the two-tier approach are described in Ref. 1 and 2. Lately, there have been occasions in which the use of either the Approximate Method (AM) or MCNP3 calculations indicated potential facility impacts. For example, one or both methods may indicate that a 12-rad zone extends outside of relatively thick shielding, or extends to the roof of a facility, or extends through shielding to part of a stairwell. In such cases, a criticality alarm system may have to be installed to protect workers in a small, localized area from a potential dose that is not substantially greater than 12 rad in air. But, is the potential dose really greater than 12 rad in air? A subcommittee was appointed to look into the two 12-rad zone calculation methods for the purpose of identifying items contributing to over-conservatism and under-conservatism, and to recommend a path forward

  12. Coupled core criticality calculations with control rods located in the central reflector region

    Energy Technology Data Exchange (ETDEWEB)

    Sobhy, M [Reactor depatrment, nuclear research center, Inshaas (Egypt)

    1995-10-01

    The reactivity of a coupled core is controlled by a set of control rods distributed in the central reflector region. The reactor contains two compact cores cooled and moderated by light water. Control rods are designed to have reactivity worths sufficient to start, control and shutdown the coupled system. Each core in a coupled system is in subcritical conditions without any absorber then each core needs to the other core to fulfill nuclear chain reaction and to approach the criticality. In this case, each core is considered clean which is suitable for research reactor with low flux disturbance and better neutron economy, in addition to the advantage of disappearing the cut corner fuel baskets. This facilitate the in core fuel management with identical fuel baskets. Hot spots will disappear. This leads to a good heat transfer process. the excess reactivity and the shutdown margin are calculated for some of reflector as coupling region gives sufficient area for coupled core are calculated cost. The fluctuations of reactivity for coupled core are calculated by noise analysis technique and compared with that for rode core. The results show low reactivity perturbation associated with coupled core.

  13. Verification of ANISN-F by calculating the neutron distribution from a Ra-Be source in water as well as by simple criticality calculations

    International Nuclear Information System (INIS)

    Etemad, M.A.

    1981-04-01

    The one dimensional discrete ordinates code ANISN-F was used to calculate the thermal neutron flux distribution in water from a Ra-Be neutron source. The calculations were performed in order to investigate the different possibilities of the code as well as to verify the results of the calculations in terms of comparisons to corresponding experimental data. Two different group cross section libraries were used in the calculations and conclusions were drawn on the adequacy of these libraries for a fixed source type calculation. Furthermore, critically calculations were performed for an infinite homogeneous slab of multiplying material using different angular and spatial approximations. The results of these calculations were then compared to the corresponding results previously obtained at this department by a different method and a different code. (author)

  14. Critical Causes and Consequences of Construction Project Interruption: Client, Contractor, Consultant and External Factors Standpoints

    Directory of Open Access Journals (Sweden)

    Nurbasirah Mohamed Alias

    2013-09-01

    Full Text Available It should be pointed out that the dilemma of interruption and stoppage in the construction industry is a large-scale trend. Construction interruption and stoppage is considered one of the most recurring problems in the construction industry. Construction Interruption and stoppage is always measured as costly to all parties concerned in the projects and very often it will result in clash, claims, total desertion and much difficult for the feasibility and it slows the growth of construction sector. The objective of the research work that underpins this paper was to investigate the Factors and Consequences of construction projects on local contractors. A construction project is commonly acknowledged as successful when the aim of the project is achieved in terms of predetermined objectives that are mainly completed the project on time, within budget and specified quality in accordance with the specifications and to stakeholders’ satisfaction. One of the most important problems that may arise in the construction project is delays and the magnitude of these delays varies considerably from project to project. According to delay categories that were contractor related, client related, consultant related, labour related and external related, the study revealed the six major effects of delay that were time overrun, cost overrun, dispute, arbitration, total abandonment, and litigation.

  15. Criticality Model

    International Nuclear Information System (INIS)

    Alsaed, A.

    2004-01-01

    The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality

  16. Global phase equilibrium calculations: Critical lines, critical end points and liquid-liquid-vapour equilibrium in binary mixtures

    DEFF Research Database (Denmark)

    Cismondi, Martin; Michelsen, Michael Locht

    2007-01-01

    A general strategy for global phase equilibrium calculations (GPEC) in binary mixtures is presented in this work along with specific methods for calculation of the different parts involved. A Newton procedure using composition, temperature and Volume as independent variables is used for calculation...

  17. Critical evaluation of the external occupational exposure in nuclear medicine services in Brazil

    International Nuclear Information System (INIS)

    Kubo, Ana Luiza Silva Lima

    2016-01-01

    Currently in Brazil (2016), there are 421 Nuclear Medicine Services (NMS). In nuclear medicine, the possibility of occupational internal contamination and external exposure is unavoidable. The chest individual monitoring, to estimate the effective dose, is mandatory, but the extremity monitoring is not always made. The aim of this study was to conduct a survey of data for external exposure of NMS professionals in Brazil from 1987 to 2010, analysing them in terms of trends and comparing them with measurements carried out in this work and in other countries. Although most of the NMS is still located in large urban centres (54% in the Southeast region), there is no state without any NMS. The increasing number of NMS has generated the need for more professionals. In the year 1987, they were 755 workers and, in 2010, 4134, with the following distribution of specialties: 29% of Nuclear Medicine Technicians (NMT), 23% of Nursing professionals, 29% of Physicians and 3% of Physicists. The average annual effective dose reached more than 3.0 mSv in some regions of the country, from 1987 to 2010, but tends to 1.0 mSv in 2010. The highest doses, as expected, are received by NMT and Nursing. The professionals who handle radiopharmaceuticals have their hands much more exposed than the chest. During 2010, only 31% of NMT and 16% of Nursing used extremity dosimeters as compared to chest dosimeters. The data from the measurements indicate that not all individual dosimeters are used properly. Generally, both in the measurements as in national registries, the hand doses were higher for professionals who prepared the radiopharmaceutical (NMT) than those who injected (Nursing). The value measured by chest dosimeters can be used to estimate the equivalent dose to the eye lenses, except for NMT at preparation practices at conventional NMS, where the equivalent dose of the lens is about 2 times higher than the dose at the chest. The most exposed areas of the hands are the tips of the index

  18. Study of the External Neutron Source Effect on TRU Burning in a Sub-critical Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zafar, Zafar Iqbal; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    One of the drawback points of nuclear power is the production of highly radioactive and long lasting waste isotopes during power production. Therefore, most important design requirement of future nuclear option should have a potential to burn selectively long-lived fission products (LLFP) and long-lived minor actinides (LLMA). However, there is no way to burn them selectively in the reactor core. Practical method of waste transmutation should rely on selective separation of them from spent nuclear fuel of power plants. Under the proliferation concern, direct separation of trans-uranic isotopes (TRU) from pyro-reprocessing plant became a feasible option in our country. Even though social political agreement is not matured as well as technical feasibility, current study is done based on basic assumptions; TRU and LLFP is separated from spent fuel of nuclear power plants. The remaining neutrons (among the external 3%) very few in number (less than 1% in any case) being very energetic (above three MeV or so) do cause much more fissions per neutron than their counterparts but, because of their overall low population they do not have any significant and decisive influence in the overall reactor performance. Currently, entire study is limited to the source neutron energy of 20 MeV only. In future, it is expected to get reasonably plausible fixed source dependent difference in the TRU burning by using tabulated data for the neutrons of higher energy (up to 250 MeV at least). Secondly, a clearer picture is expected if the TRU loading was increased from the current value of 133 kg to few metric tons, as is the case in most of the existing reactors.

  19. Calculation of multiplication factors regarding criticality aiming at the storage of fissile material

    International Nuclear Information System (INIS)

    Lima Barros, M. de.

    1982-04-01

    The multiplication factors of several systems with low enrichment, 3,5% and 3,2% in the isotope 235 U, aiming at the storage of fuel of ANGRA-I and ANGRA II, through the method of Monte Carlo, by the computacional code KENO-IV and the library of section of cross Hansen - Roach with 16 groups of energy. The method of Monte Carlo is specially suitable to the calculation of the factor of multiplication, because it is one of the most acurate models of solution and allows the description of complex tridimensional systems. Various tests of sensibility of this method have been done in order to present the most convenient way of working with KENO-IV code. The safety on criticality of stores of fissile material of the 'Fabrica de Elementos Combustiveis ', has been analyzed through the method of Monte Carlo. (Author) [pt

  20. Critical channel power calculation for nominal operation in the CNE (Embalse nuclear power plant): sensitivity study

    International Nuclear Information System (INIS)

    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)

  1. Riding Bare-Back on unstructured meshes for 21. century criticality calculations - 244

    International Nuclear Information System (INIS)

    Kelley, K.C.; Martz, R.L.; Crane, D.L.

    2010-01-01

    MCNP has a new capability that permits tracking of neutrons and photons on an unstructured mesh which is embedded as a mesh universe within its legacy geometry capability. The mesh geometry is created through Abaqus/CAE using its solid modeling capabilities. Transport results are calculated for mesh elements through a path length estimator while element to element tracking is performed on the mesh. The results from MCNP can be exported to Abaqus/CAE for visualization or other-physics analysis. The simple Godiva criticality benchmark problem was tested with this new mesh capability. Computer run time is proportional to the number of mesh elements used. Both first and second order polyhedrons are used. Models that used second order polyhedrons produced slightly better results without significantly increasing computer run time. Models that used first order hexahedrons had shorter runtimes than models that used first order tetrahedrons. (authors)

  2. Neutron flux calculations for criticality safety analysis using the narrow resonance approximations. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.

  3. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.; Parks, C.V. [Oak Ridge National Lab., TN (United States); Brady, M.C. [Sandia National Labs., Las Vegas, NV (United States)

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155.

  4. OECD/NEA burnup credit calculational criticality benchmark Phase I-B results

    International Nuclear Information System (INIS)

    DeHart, M.D.; Parks, C.V.; Brady, M.C.

    1996-06-01

    In most countries, criticality analysis of LWR fuel stored in racks and casks has assumed that the fuel is fresh with the maximum allowable initial enrichment. This assumption has led to the design of widely spaced and/or highly poisoned storage and transport arrays. If credit is assumed for fuel burnup, initial enrichment limitations can be raised in existing systems, and more compact and economical arrays can be designed. Such reliance on the reduced reactivity of spent fuel for criticality control is referred to as burnup credit. The Burnup Credit Working Group, formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development, has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods agree to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods agree within 11% about the average for all fission products studied. Most deviations are less than 10%, and many are less than 5%. The exceptions are Sm 149, Sm 151, and Gd 155

  5. Validating criticality calculations for spent fuel with 252Cf-source-driven noise measurements

    International Nuclear Information System (INIS)

    Mihalczo, J.T.; Krass, A.W.; Valentine, T.E.

    1992-01-01

    The 252 Cf-Source-driven noise analysis method can be used for measuring the subcritical neutron multiplication factor k of arrays of spent light water reactor (LWR) fuel. This type of measurement provides a parameter that is directly related to the criticality state of arrays of LWR fuel. Measurements of this parameter can verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. The practicality of a measurement depends on the ability to install the hardware required to perform the measurement. Source chambers containing the 252 Cf at the required source intensity for this application have been constructed and have operated successfully for ∼10 years and can be fabricated to fit into control rod guide tubes of PWR fuel elements. Fission counters especially developed for spent-fuel measurements are available that would allow measurements of a special 3 x 3 spent fuel array and a typical burnup credit rail cask with spent fuel in unborated water. Adding a moderator around these fission counters would allow measurements with the typical burnup credit rail cask with borated water and the special 3 x 3 array with borated water. The recent work of Ficaro on modifying the KENO Va code to calculate by the Monte Carlo method the time sequences of pulses at two detectors near a fissile assembly from the fission chain multiplication process, initiated by a 252 Cf source in the assembly allows a direct computer calculation of the noise analysis data from this measurement method

  6. A simplified treatment of radial enrichment distributions of LWR fuel assemblies in criticality calculations

    International Nuclear Information System (INIS)

    Hennebach, M.; Schnorrenberg, N.

    2008-01-01

    Criticality safety assessments are usually performed for fuel assembly models that are as generic as possible to encompass small modifications in geometry that have no impact on criticality. Dealing with different radial enrichment distributions for a fuel assembly type, which is especially important for BWR fuel, poses more of a challenge, since this characteristic is rather obviously influencing the neutronic behaviour of the system. Nevertheless, the large variability of enrichment distributions makes it very desirable and even necessary to treat them in a generalized way, both to keep the criticality safety assessment from becoming too unwieldy and to avoid having to extend it every time a new variation comes up. To be viable, such a generic treatment has to be demonstrably covering, i.e. lead to a higher effective neutron multiplication factor k eff than any of the radial enrichment distributions it represents. Averaging the enrichment evenly over the fuel rods of the assembly is a general and simple approach, and under reactor conditions, it is also a covering assumption: the graded distribution is introduced to achieve a linear power distribution, therefore reducing the enrichment of the better moderated rods at the edge of the assembly. With an even distribution of the average enrichment over all rods, these wellmoderated rods will cause increased fission rates at the assembly edges and a rise in k eff . Since the moderator conditions in a spent nuclear fuel cask differ strongly from a reactor even when considering optimal moderation, the proof that a uniform enrichment distribution is a covering assumption compared with detailed enrichment distributions has to be cask-specific. In this report, a method for making that proof is presented along with results for fuel assemblies from BWR reactors. All results are from three-dimensional Monte Carlo calculations with the SCALE 5.1 code package [1], using a 44-group neutron crosssection library based on ENDF

  7. A case study and critical assessment in calculating power usage effectiveness for a data centre

    International Nuclear Information System (INIS)

    Brady, Gemma A.; Kapur, Nikil; Summers, Jonathan L.; Thompson, Harvey M.

    2013-01-01

    Highlights: • A case study PUE calculation is carried out on a data centre by using open source specifications. • The PUE metric does not drive improvements in the efficiencies of IT processes. • The PUE does not fairly represent energy use; an increase in IT load can lead to a decrease in the PUE. • Once a low PUE is achieved, power supply efficiency and IT load have the greatest impact on its value. - Abstract: Metrics commonly used to assess the energy efficiency of data centres are analysed through performing and critiquing a case study calculation of energy efficiency. Specifically, the metric Power Usage Effectiveness (PUE), which has become a de facto standard within the data centre industry, will be assessed. This is achieved by using open source specifications for a data centre in Prineville, Oregon, USA provided by the Open Compute Project launched by the social networking company Facebook. The usefulness of the PUE metric to the IT industry is critically assessed and it is found that whilst it is important for encouraging lower energy consumption in data centres, it does not represent an unambiguous measure of energy efficiency

  8. External glass peening of zircaloy calandria tubes to increase the critical heat flux

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Coleman, C.E.; Nitheanandan, T.; Kroeger, V.D.; Moyer, R.G.; Sanderson, D.B.; Root, J.H.; Rogge, R.B.

    1997-12-01

    Glass-peening the outside surfaces of Zircaloy calandria tubes increases the nucleation sites available for boiling heat transfer and has been demonstrated to enhance the critical heat flux (CHF) in pool-boiling experiments. The objective of this study is to optimise the heat-transfer enhancement by glass peening while ensuring that the microstructure of the peened tube is acceptable for reactor use. Pool-boiling tests were done using small Zircaloy tubes with as-received ('smooth') surfaces and variously peened surfaces, to evaluate two peening parameters, glass-bead size and the coverage of peened surface. Our results showed that the maximum enhancement of CHF (by 60% compared with as-received tubes) was obtained using a glass-bead size of 90-125 μm with a coverage of 100%. The CHF enhancement was found to be insensitive to glass-bead size over a wide range (from 60-90 μm to 125-180 μm). Using a fixed glass-bead size of 125-180 μm to evaluate the influence of peening coverage, the maximum effect on the CHF response was obtained with a coverage of 1 00%. The microstructures of the peened tubes were evaluated using light microscopy, X-ray and neutron diffraction, and mechanical tests. After peening, the microstructure in the subsurface layer (-30 μm) consisted of deformed α-Zr grains, and the crystallographic texture of the grains changed slightly. After stress-relieving at 500 degrees C for 1 h, some recrystallisation had occurred and the residual strains remaining in the tube were low. The tensile and burst properties of glass-peened and stress-relieved tubes were similar to those of as-received tubes. The microstructures introduced by peening and stress relieving were judged to have little effect on creep and growth behaviour. Since there are no deleterious consequences of the glass-peening treatment, the peened and stress-relieved tubes are found to be acceptable for reactor use. (author)

  9. Preliminary Criticality Calculation on Conceptual Deep Borehole Disposal System for Trans-metal Waste during Operational Phase

    International Nuclear Information System (INIS)

    Kim, In Young; Choi, Heui Joo; Cho, Dong Geun

    2013-01-01

    The primary function of any repository is to prevent spreading of dangerous materials into surrounding environment. In the case of high-level radioactive waste repository, radioactive material must be isolated and retarded during sufficient decay time to minimize radiation hazard to human and surrounding environment. Sub-criticality of disposal canister and whole disposal system is minimum requisite to prevent multiplication of radiation hazard. In this study, criticality of disposal canister and DBD system for trans-metal waste is calculated to check compliance of sub-criticality. Preliminary calculation on criticality of conceptual deep borehole disposal system and its canister for trans-metal waste during operational phase is conducted in this study. Calculated criticalities at every temperature are under sub-criticalities and criticalities of canister and DBD system considering temperature are expected to become 0.34932 and 0.37618 approximately. There are obvious limitations in this study. To obtain reliable data, exact elementary composition of each component, system component temperature must be specified and applied, and then proper cross section according to each component temperature must be adopted. However, many assumptions, for example simplified elementary concentration and isothermal component temperature, are adopted in this study. Improvement of these data must be conducted in the future work to progress reliability. And, post closure criticality analyses including geo, thermal, hydro, mechanical, chemical mechanism, especially fissile material re-deposition by precipitation and sorption, must be considered to ascertain criticality safety of DBD system as a future work

  10. Manual for calculating critical loads of heavy metals for soils and surface waters; preliminary guidelines for environmental quality criteria, calculation methods and input data

    NARCIS (Netherlands)

    Vries, de W.; Bakker, D.J.

    1996-01-01

    Methodologies are described for calculating critical loads of lead, cadmium, copper, zinc, nickel, chromium and mercury for soils and surface waters. The aspects which are discussed are: selection of a computation model, determination of environmental-quality criteria for the metals, collection of

  11. Minimum critical values of uranyl and plutonium nitrate solutions calculated by various routes of the french criticality codes system CRISTAL using the new isopiestic nitrate density law

    International Nuclear Information System (INIS)

    Anno, Jacques; Rouyer, Veronique; Leclaire, Nicolas

    2003-01-01

    This paper provides for various cases of 235 U enrichment or Pu isotopic vectors, and different reflectors, new minimum critical values of uranyl nitrate and plutonium nitrate solutions (H + =0) obtained by the standard IRSN calculation route and the new isopiestic density laws. Comparisons are also made with other more accurate routes showing that the standard one's results are most often conservative and usable for criticality safety assessments. (author)

  12. Development of Nb3Sn cabled conductor by external diffusion process and the effect of strain on the critical current

    International Nuclear Information System (INIS)

    Pasztor, G.; Ekin, J.W.

    1984-01-01

    The planned extension of the High Field Test Facility SULTAN at SIN to provide 12 T operation in a 60 cm bore has led to an increased involvements on the part of SIN in the development of reliable A15 multifilamentary conductors. It is the purpose of this paper to describe the development of stranded Nb 3 Sn conductors using the external diffusion technique. Although not fully optimized, the primary cables have high overall critical current densities. Problems associated with the diffusion of the tin into the copper matrix, such as tin coalescence and development of Kirkendall porosity were successfully solved using small diameter wires and by appropriate diffusion and reaction heat treatment conditions. The elastic strain sensitivity of the critical current of a previously developed cable was comparable to that of bronze processed monolithic Nb 3 Sn, while the irreversible strain limit of 1.2% was significantly higher, On the other hand a longer current transfer length (about 4 times that of bronze processed monolithic conductors) was found

  13. Basis for calculating body equivalent doses after external radiation exposure. 3. rev. and enl. ed.; Berechnungsgrundlage fuer die Ermittlung von Koerper-Aequivalentdosen bei aeusserer Strahlenexposition

    Energy Technology Data Exchange (ETDEWEB)

    Sarenio, O. (comp.) [Geschaeftsstelle der Strahlenschutzkommission beim Bundesamt fuer Strahlenschutz, Bonn (Germany)

    2017-07-01

    The book on the basis for calculating body equivalent doses after external radiation exposure includes the following issues: introduction covering the scope of coverage and body equivalent doses for radiation protection, terminology, photon radiation, neutron radiation, electron radiation, mixed radiation fields and the estimation of body equivalent doses for skin surface contamination.

  14. External validation of the Intensive Care National Audit & Research Centre (ICNARC) risk prediction model in critical care units in Scotland.

    Science.gov (United States)

    Harrison, David A; Lone, Nazir I; Haddow, Catriona; MacGillivray, Moranne; Khan, Angela; Cook, Brian; Rowan, Kathryn M

    2014-01-01

    Risk prediction models are used in critical care for risk stratification, summarising and communicating risk, supporting clinical decision-making and benchmarking performance. However, they require validation before they can be used with confidence, ideally using independently collected data from a different source to that used to develop the model. The aim of this study was to validate the Intensive Care National Audit & Research Centre (ICNARC) model using independently collected data from critical care units in Scotland. Data were extracted from the Scottish Intensive Care Society Audit Group (SICSAG) database for the years 2007 to 2009. Recoding and mapping of variables was performed, as required, to apply the ICNARC model (2009 recalibration) to the SICSAG data using standard computer algorithms. The performance of the ICNARC model was assessed for discrimination, calibration and overall fit and compared with that of the Acute Physiology And Chronic Health Evaluation (APACHE) II model. There were 29,626 admissions to 24 adult, general critical care units in Scotland between 1 January 2007 and 31 December 2009. After exclusions, 23,269 admissions were included in the analysis. The ICNARC model outperformed APACHE II on measures of discrimination (c index 0.848 versus 0.806), calibration (Hosmer-Lemeshow chi-squared statistic 18.8 versus 214) and overall fit (Brier's score 0.140 versus 0.157; Shapiro's R 0.652 versus 0.621). Model performance was consistent across the three years studied. The ICNARC model performed well when validated in an external population to that in which it was developed, using independently collected data.

  15. Automatic treatment of the variance estimation bias in TRIPOLI-4 criticality calculations

    International Nuclear Information System (INIS)

    Dumonteil, E.; Malvagi, F.

    2012-01-01

    The central limit (CLT) theorem States conditions under which the mean of a sufficiently large number of independent random variables, each with finite mean and variance, will be approximately normally distributed. The use of Monte Carlo transport codes, such as Tripoli4, relies on those conditions. While these are verified in protection applications (the cycles provide independent measurements of fluxes and related quantities), the hypothesis of independent estimates/cycles is broken in criticality mode. Indeed the power iteration technique used in this mode couples a generation to its progeny. Often, after what is called 'source convergence' this coupling almost disappears (the solution is closed to equilibrium) but for loosely coupled systems, such as for PWR or large nuclear cores, the equilibrium is never found, or at least may take time to reach, and the variance estimation such as allowed by the CLT is under-evaluated. In this paper we first propose, by the mean of two different methods, to evaluate the typical correlation length, as measured in cycles number, and then use this information to diagnose correlation problems and to provide an improved variance estimation. Those two methods are based on Fourier spectral decomposition and on the lag k autocorrelation calculation. A theoretical modeling of the autocorrelation function, based on Gauss-Markov stochastic processes, will also be presented. Tests will be performed with Tripoli4 on a PWR pin cell. (authors)

  16. A power spectrum approach to tally convergence in Monte Carlo criticality calculation

    International Nuclear Information System (INIS)

    Ueki, Taro

    2017-01-01

    In Monte Carlo criticality calculation, confidence interval estimation is based on the central limit theorem (CLT) for a series of tallies from generations in equilibrium. A fundamental assertion resulting from CLT is the convergence in distribution (CID) of the interpolated standardized time series (ISTS) of tallies. In this work, the spectral analysis of ISTS has been conducted in order to assess the convergence of tallies in terms of CID. Numerical results obtained indicate that the power spectrum of ISTS is equal to the theoretically predicted power spectrum of Brownian motion for tallies of effective neutron multiplication factor; on the other hand, the power spectrum of ISTS of a strongly correlated series of tallies from local powers fluctuates wildly while maintaining the spectral form of fractional Brownian motion. The latter result is the evidence of a case where a series of tallies are away from CID, while the spectral form supports normality assumption on the sample mean. It is also demonstrated that one can make the unbiased estimation of the standard deviation of sample mean well before CID occurs. (author)

  17. Automatic whole core depletion and criticality calculations by MCNPX 2.7.0

    International Nuclear Information System (INIS)

    Kalcheva, S.; Koonen, E.

    2012-01-01

    Different approaches to perform automatic whole core criticality and depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy. (authors)

  18. Automatic treatment of the variance estimation bias in TRIPOLI-4 criticality calculations

    Energy Technology Data Exchange (ETDEWEB)

    Dumonteil, E.; Malvagi, F. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA SACLAY DEN, Laboratoire de Transport Stochastique et Deterministe, 91191 Gif-sur-Yvette (France)

    2012-07-01

    The central limit (CLT) theorem States conditions under which the mean of a sufficiently large number of independent random variables, each with finite mean and variance, will be approximately normally distributed. The use of Monte Carlo transport codes, such as Tripoli4, relies on those conditions. While these are verified in protection applications (the cycles provide independent measurements of fluxes and related quantities), the hypothesis of independent estimates/cycles is broken in criticality mode. Indeed the power iteration technique used in this mode couples a generation to its progeny. Often, after what is called 'source convergence' this coupling almost disappears (the solution is closed to equilibrium) but for loosely coupled systems, such as for PWR or large nuclear cores, the equilibrium is never found, or at least may take time to reach, and the variance estimation such as allowed by the CLT is under-evaluated. In this paper we first propose, by the mean of two different methods, to evaluate the typical correlation length, as measured in cycles number, and then use this information to diagnose correlation problems and to provide an improved variance estimation. Those two methods are based on Fourier spectral decomposition and on the lag k autocorrelation calculation. A theoretical modeling of the autocorrelation function, based on Gauss-Markov stochastic processes, will also be presented. Tests will be performed with Tripoli4 on a PWR pin cell. (authors)

  19. Assessment of the available {sup 233}U cross-section evaluations in the calculation of critical benchmark experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.; Wright, R.Q.

    1996-10-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U.S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the S{sub n} transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  20. Assessment of the Available (Sup 233)U Cross Sections Evaluations in the Calculation of Critical Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Leal, L.C.

    1993-01-01

    In this report we investigate the adequacy of the available {sup 233}U cross-section data for calculation of experimental critical systems. The {sup 233}U evaluations provided in two evaluated nuclear data libraries, the U. S. Data Bank [ENDF/B (Evaluated Nuclear Data Files)] and the Japanese Data Bank [JENDL (Japanese Evaluated Nuclear Data Library)] are examined. Calculations were performed for six thermal and ten fast experimental critical systems using the Sn transport XSDRNPM code. To verify the performance of the {sup 233}U cross-section data for nuclear criticality safety application in which the neutron energy spectrum is predominantly in the epithermal energy range, calculations of four numerical benchmark systems with energy spectra in the intermediate energy range were done. These calculations serve only as an indication of the difference in calculated results that may be expected when the two {sup 233}U cross-section evaluations are used for problems with neutron spectra in the intermediate energy range. Additionally, comparisons of experimental and calculated central fission rate ratios were also made. The study has suggested that an ad hoc {sup 233}U evaluation based on the JENDL library provides better overall results for both fast and thermal experimental critical systems.

  1. Validating MCNP5 libraries and tracking the reason for differences between libraries in criticality calculations

    International Nuclear Information System (INIS)

    Hossny, K.

    2015-01-01

    The purpose of this work is to validate MCNP5 libraries by simulating 4 detailed benchmark experiments and comparing MCNP5 results (each library) with the experimental results and also the previously validated codes for the same experiments MORET 4.A coupled with APOLLO2 (France), and MONK8 (UK). The reasons for difference between libraries are also investigated in this work. Investigating the reason for the differences between libraries will be done by specifying a different library for specific part (clad, fuel, light water) and checking the result deviation than the previously calculated result (with all parts of the same library). The investigated benchmark experiments are of single fuel rods arrays that are water-moderated and water-reflected. Rods contained low-enriched (4.738 wt.% 92 235 U)uranium dioxide (UO 2 ) fuel were clad with aluminum alloy AGS. These experiments were subcritical approaches extrapolated to critical, with the multiplication factor reached being very close to 1.000 (within 0.1%); the subcritical approach parameter was the water level. The studied four cases differ from each other in pitch, number of fuel rods and of course critical height of water. The results show that although library ENDF/B-IV lacks light water treatment card, however its results can be reliable as light water treatment library does not have significant differences from library to another, so it will not be necessary to specify light water treatment card. The main reason for differences between ENDF/B-V and ENDF/B-VI is light water material, especially the Hydrogen element. Specifying the library of Uranium is necessary in case of using library ENDF/B-IV. On the other hand it is not necessary to specify library of cladding material whatever the used library. Validated libraries are ENDF/BIV, ENDF/B-V and ENDF/B-VI with codes in MCNP 42C, 50C and 60C respectively. The presentation slides have been added to the article

  2. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  3. Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

    International Nuclear Information System (INIS)

    Bessis, J.

    1977-01-01

    The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

  4. Burnup credit calculations for criticality safety justification for RBMK-1000 spent fuel of transport and storage systems

    Directory of Open Access Journals (Sweden)

    V. V. Galchenko

    2010-12-01

    Full Text Available In present paper the burnup credit calculations for TK-8 transport container and SVJP-1 spent fuel storage fa-cility of pool type with RBMK-1000 spent fuel during 100-years of cooling time were performed for criticality safety analysis purpose using MCNP and SCALE codes. Only actinides were taken into account for these critical systems. Two approaches were analyzed with isotopes distribution calculations along fuel assembly height and without it. The results show that subcriticality margin is increased considerably using burnup credit and isotopes distribution along fuel assembly height made this value more reasonable.

  5. Critical excitation method for calculating earthquake effects on nuclear plant structures: an assessment study. Technical report

    International Nuclear Information System (INIS)

    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

  6. Test calculations of physical parameters of the TRX,BETTIS and MIT critical assemblies according to the TRIFON program

    International Nuclear Information System (INIS)

    Kochurov, B.P.

    1980-01-01

    Results of calculations of physical parameters characterizing the TRX, MIT and BETTIS critical assemblies obtained according to the program TRIFON are presented. The program TRIFON permits to calculate the space-energy neutron distribution in the multigroup approximation in a multizone cylindrical cell. Results of comparison of the TRX, BETTIS and MIT crytical assembly parameters with experimental data and calculational results according to the Monte Carlo method are presented as well. Deviations of the parameters are in the range of 1.5-2 of experimental errors. Data on the interference of uranium 238 levels in the resonant neutron absorption in the cell are given [ru

  7. 3-D flux distribution and criticality calculation of TRIGA Mark-II

    International Nuclear Information System (INIS)

    Can, B.

    1982-01-01

    In this work, the static calculation of the (I.T.U. TRIGA Mark-II) flux distribution has been made. The three dimensional, r-θ-z, representation of the core has been used. In this representation, for different configuration, the flux distribution has been calculated depending on two group theory. The thermal-hydraulics, the poisoning effects have been ignored. The calculations have been made by using the three dimensional and multigroup code CAN. (author)

  8. The Prognosis of Political Stability of the Russian Federation on the Basis of Calculation of the Index of National External Economic Stability

    Directory of Open Access Journals (Sweden)

    Владимир Геннадьевич Иванов

    2012-12-01

    Full Text Available The article contains the development of ideas presented in the previous issue of the bulletin. On the basis of the proposed by V.G. Ivanov methodology of calculation of the index of national external economic stability there has been prepared the short- mid-term prognosis of the level of stability of the Russian political regime. With a glance to the specificity of the development of the Russian Federation the methodology of calculation of the deflator of the referred index has been worked out as well.

  9. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP

    International Nuclear Information System (INIS)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R.

    2013-10-01

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  10. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    International Nuclear Information System (INIS)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun

    2009-01-01

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  11. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun [KAERI, Daejeon (Korea, Republic of)

    2009-01-15

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  12. Criticality calculations by source-collision iteration technique for cylindrical systems

    International Nuclear Information System (INIS)

    Sundaram, V.K.; Gopinath, D.V.

    1977-01-01

    A fast-converging iterative technique is presented which uses first collision probabilities developed for obtaining criticality parameters in two-region cylindrical systems with multigroup structure in energy of the neutrons. The space transmission matrix is obtained part analytically and part numerically through evaluation of a single-fold integral. Critical dimensions for condensed systems of uranium and plutonium computed using this method are presented and compared with published values

  13. COMPARISON OF GKS CALCULATED CRITICAL ION TEMPERATURE GRADIENTS AND ITG GROWTH RATES TO DIII-D MEASURED GRADIENTS AND DIFFUSIVITIES

    International Nuclear Information System (INIS)

    BAKER, DR; STAEBLER, GM; PETTY, CC; GREENFIELD, CM; LUCE, TC

    2003-01-01

    OAK-B135 The gyrokinetic equations predict that various drift type waves or modes can be unstable in a tokamak. For some of these modes, such as the ion temperature gradient (ITG) mode and the electron temperature gradient mode, there exists a critical gradient, above which the mode is unstable. Since the existence of unstable modes can cause increased transport, plasmas which are centrally heated tend to increase in temperature gradient until the modes become unstable. Under some conditions the increased transport can fix the gradient at the critical value. here they present a comparison between the measured ion temperature gradients and the critical gradient as calculated by a gyrokinetic linear stability (GKS) code. They also present the maximum linear growth rate as calculated by this code for comparison to experimentally derived transport coefficients. The results show that for low confinement mode (L-mode) discharges, the measured ion temperature gradient is significantly greater than the GKS calculated critical gradient over a large region of the plasma. This is the same region of the plasma where the ion thermal diffusivity is large. For high confinement mode (H-mode) discharges the ion temperature gradient is closer to the critical gradient, but often still greater than the critical gradient over some region. For the best H-mode discharges, the ion temperature is less than or equal to the critical gradient over the whole plasma. In general they find that the position in the plasma where the ion thermal diffusivity starts to increase rapidly is where the maximum linear growth rate is greater than the E x B shearing rate

  14. The calculation of dose from external photon exposures using reference human phantoms and Monte-Carlo methods. Pt. 1

    International Nuclear Information System (INIS)

    Kramer, R.; Zankl, M.; Williams, G.; Drexler, G.

    1982-12-01

    By the help of a Monte-Carlo program the dose that single organs, organ groups and bigger or smaller parts of body would receive on an average, caused by an irradiation definitely fixed by the geometry of irradiation and photon energy, can be determined. Thus the phantom in connection with the Monte-Carlo program can be used for several considerations as for example - calculation of dose from occupational exposures - calculation of dose from diagnostic procedures - calculation of dose from radiotherapy procedures. (orig.)

  15. Multigroup transport calculations of critical and fuel assemblies with taking into account the scattering anisotropy

    International Nuclear Information System (INIS)

    Rubin, I.E.; Dneprovskaya, N.M.

    2005-01-01

    A technique for calculation of reactor lattices by means of the transmission probabilities with taking into account the scattering anisotropy is generalized for the multigroup case. The errors of the calculated multiplication coefficients and energy release distributions do noe exceed practically the errors, of these values, obtained by the Monte Carlo method. The proposed method is most effective when determining the small difference effects [ru

  16. Combining external and internal mixing representation of atmospheric aerosol for optical properties calculations: focus on absorption properties over Europe and North America using AERONET observations and AQMEII simulations

    Science.gov (United States)

    Curci, Gabriele

    2017-04-01

    The calculation of optical properties from knowledge of the composition and abundance of atmospheric aerosol implies a certain number of assumptions. First and if not known or explicitly simulated, a size distribution must be assigned to each aerosol component (e.g. sulfate-like inorganic ions, organic and back carbon, soil dust, sea salt). Second, physical-chemical properties such as the shape, density, complex refractive index, and hygroscopic factors must be associated to each aerosol species. Third, a representation of how the aerosol species combine together must be made: among those, the most popular are the assumptions of external mixing, in which each particle is assumed to be formed of a single compound and the optical properties may be calculated separately for each species, or of internal core-shell arrangement, in which each particle consists of a water-insoluble core coated with a water-soluble shell and that requires more elaborate calculations for optical properties. Previous work found that the assumption on the mixing state (external or core-shell internal) is the one that introduces the highest uncertainty, quantified in about 30% uncertainty on the calculation of monthly mean aerosol optical depth (AOD) and single-scattering albedo (SSA). The external mixing assumption is generally more reasonable for freshly emitted aerosol, while the internal mixing case is associated with aged aerosol that had the time to form the coating around the core. Both approximations are thus regarded as valid, but in general a combination of the two mixing states may be expected in a given air mass. In this work, we test a simple empirical parameterization of the fraction of internally mixed particles (F_in) in a generic air mass. The F_in fraction is calculated in two alternative ways, one exploiting the NOz to NOx ratio (proxy of the photochemical aging), and the other using the relative abundance of black carbon with respect to other aerosol components (proxy of

  17. Requirements on waste forms for the planned Konrad repository based on criticality calculations

    International Nuclear Information System (INIS)

    Berg, H.P.

    1988-02-01

    In the framework of the safety analyses for the planned Konrad repository it has been investigated whether a criticality incident may be possible during the operational phase or in the post-operational phase. The analysis has shown that the criticality safety is ensured by limitation of a mass concentration of the fissile material in the waste form group and by determination of a maximum permissible mass of fissile material per waste package. The resulting requirements of the waste packages, including a mixture in the cross-section of an emplacement room, are explained. (orig.) [de

  18. Application of the Bowring correlation for calculating the critical heat flux

    International Nuclear Information System (INIS)

    Borges, R.C.; Freitas, R.L.

    1986-01-01

    The evaluation of the critical heat flux is of great importance for the nuclear reactor project, because it permits the verification of the safety margin with respect to fuel rod damage. This work presents a comparison of the original critical heat flux correlation proposed by Bowring with an alternative form derived from it presented in several papers. Very different results have been encountered from the application of the two correlation forms. Therefore, a criterious choice of the correlation form must be done avoid the violation of the project's safety margin. (Author) [pt

  19. SARTEMP2 - A computer program to calculate power and temperatures in a transport flask during a criticality accident

    International Nuclear Information System (INIS)

    Shaw, P.M.

    1983-04-01

    The computer code SARTEMP2, an extended version of the original SARTEMP program, which calculates the power and temperatures in a transport flask during a hypothetical criticality accident is described. The accident arises, it is assumed, during the refilling of the flask with water, bringing the system to delayed critical. As the water level continues to rise, reactivity is added causing the power to rise, and thus temperatures in the fuel, clad and water to increase. The point kinetics equations are coupled to the one-dimensional heat conduction equation. The model used, the method of solution of the equations and the input data required are given. (author)

  20. Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.

    2000-01-01

    The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

  1. The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 4

    International Nuclear Information System (INIS)

    Williams, G.; Zankl, M.; Drexler, G.

    1984-12-01

    This report considers the contribution from scattered radiation to the dose to organs and tissues which lie outside the useful therapy beams. The results presented are the product of Monte Carlo studies used to determine the tissue doses due to internal scattering of the useful beams only. General cases are calculated in which central target volumes in the trunk are treated with 10 x 14 cm 2 and 14 x 14 cm 2 fields from 200 kV, Co-60, 8 MV and 25 MV therapy equipment. Target volumes in the neck are considered to be treated with 5 x 5 cm 2 fields. Different treatment plans are calculated including rotational therapy. Also two specific cases are more fully analysed, namely for Ankylosing Spondylitis and central abdomen malignant disease in the region of the head of the pancreas. The calculated organ doses are presented in tables as a percentage of the target volume dose. (orig.)

  2. Summary and recommendations of a National Cancer Institute workshop on issues limiting the clinical use of Monte Carlo dose calculation algorithms for megavoltage external beam radiation therapy

    International Nuclear Information System (INIS)

    Fraass, Benedick A.; Smathers, James; Deye, James

    2003-01-01

    Due to the significant interest in Monte Carlo dose calculations for external beam megavoltage radiation therapy from both the research and commercial communities, a workshop was held in October 2001 to assess the status of this computational method with regard to use for clinical treatment planning. The Radiation Research Program of the National Cancer Institute, in conjunction with the Nuclear Data and Analysis Group at the Oak Ridge National Laboratory, gathered a group of experts in clinical radiation therapy treatment planning and Monte Carlo dose calculations, and examined issues involved in clinical implementation of Monte Carlo dose calculation methods in clinical radiotherapy. The workshop examined the current status of Monte Carlo algorithms, the rationale for using Monte Carlo, algorithmic concerns, clinical issues, and verification methodologies. Based on these discussions, the workshop developed recommendations for future NCI-funded research and development efforts. This paper briefly summarizes the issues presented at the workshop and the recommendations developed by the group

  3. Monitor unit calculations for external photon and electron beams: Report of the AAPM Therapy Physics Committee Task Group No. 71

    Energy Technology Data Exchange (ETDEWEB)

    Gibbons, John P., E-mail: john.gibbons@marybird.com [Department of Physics, Mary Bird Perkins Cancer Center, Baton Rouge, Louisiana 70809 (United States); Antolak, John A. [Department of Radiation Oncology, Mayo Clinic, Rochester, Minnesota 55905 (United States); Followill, David S. [Department of Radiation Physics, UT M.D. Anderson Cancer Center, Houston, Texas 77030 (United States); Huq, M. Saiful [Department of Radiation Oncology, University of Pittsburgh Cancer Institute, Pittsburgh, Pennsylvania 15232 (United States); Klein, Eric E. [Department of Radiation Oncology, Washington University School of Medicine, St. Louis, Missouri 63110 (United States); Lam, Kwok L. [Department of Radiation Oncology, University of Michigan, Ann Arbor, Michigan 48109 (United States); Palta, Jatinder R. [Department of Radiation Oncology, Virginia Commonwealth University, Richmond, Virginia 23298 (United States); Roback, Donald M. [Department of Radiation Oncology, Cancer Centers of North Carolina, Raleigh, North Carolina 27607 (United States); Reid, Mark [Department of Medical Physics, Fletcher-Allen Health Care, Burlington, Vermont 05401 (United States); Khan, Faiz M. [Department of Radiation Oncology, University of Minnesota, Minneapolis, Minnesota 55455 (United States)

    2014-03-15

    A protocol is presented for the calculation of monitor units (MU) for photon and electron beams, delivered with and without beam modifiers, for constant source-surface distance (SSD) and source-axis distance (SAD) setups. This protocol was written by Task Group 71 of the Therapy Physics Committee of the American Association of Physicists in Medicine (AAPM) and has been formally approved by the AAPM for clinical use. The protocol defines the nomenclature for the dosimetric quantities used in these calculations, along with instructions for their determination and measurement. Calculations are made using the dose per MU under normalization conditions, D{sub 0}{sup ′}, that is determined for each user's photon and electron beams. For electron beams, the depth of normalization is taken to be the depth of maximum dose along the central axis for the same field incident on a water phantom at the same SSD, where D{sub 0}{sup ′} = 1 cGy/MU. For photon beams, this task group recommends that a normalization depth of 10 cm be selected, where an energy-dependent D{sub 0}{sup ′} ≤ 1 cGy/MU is required. This recommendation differs from the more common approach of a normalization depth of d{sub m}, with D{sub 0}{sup ′} = 1 cGy/MU, although both systems are acceptable within the current protocol. For photon beams, the formalism includes the use of blocked fields, physical or dynamic wedges, and (static) multileaf collimation. No formalism is provided for intensity modulated radiation therapy calculations, although some general considerations and a review of current calculation techniques are included. For electron beams, the formalism provides for calculations at the standard and extended SSDs using either an effective SSD or an air-gap correction factor. Example tables and problems are included to illustrate the basic concepts within the presented formalism.

  4. Calculations of critical micelle concentration by dissipative particle dynamics simulations: the role of chain rigidity.

    Science.gov (United States)

    Lee, Ming-Tsung; Vishnyakov, Aleksey; Neimark, Alexander V

    2013-09-05

    Micelle formation in surfactant solutions is a self-assembly process governed by complex interplay of solvent-mediated interactions between hydrophilic and hydrophobic groups, which are commonly called heads and tails. However, the head-tail repulsion is not the only factor affecting the micelle formation. For the first time, we present a systematic study of the effect of chain rigidity on critical micelle concentration and micelle size, which is performed with the dissipative particle dynamics simulation method. Rigidity of the coarse-grained surfactant molecule was controlled by the harmonic bonds set between the second-neighbor beads. Compared to flexible molecules with the nearest-neighbor bonds being the only type of bonded interactions, rigid molecules exhibited a lower critical micelle concentration and formed larger and better-defined micelles. By varying the strength of head-tail repulsion and the chain rigidity, we constructed two-dimensional diagrams presenting how the critical micelle concentration and aggregation number depend on these parameters. We found that the solutions of flexible and rigid molecules that exhibited approximately the same critical micelle concentration could differ substantially in the micelle size and shape depending on the chain rigidity. With the increase of surfactant concentration, primary micelles of more rigid molecules were found less keen to agglomeration and formation of nonspherical aggregates characteristic of flexible molecules.

  5. Calculation of critical fault recovery time for nonlinear systems based on region of attraction analysis

    DEFF Research Database (Denmark)

    Tabatabaeipour, Mojtaba; Blanke, Mogens

    2014-01-01

    of a system. It must be guaranteed that the trajectory of a system subject to fault remains in the region of attraction (ROA) of the post-fault system during this time. This paper proposes a new algorithm to compute the critical fault recovery time for nonlinear systems with polynomial vector elds using sum...

  6. Criticality calculations of the HTR-10 pebble-bed reactor with SCALE6/CSAS6 and MCNP5

    International Nuclear Information System (INIS)

    Wang, Meng-Jen; Sheu, Rong-Jiun; Peir, Jinn-Jer; Liang, Jenq-Horng

    2014-01-01

    Highlights: • Comparisons of the HTR-10 criticality calculations with SCALE6/CSAS6 and MCNP5 were performed. • The DOUBLEHET unit-cell treatment provides the best k eff estimation among PBR criticality calculations using SCALE6. • The continuous-energy SCALE6 calculations present a non-negligible discrepancy with MCNP5 in three PBR cases. - Abstract: HTR-10 is a 10 MWt prototype pebble-bed reactor (PBR) that presents a doubly heterogeneous geometry for neutronics calculations. An appropriate unit-cell treatment for the associated fuel elements is vital for creating problem-dependent multigroup cross sections. Considering four unit-cell options for resonance self-shielding correction in SCALE6, a series of HTR-10 core models were established using the CSAS6 sequence to systematically investigate how they affected the computational accuracy and efficiency of PBR criticality calculations. Three core configurations, which ranged from simplified infinite lattices to a detailed geometry, were examined. Based on the same ENDF/B-VII.0 cross-section library, multigroup results were evaluated by comparing with continuous-energy SCALE6/CSAS6 and MCNP5 calculations. The comparison indicated that the INFHOMMEDIUM results overestimated the effective multiplication factor (k eff ) by about 2800 pcm, whereas the LATTICECELL and MULTIREGION treatments overestimated k eff values with similar biases at approximately 470–680 pcm. The DOUBLEHET results attained further improvement, reducing the k eff overestimation to approximately 280 pcm. The comparison yielded two unexpected problems from using SCALE6/CSAS6 in HTR-10 criticality calculations. In particular, the continuous-energy CSAS6 calculations in this study present a non-negligible discrepancy with MCNP5, potentially causing a k eff value overestimate of approximately 680 pcm. Notably, using a cell-weighted mixture instead of an explicit model of individual TRISO particles in the pebble fuel zone does not shorten the

  7. Comparison of statistical evaluation of criticality calculations for reactors VENUS-F and ALFRED

    Directory of Open Access Journals (Sweden)

    Janczyszyn Jerzy

    2017-01-01

    Full Text Available Limitations of correct evaluation of keff in Monte Carlo calculations, claimed in literature, apart from the nuclear data uncertainty, need to be addressed more thoroughly. Respective doubts concern: the proper number of discarded initial cycles, the sufficient number of neutrons in a cycle and the recognition and dealing with the keff bias. Calculations were performed to provide more information on these points with the use of the MCB code, solely for fast cores. We present applied methods and results, such as: calculation results for stability of variance, relation between standard deviation reported by MCNP and this from the dispersion of multiple independent keff values, second order standard deviations obtained from different numbers of grouped results. All obtained results for numbers of discarded initial cycles from 0 to 3000 were analysed leading for interesting conclusions.

  8. The graphics calculator in mathematics education: A critical review of recent research

    Science.gov (United States)

    Penglase, Marina; Arnold, Stephen

    1996-04-01

    The graphics calculator, sometimes referred to as the "super calculator," has sparked great interest among mathematics educators. Considered by many to be a tool which has the potential to revolutionise mathematics education, a significant amount of research has been conducted into its effectiveness as a tool for instruction and learning within precalculus and calculus courses, specifically in the study of functions, graphing and modelling. Some results suggest that these devices (a) can facilitate the learning of functions and graphing concepts and the development of spatial visualisation skills; (b) promote mathematical investigation and exploration; and (c) encourage a shift in emphasis from algebraic manipulation and proof to graphical investigation and examination of the relationship between graphical, algebraic and geometric representations. Other studies, however, indicate that there is still a need for manipulative techniques in the learning of function and graphing concepts, that the use of graphics calculators may not facilitate the learning of particular precalculus topics, and that some "de-skilling" may occur, especially among males. It is the contention of this paper, however, that much of the research in this new and important field fails to provide clear guidance or even to inform debate in adequate ways regarding the role of graphics calculators in mathematics teaching and learning. By failing to distinguish the role of the tool from that of the instructional process, many studies reviewed could be more appropriately classified as "program evaluations" rather than as research on the graphics calculator per se. Further, claims regarding the effectiveness of the graphics calculator as a tool for learning frequently fail to recognise that judgments of effectiveness result directly from existing assumptions regarding both assessment practice and student "achievement."

  9. A Method of Calculating Critical Depth of Burial of Explosive Charges to Generate Bulging and Cratering in Rock

    Directory of Open Access Journals (Sweden)

    Mingyang Wang

    2016-01-01

    Full Text Available For underground explosions, a thin to medium thickness layer near the cavity of an explosion can be considered a theoretical shell structure. Detonation products transmit the effective energy of explosives to this shell which can expand thus leading to irreversible deformation of the surrounding medium. Based on mass conservation, incompressible conditions, and boundary conditions, the possible kinematic velocity fields in the plastic zone are established. Based on limit equilibrium theory, this work built equations of material resistance corresponding to different possible kinematic velocity fields. Combined with initial conditions and boundary conditions, equations of motion and material resistance are solved, respectively. It is found that critical depth of burial is positively related to a dimensionless impact factor, which reflects the characteristics of the explosives and the surrounding medium. Finally, an example is given, which suggests that this method is capable of calculating the critical depth of burial and the calculated results are consistent with empirical results.

  10. Critical study of Weyl anomaly calculation in the string by Fujikawa method

    International Nuclear Information System (INIS)

    Dalmazi, D.

    1987-01-01

    The Fujikawa's evaluation of Weyl's anomaly in the bosonic string is carefully analyzed, paying special attention to the covariance under conformal transformation. The choice of operators used to calculate the variation of path integral measure under Weyl's transformation is discussed in detail. (author) [pt

  11. Experimental study and calculations of the near critical behavior of a synthetic fluid in nitrogen injection

    International Nuclear Information System (INIS)

    Coronado Parra, Carlos Alberto; Escobar Remolina, Juan Carlos M

    2005-01-01

    In recent years, the use of nitrogen has increased as gas injection to recover oil fluids near the critical point. The behavior of hydrocarbon mixture phases in the critical region shows very interesting complex phenomena when facing a recovery project with nitrogen. Therefore, it is important to have experimental information of the PVTx thermodynamic variable, often scarce, for this type of critical phenomena. This paper reports the experimental measures of the volumetric behavior and phases of synthetic fluid in a nitrogen injection process. The experiment was performed at laboratory scale, and it obtained variations on the saturation pressure, gas oil ratio, density and composition of the hydrocarbon phase when nitrogen was injected at molars of 10,20,30 and 40% on different volumetric portions of the mother sample. In addition, the data obtained experimentally was used to demonstrate the capacity of tune to compositional models. The data provided represents a valuable contribution to the understanding of phenomena associated with retrograde and near critical regions, as well as their use in tuning and developing more elaborate models such as Cubic Equations of State (EOS). It is worth highlighting the importance of this data in the potential processes of nitrogen, CO 2 , and lean gas injection, which require knowledge of the gas-oil ratio, saturation pressures, density and composition of the fluid in current production. The identification of the phenomena shown, represent a potential application to the modeling of displacements and maintaining the pressure in the improved recovery when scaling up the laboratory data to the field / reservoir conditions

  12. CIRCEE: a software for isodose calculations of criticality risk in evolutional environment

    International Nuclear Information System (INIS)

    Masse, A.; Abeguile, F.

    2010-01-01

    The authors briefly present the CIRCEE software which has been developed to assess doses, dose rates and ambient dose equivalent rates induced by neutrons and gamma radiations produced by fission and secondary gamma radiations through 3D public works in a nuclear installation almost in real time. It is meant to be used in case of criticality accident. They present the methodology, the scope and domain of application, the software operating mode

  13. Comments on the use of the Monte Carlo method for criticality calculations

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1975-01-01

    As evidenced by recent papers given at Nuclear Criticality Safety Division meetings, the use of the Monte Carlo method has become a very popular computational tool. The ease of use has undoubtably been a primary reason for this popularity. This ease of use, however, may lead to a false sense of security when using the method. Guidance on the effective use of the method and some suggestions on how to avoid some of the pitfalls that can occur are presented

  14. A comparison of reaction rate calculations using Endf/B-VII with critical assembly measurements

    International Nuclear Information System (INIS)

    Wilkerson, C.; Mac Innes, M.; Barr, D.; Trellue, H.; MacFarlane, R.; Chadwick, M.

    2008-01-01

    We present critical assembly reaction rate data, and modeling of the same using the recently released Endf/B-VII library. While some of the experimental measurements were performed as long as 50 years ago, the results have not been widely used/available outside of Los Alamos. Over the years, a variety of target foils were fabricated and placed in differing neutron spectrum/fluence environments within critical assemblies. Neutron-induced reactions such as (n,γ), (n,2n), and (n,f) on these targets were measured, typically referenced to 235 U(n,f) or 239 Pu(n,f). Because the cross section for the latter reactions are now well known, these experiments provide a rich data set for testing evaluated cross sections. Due to the large variety of critical assemblies that were historically available at Los Alamos, it was possible to make measurements in spectral environments ranging from hard (Pu Jezebel, center of Pu Flattop) through intermediate (Big Ten) to degraded (reflector region of Flattop). This broad range of configurations allows us to test both the cross section magnitudes and their energy dependencies. We will present data, along with reaction rate predictions using primarily MCNP5 in conjunction with Endf/B-VII, for a number of target nuclei, including iridium, isotopes of uranium (e.g., 233, 235, 237, 238), neptunium (237), plutonium (239), and americium (241). (authors)

  15. Guidelines for planning interventions against external exposure in industrial area after a nuclear accident. Pt. 2. Calculation of doses using Monte Carlo method

    International Nuclear Information System (INIS)

    Kis, Z.; Eged, K.; Meckbach, R.; Mueller, H.

    2003-01-01

    Countermeasures being different from the usual urban ones and largely applicable in industrial area are collected and evaluated in a separate report. The industrial area is defined here as such an area where productive and/or commercial activity is carried out. A good example is a supermarket or a factory. Based on the history of calculation models it is unambiguous that the Monte Carlo based simulation is the perspective to the dose assessment from external exposures in such a complex environment. A method of the calculation of doses from external exposures in urban-industrial environment is presented. Moreover, this report gives a summary about the time dependence of the source strengths relative to a reference surface and a short overview about the mechanical and chemical intervention techniques which can be applied in this area. Using a hypothetical scenario (a supermarket area contaminated by 137 Cs) the details of an exemplary calculation are given directly addressing the dose and averted dose blocks of the templates of industrial countermeasures. In addition, a sensitivity analysis of the results is presented. (orig.)

  16. The comparison of MCNP perturbation technique with MCNP difference method in critical calculation

    International Nuclear Information System (INIS)

    Liu Bin; Lv Xuefeng; Zhao Wei; Wang Kai; Tu Jing; Ouyang Xiaoping

    2010-01-01

    For a nuclear fission system, we calculated Δk eff , which arise from system material composition changes, by two different approaches, the MCNP perturbation technique and the MCNP difference method. For every material composition change, we made four different runs, each run with different cycles or each cycle generating different neutrons, then we compared the two Δk eff that are obtained by two different approaches. As a material composition change in any particular cell of the nuclear fission system is small compared to the material compositions in the whole nuclear fission system, in other words, this composition change can be treated as a small perturbation, the Δk eff results obtained from the MCNP perturbation technique are much quicker, much more efficient and reliable than the results from the MCNP difference method. When a material composition change in any particular cell of the nuclear fission system is significant compared to the material compositions in the whole nuclear fission system, both the MCNP perturbation technique and the MCNP difference method can give satisfactory results. But for the run with the same cycles and each cycle generating the same neutrons, the results obtained from the MCNP perturbation technique are systemically less than the results obtained from the MCNP difference method. To further confirm our calculation results from the MCNP4C, we run the exact same MCNP4C input file in MCNP5, the calculation results from MCNP5 are the same as the calculation results from MCNP4C. We need caution when using the MCNP perturbation technique to calculate the Δk eff as the material composition change is large compared to the material compositions in the whole nuclear fission system, even though the material composition changes of any particular cell of the fission system still meet the criteria of MCNP perturbation technique.

  17. Validation of KENO V.a for criticality safety calculations involving WR-1 fast-neutron fuel arrangements

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I. C.

    1991-07-15

    The KENO V.a criticality safety code, used with the SCALE 27-energy-group ENDF/B-IV-based cross-section library, has been validated for low-enriched uranium carbide (UC) WR-1 fast-neutron (FN) fuel arrangements. Because of a lack of relevant experimental data for UC fuel in the published literature, the validation is based primarily on calculational comparisons with critical experiments for fuel types with a range of enrichments and densities that cover those of the FN UC fuel. The ability of KENO V.a to handle the unique annular pin arrangement of the WR-1 FN fuel bundle was established using a comparison with the MCNP3B code used with a continuous-energy ENDF/B-V-based cross-section library. This report is part of the AECL--10146 report series documenting the validation of the KENO V.a criticality safety code.

  18. 临界水深计算方法的研究%Research on calculation method for critical water depth

    Institute of Scientific and Technical Information of China (English)

    王功

    2011-01-01

    总结了渠道临界水深常见的计算方法,分析了过水断面比能曲线的特性,根据渠道临界水深的定义,利用计算机软件编程技术可以解决大量繁琐计算的特点,求解了明渠临界水深,并且分析与总结了用定义法解决工程计算的意义.%Firstly, common calculation methods for the channel critical depth has been summarized, and the characteristics of specific energy curve of flow cross-section have been analyzed in this paper.By using computer software programming technology that can solve the massive trival calculation, based on the definition of the channel critical depth, the critical water depth was solved.And the significance of using the definition method to solve engineering calculation has been analysed and summarized.

  19. The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 6

    International Nuclear Information System (INIS)

    Zankl, M.; Panzer, W.; Drexler, G.

    1991-11-01

    Computed tomography (CT) is a technique which offers a high diagnostic capability; however, the dose to the patient is high compared to conventional radiography. This report provides a catalogue of organ doses resulting from CT examinations. The organ doses were calculated for the type of CT scanners most commonly used in the FRG and for three different radiation qualities. For the dose calculations, the patients were represented by the adult mathematical phantoms Adam and Eva. The radiation transport in the body was simulated using a Monte Carlo method. The doses were calculated as conversion factors of mean organ doses per air kerma free in air on the axis of rotation. Mean organ dose conversion factors are given per organ and per single CT slice of 1 cm width. The mean dose to an organ resulting from a particular CT examination can be estimated by summing up the contribution to the organ dose from each relevant slice. In order to facilitate the selection of the appropriate slices, a table is given which relates the mathematical phantoms' coordinates to certain anatomical landmarks in the human body. (orig.)

  20. The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 3

    International Nuclear Information System (INIS)

    Drexler, G.; Panzer, W.; Widenmann, L.; Williams, G.; Zankl, M.

    1984-03-01

    This report gives tables of conversion factors for the calculation of organ doses from technical parameters of typical radiographic techniques. These conversion factors were calculated using a male and a female mathematical human phantom and an efficient Monte Carlo programme that determines the mean organ doses from the energy deposited in each organ. Each diagnostic X-ray examination is studied using three X-ray spectra resulting from three different high tension values. The conversion factors per unit entrance air dose in free air are given for sixteen organs and for the entrance and exit surface skin doses. The tables are actually valid only for the given parameters such as phantom dimensions, source-to-skin distance, projection and X-ray quality. This, of course, gives rise to some uncertainty when dealing with the individual technique and patient. The uncertainty in organ dose of adult patients, however, should not be very large, if the calculation is based on a similar geometry, and before all, on the actually administered entrance air dose in the selected high tension range according to the patient parameters. (orig.)

  1. MCNP apply in calculating reactor critical coefficient Keff under the changing of the burnable poison rod

    International Nuclear Information System (INIS)

    Wang Xinghua; Zhou Sichun; Zhang Qingxian; Zhao Feng; Liu Jun; Zhu Jian

    2013-01-01

    Taking Qinshan nuclear power plant as an example, in this paper, Monte Carlo method was used in the MCNP procedures for the establishment of nuclear power station simulation model, construct the reactor pressure vessel and vessel core component composition and arrangement, KCODE card was used to calculate the effect of the number and the location of burnable poison control rod factor K eff by the boron acid. The calculation results show that, with the increasing in the number of burnable poison control rod value-added factor K eff shown a downward trend, and with the burnable poison control rod from the dense to sparse, which K eff will be decreasing slowly. This condition is consistent with the theoretical. (authors)

  2. Implementation of CTRLPOS, a VENTURE module for control rod position criticality searches, control rod worth curve calculations, and general criticality searches

    Energy Technology Data Exchange (ETDEWEB)

    Smith, L.A.; Renier, J.P.

    1994-06-01

    A module in the VENTURE reactor analysis code system, CTRLPOS, is developed to position control rods and perform control rod position criticality searches. The module is variably dimensioned so that calculations can be performed with any number of control rod banks each having any number of control rods. CTRLPOS can also calculate control rod worth curves for a single control rod or a bank of control rods. Control rod depletion can be calculated to provide radiation source terms. These radiation source terms can be used to predict radiation doses to personnel and estimate the shielding and long-term storage requirements for spent control rods. All of these operations are completely automated. The numerous features of the module are discussed in detail. The necessary input data for the CTRLPOS module is explained. Several sample problems are presented to show the flexibility of the module. The results presented with the sample problems show that the CTRLPOS module is a powerful tool which allows a wide variety of calculations to be easily performed.

  3. Timing comparison of two-dimensional discrete-ordinates codes for criticality calculations

    International Nuclear Information System (INIS)

    Miller, W.F. Jr.; Alcouffe, R.E.; Bosler, G.E.; Brinkley, F.W. Jr.; O'dell, R.D.

    1979-01-01

    The authors compare two-dimensional discrete-ordinates neutron transport computer codes to solve reactor criticality problems. The fundamental interest is in determining which code requires the minimum Central Processing Unit (CPU) time for a given numerical model of a reasonably realistic fast reactor core and peripherals. The computer codes considered are the most advanced available and, in three cases, are not officially released. The conclusion, based on the study of four fast reactor core models, is that for this class of problems the diffusion synthetic accelerated version of TWOTRAN, labeled TWOTRAN-DA, is superior to the other codes in terms of CPU requirements

  4. BCG: a code for calculating pointwise neutron spectra and criticality in fast reactor cells

    International Nuclear Information System (INIS)

    Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.

    1988-02-01

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is presented. The code solves the unidimensional neutron transport equation together with interface current relations at each energy in an unionized grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstructed total microscopic cross sections of the atomic species in the cell. Results for a defined sample problem illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt

  5. A nonparametric approach to calculate critical micelle concentrations: the local polynomial regression method

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Fontan, J.L.; Costa, J.; Ruso, J.M.; Prieto, G. [Dept. of Applied Physics, Univ. of Santiago de Compostela, Santiago de Compostela (Spain); Sarmiento, F. [Dept. of Mathematics, Faculty of Informatics, Univ. of A Coruna, A Coruna (Spain)

    2004-02-01

    The application of a statistical method, the local polynomial regression method, (LPRM), based on a nonparametric estimation of the regression function to determine the critical micelle concentration (cmc) is presented. The method is extremely flexible because it does not impose any parametric model on the subjacent structure of the data but rather allows the data to speak for themselves. Good concordance of cmc values with those obtained by other methods was found for systems in which the variation of a measured physical property with concentration showed an abrupt change. When this variation was slow, discrepancies between the values obtained by LPRM and others methods were found. (orig.)

  6. BCG: a computer code for calculating neutron spectra and criticality in cells of fast reactors

    International Nuclear Information System (INIS)

    Leite, S.B.; Caldeira, A.D.; Garcia, R.D.M.

    1988-01-01

    The BCG code for determining the space and energy neutron flux distribution and criticality of fast reactor cylindrical cells is discussed. The code solves the unidimensional neutron transport equation together with interface current relations at each energy point in an unionized energy grid prepared for the cell and at an arbitrary number of spatial zones. While the spatial resolution is user specified, the energy dependence of the flux distribution is resolved according to the degree of variation in the reconstruced total microscopic cross sections of the atomic species in the cell. Results for a simplified fuel cell illustrate the high resolution and accuracy that can be obtained with the code. (author) [pt

  7. Comparison of Hansen--Roach and ENDF/B-IV cross sections for 233U criticality calculations

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1976-01-01

    A comparison is made between criticality calculations performed using ENDF/B-IV cross sections and the 16-group Hansen-- Roach library at ORNL. The area investigated is homogeneous systems of highly enriched 233 U in simple geometries. Calculations are compared with experimental data for a wide range of H/ 233 U ratios. Results show that calculations of k/sub eff/ made with the Hansen--Roach cross sections agree within 1.5 percent for the experiments considered. Results using ENDF/B-IV cross sections were in good agreement for well-thermalized systems, but discrepancies up to 7 percent in k/sub eff/ were observed in fast and epithermal systems

  8. CRITICALITY CALCULATION FOR THE MOST REACTIVE DEGRADED CONFIGURATIONS OF THE FFTF SNF CODISPOSAL WP CONTAINING AN INTACT IDENT-69 CONTAINER

    International Nuclear Information System (INIS)

    D.R. Moscalu

    2002-01-01

    The objective of this calculation is to perform additional degraded mode criticality evaluations of the Department of Energy's (DOE) Fast Flux Test Facility (FFTF) Spent Nuclear Fuel (SNF) codisposed in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP). The scope of this calculation is limited to the most reactive degraded configurations of the codisposal WP with an almost intact Ident-69 container (breached and flooded but otherwise non-degraded) containing intact FFTF SNF pins. The configurations have been identified in a previous analysis (CRWMS M andO 1999a) and the present evaluations include additional relevant information that was left out of the original calculations. The additional information describes the exact distribution of fissile material in each container (DOE 2002a). The effects of the changes that have been included in the baseline design of the codisposal WP (CRWMS M andO 2000) are also investigated. The calculation determines the effective neutron multiplication factor (k eff ) for selected degraded mode internal configurations of the codisposal waste package. These calculations will support the demonstration of the technical viability of the design solution adopted for disposing of MOX (FFTF) spent nuclear fuel in the potential repository. This calculation is subject to the Quality Assurance Requirements and Description (QARD) (DOE 2002b) per the activity evaluation under work package number P6212310M2 in the technical work plan TWP-MGR-MD-0000101 (BSC 2002)

  9. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    1995-05-01

    The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

  10. Finite element transport methods for criticality calculations - current status and potential applications

    International Nuclear Information System (INIS)

    Oliveira, C.R.E. de; Goddard, A.

    1991-01-01

    In this paper we review the current status of the finite element method applied to the solution of the neutron transport equation and we discuss its potential role in the field of criticality safety. We show that the method's ability in handling complex, irregular geometry in two- and three-dimensions coupled with its accurate solutions potentially renders it an attractive alternative to the longer-established Monte Carlo method. Details of the most favoured form of the method - that which combines finite elements in space and spherical harmonics in angle - are presented. This form of the method, which has been extensively investigated over the last decade by research groups at the University of London, has been numerically implemented in the finite element code EVENT. The code has among its main features the capability of solving fixed source eigenvalue and time-dependent complex geometry problems in two- and three-dimensions. Other features of the code include anisotropic up- and down-scatter, direct and/or adjoint solutions and access to standard data libraries. Numerical examples, ranging from simple criticality benchmark studies to the analysis of idealised three-dimensional reactor cores, are presented to demonstrate the potential of the method. (author)

  11. Computerization of effluent management and external dose calculation using the 'ODCM' methodology applied to Almaraz-NPP

    International Nuclear Information System (INIS)

    Garcia Gutierrez, M.E.; Sustacha Duo, D.

    1993-01-01

    The ODCM (Offsite Dose Calculation Manual), the official operational document for all nuclear power plants develops the details for the technical specifications for discharges and governs their practical application. The use of ODCM methodology for managing and controlling data associated with radioactive discharges, as well as the subsequent processing of this data to assess the radiological impact, requires and generates a large volume of data, which demands the frequent application of laborious and complex calculation processes, making computerization necessary. The computer application created for Almaraz NPP has the capacity to store and manage data on all discharges, evaluate their effects, presents reports and copies the information to be sent periodically to the CSN (Spanish Nuclear Regulatory Commission) on a magnetic tape. The radiological impact of an actual or possible discharge can be evaluated at anytime and, furthermore, general or particular reports and graphs on the discharges and doses over time can be readily obtained. The application is run on a personal computer under a relational database management system. This interactive application is based on menus and windows. (author)

  12. ESCLOUD: A computer program to calculate the air concentration, deposition rate and external dose rate from a continuous discharge of radioactive material to atmosphere

    International Nuclear Information System (INIS)

    Jones, J.A.

    1980-03-01

    Radioactive material may be discharged to atmosphere in small quantities during the normal operation of a nuclear installation as part of a considered waste management practice. Estimates of the individual and collective dose equivalent rates resulting from such a discharge are required in a number of contexts: for example, in assessing compliance with dose limits, in estimating the radiological impact of the discharge and as an input into optimisation studies. The suite of programs which has been developed to undertake such calculations is made up of a number of independent modules one of which, ESCLOUD, is described in this report. The ESCLOUD program evaluates, as a function of distance and direction from the release point, the air concentration, deposition rate and external β and γ doses from airborne and deposited activity. The air concentration and deposition rate can be used as input to other modules for calculating inhalation and ingestion doses. (author)

  13. The critical role of logarithmic transformation in Nernstian equilibrium potential calculations.

    Science.gov (United States)

    Sawyer, Jemima E R; Hennebry, James E; Revill, Alexander; Brown, Angus M

    2017-06-01

    The membrane potential, arising from uneven distribution of ions across cell membranes containing selectively permeable ion channels, is of fundamental importance to cell signaling. The necessity of maintaining the membrane potential may be appreciated by expressing Ohm's law as current = voltage/resistance and recognizing that no current flows when voltage = 0, i.e., transmembrane voltage gradients, created by uneven transmembrane ion concentrations, are an absolute requirement for the generation of currents that precipitate the action and synaptic potentials that consume >80% of the brain's energy budget and underlie the electrical activity that defines brain function. The concept of the equilibrium potential is vital to understanding the origins of the membrane potential. The equilibrium potential defines a potential at which there is no net transmembrane ion flux, where the work created by the concentration gradient is balanced by the transmembrane voltage difference, and derives from a relationship describing the work done by the diffusion of ions down a concentration gradient. The Nernst equation predicts the equilibrium potential and, as such, is fundamental to understanding the interplay between transmembrane ion concentrations and equilibrium potentials. Logarithmic transformation of the ratio of internal and external ion concentrations lies at the heart of the Nernst equation, but most undergraduate neuroscience students have little understanding of the logarithmic function. To compound this, no current undergraduate neuroscience textbooks describe the effect of logarithmic transformation in appreciable detail, leaving the majority of students with little insight into how ion concentrations determine, or how ion perturbations alter, the membrane potential. Copyright © 2017 the American Physiological Society.

  14. Independent Calculations for the SR Can Assessment. External review contribution in support of SKI's and SSI's review of SR-Can

    Energy Technology Data Exchange (ETDEWEB)

    Maul, Philip; Robinson, Peter; Bond, Alex; Benbow, Steven (Quintessa Limited, Henley-on-Thames (GB))

    2008-03-15

    SKB has published the SR-Can assessment of a deep repository for spent fuel at either the Forsmark or Laxemar sites. This is the final assessment prior to a formal regulatory submission. A number of independent calculations have been undertaken in support of SKI's review of SR-Can. The types of calculations are: 1. direct checks of specified SKB calculations; 2. reproduction of SKB computer calculations with independent codes, to ensure that what SKB has done is properly understood, and to check that the calculations are properly documented; and 3. independent calculations to investigate particular aspects of the safety case. The data used by SKB in its Performance Assessment calculations have not been subject to detailed review. The independent calculations provide information on: 1. where independent calculations have been able to provide support for the arguments put forward by SKB; 2. areas where insufficient information has been provided by SKB to enable a third party to reproduce the SR-Can calculations; and 3. areas where calculations lead to questions about the validity of SKB's arguments. The timescale for the production of the present report has been determined by the timescales for SKI's review of the SR-Can assessment. As a result, some of the independent calculations referred to have not been fully documented, and this will be carried out in 2008. The following conclusions have been drawn. 1. SKB has worked hard to respond to criticisms of previous performance assessments, and SR-Can is an impressive piece of work. 2. In several areas either insufficient or inconsistent information has been presented so that a full reproduction of SKB's calculations has not been possible. This is an important area where SKB will need to improve the presentation of its assessment for SR-Site. 3. There are several areas where SKB's description of post-closure repository evolution needs to be further reviewed. Overall SKB have given only limited

  15. Main factors affecting strong ground motion calculations: Critical review and assessment

    International Nuclear Information System (INIS)

    Mohammadioun, B.; Pecker, A.

    1990-01-01

    In the interests of guarding lives and property against the effects of earthquakes, building codes are frequently enforced when erecting conventional structures, usually calling for simple, static calculations. Where more vulnerable facilities are involved, the failure of which, or of parts of which, could subject the environment to harmful substances, more sophisticated methods are used to compute or verify their design, often accompanied by safety margins intended to compensate for uncertainties encountered at various stages of the analysis that begins with input seismic data and culminates with an effective anti-seismic design. The forthcoming discussion will deal with what is known of the characteristics of strong ground motion, highly variable according to context, without entering into the problems raised by seismotectonic studies, which actually constitute the first aspect that must be addressed when performing such an analysis. Our conclusion will be devoted to cogent R and D work in this area

  16. Main factors affecting strong ground motion calculations: Critical review and assessment

    Energy Technology Data Exchange (ETDEWEB)

    Mohammadioun, B [DAS/SASC (France); Pecker, A [Societe Geodynamique et Structure (France)

    1990-07-01

    In the interests of guarding lives and property against the effects of earthquakes, building codes are frequently enforced when erecting conventional structures, usually calling for simple, static calculations. Where more vulnerable facilities are involved, the failure of which, or of parts of which, could subject the environment to harmful substances, more sophisticated methods are used to compute or verify their design, often accompanied by safety margins intended to compensate for uncertainties encountered at various stages of the analysis that begins with input seismic data and culminates with an effective anti-seismic design. The forthcoming discussion will deal with what is known of the characteristics of strong ground motion, highly variable according to context, without entering into the problems raised by seismotectonic studies, which actually constitute the first aspect that must be addressed when performing such an analysis. Our conclusion will be devoted to cogent R and D work in this area.

  17. MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution

  18. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are {sup 149}Sm, {sup 151}Sm, and {sup 155}Gd.

  19. OECD/NEA Burnup Credit Calculational Criticality Benchmark Phase I-B Results

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1993-01-01

    Burnup credit is an ongoing technical concern for many countries that operate commercial nuclear power reactors. In a multinational cooperative effort to resolve burnup credit issues, a Burnup Credit Working Group has been formed under the auspices of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. This working group has established a set of well-defined calculational benchmarks designed to study significant aspects of burnup credit computational methods. These benchmarks are intended to provide a means for the intercomparison of computer codes, methods, and data applied in spent fuel analysis. The benchmarks have been divided into multiple phases, each phase focusing on a particular feature of burnup credit analysis. This report summarizes the results and findings of the Phase I-B benchmark, which was proposed to provide a comparison of the ability of different code systems and data libraries to perform depletion analysis for the prediction of spent fuel isotopic concentrations. Results included here represent 21 different sets of calculations submitted by 16 different organizations worldwide, and are based on a limited set of nuclides determined to have the most important effect on the neutron multiplication factor of light-water-reactor spent fuel. A comparison of all sets of results demonstrates that most methods are in agreement to within 10% in the ability to estimate the spent fuel concentrations of most actinides. All methods are within 11% agreement about the average for all fission products studied. Furthermore, most deviations are less than 10%, and many are less than 5%. The exceptions are 149 Sm, 151 Sm, and 155 Gd

  20. Calculated critical parameters in simple geometries for oxide and nitrate water mixtures of U-233, U-235 and Pu-239 with thorium. Final report

    International Nuclear Information System (INIS)

    Converse, W.E.; Bierman, S.R.

    1979-11-01

    Calculations have been performed on water mixtures of oxides and nitrates of 233 U, 235 U, and 239 Pu with chemically similar thorium compounds to determine critical dimensions for simple geometries (sphere, cylinder, and slab). Uranium enrichments calculated were 100%, 20%, 10%, and 5%; plutonium calculations assumed 100% 239 Pu. Thorium to uranium or plutonium weight ratios (Th: U or Pu) calculated were 0, 1, 4, and 8. Both bare and full water reflection conditions were calculated. The results of the calculations are plotted showing a critical dimension versus the uranium or plutonium concentration. Plots of K-infinity and material buckling for each material type are also shown

  1. Benchmark calculation of SCALE-PC 4.3 CSAS6 module and burnup credit criticality analysis

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Hee Sung; Ro, Seong Gy; Shin, Young Joon; Kim, Ik Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-12-01

    Calculation biases of SCALE-PC CSAS6 module for PWR spent fuel, metallized spent fuel and solution of nuclear materials have been determined on the basis of the benchmark to be 0.01100, 0.02650 and 0.00997, respectively. With the aid of the code system, nuclear criticality safety analysis for the spent fuel storage pool has been carried out to determine the minimum burnup of spent fuel required for safe storage. The criticality safety analysis is performed using three types of isotopic composition of spent fuel: ORIGEN2-calculated isotopic compositions; the conservative inventory obtained from the multiplication of ORIGEN2-calculated isotopic compositions by isotopic correction factors; the conservative inventory of only U, Pu and {sup 241}Am. The results show that the minimum burnup for three cases are 990,6190 and 7270 MWd/tU, respectively in the case of 5.0 wt% initial enriched spent fuel. (author). 74 refs., 68 figs., 35 tabs.

  2. Intercomparison of radiotherapy treatment planning systems using calculated and measured dose distributions for external photon and electron beams

    International Nuclear Information System (INIS)

    Kosunen, A.; Jaervinen, H.; Vatnitskij, S.; Ermakov, I.; Chervjakov, A.; Kulmala, J.; Pitkaenen, M.; Vaeyrynen, T.; Vaeaenaenen, A.

    1991-02-01

    The requirement of 5 % overall accuracy for the target absorbed dose in radiotherapy implies that the accuracy of the relative dose calculation should be within only a few per cent. According to the recommendation by the International Commission on radiation units and measurements (ICRU), a computer-produced dose distribution can be considered to be accurate enough if it differs from the results of relative dose measurements by less than 2 %, or 2 mm in the position of isodose curves involving very steep dose gradients. In this study five treatment planning systems, currently used by the hospitals in Finland or in the USSR, were intercompared with respect to the above requirement. Five typical cases of irradiation were selected: regular fields, oblique incidence, irregular field, wedge field and inhomogeneity in a water equivalent phantom. Complete dose distributions were used for the intercomparison, and the beam data for each TPS was that pertaining to the beam where the comparative relative measurements were performed. The results indicate that the dose distributions produced by different TPS:s can differ from each other as well as from the measured dose distributions up to a level which is not acceptable in terms of the above requirement. Greatest differences seem to be related to the omission or undue consideration of the scatter components of the beam. A suitable quality assurance program for the systematic testing of the performance of the treatment planning systems could be based on a selection of tests as used in this study.(orig.)

  3. Track 4: basic nuclear science variance reduction for Monte Carlo criticality simulations. 6. Variational Variance Reduction for Monte Carlo Criticality Calculations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.; Larsen, Edward W.

    2001-01-01

    Recently, it has been shown that the figure of merit (FOM) of Monte Carlo source-detector problems can be enhanced by using a variational rather than a direct functional to estimate the detector response. The direct functional, which is traditionally employed in Monte Carlo simulations, requires an estimate of the solution of the forward problem within the detector region. The variational functional is theoretically more accurate than the direct functional, but it requires estimates of the solutions of the forward and adjoint source-detector problems over the entire phase-space of the problem. In recent work, we have performed Monte Carlo simulations using the variational functional by (a) approximating the adjoint solution deterministically and representing this solution as a function in phase-space and (b) estimating the forward solution using Monte Carlo. We have called this general procedure variational variance reduction (VVR). The VVR method is more computationally expensive per history than traditional Monte Carlo because extra information must be tallied and processed. However, the variational functional yields a more accurate estimate of the detector response. Our simulations have shown that the VVR reduction in variance usually outweighs the increase in cost, resulting in an increased FOM. In recent work on source-detector problems, we have calculated the adjoint solution deterministically and represented this solution as a linear-in-angle, histogram-in-space function. This procedure has several advantages over previous implementations: (a) it requires much less adjoint information to be stored and (b) it is highly efficient for diffusive problems, due to the accurate linear-in-angle representation of the adjoint solution. (Traditional variance-reduction methods perform poorly for diffusive problems.) Here, we extend this VVR method to Monte Carlo criticality calculations, which are often diffusive and difficult for traditional variance-reduction methods

  4. OECD/NEA expert group on uncertainty analysis for criticality safety assessment: Results of benchmark on sensitivity calculation (phase III)

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, T.; Laville, C. [Institut de Radioprotection et de Surete Nucleaire IRSN, BP 17, 92262 Fontenay aux Roses (France); Dyrda, J. [Atomic Weapons Establishment AWE, Aldermaston, Reading, RG7 4PR (United Kingdom); Mennerdahl, D. [E Mennerdahl Systems EMS, Starvaegen 12, 18357 Taeby (Sweden); Golovko, Y.; Raskach, K.; Tsiboulia, A. [Inst. for Physics and Power Engineering IPPE, 1, Bondarenko sq., 249033 Obninsk (Russian Federation); Lee, G. S.; Woo, S. W. [Korea Inst. of Nuclear Safety KINS, 62 Gwahak-ro, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Bidaud, A.; Sabouri, P. [Laboratoire de Physique Subatomique et de Cosmologie LPSC, CNRS-IN2P3/UJF/INPG, Grenoble (France); Patel, A. [U.S. Nuclear Regulatory Commission (NRC), Washington, DC 20555-0001 (United States); Bledsoe, K.; Rearden, B. [Oak Ridge National Laboratory ORNL, M.S. 6170, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Gulliford, J.; Michel-Sendis, F. [OECD/NEA, 12, Bd des Iles, 92130 Issy-les-Moulineaux (France)

    2012-07-01

    The sensitivities of the k{sub eff} eigenvalue to neutron cross sections have become commonly used in similarity studies and as part of the validation algorithm for criticality safety assessments. To test calculations of the sensitivity coefficients, a benchmark study (Phase III) has been established by the OECD-NEA/WPNCS/EG UACSA (Expert Group on Uncertainty Analysis for Criticality Safety Assessment). This paper presents some sensitivity results generated by the benchmark participants using various computational tools based upon different computational methods: SCALE/TSUNAMI-3D and -1D, MONK, APOLLO2-MORET 5, DRAGON-SUSD3D and MMKKENO. The study demonstrates the performance of the tools. It also illustrates how model simplifications impact the sensitivity results and demonstrates the importance of 'implicit' (self-shielding) sensitivities. This work has been a useful step towards verification of the existing and developed sensitivity analysis methods. (authors)

  5. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Volkov, Yu. N. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  6. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    International Nuclear Information System (INIS)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A.; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V.; Volkov, Yu. N.

    2016-01-01

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  7. Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

    International Nuclear Information System (INIS)

    Rafhayudi Jamro; Redzuwan Yahaya; Abdul Aziz Mohamed; Eid Abdel-Munem; Megat Harun Al-Rashid; Julia Abdul Karim; Ikki Kurniawan; Hafizal Yazid; Azraf Azman; Shukri Mohd

    2008-01-01

    A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (k e ff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k e ff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of k e ff from MCNP5 were obtained when the control fuel rods were fully inserted, the k e ff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of k e ff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

  8. Calculating the metabolizable energy of macronutrients: a critical review of Atwater's results.

    Science.gov (United States)

    Sánchez-Peña, M Judith; Márquez-Sandoval, Fabiola; Ramírez-Anguiano, Ana C; Velasco-Ramírez, Sandra F; Macedo-Ojeda, Gabriela; González-Ortiz, Luis J

    2017-01-01

    The current values for metabolizable energy of macronutrients were proposed in 1910. Since then, however, efforts to revise these values have been practically absent, creating a crucial need to carry out a critical analysis of the experimental methodology and results that form the basis of these values. Presented here is an exhaustive analysis of Atwater's work on this topic, showing evidence of considerable weaknesses that compromise the validity of his results. These weaknesses include the following: (1) the doubtful representativeness of Atwater's subjects, their activity patterns, and their diets; (2) the extremely short duration of the experiments; (3) the uncertainty about which fecal and urinary excretions contain the residues of each ingested food; (4) the uncertainty about whether or not the required nitrogen balance in individuals was reached during experiments; (5) the numerous experiments carried out without valid preliminary experiments; (6) the imprecision affecting Atwater's experimental measurements; and (7) the numerous assumptions and approximations, along with the lack of information, characterizing Atwater's studies. This review presents specific guidelines for establishing new experimental procedures to estimate more precise and/or more accurate values for the metabolizable energy of macronutrients. The importance of estimating these values in light of their possible dependence on certain nutritional parameters and/or physical activity patterns of individuals is emphasized. The use of more precise values would allow better management of the current overweight and obesity epidemic. © The Author(s) 2016. Published by Oxford University Press on behalf of the International Life Sciences Institute. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

  9. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Barranco R, F.

    2015-01-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  10. A Critical Review of Validation, Blind Testing, and Real- World Use of Alchemical Protein-Ligand Binding Free Energy Calculations.

    Science.gov (United States)

    Abel, Robert; Wang, Lingle; Mobley, David L; Friesner, Richard A

    2017-01-01

    Protein-ligand binding is among the most fundamental phenomena underlying all molecular biology, and a greater ability to more accurately and robustly predict the binding free energy of a small molecule ligand for its cognate protein is expected to have vast consequences for improving the efficiency of pharmaceutical drug discovery. We briefly reviewed a number of scientific and technical advances that have enabled alchemical free energy calculations to recently emerge as a preferred approach, and critically considered proper validation and effective use of these techniques. In particular, we characterized a selection bias effect which may be important in prospective free energy calculations, and introduced a strategy to improve the accuracy of the free energy predictions. Copyright© Bentham Science Publishers; For any queries, please email at epub@benthamscience.org.

  11. The Internal, External, and Diagnostic Validity of Sluggish Cognitive Tempo: A Meta-Analysis and Critical Review

    Science.gov (United States)

    Becker, Stephen P.; Leopold, Daniel R.; Burns, G. Leonard; Jarrett, Matthew A.; Langberg, Joshua M.; Marshall, Stephen A.; McBurnett, Keith; Waschbusch, Daniel A.; Willcutt, Erik G.

    2015-01-01

    Objective To conduct the first meta-analysis evaluating the internal and external validity of the sluggish cognitive tempo (SCT) construct as related to or distinct from attention-deficit/hyperactivity disorder (ADHD) and as associated with functional impairment and neuropsychological functioning. Method Electronic databases were searched through September 2015 for studies examining the factor structure and/or correlates of SCT in children or adults. The search procedures identified 73 papers. The core SCT behaviors included across studies, as well as factor loadings and reliability estimates, were reviewed to evaluate internal validity. Pooled correlation effect sizes using random effects models were used to evaluate SCT in relation to external validity domains (i.e., demographics, other psychopathologies, functional impairment, and neuropsychological functioning). Results Strong support was found for the internal validity of the SCT construct. Specifically, across factor analytic studies including over 19,000 individuals, 13 SCT items loaded consistently on an SCT factor as opposed to an ADHD factor. Findings also support the reliability (i.e., internal consistency, test-retest reliability, inter-rater reliability) of SCT. In terms of external validity, there is some indication that SCT may increase with age (r = 0.11) and be associated with lower socioeconomic status (r = 0.10). Modest (potentially negligible) support was found for SCT symptoms being higher in males than females in children (r = 0.05) but not adults. SCT is more strongly associated with ADHD inattention (r = 0.63 in children, r = 0.72 in adults) than with ADHD hyperactivity-impulsivity (r = 0.32 in children, r = 0.46 in adults), and it likewise appears that SCT is more strongly associated with internalizing symptoms than with externalizing symptoms. SCT is associated with significant global, social, and academic impairment (rs = 0.38–0.44). Effects for neuropsychological functioning are mixed

  12. A practical method to calculate head scatter factors in wedged rectangular and irregular MLC shaped beams for external and internal wedges

    International Nuclear Information System (INIS)

    Georg, Dietmar; Olofsson, Joergen; Kuenzler, Thomas; Aiginger, Hannes; Karlsson, Mikael

    2004-01-01

    Factor based methods for absorbed dose or monitor unit calculations are often based on separate data sets for open and wedged beams. The determination of basic beam parameters can be rather time consuming, unless equivalent square methods are applied. When considering irregular wedged beams shaped with a multileaf collimator, parametrization methods for dosimetric quantities, e.g. output ratios or wedge factors as a function of field size and shape, become even more important. A practical method is presented to derive wedged output ratios in air (S c,w ) for any rectangular field and for any irregular MLC shaped beam. This method was based on open field output ratios in air (S c ) for a field with the same collimator setting, and a relation f w between S c,w and S c . The relation f w can be determined from measured output ratios in air for a few open and wedged fields including the maximum wedged field size. The function f w and its parametrization were dependent on wedge angle and treatment head design, i.e. they were different for internal and external wedges. The proposed method was tested for rectangular wedged fields on three accelerators with internal wedges (GE, Elekta, BBC) and two accelerators with external wedges (Varian). For symmetric regular beams the average deviation between calculated and measured S c,w /S c ratios was 0.3% for external wedges and about 0.6% for internal wedges. Maximum deviations of 1.8% were obtained for elongated rectangular fields on the GE and ELEKTA linacs with an internal wedge. The same accuracy was achieved for irregular MLC shaped wedged beams on the accelerators with MLC and internal wedges (GE and Elekta), with an average deviation <1% for the fields tested. The proposed method to determine output ratios in air for wedged beams from output ratios of open beams, combined with equivalent square approaches, can be easily integrated in empirical or semi-empirical methods for monitor unit calculations

  13. A simple and fast physics-based analytical method to calculate therapeutic and stray doses from external beam, megavoltage x-ray therapy.

    Science.gov (United States)

    Jagetic, Lydia J; Newhauser, Wayne D

    2015-06-21

    State-of-the-art radiotherapy treatment planning systems provide reliable estimates of the therapeutic radiation but are known to underestimate or neglect the stray radiation exposures. Most commonly, stray radiation exposures are reconstructed using empirical formulas or lookup tables. The purpose of this study was to develop the basic physics of a model capable of calculating the total absorbed dose both inside and outside of the therapeutic radiation beam for external beam photon therapy. The model was developed using measurements of total absorbed dose in a water-box phantom from a 6 MV medical linear accelerator to calculate dose profiles in both the in-plane and cross-plane direction for a variety of square field sizes and depths in water. The water-box phantom facilitated development of the basic physical aspects of the model. RMS discrepancies between measured and calculated total absorbed dose values in water were less than 9.3% for all fields studied. Computation times for 10 million dose points within a homogeneous phantom were approximately 4 min. These results suggest that the basic physics of the model are sufficiently simple, fast, and accurate to serve as a foundation for a variety of clinical and research applications, some of which may require that the model be extended or simplified based on the needs of the user. A potentially important advantage of a physics-based approach is that the model is more readily adaptable to a wide variety of treatment units and treatment techniques than with empirical models.

  14. Ab initio supercell calculations of the (0001) α-Cr2O3 surface with a partially or totally Al-substituted external layer

    International Nuclear Information System (INIS)

    Sun Jizhong; Stirner, Thomas

    2009-01-01

    Ab initio supercell calculations employing the periodic Hartree-Fock formalism are presented of the (0001) α-Cr 2 O 3 surface with a partially or totally Al-substituted external layer. In the simulations a fraction of the Cr atoms at the surface of the chromia slab are replaced by Al atoms, and the Al surface coverage is varied between zero (pure chromia) and 100% (Al-terminated chromia). The surface Al atoms are found to relax inwards considerably, with the magnitude of the relaxation decreasing with increasing Al surface coverage. The calculations also reveal that the surface energy of the slab decreases with increasing Al coverage. Finally, the electronic properties at the surface of the Al-substituted (0001) α-Cr 2 O 3 slabs are investigated. Here the calculations show that the substitution of Cr by Al gives rise to an increase in the covalency of the Al-O bonds compared to slabs of pure alumina. In contrast, the influence of the surface Al atoms on the electrostatic potential in the (0001) plane of metal ions is relatively small. These findings support the utilisation of α-chromia substrates for the templated growth of α-alumina, which is consistent with recent experiments.

  15. A framework for the system-of-systems analysis of the risk for a safety-critical plant exposed to external events

    International Nuclear Information System (INIS)

    Zio, E.; Ferrario, E.

    2013-01-01

    We consider a critical plant exposed to risk from external events. We propose an original framework of analysis, which extends the boundaries of the study to the interdependent infrastructures which support the plant. For the purpose of clearly illustrating the conceptual framework of system-of-systems analysis, we work out a case study of seismic risk for a nuclear power plant embedded in the connected power and water distribution, and transportation networks which support its operation. The technical details of the systems considered (including the nuclear power plant) are highly simplified, in order to preserve the purpose of illustrating the conceptual, methodological framework of analysis. Yet, as an example of the approaches that can be used to perform the analysis within the proposed framework, we consider the Muir Web as system analysis tool to build the system-of-systems model and Monte Carlo simulation for the quantitative evaluation of the model. The numerical exercise, albeit performed on a simplified case study, serves the purpose of showing the opportunity of accounting for the contribution of the interdependent infrastructure systems to the safety of a critical plant. This is relevant as it can lead to considerations with respect to the decision making related to safety critical-issues. -- Highlights: ► We consider a critical plant exposed to risk from external events. ► We consider also the interdependent infrastructures that support the plant. ► We use Muir Web as system analysis tool to build the system-of-systems model. ► We use Monte Carlo simulation for the quantitative evaluation of the model. ► We find that the interdependent infrastructures should be considered as they can be a support for the critical plant safety

  16. Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

    International Nuclear Information System (INIS)

    Martinez-Gonzalez, Jesus S.; Ade, Brian J.; Bowman, Stephen M.; Gauld, Ian C.; Ilas, Germina; Marshall, William BJ J.

    2015-01-01

    Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

  17. On the contribution of external cost calculations to energy system governance: the case of a potential large-scale nuclear accident

    NARCIS (Netherlands)

    Laes, E.; Meskens, G.; van der Sluijs, J.P.|info:eu-repo/dai/nl/073427489

    2011-01-01

    The European Commission's ExternE Project has made major advances in the quantification of external costs of electricity. Although some impacts cannot be valued, important conclusions are possible. This paper outlines some provisional implications for energy policy. External costs are technology

  18. INDAR: a computer code for the calculation of critical group radiation exposure from routine discharges of radioactivity to seas and estuaries - description and users' guide

    International Nuclear Information System (INIS)

    Maul, P.R.; Wilmott, S.

    1988-12-01

    The computer program INDAR enables detailed estimates to be made of critical group radiation exposure arising from routine discharges of radioactivity for coastal sites where the discharge is close to the shore and the shoreline is reasonably straight, and for estuarine sites where radioactivity is rapidly mixed across the width of the estuary. Important processes which can be taken into account include the turbulence generated by the discharge, the effects of a sloping sea bed and the variation with time of the lateral dispersion coefficient. The significance of the timing of discharges can also be assessed. INDAR uses physically meaningful hydrographic parameters directly. For most sites the most important exposure pathways are seafood consumption, external exposure over estuarine sediments and beaches, and the handling of fishing gear. As well as for these primary pathways, INDAR enables direct calculations to be made for some additional exposure pathways. The secondary pathways considered are seaweed consumption, swimming, the handling of materials other than fishing gear and the inhalation of activity. (author)

  19. Research and design calculation of multipurpose critical assembly using moderated light water and low enriched fuel from 1.6 to 5.0% U-235

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Vo Doan Hai Dang; Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Minh Tuan; Nguyen Manh Hung; Pham Quang Huy; Tran Quoc Duong; Tran Tri Vien

    2015-01-01

    Basing on the idea in ??using fuel of nuclear power plants such as PWR (AP-1000) and VVER-1000 with light water as moderation, design calculation of critical assembly was performed to confirm the possibility of using these fuels. Designed critical assembly has simple structure consisting of low enriched fuel from 1.6% to 5% U-235; water has functions as cooling, biological protection and control. Critical assembly is operated at nominal power 100 W with fuel pitch about 2.0 cm. Applications of the critical assembly are quite abundant in basic research, education and training with low investment cost compare with research reactor and easy in operation. So critical assembly can be used for university or training centre for nuclear engineering training. Main objectives of the project are: design calculation in neutronics, thermal hydraulics and safety analysis for critical configuration benchmarks using low enriched fuel; design in mechanical and auxiliary systems for critical assembly; determine technical specifications and estimate construction, installation cost of critical assembly. The process of design, fabrication, installation and construction of critical assembly will be considered with different implementation phases and localization capabilities in installation of critical assembly is highly feasibility. Cost estimation of construction and installation of critical assembly was implemented and showed that investment cost for critical assembly is much lower than research reactor and most of components, systems of critical assembly can be localized with current technique quality of the country. (author)

  20. Effect of Transition From a Unit-Based Team to External Transport Team for a Pediatric Critical Care Unit.

    Science.gov (United States)

    Cummings, Brian M; Kaliannan, Kanakaraju; Yager, Phoebe H; Noviski, Natan

    2017-12-01

    Pediatric hospitals must consider staff, training, and direct costs required to maintain a pediatric specialized transport team, balanced with indirect potential benefits of marketing and referral volume. The effect of transitioning a unit-based transport team to an external service on the pediatric intensive care unit (PICU) is unknown, but information is needed as hospital systems focus on population management. We examined the impact on PICU transports after transition to an external transport vendor. Single-center retrospective review performed of PICU admissions, referrals, and transfers during baseline, post-, and maintenance period with a total of 9-year follow-up. Transfer volume was analyzed during pre-, post-, and maintenance phase with descriptive statistics and statistical process control charts from 1999 to 2012. Total PICU admissions increased with an annual growth rate of 3.7%, with mean annual 626 admissions prior to implementation to the mean of 890 admissions at the end of period, P < .001. The proportion of transport to total admissions decreased from 27% to 21%, but mean annual transports were unchanged, 175 to 183, P = .6, and mean referrals were similar, 186 to 203, P = .8. Seasonal changes in transport volume remained as a predominant source of variability. Annual transport refusals increased initially in the postimplementation phase, mean 11 versus 33, P < .03, but similar to baseline in the maintenance phase, mean 20/year, P = .07. Patient refusals were due to bed and staffing constraints, with 7% due to the lack of transport vendor availability. In a transition to a regional transport service, PICU transport volume was maintained in the long-term follow-up and total PICU admissions increased. Further research on the direct and indirect impact of transport regionalization is needed to determine the optimal cost-benefit and quality of care as health-care systems focus on population management.

  1. Calculation of critical heat transfer in horizontal evaporator pipes in cooling systems of high-rise buildings

    Science.gov (United States)

    Aksenov, Andrey; Malysheva, Anna

    2018-03-01

    An exact calculation of the heat exchange of evaporative surfaces is possible only if the physical processes of hydrodynamics of two-phase flows are considered in detail. Especially this task is relevant for the design of refrigeration supply systems for high-rise buildings, where powerful refrigeration equipment and branched networks of refrigerants are used. On the basis of experimental studies and developed mathematical model of asymmetric dispersed-annular flow of steam-water flow in horizontal steam-generating pipes, a calculation formula has been obtained for determining the boundaries of the zone of improved heat transfer and the critical value of the heat flux density. A new theoretical approach to the solution of the problem of the flow structure of a two-phase flow is proposed. The applied method of dissipative characteristics of a two-phase flow in pipes and the principle of a minimum rate of entropy increase in stabilized flows made it possible to obtain formulas that directly reflect the influence of the viscous characteristics of the gas and liquid media on their distribution in the flow. The study showed a significant effect of gravitational forces on the nature of the phase distribution in the cross section of the evaporative tubes. At a mass velocity of a two-phase flow less than 700 kg / m2s, the volume content of the liquid phase near the upper outer generating lines of the tube is almost an order of magnitude lower than the lower one. The calculation of the heat transfer crisis in horizontal evaporative tubes is obtained. The calculated dependence is in good agreement with the experimental data of the author and a number of foreign researchers. The formula generalizes the experimental data for pipes with the diameter of 6-40 mm in the pressure of 2-7 MPa.

  2. Numerical calculation procedure for criticality parameters of the two-zone reflected reactor with flat central zone

    International Nuclear Information System (INIS)

    Bosevski, T.; Strugar, P.

    1966-10-01

    In determining the criticality parameters of a two-zone reactor with flat central zone one encounters a numerical problem requiring the solution of a system of two non-linear equations. To solve them the Newton method, which proved convenient, was used n this work. By comparing our results with those reported one obtains about 5% smaller values of both the radius of the flat zone and of the radial buckling of the outer zone. This discrepancy probably results from some approximations used in solving the same system of equations used in solving the same system of equations where the procedure form was applied, whereas the calculation time is by one order of magnitude smaller

  3. Criticality and shielding calculations of an interim dry storage system for the spent fuel from Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    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

  4. Dosimetric quality control of treatment planning systems in external radiation therapy using Digital Test Objects calculated by PENELOPE Monte-Carlo simulations

    International Nuclear Information System (INIS)

    Ben Hdech, Yassine

    2011-01-01

    To ensure the required accuracy and prevent from mis-administration, cancer treatments, by external radiation therapy are simulated on Treatment Planning System or TPS before radiation delivery in order to ensure that the prescription is achieved both in terms of target volumes coverage and healthy tissues protection. The TPS calculates the patient dose distribution and the treatment time per beam required to deliver the prescribed dose. TPS is a key system in the decision process of treatment by radiation therapy. It is therefore essential that the TPS be subject to a thorough check of its performance (quality control or QC) and in particular its ability to accurately compute dose distributions for patients in all clinical situations that be met. The 'traditional' methods recommended to carry out dosimetric CQ of algorithms implemented in the TPS are based on comparisons between dose distributions calculated with the TPS and dose measured in physical test objects (PTO) using the treatment machine. In this thesis we propose to substitute the reference dosimetric measurements performed in OTP by benchmark dose calculations in Digital Test Objects using PENELOPE Monte-Carlo code. This method has three advantages: (i) it allows simulation in situations close to the clinic and often too complex to be experimentally feasible; (ii) due to the digital form of reference data the QC process may be automated; (iii) it allows a comprehensive TPS CQ without hindering the use of an equipment devoted primarily to patients treatments. This new method of CQ has been tested successfully on the Eclipse TPS from Varian Medical Systems Company. (author) [fr

  5. Studies of criticality Monte Carlo method convergence: use of a deterministic calculation and automated detection of the transient

    International Nuclear Information System (INIS)

    Jinaphanh, A.

    2012-01-01

    Monte Carlo criticality calculation allows to estimate the effective multiplication factor as well as local quantities such as local reaction rates. Some configurations presenting weak neutronic coupling (high burn up profile, complete reactor core,...) may induce biased estimations for k eff or reaction rates. In order to improve robustness of the iterative Monte Carlo methods, a coupling with a deterministic code was studied. An adjoint flux is obtained by a deterministic calculation and then used in the Monte Carlo. The initial guess is then automated, the sampling of fission sites is modified and the random walk of neutrons is modified using splitting and russian roulette strategies. An automated convergence detection method has been developed. It locates and suppresses the transient due to the initialization in an output series, applied here to k eff and Shannon entropy. It relies on modeling stationary series by an order 1 auto regressive process and applying statistical tests based on a Student Bridge statistics. This method can easily be extended to every output of an iterative Monte Carlo. Methods developed in this thesis are tested on different test cases. (author)

  6. Calculation of critical level value for radioactivity detection in gamma spectrometric analysis on the base of semiconductor detectors under the Chernobyl' conditions in 1986-1987

    International Nuclear Information System (INIS)

    Glazunov, V.O.; Rusyaev, R.V.

    1989-01-01

    The problem of determination of radioactivity critical level in a sample by means of gamma spectrometer with semiconductor detector is studied theoretically. The formula for critical level, which shows that it is necessary to know the background pulse counting rate in order to determine the minimum gamma photon pulse counting rates, is derived. Calculations of critical level for the Chernobyl' conditions in time period from October 1986 till July 1987 are made. 8 refs.; 7 figs.; 17 tabs

  7. Classification of criticality calculations with correlation coefficient method and its application to OECD/NEA burnup credit benchmarks phase III-A and II-A

    International Nuclear Information System (INIS)

    Okuno, Hiroshi

    2003-01-01

    A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the Nuclear Energy Agency of the Organisation for Economic Cooperation and Development (OECD/NEA). Phase III-A benchmark was a series of criticality calculations for irradiated Boiling Water Reactor (BWR) fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated Pressurized Water Reactor (PWR) fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark calculation results were classified according to the criterion that the values of the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of benchmark calculation results belonged to the same group, one calculation result was found predictable from the other. An example was shown for each of the Benchmarks. While the evaluated nuclear data seemed the main factor for the classification, further investigations were required for finding other factors. (author)

  8. The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 7. Organ doses due to parallel and environmental exposure geometries

    Energy Technology Data Exchange (ETDEWEB)

    Zankl, M. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Drexler, G. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Petoussi-Henss, N. [GSF - Forschungszentrum fuer Umwelt und Gesundheit Neuherberg GmbH, Oberschleissheim (Germany). Inst. fuer Strahlenschutz; Saito, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    1997-03-01

    This report presents a tabulation of organ and tissue equivalent dose as well as effective dose conversion coefficients, normalised to air kerma free in air, for occupational exposures and environmental exposures of the public to external photon radiation. For occupational exposures, whole-body irradiation with idealised geometries, i.e. broad parallel beams and fully isotropic radiation incidence, is considered. The directions of incidence for the parallel beams are anterior-posterior, posterior-anterior, left lateral, right lateral and a full 360 rotation around the body`s longitudinal axis. The influence of beam divergence on the body doses is also considered as well as the dependence of effective dose on the angle of radiation incidence. Regarding exposure of the public to environmental sources, three source geometries are considered: exposure from a radioactive cloud, from ground contamination and from the natural radionuclides distributed homogeneously in the ground. The precise angular and energy distributions of the gamma rays incident on the human body were taken into account. The organ dose conversion coefficients given in this catalogue were calculated using a Monte Carlo code simulating the photon transport in mathematical models of an adult male and an adult female, respectively. Conversion coefficients are given for the equivalent dose of 23 organs and tissues as well as for effective dose and the equivalent dose of the so-called `remainder`. The organ equivalent dose conversion coefficients are given separately for the adult male and female models and - as arithmetic mean of the conversion coefficients of both - for an average adult. Fitted data of the coefficients are presented in tables; the primary raw data as resulting from the Monte Carlo calculation are shown in figures together with the fitted data. (orig.)

  9. The calculation of dose from external photon exposures using reference human phantoms and Monte Carlo methods. Pt. 7. Organ doses due to parallel and environmental exposure geometries

    International Nuclear Information System (INIS)

    Zankl, M.

    1997-03-01

    This report presents a tabulation of organ and tissue equivalent dose as well as effective dose conversion coefficients, normalised to air kerma free in air, for occupational exposures and environmental exposures of the public to external photon radiation. For occupational exposures, whole-body irradiation with idealised geometries, i.e. broad parallel beams and fully isotropic radiation incidence, is considered. The directions of incidence for the parallel beams are anterior-posterior, posterior-anterior, left lateral, right lateral and a full 360 rotation around the body's longitudinal axis. The influence of beam divergence on the body doses is also considered as well as the dependence of effective dose on the angle of radiation incidence. Regarding exposure of the public to environmental sources, three source geometries are considered: exposure from a radioactive cloud, from ground contamination and from the natural radionuclides distributed homogeneously in the ground. The precise angular and energy distributions of the gamma rays incident on the human body were taken into account. The organ dose conversion coefficients given in this catalogue were calculated using a Monte Carlo code simulating the photon transport in mathematical models of an adult male and an adult female, respectively. Conversion coefficients are given for the equivalent dose of 23 organs and tissues as well as for effective dose and the equivalent dose of the so-called 'remainder'. The organ equivalent dose conversion coefficients are given separately for the adult male and female models and - as arithmetic mean of the conversion coefficients of both - for an average adult. Fitted data of the coefficients are presented in tables; the primary raw data as resulting from the Monte Carlo calculation are shown in figures together with the fitted data. (orig.)

  10. HTTR criticality calculations with SCALE6: Studies of various geometric and unit-cell options in modeling

    Energy Technology Data Exchange (ETDEWEB)

    Wang, J. Y.; Chiang, M. H.; Sheu, R. J.; Liu, Y. W. H. [Inst. of Nuclear Engineering and Science, National Tsing Hua Univ., Hsinchu 30013, Taiwan (China)

    2012-07-01

    The fuel element of the High Temperature Engineering Test Reactor (HTTR) presents a doubly heterogeneous geometry, where tiny TRISO fuel particles dispersed in a graphite matrix form the fuel region of a cylindrical fuel rod, and a number of fuel rods together with moderator or reflector then constitute the lattice design of the core. In this study, a series of full-core HTTR criticality calculations were performed with the SCALE6 code system using various geometric and unit-cell options in order to systematically investigate their effects on neutronic analysis. Two geometric descriptions (ARRAY or HOLE) in SCALE6 can be used to construct a complicated and repeated model. The result shows that eliminating the use of HOLE in the HTTR geometric model can save the computation time by a factor of 4. Four unit-cell treatments for resonance self-shielding corrections in SCALE6 were tested to create problem-specific multigroup cross sections for the HTTR core model. Based on the same ENDF/B-VII cross-section library, their results were evaluated by comparing with continuous-energy calculations. The comparison indicates that the INFHOMMEDIUM result overestimates the system multiplication factor (k{sub eff}) by 55 mk, whereas the LATTICECELL and MULTIREGION treatments predict the k{sub eff} values with similar biases of approximately 10 mk overestimation. The DOUBLEHET result shows a more satisfactory agreement, about 4.2 mk underestimation in the k{sub eff} value. In addition, using cell-weighted cross sections instead of an explicit modeling of TRISO particles in fuel region can further reduce the computation time by a factor of 5 without sacrificing accuracy. (authors)

  11. Adult echocardiographic nomograms: overview, critical review and creation of a software for automatic, fast and easy calculation of normal values.

    Science.gov (United States)

    Cantinotti, Massimiliano; Giordano, Raffaele; Paterni, Marco; Saura, Daniel; Scalese, Marco; Franchi, Eliana; Assanta, Nadia; Koestenberg, Martin; Dulgheru, Raluca; Sugimoto, Tadafumi; Bernard, Anne; Caballero, Luis; Lancellotti, Patrizio

    2017-12-01

    There is a crescent interest on normal adult echocardiographic values and the introduction of new deformation imaging and 3D parameters pose the issue of normative data. A multitude of nomograms has been recently published, however data are often fragmentary, difficult to find, and their strengths/limitations have been never evaluated. (I) to provide a review of current echocardiographic nomograms; (II) to generate a tool for easy and fast access to these data. A literature search was conducted accessing the National Library of Medicine using the keywords: 2D/3D echocardiography, strain, left/right ventricle, atrial, mitral/tricuspid valve, aorta, reference values/nomograms/normal values. Adding the following keywords, the results were further refined: range/intervals, myocardial velocity, strain rate and speckle tracking. Forty one published studies were included. Our study reveals that for several of 2D/3D parameters sufficient normative data exist, however, a few limitations still persist. For some basic parameters (i.e., mitral/tricuspid/pulmonary valves, great vessels) and for 3D valves data are scarce. There is a lack of studies evaluating ethnic differences. Data have been generally expressed as mean values normalised for gender and age instead of computing models incorporating different variables (age/gender/body sizes) to calculate z scores. To summarize results a software ( Echocardio-Normal Values ) who automatically calculate range of normality for a broad range of echocardiographic measurements according to age/gender/weight/height, has been generated. We provide an up-to-date and critical review of strengths/limitation of current adult echocardiographic nomograms. Furthermore we generated a software for automatic, easy and fast access to multiple echocardiographic normative data.

  12. OECD/NEA burnup credit criticality benchmarks phase IIIB: Burnup calculations of BWR fuel assemblies for storage and transport

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of ±10% relative to the average, although some results, esp. 155 Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k ∞ also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  13. OECD/NEA burnup credit criticality benchmarks phase IIIB. Burnup calculations of BWR fuel assemblies for storage and transport

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi; Naito, Yoshitaka; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-02-01

    The report describes the final results of the Phase IIIB Benchmark conducted by the Expert Group on Burnup Credit Criticality Safety under the auspices of the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD). The Benchmark was intended to compare the predictability of current computer code and data library combinations for the atomic number densities of an irradiated PWR fuel assembly model. The fuel assembly was irradiated under specific power of 25.6 MW/tHM up to 40 GWd/tHM and cooled for five years. The void fraction was assumed to be uniform throughout the channel box and constant, at 0, 40 and 70%, during burnup. In total, 16 results were submitted from 13 institutes of 7 countries. The calculated atomic number densities of 12 actinides and 20 fission product nuclides were found to be for the most part within a range of {+-}10% relative to the average, although some results, esp. {sup 155}Eu and gadolinium isotopes, exceeded the band, which will require further investigation. Pin-wise burnup results agreed well among the participants. The results in the infinite neutron multiplication factor k{sub {infinity}} also accorded well with each other for void fractions of 0 and 40%; however some results deviated from the averaged value noticeably for the void fraction of 70%. (author)

  14. Literature research concerning alternative methods for validation of criticality calculation systems; Literaturrecherche zu alternativen Daten und Methoden zur Validierung von Kritikalitaetsrechensystemen

    Energy Technology Data Exchange (ETDEWEB)

    Behler, Matthias

    2016-05-15

    Beside radiochemical analysis of irradiated fuel and critical experiments, which has become a well-established basis for the validation of depletion code and criticality codes respectively, also results of oscillation experiments or the operating conditions of power reactor and research reactors can provide useful information for the validation of the above mentioned codes. Based on a literature review the potential of the utilization of oscillation experiment measurements for the validation of criticality codes is estimated. It is found that the reactivity measurements for actinides and fission products within the CERES program on the reactors DIMPLE (Winfrith, UK) and MINERVE (Cadarache, France) can give a valuable addition to the commonly used critical experiments for criticality code validation. However, there are approaches but yet no generally satisfactory solution for integrating the reactivity measurements in a quantitative bias determination for the neutron multiplication factor of typical application cases including irradiated spent fuel outside reactor cores, calculated using common criticality codes.

  15. Validation of KENO V.a. and two cross-section libraries for criticality calculations of low-enriched uranium systems

    International Nuclear Information System (INIS)

    Easter, M.E.

    1985-07-01

    The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had 235 U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded

  16. Calculation methodology validation. Pt. 2/01-R. Calculation of the multiplication factor for eight experiments with a critical set of nineteen VVER-440 fuel assemblies

    International Nuclear Information System (INIS)

    Kyncl, J.

    2001-04-01

    Comparison calculations were performed for 8 experiments accomplished in 2000 on the LR-0 reactor. The MCNP4a code was applied using effective cross section data in the continuous representation as per the ENDF/B-VI library. (P.A.)

  17. The measurement and calculation of the kinetic parameter {beta}{sub eff}/{Lambda} of a small high-temperature like, critical system

    Energy Technology Data Exchange (ETDEWEB)

    Wallerbos, E.J.M.; Hoogenboom, J.E. [Interfaculty Reactor Inst., Delft Univ. of Technology, Delft (Netherlands)

    1998-01-01

    This paper demonstrates that it is well possible to determine the kinetic parameter {beta}{sub eff}/{Lambda} in a neutronically very slow system by means of noise measurements in the critical state. The advantages of this technique are that it can be conducted in a critical reactor directly, and that no special measurement equipment is needed. The comparison to calculated values for four configurations, which differ in the amount of moderation in the core region, shows a satisfactory agreement. (author)

  18. On the implementation of new versions of the algorithms of calculation of dose absorbed in radiotherapy external; Sobre la implementacion de nuevas versiones de los algoritmos de calculo de dosis absorbida en radioterapia externa

    Energy Technology Data Exchange (ETDEWEB)

    Latorre-Musoll, A.; Carrasco de Fez, P.; Lizondo Gisbert, M.; Jordi-Ollero, O.; Jornet Sala, N.; Eudaldo Puell, T.; Ruiz Martinez, A.; Ribas Morales, M.

    2015-07-01

    The changes of version of the algorithms of calculation of dose absorbed in radiotherapy external should implement in a time reduced due to the pressure care. A set reduced of checks could pass by high discrepancies significant between the stones and the measures experimental, as illustrate in this work. (Author)

  19. Regional Externalities

    NARCIS (Netherlands)

    Heijman, W.J.M.

    2007-01-01

    The book offers practical and theoretical insights in regional externalities. Regional externalities are a specific subset of externalities that can be defined as externalities where space plays a dominant role. This class of externalities can be divided into three categories: (1) externalities

  20. Effects of internal and external scatter on the build-up characteristics of Monte Carlo calculated absorbed dose for electron irradiation

    International Nuclear Information System (INIS)

    Lin, H.; Wu, DS.; Wu, AD.

    2005-01-01

    The effects of internal and external scatter on surface, build-up and depth dose characteristics simulated by Monte Carlo code EGSnrc for varying field size and SSD for a 10 MeV monoenergetic electron beam with and without an accelerator model are extensively studied in this paper. In particular, sub-millimetre surface PDD was investigated. The percentage depth doses affected significantly by the external scatter show a larger build-up dose. A forward shifted Dmax depth and a sharper fall-off region compared to PDDs with only internal scatter considered. The surface dose with both internal and external scatter shows a marked decrease at 110 cm SSD, and then slight further changes with the increasing SSD since few external scattered particles from accelerator model can reach the phantom for large SSDs. The sharp PDD increase for the 5 cm x 5 cm field compared to other fields seen when only internal scatter is considered is significantly less when external scatter is also present. The effect of external scatter on surface PDD is more pronounced for large fields than small fields (5 cm x 5 cm field)

  1. Critical and subcritical mass calculations of curium-243 to -247 based on JENDL-3.2 for revision of ANSI/ANS-8.15

    International Nuclear Information System (INIS)

    Okuno, Hiroshi

    2002-01-01

    Critical and subcritical masses were calculated for a sphere of five curium isotopes from 243 Cm to 247 Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%Δk/k relative differences in the neutron multiplication factor for a homogeneous mixture of 243 Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for 246 Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized. (author)

  2. Critical and subcritical mass calculations of curium-243 to -247 based on JENDL-3.2 for revision of ANSI/ANS-8.15

    Energy Technology Data Exchange (ETDEWEB)

    Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawasaki, Hiromitsu [CRC Solutions Corporation, Hitachinaka, Ibaraki (Japan)

    2002-10-01

    Critical and subcritical masses were calculated for a sphere of five curium isotopes from {sup 243}Cm to {sup 247}Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%{delta}k/k relative differences in the neutron multiplication factor for a homogeneous mixture of {sup 243}Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for {sup 246}Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized. (author)

  3. Critical experiments and nuclear calculations - LAMPRE-I; Experiences critiques et calculs nucleaires concernant le LAMPRE-I; Kriticheskie opyty i yadernye raschety - LAMPRE-I; Experimentos criticos u calculos nucleares relativos al LAMPRE-I

    Energy Technology Data Exchange (ETDEWEB)

    Battat, M E [Los Alamos Scientific Laboratory, University of California, Los Alamos, NM (United States)

    1962-03-15

    As part of a programme to develop plutonium fuels for fast-breeder reactors, the Los Alamos Scientific Laboratory has constructed and is operating a 1-MW sodium-cooled test reactor whose core contains a molten alloy of plutonium andiron (90 at. % Pu, 10 at. % Fe, m.p. 410 deg. C). Reactivity control is provided by the use of a stainless-steel reflector and four nickel control-rods located external to the core. Experiments have been performed at core temperatures (isothermal) of 80, 160 and 480 deg. C to determine critical mass and reflector worth at each of these temperatures. Control-rod worths, from period measurements, and temperature coefficient of reactivity were also measured. Calculations have been made, using the S{sub n} method for solving the neutron transport problem, to determine the basic nuclear parameters of the system. The comparison between calculated and measured values of parameters such as temperature coefficient, control-element worths, and critical mass is also of interest in evaluating the reliability of the design calculations. (author) [French] Un reacteur d'essais de 1 MW refroidi au sodium, dont le coeur contient un alliage fondu de plutonium et defer (90 at. % Pu, 10 at. % Fe, p. f. 410 deg. C), a ete construit et est en fonctionnement au Laboratoire scientifique de Los Alamos, dans le cadre d'un programme d'etudes sur les combustibles au plutonium pour reacteurs surgenerateurs a neutrons rapides. Le controle de la reactivite est assure au moyen d'un reflecteur en acier inoxydable et de quatre barres de controle en nickel, a l'exterieur du coeur. On a fait des experiences a des temperatures du coeur de 80, 160 et 480 deg. C afin de determiner la masse critique et la quantite de reflecteur qui correspond a chacune de ces temperatures. On a aussi mesure l'efficacite des barres de controle, a partir de mesures de periode, ainsi que le coefficient thermique de reactivite. Afin de determiner les parametres nucleaires de base du reacteur, on a

  4. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    International Nuclear Information System (INIS)

    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

  5. Nuclear criticality safety experiments, calculations, and analyses - 1958 to 1982. Volume 2. Summaries. Complilation of papers from the Transactions of the American Nuclear Society

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, B.L.; Hampel, V.E.

    1982-10-21

    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.

  6. Calculation methods for determining dose equivalent

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Tanner, J.E.; Scherpelz, R.I.; Hadlock, D.E.

    1988-01-01

    A series of calculations of neutron fluence as a function of energy in an anthropomorphic phantom was performed to develop a system for determining effective dose equivalent for external radiation sources. critical organ dose equivalents are calculated and effective dose equivalents are determined using ICRP-26 methods. Quality factors based on both present definitions and ICRP-40 definitions are used in the analysis. The results of these calculations are presented and discussed

  7. KENO-VI Primer: A Primer for Criticality Calculations with SCALE/KENO-VI Using GeeWiz

    International Nuclear Information System (INIS)

    Bowman, Stephen M.

    2008-01-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system developed at Oak Ridge National Laboratory is widely used and accepted around the world for criticality safety analyses. The well-known KENO-VI three-dimensional Monte Carlo criticality computer code is one of the primary criticality safety analysis tools in SCALE. The KENO-VI primer is designed to help a new user understand and use the SCALE/KENO-VI Monte Carlo code for nuclear criticality safety analyses. It assumes that the user has a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with SCALE/KENO-VI in particular. The primer is designed to teach by example, with each example illustrating two or three features of SCALE/KENO-VI that are useful in criticality analyses. The primer is based on SCALE 6, which includes the Graphically Enhanced Editing Wizard (GeeWiz) Windows user interface. Each example uses GeeWiz to provide the framework for preparing input data and viewing output results. Starting with a Quickstart section, the primer gives an overview of the basic requirements for SCALE/KENO-VI input and allows the user to quickly run a simple criticality problem with SCALE/KENO-VI. The sections that follow Quickstart include a list of basic objectives at the beginning that identifies the goal of the section and the individual SCALE/KENO-VI features that are covered in detail in the sample problems in that section. Upon completion of the primer, a new user should be comfortable using GeeWiz to set up criticality problems in SCALE/KENO-VI. The primer provides a starting point for the criticality safety analyst who uses SCALE/KENO-VI. Complete descriptions are provided in the SCALE/KENO-VI manual. Although the primer is self-contained, it is intended as a companion volume to the SCALE/KENO-VI documentation. (The SCALE manual is provided on the SCALE installation DVD.) The primer provides specific examples of

  8. Calculation of normal tissue complication probability and dose-volume histogram reduction schemes for tissues with a critical element architecture

    International Nuclear Information System (INIS)

    Niemierko, Andrzej; Goitein, Michael

    1991-01-01

    The authors investigate a model of normal tissue complication probability for tissues that may be represented by a critical element architecture. They derive formulas for complication probability that apply to both a partial volume irradiation and to an arbitrary inhomogeneous dose distribution. The dose-volume isoeffect relationship which is a consequence of a critical element architecture is discussed and compared to the empirical power law relationship. A dose-volume histogram reduction scheme for a 'pure' critical element model is derived. In addition, a point-based algorithm which does not require precomputation of a dose-volume histogram is derived. The existing published dose-volume histogram reduction algorithms are analyzed. The authors show that the existing algorithms, developed empirically without an explicit biophysical model, have a close relationship to the critical element model at low levels of complication probability. However, it is also showed that they have aspects which are not compatible with a critical element model and the authors propose a modification to one of them to circumvent its restriction to low complication probabilities. (author). 26 refs.; 7 figs

  9. Calculation of the critical buckling of a lattice based on the integral form of the transport equation

    International Nuclear Information System (INIS)

    Benoist, P.

    1990-06-01

    The migration area, which relates the buckling to the multiplication factor, can be calculated by means of the Deniz formula. This formula involves the direct and adjoint angular fluxes. It is shown in this note that it is possible, using the integral form of the transport equation, to establish an equivalent formula in which only angle-integrated quantities appear. This formulation is more suitable for the calculation by the collision probably method [fr

  10. Experimental data and calculation studies of critical heat fluxes at local disturbances of geometry of WWER fuel assemblies

    International Nuclear Information System (INIS)

    Kobzar, L.L.; Oleksyuk, D.A.

    2001-01-01

    The results of experiments executed in RRC 'Kurchatov Institute on the thermal-physical critical facility SVD are presented herein. The experiments modeled the drawing of two fuel rods to each other till touching WWER-1000 reactor in FA. The experimental model is a 7-rod bundle with the heated length of 1 m. The primary goal of experiments was to acquire the quantitative factors of the reduction in the critical heat fluxes as contrasted to the basic model (without disturbances of FA geometry) at the expense of local disturbance of a rod bundle geometry. As it follows from the experiment, the effect of decrease of the critical heat rate depends on combination of regime parameters and it makes 15% in the most unfavorable case (Authors)

  11. Development of a three dimensional homogeneous calculation model for the BFS-62 critical experiment. Preparation of adjusted equivalent measured values for sodium void reactivity values. Final report

    International Nuclear Information System (INIS)

    Manturov, G.; Semenov, M.; Seregin, A.; Lykova, L.

    2004-01-01

    The BFS-62 critical experiments are currently used as 'benchmark' for verification of IPPE codes and nuclear data, which have been used in the study of loading a significant amount of Pu in fast reactors. The BFS-62 experiments have been performed at BFS-2 critical facility of IPPE (Obninsk). The experimental program has been arranged in such a way that the effect of replacement of uranium dioxied blanket by the steel reflector as well as the effect of replacing UOX by MOX on the main characteristics of the reactor model was studied. Wide experimental program, including measurements of the criticality-keff, spectral indices, radial and axial fission rate distributions, control rod mock-up worth, sodium void reactivity effect SVRE and some other important nuclear physics parameters, was fulfilled in the core. Series of 4 BFS-62 critical assemblies have been designed for studying the changes in BN-600 reactor physics from existing state to hybrid core. All the assemblies are modeling the reactor state prior to refueling, i.e. with all control rod mock-ups withdrawn from the core. The following items are chosen for the analysis in this report: Description of the critical assembly BFS-62-3A as the 3rd assembly in a series of 4 BFS critical assemblies studying BN-600 reactor with MOX-UOX hybrid zone and steel reflector; Development of a 3D homogeneous calculation model for the BFS-62-3A critical experiment as the mock-up of BN-600 reactor with hybrid zone and steel reflector; Evaluation of measured nuclear physics parameters keff and SVRE (sodium void reactivity effect); Preparation of adjusted equivalent measured values for keff and SVRE. Main series of calculations are performed using 3D HEX-Z diffusion code TRIGEX in 26 groups, with the ABBN-93 cross-section set. In addition, precise calculations are made, in 299 groups and Ps-approximation in scattering, by Monte-Carlo code MMKKENO and discrete ordinate code TWODANT. All calculations are based on the common system

  12. Group additivity calculations of the thermodynamic properties of unfolded proteins in aqueous solution: a critical comparison of peptide-based and HKF models.

    Science.gov (United States)

    Hakin, A W; Hedwig, G R

    2001-02-15

    A recent paper in this journal [Amend and Helgeson, Biophys. Chem. 84 (2000) 105] presented a new group additivity model to calculate various thermodynamic properties of unfolded proteins in aqueous solution. The parameters given for the revised Helgeson-Kirkham-Flowers (HKF) equations of state for all the constituent groups of unfolded proteins can be used, in principle, to calculate the partial molar heat capacity, C(o)p.2, and volume, V2(0), at infinite dilution of any polypeptide. Calculations of the values of C(o)p.2 and V2(0) for several polypeptides have been carried out to test the predictive utility of the HKF group additivity model. The results obtained are in very poor agreement with experimental data, and also with results calculated using a peptide-based group additivity model. A critical assessment of these two additivity models is presented.

  13. Quality control of the treatment planning systems dose calculations in external radiation therapy using the Penelope Monte Carlo code; Controle qualite des systemes de planification dosimetrique des traitements en radiotherapie externe au moyen du code Monte-Carlo Penelope

    Energy Technology Data Exchange (ETDEWEB)

    Blazy-Aubignac, L

    2007-09-15

    The treatment planning systems (T.P.S.) occupy a key position in the radiotherapy service: they realize the projected calculation of the dose distribution and the treatment duration. Traditionally, the quality control of the calculated distribution doses relies on their comparisons with dose distributions measured under the device of treatment. This thesis proposes to substitute these dosimetry measures to the profile of reference dosimetry calculations got by the Penelope Monte-Carlo code. The Monte-Carlo simulations give a broad choice of test configurations and allow to envisage a quality control of dosimetry aspects of T.P.S. without monopolizing the treatment devices. This quality control, based on the Monte-Carlo simulations has been tested on a clinical T.P.S. and has allowed to simplify the quality procedures of the T.P.S.. This quality control, in depth, more precise and simpler to implement could be generalized to every center of radiotherapy. (N.C.)

  14. Results of calculations of external gamma radiation exposure rates from local fallout and the related radionuclide compositions of two hypothetical 1-MT nuclear bursts. Final report

    International Nuclear Information System (INIS)

    Hicks, H.

    1984-12-01

    This report presents data on calculated gamma radiation exposure rates and local surface deposition of related radionuclides resulting from two hypothetical 1-Mt nuclear bursts. Calculations are made of the debris from two types of bombs: one containing 235 U as a fissionable material (designated oralloy), the other containing 238 U (designated tuballoy). 4 references

  15. Criticality calculation of the nuclear material warehouse of the ININ; Calculo de criticidad del almacen del material nuclear del ININ

    Energy Technology Data Exchange (ETDEWEB)

    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)

  16. Application of the single-channel continuous synthesis method to criticity and power distribution calculations in thermal reactors

    International Nuclear Information System (INIS)

    Medrano Asensio, Gregorio.

    1976-06-01

    A detailed power distribution calculation in a large power reactor requires the solution of the multigroup 3D diffusion equations. Using the finite difference method, this computation is too expensive to be performed for design purposes. This work is devoted to the single channel continous synthesis method: the choice of the trial functions and the determination of the mixing functions are discussed in details; 2D and 3D results are presented. The method is applied to the calculation of the IAEA ''Benchmark'' reactor and the results obtained are compared with a finite element resolution and with published results [fr

  17. Neutronic calculations with transport and diffusion computer codes for light water moderated critical with UO2 enriched at 4,75% as fuel

    International Nuclear Information System (INIS)

    Sabundjian, G.; Nakata, H.

    1983-02-01

    The neutronic calculational procedure in a 4,75% w/O enriched UO 2 fueled light water moderated critical assembly was tested, using the transport codes and diffusin code available at the Instituto de Pesquisas Energeticas e Nucleares. The results of the tested codes, LEOPARD, CITHAMMER, LASER, GELS and CITATION, were found to be satisfatory and only a slight advantage is presented by CITHAMMER code. (Author) [pt

  18. Eilenberger equation for rotating superfluid 3He and calculation of the upper critical angular velocity Ω/sub c/2

    International Nuclear Information System (INIS)

    Schopohl, N.

    1980-01-01

    On the basis of Gorkov's formulation of superconductivity theory, generalized Eilenberger equations are derived which apply to rotating superfluid 3 He in the presence of a magnetic field h and finite superflow v. In analyogy to conventional type II superconductors, the possibility of vortex solutions in discussed. An implicit equation determining the upper critical angular velocity Ω/sub c/2 as a function of temperature T, magnetic field h, and superflow v parallel to the rotation axis is-inferred from the linearized Eilenberger equations. In contrast to the case of slowly rotating 3 He-A, the solution of the eigenvalue problem determining the order parameter Δ near the the upper critical angular velocity admits no coreless vortex no coreless solutions. The space-dependent amplitude of the order parameter is analogous to Abrikosov's vortex array solution, while the spin-orbit part is given either by a polar-state type or an Anderson-Brinkman-Morel (ABM)-state-type eigensolution. Among the possible eigensolutions the polar-state type yields for vanishing superflow v the highest critical rotation frequency. For finite superflow v parallel to the rotation axis, however, the ABM-state-type solution is stabilized in comparison to the polar state for Vertical BarvVertical Bar> or approx. =0.2π(Tc/sub c/0/T/sub F/)v/sub f/ at zero temperature

  19. Critical Length Criterion and the Arc Chain Model for Calculating the Arcing Time of the Secondary Arc Related to AC Transmission Lines

    International Nuclear Information System (INIS)

    Cong Haoxi; Li Qingmin; Xing Jinyuan; Li Jinsong; Chen Qiang

    2015-01-01

    The prompt extinction of the secondary arc is critical to the single-phase reclosing of AC transmission lines, including half-wavelength power transmission lines. In this paper, a low-voltage physical experimental platform was established and the motion process of the secondary arc was recorded by a high-speed camera. It was found that the arcing time of the secondary arc rendered a close relationship with its arc length. Through the input and output power energy analysis of the secondary arc, a new critical length criterion for the arcing time was proposed. The arc chain model was then adopted to calculate the arcing time with both the traditional and the proposed critical length criteria, and the simulation results were compared with the experimental data. The study showed that the arcing time calculated from the new critical length criterion gave more accurate results, which can provide a reliable criterion in term of arcing time for modeling and simulation of the secondary arc related with power transmission lines. (paper)

  20. Results of calculations of external gamma radiation exposure rates from fallout and the related radionuclide compositions. Operation Tumbler-Snapper, 1952

    International Nuclear Information System (INIS)

    Hicks, H.G.

    1981-07-01

    This report presents data on calculated gamma radiation exposure rates and ground deposition of related radionuclides resulting from Events that deposited detectable radioactivity outside the Nevada Test Site complex

  1. MCNP calculations for criticality-safety benchmarks with ENDF/B-V and ENDF/B-VI libraries

    International Nuclear Information System (INIS)

    Iverson, J.L.; Mosteller, R.D.

    1995-01-01

    The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k eff are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of 233 U

  2. Development and Validation of NODAL-LAMBDA Program for the Calculation of the Sub-criticality of LAMDA MODES By Nodal Methods in BWR reactors

    International Nuclear Information System (INIS)

    Munoz-Cobo, J. L.; Merino, R.; Escriva, A.; Melara, J.; Concejal, A.

    2014-01-01

    We have developed a 3D code with two energy groups and diffusion theory that is capable of calculating eigenvalues lambda of a BWR reactor using nodal methods and boundary conditions that calculates ALBEDO NODAL-LAMBDA from the properties of the reflector code itself. The code calculates the sub-criticality of the first harmonic, which is involved in the stability against oscillations reactor out of phase, and which is needed for calculating the decay rate for data out of phase oscillations. The code is very fast and in a few seconds is able to make a calculation of the first eigenvalues and eigenvectors, discretized solving the problem with different matrix elements zero. The code uses the LAPACK and ARPACK libraries. It was necessary to modify the LAPACK library to perform various operations with five non-diagonal matrices simultaneously in order to reduce the number of calls to bookstores and simplify the procedure for calculating the matrices in compressed format CSR. The code is validated by comparing it with the results for SIMULATE different cases and making 3D BENCHMAR of the IAEA. (Author)

  3. Application of the “best representativity” method to a PWR fuel calculation using the critical experiments at the Toshiba NCA facility

    International Nuclear Information System (INIS)

    Umano, Takuya; Yoshioka, Kenichi; Obara, Toru

    2015-01-01

    Highlights: • Calculation procedures are easier than those of the cross-section adjustment method. • In addition, few cases of experimental results, two or three cases for example, can be well managed with the method. • Different from the bias factor method, a representativity factor is simultaneously obtained to know the whole quality of utilized experiments. • After easier process of calculations, it is possible to obtain a correction value of a particular physical property with the method. • Our proposed method was considered to be applicable to the correction of the infinite neutron multiplication factor in LWR studies. - Abstract: To judge the applicability of a critical experiment, it is necessary to confirm the similarities of the experiment with actual reactor conditions or equipment. The concept of the “representativity factor” has been well adopted since the late 1970’s, particularly for fast breeder reactors (FBRs) and future reactor studies. In our previous study, we extended this concept to the design of a light water reactor (LWR) system, and derived mathematical formulas for a new numerical evaluation method to correct a physical property of a target system. This method is different from the cross-section adjustment method and the bias factor method. For the first qualification of the method, sample calculations were carried out to correct the effective neutron multiplication factor through critical experiments at the Toshiba Nuclear Critical Assembly (NCA) facility. We also compared the result with that of the Product of Exponentiated experimental values method (PE method) of the extended bias factor methods. A good agreement was observed. The purpose of this study was to demonstrate the applicability of the method to the infinite neutron multiplication factor. Using the method and three kinds of critical experiments of NCA, calculations were performed to correct the infinite neutron multiplication factor of a pressurized water

  4. Effectiveness of teaching automated external defibrillators use using a traditional classroom instruction versus self-instruction video in non-critical care nurses

    Directory of Open Access Journals (Sweden)

    Ismail M. Saiboon

    2016-04-01

    Full Text Available Objectives: To evaluate the effectiveness and retention of learning automated external defibrillator (AED usage taught through a traditional classroom instruction (TCI method versus a novel self instructed video (SIV technique in non-critical care nurses (NCCN. Methods: A prospective single-blind randomized study was conducted over 7 months (April-October 2014 at the Universiti Kebangsaan Malaysia Medical Center, Kuala Lampur, Malaysia. Eighty nurses were randomized into either TCI or SIV instructional techniques. We assessed knowledge, skill and confidence level at baseline, immediate and 6-months post-intervention. Knowledge and confidence were assessed via questionnaire; skill was assessed by a calibrated and blinded independent assessor using an objective structured clinical examination (OSCE method. Results: Pre-test mean scores for knowledge in the TCI group was 10.87 ± 2.34, and for the SIV group was 10.37 ± 1.85 (maximum achievable score 20.00; 4.05 ± 2.87 in the TCI and 3.71 ± 2.66 in the SIV (maximum score 11.00 in the OSCE evaluation and 9.54 ± 3.65 in the TCI and 8.56 ± 3.47 in the SIV (maximum score 25.00 in the individual’s personal confidence level. Both methods increased the mean scores significantly during immediate post-intervention (0-month. At 6-months, the TCI group scored lower than the SIV group in all aspects 11.13 ± 2.70 versus 12.95 ± 2.26 (p=0.03 in knowledge, 7.27 ± 1.62 versus 7.68 ± 1.73 (p=0.47 in the OSCE, and 16.40 ± 2.72 versus 18.82 ± 3.40 (p=0.03 in confidence level. Conclusion: In NCCN’s, SIV is as good as TCI in providing the knowledge, competency, and confidence in performing AED defibrillation.

  5. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

    Energy Technology Data Exchange (ETDEWEB)

    Aggery, A

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  6. On the method of heat exchange calculation for critical and postcritical regimes in sodium-water steam generators

    International Nuclear Information System (INIS)

    Khudasko, V.V.; Kardash, D.Yu.; Grachev, N.S.

    1986-01-01

    Technique for calculating heat exchange in sodium-water steam generators with provisions for steam-water flow non-equilibrium character and moisture additional evaporation in pipes is suggested. Zone of heat exchange crisis representing the zone of transition from developed boiling to postcritical zone is considered. Comparison of estimated and experimental data performed for the following ranges of steame generator parameters: pressure p=7.8-14.0 MPa, coolant flow rate ρw=350-1000 kg/(m 2 xs), inlet sodium temperature T=590-825 K shows their good agreement

  7. Comparison of MCNP and WIMS-AECL/RFSP calculations against critical heavy water experiments in ZED-2 with CANFLEX-LVRF and CANFLEX-LEU fuels

    International Nuclear Information System (INIS)

    Bromley, B. P.; Watts, D. G.; Pencer, J.; Zeller, M.; Dweiri, Y.

    2009-01-01

    This paper summarizes calculations of MCNP5 and WIMS-AECL/RFSP compared against measurements in coolant void substitution experiments in the ZED-2 critical facility with CANFLEX R-LEU/RU (Low Enriched Uranium, Recovered Uranium) reference fuels and CANFLEX-LVRF (Low Void Reactivity Fuel) test fuel, and H 2 O/air coolants. Both codes are tested for the prediction of the change in reactivity with complete voiding of all fuel channels, and that for a checkerboard voiding pattern. Understanding these phenomena is important for the ACR-1000 R reactor. Comparisons are also made for the prediction of the axial and radial neutron flux distributions, as measured by copper foil activation. The experimental data for these comparisons were obtained from critical mixed lattice / substitution experiments in AECL's ZED-2 critical facility using CANFLEX-LEU/RU and CANFLEX-LVRF fuel in a 24-cm square lattice pitch at 25 degrees C. Substitution analyses were performed to isolate the properties (buckling, bare critical lattice dimensions) of the CANFLEX-LVRF fuel. This data was then used to further test the lattice physics codes. These comparisons establish biases/uncertainties and errors in the calculation of k eff , coolant void reactivity, checkerboard coolant void reactivity, and flux distributions. Results show small to modest biases in void reactivity and very good agreement for flux distributions. The importance of boundary conditions and the modeling of un-moderated fuel in the critical experiments are demonstrated. This comparison study provides data that supports code validation and gives good confidence in the reactor physics tools used in the design and safety analysis of the ACR-1000 reactor. (authors)

  8. Comparison of IBM-2 calculations with X(5) critical point symmetry for low lying states in 128-140Nd

    International Nuclear Information System (INIS)

    Uluer, I.; Olgun, D.; Inan, S.; Tuerkan, N.

    2006-01-01

    The X(5) would take place when moving continuously from the pure U(5) symmetry to the SU(3) symmetry and it implies a definite relations among the level energies and among the E2 transition strengths. It was recently shown that a signature of phase transition is observed in the chain of Sm, Mo and Nd isotopes, where 1 52Sm, 1 04Mo and 1 50Nd display the predicted features of the X(5) symmetry and mark therefore the critical point. However, more detailed studies and experiments are needed to get ideas about this signature. Without entering into detail we have firstly compared the results obtained in our previous study of 1 28- 1 40Nd with that of the limits in X(5) symmetry and then given a clear description about the validity of the Hamiltonian parameters used in the study. At the end, we have concluded that some of Nd isotopes display X(5) symmetry features

  9. KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

    International Nuclear Information System (INIS)

    Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

    2001-01-01

    The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

  10. Morphine-augmented cholescintigraphy and acute a calculous cholecystitis in critically ill patients: interest and new way of interpreting

    International Nuclear Information System (INIS)

    Emptaz, A.; Prevot, N.; Dubois, F.; Mahul, P.; Mariat, G.; Jospe, R.; Auboyer, C.; Cuilleron, M.

    2005-01-01

    Introduction: Acute acalculous cholecystitis (AAC) is a serious disease, difficult to diagnose in critically ill patients. The aim of the study was to evaluate the diagnostic performances of abdominal ultrasonography (US) and morphine-augmented cholescintigraphy (MC) and to improve diagnostic strategy in patients of intensive care unit (ICU) with suspected AA C. Methods: We retrospectively studied 82 consecutive ICU patients with suspected AA C. US was positive if the triad of gallbladder distension, gallbladder wall thickening and sludge was found. MC was positive if the gallbladder remained non-visualized after morphine injection. In a second time, other scintigraphic criteria of interpretation were tested, according to the visualization of the gallbladder before or after morphine administration. Treatment was decided on the basis of clinical, laboratory and imaging data. Results: The diagnosis of AAC was retained in 34 patients. US and MC had respectively for the diagnosis of AAC a sensitivity of 20.6 and 70.6%, and a specificity of 95.8 and 100%. Interpreting the MC as positive if the gallbladder remains non-visualized after morphine, as negative if it appears before, and as non-conclusive if visualized after, makes it possible to define respectively patients with high probability (100%), with low probability (7.5%) or with intermediate probability (39%) of AAC. Conclusions: MC is better than US for diagnosing AAC in critically ill patients, having in particular excellent specificity using the classical criteria of interpretation. MC must be thus performed in patients at risk for AAC, determined with clinical, laboratory and eventually echographic findings. To decrease false negative rate of MC, a probability categorical classification is proposed to improve patients' care. (author)

  11. Proposal on the accelerator driven molten-salt reactor (ATW-concept) benchmark calculation (stage-1 without an external neutron sources)

    International Nuclear Information System (INIS)

    Svarny, J.; Mikolas, P.

    1999-01-01

    The simple model of two component concept of the ATW (graphite + molten salt system) was found. The main purpose of this benchmark will be not only to provide the basic characteristics of given ADS but also to test codes in calculations of the rate of transmutation waste and to evaluate basic kinetics parameters and reactivity effects. (Authors)

  12. Externalities - an analysis using the EU ExternE-results

    International Nuclear Information System (INIS)

    2003-10-01

    The EU project ExternE quantified the externalities for the different energy technologies. In this work, the ExternE results are used in a MARKAL-analysis for the Nordic countries. The analysis does not go into detail, but gives some interesting indications: The external costs are not fully covered in the Nordic energy systems, the present taxes and charges are not high enough. The emissions from the energy systems would be strongly reduced, if taxes/environmental charges were set at the level ExternE calculate. The emissions from power production would be reduced most. Renewable energy sources and natural gas dominate the energy systems in the ExternE case

  13. Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2001-01-11

    Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

  14. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations; Alize 3 - premiere experience critique pour le reacteur a haut flux franco-allemand. Calculs

    Energy Technology Data Exchange (ETDEWEB)

    Scharmer, K [Commissariat a l' Energie Atomique, Dir. des Piles Atomiques, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The results of experiments in the light water cooled D{sub 2}O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k{sub eff} was smaller than 0.5 per cent {delta}k/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D{sub 2}O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [French] Les resultats des experiences faites dans la maquette critique ALIZE III, refrigeree a l'eau legere et reflechie par l'eau lourde, ont ete compares aux calculs. On a utilise un modele de la theorie de diffusion a trois groupes rapides et epithermiques et deux groupes thermiques qui se recouvrent. Ce modele a permis de calculer la distribution de puissance dans le coeur en bon accord avec les mesures, meme dans le cas d'une forte variation du spectre des neutrons dans le coeur. L'erreur entre k{sub eff} calcule et mesure etait inferieure a 0,5 pour cent {delta}k/k. Le coefficient de vide et des materiaux de structure, la reactivite des barres 'noires', les variations du spectre (rapport Cd, rapport Pu/U) et la fraction des photo-neutrons retardes sont egalement calcules. Les mesures de reactivite et de perturbation de flux dans le reflecteur, dues aux canaux, ont ete interpretees du point de vue d'un arrangement optimum des canaux pour le Reacteur a Haut Flux Franco-Allemand. (auteur)

  15. MCNP calculation of the critical H_3BO_3 concentrations for the first fuel loading into the reactor core of NPP MO-3-4 units

    International Nuclear Information System (INIS)

    Vrban, B.; Lueley, J.; Farkas, G.; Hascik, J.; Hinca, R.; Petriska, M.; Slugen, V.

    2012-01-01

    The purpose of the analysis was the determination of critical H_3BO_3 concentrations for the first fuel loading into the reactor core of MO34 units using 2"n"d generation fuel during the first start-up of new unit using calculation code MCNP 1.60. H_3BO_3 concentrations were computed for the given temperature of the primary circuit and position of the 6"t"h safety control rod group. Because of the very first start-up of these units, detailed analyses of active-core parameters are required by National Regulatory Authority and needed for safe operation of nuclear facility. (authors)

  16. Alecto 1 - criticality experiment on a solution of plutonium and of uranium 235. Experimental results and calculations on tank number 2 ({phi} 300 mm); Alecto 1 - experience de criticite sur une solution de plutonium et d'uranium enrichi a 90 pour cent. Resultats experimentaux et calculs concernant la cuve no. 2 ({phi} = 300 mm)

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, J G; Brunet, J P; Clouet D' Orval, Ch; Kremser, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Experiments on plutonium and 90 pour cent enriched uranium solutions have been made in the Alecto reactor with a tank of external diameter 300 mm. Various geometries Lave been tested, for variable concentrations of fissionable salts. The critical mass was studied as a function of the concentration in various reflector conditions (water, concrete, wood) and the experimental values were compared with calculated values. The effects of cadmium as a reflector and of the stainless steel tank were also studied. Lastly were carried out measurements of {beta}/{tau}, ratio of the effective fraction of delayed neutrons to the average lifetime of the neutrons in the reactor. (authors) [French] Des experiences sur des solutions de plutonium et d'uranium enrichi a 90 pour cent ont ete effectuees dans le reacteur Alecto, avec une cuve de diametre exterieur 300 mm. Diverses configurations geometriques ont ete realisees, pour des concentrations variables du sel fissile. On a etudie la masse critique en fonction de la concentration, dans plusieurs conditions de reflexion (eau, beton, bois), et on a compare les resultats experimentaux aux valeurs donnees par le calcul. On a egalement etudie l'influence du cadmium comme reflecteur et celle de la cuve d'acier inoxydable. Enfin on a effectue des mesures de {beta}/{tau}, rapport de la proportion effective des neutrons retardes au temps de vie moyen des neutrons dans la pile. (auteurs)

  17. Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

    International Nuclear Information System (INIS)

    Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

    2000-01-01

    Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

  18. Metrological and treatment planning improvements on external beam radiotherapy. Detector size effect and dose calculation in low-density media (in Spanish)

    International Nuclear Information System (INIS)

    Garcia-Vicente, Feliciano

    2004-01-01

    The objective of this thesis is the improvement of the measurement and calculation accuracy for radiation therapy fields. Basically, it deals with two questions: the detector size effect and the heterogeneity dose calculation. The author analyzes both the metrological and computational effects and its clinical implications by simulation of the radiotherapy treatments in a treatment planning system. The detector size effect leads up to smoothing of the radiation profile increasing the penumbra (20%-80%) and beam fringe (50%-90%) values with the consequent clinical effect of over-irradiation of the organs at risk close to the planning target volume (PTV). In this thesis this problem is analyzed finding mathematical solutions based on profile deconvolution or the use of radiation detectors of adequate size. On the other side, the author analyzes the dose computation on heterogeneous media by the superposition algorithms versus classical algorithms. The derived conclusion from this thesis is that in locations like lung and breast, the classical algorithms lead to a significant underdosage of the PTV with an important decrease of tumor control probability (TCP). On this basis, the author does not recommend the clinical use of these algorithms in the mentioned tumor locations

  19. Coupled calculation of external heat transfer and material temperatures of convection-cooled turbine blades. Final report; Gekoppelte Berechnung des aeusseren Waermeuebergangs und der Materialtemperaturen konvektionsgekuehlter Turbinenschaufeln. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Heselhaus, A.

    1997-05-01

    In this work a hybrid program system consisting of a 3D finite-volume Navier-Stokes flow solver and a 3D finite-element heat conduction solver has been developed. It enables the coupled calculation of structure temperatures in diabatic solid/fluid configurations. The grids of both the finite element and the finite volume computational domain may be completely independent. The coupled program fully resolves the thermal interaction between heat transfer and the resulting material temperatures. The developed coupling algorithm is numerically stable, conservative and works without the need to define ambient temperatures in the flowfield. This allows for the simulation of any solid/fluid configuration. When simulating combined blade/endwall cooling or filmcooling, only a coupled procedure is capable to completely account for the interaction between all relevant thermal parameters. It is found that the coupled calculation of convective cooling in a realistic guide vane leads locally to 45 K higher and 107 K lower blade temperatures than the uncoupled calculation. This shows that accounting for the thermal interaction between the flow and the structure offers both potential to save cooling air and a lower margin of safety when designing cooling systems close to the thermal limits of the blade material. (orig.) [Deutsch] Im Rahmen der vorliegenden Arbeit wurde ein Verfahren zur Berechnung der Temperaturverteilung in diabat umstroemten Koerpern entwickelt, bei dem ein 3D-Finite Volumen Navier-Stokes Stroemungsloeser und ein 3D-Finite Elemente Waermeleitungsloeser zu einem hybriden Programmsystem gekoppelt werden. Dabei besteht die Moeglichkeit, voellig unabhaengige Rechennetze fuer Stroemung und Struktur zu verwenden. Mit dem gekoppelten Verfahren kann die Wechselwirkung zwischen resultierenden Materialtemperaturen und dem davon rueck-beeinflussten Waermeuebergang beruecksichtigt werden. Weiterhin ist der hier entwickelte, stabile und konservative Kopplungsalgorithmus nicht

  20. Monte Carlo calculations of the response of an external detector to a source of photons in the lungs of a heterogeneous phantom

    International Nuclear Information System (INIS)

    Bhati, S.; Sharma, R.C.; Somasundaram, S.

    1979-01-01

    A computer program to calculate the response of a 20 cm dia phoswich (3mm thick NaI(Tl) primary detector) to a source of low-energy photons distributed in the lungs of a heterogeneous (MIRD) phantom, approximating ICRP Reference Man, has been developed. Monte Carlo techniques are employed to generate photons and trace their fates in the thorax of MIRD phantom. The acceptable points of photon interactions in skeletal, lung and ordinary tissue are determined by Coleman technique. The photon interactions considered are photoelectric and Compton. The calculations yield the exit photon energy spectrum which is smeared with experimentally determined Gaussian resolution function to convert into pulse-height spectrum observable with the detector. The computer program has provisions for incorporating the effects of iodine K x-ray escape as well as variable intrinsic efficiency of the detector. Computed calibration factors (cpm/μCi integrated over the full spectrum) are given for the phoswich located centrally over and in contact with the chest for several low-energy photon sources distributed uniformly or as points in the lungs of the phantom. The radionuclides considered are 238 Pu, 239 Pu, 241 Am, 244 Cm, 246 Cm, 250 Cf and 103 Pd. Examples of generated exit photon and the corresponding pulse-height spectra are included. The spectral changes observed in these generated spectra, which are also discerned in experimental pulse-height spectra, are discussed in detail. Thus, photopeak energies of 18.4 and 55.5 KeV for Usub(L) x-rays and 241 Am gamma-rays respectively have been observed. It is shown that consideration of the total (i.e. both uncollided and those escaping after collision instead of the uncollided alone) flux of escaping photons improves the calibration factors by about 50% for 239 Pu, 70% for 103 Pd and as much as 340% for 241 Am gamma-rays. In addition, calibration factors are calculated for point 239 Pu sources located at different sites in the phantom lungs

  1. Criticality calculations on pebble-bed HTR-PROTEUS configuration as a validation for the pseudo-scattering tracking method implemented in the MORET 5 Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Forestier, Benoit; Miss, Joachim; Bernard, Franck; Dorval, Aurelien [Institut de Radioprotection et Surete Nucleaire, Fontenay aux Roses (France); Jacquet, Olivier [Independent consultant (France); Verboomen, Bernard [Belgian Nuclear Research Center - SCK-CEN (Belgium)

    2008-07-01

    The MORET code is a three dimensional Monte Carlo criticality code. It is designed to calculate the effective multiplication factor (k{sub eff}) of any geometrical configuration as well as the reaction rates in the various volumes and the neutron leakage out of the system. A recent development for the MORET code consists of the implementation of an alternate neutron tracking method, known as the pseudo-scattering tracking method. This method has been successfully implemented in the MORET code and its performances have been tested by mean of an extensive parametric study on very simple geometrical configurations. In this context, the goal of the present work is to validate the pseudo-scattering method against realistic configurations. In this perspective, pebble-bed cores are particularly well-adapted cases to model, as they exhibit large amount of volumes stochastically arranged on two different levels (the pebbles in the core and the TRISO particles inside each pebble). This paper will introduce the techniques and methods used to model pebble-bed cores in a realistic way. The results of the criticality calculations, as well as the pseudo-scattering tracking method performance in terms of computation time, will also be presented. (authors)

  2. Calculation for shielding based on the new law in the nuclear medicine facilities. Calculation methods of effective dose concerning the external and internal exposures and of radioisotope concentration concerning the exhaust gas drainage

    International Nuclear Information System (INIS)

    Ohba, Hisateru; Takeda, Hiromitsu; Asanuma, Osamu

    2001-01-01

    Following the revision of the law which incorporated the ICRP 1990 Recommendation, the medical law enforcement rule and related notices are also revised and enforced from April 1, 2001. Revised points related with the nuclear medicine facilities involve the reported items (addition of the scheduled maximum amount to be used in the next 3 months), change of dose limits at the boundary of the controlled area (from 300 μSv/w to 1.3 mSv/3 m), change of density limits in air, exhausted air and drainage, change of evaluation of radioisotope density in air (from average density during 8 hr to 1 week), change of exposure dose limits in medical workers and change of calculation method of effective dose due to internal exposure. This paper concerns the calculation methods for above and their concepts in nuclear medicine facilities in Hokkaido area. Numerical data for shielding and conditions of the facilities for clinical practice including diagnostic nuclide are taken into consideration and the actual paper forms for these items are also shown. (K.H.)

  3. Evaluation of the external radiation exposure dosimetry and calculation of maximum permissible concentration values for airborne materials containing 18F, 15O, 13N, 11C and 133Xe

    International Nuclear Information System (INIS)

    Piltingsrud, H.V.; Gels, G.L.

    1985-01-01

    To better understand the dose equivalent (D.E.) rates produced by airborne releases of gaseous positron-emitting radionuclides under various conditions of cloud size, a study of the external radiation exposure dosimetry of these radionuclides, as well as negatron, gamma and x-ray emitting 133Xe, was undertaken. This included a calculation of the contributions to D.E. as a function of cloud radii, at tissue depths of 0.07 mm (skin), 3 mm (lens of eye) and 10 mm (whole body) from both the particulate and photon radiations emitted by these radionuclides. Estimates of maximum permissible concentration (MPC) values were also calculated based on the calculated D.E. rates and current regulations for personnel radiation protection (CFR84). Three continuous air monitors, designed for use with 133Xe, were evaluated for applications in monitoring air concentrations of the selected positron emitters. The results indicate that for a given radionuclide and for a cloud greater than a certain radius, personnel radiation dosimeters must respond acceptably to only the photon radiations emitted by the radionuclide to provide acceptable personnel dosimetry. For clouds under that radius, personnel radiation dosimeters must also respond acceptably to the positron or negatron radiations to provide acceptable personnel dosimetry. It was found that two out of the three air concentration monitors may be useful for monitoring air concentrations of the selected positron emitters

  4. Critical mass flux for flaming ignition of dead, dry wood as a function of external radiant heat flux and oxidizer flow velocity

    Science.gov (United States)

    Sara McAllister; Mark Finney; Jack Cohen

    2010-01-01

    Extreme weather often contributes to crown fires, where the fire spreads from one tree crown to the next as a series of piloted ignitions. An important aspect in predicting crown fires is understanding the ignition of fuel particles. The ignition criterion considered in this work is the critical mass flux criterion – that a sufficient amount of pyrolysis gases must be...

  5. External Validation and Recalibration of Risk Prediction Models for Acute Traumatic Brain Injury among Critically Ill Adult Patients in the United Kingdom.

    Science.gov (United States)

    Harrison, David A; Griggs, Kathryn A; Prabhu, Gita; Gomes, Manuel; Lecky, Fiona E; Hutchinson, Peter J A; Menon, David K; Rowan, Kathryn M

    2015-10-01

    This study validates risk prediction models for acute traumatic brain injury (TBI) in critical care units in the United Kingdom and recalibrates the models to this population. The Risk Adjustment In Neurocritical care (RAIN) Study was a prospective, observational cohort study in 67 adult critical care units. Adult patients admitted to critical care following acute TBI with a last pre-sedation Glasgow Coma Scale score of less than 15 were recruited. The primary outcomes were mortality and unfavorable outcome (death or severe disability, assessed using the Extended Glasgow Outcome Scale) at six months following TBI. Of 3626 critical care unit admissions, 2975 were analyzed. Following imputation of missing outcomes, mortality at six months was 25.7% and unfavorable outcome 57.4%. Ten risk prediction models were validated from Hukkelhoven and colleagues, the Medical Research Council (MRC) Corticosteroid Randomisation After Significant Head Injury (CRASH) Trial Collaborators, and the International Mission for Prognosis and Analysis of Clinical Trials in TBI (IMPACT) group. The model with the best discrimination was the IMPACT "Lab" model (C index, 0.779 for mortality and 0.713 for unfavorable outcome). This model was well calibrated for mortality at six months but substantially under-predicted the risk of unfavorable outcome. Recalibration of the models resulted in small improvements in discrimination and excellent calibration for all models. The risk prediction models demonstrated sufficient statistical performance to support their use in research and audit but fell below the level required to guide individual patient decision-making. The published models for unfavorable outcome at six months had poor calibration in the UK critical care setting and the models recalibrated to this setting should be used in future research.

  6. Development of a method for high temperature reactor calculations tested at the critical facility Kahter using the program system RSYST. Entwicklung einer Rechenmethode zur HTR-Auslegung im Rahmen des Programmsystems RSYST und deren Erprobung an der kritischen Anlage 'Kahter'

    Energy Technology Data Exchange (ETDEWEB)

    Nabi, R

    1979-08-15

    In this report the neutron- and reactor physical aspects of the high temperature pebble bed reactor are studied. For this purpose appropriate HTR-nuclear data sets are generated and applied in a calculation model, which is developed on the basis of neutron transport and diffusion theory. This model includes the complete reactor calculation for determination of neutron flux, reactivity and reaction rates. This reactor calculation is based on following: evaluation of resonance absorption in double heterogeneity, cell calculation in spherical geometry, zone spectral calculation and subsequent 2-dimensional diffusion calculation. All calculations are performed in the modular program system RSYST, which accommodates simplified treatment of reactor physics problems through its data transfer and treatment techniques and through its calculations control features. In this report the neutron- and reactor physical aspects of the high temperature pebble bed reactor are studied. For this purpose appropriate HTR-nuclear data sets are generated and applied in a calculation model, which is developed on the basis of neutron transport and diffusion theory. This model includes the complete reactor calculation for determination of neutron flux, reactivity and reaction rates. This reactor calculation is based on following: evaluation of resonance absorption in double heterogeneity, cell calculation in spherical geometry, zone spectral calculation and subsequent 2-dimensional diffusion calculation. All calculations are performed in the modular program system RSYST, which accommodates simplified treatment of reactor physics problems through its data transfer and treatment techniques and through its calculations control features. The results of the calculations are compared with measured values of different core configurations of the critical facility for the high temperature pebble bed reactor (KAHTER). This comparison shows how a critical facility is used to verify and to adjust

  7. Validation of Neutron Calculation Codes and Models by means of benchmark cases in the frame of the Binational Commission of Nuclear Energy. Criticality Experiments

    International Nuclear Information System (INIS)

    Dos Santos, Adimir; Siqueira, Paulo de Tarso D.; Andrade e Silva, Graciete Simões; Grant, Carlos; Tarazaga, Ariel E.; Barberis, Claudia

    2013-01-01

    In year 2008 the Atomic Energy National Commission (CNEA) of Argentina, and the Brazilian Institute of Energetic and Nuclear Research (IPEN), under the frame of Nuclear Energy Argentine Brazilian Agreement (COBEN), among many others, included the project “Validation and Verification of Calculation Methods used for Research and Experimental Reactors . At this time, it was established that the validation was to be performed with models implemented in the deterministic codes HUEMUL and PUMA (cell and reactor codes) developed by CNEA and those ones implemented in MCNP by CNEA and IPEN. The necessary data for these validations would correspond to theoretical-experimental reference cases in the research reactor IPEN/MB-01 located in São Paulo, Brazil. The staff of the group Reactor and Nuclear Power Studies (SERC) of CNEA, from the argentine side, performed calculations with deterministic models (HUEMUL-PUMA) and probabilistic methods (MCNP) modeling a great number of physical situations of de reactor, which previously have been studied and modeled by members of the Center of Nuclear Engineering of the IPEN, whose results were extensively provided to CNEA. In this paper results for critical configurations are shown. (author)

  8. Critical evaluation of the external occupational exposure in nuclear medicine services in Brazil; Avaliacao critica da exposicao ocupacional externa nos servicos de medicina nuclear no Brasil

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, Ana Luiza Silva Lima

    2016-07-01

    Currently in Brazil (2016), there are 421 Nuclear Medicine Services (NMS). In nuclear medicine, the possibility of occupational internal contamination and external exposure is unavoidable. The chest individual monitoring, to estimate the effective dose, is mandatory, but the extremity monitoring is not always made. The aim of this study was to conduct a survey of data for external exposure of NMS professionals in Brazil from 1987 to 2010, analysing them in terms of trends and comparing them with measurements carried out in this work and in other countries. Although most of the NMS is still located in large urban centres (54% in the Southeast region), there is no state without any NMS. The increasing number of NMS has generated the need for more professionals. In the year 1987, they were 755 workers and, in 2010, 4134, with the following distribution of specialties: 29% of Nuclear Medicine Technicians (NMT), 23% of Nursing professionals, 29% of Physicians and 3% of Physicists. The average annual effective dose reached more than 3.0 mSv in some regions of the country, from 1987 to 2010, but tends to 1.0 mSv in 2010. The highest doses, as expected, are received by NMT and Nursing. The professionals who handle radiopharmaceuticals have their hands much more exposed than the chest. During 2010, only 31% of NMT and 16% of Nursing used extremity dosimeters as compared to chest dosimeters. The data from the measurements indicate that not all individual dosimeters are used properly. Generally, both in the measurements as in national registries, the hand doses were higher for professionals who prepared the radiopharmaceutical (NMT) than those who injected (Nursing). The value measured by chest dosimeters can be used to estimate the equivalent dose to the eye lenses, except for NMT at preparation practices at conventional NMS, where the equivalent dose of the lens is about 2 times higher than the dose at the chest. The most exposed areas of the hands are the tips of the index

  9. Retrofitting Systems for External Walls

    DEFF Research Database (Denmark)

    Rose, Jørgen

    1997-01-01

    In this report, 9 different external and internal retrofitting systems are analyzed using numerical calculations. The analysis focuses on the thermal bridge effects in the different systems, and on this basis it is discussed whether internal or external retrofitting has the most advantages...

  10. On Ising - Onsager problem in external magnetic field

    International Nuclear Information System (INIS)

    Kochmanski, M.S.

    1997-01-01

    In this paper a new approach to solving the Ising - Onsager problem in external magnetic field is investigated. The expression for free energy on one Ising spin in external field both for the two dimensional and three dimensional Ising model with interaction of the nearest neighbors are derived. The representations of free energy being expressed by multidimensional integrals of Gauss type with the appropriate dimensionality are shown. Possibility of calculating the integrals and the critical indices on the base of the derived representations for free energy is investigated

  11. Geochemistry Model Validation Report: External Accumulation Model

    International Nuclear Information System (INIS)

    Zarrabi, K.

    2001-01-01

    The purpose of this Analysis and Modeling Report (AMR) is to validate the External Accumulation Model that predicts accumulation of fissile materials in fractures and lithophysae in the rock beneath a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. (Lithophysae are voids in the rock having concentric shells of finely crystalline alkali feldspar, quartz, and other materials that were formed due to entrapped gas that later escaped, DOE 1998, p. A-25.) The intended use of this model is to estimate the quantities of external accumulation of fissile material for use in external criticality risk assessments for different types of degrading WPs: U.S. Department of Energy (DOE) Spent Nuclear Fuel (SNF) codisposed with High Level Waste (HLW) glass, commercial SNF, and Immobilized Plutonium Ceramic (Pu-ceramic) codisposed with HLW glass. The scope of the model validation is to (1) describe the model and the parameters used to develop the model, (2) provide rationale for selection of the parameters by comparisons with measured values, and (3) demonstrate that the parameters chosen are the most conservative selection for external criticality risk calculations. To demonstrate the applicability of the model, a Pu-ceramic WP is used as an example. The model begins with a source term from separately documented EQ6 calculations; where the source term is defined as the composition versus time of the water flowing out of a breached waste package (WP). Next, PHREEQC, is used to simulate the transport and interaction of the source term with the resident water and fractured tuff below the repository. In these simulations the primary mechanism for accumulation is mixing of the high pH, actinide-laden source term with resident water; thus lowering the pH values sufficiently for fissile minerals to become insoluble and precipitate. In the final section of the model, the outputs from PHREEQC, are processed to produce mass of accumulation

  12. ExternE: Externalities of energy Vol. 4. Oil and gas

    International Nuclear Information System (INIS)

    Friedrich, R.; Krewitt, W.; Mayerhofer, P.

    1995-01-01

    Awareness of the environmental damage resulting from human activity, particularly commencing energy use, has grown greatly in recent years. Effects such as global warming, ozone depletion and acid rain are now the subjects of much research and public debate. It is now known that these and other effects damage a wide range of receptors, including human health, forests, crops, freshwater ecosystems and buildings. Such damages are typically not accounted for by the producers and consumers of the good in question (in this case energy). They are thus referred to as 'external costs' or 'externalities', to distinguish them from the private costs which account for the construction of plant, cost of fuel, wages, etc. In recent years there has been a growing interest in the assessment of the environmental and health impacts of energy, and the related external costs. This concern is driven by a number of different factors: the need to integrate environmental concerns in decision making over the choice between different fuels and energy technologies; the need to evaluate the costs and benefits of stricter environmental standards; increased attention to the use of economic instruments for environmental policy, the need to develop overall indicators of environmental performance of different technologies; major changes in the energy sector, including privatisation, liberalisation of markets, reduction of subsidies, etc. An agreed methodology for calculation and integration of external costs has not been established. Earlier work is typically of a preliminary nature and tends to be deficient with respect to both the methods employed and the quality of models and data used. In consequence of this a collaborative project, the EC/US Fuel Cycles Study, was established between Directorate General XLI (Science, Research and Technology) of the European Commission and the United States Department of Energy. This ran for the period 1991 to 1993, and good agreement on a variety of

  13. Environmental externalities related to power production on biogas and natural gas based on the EU ExternE methodology

    DEFF Research Database (Denmark)

    Nielsen, Per Sieverts; Ibsen, Liselotte Schleisner

    1998-01-01

    This paper assesses the environmental impacts and external costs from selected electricity generation systems in Denmark. The assessment is carried out as part of the ExternE National Implementation, which is the second phase of the ExternE project and involves case studies from all Western...... European countries. The project use a “bottom-up” methodology to evaluate the external costs associated with a wide range of different fuel cycles. The project has identified priority impacts, where most are impacts from air emissions. Externalities due to atmospheric emissions are calculated through...

  14. Burnout calculation

    International Nuclear Information System (INIS)

    Li, D.

    1980-01-01

    Reviewed is the effect of heat flux of different system parameters on critical density in order to give an initial view on the value of several parameters. A thorough analysis of different equations is carried out to calculate burnout is steam-water flows in uniformly heated tubes, annular, and rectangular channels and rod bundles. Effect of heat flux density distribution and flux twisting on burnout and storage determination according to burnout are commended [ru

  15. Accumulation of nitrogen - a critical parameter for the calculation of load limits from nitrogen in forests; Akkumulering av nitrogen - en kritisk parameter for beregning av taalegrenser for nitrogen i skog

    Energy Technology Data Exchange (ETDEWEB)

    Sogn, T A; Stuanes, A O; Abrahamsen, G

    1996-01-01

    The conference paper deals with the accumulation of nitrogen in forests in Norway. The level of accumulation is a critical factor for the calculation of load limits. The paper compares the average rapidity values of accumulation since the last glacial age with the calculated values from the more short-lasting period based on data from surveying programs of the State Pollution Control Authority, manuring experiments, and other relevant research programs in this field. 8 refs., 1 fig., 1 tab.

  16. Commissioning of radiotherapy treatment planning systems: Testing for typical external beam treatment techniques. Report of the Coordinated Research Project (CRP) on Development of Procedures for Quality Assurance of Dosimetry Calculations in Radiotherapy

    International Nuclear Information System (INIS)

    2008-01-01

    users at the hospitals. Commissioning is one of the most important parts of the entire QA programme for both the RTPS and the planning process. Commissioning involves testing of system functions, documentation of the different capabilities and verification of the ability of the dose calculation algorithms to reproduce measured dose calculations. The current report is limited to treatment simulation tests for external high-energy photon beams that are performed prior to clinical use of RTPS. The report deals with the verification of the dose calculations through commissioning tests that cover typical treatment techniques only. This report also summarizes the results of a pilot study of the clinical commissioning recommendations that was performed by the participants of the Coordinated Research Project at their home institutions. The summary of the pilot study is available to medical physicists as an example of the implementation of the clinical commissioning procedures for RTPSs at their hospitals. Issues related to intensity modulated radiation therapy (IMRT) or other specialized techniques such as stereotactic radiosurgery are not addressed in this clinical commissioning report. While recognizing the specific scope of this report, this publication is useful to the purchasers of RTPSs in any country although they may have to perform tests beyond those described in this report to meet the needs of specialized techniques that have not been addressed here

  17. Restoration of symmetry by temperature effect under influence of external electro magnetic field in gauge theory

    International Nuclear Information System (INIS)

    Aquino, V.M. de.

    1987-01-01

    We have analysed, within a semi classical approach, the influence of external electromagnetic field on phase transitions in gauge theories. The critical temperature was calculated for an Abelian case, scalar electrodynamics, and for an non Abelian case, the Weinberg Salam model. (author)

  18. The ExternE project: methodology, objectives and limitations

    International Nuclear Information System (INIS)

    Rabl, A.; Spadaro, J.V.

    2002-01-01

    This paper presents a summary of recent studies on external costs of energy systems, in particular the ExternE (External Costs of Energy) Project of the European Commission. To evaluate the impact and damage cost of a pollutant, one needs to carry out an impact pathway analysis; this involves the calculation of increased pollutant concentrations in all affected regions due to an incremental emission (e.g. μg/m 3 of particles, using models of atmospheric dispersion and chemistry), followed by the calculation of physical impacts (e.g. number of cases of asthma due to these particles, using a dose-response function). The entire so-called fuel chain (or fuel cycle) is evaluated and compared on the basis of delivered end use energy. Even though the uncertainties are large, the results provide substantial evidence that the classical air pollutants (particles, NO x and SO x ) from the combustion of fossil fuels impose a heavy toll, in addition to the cost of global warming. The external costs are especially large for coal; even for 'good current technology' they may be comparable to the price of electricity. For natural gas the external costs are about a third to a half of coal. The external costs of nuclear are small compared to the price of electricity (at most a few %), and so are the external costs of most renewable energy systems. (authors)

  19. ExternE: Externalities of energy Vol. 1. Summary

    International Nuclear Information System (INIS)

    Holland, M.; Berry, J.

    1995-01-01

    There is a growing requirement for policy analysts to take account of the environment in their decision making and to undertake the specified cost-benefit analysis. Within the European Union this is reflected in the 5th Environmental Action Programme, and the Commission's White Paper entitled 'Growth, competitiveness, employment and the ways forward to the 21st century'. This has led to a need for evaluation of environmental externalities. The ExternE Project commenced in 1991 as the European part of a collaborative study between the European Commission and the US Department of Energy. It aims to be the first systematic approach to the evaluation of external costs of a wide range of different fuel cycles. The project will result in an operational accounting framework for the quantification and monetarisation of priority environmental and other externalities. This framework will allow the calculation of the marginal external costs and benefits for specific power plants, at specific sites using specified technologies. There are three major phases in the project. Phase 1 was undertaken in collaboration with the US Department of Energy. In this phase the teams jointly developed the conceptual approach and methodology and shared scientific information for application to a number of fuel cycles. On the European side work concentrated on the nuclear and coal fuel cycles which together were expected to raise many of the fundamental issues in fuel cycle analysis. The project is currently nearing completion of Phase 2. During this phase the methodology has been applied to a wide range of different fossil, nuclear and renewable fuel cycles for power generation and energy conservation options. Also a series of National Implementation Programmes is underway in which the methodology and accounting framework are being applied to reference sites throughout Europe. In addition the general methodology is being extended to address the evaluation of externalities associated with

  20. Externalities of fuel cycles 'ExternE' project. Summary report

    International Nuclear Information System (INIS)

    Holland, M.; Berry, J.

    1994-01-01

    There is a growing requirement for policy analysts to take account of the environment in their decision making and to undertake the specified cost-benefit analysis. Within the European Union this is reflected in the 5th Environmental Action Programme, and the Commission's White Paper entitled 'Growth, competitiveness, employment and the ways forward to the 21st century'. This has led to a need for evaluation of environmental externalities. The ExternE Project commenced in 1991 as the European part of a collaborative study between the European Commission and the US Department of Energy. It aims to be the first systematic approach to the evaluation of external costs of a wide range of different fuel cycles. The project will result in an operational accounting framework for the quantification and monetarisation of priority environmental and other externalities. This framework will allow the calculation of the marginal external costs and benefits for specific power plants, at specific sites using specified technologies. There are three major phases in the project. Phase I was undertaken in collaboration with the US Department of Energy. In this phase the teams jointly developed the conceptual approach and methodology and shared scientific information for application to a number of fuel cycles. On the European side work concentrated on the nuclear and coal fuel cycles which together were expected to raise many of the fundamental issues in fuel cycle analysis. The project is currently nearing completion of Phase 2. During this phase the methodology has been applied to a wide range of different fossil, nuclear and renewable fuel cycles for power generation and energy conservation options. Also a series of National Implementation Programmes are underway in which the methodology and accounting framework are being applied to reference sites throughout Europe. In addition the general methodology is being extended to address the evaluation of externalities associated with

  1. ExternE: Externalities of energy Vol. 3. Coal and lignite

    International Nuclear Information System (INIS)

    Berry, J.; Holland, M.; Lee, D.

    1995-01-01

    Awareness of the environmental damage resulting from human activity, particularly concerning energy use, has grown greatly in recent years. Effects such as global warming, ozone depletion and acid rain are now the subjects of much research and public debate. It is now known that these and other effects damage a wide range of receptors, including human health, forests, crops, freshwater ecosystems and buildings. Such damages are typically not accounted for by the producers and consumers of the good in question (in this case energy). They are thus referred to as 'external costs' or 'externalities', to distinguish them from the private costs which account for the construction of plant, cost of fuel, wages, etc. In recent years there has been a growing interest in the assessment of the environmental and health impacts of energy, and the related external costs. This concern is driven by a number of different factors: the need to integrate environmental concerns in decision making over the choice between different fuels and energy technologies; the need to evaluate the costs and benefits of stricter environmental standards; increased attention to the use of economic instruments for environmental policy; the need to develop overall indicators of environmental performance of different technologies; major changes in the energy sector, including privatisation, liberalisation of markets, reduction of subsidies, etc. An agreed methodology for calculation and integration of external costs has not been established. Earlier work is typically of a preliminary nature and tends to be deficient with respect to both the methods employed and the quality of models and data used. In consequence of this a collaborative project, the EC/US Fuel Cycles Study, was established between Directorate General XII (Science, Research and Technology) of the European Commission and the United States Department of Energy. This ran for the period 1991 to 1993, and good agreement on a variety of

  2. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  3. KENO V.a Primer: A Primer for Criticality Calculations with SCALE/KENO V.a Using CSPAN for Input

    International Nuclear Information System (INIS)

    Busch, R.D.

    2003-01-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) computer software system developed at Oak Ridge National Laboratory (ORNL) is widely used and accepted around the world for criticality safety analyses. The well-known KENO V.a three-dimensional Monte Carlo criticality computer code is the primary criticality safety analysis tool in SCALE. The KENO V.a primer is designed to help a new user understand and use the SCALE/KENO V.a Monte Carlo code for nuclear criticality safety analyses. It assumes that the user has a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with SCALE/KENO V.a in particular. The primer is designed to teach by example, with each example illustrating two or three features of SCALE/KENO V.a that are useful in criticality analyses. The primer is based on SCALE 4.4a, which includes the Criticality Safety Processor for Analysis (CSPAN) input processor for Windows personal computers (PCs). A second edition of the primer, which uses the new KENO Visual Editor, is currently under development at ORNL and is planned for publication in late 2003. Each example in this first edition of the primer uses CSPAN to provide the framework for data input. Starting with a Quickstart section, the primer gives an overview of the basic requirements for SCALE/KENO V.a input and allows the user to quickly run a simple criticality problem with SCALE/KENO V.a. The sections that follow Quickstart include a list of basic objectives at the beginning that identifies the goal of the section and the individual SCALE/KENO V.a features which are covered in detail in the example problems in that section. Upon completion of the primer, a new user should be comfortable using CSPAN to set up criticality problems in SCALE/KENO V.a

  4. External costs of electricity

    International Nuclear Information System (INIS)

    Rabl, A.; Spadaro, J.V.

    2005-01-01

    This article presents a synthesis of the ExternE project (External costs of Energy) of the European community about the external costs of power generation. Pollution impacts are calculated using an 'impact pathways' analysis, i.e. an analysis of the emission - dispersion - dose-response function - cost evaluation chain. Results are presented for different fuel cycles (with several technological variants) with their confidence intervals. The environmental impact costs are particularly high for coal: for instance, in France, for coal-fired power plants it is of the same order as the electricity retail price. For natural gas, this cost is about a third of the one for coal. On the contrary, the environmental impact costs for nuclear and renewable energies are low, typically of few per cent of the electricity price. The main part of these costs corresponds to the sanitary impacts, in particular the untimely mortality. In order to avoid any controversy about the cost evaluation of mortality, the reduction of the expectation of life due to the different fuel cycles is also indicated and the risks linked with nuclear energy are presented using several comparisons. (J.S.)

  5. Verification OFENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries for criticality calculations using NEA/NSC benchmarks

    International Nuclear Information System (INIS)

    Bouhaddane, A.; Farkas, G.; Hascik, J.; Slugen, V.

    2015-01-01

    The paper presents verification of selected nuclear data libraries with the aim to apply them to fast reactor calculations. More precise results were achieved for thermal neutrons calculations. This corresponds with the demand for more precise nuclear data for fast reactors. However, fast neutron calculations show some consistency, in particular between ENDF-B/VII.1 and JENDL-4.0 nuclear data libraries. The results support the idea to prefer using newer ENDF-B/VII.1 instead of the previous version ENDF-B/VII.0. Certainly, there are still some issues to be addressed and there is potential to gain more conclusive results. Although, application of ENDF-B/VII.1 and JENDL-4.0 is expected for further calculations. (authors)

  6. Energy and externality environmental regional model

    International Nuclear Information System (INIS)

    Baldi, L.; Bianchi, A.; Peri, M.

    2000-01-01

    The use of environmental externalities in both territorial management and the direction of energy and environment, faces the difficulties arising from their calculation. The so-called MACBET regional model, which has been constructed for Lombardy, is a first brand new attempt to overcome them. MACBET is a calculation model to assess environmental and employment externalities connected to energy use [it

  7. Calculation methods for determining dose equivalent

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Tanner, J.E.; Scherpelz, R.I.; Hadlock, D.E.

    1987-11-01

    A series of calculations of neutron fluence as a function of energy in an anthropomorphic phantom was performed to develop a system for determining effective dose equivalent for external radiation sources. Critical organ dose equivalents are calculated and effective dose equivalents are determined using ICRP-26 [1] methods. Quality factors based on both present definitions and ICRP-40 definitions are used in the analysis. The results of these calculations are presented and discussed. The effective dose equivalent determined using ICRP-26 methods is significantly smaller than the dose equivalent determined by traditional methods. No existing personnel dosimeter or health physics instrument can determine effective dose equivalent. At the present time, the conversion of dosimeter response to dose equivalent is based on calculations for maximal or ''cap'' values using homogeneous spherical or cylindrical phantoms. The evaluated dose equivalent is, therefore, a poor approximation of the effective dose equivalent as defined by ICRP Publication 26. 3 refs., 2 figs., 1 tab

  8. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel

    International Nuclear Information System (INIS)

    Lara, Rafael G.; Maiorino, Jose R.

    2013-01-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others

  9. The software program Peridose to calculate the fetal dose or dose to other critical structures outside the target area in radiation therapy

    International Nuclear Information System (INIS)

    Giessen, P.H. van der

    2001-01-01

    An accurate estimate of the dose outside the target area is of utmost importance when pregnant patients have to undergo radiotherapy, something that occurs in every radiotherapy department once in a while. Such peripheral doses (PD) are also of interest for late effects risk estimations for doses to specific organs as well as estimations of dose to pacemakers. A software program, Peridose, is described to allow easy calculation of this peripheral dose. The calculation is based on data from many publications on peripheral dose measurements, including those by the author. Clinical measurements have shown that by using data averaged over many measurements and different machine types PDs can be estimated with an accuracy of ± 60% (2 standard deviations). The program allows easy and fairly accurate estimates of peripheral doses in patients. Further development to overcome some of the constraints and limitations is desirable. The use of average data is to be preferred if general applicability is to be maintained. (author)

  10. Use of the ''Lagrangian and Eulerian points of view'' in the transient critical heat flux calculations for BWR rod bundles and experimental verifications

    International Nuclear Information System (INIS)

    Marinelli, V.; Pellei, A.; Vallero, P.; Vitanza, C.

    1975-01-01

    The calculations performed in comparison of the ''Lagrangian point of view'', by means of the DOLCE computer code with the local space--time approach of the ''Eulerian point of view'' indicate that the two methods give substantially equivalent results and predict satisfactorily the onset of the transient CHF for the Centro Informazioni Studi Esperienze annuli experimental data and General Electric Company 16-rod bundles data under typical boiling water reactor transients, including loss-of-coolant accident simulations. 9 references

  11. SU-C-BRC-03: Development of a Novel Strategy for On-Demand Monte Carlo and Deterministic Dose Calculation Treatment Planning and Optimization for External Beam Photon and Particle Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Y M; Bush, K; Han, B; Xing, L [Department of Radiation Oncology, Stanford University School of Medicine, Stanford, CA (United States)

    2016-06-15

    Purpose: Accurate and fast dose calculation is a prerequisite of precision radiation therapy in modern photon and particle therapy. While Monte Carlo (MC) dose calculation provides high dosimetric accuracy, the drastically increased computational time hinders its routine use. Deterministic dose calculation methods are fast, but problematic in the presence of tissue density inhomogeneity. We leverage the useful features of deterministic methods and MC to develop a hybrid dose calculation platform with autonomous utilization of MC and deterministic calculation depending on the local geometry, for optimal accuracy and speed. Methods: Our platform utilizes a Geant4 based “localized Monte Carlo” (LMC) method that isolates MC dose calculations only to volumes that have potential for dosimetric inaccuracy. In our approach, additional structures are created encompassing heterogeneous volumes. Deterministic methods calculate dose and energy fluence up to the volume surfaces, where the energy fluence distribution is sampled into discrete histories and transported using MC. Histories exiting the volume are converted back into energy fluence, and transported deterministically. By matching boundary conditions at both interfaces, deterministic dose calculation account for dose perturbations “downstream” of localized heterogeneities. Hybrid dose calculation was performed for water and anthropomorphic phantoms. Results: We achieved <1% agreement between deterministic and MC calculations in the water benchmark for photon and proton beams, and dose differences of 2%–15% could be observed in heterogeneous phantoms. The saving in computational time (a factor ∼4–7 compared to a full Monte Carlo dose calculation) was found to be approximately proportional to the volume of the heterogeneous region. Conclusion: Our hybrid dose calculation approach takes advantage of the computational efficiency of deterministic method and accuracy of MC, providing a practical tool for high

  12. SU-C-BRC-03: Development of a Novel Strategy for On-Demand Monte Carlo and Deterministic Dose Calculation Treatment Planning and Optimization for External Beam Photon and Particle Therapy

    International Nuclear Information System (INIS)

    Yang, Y M; Bush, K; Han, B; Xing, L

    2016-01-01

    Purpose: Accurate and fast dose calculation is a prerequisite of precision radiation therapy in modern photon and particle therapy. While Monte Carlo (MC) dose calculation provides high dosimetric accuracy, the drastically increased computational time hinders its routine use. Deterministic dose calculation methods are fast, but problematic in the presence of tissue density inhomogeneity. We leverage the useful features of deterministic methods and MC to develop a hybrid dose calculation platform with autonomous utilization of MC and deterministic calculation depending on the local geometry, for optimal accuracy and speed. Methods: Our platform utilizes a Geant4 based “localized Monte Carlo” (LMC) method that isolates MC dose calculations only to volumes that have potential for dosimetric inaccuracy. In our approach, additional structures are created encompassing heterogeneous volumes. Deterministic methods calculate dose and energy fluence up to the volume surfaces, where the energy fluence distribution is sampled into discrete histories and transported using MC. Histories exiting the volume are converted back into energy fluence, and transported deterministically. By matching boundary conditions at both interfaces, deterministic dose calculation account for dose perturbations “downstream” of localized heterogeneities. Hybrid dose calculation was performed for water and anthropomorphic phantoms. Results: We achieved <1% agreement between deterministic and MC calculations in the water benchmark for photon and proton beams, and dose differences of 2%–15% could be observed in heterogeneous phantoms. The saving in computational time (a factor ∼4–7 compared to a full Monte Carlo dose calculation) was found to be approximately proportional to the volume of the heterogeneous region. Conclusion: Our hybrid dose calculation approach takes advantage of the computational efficiency of deterministic method and accuracy of MC, providing a practical tool for high

  13. Generation of the problem-dependent data libraries for IFIN-HH WWR-S spent fuel storage criticality and dose calculation

    International Nuclear Information System (INIS)

    Ene, Daniela; Tigau, F.

    1998-01-01

    The methods used for the radioactivity inventory calculation and dose evaluation of the fuel elements irradiated in the WWR-S IFIN-HH reactor are discussed in this work. A particular attention is paid to the processed problem-dependent nuclear libraries. SAS2H, a complex sequence of the SCALE-4.3 code system containing the modules BONAMI - NITAWL - XSDRNPM - COUPLE - ORIGEN-S - XSDOSE, has been assimilated on the IFIN-HH computer and applied to update the ORIGEN-S libraries by producing problem-dependent processed data libraries needed to perform the depletion and shielding analysis. This sequence uses one of the eight associated data libraries of the SCALE-4.3 system according to the choice of the user. The method consists in the following analysis processes: i) lattice cell neutron analysis to produce the flux weighting spectrum for activation library updating; ii) update of the nuclear data constants of the ORIGEN-S libraries; iii) depletion and decay analysis for a specified fuel assembly and irradiation history in order to generate gamma and neutron source strength and spectra. iv) one-dimensional radial shielding calculation for the evaluation of the angular neutron and gamma flux at the surface of a spent fuel shipping cask and further calculation of the dose rates at various points outside the cask. An efficient alternative of the calculation sequence mentioned above is the ARP (Automatic Rapid Processing) method conceived in order to generate independently ORIGEN-S libraries and to reduce substantially the running time. The substance of this method is the generation of the problem-dependent libraries from basis libraries a priori created by SAS2H for specific fuel assembly type and further interpolation of two independent variables, enrichment and burnup. Specific applications concerning WWR-S spent fuel were performed: i) generation of three problem-dependent libraries for the S-36 fuel assembly taking into account the maximum value of the burnup of this

  14. Neutrophil and Monocyte CD64 and CD163 Expression in Critically Ill Neonates and Children with Sepsis: Comparison of Fluorescence Intensities and Calculated Indexes

    Directory of Open Access Journals (Sweden)

    Mojca Groselj-Grenc

    2008-01-01

    Full Text Available Objective. To evaluate the expression of CD64 and CD163 on neutrophils and monocytes in SIRS with/without sepsis and to compare the diagnostic accuracy of CD64 and CD163 molecules expression determined as (1 mean fluorescence intensities (MFI of CD64 and CD163; and (2 the ratio (index of linearized MFI to the fluorescence signal of standardized beads. Patients and methods. Fifty-six critically ill neonates and children with systemic inflammatory response syndrome (SIRS and suspected sepsis, classified into two groups: SIRS with sepsis (n=29 and SIRS without sepsis (n=27. Results. CD64 and CD163 MFI measured on neutrophils and monocytes were elevated in patients with SIRS with sepsis. Diagnostic accuracy of indexes was equal to diagnostic accuracy of MFI for CD64 on neutrophils (0.833 versus 0.854 for day 0 and 0.975 versus 0.983 for day 1 and monocytes (0.811 versus 0.865 for day 0 and 0.825 versus 0.858 for day 1, and CD163 on neutrophils (0.595 versus 0.655 for day 0 and 0.677 versus 0.750 for day 1, but not for CD163 on monocytes. Conclusion. CD64 MFI, CD163 MFI, CD64 indexes for neutrophils and monocytes, and CD163 index for neutrophils can all be used for discrimination of SIRS and sepsis in critically ill neonates and children. CD64 index for neutrophils, however, is superior to all other markers.

  15. Energy security externalities and fuel cycle comparisons

    International Nuclear Information System (INIS)

    Bohi, D.; Toman, M.

    1994-01-01

    Externalities related to 'energy security' may be one way in which the full social costs of energy use diverge from the market prices of energy commodities. Such divergences need to be included in reckoning the full costs of different fuel cycles. In this paper we critically examine potential externalities related to energy security and issues related to the measurement of 2 these externalities, in the context of fuel cycle comparisons

  16. Energy security externalities and fuel cycle comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Bohi, D; Toman, M

    1994-07-01

    Externalities related to 'energy security' may be one way in which the full social costs of energy use diverge from the market prices of energy commodities. Such divergences need to be included in reckoning the full costs of different fuel cycles. In this paper we critically examine potential externalities related to energy security and issues related to the measurement of 2 these externalities, in the context of fuel cycle comparisons.

  17. An updated nuclear criticality slide rule

    International Nuclear Information System (INIS)

    Hopper, C.M.; Broadhead, B.L.

    1998-04-01

    This Volume 2 contains the functional version of the updated nuclear criticality slide rule (more accurately, sliding graphs) that is referenced in An Updated Nuclear Criticality Slide Rule: Technical Basis, NUREG/CR-6504, Vol. 1 (ORNL/TM-13322/V1). This functional slide rule provides a readily usable open-quotes in-handclose quotes method for estimating pertinent nuclear criticality accident information from sliding graphs, thereby permitting (1) the rapid estimation of pertinent criticality accident information without laborious or sophisticated calculations in a nuclear criticality emergency situation, (2) the appraisal of potential fission yields and external personnel radiation exposures for facility safety analyses, and (3) a technical basis for emergency preparedness and training programs at nonreactor nuclear facilities. The slide rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, distance from the fission source, and time-after criticality accidents for five different critical systems. Another sliding graph permits the estimation of critical solution fission yields based upon fissile material concentration, critical vessel geometry, and solution addition rate. Another graph provides neutron and gamma dose-reduction factors for water, steel, and concrete. Graphs from historic documents are provided as references for estimating critical parameters of various fissile material systems. Conversion factors for various English and metric units are provided for quick reference

  18. The order parameter and susceptibility of the 3D Ising-like system in an external field near the phase transition point

    Directory of Open Access Journals (Sweden)

    M.P. Kozlovskii

    2010-01-01

    Full Text Available The present work is devoted to the investigation of the 3D Ising-like model in the presence of an external field in the vicinity of critical point. The method of collective variables is used. General expressions for the order parameter and susceptibility are calculated as functions of temperature and the external field as well as scaling functions of that are explicitly obtained. The results are compared with the ones obtained within the framework of parametric representation of the equation of state and Monte Carlo simulations. New expression for the exit point from critical regime of the order parameter fluctuations is proposed and used for the calculation.

  19. Petroleum - as far as available. Analysis and critical model calculation up to the year 2200. Erdoel - soweit verfuegbar. Analyse und kritische Modellrechnung bis zum Jahr 2200

    Energy Technology Data Exchange (ETDEWEB)

    Stahmer, A M

    1979-01-01

    An outline is given on the most important oil producing countries and on the petroleum reserves available there as they developed from 1960 to 1978. For the last few years, the figures at hand show a distinct decline of new discoveries i.e. of proved reserves, compared with consumption. Prospective development factors are critically examined, and a model of potential petroleum supply is developed covering the period up to the year 2200 and requiring the adjustment of production and consumption. Until 1986, the model shows a plateau of international oil production (and of consumption) amounting to 3 billion tons per year on an average, and subsequently a constant decline of output. The international average consumption being kept at a relatively high level is to give the world economy the opportunity to develop alternative energy sources. From this detailed statement follows automatically the demand to increase exploratory activities and to lead an optimal consumption policy. In a separate section deposits of heavy oil sands and oil shales and respective production technologies are dealt with as well as their potential contribution to meet energy needs.

  20. Critical dynamics

    International Nuclear Information System (INIS)

    Dekker, H.

    1980-01-01

    It is shown how to solve the master equation for a Markov process including a critical point by means of successive approximations in terms of a small parameter. A critical point occurs if, by adjusting an externally controlled quantity, the system shows a transition from normal monostable to bistable behaviour. The fundamental idea of the theory is to separate the master equation into its proper irreducible part and a corrective remainder. The irreducible or zeroth order stochastic approximation will be a relatively simple Fokker-Planck equation that contains the essential features of the process. Once the solution of this irreducible equation is known, the higher order corrections in the original master equation can be incorporated in a systematic manner. (Auth.)

  1. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  2. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

    A definition of external dosimetry r external sources dosimetry,physical and mathematical treatment of the interaction of gamma radiation with a minimal area in that direction. Concept of attenuation coefficient, cumulated effect by polyenergetic sources, exposition rate, units, cumulated dose,shielding, foton shielding, depth calculation, materials used for shielding.Beta shielding, consideration of range and maximum β energy , low stopping radiation by use of low Z shielding. Tables for β energy of β emitters, I (tau) factor, energy-range curves for β emitters in aqueous media, gamma attenuation factors for U, W and Pb. Y factor for bone tissue,muscle and air, build-up factors

  3. Plasma Vertical Control with Internal and External Coils in Nest Step Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Kessel; P. Heitzenroeder; C. Jun

    2000-11-03

    Vertical stability and control are examined for a tokamak configuration intended to be a generic representation of next step devices. Vertical stability calculations show that a critical resistive wall location can be determined for realistic structures, and that the introduction of small amounts of low resistivity material to an all steel structure can significantly reduce the vertical instability growth rate. Vertical control simulations show that internal control coils require significantly less feedback power than external coils, and that low resistivity materials can allow very low feedback powers or coils to be located externally with reasonable feedback powers.

  4. Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Costa da Silva, Adilson; Carvalho da Silva, Fernando [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil); Senra Martinez, Aquilino, E-mail: aquilino@lmp.ufrj.br [COPPE/UFRJ, Programa de Engenharia Nuclear, Caixa Postal 68509, 21941-914, Rio de Janeiro (Brazil)

    2011-07-15

    Highlights: > We proposed a new neutron diffusion hybrid equation with external neutron source. > A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. > 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

  5. Prediction of the neutrons subcritical multiplication using the diffusion hybrid equation with external neutron sources

    International Nuclear Information System (INIS)

    Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

    2011-01-01

    Highlights: → We proposed a new neutron diffusion hybrid equation with external neutron source. → A coarse mesh finite difference method for the adjoint flux and reactivity calculation was developed. → 1/M curve to predict the criticality condition is used. - Abstract: We used the neutron diffusion hybrid equation, in cartesian geometry with external neutron sources to predict the subcritical multiplication of neutrons in a pressurized water reactor, using a 1/M curve to predict the criticality condition. A Coarse Mesh Finite Difference Method was developed for the adjoint flux calculation and to obtain the reactivity values of the reactor. The results obtained were compared with benchmark values in order to validate the methodology presented in this paper.

  6. Calculation of the power factor using the neutron diffusion hybrid equation

    International Nuclear Information System (INIS)

    Costa da Silva, Adilson; Carvalho da Silva, Fernando; Senra Martinez, Aquilino

    2013-01-01

    Highlights: ► A neutron diffusion hybrid equation with an external neutron source was used. ► Nodal expansion method to obtain the neutron flux was used. ► Nuclear power factors in each fuel element in the reactor core were calculated. ► The results obtained were very accurate. -- Abstract: In this paper, we used a neutron diffusion hybrid equation with an external neutron source to calculate nuclear power factors in each fuel element in the reactor core. We used the nodal expansion method to obtain the neutron flux for a given control rods bank position. The results were compared with results obtained for eigenvalue problem near criticality condition and fixed source problem during the start-up of the reactor, where external neutron sources are extremely important for the stabilization of external neutron detectors.

  7. Critical Behavior of Light in Mode-Locked Lasers

    Science.gov (United States)

    Weill, Rafi; Rosen, Amir; Gordon, Ariel; Gat, Omri; Fischer, Baruch

    2005-06-01

    Light is shown to exhibit critical and tricritical behavior in passively mode-locked lasers with externally injected pulses. It is a first and unique example of critical phenomena in a one-dimensional many-body light-mode system. The phase diagrams consist of regimes with continuous wave, driven parapulses, spontaneous pulses via mode condensation, and heterogeneous pulses, separated by phase transition lines that terminate with critical or tricritical points. Enhanced non-Gaussian fluctuations and collective dynamics are present at the critical and tricritical points, showing a mode system analog of the critical opalescence phenomenon. The critical exponents are calculated and shown to comply with the mean field theory, which is rigorous in the light system.

  8. Radioactive cloud dose calculations

    International Nuclear Information System (INIS)

    Healy, J.W.

    1984-01-01

    Radiological dosage principles, as well as methods for calculating external and internal dose rates, following dispersion and deposition of radioactive materials in the atmosphere are described. Emphasis has been placed on analytical solutions that are appropriate for hand calculations. In addition, the methods for calculating dose rates from ingestion are discussed. A brief description of several computer programs are included for information on radionuclides. There has been no attempt to be comprehensive, and only a sampling of programs has been selected to illustrate the variety available

  9. Regulating multiple externalities

    DEFF Research Database (Denmark)

    Waldo, Staffan; Jensen, Frank; Nielsen, Max

    2016-01-01

    Open access is a well-known externality problem in fisheries causing excess capacity and overfishing. Due to global warming, externality problems from CO2 emissions have gained increased interest. With two externality problems, a first-best optimum can be achieved by using two regulatory instrume......Open access is a well-known externality problem in fisheries causing excess capacity and overfishing. Due to global warming, externality problems from CO2 emissions have gained increased interest. With two externality problems, a first-best optimum can be achieved by using two regulatory...

  10. ExternE National Implementation Finland

    Energy Technology Data Exchange (ETDEWEB)

    Pingoud, K; Maelkki, H; Wihersaari, M; Pirilae, P [VTT Energy, Espoo (Finland); Hongisto, M [Imatran Voima Oy, Vantaa (Finland); Siitonen, S [Ekono Energy Ltd, Espoo (Finland); Johansson, M [Finnish Environment Institute, Helsinki (Finland)

    1999-07-01

    ExternE National Implementation is a continuation of the ExternE Project, funded in part by the European Commission's Joule III Programme. This study is the result of the ExternE National Implementation Project for Finland. Three fuel cycles were selected for the Finnish study: coal, peat and wood-derived biomass, which together are responsible for about 40% of total electricity generation in Finland and about 75% of the non-nuclear fuel based generation. The estimated external costs or damages were dominated by the global warming (GW) impacts in the coal and peat fuel cycles, but knowledge of the true GW impacts is still uncertain. From among other impacts that were valued in monetary terms the human health damages due to airborne emissions dominated in all the three fuel cycles. Monetary valuation for ecosystem impacts is not possible using the ExternE methodology at present. The Meri-Pori power station representing the coal fuel cycle is one of the world's cleanest and most efficient coal-fired power plants with a condensing turbine. The coal is imported mainly from Poland. The estimated health damages were about 4 mECU/kWh, crop damages an order of magnitude lower and damages caused to building materials two orders of magnitude lower. The power stations of the peat and biomass fuel cycles are of CHP type, generating electricity and heat for the district heating systems of two cities. Their fuels are of domestic origin. The estimated health damages allocated to electricity generation were about 5 and 6 mECU/kWh, respectively. The estimates were case-specific and thus an generalisation of the results to the whole electricity generation in Finland is unrealistic. Despite the uncertainties and limitations of the methodology, it is a promising tool in the comparison of similar kinds of fuel cycles, new power plants and pollution abatement technologies and different plant locations with each other. (orig.)

  11. ExternE National Implementation Finland

    International Nuclear Information System (INIS)

    Pingoud, K.; Maelkki, H.; Wihersaari, M.; Pirilae, P.; Hongisto, M.; Siitonen, S.; Johansson, M.

    1999-01-01

    ExternE National Implementation is a continuation of the ExternE Project, funded in part by the European Commission's Joule III Programme. This study is the result of the ExternE National Implementation Project for Finland. Three fuel cycles were selected for the Finnish study: coal, peat and wood-derived biomass, which together are responsible for about 40% of total electricity generation in Finland and about 75% of the non-nuclear fuel based generation. The estimated external costs or damages were dominated by the global warming (GW) impacts in the coal and peat fuel cycles, but knowledge of the true GW impacts is still uncertain. From among other impacts that were valued in monetary terms the human health damages due to airborne emissions dominated in all the three fuel cycles. Monetary valuation for ecosystem impacts is not possible using the ExternE methodology at present. The Meri-Pori power station representing the coal fuel cycle is one of the world's cleanest and most efficient coal-fired power plants with a condensing turbine. The coal is imported mainly from Poland. The estimated health damages were about 4 mECU/kWh, crop damages an order of magnitude lower and damages caused to building materials two orders of magnitude lower. The power stations of the peat and biomass fuel cycles are of CHP type, generating electricity and heat for the district heating systems of two cities. Their fuels are of domestic origin. The estimated health damages allocated to electricity generation were about 5 and 6 mECU/kWh, respectively. The estimates were case-specific and thus an generalisation of the results to the whole electricity generation in Finland is unrealistic. Despite the uncertainties and limitations of the methodology, it is a promising tool in the comparison of similar kinds of fuel cycles, new power plants and pollution abatement technologies and different plant locations with each other. (orig.)

  12. ExternE National Implementation Finland

    Energy Technology Data Exchange (ETDEWEB)

    Pingoud, K.; Maelkki, H.; Wihersaari, M.; Pirilae, P. [VTT Energy, Espoo (Finland); Hongisto, M. [Imatran Voima Oy, Vantaa (Finland); Siitonen, S. [Ekono Energy Ltd, Espoo (Finland); Johansson, M. [Finnish Environment Institute, Helsinki (Finland)

    1999-07-01

    ExternE National Implementation is a continuation of the ExternE Project, funded in part by the European Commission's Joule III Programme. This study is the result of the ExternE National Implementation Project for Finland. Three fuel cycles were selected for the Finnish study: coal, peat and wood-derived biomass, which together are responsible for about 40% of total electricity generation in Finland and about 75% of the non-nuclear fuel based generation. The estimated external costs or damages were dominated by the global warming (GW) impacts in the coal and peat fuel cycles, but knowledge of the true GW impacts is still uncertain. From among other impacts that were valued in monetary terms the human health damages due to airborne emissions dominated in all the three fuel cycles. Monetary valuation for ecosystem impacts is not possible using the ExternE methodology at present. The Meri-Pori power station representing the coal fuel cycle is one of the world's cleanest and most efficient coal-fired power plants with a condensing turbine. The coal is imported mainly from Poland. The estimated health damages were about 4 mECU/kWh, crop damages an order of magnitude lower and damages caused to building materials two orders of magnitude lower. The power stations of the peat and biomass fuel cycles are of CHP type, generating electricity and heat for the district heating systems of two cities. Their fuels are of domestic origin. The estimated health damages allocated to electricity generation were about 5 and 6 mECU/kWh, respectively. The estimates were case-specific and thus an generalisation of the results to the whole electricity generation in Finland is unrealistic. Despite the uncertainties and limitations of the methodology, it is a promising tool in the comparison of similar kinds of fuel cycles, new power plants and pollution abatement technologies and different plant locations with each other. (orig.)

  13. Estimation of Externalities for Juragua Nuclear Project

    International Nuclear Information System (INIS)

    Mora, H. R.; Carbonell, L. T.

    2002-01-01

    Estimation of externalities allows taking into account environmental impacts due to any activity in total costs calculation. In the present work, the external costs of electricity generation from nuclear energy were calculated considering three scenarios: normal operation (routine releases), accident situation and solid waste disposal. A comparison between these results and those obtained for electricity generation from fossil fuels was made. IAEA proposals of Simplified methodologies were used for externality calculations. The Juragua project was selected as a study case; it is based in two energetic blocks both PWR, VVER 440/318 type with a plant capacity of 417 MWe each. Four impact ways were considered for all scenarios: (1) Inhalation of radionuclides in the air, (2) External irradiation from radionuclides immersed in clouds, (3) External irradiation from deposited radionuclides and (4) Ingestion of radionuclides in agricultural products. Besides, two impact categories (local and regional) for all scenarios were considered. The total cost of externalities was 0.01425 c/kWh, value smaller than the one obtained for electricity generation from fossil fuel (0.256 c/kWh). For the normal operation scenario, the external cost calculated was 0.00112 c/kWh, for accident situation 0.01103 c/kWh, and for the solid wastes management scenario 0.0021 c/kWh. The high value obtained for solid waste disposal scenario is due to repository placement features. (author)

  14. External Otitis (Swimmer's Ear)

    Science.gov (United States)

    ... otitis. Fungal external otitis (otomycosis), typically caused by Aspergillus niger or Candida albicans, is less common. Boils are ... in the ear. Fungal external otitis caused by Aspergillus niger usually causes grayish black or yellow dots (called ...

  15. ExternE: Externalities of energy Vol. 2. Methodology

    International Nuclear Information System (INIS)

    Berry, J.; Holland, M.; Watkiss, P.

    1995-01-01

    This report describes the methodology used by the ExternE Project of the European Commission (DGXII) JOULE Programme for assessment of the external costs of energy. It is one of a series of reports describing analysis of nuclear, fossil and renewable fuel cycles for assessment of the externalities associated with electricity generation. Part I of the report deals with analysis of impacts, and Part II with the economic valuation of those impacts. Analysis is conducted on a marginal basis, to allow the effect of an incremental investment in a given technology to be quantified. Attention has been paid to the specificity of results with respect to the location of fuel cycle activities, the precise technologies used, and the type and source of fuel. The main advantages of this detailed approach are as follows: It takes full and proper account of the variability of impacts that might result from different power projects; It is more transparent than analysis based on hypothetically 'representative' cases for each of the different fuel cycles; It provides a framework for consistent comparison between fuel cycles. A wide variety of impacts have been considered. These include the effects of air pollution on the natural and human environment, consequences of accidents in the workplace, impacts of noise and visual intrusion on amenity, and the effects of climate change arising from the release of greenhouse gases. Wherever possible we have used the 'impact pathway' or 'damage function' approach to follow the analysis from identification of burdens (e.g. emissions) through to impact assessment and then valuation in monetary terms. This has required a detailed knowledge of the technologies involved, pollutant dispersion, analysis of effects on human and environmental health, and economics. In view of this the project brought together a multi-disciplinary team with experts from many European countries and the USA. The spatial and temporal ranges considered in the analysis are

  16. Utilization of Keno system for criticality calculation

    International Nuclear Information System (INIS)

    Maragni, M.G.

    1990-01-01

    Several studies involving benchmarks have been performed with the KENO-IV code in order to utilize it in a more efficient way at IPEN-COPESP. The influence of different cross section libraries has been verifed. The Hansen-Roach library produced better results for fast systems, while GAMTEC-II code was more efficient for thermal systems. For reflectors it has been shown that the differential albedo and automatic reflection options are more appropriate for infinite and finite reflectors, respectively. A number of histories greater than 30.000 did not seem to improve the results. Plutonium systems should be treated with special care. (author) [pt

  17. ASH External Web Portal (External Portal) -

    Data.gov (United States)

    Department of Transportation — The ASH External Web Portal is a web-based portal that provides single sign-on functionality, making the web portal a single location from which to be authenticated...

  18. Computerization of effluent management and external dose calculation using the 'ODCM' methodology applied to Almaraz-NPP; Mecanizacion informatica de la gestion de efluentes y calculo de dosis al exterior usando la metodologia MCDE aplicada a C.N. Almaraz

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Gutierrez, M E [Empresarios Agrupados, A.I.E., Madrid (Spain); Sustacha Duo, D [C.N. Almaraz, Caceres (Spain)

    1993-12-15

    The ODCM (Offsite Dose Calculation Manual), the official operational document for all nuclear power plants develops the details for the technical specifications for discharges and governs their practical application. The use of ODCM methodology for managing and controlling data associated with radioactive discharges, as well as the subsequent processing of this data to assess the radiological impact, requires and generates a large volume of data, which demands the frequent application of laborious and complex calculation processes, making computerization necessary. The computer application created for Almaraz NPP has the capacity to store and manage data on all discharges, evaluate their effects, presents reports and copies the information to be sent periodically to the CSN (Spanish Nuclear Regulatory Commission) on a magnetic tape. The radiological impact of an actual or possible discharge can be evaluated at anytime and, furthermore, general or particular reports and graphs on the discharges and doses over time can be readily obtained. The application is run on a personal computer under a relational database management system. This interactive application is based on menus and windows. (author)

  19. Critical heat fluxes and liquid distribution in annular channels in the dispersion-annular flow

    International Nuclear Information System (INIS)

    Boltenko, Eh.A.; Pomet'ko, R.S.

    1984-01-01

    On the basis of using the dependence of intensity of total mass transfer between the flux nucleus and wall film obtained for tubes with uniform heat release and taking into account the peculiarities of mass transfer between the flux nucleus and wall film in annular channels the technique for calculating the liquid distribution and critical capacity of annular channels with internal, external and bilateral heating at uniform and non-uniform heat release over the length is proposed. The calculation of annular channels critical capacity according to the suggested technique is performed. A satisfactory agreement of calculation results with the experimental data is attained

  20. Development of mathematical pediatric phantoms for internal dose calculations: designs, limitations, and prospects

    International Nuclear Information System (INIS)

    Cristy, M.

    1980-01-01

    Mathematical phantoms of the human body at various ages are employed with Monte Carlo radiation transport codes for calculation of photon specific absorbed fractions. The author has developed a pediatric phantom series based on the design of the adult phantom, but with explicit equations for each organ so that organ sizes and marrow distributions could be assigned properly. Since the phantoms comprise simple geometric shapes, predictive dose capability is limited when geometry is critical to the calculation. Hence, there is a demand for better phantom design in situations where geometry is critical, such as for external irradiation or for internal emitters with low energy photons. Recent advances in computerized axial tomography (CAT) present the potential for derivation of anatomical information, which is so critical to development of phantoms, and ongoing developmental work on compuer architecture to handle large arrays for Monte Carlo calculations should make complex-geometry dose calculations economically feasible within this decade

  1. Analysis of Stiffened Penstock External Pressure Stability Based on Immune Algorithm and Neural Network

    Directory of Open Access Journals (Sweden)

    Wensheng Dong

    2014-01-01

    Full Text Available The critical external pressure stability calculation of stiffened penstock in the hydroelectric power station is very important work for penstock design. At present, different assumptions and boundary simplification are adopted by different calculation methods which sometimes cause huge differences too. In this paper, we present an immune based artificial neural network model via the model and stability theory of elastic ring, we study effects of some factors (such as pipe diameter, pipe wall thickness, sectional size of stiffening ring, and spacing between stiffening rings on penstock critical external pressure during huge thin-wall procedure of penstock. The results reveal that the variation of diameter and wall thickness can lead to sharp variation of penstock external pressure bearing capacity and then give the change interval of it. This paper presents an optimizing design method to optimize sectional size and spacing of stiffening rings and to determine penstock bearing capacity coordinate with the bearing capacity of stiffening rings and penstock external pressure stability coordinate with its strength safety. As a practical example, the simulation results illustrate that the method presented in this paper is available and can efficiently overcome inherent defects of BP neural network.

  2. External Validation of a Prediction Model for Successful External Cephalic Version

    NARCIS (Netherlands)

    de Hundt, Marcella; Vlemmix, Floortje; Kok, Marjolein; van der Steeg, Jan W.; Bais, Joke M.; Mol, Ben W.; van der Post, Joris A.

    2012-01-01

    We sought external validation of a prediction model for the probability of a successful external cephalic version (ECV). We evaluated the performance of the prediction model with calibration and discrimination. For clinical practice, we developed a score chart to calculate the probability of a

  3. Declination Calculator

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Declination is calculated using the current International Geomagnetic Reference Field (IGRF) model. Declination is calculated using the current World Magnetic Model...

  4. Research on cooling of ultra high critical heat flux with external flow boiling of water. Challenge to achieve ultra high critical heat flux and improvement in estimation of critical heat flux. JAERI's nuclear research promotion program, H11-004 (Contract research)

    International Nuclear Information System (INIS)

    Monde, Masanori; Mitsutake, Yuichi; Ishida, Kenji; Hino, Ryutaro

    2003-03-01

    An ultra high critical heat flux (CHF) has been challenged with a highly subcooled water jet impinging on a small rectangular heated surface. Major objective of the study is to achieve an ultra high heat flux cooling as large as 100 MW/m 2 and to establish an accurate estimation method of the CHF. The experiments were carried out over the experimental range; a fixed jet diameter of 2 mm, jet velocity of 5 - 35 m/s, degree of subcooling of 80 - 170 K and system pressure of 0.1 - 1.0 MPa. The rectangular heated surface with a thin nickel foil of 0.03 - 0.3 mm in thickness, 5 and 10 mm in length, and 4 mm in width and heated by a direct current. Effects of thickness of heater wall, jet velocity and subcooling on the CHF were experimentally elucidated. The experimental results show that the CHF decreases about 50% as the heater thickness, namely heat capacity of heater decreases. Characteristics of the CHF with heater length of 10 mm are correlated within ±20% by the generalized correlation of subcooled CHF proposed by the authors. However, the CHF with the shorter heater length of 5 mm shows large deviation of -40% especially at lower subcooling and higher velocity. The maximum CHF of 212 MW/m 2 was achieved at the subcooling of 151 K, the jet velocity of 35 m/s and system pressure of 0.5 MPa. The maximum CHF under atmospheric pressure approaches to 48% of the ultimate maximum heat flux given by the assumptions that vapor molecules leave a liquid-vapor interface at the average speed of a Boltzman-Maxwellian gas and any molecules returning to the interface are not permitted. The ratio of the CHF and ultimate maximum heat flux was considerably enhanced from the existing record of 30%. This study can give the feasibility of ultra high heat flux removal facing in a development of components such as a diverter of a fusion reactor. (author)

  5. RIFIFI: Analytical calculation method of the critical condition and flux in a varied regions reactor by two-group theory and one dimension developed for the Mercury-Ferranti computer; Rififi: methode de calcul analytique de la condition critique et des flux d'une pile a regions variees en theorie a deux groupes et a une dimension programmee pour le calculateur electronique Mercury (Ferranti)

    Energy Technology Data Exchange (ETDEWEB)

    Amouyal, A; Bacher, P; Lago, B; Mengin, F L; Parker, E [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1960-07-01

    The calculation method presented in this report has been developed for the Mercury-Ferranti computer of the C.E.N.S. This calculation method allows to resolve the diffusion equations and continuity equations of flux and flow with two groups of neutrons and one dimension in spherical, cylindrical and linear geometry. In the cylindrical and linear configurations, we can take the height and extrapolated radius into account. The critical condition can be realised by varying linearly one or more parameters: k{sub {infinity}}, medium frontier, height or extrapolated radius. The calculation method enables also to calculate the flux, adjoint flux and various integrals. In the first part, it explains what is needed to know before using the method: data presentation, method possibilities, results presentation with some information about restrictions, accuracy and calculation time. The complete formulation of the calculation method is given in the second part. (M.P.)

  6. CriticalEd

    DEFF Research Database (Denmark)

    Kjellberg, Caspar Mølholt; Meredith, David

    2014-01-01

    . Since the comments are not input sequentially, with regard to position, but in arbitrary order, this list must be sorted by copy/pasting the rows into place—an error-prone and time-consuming process. Scholars who produce critical editions typically use off-the-shelf music notation software......The best text method is commonly applied among music scholars engaged in producing critical editions. In this method, a comment list is compiled, consisting of variant readings and editorial emendations. This list is maintained by inserting the comments into a document as the changes are made......, consisting of a Sibelius plug-in, a cross-platform application, called CriticalEd, and a REST-based solution, which handles data storage/retrieval. A prototype has been tested at the Danish Centre for Music Publication, and the results suggest that the system could greatly improve the efficiency...

  7. External radiation surveillance

    International Nuclear Information System (INIS)

    Antonio, E.J.

    1995-01-01

    This section of the 1994 Hanford Site Environmental Report describes how external radiation was measured, how surveys were performed, and the results of these measurements and surveys. External radiation exposure rates were measured at locations on and off the Hanford Site using thermoluminescent dosimeters (TLD). External radiation and contamination surveys were also performed with portable radiation survey instruments at locations on and around the Hanford Site

  8. External radiation surveillance

    Energy Technology Data Exchange (ETDEWEB)

    Antonio, E.J.

    1995-06-01

    This section of the 1994 Hanford Site Environmental Report describes how external radiation was measured, how surveys were performed, and the results of these measurements and surveys. External radiation exposure rates were measured at locations on and off the Hanford Site using thermoluminescent dosimeters (TLD). External radiation and contamination surveys were also performed with portable radiation survey instruments at locations on and around the Hanford Site.

  9. External events analysis of the Ignalina Nuclear Power Plant

    International Nuclear Information System (INIS)

    Liaukonis, Mindaugas; Augutis, Juozas

    1999-01-01

    This paper presents analysis of external events impact on the safe operation of the Ignalina Nuclear Power Plant (INPP) safety systems. Analysis was based on the probabilistic estimation and modelling of the external hazards. The screening criteria were applied to the number of external hazards. The following external events such as aircraft failure on the INPP, external flooding, fire, extreme winds requiring further bounding study were analysed. Mathematical models were developed and event probabilities were calculated. External events analysis showed rather limited external events danger to Ignalina NPP. Results of the analysis were compared to analogous analysis in western NPPs and no great differences were specified. Calculations performed show that external events can not significantly influence the safety level of the Ignalina NPP operation. (author)

  10. A note on numerical solution to the problem of criticality

    International Nuclear Information System (INIS)

    Kyncl, J.

    2002-01-01

    The contribution deals with numerical solution to the problem of criticality for neutron transport equation by the external source iteration method. Especially, the speed of convergence is examined. It is shown that if neutron absorption in the medium considered is high and if the space region occupied by the medium is large then a slow convergence of the iterations can be expected. This expectation is confirmed by results to CB4 benchmark obtained by MCNP code. Besides the results presented some questions concerning applications of them to criticality calculations are pointed out (Author)

  11. Critical energy of superconducting composites

    International Nuclear Information System (INIS)

    Jayakumar, R.

    1987-01-01

    The stability of superconducting composites is studied in one-dimensional geometry and critical quench energies are calculated by solving for the steady state temperature profile which gives the minimum energy. The present calculations give lower values for the critical energy than previous estimates. The calculations are shown to be applicable to both direct cooled and impregnated conductors. Critical energies are also calculated including the effect of temperature dependence of conductor properties. (author)

  12. AIRGAMMA, External Gamma-Ray Exposure from Radioactive Cloud

    International Nuclear Information System (INIS)

    Hidaka, Akihide; Iijima, Tshinori

    1989-01-01

    1 - Description of program or function: AIRGAMMA calculates quickly the external exposure to gamma rays from a radioactive cloud. 2 - Method of solution: The external exposure is calculated by interpolating the normalized doses providing on the basis of the Gaussian plume model. 3 - Restrictions on the complexity of the problem: Memory requirement is 30 Kbytes

  13. Towards a beyond 1 GHz solid-state nuclear magnetic resonance: External lock operation in an external current mode for a 500 MHz nuclear magnetic resonance

    International Nuclear Information System (INIS)

    Takahashi, Masato; Maeda, Hideaki; Ebisawa, Yusuke; Tennmei, Konosuke; Yanagisawa, Yoshinori; Nakagome, Hideki; Hosono, Masami; Takasugi, Kenji; Hase, Takashi; Miyazaki, Takayoshi; Fujito, Teruaki; Kiyoshi, Tsukasa; Yamazaki, Toshio

    2012-01-01

    Achieving a higher magnetic field is important for solid-state nuclear magnetic resonance (NMR). But a conventional low temperature superconducting (LTS) magnet cannot exceed 1 GHz (23.5 T) due to the critical magnetic field. Thus, we started a project to replace the Nb 3 Sn innermost coil of an existing 920 MHz NMR (21.6 T) with a Bi-2223 high temperature superconducting (HTS) innermost coil. Unfortunately, the HTS magnet cannot be operated in persistent current mode; an external dc power supply is required to operate the NMR magnet, causing magnetic field fluctuations. These fluctuations can be stabilized by a field-frequency lock system based on an external NMR detection coil. We demonstrate here such a field-frequency lock system in a 500 MHz LTS NMR magnet operated in an external current mode. The system uses a 7 Li sample in a microcoil as external NMR detection system. The required field compensation is calculated from the frequency of the FID as measured with a frequency counter. The system detects the FID signal, determining the FID frequency, and calculates the required compensation coil current to stabilize the sample magnetic field. The magnetic field was stabilized at 0.05 ppm/3 h for magnetic field fluctuations of around 10 ppm. This method is especially effective for a magnet with large magnetic field fluctuations. The magnetic field of the compensation coil is relatively inhomogeneous in these cases and the inhomogeneity of the compensation coil can be taken into account.

  14. ExternE: Externalities of energy Vol. 5. Nuclear

    International Nuclear Information System (INIS)

    Dreicer, M.; Tort, V.; Manen, P.

    1995-01-01

    Since the early 1970s, there has been increased interest in the environmental impacts that are caused by the generation of electricity. The comparative risk assessment studies at that time used mainly deaths and injuries as impact indicators. By the end of the 1980s studies changed to the assessment of the costs imposed on society and the environment that were not included in the market price of the energy produced, the so-called external costs. The preliminary studies that were published set the conceptual basis, grounded in neo-classical economics, for the valuation of the health and environmental impacts that could be assessed. As a consequence of the many questions raised by the methodologies employed by these early studies, Directorate General XII (DG XII) of the Commission of the European Communities established a collaborative research programme with the United States Department of Energy to identify an appropriate methodology for this type of work. Following the completion of this collaboration, the DG XII programme has continued as the ExternE project. The main objective of the work carried out at CEPN was to develop an impact pathway methodology for the nuclear fuel cycle that would be consistent with the methodologies developed for other fuel cycles, without loosing the nuclear-specific techniques required for a proper evaluation. In this way, comparisons between the different fuel cycles would be possible. This report presents the methodology and demonstration of the results in the context of the French nuclear fuel cycle. The United States team at Oak Ridge National Laboratory has previously issued a draft report on the results of their assessment. The French fuel cycle was broken down into 8 separate stages. Reference sites and 1990s technology were chosen to represent the total nuclear fuel cycle, as it exists today. In addition, the transportation of material between the sites was considered. The facilities are assessed for routine operation, except

  15. Critical scattering by bubbles

    International Nuclear Information System (INIS)

    Fiedler-Ferrari, N.; Nussenzveig, H.M.

    1986-11-01

    We apply the complex angular momentum theory to the problem of the critical scattering of light by spherical cavities in the high frequency limit (permittivity greater than the external media) (e.g, air bubble in water) (M.W.O.) [pt

  16. Criticality handbook. Pt. 1

    International Nuclear Information System (INIS)

    Heinicke, W.; Krug, H.; Thomas, W.; Weber, W.; Gmal, B.

    1985-12-01

    The GRS Criticality Handbook is intended as a source of information on criticality problems for the persons concerned in industry, authorities, or research laboratories. It is to serve as a guide allowing quick and appropriate evaluation of criticality problems during design or erection of nuclear installations. This present issue replaces the one published in 1979, presenting revised and new data in a modified construction, but within the framework of the proven basic structure of the Handbook. Some fundamental knowledge is required of criticality problems and the relevant terms and definitions of nuclear safety, in order to fully deploy the information given. Part 1 of the Handbook therefore first introduces terminology and definitions, followed by experimental methods and calculation models for criticality calculations. The next chapters deal with the function and efficiency of neutron reflectors and neutron absorbers, measuring methods for criticality monitoring, organisational safety measures, and criticality accidents and their subsequent analysis. (orig./HP) [de

  17. KENO, Multigroup P1 Scattering Monte-Carlo Transport Calculation for Criticality, Keff, Flux in 3-D. KENO-5, SCALE-1 Module with Pn Scattering, Super-grouping, Diffusion Albedo Reflection

    International Nuclear Information System (INIS)

    Petrie, L.M.; Landers, N.F.

    2001-01-01

    1 - Description of problem or function: KENO is a multigroup, Monte Carlo criticality code containing a special geometry package which allows easy description of systems composed of cylinders, spheres, and cuboids (rectangular parallelepipeds) arranged in any order with only one restriction. They cannot be rotated or translated. Each geometrical region must be described as completely enclosing all regions interior to it. For systems not describable using this special geometry package, the program can use the generalized geometry package (GEOM) developed for the O5R Monte Carlo code. It allows any system that can be described by a collection of planes and/or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. The entire problem can be mocked up in generalized geometry, or one generalized geometry unit or box type can be used alone or in combination with standard KENO units or box types. Rectangular arrays of fissile units are allowed with or without external reflector regions. Output from KENO consists of k eff for the system plus an estimate of its standard deviation and the leakage, absorption, and fissions for each energy group plus the totals for all groups. Flux as a function of energy group and region and fission densities as a function of region are optional output. KENO-4: Added features include a neutron balance edit, PICTURE routines to check the input geometry, and a random number sequencing subroutine written in FORTRAN-4. 2 - Method of solution: The scattering treatment used in KENO assumes that the differential neutron scattering cross section can be represented by a P1 Legendre polynomial. Absorption of neutrons in KENO is not allowed. Instead, at each collision point of a neutron tracking history the weight of the neutron is reduced by the absorption probability. When the neutron weight has been reduced below a specified point for the region in which the collision occurs, Russian roulette is played to determine if the

  18. Critical Axial Load

    Directory of Open Access Journals (Sweden)

    Walt Wells

    2008-01-01

    Full Text Available Our objective in this paper is to solve a second order differential equation for a long, simply supported column member subjected to a lateral axial load using Heun's numerical method. We will use the solution to find the critical load at which the column member will fail due to buckling. We will calculate this load using Euler's derived analytical approach for an exact solution, as well as Euler's Numerical Method. We will then compare the three calculated values to see how much they deviate from one another. During the critical load calculation, it will be necessary to calculate the moment of inertia for the column member.

  19. Computation of high Reynolds number internal/external flows

    Science.gov (United States)

    Cline, M. C.; Wilmoth, R. G.

    1981-01-01

    A general, user oriented computer program, called VNAP2, was developed to calculate high Reynolds number, internal/ external flows. The VNAP2 program solves the two dimensional, time dependent Navier-Stokes equations. The turbulence is modeled with either a mixing-length, a one transport equation, or a two transport equation model. Interior grid points are computed using the explicit MacCormack Scheme with special procedures to speed up the calculation in the fine grid. All boundary conditions are calculated using a reference plane characteristic scheme with the viscous terms treated as source terms. Several internal, external, and internal/external flow calculations are presented.

  20. Computation of high Reynolds number internal/external flows

    International Nuclear Information System (INIS)

    Cline, M.C.; Wilmoth, R.G.

    1981-01-01

    A general, user oriented computer program, called VNAP2, has been developed to calculate high Reynolds number, internal/external flows. VNAP2 solves the two-dimensional, time-dependent Navier-Stokes equations. The turbulence is modeled with either a mixing-length, a one transport equation, or a two transport equation model. Interior grid points are computed using the explicit MacCormack scheme with special procedures to speed up the calculation in the fine grid. All boundary conditions are calculated using a reference plane characteristic scheme with the viscous terms treated as source terms. Several internal, external, and internal/external flow calculations are presented