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

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)

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)

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.

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

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.

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.

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

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)

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)

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

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)

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)

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

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

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)

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

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)

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.

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

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.

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)

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

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)

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

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

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)

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.

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

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.

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.

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)

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

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

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.

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.

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.

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

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

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)

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)

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

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

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

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)

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)

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.

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.

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

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

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

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)

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

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.

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

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

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

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.

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

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

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…

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

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.)

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.)

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

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.)

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)

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

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

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.

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

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

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

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)

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

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

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

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

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)

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.

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.

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

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)

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

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

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.

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)

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

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.

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

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.

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)

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

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.

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)

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)

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)

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

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

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

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

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

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.

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)

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.

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

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

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

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)

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

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.

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)

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...

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...

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

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.

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

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

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.

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)

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

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.

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.

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

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%

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)

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)

Mathematical model of voltage-current characteristics of Bi(2223)/Ag magnets under an external magnetic field

CERN Document Server

Pitel, J; Lehtonen, J; Kovács, P

2002-01-01

We have developed a mathematical model, which enables us to predict the voltage-current V(I) characteristics of a solenoidal high-temperature superconductor (HTS) magnet subjected to an external magnetic field parallel to the magnet axis. The model takes into account the anisotropy in the critical current-magnetic field (I sub c (B)) characteristic and the n-value of Bi(2223)Ag multifilamentary tape at 20 K. From the power law between the electric field and the ratio of the operating and critical currents, the voltage on the magnet terminals is calculated by integrating the contributions of individual turns. The critical current of each turn, at given values of operating current and external magnetic field, is obtained by simple linear interpolation between the two suitable points of the I sub c (B) characteristic, which corresponds to the angle alpha between the vector of the resulting magnetic flux density and the broad tape face. In fact, the model is valid for any value and orientation of external magneti...

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.

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.

Differences between two definitions of the critical current of HTS coils

International Nuclear Information System (INIS)

Pitel, Jozef

2013-01-01

Definition of the critical current of a coil made of anisotropic high temperature superconducting conductor is rather complicated and ambiguous, since the magnetic field generated across the winding can differ considerably in relation to both its magnitude and orientation. Two definitions of the critical current of such coils are discussed. The first definition, very often used in calculations to analyze the current carrying capacity, electric field and power dissipation of individual turns, represents an operating current at which an electric field of 1 μV cm −1 appears on one turn. The second definition represents an integral approach, and is used in experiments. This definition introduces the critical current of the coil as an operating current at which an average electric field E s , usually 0.1 μV cm −1 , appears on coil terminals. As an example, the distribution of the critical current and electric field of individual turns in the winding of a BSCCO model coil was analyzed. Critical currents of the coil as a function of an external magnetic field parallel with the coil axis were calculated according to both definitions. The results show that the first definition, which characterizes the winding at the local level, is suitable for HTS coils either operating in self-field or in a low external field, because the differences between the critical currents and n-indices of individual turns are considerable. The second criterion is suitable for the HTS coils operating in high fields, i.e. like high field insert coils. The self-field of a high field insert coil is negligible if the external field is high. As a result, the critical currents of all turns are almost identical, and the anisotropy in I c (B) characteristic plays practically no role. Rather unexpected behavior of the voltage–current characteristic of the model coil is predicted if an external field is applied. (paper)

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

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.

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)

Critical behavior of mean-field hadronic models for warm nuclear matter

International Nuclear Information System (INIS)

Silva, J.B.; Lourenco, O.; Delfino, A.; Martins, J.S. Sa; Dutra, M.

2008-01-01

We study a set of hadronic mean-field models in the liquid-gas phase transition regime and calculate their critical parameters. The discussion is unified by scaling the coexistence curves in terms of these critical parameters. We study the models close to spinodal points, where they also present critical behavior. Inspired by signals of criticality shown in fragmentation experiments, we analyze two different scenarios in which such behavior would be expected: (i) the stability limits of a metastable system with vanishing external pressure; and (ii) the critical point of a gas-liquid phase equilibrium system for which the Maxwell construction applies. Spinodal and coexistence curves show the regions in which model dependence arises. Unexpectedly, this model dependence does not manifest if one calculates the thermal incompressibility of the models

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.

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)

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)

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.

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

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.

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

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 ( .

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

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).

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

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

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)

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

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.

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.

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

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)

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

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

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

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.

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

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

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

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

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)

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

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)

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)

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)

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)

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.

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.

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)

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)

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.

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

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

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

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.

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.

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

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

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)

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

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)

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

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

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

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

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

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...

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.

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.)

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

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

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

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...

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.)

A method to combine three dimensional dose distributions for external beam and brachytherapy radiation treatments for gynecological neoplasms

International Nuclear Information System (INIS)

Narayana, V.; Sahijdak, W.M.; Orton, C.G.

1997-01-01

Purpose: Radiation treatment of gynecological neoplasms, such as cervical carcinoma, usually combines external radiation therapy with one or more intracavitary brachytherapy applications. Although the dose from external beam radiation therapy and brachytherapy can be calculated and displayed in 3D individually, the dose distributions are not combined. At most, combined point doses are calculated for select points using various time-dose models. In this study, we present a methodology to combine external beam and brachytherapy treatments for gynecological neoplasms. Material and Methods: Three dimensional bio-effect treatment planning to obtain complication probability has been outlined. CT scans of the patient's pelvis with the gynecological applicator in place are used to outline normal tissue and tumor volumes. 3D external beam and brachytherapy treatment plans are developed separately and an external beam dose matrix and a brachytherapy dose matrix was calculated. The dose in each voxel was assumed to be homogeneous. The physical dose in each voxel of the dose matrix was then converted into extrapolated response dose (ERD) based on the linear quadratic model that accounts for the dose per fraction, number of fractions, dose rate, and complete or incomplete repair of sublethal damage (time between fractions). The net biological dose delivered was obtained by summing the ERD grids from external beam and brachytherapy since there was complete repair of sublethal damage between external beam and brachytherapy treatments. The normal tissue complication probability and tumor control probability were obtained using the biological dose matrix based on the critical element model. Results: The outlined method of combining external beam and brachytherapy treatments was implemented on gynecological treatments using an applicator for brachytherapy treatments. Conclusion: Implementation of the biological dose calculation that combine different modalities is extremely useful

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)

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)

Externalities of fuel cycles 'ExternE' project. Coal fuel cycle. Estimation of physical impacts and monetary valuation for priority impact pathways

International Nuclear Information System (INIS)

Berry, J.E.; Holland, M.R.; Watkiss, P.R.

1994-01-01

Background to the ExternE Project 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

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)

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

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

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)

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.

Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

International Nuclear Information System (INIS)

Anspaugh, L.R.; Church, B.W.

1986-01-01

In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, UT; Ely, NV; and Las Vegas, NV. Three events, HARRY (19 May 1953), BEE (22 March 1955), and SMOKY (31 August 1957), accounted for more than half the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of infinite exposure, estimated exposure, and 1-yr effective biological exposure are explained

Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

International Nuclear Information System (INIS)

Anspaugh, L.R.; Church, B.W.

1985-12-01

In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, Utah; Ely, Nevada; and Las Vegas, Nevada. Three events, HARRY (May 19, 1953), BEE (March 22, 1955), and SMOKY (August 31, 1957), accounted for over half of the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of ''infinite exposure,'' ''estimated exposure,'' and ''one year effective biological exposure'' are explained. 4 figs., 7 tabs

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

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)

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

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.

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)

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

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

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.)

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)

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)

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.

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.

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

Research on neutron source multiplication method in nuclear critical safety

International Nuclear Information System (INIS)

Zhu Qingfu; Shi Yongqian; Hu Dingsheng

2005-01-01

The paper concerns in the neutron source multiplication method research in nuclear critical safety. Based on the neutron diffusion equation with external neutron source the effective sub-critical multiplication factor k s is deduced, and k s is different to the effective neutron multiplication factor k eff in the case of sub-critical system with external neutron source. The verification experiment on the sub-critical system indicates that the parameter measured with neutron source multiplication method is k s , and k s is related to the external neutron source position in sub-critical system and external neutron source spectrum. The relation between k s and k eff and the effect of them on nuclear critical safety is discussed. (author)

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)

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)

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.

Catalog and history of the experiments of criticality Saclay (1958-1964) Valduc / Building 10 (1964-2003)

International Nuclear Information System (INIS)

Poullot, G.; Dumont, V.; Anno, J.; Cousinou, P.; Grivot, P.; Girault, E.; Fouillaud, P.; Barbry, F.

2003-01-01

The group ' International Criticality Safety Evaluation Benchmark evaluation project ' (I.C.S.B.E.P.) has for aim to supply to the international community experiments of benchmarks criticality, of certified quality, used to guarantee the qualification of criticality calculation codes. Have been defined: a structure of experiments classification, a format of standard presentation, a structure of work with evaluation, internal and external checks, presentation in plenary session. After favourable opinion of the work group, the synthesis document called evaluation is integrated to the general report I.C.S.B.E.P. (N.C.)

Critical group doses arising from routine aquatic discharges of activity from the Heysham 2 nuclear power station

International Nuclear Information System (INIS)

Maul, P.R.

1986-07-01

An assessment of critical group doses arising from routine discharges of activity to sea from the Heysham 2 nuclear power station has been undertaken using the CODAR2 computer program. The largest critical group dose rate was calculated to be 50 μSv/y, about 80 per cent of which arose from the external exposure of individuals occupying intertidal sediments of the Lune Estuary from the single nuclide 60 Co. CODAR2 employs a single well mixed local compartment for critical group calculations and can not include the effects of processes varying over the compartment. A more detailed study of the critical group exposure has been carried out using recently developed methods to assess the uncertainties and pessimisms involved in the CODAR2 calculations. It is concluded that these calculations are pessimistic by about a factor of 2, with a reference calculation using the more detailed methods giving a dose rate of 23 μSv/y for the Lune Estuary group. Further reductions in the dose estimate might be possible if site specific measurements of the concentration factor for Co on estuarine sediments and the sedimentation rate in the Lune Estuary could be made. (author)

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

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.

Historical estimates of external gamma exposure and collective external gamma exposure from testing at the Nevada Test Site. I. Test series through HARDTACK II, 1958

Energy Technology Data Exchange (ETDEWEB)

Anspaugh, L.R.; Church, B.W.

1985-12-01

In 1959, the Test Manager's Committee to Establish Fallout Doses calculated estimated external gamma exposure at populated locations based upon measurements of external gamma-exposure rate. Using these calculations and estimates of population, we have tabulated the collective estimated external gamma exposures for communities within established fallout patterns. The total collective estimated external gamma exposure is 85,000 person-R. The greatest collective exposures occurred in three general areas: Saint George, Utah; Ely, Nevada; and Las Vegas, Nevada. Three events, HARRY (May 19, 1953), BEE (March 22, 1955), and SMOKY (August 31, 1957), accounted for over half of the total collective estimated external gamma exposure. The bases of the calculational models for external gamma exposure of ''infinite exposure,'' ''estimated exposure,'' and ''one year effective biological exposure'' are explained. 4 figs., 7 tabs.

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)

Surface critical behavior and scaling functions for the three-dimensional mean spherical model

Energy Technology Data Exchange (ETDEWEB)

Amin, Magdy E. [Mathematics Department, Ar' ar Teacher College, Kingdom of Saudi Arabia (Saudi Arabia) and Mathematics Department, Faculty of Science, Minia University (Egypt)]. E-mail: aminmagdy@yahoo.com

2006-10-09

The d-dimensional mean spherical model on a fully finite L{sup d} simple cubic lattice with Neumann-Dirichlet boundary conditions is considered in the presence of a surface external fields acting at the surfaces bounding the system. Exact calculations are evaluated for the fully finite system and in the case of a film geometry Lx{approx}{sup d-1}. Critical finite-size scaling functions both for the specific heat and the mean-square magnetization are derived and investigated close to and below the bulk critical temperature K{sub c}.

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.

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

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.)

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....

Behaviour of the order parameter of the simple magnet in an external field

Directory of Open Access Journals (Sweden)

M.P.Kozlovskii

2005-01-01

Full Text Available The effect of a homogeneous external field on the three-dimensional uniaxial magnet behaviour near the critical point is investigated within the framework of the nonperturbative collective variables method using the ρ4 model. The research is carried out for the low-temperature region. The analytic explicit expressions for the free energy, average spin moment and susceptibility are obtained for weak and strong fields in comparison with the field value belonging to the pseudocritical line. The calculations are performed on the microscopic level without any adjusting parameters. It is established that the long-wave fluctuations of the order parameter play a crucial role in forming a crossover between the temperature-dependence and field-dependence critical behaviour of the system.

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).

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

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)

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

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)

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

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)

The Analytical Potential Energy Function of NH Radical Molecule in External Electric Field

International Nuclear Information System (INIS)

Wu Dong-Lan; Tan Bin; Wan Hui-Jun; Xie An-Dong; Ding Da-Jun

2015-01-01

The geometric structures of an NH radical in different external electric fields are optimized by using the density functional B3P86/cc-PV5Z method, and the bond lengths, dipole moments, vibration frequencies and IR spectrum are obtained. The potential energy curves are gained by the CCSD (T) method with the same basis set. These results indicate that the physical property parameters and potential energy curves may change with the external electric field, especially in the reverse direction electric field. The potential energy function of zero field is fitted by the Morse potential, and the fitting parameters are in good accordance with the experimental data. The potential energy functions of different external electric fields are fitted adopting the constructed potential model. The fitted critical dissociation electric parameters are shown to be consistent with the numerical calculation, and the relative errors are only 0.27% and 6.61%, hence the constructed model is reliable and accurate. The present results provide an important reference for further study of the molecular spectrum, dynamics and molecular cooling with Stark effect. (paper)

Externalities of fuel cycles 'ExternE' project. Wind fuel cycle. Estimation of physical impacts and monetary valuation for priority impact pathways

International Nuclear Information System (INIS)

Eyre, N.

1994-01-01

electricity use, although they are of a somewhat preliminary nature. The task of calculating the externalities is rendered difficult by a range of problems. It is particularly important for renewables that the site and project specificity of many impacts are recognised. In consequence of this, the ExternE Project of Directorate General XII (Science, Research and Technology) of the European Commission has been established to identify the most appropriate methodology for this type of work. The current programme has 4 principal objectives: 1) To create a unified conceptual design for quantifying the various social costs associated with the production and consumption of energy from different fuel sources. 2) To demonstrate an accounting framework that can be used to estimate the external costs that result from the incremental use of different fuel types. 3) To use this information in an international comparative analysis of fuel types. 4) To identify critical methodological issues and information needs. The objective of this particular study is to develop the methodology for wind energy and to implement it at two sites in the United Kingdom. Modern wind turbines bear little resemblance to the medieval designs of windmills characteristic of many European landscapes. The aerodynamic systems have been developed using aviation technology. Current designs typically have two or three blades, manufactured from glass reinforced elastic or wood epoxy, and rotate at 0.5 to 0.8 Hz. The most cost effective sizes currently have rotor diameters of approximately 30 metres, giving an output of a few hundreds of kW. The commercial wind energy programme in the UK is relatively recent, having been stimulated in England and Wales through the mechanism of the Non-Fossil Fuel Obligation, originally designed to subsidise nuclear electricity production. There is now well over 100 MW of capacity in place, mostly in wind farms situated in hill or coastal areas where mean wind speeds are in excess of 7m

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

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.

Cervix Regression and Motion During the Course of External Beam Chemoradiation for Cervical Cancer

International Nuclear Information System (INIS)

Beadle, Beth M.; Jhingran, Anuja; Salehpour, Mohammad; Sam, Marianne; Iyer, Revathy B.; Eifel, Patricia J.

2009-01-01

Purpose: To evaluate the magnitude of cervix regression and motion during external beam chemoradiation for cervical cancer. Methods and Materials: Sixteen patients with cervical cancer underwent computed tomography scanning before, weekly during, and after conventional chemoradiation. Cervix volumes were calculated to determine the extent of cervix regression. Changes in the center of mass and perimeter of the cervix between scans were used to determine the magnitude of cervix motion. Maximum cervix position changes were calculated for each patient, and mean maximum changes were calculated for the group. Results: Mean cervical volumes before and after 45 Gy of external beam irradiation were 97.0 and 31.9 cc, respectively; mean volume reduction was 62.3%. Mean maximum changes in the center of mass of the cervix were 2.1, 1.6, and 0.82 cm in the superior-inferior, anterior-posterior, and right-left lateral dimensions, respectively. Mean maximum changes in the perimeter of the cervix were 2.3 and 1.3 cm in the superior and inferior, 1.7 and 1.8 cm in the anterior and posterior, and 0.76 and 0.94 cm in the right and left lateral directions, respectively. Conclusions: Cervix regression and internal organ motion contribute to marked interfraction variations in the intrapelvic position of the cervical target in patients receiving chemoradiation for cervical cancer. Failure to take these variations into account during the application of highly conformal external beam radiation techniques poses a theoretical risk of underdosing the target or overdosing adjacent critical structures

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.

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.

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

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

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

Critical currents and energy dissipation in Y1Ba2Cu3O7-δ films

International Nuclear Information System (INIS)

Windte, C.V.

1993-09-01

The critical current density j c of patterned YBaCuO thin film samples and the electric field E seen by the superconductor were calculated from data of magnetization loops, external magnetic field sweeps and relaxation, respectively. Data were taken at 77 K up to external magnetic fields of 150 mT with a sensitive Hall probe detecting the magnetic fields of the screening currents of the sample. The E(j) characteristic could be determined over six decades of magnitude of E (10 -7 μV/cm -1 μV/cm). These characteristics permit the comparison with theoretical pictures: They are incompatible with Kim-Andersons flux-creep and a power law behaviour. The characteristics are compatible with the vortex-glass or the collective pinning ansatz. There exists a limiting critical current beyond no known theoretical picture describes the E(j) characteristic. (orig.) [de

Critical Dynamics : The Expansion of the Master Equation Including a Critical Point

NARCIS (Netherlands)

Dekker, H.

1980-01-01

In this thesis 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

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

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

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.

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)

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)

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.

THE USE OF THE FINITE DIFFERENCE METHOD FOR CALCULATION OF ELECTRONIC STATES IN MIS-STRUCTURE WITH SINGLE DONOR 1

Directory of Open Access Journals (Sweden)

E. A. Levchuk

2018-01-01

Full Text Available Numerical modeling of electronic state evolution due to non-uniform external electric field in the structure metal-insulator-semiconductor with solitary donor center is carried out. Considering a nanometer disc-shaped gate as a source of the electric field, the problem for the Laplace equation in multilayered medium is solved numerically to determine the distribution of the gate potential. The energy spectrum of a bound electron is calculated from the problem for the stationary Schrödinger equation. Finite difference schemes are constructed to solve both the problems. Difference scheme for the Schrödinger equation takes into account cusp condition for the wave function at the donor location. To solve the problem for the Laplace equation, asymptotic boundary conditions for approximating the external field potential at large distances from the gate in different layers are suggested. These conditions allow to reduce the calculation domain for the electrostatic problem essentially. The effect of the boundary conditions on the accuracy of calculating the potential and energies is investigated. Using the developed difference schemes, the dependences of the energy spectrum of the bound electron on the gate potential are calculated, and the values of critical potential at which the wave function of the electron is relocated are determined. It has been found on the basis of calculation results, that governing parameter for the description of electronic behavior is the potential difference between the donor and semiconductor surface. It has been shown that critical potential difference does not depend on dielectric thickness and permittivity.

Three-Dimensional Non-Fermi-Liquid Behavior from One-Dimensional Quantum Critical Local Moments

Science.gov (United States)

Classen, Laura; Zaliznyak, Igor; Tsvelik, Alexei M.

2018-04-01

We study the temperature dependence of the electrical resistivity in a system composed of critical spin chains interacting with three-dimensional conduction electrons and driven to criticality via an external magnetic field. The relevant experimental system is Yb2 Pt2 Pb , a metal where itinerant electrons coexist with localized moments of Yb ions which can be described in terms of effective S =1 /2 spins with a dominantly one-dimensional exchange interaction. The spin subsystem becomes critical in a relatively weak magnetic field, where it behaves like a Luttinger liquid. We theoretically examine a Kondo lattice with different effective space dimensionalities of the two interacting subsystems. We characterize the corresponding non-Fermi liquid behavior due to the spin criticality by calculating the electronic relaxation rate and the dc resistivity and establish its quasilinear temperature dependence.

Risk and sensitivity analysis in relation to external events

International Nuclear Information System (INIS)

Alzbutas, R.; Urbonas, R.; Augutis, J.

2001-01-01

This paper presents risk and sensitivity analysis of external events impacts on the safe operation in general and in particular the Ignalina Nuclear Power Plant safety systems. Analysis is based on the deterministic and probabilistic assumptions and assessment of the external hazards. The real statistic data are used as well as initial external event simulation. The preliminary screening criteria are applied. The analysis of external event impact on the NPP safe operation, assessment of the event occurrence, sensitivity analysis, and recommendations for safety improvements are performed for investigated external hazards. Such events as aircraft crash, extreme rains and winds, forest fire and flying parts of the turbine are analysed. The models are developed and probabilities are calculated. As an example for sensitivity analysis the model of aircraft impact is presented. The sensitivity analysis takes into account the uncertainty features raised by external event and its model. Even in case when the external events analysis show rather limited danger, the sensitivity analysis can determine the highest influence causes. These possible variations in future can be significant for safety level and risk based decisions. Calculations show that external events cannot significantly influence the safety level of the Ignalina NPP operation, however the events occurrence and propagation can be sufficiently uncertain.(author)

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)

External costs of electricity; Les couts externes de l'electricite

Energy Technology Data Exchange (ETDEWEB)

Rabl, A. [Ecole des Mines de Paris, 75 (France); Spadaro, J.V. [International Atomic Energy Agency (IAEA), Vienna (Austria)

2005-07-01

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.)

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

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

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.

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

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)

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)

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)

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)

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

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)

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)

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

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.

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.

Problems of quantum electrodynamics with external field creating pairs

International Nuclear Information System (INIS)

Fradkin, E.S.; Gitman, D.M.

1979-11-01

This paper is a preliminary version of a review of the results obtained by the authors and their collaborators which mainly concern problems of quantum electrodynamics with the pair-creating external field. In this paper the Furry picture is constructed for quantum electrodynamics with the pair-creating external field. It is shown, that various Green functions in the external field arise in the theory in a natural way. Special features of usage of the unitarity conditions for calculating the total probabilities of transitions are discussed. Perturbation theory for determining the mean electromagnetic field is constructed. Effective Lagrangians for pair-creating fields are built. One of the possible ways to introduce external field in quantum electrodynamics is considered. All the Green functions arising in the theory suggested are calculated for a constant field and a plane wave field. For the case of the electric field the total probability of creation of pairs from the vacuum accompanied by the photon irradiation and the total probability of transition from a single-electron state accompanied by the photon irradiation and creation of pairs are obtained by using the formulated rules for calculating the total probabilities of transitions. (author)

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)

Development of hybrid core calculation system using 2-D full-core heterogeneous transport calculation and 3-D advanced nodal calculation

International Nuclear Information System (INIS)

Sugimura, Naoki; Mori, Masaaki; Hijiya, Masayuki; Ushio, Tadashi; Arakawa, Yasushi

2004-01-01

This paper presents the Hybrid Core Calculation System which is a very rigorous but a practical calculation system applicable to best estimate core design calculations taking advantage of the recent remarkable progress of computers. The basic idea of this system is to generate the correction factors for assembly homogenized cross sections, discontinuity factors, etc. by comparing the CASMO-4 and SIMULATE-3 2-D core calculation results under the consistent calculation condition and then apply them for SIMULATE-3 3-D calculation. The CASMO-4 2-D heterogeneous core calculation is performed for each depletion step with the core conditions previously determined by ordinary SIMULATE-3 core calculation to avoid time consuming iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. The final SIMULATE-3 3-D calculation using the correction factors is performed with iterative calculations searching for the critical boron concentrations while treating the thermal hydraulic feedback. (author)

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

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

A new parallel algorithm for simulation of spin glasses on scales of space-time periods of external fields with consideration of relaxation effects

International Nuclear Information System (INIS)

Gevorkyan, A.S.; Abajyan, H.G.

2011-01-01

We have investigated the statistical properties of an ensemble of disordered 1D spatial spin chains (SSCs) of finite length, placed in an external field, with consideration of relaxation effects. The short-range interaction complex-classical Hamiltonian was first used for solving this problem. A system of recurrent equations is obtained on the nodes of the spin-chain lattice. An efficient mathematical algorithm is developed on the basis of these equations with consideration of the advanced Sylvester conditions which allow step by step construct a huge number of stable spin chains in parallel. The distribution functions of different parameters of spin-glass system are constructed from the first principles of the complex classical mechanics by analyzing the calculation results of the 1D SSCs ensemble. It is shown that the behavior of the parameter distributions is quite different depending on the external fields. The energy ensembles and constants of spin-spin interactions are changed smoothly depending on the external field in the limit of statistical equilibrium, while some of them such as the mean value of polarizations of ensemble and parameters of its orderings are frustrated. We have also studied some critical properties of the ensemble of such catastrophes in the Clausius-Mossotti equation depending on the value of the external field. We have shown that the generalized complex-classical approach excludes these catastrophes allowing one to organize continuous parallel computing on the whole region of values of the external field including critical points. A new representation of the partition function based on these investigations is suggested. As opposed to usual definition, this function is a complex one and its derivatives are everywhere defined, including critical points

External legitimation in international new ventures

DEFF Research Database (Denmark)

Turcan, Romeo V.

2012-01-01

This paper explores within the framework of new venture legitimation how and why international new ventures acquire external legitimacy and strive for survival in the face of critical events. Following a longitudinal multiple-case study methodology that was adopted for the purpose of theory...... building, the paper introduces the typology of captivity, and the four types that have emerged: captive industry supplier, captive dyadic partner, captive market leader, and free market leader. The effects of captivity types on the acquisition of external legitimacy and its survival, on reaching legitimacy...

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)

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)

Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

International Nuclear Information System (INIS)

Hwang, J. S.; Suh, K. Y.

2009-01-01

The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

External radiation exposure after deposition of man-made radionuclides

International Nuclear Information System (INIS)

Jacob, P.

1991-01-01

The first step in assessing the external radiation exposure of the population is the determination of the gamma dose rate over meadows, which are used as reference points for various reasons. The second step is the description of external radiation exposures in urban and rural environments. The relation to the radiation exposure in a meadow is a function of the radionuclide distribution, i.e. the type of deposition. Finally, a simple method of calculating external radiation exposure is developed on the basis of recent findings. The method is compared with the method used in the UNSCEAR report for calculating radiation exposures after Chernobyl and with the method described in the AVV (General Administrative Regulation) of the Radiation Protection Ordinance. (orig./HP) [de

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.

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

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)

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.)

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

14 CFR 29.865 - External loads.

Science.gov (United States)

2010-01-01

... 200 volts per meter. (iii) Be protected against any failure that could be induced by a failure mode of... operational envelope without hazard to the rotorcraft during normal flight conditions. In addition, these...-load combinations to be used for nonhuman external cargo except for the failure of critical structural...

Effects of mixing and stirring on the critical behaviour

International Nuclear Information System (INIS)

Antonov, N V; Hnatich, Michal; Honkonen, Juha

2006-01-01

Stochastic dynamics of a nonconserved scalar order parameter near its critical point, subject to random stirring and mixing, is studied using the field-theoretic renormalization group. The stirring and mixing are modelled by a random external Gaussian noise with the correlation function ∼δ(t - t')k 4-d-y and the divergence-free (due to incompressibility) velocity field, governed by the stochastic Navier-Stokes equation with a random Gaussian force with the correlation function ∝ δ(t-t')k 4-d-y' . Depending on the relations between the exponents y and y' and the space dimensionality d, the model reveals several types of scaling regimes. Some of them are well known (model A of equilibrium critical dynamics and linear passive scalar field advected by a random turbulent flow), but there are three new non-equilibrium regimes (universality classes) associated with new nontrivial fixed points of the renormalization group equations. The corresponding critical dimensions are calculated in the two-loop approximation (second order of the triple expansion in y, y' and ε = 4 - d)

Criticality studies; Etudes de criticite

Energy Technology Data Exchange (ETDEWEB)

Breton, D; Lecorche, P; Clouet d' Orval, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Criticality studies made at the Commissariat a l'Energie atomique deal on the one hand with experiments on plutonium and uranium solutions, on the other hand with theoretical work on the development and use of computation, methods for the resolution of problems concerning the nuclear safety of chemical and metallurgical plants. I - Since 1958 the experimental studies have dealt with homogeneous media constituted by a fissile salt dissolved in light water. Developed using the reactor Proserpine, the experiments have been carried on at Saclay on the Alecto assemblies where solutions of plutonium or of 90 p.100 - enriched uranium can be made critical. The results already obtained relate to critical masses of cylindrical tanks of diameters from 20 to 50 cm. reflected in several ways (water, concrete, etc. . ) at concentrations up to 100 g/liter. Physical measurements (spectra, reactor noises) and interaction measurements complete the results. Other experiments relating to plutonium solutions were begun in 1963, at the Valduc Center. They deal with the study of critical masses of annular vessels of external diameter 50 cm and internal diameter varying from 10 to 30 cm. These vessels can be water reflected internally, externally, or both. Two of these vessels have been studied in interaction for various geometries. Slabs of various thicknesses were also studied. II - The studies thus undertaken allowed the development of methods of computation which have been tested on several experiments. Particular use has been made of the possibilities of calculations based on transport theory and on Monte Carlo methods. All these theoretical studies are applied to the design and control of industrial plants from the point of view of safety. (authors) [French] Les etudes de criticite effectuees au CEA comportent d'une part des experiences sur des solutions de plutonium et d'uranium enrichi, d'autre part des travaux theoriques portant sur la mise au point et l'exploitation de methodes

A Criticality Safety Study on Storing Unirradiated Cintichem-Type Targets at Sandia National Laboratories

International Nuclear Information System (INIS)

Romero, D.J.; Parma, E.J.; Busch, R.D.

1999-01-01

This criticality safety analysis is performed to determine the effective multiplication factor (k eff ) for a storage cabinet filled with unirradiated Cintichem-type targets. These targets will be used to produce 99 Mo at Sandia National Laboratories and will be stored on-site prior to irradiation in the Annular Core Research Reactor. The analysis consisted of using the Monte Carlo code MCNP (Version 4A) to model and predict the k eff for the proposed dry storage configuration under credible loss of geometry and moderator control. Effects of target pitch, non-uniform loading, and target internal/external flooding are evaluated. Further studies were done with deterministic methods to verify the results obtained from MCNP and to obtain a clearer understanding of the parameters affecting system criticality. The diffusion accelerated neutral particle transport code ONEDANT was used to model the target in a one-dimensional, infinite half-slab geometry and determine the critical slab thickness. Hand calculations were also completed to determine the critical slab thickness with modified one-group, and one-group, two region approximations. Results obtained from ONEDANT and the hand calculations were compared to applicable cases in a commonly used criticality safety analysis handbook. Overall, the critical slab thicknesses obtained in the deterministic analysis were much larger than the dimensions of the cabinet and further support the predictions by MCNP that a critical system cannot be attained for the base case or in conditions where loss of geometry and moderation control occur

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.

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)

Energy dependence of critical state of single-component systems

International Nuclear Information System (INIS)

Volchenkova, R.A.

1985-01-01

Equations of critical states of the single-component systems: Psub(cr)(/Psub(o)=(Tsub(cr)/Tsub(o))x0.73, Tsub(cr)=K(Tsub(boil))sup(1.116) and Hsub(cr)(/Hsub(B)=Tsub(sr)/Tsub(B))sup(1.48) where Tsub(B)=1K, Hsub(B)-2 kcal/g-at, K-dimension factor are presented. It is shown that the revealed dependence Hsub(cr)=H(Tsub(cr)) is an energy boundary of a liquid-vapour phase state of the single-component systems beyond limits of which difference between liquid and vapour phases vanishes in increasing the system energy content. The given equations of state are true for all the single-component systems and permit to consider physicomechanical properties of substances in dynamic state depending on external conditions. Critical temperatures and dependences for elements from the most fusible He to infusible W and Re have been calculated

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)

A comparative study of formal methods for safety critical software in nuclear power plant

International Nuclear Information System (INIS)

Sohn, Se Do; Seong Poong Hyun

2000-01-01

The requirement of ultra high reliability of the safety critical software can not be demonstrated by testing alone. The specification based on formal method is recommended for safety system software. But there exist various kinds of formal methods, and this variety of formal method is recognized as an obstacle to the wide use of formal method. In this paper six different formal method have been applied to the same part of the functional requirements that is calculation algorithm intensive. The specification results were compared against the criteria that is derived from the characteristics that good software requirements specifications should have and regulatory body recommends to have. The application experience shows that the critical characteristics should be defined first, then appropriate method has to e selected. In our case, the Software Cost Reduction method was recommended for internal condition or calculation algorithm checking, and state chart method is recommended for the external behavioral description. (author)

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)

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)

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

External parallel sorting with multiprocessor computers

International Nuclear Information System (INIS)

Comanceau, S.I.

1984-01-01

This article describes methods of external sorting in which the entire main computer memory is used for the internal sorting of entries, forming out of them sorted segments of the greatest possible size, and outputting them to external memories. The obtained segments are merged into larger segments until all entries form one ordered segment. The described methods are suitable for sequential files stored on magnetic tape. The needs of the sorting algorithm can be met by using the relatively slow peripheral storage devices (e.g., tapes, disks, drums). The efficiency of the external sorting methods is determined by calculating the total sorting time as a function of the number of entries to be sorted and the number of parallel processors participating in the sorting process

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.

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

Method for calculating individual equivalent doses and cumulative dose of population in the vicinity of nuclear power plant site

International Nuclear Information System (INIS)

Namestek, L.; Khorvat, D; Shvets, J.; Kunz, Eh.

1976-01-01

A method of calculating the doses of external and internal person irradiation in the nuclear power plant vicinity under conditions of normal operation and accident situations has been described. The main difference between the above method and methods used up to now is the use of a new antropomorphous representation of a human body model together with all the organs. The antropomorphous model of human body and its organs is determined as a set of simple solids, coordinates of disposistion of the solids, sizes, masses, densities and composition corresponding the genuine organs. The use of the Monte-Carlo method is the second difference. The results of the calculations according to the model suggested can be used for determination: a critical group of inhabitans under conditions of normal plant operation; groups of inhabitants most subjected to irradiation in the case of possible accident; a critical sector with a maximum collective dose in the case of an accident; a critical radioisotope favouring the greatest contribution to an individual equivalent dose; critical irradiation ways promoting a maximum contribution to individual equivalent doses; cumulative collective doses for the whole region or for a chosen part of the region permitting to estimate a population dose. The consequent method evoluation suggests the development of separate units of the calculationg program, critical application and the selection of input data of physical, plysiological and ecological character and improvement of the calculated program for the separate concrete events [ru

Metallic magnets without inversion symmetry and antiferromagnetic quantum critical points

Energy Technology Data Exchange (ETDEWEB)

Fischer, I.A.

2006-07-01

This thesis focusses on two classes of systems that exhibit non-Fermi liquid behaviour in experiments: we investigated aspects of chiral ferromagnets and of antiferromagnetic metals close to a quantum critical point. In chiral ferromagnets, the absence of inversion symmetry makes spin-orbit coupling possible, which leads to a helical modulation of the ferromagnetically ordered state. We studied the motion of electrons in the magnetically ordered state of a metal without inversion symmetry by calculating their generic band-structure. We found that spin-orbit coupling, although weak, has a profound effect on the shape of the Fermi surface: On a large portion of the Fermi surface the electron motion parallel to the helix practically stops. Signatures of this effect can be expected to show up in measurements of the anomalous Hall effect. Recent neutron scattering experiments uncovered the existence of a peculiar kind of partial order in a region of the phase diagram adjacent to the ordered state of the chiral ferromagnet MnSi. Starting from the premise that this partially ordered state is a thermodynamically distinct phase, we investigated an extended Ginzburg-Landau theory for chiral ferromagnets. In a certain parameter regime of the Ginzburg-Landau theory we identified crystalline phases that are reminiscent of the so-called blue phases in liquid crystals. Many antiferromagnetic heavy-fermion systems can be tuned into a regime where they exhibit non-Fermi liquid exponents in the temperature dependence of thermodynamic quantities such as the specific heat capacity; this behaviour could be due to a quantum critical point. If the quantum critical behaviour is field-induced, the external field does not only suppress antiferromagnetism but also induces spin precession and thereby influences the dynamics of the order parameter. We investigated the quantum critical behavior of clean antiferromagnetic metals subject to a static, spatially uniform external magnetic field. We

Years of life lost due to external radiation exposure

International Nuclear Information System (INIS)

Raicevic, J.J.; Merkle, J.M.; Ehrhardt, J.; Ninkovic, M.M.

2002-01-01

A new approach for calculation of the years of life lost per excess death (YLL) due to stochastic health effects is applied to external exposure pathways. The short-term external exposures are due to the passage of radioactive cloud (CL) and due to the skin and clothes contamination (SK). The long-term external exposure is the one from the radioactive material deposited on ground (GR). Three nuclides, 131 I , 137 Cs and 239 Pu with extremely wide range of the half-life are considered to examine its possible influence on the calculated YLL values. For each of these nuclides, the YLL is found as a decreasing function of the age at exposure and presented graphically in this paper. Another negative correlation is established between the fully averaged YLL and the duration of the nuclide's half-life has been found for protracted exposure (GR). On the other hand, the YLL for the short-term external exposures (CL and SK) practically does not depend on the nuclide's half-life. In addition, a weak YLL dependence of the dose was commented. (author)

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

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.

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.

Ensuring the validity of calculated subcritical limits

International Nuclear Information System (INIS)

Clark, H.K.

1977-01-01

The care taken at the Savannah River Laboratory and Plant to ensure the validity of calculated subcritical limits is described. Close attention is given to ANSI N16.1-1975, ''Validation of Calculational Methods for Nuclear Criticality Safety.'' The computer codes used for criticality safety computations, which are listed and are briefly described, have been placed in the SRL JOSHUA system to facilitate calculation and to reduce input errors. A driver module, KOKO, simplifies and standardizes input and links the codes together in various ways. For any criticality safety evaluation, correlations of the calculational methods are made with experiment to establish bias. Occasionally subcritical experiments are performed expressly to provide benchmarks. Calculated subcritical limits contain an adequate but not excessive margin to allow for uncertainty in the bias. The final step in any criticality safety evaluation is the writing of a report describing the calculations and justifying the margin

A new external microbeam system in Fudan University

International Nuclear Information System (INIS)

Zheng, Y.; Shen, H.; Li, Y.Q.; Li, X.Y.; Yang, M.J.; Mi, Y.

2013-01-01

A cost-effective and removable external beam system is set up based on the Oxford Microbeam system in Fudan University. In our external beam system, 7.5-μm-thick Kapton film is used as exit window with a diameter of 3.5 mm. The spatial resolution is about 18 μm full width at half maximum (FWHM) on a copper grid. As an example, calcium distribution in otolith is present by the external micro-PIXE. In addition, little change can be done to the external system mentioned above for radiobiology experiments. The exit window can be changed from the focal plane to the observation window of vacuum chamber. By calculation, the beam spot size can reach less than 30 μm. Since the Oxford type octagonal target chamber is popular among the nuclear microprobe facilities, this method can be provided to easily replace the in-vacuum system with the external system, extending the in-vacuum analysis to external beam analysis

A new external microbeam system in Fudan University

Energy Technology Data Exchange (ETDEWEB)

Zheng, Y.; Shen, H., E-mail: haoshen@fudan.edu.cn; Li, Y.Q.; Li, X.Y.; Yang, M.J.; Mi, Y.

2013-07-01

A cost-effective and removable external beam system is set up based on the Oxford Microbeam system in Fudan University. In our external beam system, 7.5-μm-thick Kapton film is used as exit window with a diameter of 3.5 mm. The spatial resolution is about 18 μm full width at half maximum (FWHM) on a copper grid. As an example, calcium distribution in otolith is present by the external micro-PIXE. In addition, little change can be done to the external system mentioned above for radiobiology experiments. The exit window can be changed from the focal plane to the observation window of vacuum chamber. By calculation, the beam spot size can reach less than 30 μm. Since the Oxford type octagonal target chamber is popular among the nuclear microprobe facilities, this method can be provided to easily replace the in-vacuum system with the external system, extending the in-vacuum analysis to external beam analysis.

A new external microbeam system in Fudan University

Energy Technology Data Exchange (ETDEWEB)

Zheng, Y. [Applied Ion Beam Physics Laboratory, Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Shen, H., E-mail: haoshen@fudan.edu.cn [Applied Ion Beam Physics Laboratory, Institute of Modern Physics, Fudan University, Shanghai 200433 (China); Li, Y.Q.; Li, X.Y.; Yang, M.J.; Mi, Y. [Applied Ion Beam Physics Laboratory, Institute of Modern Physics, Fudan University, Shanghai 200433 (China)

2013-02-01

A cost-effective and removable external beam system is set up based on the Oxford Microbeam system in Fudan University. In our external beam system, 7.5-μm-thick Kapton film is used as exit window with a diameter of 3.5 mm. The spatial resolution is about 18 μm full width at half maximum (FWHM) on a copper grid. As an example, calcium distribution in otolith is present by the external micro-PIXE. In addition, little change can be done to the external system mentioned above for radiobiology experiments. The exit window can be changed from the focal plane to the observation window of vacuum chamber. By calculation, the beam spot size can reach less than 30 μm. Since the Oxford type octagonal target chamber is popular among the nuclear microprobe facilities, this method can be provided to easily replace the in-vacuum system with the external system, extending the in-vacuum analysis to external beam analysis.

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%

Performance Targets and External Market Prices

DEFF Research Database (Denmark)

Hansen, Allan; Friis, Ivar; Vámosi, Tamás S.

2012-01-01

the implementation of external market information in target setting – well known in transfer pricing, relative performance evaluation, beyond budgeting, target costing, piece rates systems and value based management – relate to challenging motivation and information problem. The analysis and discussion of those...... problems, in particular those related to accounting for the internal performance (that are going to be compared with the external target), calculating the ‘inside’ costs and defining controllability, contributes to the management accounting as well as the piece-rate literature.......In this paper we explore the processes of ‘bringing the market inside the firm’ to set performance targets and benchmark production workers productivity. We analyze attempts to use external suppliers’ bids in target setting in a Danish manufacturing company. The case study illustrates how...

Critical enrichment and critical density of infinite systems for nuclear criticality safety evaluation

International Nuclear Information System (INIS)

Naito, Yoshitaka; Koyama, Takashi; Komuro, Yuichi

1986-03-01

Critical enrichment and critical density of homogenous infinite systems, such as U-H 2 O, UO 2 -H 2 O, UO 2 F 2 aqueous solution, UO 2 (NO 3 ) 2 aqueous solution, Pu-H 2 O, PuO 2 -H 2 O, Pu(NO 3 ) 4 aqueous solution and PuO 2 ·UO 2 -H 2 O, were calculated with the criticality safety evaluation computer code system JACS for nuclear criticality safety evaluation on fuel facilities. The computed results were compared with the data described in European and American criticality handbooks and showed good agreement with each other. (author)

Inverse kinetics for subcritical systems with external neutron source

International Nuclear Information System (INIS)

Carvalho Gonçalves, Wemerson de; Martinez, Aquilino Senra; Carvalho da Silva, Fernando

2017-01-01

Highlights: • It was developed formalism for reactivity calculation. • The importance function is related to the system subcriticality. • The importance function is also related with the value of the external source. • The equations were analyzed for seven different levels of sub criticality. • The results are physically consistent with others formalism discussed in the paper. - Abstract: Nuclear reactor reactivity is one of the most important properties since it is directly related to the reactor control during the power operation. This reactivity is influenced by the neutron behavior in the reactor core. The time-dependent neutrons behavior in response to any change in material composition is important for the reactor operation safety. Transient changes may occur during the reactor startup or shutdown and due to accidental disturbances of the reactor operation. Therefore, it is very important to predict the time-dependent neutron behavior population induced by changes in neutron multiplication. Reactivity determination in subcritical systems driven by an external neutron source can be obtained through the solution of the inverse kinetics equation for subcritical nuclear reactors. The main purpose of this paper is to find the solution of the inverse kinetics equation the main purpose of this paper is to device the inverse kinetics equations for subcritical systems based in a previous paper published by the authors (Gonçalves et al., 2015) and by (Gandini and Salvatores, 2002; Dulla et al., 2006). The solutions of those equations were also obtained. Formulations presented in this paper were tested for seven different values of k eff with external neutrons source constant in time and for a powers ratio varying exponentially over time.

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

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

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.

MCNPTM criticality primer and training experiences

International Nuclear Information System (INIS)

Briesmeister, J.; Forster, R.A.; Busch, R.

1995-01-01

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, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (k eff ). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple k eff calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP

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

Critical point predication device

International Nuclear Information System (INIS)

Matsumura, Kazuhiko; Kariyama, Koji.

1996-01-01

An operation for predicting a critical point by using a existent reverse multiplication method has been complicated, and an effective multiplication factor could not be plotted directly to degrade the accuracy for the prediction. The present invention comprises a detector counting memory section for memorizing the counting sent from a power detector which monitors the reactor power, a reverse multiplication factor calculation section for calculating the reverse multiplication factor based on initial countings and current countings of the power detector, and a critical point prediction section for predicting the criticality by the reverse multiplication method relative to effective multiplication factors corresponding to the state of the reactor core previously determined depending on the cases. In addition, a reactor core characteristic calculation section is added for analyzing an effective multiplication factor depending on the state of the reactor core. Then, if the margin up to the criticality is reduced to lower than a predetermined value during critical operation, an alarm is generated to stop the critical operation when generation of a period of more than a predetermined value predicted by succeeding critical operation. With such procedures, forecasting for the critical point can be easily predicted upon critical operation to greatly mitigate an operator's burden and improve handling for the operation. (N.H.)

Thermodynamic properties of open noncritical string in external electromagnetic field

International Nuclear Information System (INIS)

Lichtzier, I.M.; Odintsov, S.D.; Bytsenko, A.A.

1991-01-01

We investigate the thermodynamics of open noncritical string (charged and neutral) in an external constant magnetic field. The free energy and Hagedorn temperature are calculated. It is shown that Hagedorn temperature is the same as in the absence of constant magnetic field. We present also the expressions for the free energy and Hagedorn temperature of the neutral open noncritical string in an external constant electromagnetic field. In this case Hagedorn temperature depends on the external electric field. (author)

External and internal standards in the single-isotope derivative (radioenzymatic) measurement of plasma norepinephrine and epinephrine

International Nuclear Information System (INIS)

Shah, S.D.; Clutter, W.E.; Cryer, P.E.

1985-01-01

In plasma from normal humans (n = 9, 35 samples) and from patients with diabetes mellitus (n = 12, 24 samples) single-isotope derivative (radioenzymatic) plasma norepinephrine and epinephrine concentrations calculated from external standard curves constructed in a normal plasma pool were identical to those calculated from internal standards added to an aliquot of each plasma sample. In plasma from patients with end-stage renal failure receiving long-term dialysis (n = 34, 109 samples), competitive catechol-O-methyltransferase (COMT) inhibitory activity resulted in a systematic error when external standards in a normal plasma pool were used, as reported previously; values so calculated averaged 21% (+/- 12%, SD) lower than those calculated from internal standards. However, when external standard curves were constructed in plasma from a given patient with renal failure and used to calculate that patient's values, or in a renal failure plasma pool and used to calculate all renal failure values, norepinephrine and epinephrine concentrations were not significantly different from those calculated from internal standards. We conclude: (1) External standard curves constructed in plasma from a given patient with renal failure can be used to measure norepinephrine and epinephrine in plasma from that patient; further, external standards in a renal failure plasma pool can be used for assays in patients with end-stage renal failure receiving long-term dialysis. (2) Major COMT inhibitory activity is not present commonly if samples from patients with renal failure are excluded. Thus, it would appear that external standard curves constructed in normal plasma can be used to measure norepinephrine and epinephrine precisely in samples from persons who do not have renal failure

Analytical electromagnetic calculations for practical superconductors

International Nuclear Information System (INIS)

Tominaka, T

2008-01-01

The current distributions of twisted long filamentary composites are studied during the current sweep and under the transverse and longitudinal external field, using the inductance matrix among superconducting helical filaments and the inductive voltage between filaments and the external coil in the circuit equation. The self- and mutual inductances of helical conductors with a uniform helical current density are approximately calculated from the analytical expressions of long helical thin conductors. In addition, the magnetic field and the vector potential distributions of a twisted superconducting composite are also obtained by the numerical integral calculation over the cross section of the analytical expressions for the magnetic field and vector potential due to an infinitely long helical conductor

Davidson's Externalism and Swampman's Troublesome Biography

Directory of Open Access Journals (Sweden)

André Leclerc

2005-12-01

Full Text Available After the seminal works of Putnam (1975, Burge (1979, and Kripke (1982, the next important contribution to externalism is certainly Davidson’s (mainly 1987, 1988, 1989, 2001. By criticizing the posi-tions of these philosophers, Davidson elaborated his own brand of exter-nalism. We shall first present some features of Davidson’s externalism (the importance of historical-causal connections for the foundation of language and thought, for the explanation of how language can be learned, and how attitudes can be identified by the interpreter, and fi-nally how mental content is determined by appealing to the idea of trian-gulation, to prepare the discussion of a few problems. We then discuss two questions in Davidson’s externalism. First, how to reconcile the fact that external factors determine mental content, as Putnam, Burge and Davidson himself argued convincingly, with token-physicalism, the thesis that mental events are identical with physical events occurring “in the head” (or the thesis that mental events supervenes locally on brain ac-tivities? The second main problem is how to reconcile the first person authority with some prima facie consequences of externalism, mainly that we should know the relevant parts of our (natural and social envi-ronment in order to know the content of our own thoughts? We argue that Davidson’s answer to the first question is not successful, while his answer to the second was a breakthrough.

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

Years of life lost due to external radiation exposure

Directory of Open Access Journals (Sweden)

Raičević Jagoš J.

2004-01-01

Full Text Available In this paper a new approach for calculation of the years of life lost per excess death due to stochastic health effects is applied to external exposure pathways. The short-term external exposures are due to the passage of radioactive cloud and due to the skin and clothes contamination. The long-term external exposure is the one from the radioactive material deposited on the ground (groundshine. Three nuclides, 131I, 137Cs, and 239Pu, and with the extremely wide range of half-life are considered in order to examine their possible influence on the calculated values of years of life lost. For each of these nuclides, the number of years of life lost has been found as a decreasing function of the age at the expo sure and presented graphically in this paper. For protracted exposures, the fully averaged number of years of life lost is negative correlated with the nuclide’s half-life. On the other hand, the short-term external exposures do not depend on the nuclide’s half-life. In addition, a weak years of life lost dependence of the dose has been commented.

ExternE transport methodology for external cost evaluation of air pollution

DEFF Research Database (Denmark)

Jensen, S. S.; Berkowicz, R.; Brandt, J.

The report describes how the human exposure estimates based on NERI's human exposure modelling system (AirGIS) can improve the Danish data used for exposure factors in the ExternE Transport methodology. Initially, a brief description of the ExternE Tranport methodology is given and it is summarised...

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

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

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.

An external irradiation treatment planning system installed on a personal computer. 137

International Nuclear Information System (INIS)

Kunieda, Etso; Ogawa, Koichi; Mita, Kazumasa; Sekiguchi, Kozo; Wada, Tadashi; Hashimoto, Shyozo

1987-01-01

A compact and practical treatment planning system for external photon therapy has been developed for use on a desk-top personal computer. The system calculates the dose distributions of inhomogeneous density fields by using the CT-value of each pixel and displays isodose curves on the CRT superimposed on the gray scale CT image or on a hard-copy. Inhomogeneity correction is based on the TAR method where the path length from the calculation point to the surface is determined by summing up electron density derived from the CT-values of pixels on the path. Wedge filter correction is also available by using stored geometric data. The contour of patients is acquired by tracting the CT image on the light panel of the digitizer, or directly from the digital CT data. Though some critical parts of the programs are written in machine language, the system is mostly in BASIC and C languages. The minimum required hardware consists of a MS-DOS based personal computer, a color CRT display, an 8 inch floppy disk drive and a digitizer. They are generally available in Japan at reasonable cost. Tests were carried out in homogeneous and inhomogeneous density phantoms to evaluate the accuracy of the acquired dosage, and showed reasonable results compared with other commercially available treatment planning systems. The overall calculation time is satisfactory for multiple beam calculations. 5 refs.; 3 figs

External radiation exposure of the public

International Nuclear Information System (INIS)

Mehl, J.

1977-01-01

Results of several ten thousand measurements on external radiation (outside buildings, in living rooms) are used for illustrating by isodose charts covering the total area of the Federal Republic of Germany the exposure of the public from external radiation originating from natural radiation of the environment. Results of calculations on external radiation exposure of the public due to releases of radioactivity in air from nuclear installations are used for illustrating by coloured isodose charts the exposure of the public in the plant site vicinity. From comparison of the exposure levels it becomes obvious that if exposure levels of several 10 mrem per year are considered to be of real concern to public health, control of natural radoactivity in the environment of man would require more attention than present and foreseeable releases of radioactivity in air from nuclear inst

Experimental performance and results of the critical pebble bed facility KAHTER

Energy Technology Data Exchange (ETDEWEB)

Krings, F. J.; Drueke, V.; Kirch, N.; Neef, R. D.

1974-10-15

The paper provides a description and results of critical experiments performed in KAHTER fueled with pebbles containing coated particles of HEU/Th oxide with a ratio of uranium-to-thorium of 1.1:5. Core loadings with varying amounts of fuel and solid graphite pebbles were tested with fuel-to-graphite pebble ratios of 3:1, 1:1, and 1:3. Tests included criticality for various fuel loadings with all control rods removed, control rod worths for reflector-mounted control as single rods and in a bank and control worths for a central control rod, reaction rates by flux wire activations (Dy, Mn, In, Au, and U-235) and detector measurements (BF3 and fission chamber), simulated xenon stability testing using the motions of a Cf-252 source and Cd-absorber observed by an externally-located BF3 detector, and the reactivity worth of a Hf burnable absorber. For calculations of the room-temperature zero-power critical experiment, the values for nitrogen and hydrogen contents of the graphite were taken from previous experiments in CESAR.

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.

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.

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)

External cephalic version-related risks: a meta-analysis.

Science.gov (United States)

Grootscholten, Kim; Kok, Marjolein; Oei, S Guid; Mol, Ben W J; van der Post, Joris A

2008-11-01

To systematically review the literature on external cephalic version-related complications and to assess if the outcome of a version attempt is related to complications. In March 2007 we searched MEDLINE, EMBASE, and the Cochrane Central Register of Controlled Trials. Studies reporting on complications from an external cephalic version attempt for singleton breech pregnancies after 36 weeks of pregnancy were selected. We calculated odds ratios (ORs) from studies that reported both on complications as well as on the position of the fetus immediately after the procedure. We found 84 studies, reporting on 12,955 version attempts that reported on external cephalic version-related complications. The pooled complication rate was 6.1% (95% CI 4.7-7.8), 0.24% for serious complications (95% confidence interval [CI] 0.17-0.34) and 0.35% for emergency cesarean deliveries (95% CI 0.26-0.47). Complications were not related to external cephalic version outcome (OR 1.2 (95% CI 0.93-1.7). External cephalic version is a safe procedure. Complications are not related to the fetal position after external cephalic version.

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).

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).

LCEs for Naval Reactor Benchmark Calculations

International Nuclear Information System (INIS)

W.J. Anderson

1999-01-01

The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k eff ) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository

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)

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)

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.

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

Catalog and history of the experiments of criticality Saclay (1958-1964) Valduc / Building 10 (1964-2003); Catalogue et historique des experiences de criticite Saclay (1958 - 1964) Valduc / Batiment 10 (1964-2003)

Energy Technology Data Exchange (ETDEWEB)

Poullot, G.; Dumont, V.; Anno, J.; Cousinou, P. [Institut de Radioprotection et de Surete Nucleaire (IRSN), 92 - Fontenay aux Roses (France); Grivot, P.; Girault, E.; Fouillaud, P.; Barbry, F. [CEA Valduc, 21 - Is-sur-Tille (France)

2003-07-01

The group ' International Criticality Safety Evaluation Benchmark evaluation project ' (I.C.S.B.E.P.) has for aim to supply to the international community experiments of benchmarks criticality, of certified quality, used to guarantee the qualification of criticality calculation codes. Have been defined: a structure of experiments classification, a format of standard presentation, a structure of work with evaluation, internal and external checks, presentation in plenary session. After favourable opinion of the work group, the synthesis document called evaluation is integrated to the general report I.C.S.B.E.P. (N.C.)

Electricity generation and environmental externalities: Case studies, September 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-28

Electricity constitutes a critical input in sustaining the Nation`s economic growth and development and the well-being of its inhabitants. However, there are byproducts of electricity production that have an undesirable effect on the environment. Most of these are emissions introduced by the combustion of fossil fuels, which accounts for nearly 70 percent of the total electricity generated in the United States. The environmental impacts (or damages) caused by these emissions are labeled environmental ``externalities.`` Included in the generic term ``externality`` are benefits or costs resulting as an unintended byproduct of an economic activity that accrue to someone other than the parties involved in the activity. This report provides an overview of the economic foundation of externalities, the Federal and State regulatory approaches, and case studies of the impacts of the externality policies adopted by three States.

A comparison of internal versus external risk-adjustment for monitoring clinical outcomes

NARCIS (Netherlands)

Koetsier, Antonie; de Keizer, Nicolette; Peek, Niels

2011-01-01

Internal and external prognostic models can be used to calculate severity of illness adjusted mortality risks. However, it is unclear what the consequences are of using an external model instead of an internal model when monitoring an institution's clinical performance. Theoretically, using an

Application of voxel phantoms and Monte Carlo methods to internal and external dosimetry

International Nuclear Information System (INIS)

Hunt, J.G.; Santos, D. de S.; Silva, F.C. da; Dantas, B.M.; Azeredo, A.; Malatova, I.; Foltanova, S.

2000-01-01

Voxel phantoms and the Monte Carlo technique are applied to the areas of calibration of in vivo measurement systems, Specific Effective Energy calculations, and dose calculations due to external sources of radiation. The main advantages of the use of voxel phantoms is their high level of detail of body structures, and the ease with which their physical dimensions can be changed. For the simulation of in vivo measurement systems for calibration purposes, a voxel phantom with a format of 871 'slices' each of 277 x 148 picture elements was used. The Monte Carlo technique is used to simulate the tissue contamination, to transport the photons through the tissues and to simulate the detection of the radiation. For benchmarking, the program was applied to obtain calibration factors for the in vivo measurement of 241 Am, U nat and 137 Cs deposited in various tissues or in the whole body, as measured with a NaI or Gernlanium detector. The calculated and real activities in all cases were found to be in good agreement. For the calculation of Specific Effective Energies (SEEs) and the calculation of dose received from external sources, the Yale voxel phantom with a format of 493 slices' each of 87 x 147 picture elements was used. The Monte Carlo program was developed to calculate external doses due to environmental, occupational or accidental exposures. The program calculates tissue and effective dose for the following geometries: cloud immersion, ground contamination, X-ray irradiation, point source irradiation or others. The benchmarking results for the external source are in good agreement with the measured values. The results obtained for the SEEs are compatible with the ICRP values. (author)

Integration of risk aversion in the evaluation of the external cost of a nuclear accident

International Nuclear Information System (INIS)

Eeckhoudt, L.; Schieber, C.; Schneider, Th.

1998-01-01

Full text of publication follows: the external costs of fuel cycles used in the production of electricity are those imposed on society and environment that are not accounted for by the producers and consumers of energy. Within the evaluation of the external cost of the nuclear fuel cycle, the evaluation of a nuclear accident has to be addressed. For this purpose, the basic approach consists in calculating the expected value of various occident scenarios. the main criticism of this approach is that there is a discrepancy between the social acceptability of the risk and the average monetary value which corresponds in principle to the compensation of the consequences for each individual of the population affected by the accident. The aim of this paper is to propose a methodology for the integration of risk aversion, relying on the expected utility approach, as well as a numerical application based on the French data for the external cost of a nuclear accident. Although a huge range of values has been published for the relative risk aversion coefficient, it seems reasonable to adopt a value of 2 for the specific case of nuclear accident. This leads to an estimated multiplying coefficient approximately equal to 20 to be applied to the expected external cost of a nuclear accident corresponding to a release of about 1% of the core. In this case, the external cost of the nuclear accident is estimated to 0.046 mECU/kWh (i.e. about 50% of the total external costs of the nuclear fuel cycle estimated at 0.1 mECU/kWh with a 3% discount rate), instead of 0.0023 mECU/kWh without taking into account risk aversion. (authors)

External field-induced chaos in classical and quantum Hamiltonian systems

International Nuclear Information System (INIS)

Lin, W.C.

1986-01-01

Classical nonlinear nonintegrable systems exhibit dense sets of resonance zones in phase space. Global chaotic motion appears when neighboring resonance zones overlap. The chaotic motion signifies the destruction of a quasi constant of motion. The motion of a particle, trapped in one of the wells of a sinusoidal, potential driven by a monochromatic external field was studied. Global chaotic behavior sets in when the amplitude of the external field reaches a critical value. The particle then escapes the well. The critical values are found to be in good agreement with a resonance overlap criterion rather than a renormalization-group scheme. A similar system was then studied, but with the particle being confined in an infinite square well potential instead. A stochastic layer is found in the low-energy part of the phase space. The resonance zone structure is found to be in excellent agreement with predictions. The critical values for the onset of global chaotic behavior are found to be in excellent agreement with the renormalization group scheme. The quantum version of the second model above was then considered. In a similar fashion, the external field induces quantum resonance zones. The spectral statistics were computed, and a transition of statistics from Poissonian to Wigner-like was found as overlap of quantum resonances occurs. This also signifies the destruction of a quasi-constant of motion

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

Magnetocaloric effect and its implementation in critical behaviour ...

Indian Academy of Sciences (India)

Model; manganites; magnetization; magnetocaloric effect; critical exponent. 1. Introduction. Large number of magnetocaloric effect (MCE) materials have attracted much ... external magnetic field, which is advantageous for applica- tion as magnetic ... of the magnetic phase transition and critical behaviour can be obtained by ...

Externalities of energy. Swedish implementation of the ExternE methodology

International Nuclear Information System (INIS)

Nilsson, Maans; Gullberg, M.

1998-01-01

The growing interest for developing economic instruments for efficient environmental policies has opened up a large area of multi-disciplinary research. ExternE is an example of this research, combining disciplines such as engineering, ecology, immunology and economics expertise to create new knowledge about how environmental pressures from energy production affect our nature and society. The ExternE Project aims to identify and, as far as possible quantify the externalities of energy production in Europe. The Stockholm Environment Institute has carried out a preliminary aggregation: -Coal Fuel Cycle: centred around Vaesteraas Kraftvaermeverk, Vaesteraas. This is the largest co-generation plant in Sweden, with four blocks and a maximum co-generation output of 520 MW electricity and 950 MW heat. The analysis is carried out on boiler B4. -Biomass Fuel Cycle: centred around Haendeloeverket, Norrkoeping. This plant predominately burns forestry residues, but a variety of fuels are combusted. Haendeloeverket has an installed capacity of 100 MW electricity and 375 MW heat, in a total of three boilers and two back-pressure turbines. The analysis is carried out on boiler P13. -Hydro Fuel Cycle: Klippens Kraftstation, Storuman. Built in 1990-1994, it is the youngest hydro power station in Sweden. It has been designed and built with significant efforts to account for and protect environmental values. Installed capacity is 28 MW. The environmental impact assessment from the construction of this plant is carried out, but the evaluation is still not finalized. The preliminary aggregation aimed to test whether ExternE results could be used to make estimates for the entire Swedish electricity production system. Hence, national results as well as results from other partner countries in ExternE has been applied

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.

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

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

Fourier series analysis of a cylindrical pressure vessel subjected to axial end load and external pressure

International Nuclear Information System (INIS)

Brar, Gurinder Singh; Hari, Yogeshwar; Williams, Dennis K.

2013-01-01

This paper presents the comparison of a reliability technique that employs a Fourier series representation of random axisymmetric and asymmetric imperfections in a cylindrical pressure vessel subjected to an axial end load and external pressure, with evaluations prescribed by the ASME Boiler and Pressure Vessel Code, Section VIII, Division 2 Rules. The ultimate goal of the reliability technique described herein is to predict the critical buckling load associated with the subject cylindrical pressure vessel. Initial geometric imperfections are shown to have a significant effect on the calculated load carrying capacity of the vessel. Fourier decomposition was employed to interpret imperfections as structural features that can be easily related to various other types of defined imperfections. The initial functional description of the imperfections consists of an axisymmetric portion and a deviant portion, which are availed in the form of a double Fourier series. Fifty simulated shells generated by the Monte Carlo technique are employed in the final prediction of the critical buckling load. The representation of initial geometrical imperfections in the cylindrical pressure vessel requires the determination of respective Fourier coefficients. Multi-mode analyses are expanded to evaluate a large number of potential buckling modes for both predefined geometries in combination with asymmetric imperfections as a function of position within the given cylindrical shell. The probability of the ultimate buckling stress exceeding a predefined threshold stress is also calculated. The method and results described herein are in stark contrast to the “knockdown factor” approach as applied to compressive stress evaluations currently utilized in industry. Further effort is needed to improve on the current design rules regarding column buckling of large diameter pressure vessels subjected to an axial end load and external pressure designed in accordance with ASME Boiler and

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

Modal response of interior mass based upon external measurements

International Nuclear Information System (INIS)

Chow, C T; Eli, M; Jorgensen, B R; Woehrle, T.

1999-01-01

Modal response testing has been used to predict the motion of interior masses of a system in which only external instrumentation is allowed. Testing of this form may occasionally be necessary in validation of a computer model, but also has potential as a tool for validating individual assemblies in a QA process. Examination of the external frequency response and mode shapes can offer insight into interior response. The interpretation of these results is improved through parallel analytical solutions. A simple, three-mass model has been examined experimentally and analytically to demonstrate modal theory. These results show the limitations of the external measurement in predicting internal response due to transmissibility. A procedure for utilizing external testing is described. The question posed through this research is whether or not modal correlation analysis can be adapted for use in systems for which instrumentation of critical components is missing

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.

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

CRITICALITY CURVES FOR PLUTONIUM HYDRAULIC FLUID MIXTURES

International Nuclear Information System (INIS)

WITTEKIND WD

2007-01-01

This Calculation Note performs and documents MCNP criticality calculations for plutonium (100% 239 Pu) hydraulic fluid mixtures. Spherical geometry was used for these generalized criticality safety calculations and three geometries of neutron reflection are: (sm b ullet)bare, (sm b ullet)1 inch of hydraulic fluid, or (sm b ullet)12 inches of hydraulic fluid. This document shows the critical volume and critical mass for various concentrations of plutonium in hydraulic fluid. Between 1 and 2 gallons of hydraulic fluid were discovered in the bottom of HA-23S. This HA-23S hydraulic fluid was reported by engineering to be Fyrquel 220. The hydraulic fluid in GLovebox HA-23S is Fyrquel 220 which contains phosphorus. Critical spherical geometry in air is calculated with 0 in., 1 in., or 12 inches hydraulic fluid reflection

No need for external orthogonality in subsystem density-functional theory.

Science.gov (United States)

Unsleber, Jan P; Neugebauer, Johannes; Jacob, Christoph R

2016-08-03

Recent reports on the necessity of using externally orthogonal orbitals in subsystem density-functional theory (SDFT) [Annu. Rep. Comput. Chem., 8, 2012, 53; J. Phys. Chem. A, 118, 2014, 9182] are re-investigated. We show that in the basis-set limit, supermolecular Kohn-Sham-DFT (KS-DFT) densities can exactly be represented as a sum of subsystem densities, even if the subsystem orbitals are not externally orthogonal. This is illustrated using both an analytical example and in basis-set free numerical calculations for an atomic test case. We further show that even with finite basis sets, SDFT calculations using accurate reconstructed potentials can closely approach the supermolecular KS-DFT density, and that the deviations between SDFT and KS-DFT decrease as the basis-set limit is approached. Our results demonstrate that formally, there is no need to enforce external orthogonality in SDFT, even though this might be a useful strategy when developing projection-based DFT embedding schemes.

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

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

External dose reconstruction in tooth enamel of Techa riverside residents

Energy Technology Data Exchange (ETDEWEB)

Shishkina, E.A.; Volchkova, A.Yu.; Krivoschapov, V.A.; Degteva, M.O. [Urals Research Center for Radiation Medicine, Chelyabinsk (Russian Federation); Timofeev, Y.S.; Zalyapin, V.I. [Southern Urals State University, Chelyabinsk (Russian Federation); Fattibene, P.; Della Monaca, S.; De Coste, V. [Istituto Superiore di Sanita e Istituto Nazionale di Fisica Nucleare, Rome (Italy); Wieser, A. [Helmholtz Zentrum Muenchen, German Research Centre for Environmental Health, Neuherberg (Germany); Ivanov, D.V. [M.N. Mikheev Institute of Metal Physics, Ural Division of the Russian Academy of Sciences, Ekaterinburg (Russian Federation); Ural Federal University, Yekaterinburg (Russian Federation); Anspaugh, L.R. [University of Utah, Salt Lake City, UT (United States)

2016-11-15

This study summarizes the 20-year efforts for dose reconstruction in tooth enamel of the Techa riverside residents exposed to ionizing radiation as a result of radionuclide releases into the river in 1949-1956. It represents the first combined analysis of all the data available on EPR dosimetry with teeth of permanent residents of the Techa riverside territory. Results of electron paramagnetic resonance (EPR) measurements of 302 teeth donated by 173 individuals living permanently in Techa riverside settlements over the period of 1950-1952 were analyzed. These people were residents of villages located at the free-flowing river stream or at the banks of stagnant reservoirs such as ponds or blind river forks. Cumulative absorbed doses measured using EPR are from several sources of exposure, viz., background radiation, internal exposure due to bone-seeking radionuclides ({sup 89}Sr, {sup 90}Sr/{sup 90}Y), internal exposure due to {sup 137}Cs/{sup 137m}Ba incorporated in soft tissues, and anthropogenic external exposure. The purpose of the present study was to evaluate the contribution of different sources of enamel exposure and to deduce external doses to be used for validation of the Techa River Dosimetry System (TRDS). Since various EPR methods were used, harmonization of these methods was critical. Overall, the mean cumulative background dose was found to be 63 ± 47 mGy; cumulative internal doses due to {sup 89}Sr and {sup 90}Sr/{sup 90}Y were within the range of 10-110 mGy; cumulative internal doses due to {sup 137}Cs/{sup 137m}Ba depend on the distance from the site of releases and varied from 1 mGy up to 90 mGy; mean external doses were maximum for settlements located at the banks of stagnant reservoirs (∝500 mGy); in contrast, external doses for settlements located along the free-flowing river stream did not exceed 160 mGy and decreased downstream with increasing distance from the site of release. External enamel doses calculated using the TRDS code and

Environmental external effects from wind power based on the EU ExternE methodology

DEFF Research Database (Denmark)

Ibsen, Liselotte Schleisner; Nielsen, Per Sieverts

1998-01-01

of the Danish part of the project is to implement the framework for externality evaluation, for three different power plants located in Denmark. The paper will focus on the assessment of the impacts of the whole fuel cycles for wind, natural gas and biogas. Priority areas for environmental impact assessment......The European Commission has launched a major study project, ExternE, to develop a methodology to quantify externalities. A “National Implementation Phase”, was started under the Joule II programme with the purpose of implementing the ExternE methodology in all member states. The main objective...

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.)

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.)

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.)

Plasmon instability under four external fields

International Nuclear Information System (INIS)

Pereira, R.B.; Fonseca, A.L.A.; Nunes, O.A.C.

1998-01-01

The plasmon instability in a laboratory produced plasma in the presence of four external fields, namely two laser fields, one strong magnetic field and one static electric field, is discussed. The method of unitary transformations is used to transform the problem of electron motion under the four external fields to that of an electron in the presence only of crossed electric and magnetic fields. A kinetic equation for the plasmon population is derived from which the damping (amplification) rate is calculated. We found that the joint action of the four fields results in a relatively larger amplification rate for some values of the static electric field in contrast to the case where no electric field is present. It was also found that the plasmon growth rate favors plasmon wave vectors in an extremely narrow band i.e., the plasmon instability in four external fields is a very selective mechanism for plasmon excitation. (author)

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

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

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)

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)

In-vivo dosimetric study of carcinoma of uterine cervix with FBX solution in external beam therapy

International Nuclear Information System (INIS)

Srinivas, Challapalli; Shenoy, K. Kamalaksh; Dinesh, M.; Savitha, K.S.; Kasturi, Dinesh Pai; Supe, S.S.; Nagesha, Y.N.

1999-01-01

To ensure accurate dose delivery to target site in external beam therapy and brachytherapy, various authors have conducted tests to assess the process of manual dose calculations. In vivo dosimetric measurement is one of these methods to verify these calculations. In this study, an attempt has been made to compare the manually calculated dose to dose estimated using a chemical dosimeter (FBX) solution (in-vivo method, using polypropylene vials), on 12 patients of carcinoma of uterine cervix in external beam therapy. Dose measured by FBX vial varies in the range of ± 2 to 6.75%, as compared with manual calculations. These variations seen may be attributed to the location of the vial position in the vagina, with reference to the beam axis (may not be horizontal), off axis position, manual calculation variations and reproducibility of the FBX system etc. FBX dosimetry offers itself as an in-vivo method to estimate the dose delivered to the target site in external beam therapy. (author)

Bursting pressure of autofrettaged cylinders with inclined external cracks

International Nuclear Information System (INIS)

Seifi, Rahman; Babalhavaeji, Majid

2012-01-01

Autofrettaging a pressure vessel improves its pressure capacity. This is reliable if there isn’t any crack or other type of flaws. In this paper, the effects of external surface cracks on bursting pressure of autofrettaged cylinders are studied. It is observed that bursting pressure decreases considerably (up to 30%) due to external cracks in the cylinders without autofrettage. This reduction increases for high levels of the applied autofrettage. External axial cracks have more effects than inclined cracks. Comparing experimental and numerical results show that the numerical methods can acceptably predict the bursting pressure of the autofrettaged cracked cylinders. These predictions are valid when the fracture parameter (J-Integral) is calculated from the modified equation that takes into account the effects of residual stresses. - Highlights: ► Modified J-Integral can be used for study of autofrettaged cracked cylinders. ► External axial cracks reduce considerably the pressure capacity of cylinders. ► External circumferential cracks have not considerable effects on bursting pressure. ► Autofrettage has contrary effects on external crack in compared with internal crack.

Constant external fields in gauge theory and the spin 0, 1/2, 1 path integrals

International Nuclear Information System (INIS)

Reuter, M.; Schmidt, M.G.

1996-10-01

We investigate the usefulness of the ''string-inspired technique'' for gauge theory calculations in a constant external field background. Our approach is based on Strassler's worldline path integral approach to the Bern-Kosower formalism, and on the construction of worldline (super-) Green's functions incorporating external fields as well as internal propagators. The worldline path integral representation of the gluon loop is reexamined in detail. We calculate the two-loop effective actions induced for a constant external field by a scalar and spinor loop, and the corresponding one-loop effective action in the gluon loop case. (orig.)

Optimization of External Envelope Insulation Thickness: A Parametric Study

Directory of Open Access Journals (Sweden)

Eleftheria Touloupaki

2017-02-01

Full Text Available Almost four years after the implementation deadline of the energy performance of buildings Directive recast (2010/31/EU and after being referred to the Court of Justice of the EU by the European Commission, Greece has not yet proceeded with the necessary calculations and legislative measures on the minimum, cost-optimal energy performance requirements for buildings. This paper aims to identify the optimal thickness of insulation that is cost-effective to apply in urban multi-family domestic buildings in the four climate zones of Greece. A reference building is selected in order to perform calculations over ten scenarios of external insulation thickness for each climate zone on a basic and three sensitivity analysis calculations according to the EU comparative methodology framework. The resulting energy savings for each insulation scenario are calculated, and then the cost-effectiveness of the measure is examined in financial and macroeconomic perspective for an economic lifecycle of 30 years. The results demonstrate the inadequacy of the national regulation’s current insulation limits and the externalities (funding gaps that need to be addressed in order to achieve the effective improvement of energy efficiency in Greek homes.

Online plasma calculator

Science.gov (United States)

Wisniewski, H.; Gourdain, P.-A.

2017-10-01

APOLLO is an online, Linux based plasma calculator. Users can input variables that correspond to their specific plasma, such as ion and electron densities, temperatures, and external magnetic fields. The system is based on a webserver where a FastCGI protocol computes key plasma parameters including frequencies, lengths, velocities, and dimensionless numbers. FastCGI was chosen to overcome security problems caused by JAVA-based plugins. The FastCGI also speeds up calculations over PHP based systems. APOLLO is built upon the WT library, which turns any web browser into a versatile, fast graphic user interface. All values with units are expressed in SI units except temperature, which is in electron-volts. SI units were chosen over cgs units because of the gradual shift to using SI units within the plasma community. APOLLO is intended to be a fast calculator that also provides the user with the proper equations used to calculate the plasma parameters. This system is intended to be used by undergraduates taking plasma courses as well as graduate students and researchers who need a quick reference calculation.

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

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

CONSEQUENCES OF EXTERNAL ENVIRONMENT ON ENTREPRENEURIAL MOTIVATION IN IRAN

Directory of Open Access Journals (Sweden)

Hassan Gholipour Fereidouni

2010-07-01

Full Text Available The purpose of this study is to examine importance of business environment, social status of entrepreneurs, and country external conflicts as predictors of motivation to start a business in an environment, such as that in Iran, in which the economy is highly dependent on government initiatives. Data are collected from 106 MBA students through questionnaires. Respondents are questioned regarding the perceived importance of the business environment, socio-cultural factors (social status, and country external conflicts. The results show that the importance of business environment and country external conflicts contribute considerably to the level of entrepreneurial motivation. Further, the results reveal that social status is not a critical factor in determining the level of motivation to start a business. In particular, business environment is the most important factor in predicting entrepreneurial motivation. The results contribute to the growing body of literature in entrepreneurship and provide some implications for Iranian policy makers to create favourable external environment for potential entrepreneurs.

Cluster-cell calculation using the method of generalized homogenization

International Nuclear Information System (INIS)

Laletin, N.I.; Boyarinov, V.F.

1988-01-01

The generalized-homogenization method (GHM), used for solving the neutron transfer equation, was applied to calculating the neutron distribution in the cluster cell with a series of cylindrical cells with cylindrically coaxial zones. Single-group calculations of the technological channel of the cell of an RBMK reactor were performed using GHM. The technological channel was understood to be the reactor channel, comprised of the zirconium rod, the water or steam-water mixture, the uranium dioxide fuel element, and the zirconium tube, together with the adjacent graphite layer. Calculations were performed for channels with no internal sources and with unit incoming current at the external boundary as well as for channels with internal sources and zero current at the external boundary. The PRAKTINETs program was used to calculate the symmetric neutron distributions in the microcell and in channels with homogenized annular zones. The ORAR-TsM program was used to calculate the antisymmetric distribution in the microcell. The accuracy of the calculations were compared for the two channel versions

Improvements for Monte Carlo burnup calculation

Energy Technology Data Exchange (ETDEWEB)

Shenglong, Q.; Dong, Y.; Danrong, S.; Wei, L., E-mail: qiangshenglong@tsinghua.org.cn, E-mail: d.yao@npic.ac.cn, E-mail: songdr@npic.ac.cn, E-mail: luwei@npic.ac.cn [Nuclear Power Inst. of China, Cheng Du, Si Chuan (China)

2015-07-01

Monte Carlo burnup calculation is development trend of reactor physics, there would be a lot of work to be done for engineering applications. Based on Monte Carlo burnup code MOI, non-fuel burnup calculation methods and critical search suggestions will be mentioned in this paper. For non-fuel burnup, mixed burnup mode will improve the accuracy of burnup calculation and efficiency. For critical search of control rod position, a new method called ABN based on ABA which used by MC21 will be proposed for the first time in this paper. (author)

External Cost Assessment of Nuclear Power Plant Accident considering Public Risk Aversion Behavior: the Korean Case

Energy Technology Data Exchange (ETDEWEB)

Lee, Sang Hun; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

2016-05-15

The conventional approach for monetary valuation of NPP accident consequence consists of calculating the expected value of various accident scenarios. However, the main criticism of the conventional approach is that there is a discrepancy between the social acceptability of the risk and the estimated expected value of NPP accident. Therefore, an integrated framework for the estimation of the external cost associated with an NPP accident considering the public risk aversion behavior was proposed in this study based on the constructed theoretical framework for estimating both the value of statistical life (VSL) and the risk aversion coefficient associated with an NPP accident to take account of the accident cost into the unit electricity generation cost of NPP. To estimate both parameters, an individual-level survey was conducted on a sample of 1,364 participants in Korea. Based on the collected survey responses, both parameters were estimated based on the proposed framework and the external cost of NPP accident was estimated based on the consequence analysis and considering the direct cost factors for NPP accident. Internalization of external costs into the comprehensive energy production cost has been considered as a potentially efficient policy instrument for a more sustainable energy supply and use. However, the internalization of externalities, such as public health damage, have raised a number of generic policy issues in a nuclear energy sector, with specific challenges resulting from the distinct characteristics of external cost estimation. Especially, the major challenge remained to address the public safety concerns regarding a nuclear accident, which can be specified as low-probability high-consequence accident, driven by the aspects of public risk aversion.

External Cost Assessment of Nuclear Power Plant Accident considering Public Risk Aversion Behavior: the Korean Case

International Nuclear Information System (INIS)

Lee, Sang Hun; Kang, Hyun Gook

2016-01-01

The conventional approach for monetary valuation of NPP accident consequence consists of calculating the expected value of various accident scenarios. However, the main criticism of the conventional approach is that there is a discrepancy between the social acceptability of the risk and the estimated expected value of NPP accident. Therefore, an integrated framework for the estimation of the external cost associated with an NPP accident considering the public risk aversion behavior was proposed in this study based on the constructed theoretical framework for estimating both the value of statistical life (VSL) and the risk aversion coefficient associated with an NPP accident to take account of the accident cost into the unit electricity generation cost of NPP. To estimate both parameters, an individual-level survey was conducted on a sample of 1,364 participants in Korea. Based on the collected survey responses, both parameters were estimated based on the proposed framework and the external cost of NPP accident was estimated based on the consequence analysis and considering the direct cost factors for NPP accident. Internalization of external costs into the comprehensive energy production cost has been considered as a potentially efficient policy instrument for a more sustainable energy supply and use. However, the internalization of externalities, such as public health damage, have raised a number of generic policy issues in a nuclear energy sector, with specific challenges resulting from the distinct characteristics of external cost estimation. Especially, the major challenge remained to address the public safety concerns regarding a nuclear accident, which can be specified as low-probability high-consequence accident, driven by the aspects of public risk aversion

An Improved Method to Control the Critical Parameters of a Multivariable Control System

Science.gov (United States)

Subha Hency Jims, P.; Dharmalingam, S.; Wessley, G. Jims John

2017-10-01

The role of control systems is to cope with the process deficiencies and the undesirable effect of the external disturbances. Most of the multivariable processes are highly iterative and complex in nature. Aircraft systems, Modern Power Plants, Refineries, Robotic systems are few such complex systems that involve numerous critical parameters that need to be monitored and controlled. Control of these important parameters is not only tedious and cumbersome but also is crucial from environmental, safety and quality perspective. In this paper, one such multivariable system, namely, a utility boiler has been considered. A modern power plant is a complex arrangement of pipework and machineries with numerous interacting control loops and support systems. In this paper, the calculation of controller parameters based on classical tuning concepts has been presented. The controller parameters thus obtained and employed has controlled the critical parameters of a boiler during fuel switching disturbances. The proposed method can be applied to control the critical parameters like elevator, aileron, rudder, elevator trim rudder and aileron trim, flap control systems of aircraft systems.

Criticality safety studies at VTT Energy

International Nuclear Information System (INIS)

Roine, T.; Anttila, M.

1995-01-01

At VTT Energy a compact reactor physics calculation system is applied in many kind of problems. Generation of group constants for static and dynamic core calculations, flux and dose rate calculations as well as criticality safety studies are performed basically with the same codes. In the presentation a short overview of the wide variety of criticality safety problems analyzed at VTT Energy is given. The calculation system with some illustrative examples is also described. (12 refs., 1 tab.)

Validation of dose calculation programmes for recycling

International Nuclear Information System (INIS)

Menon, Shankar; Brun-Yaba, Christine; Yu, Charley; Cheng, Jing-Jy; Williams, Alexander

2002-12-01

This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on

Validation of dose calculation programmes for recycling

Energy Technology Data Exchange (ETDEWEB)

Menon, Shankar [Menon Consulting, Nykoeping (Sweden); Brun-Yaba, Christine [Inst. de Radioprotection et Securite Nucleaire (France); Yu, Charley; Cheng, Jing-Jy [Argonne National Laboratory, IL (United States). Environmental Assessment Div.; Bjerler, Jan [Studsvik Stensand, Nykoeping (Sweden); Williams, Alexander [Dept. of Energy (United States). Office of Environmental Management

2002-12-01

This report contains the results from an international project initiated by the SSI in 1999. The primary purpose of the project was to validate some of the computer codes that are used to estimate radiation doses due to the recycling of scrap metal. The secondary purpose of the validation project was to give a quantification of the level of conservatism in clearance levels based on these codes. Specifically, the computer codes RESRAD-RECYCLE and CERISE were used to calculate radiation doses to individuals during the processing of slightly contaminated material, mainly in Studsvik, Sweden. Calculated external doses were compared with measured data from different steps of the process. The comparison of calculations and measurements shows that the computer code calculations resulted in both overestimations and underestimations of the external doses for different recycling activities. The SSI draws the conclusion that the accuracy is within one order of magnitude when experienced modellers use their programmes to calculate external radiation doses for a recycling process involving material that is mainly contaminated with cobalt-60. No errors in the codes themselves were found. Instead, the inaccuracy seems to depend mainly on the choice of some modelling parameters related to the receptor (e.g., distance, time, etc.) and simplifications made to facilitate modelling with the codes (e.g., object geometry). Clearance levels are often based on studies on enveloping scenarios that are designed to cover all realistic exposure pathways. It is obvious that for most practical cases, this gives a margin to the individual dose constraint (in the order of 10 micro sievert per year within the EC). This may be accentuated by the use of conservative assumptions when modelling the enveloping scenarios. Since there can obviously be a fairly large inaccuracy in the calculations, it seems reasonable to consider some degree of conservatism when establishing clearance levels based on

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.

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...

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.

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.)

Determination of dose received by bladder and rectum in external cervical irradiation

International Nuclear Information System (INIS)

Omer, Mohamed Ahmed Ali

2001-12-01

The cervical carcinoma is the common type of malignant tumor among sudanese females during the last years. The conventional external irradiation therapy is the common model of treatment for cervical carcinoma in (RICK). The irradiation of such cases implemented via four fields (box technique), two anterior and posterior and two lateral opposed fields, as central dose calculation, giving a dose of 5000 c Gy fractionated into 25 to 30 fractions. The parameter of the fields lie at the promontory of the sacral cephalic and at the obturators foramen caudally and laterally at the bony pelvic by one centimeter. The aim of the research is to determine the dose received by the rectum and bladder (critical organs), out of central dose calculation versus off axis dose calculation to (Day's method). The data obtained by using simulator and radiation oncological computerized system (Rocs). The results are analyzed by using statistical processing for social science program (SPSS) that shows the mean dose received by the bladder is 3821 cGy, due to central dose calculation that accompanied by an un-optimum encompassment of treatment line and 4210.6 c Gy. due to Off Axis Dose calculation dose 3324.4 c Gy and 3712.1 c Gy due to central dose calculation. The increment of dose received by the rectum and the bladder is due to utilizing of wider width of the filed size for lateral irradiation. To score the aim of International Atomic Energy Agency (IAEA) and (ALARA) principle A s Low As As Reasonable Available , we have to use the simulator to obtain the anatomical structures on the contour, or will be better to use CT. Scan for calculation of dose at the side of interest. (Author)

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

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.

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

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)

Reconciling Organisational Culture and External Quality Assurance in Higher Education

Science.gov (United States)

Naidoo, Dhaya

2013-01-01

Organisational culture and external quality assurance have both been presented as significant drivers of effectiveness, efficiency and excellence in higher education institutions. However, these assumptions have not been critically examined given the philosophical, conceptual and methodological contestations surrounding both constructs. A…

Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

2016-01-01

Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

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.

Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

Energy Technology Data Exchange (ETDEWEB)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C. [Eletrobras Termonuclear S.A. - ELETRONUCLEAR, Angra dos Reis, RJ (Brazil). Central Nuclear Almirante Alvaro Alberto], e-mail: kuramot@eletronuclear.gov.br

2009-07-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

Development of a digital reactivity meter for criticality prediction and control rod worth evaluation in pressurized water reactors

International Nuclear Information System (INIS)

Kuramoto, Renato Y.R.; Miranda, Anselmo F.; Valladares, Gastao Lommez; Prado, Adelk C.

2009-01-01

In this work, we have proposed the development of a digital reactivity meter in order to monitor subcriticality continuously during criticality approach in a PWR. A subcritical reactivity meter can provide an easy prediction of the estimated critical point prior to reactor criticality, without complicated hand calculation. Moreover, in order to reduce the interval of the Physics Tests from the economical point of view, a subcritical reactivity meter can evaluate the control rod worth from direct subcriticality measurement. In other words, count rate of Source Range (SR) detector recorded during the criticality approach could be used for subcriticality evaluation or control rod worth evaluation. Basically, a digital reactivity meter is based on the inverse solution of the kinetic equations of a reactor with the external neutron source in one-point reactor model. There are some difficulties in the direct application of a digital reactivity meter to the subcriticality measurement. When the Inverse Kinetic method is applied to a sufficiently high power level or to a core without an external neutron source, the neutron source term may be neglected. When applied to a lower power level or in the sub-critical domain, however, the source effects must be taken in account. Furthermore, some treatments are needed in using the count rate of Source Range (SR) detector as input signal to the digital reactivity meter. To overcome these difficulties, we have proposed a digital reactivity meter combined with a methodology of the modified Neutron Source Multiplication (NSM) method with correction factors for subcriticality measurements in PWR. (author)

External costs related to power production technologies. ExternE national implementation for Denmark

Energy Technology Data Exchange (ETDEWEB)

Schleisner, L; Sieverts Nielsen, P [eds.

1997-12-01

The objective of the ExternE National Implementation project has been to establish a comprehensive and comparable set of data on externalities of power generation for all EU member states and Norway. The tasks include the application of the ExternE methodology to the most important fuel cycles for each country as well as to update the already existing results, to aggregate these site- and technology-specific results to more general figures. The current report covers the detailed information concerning the ExternE methodology. Importance is attached to the computer system used in the project and the assessment of air pollution effects on health, materials and ecological effects. Also the assessment of global warming damages are described. Finally the report covers the detailed information concerning the national implementation for Denmark for an offshore wind farm and a wind farm on land, a decentralised CHP plant based on natural gas and a decentralised CHP plant base on biogas. (au) EU-JOULE 3. 79 tabs., 11 ills., 201 refs.

External costs related to power production technologies. ExternE national implementation for Denmark

International Nuclear Information System (INIS)

Schleisner, L.; Sieverts Nielsen, P.

1997-12-01

The objective of the ExternE National Implementation project has been to establish a comprehensive and comparable set of data on externalities of power generation for all EU member states and Norway. The tasks include the application of the ExternE methodology to the most important fuel cycles for each country as well as to update the already existing results, to aggregate these site- and technology-specific results to more general figures. The current report covers the detailed information concerning the ExternE methodology. Importance is attached to the computer system used in the project and the assessment of air pollution effects on health, materials and ecological effects. Also the assessment of global warming damages are described. Finally the report covers the detailed information concerning the national implementation for Denmark for an offshore wind farm and a wind farm on land, a decentralised CHP plant based on natural gas and a decentralised CHP plant base on biogas. (au) EU-JOULE 3. 79 tabs., 11 ills., 201 refs

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