Interpretation and modelling of fission product Ba and Mo releases from fuel

Science.gov (United States)

Brillant, G.

2010-02-01

The release mechanisms of two fission products (namely barium and molybdenum) in severe accident conditions are studied using the VERCORS experimental observations. Barium is observed to be mostly released under reducing conditions while molybdenum release is most observed under oxidizing conditions. As well, the volatility of some precipitates in fuel is evaluated by thermodynamic equilibrium calculations. The polymeric species (MoO 3) n are calculated to largely contribute to molybdenum partial pressure and barium volatility is greatly enhanced if the gas atmosphere is reducing. Analytical models of fission product release from fuel are proposed for barium and molybdenum. Finally, these models have been integrated in the ASTEC/ELSA code and validation calculations have been performed on several experimental tests.

Realistic methods for calculating the releases and consequences of a large LOCA

International Nuclear Information System (INIS)

Stephenson, W.; Dutton, L.M.C.; Handy, B.J.; Smedley, C.

1992-01-01

This report describes a calculational route to predict realistic radiological consequences for a successfully terminated large-loss-of-coolant accident (LOCA) at a pressurized-water reactor (PWR). All steps in the calculational route are considered. For each one, a brief comment is made on the significant differences between the methods of calculation that were identified in the benchmark studies and recommendations are made for the methods and data for carrying out realistic calculations. These are based on the best supportable methods and data and the technical basis for each recommendation is given. Where the lack of well-validated methods or data means that the most realistic method that can be justified is considered to be very conservative, the need for further research is identified. The behaviour of inorganic iodine and the removal of aerosols from the atmosphere of the reactor building are identified as areas of particular importance. Where the retention of radioactivity is sensitive to design features, these are identified and, for the most importance features, the impact of different designs on the release of activity is indicated. The predictions of the proposed model are calculated for each stage and compared with the releases of activity predicted by the licensing methods that were used in the earlier benchmark studies. The conservative nature of the latter is confirmed. Methods and data are also presented for calculating the resulting doses to members of the public of the National Radiological Protection Boards as a result of work carried out by several national bodies in the UK. Other, equally acceptable, models are used in other countries of the Community and some examples are given

Screening calculations for radioactive waste releases from non-nuclear facilities

International Nuclear Information System (INIS)

Xu, Shulan; Soederman, Ann-Louis

2009-02-01

A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 μSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 μSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment plant

Screening calculations for radioactive waste releases from non-nuclear facilities

Energy Technology Data Exchange (ETDEWEB)

Shulan Xu; Soederman, Ann-Louis

2009-02-15

A series of screening calculations have been performed to assess the potential radiological consequences of discharges of radioactive substances to the environment arising from waste from non-nuclear practices. Solid waste, as well as liquids that are not poured to the sewer, are incinerated and ashes from incineration and sludge from waste water treatment plants are disposed or reused at municipal disposal facilities. Airborne discharges refer to releases from an incineration facility and liquid discharges refer both to releases from hospitals and laboratories to the sewage system, as well as leakage from waste disposal facilities. The external exposure of workers is estimated both in the waste water treatment plant and at the disposal facility. The calculations follow the philosophy of the IAEA's safety guidance starting with a simple assessment based on very conservative assumptions which may be iteratively refined using progressively more complex models, with more realistic assumptions, as necessary. In the assessments of these types of disposal, with cautious assumptions, carried out in this report we conclude that the radiological impacts on representative individuals in the public are negligible in that they are small with respect to the target dose of 10 muSv/a. A Gaussian plume model was used to estimate the doses from airborne discharges from the incinerator and left a significant safety margin in the results considering the conservative assumptions in the calculations. For the sewage plant workers the realistic approach included a reduction in working hours and the shorter exposure time resulted in maximum doses around 10 muSv/a. The calculations for the waste disposal facility show that the doses are higher or in the range of the target dose. The excess for public exposure is mainly caused by H-3 and C-14. The assumption used in the calculation is that all of the radioactive substances sent to the incineration facility and waste water treatment

Dose Assessment Model for Chronic Atmospheric Releases of Tritium

International Nuclear Information System (INIS)

Shen Huifang; Yao Rentai

2010-01-01

An improved dose assessment model for chronic atmospheric releases of tritium was proposed. The proposed model explicitly considered two chemical forms of tritium.It was based on conservative assumption of transfer of tritiated water (HTO) from air to concentration of HTO and organic beam tritium (OBT) in vegetable and animal products.The concentration of tritium in plant products was calculated based on considering dividedly leafy plant and not leafy plant, meanwhile the concentration contribution of tritium in the different plants from the tritium in soil was taken into account.Calculating the concentration of HTO in animal products, average water fraction of animal products and the average weighted tritium concentration of ingested water based on the fraction of water supplied by each source were considered,including skin absorption, inhalation, drinking water and food.Calculating the annual doses, the ingestion doses were considered, at the same time the contribution of inhalation and skin absorption to the dose was considered. Concentrations in foodstuffs and dose of annual adult calculated with the specific activity model, NEWTRI model and the model proposed by the paper were compared. The results indicate that the model proposed by the paper can predict accurately tritium doses through the food chain from chronic atmospheric releases. (authors)

Calculation of Doses Due to Accidentally Released Plutonium From An LMFBR

Energy Technology Data Exchange (ETDEWEB)

Fish, B.R.

2001-08-07

Experimental data and analytical models that should be considered in assessing the transport properties of plutonium aerosols following a hypothetical reactor accident have been examined. Behaviors of released airborne materials within the reactor containment systems, as well as in the atmosphere near the reactor site boundaries, have been semiquantitatively predicted from experimental data and analytical models. The fundamental chemistry of plutonium as it may be applied in biological systems has been used to prepare models related to the intake and metabolism of plutonium dioxide, the fuel material of interest. Attempts have been made to calculate the possible doses from plutonium aerosols for a typical analyzed release in order to evaluate the magnitude of the internal exposure hazards that might exist in the vicinity of the reactor after a hypothetical LMFBR (Liquid-Metal Fast Breeder Reactor) accident. Intake of plutonium (using data for {sup 239}Pu as an example) and its distribution in the body were treated parametrically without regard to the details of transport pathways in the environment. To the extent possible, dose-response data and models have been reviewed, and an assessment of their adequacy has been made so that recommended or preferred practices could be developed.

Comparison of numerical models for calculating dispersion from accidental releases of pollutants

Energy Technology Data Exchange (ETDEWEB)

Pepper, D W [Savannah River Lab., Aiken, SC; Cooper, R E; Baker, A J

1982-01-01

A modular, data-based system approach has been developed to facilitate computational simulation of multi-dimensional pollutant dispersion in atmospheric, steam, estuary, and groundwater applications. This system is used to assess effects of accidental releases of pollutants to the environment. Model sophistication ranges from simple statistical to complex three-dimensional numerical methods. The system used specifies desired degree of model sophistication from a terminal. The model used depends on the particular type of problem being solved, and on a basis of merit related to computer cost. The results of prediction for several model problems are presented.

A model for the release of low-volatility fission products in oxidizing conditions

International Nuclear Information System (INIS)

Cox, D.S.; Hunt, C.E.L.; Liu, Z.; Keller, N.A.; Barrand, R.D.; O'Connor, R.F.

1991-07-01

A thermodynamic and kinetic model has been developed for calculating low-volatility fission-product releases from UO 2 at high temperatures in oxidizing conditions. Volatilization of the UO 2 matrix is assumed to be the rate controlling process. Oxidation kinetics of the UO 2 are modelled by either interfacial rate control, gas phase oxidant transport control, or solid-state diffusion of oxygen. The vapour pressure of UO 3 in equilibrium with the oxidizing fuel is calculated from thermodynamic data, and volatilization rates are determined using a model for forced convective mass transport. Low-volatility fission-product releases are calculated from the volume of vapourized fuel. Model calculations are conservative compared to experimental data for Zr, La, Ce and Nb fission-product releases from irradiated UO 2 exposed to air at 1973-2350 K. The implications of this conservatism are discussed in terms of possible rate control by processes other than convective mass transport of UO 3 . Coefficients for effective surface area (based on experimental data) and for heterogeneous rate controlling reaction kinetics are introduced to facilitate agreement between calculations and the experimental data.

Tritium transport calculations for the IFMIF Tritium Release Test Module

Energy Technology Data Exchange (ETDEWEB)

Freund, Jana, E-mail: jana.freund@kit.edu; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-10-15

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

Tritium transport calculations for the IFMIF Tritium Release Test Module

International Nuclear Information System (INIS)

Freund, Jana; Arbeiter, Frederik; Abou-Sena, Ali; Franza, Fabrizio; Kondo, Keitaro

2014-01-01

Highlights: • Delivery of material data for the tritium balance in the IFMIF Tritium Release Test Module. • Description of the topological models in TMAP and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). • Computation of release of tritium from the breeder solid material into the purge gas. • Computation of the loss of tritium over the capsule wall, rig hull, container wall and purge gas return line. - Abstract: The IFMIF Tritium Release Test Module (TRTM) is projected to measure online the tritium release from breeder ceramics and beryllium pebble beds under high energy neutron irradiation. Tritium produced in the pebble bed of TRTM is swept out continuously by a purge gas flow, but can also permeate into the module's metal structures, and can be lost by permeation to the environment. According analyses on the tritium inventory are performed to support IFMIF plant safety studies, and to support the experiment planning. This paper describes the necessary elements for calculation of the tritium transport in the Tritium Release Test Module as follows: (i) applied equations for the tritium balance, (ii) material data from literature and (iii) the topological models and the computation of the five different cases; namely release of tritium from the breeder solid material into the purge gas, loss of tritium over the capsule wall, rig hull, container wall and purge gas return line in detail. The problem of tritium transport in the TRTM has been studied and analyzed by the Tritium Migration Analysis Program (TMAP) and the adapted fusion-devoted Tritium Permeation Code (FUS-TPC). TMAP has been developed at INEEL and now exists in Version 7. FUS-TPC Code was written in MATLAB with the original purpose to study the tritium transport in Helium Cooled Lead Lithium (HCLL) blanket and in a later version the Helium Cooled Pebble Bed (HCPB) blanket by [6] (Franza, 2012). This code has been further modified to be applicable to the TRTM. Results from the

Modelling transient energy release from molten fuel coolant interaction debris

International Nuclear Information System (INIS)

Fletcher, D.F.

1984-05-01

A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)

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)

Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

International Nuclear Information System (INIS)

Verfondern, K.

1991-03-01

The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

Validation of kinetic modeling of progesterone release from polymeric membranes

Directory of Open Access Journals (Sweden)

Analia Irma Romero

2018-01-01

Full Text Available Mathematical modeling in drug release systems is fundamental in development and optimization of these systems, since it allows to predict drug release rates and to elucidate the physical transport mechanisms involved. In this paper we validate a novel mathematical model that describes progesterone (Prg controlled release from poly-3-hydroxybutyric acid (PHB membranes. A statistical analysis was conducted to compare the fitting of our model with six different models and the Akaike information criterion (AIC was used to find the equation with best-fit. A simple relation between mass and drug released rate was found, which allows predicting the effect of Prg loads on the release behavior. Our proposed model was the one with minimum AIC value, and therefore it was the one that statistically fitted better the experimental data obtained for all the Prg loads tested. Furthermore, the initial release rate was calculated and therefore, the interface mass transfer coefficient estimated and the equilibrium distribution constant of Prg between the PHB and the release medium was also determined. The results lead us to conclude that our proposed model is the one which best fits the experimental data and can be successfully used to describe Prg drug release in PHB membranes.

Considerations in modeling fission gas release during normal operation

International Nuclear Information System (INIS)

Rumble, E.T.; Lim, E.Y.; Stuart, R.G.

1977-01-01

The EPRI LWR fuel rod modeling code evaluation program analyzed seven fuel rods with experimental fission gas release data. In these cases, rod-averged burnups are less than 20,000 MWD/MTM, while the fission gas release fractions range roughly from 2 to 27%. Code results demonstrate the complexities in calculating fission gas release in certain operating regimes. Beyond this work, the behavior of a pre-pressurized PWR rod is simulated to average burnups of 40,000 MWD/MTM using GAPCON-THERMAL-2. Analysis of the sensitivity of fission gas release to power histories and release correlations indicate the strong impact that LMFBR type release correlations induce at high burnup. 15 refs

Calculation of releases of radioactive materials in gaseous effluents from nuclear-powered merchant ships (NMS-GEFF code)

International Nuclear Information System (INIS)

Cardile, F.P.; Bangart, R.L.; Collins, J.T.

1978-06-01

The Intergovernmental Maritime Consultative Organization IMCO) is currently preparing guidelines concerning the safety of nuclear-powered merchant ships. An important aspect of these guidelines is the determination of the releases of radioactive material in effluents from these ships and the control exercised by the ships over these releases. To provide a method for the determination of these releases, the NRC staff has developed a computerized model, the NMS-GEFF Code, which is described in the following chapters. The NMS-GEFF Code calculates releases of radioactive material in gaseous effluents for nuclear-powered merchant ships using pressurized water reactors

A compartment model for nuclide release calculation in the near-and far-field of a HLW repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Hahn, Pil Soo

2004-01-01

The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997, from which a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel is to be introduced by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as spent fuel and generic site characteristics in Korea was roughly envisaged in 2003. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near - and far - field components of the repository, even though sufficient information has not been available that much yet, but also to show a appropriate methodology by which both a generic and site - specific safety assessment could be performed for further in - depth development of Korea reference repository concept, nuclide release calculation study for various nuclide release cases is mandatory. To this end a similar study done and yet limited for the near - field release case has been extended to the case including far - field system by introducing some more geosphere compartments. Advective and longitudinal dispersive nuclide transports along the fracture with matrix diffusion as well as several retention mechanisms and nuclide ingrowth has been added

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

The article presents methodology of external exposure calculation for reuse of conditional released materials from decommissioning using VISIPLAN 3D ALARA planning tool. Production of rails has been used as an example application of proposed methodology within the CONRELMAT project. The article presents a methodology for determination of radiological, material, organizational and other conditions for conditionally released materials reuse to ensure that workers and public exposure does not breach the exposure limits during scenario's life cycle (preparation, construction and operation of scenario). The methodology comprises a proposal of following conditions in the view of workers and public exposure: - radionuclide limit concentration of conditionally released materials for specific scenarios and nuclide vectors, - specific deployment of conditionally released materials eventually shielding materials, workers and public during the scenario's life cycle, - organizational measures concerning time of workers or public stay in the vicinity on conditionally released materials for individual performed scenarios and nuclide vectors. The above mentioned steps of proposed methodology have been applied within the CONRELMAT project. Exposure evaluation of workers for rail production is introduced in the article as an example of this application. Exposure calculation using VISIPLAN 3D ALARA planning tool was done within several models. The most exposed profession for scenario was identified. On the basis of this result, an increase of radionuclide concentration in conditional released material was proposed more than two times to 681 Bq/kg without no additional safety or organizational measures being applied. After application of proposed safety and organizational measures (additional shielding, geometry changes and limitation of work duration) it is possible to increase concentration of radionuclide in conditional released material more than ten times to 3092 Bq/kg. Storage

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

Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

International Nuclear Information System (INIS)

Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

1978-12-01

The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment

Models for environmental impact assessments of releases of radioactive substances from CERN facilities

CERN Document Server

Vojtyla, P

2005-01-01

The document describes generic models for environmental impact assessments of releases of radioactive substances from CERN facilities. Except for few models developed in the Safety Commission, the models are based on the 1997 Swiss directive HSK-R-41 and on the 2001 IAEA Safety Report No. 19. The writing style is descriptive, facilitating the practical implementation of the models at CERN. There are four scenarios assumed for airborne releases: (1) short-term releases for release limit calculations, (2) actual short-term releases, (3) short-term releases during incidents/accidents, and (4) chronic long-term releases during the normal operation of a facility. For water releases, two scenarios are considered: (1) a release into a river, and (2) a release into a water treatment plant. The document shall be understood as a reference for specific environmental studies involving radioactive releases and as a recommendation of the Safety Commission.

Comparative calculations and validation studies with atmospheric dispersion models

International Nuclear Information System (INIS)

Paesler-Sauer, J.

1986-11-01

This report presents the results of an intercomparison of different mesoscale dispersion models and measured data of tracer experiments. The types of models taking part in the intercomparison are Gaussian-type, numerical Eulerian, and Lagrangian dispersion models. They are suited for the calculation of the atmospherical transport of radionuclides released from a nuclear installation. For the model intercomparison artificial meteorological situations were defined and corresponding arithmetical problems were formulated. For the purpose of model validation real dispersion situations of tracer experiments were used as input data for model calculations; in these cases calculated and measured time-integrated concentrations close to the ground are compared. Finally a valuation of the models concerning their efficiency in solving the problems is carried out by the aid of objective methods. (orig./HP) [de

A simple operational gas release and swelling model. Pt. 1

International Nuclear Information System (INIS)

Wood, M.H.; Matthews, J.R.

1980-01-01

A new and simple model of fission gas release and swelling has been developed for oxide nuclear fuel under operational conditions. The model, which is to be incorporated into a fuel element behaviour code, is physically based and applicable to fuel at both thermal and fast reactor ratings. In this paper we present that part of the model describing the behaviour of intragranular gas: a future paper will detail the treatment of the grain boundary gas. The results of model calculations are compared with recent experimental observations of intragranular bubble concentrations and sizes, and gas release from fuel irradiated under isothermal conditions. Good agreement is found between experiment and theory. (orig.)

Modeling a point-source release of 1,1,1-trichloroethane using EPA's SCREEN model

International Nuclear Information System (INIS)

Henriques, W.D.; Dixon, K.R.

1994-01-01

Using data from the Environmental Protection Agency's Toxic Release Inventory 1988 (EPA TRI88), pollutant concentration estimates were modeled for a point source air release of 1,1,1-trichloroethane at the Savannah River Plant located in Aiken, South Carolina. Estimates were calculating using the EPA's SCREEN model utilizing typical meteorological conditions to determine maximum impact of the plume under different mixing conditions for locations within 100 meters of the stack. Input data for the SCREEN model were then manipulated to simulate the impact of the release under urban conditions (for the purpose of assessing future landuse considerations) and under flare release options to determine if these parameters lessen or increase the probability of human or wildlife exposure to significant concentrations. The results were then compared to EPA reference concentrations (RfC) in order to assess the size of the buffer around the stack which may potentially have levels that exceed this level of safety

Discussion on the methods for calculation release limits for low-level radioactive waste

International Nuclear Information System (INIS)

Cao Fengbo; Liu Xiaochao

2012-01-01

The release request for low-level radioactive waste are briefly described in this paper. Associating with the conditions of low-level radioactive waste of some radioactive waste processing station, the methods and gist for calculating release limits for low-level radioactive waste with national release limits and annual effective dose limit for the public or the occupation are discussed. Then release limits for the low-level radioactive waste are also proposed. (authors)

ANS-5.4 fission gas release model. I. Noble gases at high temperature

International Nuclear Information System (INIS)

Noble, L.D.

1979-01-01

A correlation to describe the release of volatile radioactive fission products has been developed by the ANS Working Group (ANS 5.4) on Fuel Plenum Activity. The model for release at higher temperatures is identical in form to conventional diffusion equations, but the effective diffusion coefficient incorporates an explicit dependence upon exposure. Because applicable radioactive release data is limited, parameters in the model were determined from stable fission measurements, and calculated or measured fuel temperatures. Although the model predicts high release, particularly at higher exposures, values for many cases of interest are considerably less than the 100% assumed in some accident analyses: providing potential for removal of unnecessary conservations

Sensitivity of the engineered barrier system (EBS) release rate to alternative conceptual models of advective release from waste packages under dripping fractures

International Nuclear Information System (INIS)

Lee, J.H.; Atkins, J.E.; McNeish, J.A.; Vallikat, V.

1996-01-01

Simulations were conducted to analyze the sensitivity of the engineered barrier system (EBS) release rate to alternative conceptual models of the advective release from waste packages under dripping fractures. The first conceptual model assumed that dripping water directly contacts the waste form inside the 'failed' waste package, and radionuclides are released from the EBS by advection. The second conceptual model assumed that dripping water is diverted around the 'failed' waste package (because of the presence of corrosion products plugging the perforations) and dripping water is prevented from directly contacting the waste form. In the second model, radionuclides were assumed to transport through the perforations by diffusion, and, once outside the waste package, to be released from the EBS by advection. The second model was to incorporate more realism into the EBS release calculations. For the case with the second EBS release model, most radionuclides had significantly lower peak EBS release rates (from at least one to several orders of magnitude) than with the first EBS release model. The impacts of the alternative EBS release models were greater for the radionuclides with a low solubility (or solubility-limited radionuclides) than for the radionuclides with a high solubility (or waste form dissolution-limited radionuclides). The analyses indicated that the EBS release model representing advection through a 'failed' waste package (the first EBS release model) may be too conservative in predicting the EBS performance. One major implication from this sensitivity study was that a 'failed' waste package container with multiple perforations may still be able to perform effectively as an important barrier to radionuclide release. (author)

Calculations of individual doses for Techa River Cohort members exposed to atmospheric radioiodine from Mayak releases

Energy Technology Data Exchange (ETDEWEB)

Napier, Bruce A.; Eslinger, Paul W.; Tolstykh, Evgenia I.; Vorobiova, Marina I.; Tokareva, Elena E.; Akhramenko, Boris N.; Krivoschapov, Victor A.; Degteva, Marina O.

2017-11-01

Time-dependent thyroid doses were reconstructed for Techa River Cohort members living near the Mayak production facilities from 131I released to the atmosphere for all relevant exposure pathways. The calculational approach uses four general steps: 1) construct estimates of releases of 131I to the air from production facilities; 2) model the transport of 131I in the air and subsequent deposition on the ground and vegetation; 3) model the accumulation of 131I in soil, water, and food products (environmental media); and 4) calculate individual doses by matching appropriate lifestyle and consumption data for the individual to concentrations of 131I in environmental media. The dose calculations are implemented in a Monte Carlo framework that produces best estimates and confidence intervals of dose time-histories. The 131I contribution was 75-99% of the thyroid dose. The mean total thyroid dose for cohort members was 193 mGy and the median was 53 mGy. Thyroid doses for about 3% of cohort members were larger than 1 Gy. About 7% of children born in 1940-1950 had doses larger than 1 Gy. The uncertainty in the 131I dose estimates is low enough for this approach to be used in regional epidemiological studies.

Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

International Nuclear Information System (INIS)

Miller, C.; Little, C.A.

1982-08-01

The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases

An investigation of fission models for high-energy radiation transport calculations

International Nuclear Information System (INIS)

Armstrong, T.W.; Cloth, P.; Filges, D.; Neef, R.D.

1983-07-01

An investigation of high-energy fission models for use in the HETC code has been made. The validation work has been directed checking the accuracy of the high-energy radiation transport computer code HETC to investigate the appropriate model for routine calculations, particularly for spallation neutron source applications. Model calculations are given in terms of neutron production, fission fragment energy release, and residual nuclei production for high-energy protons incident on thin uranium targets. The effect of the fission models on neutron production from thick uranium targets is also shown. (orig.)

Underground water stress release models

Science.gov (United States)

Li, Yong; Dang, Shenjun; Lü, Shaochuan

2011-08-01

The accumulation of tectonic stress may cause earthquakes at some epochs. However, in most cases, it leads to crustal deformations. Underground water level is a sensitive indication of the crustal deformations. We incorporate the information of the underground water level into the stress release models (SRM), and obtain the underground water stress release model (USRM). We apply USRM to the earthquakes occurred at Tangshan region. The analysis shows that the underground water stress release model outperforms both Poisson model and stress release model. Monte Carlo simulation shows that the simulated seismicity by USRM is very close to the real seismicity.

Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

International Nuclear Information System (INIS)

Suh, K.Y.

1989-10-01

A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

MODELS SELECTED FOR CALCULATION OF DOSES, HEALTH EFFECTS AND ECONOMIC COSTS DUE TO ACCIDENTAL RADIONUCLIDE RELEASES FROM NUCLEAR POWER PLANTS

Energy Technology Data Exchange (ETDEWEB)

Strenge, D L; Baker, D A; Droppo, J G; McPherson, R B; Napier, B A; Nieves, L A; Soldat, J K

1980-05-01

Models are described for use in site-specific environmental consequence analysis of nuclear reactor accidents of Classes 3 through 9. The models presented relate radioactivity released to resulting doses, health effects, and costs of remedial actions. Specific models are presented for the major exposure pathways of airborne releases, waterborne releases and direct irradiation from activity within the facility buildings, such as the containment. Time-dependent atmospheric dispersion parameters, crop production parameters and other variable parameters are used in the models. The environmental effects are analyzed for several accident start times during the year.

TRAFIC, a computer program for calculating the release of metallic fission products from an HTGR core

International Nuclear Information System (INIS)

Smith, P.D.

1978-02-01

A special purpose computer program, TRAFIC, is presented for calculating the release of metallic fission products from an HTGR core. The program is based upon Fick's law of diffusion for radioactive species. One-dimensional transient diffusion calculations are performed for the coated fuel particles and for the structural graphite web. A quasi steady-state calculation is performed for the fuel rod matrix material. The model accounts for nonlinear adsorption behavior in the fuel rod gap and on the coolant hole boundary. The TRAFIC program is designed to operate in a core survey mode; that is, it performs many repetitive calculations for a large number of spatial locations in the core. This is necessary in order to obtain an accurate volume integrated release. For this reason the program has been designed with calculational efficiency as one of its main objectives. A highly efficient numerical method is used in the solution. The method makes use of the Duhamel superposition principle to eliminate interior spatial solutions from consideration. Linear response functions relating the concentrations and mass fluxes on the boundaries of a homogeneous region are derived. Multiple regions are numerically coupled through interface conditions. Algebraic elimination is used to reduce the equations as far as possible. The problem reduces to two nonlinear equations in two unknowns, which are solved using a Newton Raphson technique

Diffusion modeling of fission product release during depressurized core conduction cooldown conditions

International Nuclear Information System (INIS)

Martin, R.C.

1991-01-01

A simple model for diffusion through the silicon carbide layer of TRISO particles is applied to the data for accident condition testing of fuel spheres for the High-Temperature Reactor program of the Federal Republic of Germany (FRG). Categorization of sphere release of 137 Cs based on fast neutron fluence permits predictions of release with an accuracy comparable to that of the US/FRG accident condition fuel performance model. Calculations are also performed for 85 Kr, 90 Sr, and 110m Ag. Diffusion of cesium through SiC suggests that models of fuel failure should consider fuel performance during repeated accident condition thermal cycling. Microstructural considerations in models of fission product release are discussed. The neutron-induced segregation of silicon within the SiC structure is postulated as a mechanism for enhanced fission product release during accident conditions. As oxygen-enhanced SiC decomposition mechanism is also discussed. (author). 12 refs, 11 figs, 2 tabs

Sensitivity analysis for CORSOR models simulating fission product release in LOFT-LP-FP-2 severe accident experiment

Energy Technology Data Exchange (ETDEWEB)

Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club; Pourgol-Mohammad, Mohammad [Sahand Univ. of Technology, Tabriz (Iran, Islamic Republic of). Dept. of Mechanical Engineering; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of)

2017-03-15

This paper deals with simulation, sensitivity and uncertainty analysis of LP-FP-2 experiment of LOFT test facility. The test facility simulates the major components and system response of a pressurized water reactor during a LOCA. MELCOR code is used for predicting the fission product release from the core fuel elements in LOFT LP-FP-2 experiment. Moreover, sensitivity and uncertainty analysis is performed for different CORSOR models simulating release of fission products in severe accident calculations for nuclear power plants. The calculated values for the fission product release are compared under different modeling options to the experimental data available from the experiment. In conclusion, the performance of 8 CORSOR modeling options is assessed for available modeling alternatives in the code structure.

Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

Energy Technology Data Exchange (ETDEWEB)

Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

1995-12-31

Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

A model of propagating calcium-induced calcium release mediated by calcium diffusion

NARCIS (Netherlands)

Backx, P. H.; de Tombe, P. P.; van Deen, J. H.; Mulder, B. J.; ter Keurs, H. E.

1989-01-01

The effect of sudden local fluctuations of the free sarcoplasmic [Ca++]i in cardiac cells on calcium release and calcium uptake by the sarcoplasmic reticulum (SR) was calculated with the aid of a simplified model of SR calcium handling. The model was used to evaluate whether propagation of calcium

A guide to TIRION 4 - a computer code for calculating the consequences of releasing radioactive material to the atmosphere

International Nuclear Information System (INIS)

Fryer, L.S.

1978-12-01

TIRION 4 is the most recent program in a series designed to calculate the consequences of releasing radioactive material to the atmosphere. A brief description of the models used in the program and full details of the various control cards necessary to run TIRION 4 are given. (author)

A fission gas release model for MOX fuel and its verification

International Nuclear Information System (INIS)

Koo, Y.H.; Sohn, D.S.; Strijov, P.

2000-01-01

A fission gas release model for MOX fuel has been developed based on a model for UO 2 fuel. Using the concept of equivalent cell, the model considers the uneven distribution of Pu within the fuel matrix and a number of Pu-rich particles that could lead to a non-uniform fission rate and fission gas distribution across the fuel pellet. The model has been incorporated into a code, COSMOS, and some parametric studies were made to analyze the effect of the size and Pu content of Pu-rich agglomerates. The model was then applied to the experimental data obtained from the FIGARO program, which consisted of the base irradiation of MOX fuels in the BEZNAU-1 PWR and the subsequent irradiation of four refabricated fuel segments in the Halden reactor. The calculated gas releases show good agreement with the measured ones. In addition, the present analysis indicates that the microstructure of the MOX fuel used in the FIGARO program is such that it has produced little difference in terms of gas release compared with UO 2 fuel. (author)

Geochemistry Model Validation Report: Material Degradation and Release Model

Energy Technology Data Exchange (ETDEWEB)

H. Stockman

2001-09-28

The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing 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 model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17).

Geochemistry Model Validation Report: Material Degradation and Release Model

International Nuclear Information System (INIS)

Stockman, H.

2001-01-01

The purpose of this Analysis and Modeling Report (AMR) is to validate the Material Degradation and Release (MDR) model that predicts degradation and release of radionuclides from a degrading waste package (WP) in the potential monitored geologic repository at Yucca Mountain. This AMR is prepared according to ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 17). The intended use of the MDR model is to estimate the long-term geochemical behavior of waste packages (WPs) containing 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 model is intended to predict (1) the extent to which criticality control material, such as gadolinium (Gd), will remain in the WP after corrosion of the initial WP, (2) the extent to which fissile Pu and uranium (U) will be carried out of the degraded WP by infiltrating water, and (3) the chemical composition and amounts of minerals and other solids left in the WP. The results of the model are intended for use in criticality calculations. The scope of the model validation report is to (1) describe the MDR model, and (2) compare the modeling results with experimental studies. A test case based on a degrading Pu-ceramic WP is provided to help explain the model. This model does not directly feed the assessment of system performance. The output from this model is used by several other models, such as the configuration generator, criticality, and criticality consequence models, prior to the evaluation of system performance. This document has been prepared according to AP-3.10Q, ''Analyses and Models'' (Ref. 2), and prepared in accordance with the technical work plan (Ref. 17)

Equilibrium and kinetic models for colloid release under transient solution chemistry conditions.

Science.gov (United States)

Bradford, Scott A; Torkzaban, Saeed; Leij, Feike; Simunek, Jiri

2015-10-01

We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and kinetic, and two-site kinetic models were developed to describe various rates of colloid release. These models were subsequently applied to experimental colloid release datasets to investigate the influence of variations in ionic strength (IS), pH, cation exchange, colloid size, and water velocity on release. Various combinations of equilibrium and/or kinetic release models were needed to describe the experimental data depending on the transient conditions and colloid type. Release of Escherichia coli D21g was promoted by a decrease in solution IS and an increase in pH, similar to expected trends for a reduction in the secondary minimum and nanoscale chemical heterogeneity. The retention and release of 20nm carboxyl modified latex nanoparticles (NPs) were demonstrated to be more sensitive to the presence of Ca(2+) than D21g. Specifically, retention of NPs was greater than D21g in the presence of 2mM CaCl2 solution, and release of NPs only occurred after exchange of Ca(2+) by Na(+) and then a reduction in the solution IS. These findings highlight the limitations of conventional interaction energy calculations to describe colloid retention and release, and point to the need to consider other interactions (e.g., Born, steric, and/or hydration forces) and/or nanoscale heterogeneity. Temporal changes in the water velocity did not have a large influence on the release of D21g for the examined conditions. This insensitivity was likely due to factors that reduce the applied hydrodynamic torque and/or increase the resisting adhesive torque; e.g., macroscopic roughness and grain-grain contacts. Our analysis and models improve our understanding and ability to describe the amounts

Modeling of molten core-concrete interactions and fission-product release

International Nuclear Information System (INIS)

Norkus, J.K.; Corradini, M.L.

1991-09-01

The study of molten core-concrete interaction is important in estimating the possible consequences of a severe nuclear reactor accident. CORCON-Mod2 is a computer program which models the thermal, chemical, and physical phenomena associated with molten core-concrete interactions. Models have been added to extend and improve the modeling of these phenomena. An ideal solution chemical equilibrium methodology is presented to predict the fission-product vaporization release. Additional chemical species have been added, and the calculation of chemical equilibrium has been expanded to the oxidic layer and to the mixed layer configuration. Recent experiments performed at Argonne National Laboratory are compared to CORCON predictions of melt temperature, erosion depth, and release fraction of fission products. The results consistently underpredicted the melt temperatures and erosion rates. However, the predictions of release of Te, Ba, Sr, and U were good. A sensitivity study of the effects of initial temperature, concrete type, use of the mixing option, degree of zirconium oxidation, cavity size, and amount of control material on erosion, gas production, and release of radioactive materials was performed for a PWR and a BWR. The initial melt temperature had the greatest effect on the results of interest. Concrete type and cavity size also had important effects. 78 refs., 35 figs., 40 tabs

PWR-GALE, Radioactive Gaseous Release and Liquid Release from PWR

International Nuclear Information System (INIS)

Chandrasekaran, T.; Lee, J.Y.; Willis, C.A.

1988-01-01

1 - Description of program or function: The PWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from pressurized water reactors (PWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment. 2 - Method of solution: GALE calculates expected releases based on 1) standardized coolant activities derived from ANS Standards 18.1 Working Group recommendations, 2) release and transport mechanisms that result in the appearance of radioactive material in liquid and gaseous waste streams, 3) plant-specific design features used to reduce the quantities of radioactive materials ultimately released to the environs, and 4) information received on the operation of nuclear power plants. 3 - Restrictions on the complexity of the problem: The liquid release portion of GALE uses subroutines taken from the ORIGEN (CCC-217) to calculate radionuclide buildup and decay during collection, processing, and storage of liquid radwaste. Memory requirements for this part of the program are determined by the large nuclear data base accessed by these subroutines

Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

International Nuclear Information System (INIS)

Xhonneux, Andre; Allelein, Hans-Josef

2014-01-01

The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

Simplified analytical model to simulate radionuclide release from radioactive waste trenches

International Nuclear Information System (INIS)

Sa, Bernardete Lemes Vieira de

2001-01-01

In order to evaluate postclosure off-site doses from low-level radioactive waste disposal facilities, a computer code was developed to simulate the radionuclide released from waste form, transport through vadose zone and transport in the saturated zone. This paper describes the methodology used to model these process. The radionuclide released from the waste is calculated using a model based on first order kinetics and the transport through porous media was determined using semi-analytical solution of the mass transport equation, considering the limiting case of unidirectional convective transport with three-dimensional dispersion in an isotropic medium. The results obtained in this work were compared with other codes, showing good agreement. (author)

Dose evaluation model for radionuclides released from the spent nuclear fuel reprocessing plant in Rokkasho

International Nuclear Information System (INIS)

Hisamatsu, Shun'ichi; Iyogi, Takashi; Inaba, Jiro; Chiang, Jing-Hsien; Suwa, Hiroji; Koide, Mitsuo

2007-01-01

A dose evaluation model was developed for radionuclides released from the spent nuclear fuel reprocessing plant which is located in Rokkasho, Aomori Prefecture, and now undergoing test operation. The dose evaluation model suitable for medium- and long-term dose assessments for both prolonged and short-term releases of radionuclides to the atmosphere was developed on the PC. The ARAC-2, a particle tracing type dispersion model coupled with 3-D wind field calculation by a mass conservative model, was adopted as the atmospheric dispersion model. The terrestrial transfer model included movement in soil and groundwater as well as an agricultural and livestock farming system. The available site-specific social and environmental characteristics were incorporated in the model. Growing of the crops was also introduced and radionuclides absorbed were calculated from weight increase from the start of deposition to harvest, and transfer factors. Most of the computer code system of the models was completed by 2005, and this paper reports the results of the development. (author)

Comparison of source-term calculations using the AREST and SYVAC-Vault models: [Final report

International Nuclear Information System (INIS)

Apted, M.J.; Engel, D.W.; Garisto, N.C.; LeNeveu, D.M.

1988-07-01

A comparison of the calculated radionuclide release from a waste package in a geologic repository has been performed using the verified SYVAC-Vault Model and AREST Model. the purpose of this comparison is to further establish the credibility of these codes for predictive performance assessment and to identify improvements that may be required. A reference case for a Canadian conceptual design with spent fuel as the waste form was chosen to make an initial comparison. The results from the two models were in good agreement, including peak release rates, time to reach peak release, and long term release rates. Differences in results from the two models are attributed to differences in computational approaches. Studies of the effects of sorption, convective flow, distributed containment failure, and precipitation are identified as key areas for further comparisons and are currently in progress. 11 refs., 3 figs., 5 tabs

Models selected for calculation of doses, health effects and economic costs due to accidental radionuclide releases from nuclear power plants. Technical report

International Nuclear Information System (INIS)

Strenge, D.L.; Acharya, S.; Baker, D.A.; Droppo, J.G.; McPherson, R.B.

1980-05-01

Models are described for use in site-specific environmental consequence analysis of nuclear reactor accidents of Classes 3 through 9. The models presented relate radioactivity released to resulting doses, health effects, and costs of remedial actions. Specific models are presented for the major exposure pathways of airborne releases, waterborne releases and direct irradiation from activity within the facility buildings, such as the containment. Time-dependent atmospheric dispersion parameters, crop production parameters, and other variable parameters are used in the models. The environmental effects are analyzed for several accident start times during the year. Several remedial actions are considered

Selection of models to calculate the LLW source term

International Nuclear Information System (INIS)

Sullivan, T.M.

1991-10-01

Performance assessment of a LLW disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). In turn, many of these physical processes are influenced by the design of the disposal facility (e.g., infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This document provides a brief overview of disposal practices and reviews existing source term models as background for selecting appropriate models for estimating the source term. The selection rationale and the mathematical details of the models are presented. Finally, guidance is presented for combining the inventory data with appropriate mechanisms describing release from the disposal facility. 44 refs., 6 figs., 1 tab

Model description. NUDOS: A computer program for assessing the consequences of airborne releases of radionuclides

International Nuclear Information System (INIS)

Poley, A.D.

1996-02-01

NUDOS is a computer program that can be used to evaluate the consequences of airborne releases of radioactive material. The consequences which can be evaluated are individual dose and associated radiological risk, collective dose and the contamination of land. The code is capable of dealing with both continuous (routine) and accidental releases. For accidental releases both deterministic and probabilistic calculations can be performed, and the impact and effectiveness of emergency actions can be evaluated. This report contains a description of the models contained in NUDOS92 and the recommended values for the input parameters of these models. Additionally, a short overview is given of the future model improvement planned for the next NUDOS-version. (orig.)

Use of OND-86 recommendations for calculation of the Shelter radioactive release

International Nuclear Information System (INIS)

Bogatov, S.A.

2000-01-01

A model of radioactive release from the Shelter has been considered under current operation conditions. Integral assessment of current dust release has been done on the base of natural ventilation rate. Model predictions are consistent (20% accuracy) with experimental results of air contamination measurements at the earth surface. 12 refs., 1 tab., 6 figs

Modelling of contaminant release from a uranium mine tailings site

International Nuclear Information System (INIS)

Kahnt, Rene; Metschies, Thomas

2007-01-01

Available in abstract form only. Full text of publication follows: Uranium mining and milling continuing from the early 1960's until 1990 close to the town of Seelingstaedt in Eastern Germany resulted in 4 tailings impoundments with a total tailings volume of about 105 Mio. m 3 . Leakage from these tailings impoundments enters the underlying aquifers and is discharged into surface water streams. High concentration of salts, uranium and several heavy metals are released from the tailings. At present the tailings impoundments are reshaped and covered. For the identification of suitable remediation options predictions of the contaminant release for different remediation scenarios have to be made. A compartment model representing the tailings impoundments and the surrounding aquifers for the calculation of contaminant release and transport was set up using the software GOLDSIM. This compartment model describes the time dependent hydraulic conditions within the tailings and the surrounding aquifers taking into account hydraulic and geotechnical processes influencing the hydraulic properties of the tailings material. A simple geochemical approach taking into account sorption processes as well as retardation by applying a k d -approach was implemented to describe the contaminant release and transport within the hydraulic system. For uranium as the relevant contaminant the simple approach takes into account additional geochemical conditions influencing the mobility. Alternatively the model approach allows to include the results of detailed geochemical modelling of the individual tailings zones which is than used as source term for the modelling of the contaminant transport in the aquifer and to the receiving streams. (authors)

The Updated BaSTI Stellar Evolution Models and Isochrones. I. Solar-scaled Calculations

DEFF Research Database (Denmark)

Hidalgo, Sebastian L.; Pietrinferni, Adriano; Cassisi, Santi

2018-01-01

We present an updated release of the BaSTI (a Bag of Stellar Tracks and Isochrones) stellar model and isochrone library for a solar-scaled heavy element distribution. The main input physics that have been changed from the previous BaSTI release include the solar metal mixture, electron conduction...... to metal enrichment ratio dY/dZ = 1.31. The isochrones cover an age range between 20 Myr and 14.5 Gyr, consistently take into account the pre-main-sequence phase, and have been translated to a large number of popular photometric systems. Asteroseismic properties of the theoretical models have also been...... calculated. We compare our isochrones with results from independent databases and with several sets of observations to test the accuracy of the calculations. All stellar evolution tracks, asteroseismic properties, and isochrones are made available through a dedicated web site....

The Updated BaSTI Stellar Evolution Models and Isochrones. I. Solar-scaled Calculations

Science.gov (United States)

Hidalgo, Sebastian L.; Pietrinferni, Adriano; Cassisi, Santi; Salaris, Maurizio; Mucciarelli, Alessio; Savino, Alessandro; Aparicio, Antonio; Silva Aguirre, Victor; Verma, Kuldeep

2018-04-01

We present an updated release of the BaSTI (a Bag of Stellar Tracks and Isochrones) stellar model and isochrone library for a solar-scaled heavy element distribution. The main input physics that have been changed from the previous BaSTI release include the solar metal mixture, electron conduction opacities, a few nuclear reaction rates, bolometric corrections, and the treatment of the overshooting efficiency for shrinking convective cores. The new model calculations cover a mass range between 0.1 and 15 M ⊙, 22 initial chemical compositions between [Fe/H] = ‑3.20 and +0.45, with helium to metal enrichment ratio dY/dZ = 1.31. The isochrones cover an age range between 20 Myr and 14.5 Gyr, consistently take into account the pre-main-sequence phase, and have been translated to a large number of popular photometric systems. Asteroseismic properties of the theoretical models have also been calculated. We compare our isochrones with results from independent databases and with several sets of observations to test the accuracy of the calculations. All stellar evolution tracks, asteroseismic properties, and isochrones are made available through a dedicated web site.

Development of an atmospheric diffusion numerical model for a nuclear facility. Numerical calculation method incorporating building effects

International Nuclear Information System (INIS)

Sada, Koichi; Michioka, Takenobu; Ichikawa, Yoichi

2002-01-01

Because effluent gas is sometimes released from low positions, viz., near the ground surface and around buildings, the effects caused by buildings within the site area are not negligible for gas diffusion predictions. For these reasons, the effects caused by buildings for gas diffusion are considered under the terrain following calculation coordinate system in this report. Numerical calculation meshes on the ground surface are treated as the building with the adaptation of wall function techniques of turbulent quantities in the flow calculations using a turbulence closure model. The reflection conditions of released particles on building surfaces are taken into consideration in the diffusion calculation using the Lagrangian particle model. Obtained flow and diffusion calculation results are compared with those of wind tunnel experiments around the building. It was apparent that features observed in a wind tunnel, viz., the formation of cavity regions behind the building and the gas diffusion to the ground surface behind the building, are also obtained by numerical calculation. (author)

Analysis of in-pile tritium release experiments

International Nuclear Information System (INIS)

Kopasz, J.P.; Tam, S.W.; Johnson, C.E.

1992-01-01

The objective of this work is to characterize tritium release behavior from lithium ceramics and develop insight into the underlying tritium release mechanisms. Analysis of tritium release data from recent laboratory experiments with lithium aluminate has identified physical processes which were previously unaccounted for in tritium release models. A new model that incorporates the recent data and provides for release from multiple sites rather than only one site was developed. Calculations of tritium release using this model are in excellent agreement with the tritium release behavior reported for the MOZART experiment

Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

International Nuclear Information System (INIS)

Verfondern, K.; Mueller, D.

1991-01-01

Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

A simplified model for calculating atmospheric radionuclide transport and early health effects from nuclear reactor accidents

International Nuclear Information System (INIS)

Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

1995-01-01

During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences

Steady-state and transient fission gas release and swelling model for LIFE-4

International Nuclear Information System (INIS)

Villalobos, A.; Liu, Y.Y.; Rest, J.

1984-06-01

The fuel-pin modeling code LIFE-4 and the mechanistic fission gas behavior model FASTGRASS have been coupled and verified against gas release data from mixed-oxide fuels which were transient tested in the TREAT reactor. Design of the interface between LIFE-4 and FASTGRASS is based on an earlier coupling between an LWR version of LIFE and the GRASS-SST code. Fission gas behavior can significantly affect steady-state and transient fuel performance. FASTGRASS treats fission gas release and swelling in an internally consistent manner and simultaneously includes all major mechanisms thought to influence fission gas behavior. The FASTGRASS steady-state and transient analysis has evolved through comparisons of code predictions with fission-gas release and swelling data from both in- and ex-reactor experiments. FASTGRASS was chosen over other fission-gas behavior models because of its availability, its compatibility with the LIFE-4 calculational framework, and its predictive capability

Atmospheric dispersion models of radioactivity releases

International Nuclear Information System (INIS)

Oza, R.B.

2016-01-01

In view of the rapid industrialization in recent time, atmospheric dispersion models have become indispensible 'tools' to ensure that the effects of releases are well within the acceptable limits set by the regulatory authority. In the case of radioactive releases from the nuclear facility, though negligible in quantity and many a times not even measurable, it is required to demonstrate the compliance of these releases to the regulatory limits set by the regulatory authority by carrying out radiological impact assessment. During routine operations of nuclear facility, the releases are so low that environmental impact is usually assessed with the help of atmospheric dispersion models as it is difficult to distinguish negligible contribution of nuclear facility to relatively high natural background radiation. The accidental releases from nuclear facility, though with negligible probability of occurrence, cannot be ruled out. In such cases, the atmospheric dispersion models are of great help to emergency planners for deciding the intervention actions to minimize the consequences in public domain and also to workout strategies for the management of situation. In case of accidental conditions, the atmospheric dispersion models are also utilized for the estimation of probable quantities of radionuclides which might have got released to the atmosphere. Thus, atmospheric dispersion models are an essential tool for nuclear facility during routine operation as well as in the case of accidental conditions

Model to estimate the local radiation doses to man from the atmospheric release of radionuclides (LWBR development program)

International Nuclear Information System (INIS)

Rider, J.L.; Beal, S.K.

1977-04-01

A model was developed to estimate the radiation dose commitments received by people in the vicinity of a facility that releases radionuclides into the atmosphere. This model considers dose commitments resulting from immersion in the plume, ingestion of contaminated food, inhalation of gaseous and suspended radioactivity, and exposure to ground deposits. The dose commitments from each of these pathways is explicitly considered for each radionuclide released into the atmosphere and for each daughter of each released nuclide. Using the release rate of only the parent radionuclide, the air and ground concentrations of each daughter are calculated for each position of interest. This is considered to be a significant improvement over other models in which the concentrations of daughter radionuclides must be approximated by separate releases

Modelling vesicular release at hippocampal synapses.

Directory of Open Access Journals (Sweden)

Suhita Nadkarni

2010-11-01

Full Text Available We study local calcium dynamics leading to a vesicle fusion in a stochastic, and spatially explicit, biophysical model of the CA3-CA1 presynaptic bouton. The kinetic model for vesicle release has two calcium sensors, a sensor for fast synchronous release that lasts a few tens of milliseconds and a separate sensor for slow asynchronous release that lasts a few hundred milliseconds. A wide range of data can be accounted for consistently only when a refractory period lasting a few milliseconds between releases is included. The inclusion of a second sensor for asynchronous release with a slow unbinding site, and thereby a long memory, affects short-term plasticity by facilitating release. Our simulations also reveal a third time scale of vesicle release that is correlated with the stimulus and is distinct from the fast and the slow releases. In these detailed Monte Carlo simulations all three time scales of vesicle release are insensitive to the spatial details of the synaptic ultrastructure. Furthermore, our simulations allow us to identify features of synaptic transmission that are universal and those that are modulated by structure.

Calculation of the fuel temperature field under heat release and heat conductance transient conditions

International Nuclear Information System (INIS)

Kazakov, E.K.; Chernukhina, G.M.

1974-01-01

Results of calculation of the temperature distribution in an annular fuel element at transient thermal conductivity and heat release values are given. The calculation has been carried out by the mesh technique with the third-order boundary conditions for the inner surface assumed and with heat fluxes and temperatures at the zone boundaries to be equal. Three variants of solving the problem of a stationary temperature field are considered for failed fuel elements with clad flaking or cracks. The results obtained show the nonuniformity of the fuel element temperature field to depend strongly on the perturbation parameter at transient thermal conductivity and heat release values. In case of can flaking at a short length, the core temperature rises quickly after flaking. While evaluating superheating, one should take into account the symmetry of can flaking [ru

An evaluation of nodalization/decay heat/ volatile fission product release models in ISAAC code

Energy Technology Data Exchange (ETDEWEB)

Song, Yong Mann; Park, Soo Yong; Kim, Dong Ha

2003-03-01

An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. In case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option shows mitigated conservative results.

Continuous-release formulation for environmental doses to moving receptors

International Nuclear Information System (INIS)

Piepho, M.G.

1981-01-01

Atmospheric dispersion models frequently assume a puff release or several puff releases, each of which are described separately. A dispersion model should better describe a continuous release as more puffs are assumed, but the computational cost and bookkeeping difficulty increases with additional puffs. A new formulism is derived in this work which replaces the puff approximation. With the new continuous release formulation, radioactive dose calculations to moving receptors are more accurately calculated without any great additional computation. There are several advantages of a continuous release formulation. With this formulation, a dose rate to a moving receptor is calculated as a function of time. The dose-rate will increase (decrease) as the bulk of the release gets closer (farther) to (from) the receptor which is at position x(t), y(t). The receptor may follow any x, y trajectory as a function of time, and the dose rate will be calculated along the path

Calculation of the Fission Product Release for the HTR-10 based on its Operation History

International Nuclear Information System (INIS)

Xhonneux, A.; Druska, C.; Struth, S.; Allelein, H.-J.

2014-01-01

Since the first criticality of the HTR-10 test reactor in 2000, a rather complex operation history was performed. As the HTR-10 is the only pebble bed reactor in operation today delivering experimental data for HTR simulation codes, an attempt was made to simulate the whole reactor operation up to the presence. Special emphasis was put on the fission product release behaviour as it is an important safety aspect of such a reactor. The operation history has to be simulated with respect to the neutronics, fluid mechanics and depletion to get a detailed knowledge about the time-dependent nuclide inventory. In this paper we report about such a simulation with VSOP 99/11 and our new fission product release code STACY. While STACY (Source Term Analysis Code System) so far was able to calculate the fission product release rates in case of an equilibrium core and during transients, it now can also be applied to running-in-phases. This coupling demonstrates a first step towards an HCP Prototype. Based on the published power histogram of the HTR-10 and additional information about the fuel loading and shuffling, a coupled neutronics, fluid dynamics and depletion calculation was performed. Special emphasis was put on the complex fuel-shuffling scheme within both VSOP and STACY. The simulations have shown that the HTR-10 up to now generated about 2580 MWd while reshuffling the core about 2.3 times. Within this paper, STACY results for the equilibrium core will be compared with FRESCO-II results being published by INET. Compared to these release rates, which are based on a few user defined life histories, in this new approach the fission product release rates of Ag-110m, Cs-137, Sr-90 and I-131 have been simulated for about 4000 tracer pebbles with STACY. For the calculation of the HTR-10 operation history time-dependent release rates are being presented as well. (author)

MAXDOSE-SR: A routine release atmospheric dose model used at SRS

International Nuclear Information System (INIS)

Simpkins, A.A.

2000-01-01

MAXDOSE-SR is a PC version of the dosimetry code MAXIGASP, which was used to calculate doses to the maximally exposed offsite individual for routine atmospheric releases of radioactive material at the Savannah River Site (SRS). Complete code description, verification of models, and user's manual have been included in this report. Minimal input is required to run the program, and site specific parameters are used when possible

Modeling of fuel performance and fission product release behavior during HTTR normal operation. A comparative study of the FZJ and JAERI modeling approach

International Nuclear Information System (INIS)

Verfondern, Karl; Sumita, Junya; Ueta, Shohei; Sawa, Kazuhiro

2001-03-01

For the prediction of fuel performance and fission product release behavior in the High Temperature Engineering Test Reactor, HTTR of the Japan Atomic Energy Research Institute(JAERI), during its normal operation, calculation tools were applied as have been used at the Research Center Juelich (FZJ) in safety analyses for pebble-bed HTGR designs. Calculations were made assuming the HTTR operation with a nominal operation time of 660 efpd including a 110 efpd period with elevated fuel temperatures. Fuel performance calculations by the PANAMA code with given fuel temperature distribution in the core have shown that the additional failure level of about 5x10 -6 is expected which is about twice as much as the as-fabricated through-coatings failure level. Under the extreme safety design conditions, the predicted particle failure fraction in the core increases to about 1x10 -3 in maximum. The diffusive release of metallic fission products from the fuel primarily occurs in the core layer with the maximum fuel temperature (layer 3) whereas there is hardly any contribution from layer 1 except for the recoil fraction. Silver most easily escapes the fuel; the predicted release fractions from the fuel compacts are 10% (expected) and 50% (safety design). The figures for strontium (expected: 1.5x10 -3 ), safety design: 3.1x10 -2 ) and cesium (5.6x10 -4 , 2.9x10 -2 ) reveal as well a significant fraction to originate already from intact particles. Comparison with the calculation based on JAERI's diffusion model for cesium shows a good agreement for the release behavior from the particles. The differences in the results can be explained mainly by the different diffusion coefficients applied. The release into the coolant can not modelled because of the influence of the gap between compact and graphite sleeve lowering the release by a factor of 3 to 10. For the prediction of performance and fission product release behavior of advanced ZrC TRISO particles, more experimental work is

The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

International Nuclear Information System (INIS)

Starflinger, J.; Scholtyssek, W.; Unger, H.

1994-12-01

For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

Energy Technology Data Exchange (ETDEWEB)

Van Hove, W.; Van Laeken, K.; Bartsoen, L. [Belgatom, Brussels (Belgium)] [and others

1995-09-01

To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with time dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.

131I release from a HTGR during the LOFC accident

International Nuclear Information System (INIS)

Foley, J.E.

1975-03-01

The time-dependent release of 131 I from both the core and the containment building of a high temperature gas-cooled (HTGR) reactor during the loss of forced coolant (LOFC) accident is studied. A simplified core release model is combined with a containment building release model so that the total amount of the isotope released to the environment can be calculated. The time-dependent release of 131 I from the core during the LOFC accident is primarily a function of the time-dependent core temperatures and the failed fuel release constants. The most important factor in calculating the amount of the isotope released to the environment is the total amount released into the containment building. (U.S.)

Modeling of fission product release in integral codes

International Nuclear Information System (INIS)

Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.

2014-01-01

The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)

Modelling for radiological and radioecological consequences of an accidental radionuclide release at Sea

International Nuclear Information System (INIS)

Paschoa, A.S.

1997-01-01

Full text: Scenarios concerning accidental releases of radionuclides into water bodies can be found in the open literature, mostly in connection with nuclear power plants located either onshore or inland. However, meager attention has been given to nuclear reactors used as energy sources for propulsion at sea, which are also subject to accidents. Such potential accidents may involve the loss of part of the reactor core to the surrounding water body. In addition of the initial instantaneous releases, one can estimate delayed source terms based on the rate at which radionuclides are dissolved or leached from the solidified material, like part of the core or structural materials in contact with water. Most of such solidified material might be a mixture of uranium, zirconium, iron, calcium, silica, fission and activation products, and transuranium elements as oxides, forming a glassy type solid. Transport models were used to calculate radionuclide concentrations in water resulting from short and delayed source terms. Oceanographic data used in the calculations were taken either from the open literature or from unclassified reports of the Brazilian Navy, being, however, as generic as possible. Time-dependent concentration functions for radionuclides in aquatic food following an accidental release reflect the net result of intake and elimination processes. However, to avoid the complexities of multiple parameters involved in such processes, the model accounts only for trophic transfer of radionuclides, and yet avoids the necessity of analyzing the details of each transfer step used to determine fish, crustacea, molluscs and seaweed accumulation. Swimming and other aquatic sports are not included in the model used for dose calculations because of theirs relatively low importance in comparison with the pathways concerning ingestion of aquatic food

Release modes and processes relevant to source-term calculations at Yucca Mountain

International Nuclear Information System (INIS)

Apted, M.J.

1994-01-01

The feasibility of permanent disposal of radioactive high-level waste (HLW) in repositories located in deep geologic formations is being studied world-wide. The most credible release pathway is interaction between groundwater and nuclear waste forms, followed by migration of radionuclide-bearing groundwater to the accessible environment. Under hydrologically unsaturated conditions, vapor transport of volatile radionuclides is also possible. The near-field encompasses the waste packages composed of engineered barriers (e.g. man-made materials, such as vitrified waste forms, corrosion-resistant containers), while the far-field includes the natural barriers (e.g. host rock, hydrologic setting). Taken together, these two subsystems define a series of multiple, redundant barriers that act to assure the safe isolation of nuclear waste. In the U.S., the Department of energy (DOE) is investigating the feasibility of safe, long-term disposal of high-level nuclear waste at the Yucca Mountain site in Nevada. The proposed repository horizon is located in non-welded tuffs within the unsaturated zone (i.e. above the water table) at Yucca Mountain. The purpose of this paper is to describe the source-term models for radionuclide release from waste packages at Yucca Mountain site. The first section describes the conceptual release modes that are relevant for this site and waste package design, based on a consideration of the performance of currently proposed engineered barriers under expected and unexpected conditions. No attempt is made to asses the reasonableness nor probability of occurrence for any specific release mode. The following section reviews the waste-form characteristics that are required to model and constrain the release of radionuclides from the waste package. The next section present mathematical models for the conceptual release modes, selected from those that have been implemented into a probabilistic total system assessment code developed for the Electric Power

Comparison between environmental measurements and model calculations of radioactivity in fish at the Swedish nuclear power plants and Studsvik

International Nuclear Information System (INIS)

Karlberg, O.

1995-02-01

Doses to critical groups from the activity released from swedish reactors were modelled in 1983. In this report these calculations are compared to doses calculated (using the same assumptions as in the 1983 model) from the activity measured in the water recipient. The study shows that the model overestimates activity in biota and sediments, which was expected, since the model was constructed to be conservative. 13 refs, 5 figs, 6 tabs

Kinetic Modelling of Drug Release from Pentoxifylline Matrix Tablets based on Hydrophilic, Lipophilic and Inert Polymers

Directory of Open Access Journals (Sweden)

Mircia Eleonora

2015-12-01

Full Text Available Pentoxifylline is a xanthine derivative used in the treatment of peripheral vascular disease, which because of its pharmacokinetic and pharmacologic profile is an ideal candidate for the development of extended release formulations. The aim of this study is to present a kinetic analysis of the pentoxifylline release from different extended release tablets formulations, using mechanistic and empirical kinetic models. A number of 28 formulations were prepared and analysed; the analysed formulations differed in the nature of the matrix forming polymers (hydrophilic, lipophilic, inert and in their concentrations. Measurements were conducted in comparison with the reference product Trental 400 mg (Aventis Pharma. The conditions for the dissolution study were according to official regulations of USP 36: apparatus no. 2, dissolution medium water, volume of dissolution medium is 1,000 mL, rotation speed is 50 rpm, spectrophotometric assay at 274 nm. Six mathematical models, five mechanistic (0 orders, 1st-order release, Higuchi, Hopfenberg, Hixson-Crowell and one empirical (Peppas, were fitted to pentoxifylline dissolution profile from each pharmaceutical formulation. The representative model describing the kinetics of pentoxifylline release was the 1st-order release, and its characteristic parameters were calculated and analysed.

Forecasting consequences of accidental release: how reliable are current assessment models

International Nuclear Information System (INIS)

Rohwer, P.S.; Hoffman, F.O.; Miller, C.W.

1983-01-01

This paper focuses on uncertainties in model output used to assess accidents. We begin by reviewing the historical development of assessment models and the associated interest in uncertainties as these evolutionary processes occurred in the United States. This is followed by a description of the sources of uncertainties in assessment calculations. Types of models appropriate for assessment of accidents are identified. A summary of results from our analysis of uncertainty is provided in results obtained with current methodology for assessing routine and accidental radionuclide releases to the environment. We conclude with discussion of preferred procedures and suggested future directions to improve the state-of-the-art of radiological assessments

Modelling of motorway tunnels scenario for utilization of conditionally released radioactive materials

International Nuclear Information System (INIS)

Hrncir, Tomas; Panik, Michal; Necas, Vladimir

2011-01-01

Considerable amount of solid radioactive waste with radioactivity slightly above the limits for unconditional release is generated during the decommissioning of the nuclear installations. Conditional release deals with precisely this type of materials with activity slightly above limits in order to save considerable financial resources, which would be otherwise spend on treatment, conditioning and disposal of these materials at appropriate repository. The basic principles of conditional release as well as possibilities of reusing of the conditionally released materials are described. One of these possibilities of the reusing was chosen and application proposal of conditional release of metal waste - steel reinforcement in the concrete, which could be used for construction of motorway tunnels, was created. The computer code Visiplan 4.0 3D ALARA planning tool software was used for the calculation of effective individual dose for personnel constructing the tunnel and for critical group related to scenario. Particular models for individual scenarios of conditional release have been developed within the scope of this software code. The aim of the paper is to determine a level of the radioactivity of conditional released materials to avoid over exceeding the value of annual individual effective dose 10μSv/year established by international recommendations. (author)

Review and perspectives on spallings release models in the 1996 performance assessment for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Knowles, M.K; Hansen, F.D.; Thompson, T.W.; Schatz, J.F.; Gross, M.

2000-01-01

The Waste Isolation Pilot Plant was licensed for disposal of transuranic wastes generated by the US Department of Energy. The facility consists of a repository mined in a bedded salt formation, approximately 650 m below the surface. Regulations promulgated by the US Environmental Protection Agency require that performance assessment calculations for the repository include the possibility that an exploratory drilling operation could penetrate the waste disposal areas at some time in the future. Release of contaminated solids could reach the surface during a drilling intrusion. One of the mechanisms for release, known as spallings, can occur if gas pressures in the repository exceed the hydrostatic pressure of a column of drilling mud. Calculation of solids releases for spallings depends critically on the conceptual models for the waste, for the spallings process, and assumptions regarding driller parameters and practices. The paper presents a review of the evolution of these models during regulatory review of the Compliance Certification Application for the repository. A summary and perspectives on the implementation of conservative assumptions in model development are also provided

Review and perspectives on spallings release models in the 1996 performance assessment for the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

Knowles, M.K.; Hansen, F.D.; Thompson, T.W.; Schatz, J.F.; Gross, M.

2000-01-01

The Waste Isolation Pilot Plant was licensed for disposal of transuranic wastes generated by the US Department of Energy. The facility consists of a repository mined in a bedded salt formation, approximately 650 m below the surface. Regulations promulgated by the US Environmental Protection Agency require that performance assessment calculations for the repository include the possibility that an exploratory drilling operation could penetrate the waste disposal areas at some time in the future. Release of contaminated solids could reach the surface during a drilling intrusion. One of the mechanisms for release, known as spallings, can occur if gas pressures in the repository exceed the hydrostatic pressure of a column of drilling mud. Calculation of solids releases for spallings depends critically on the conceptual models for the waste, for the spallings process, and assumptions regarding driller parameters and practices. This paper presents a review of the evolution of these models during the regulatory review of the Compliance Certification Application for the repository. A summary and perspectives on the implementation of conservative assumptions in model development are also provided

Uncertainty analysis of multiple canister repository model by large-scale calculation

International Nuclear Information System (INIS)

Tsujimoto, K.; Okuda, H.; Ahn, J.

2007-01-01

A prototype uncertainty analysis has been made by using the multiple-canister radionuclide transport code, VR, for performance assessment for the high-level radioactive waste repository. Fractures in the host rock determine main conduit of groundwater, and thus significantly affect the magnitude of radionuclide release rates from the repository. In this study, the probability distribution function (PDF) for the number of connected canisters in the same fracture cluster that bears water flow has been determined in a Monte-Carlo fashion by running the FFDF code with assumed PDFs for fracture geometry. The uncertainty for the release rate of 237 Np from a hypothetical repository containing 100 canisters has been quantitatively evaluated by using the VR code with PDFs for the number of connected canisters and the near field rock porosity. The calculation results show that the mass transport is greatly affected by (1) the magnitude of the radionuclide source determined by the number of connected canisters by the fracture cluster, and (2) the canister concentration effect in the same fracture network. The results also show the two conflicting tendencies that the more fractures in the repository model space, the greater average value but the smaller uncertainty of the peak fractional release rate is. To perform a vast amount of calculation, we have utilized the Earth Simulator and SR8000. The multi-level hybrid programming method is applied in the optimization to exploit high performance of the Earth Simulator. The Latin Hypercube Sampling has been utilized to reduce the number of samplings in Monte-Carlo calculation. (authors)

Experiment calculated ascertainment of factors affecting the energy release in IGR reactor core

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Zhotabayev, Zh.R.

2006-01-01

Full text: At present energy supply resources problem is important. Nuclear reactors can, of course, solve this problem, but at the same time there is another issue, concerning safety exploitation of nuclear reactors. That is why, for the last seven years, such experiments as 'Investigation of the processes, conducting severe accidents with core melting' are being carried out at our IGR (impulse graphite reactor) reactor. Leaving out other difficulties of such experiments, it is necessary to notice, that such experiments require more accurate IGR core energy release calculations. The final aim of the present research is verification and correction of the existing method or creation of new method of IGR core energy release calculation. IGR reactor is unique and there is no the same reactor in the world. Therefore, application of the other research reactor methods here is quite useful. This work is based on evaluation of factors affecting core energy release (physical weight of experimental device, different configuration of reactor core, i.e. location of absorbers, initial temperature of core, etc), as well as interference of absorbers group. As it is known, energy release is a value of integral reactor power. During experiments with rays, Reactor power depends on currents of ion production chambers (IPC), located round the core. It is worth to notice that each ion production chamber (IPC) in the same start-up has its own ratio coefficient between IPC current and reactor present power. This task is complicated due to 'IPC current - reactor power' ratio coefficients, that change continuously, probably, because of new loading of experimental facility and different position of control rods. That is why, in order to try about reactor power, before every start-up, we have to re-determine the 'IPC current - reactor power' ratio coefficients for each ion production chamber (IPC). Therefore, the present work will investigate the behavior of ratio coefficient within the

Atmospheric releases from severe nuclear accidents: Environmental transport and pathways to man: Modelling of radiation doses to man from Chernobyl releases

International Nuclear Information System (INIS)

Anspaugh, L.R.; Goldman, M.; Catlin, R.J.

1987-01-01

The Chernobyl accident released a large amount of highly fractionated radioactive debris, including approximately 89 PBq of 137 Cs. We calculated the resulting collective dose commitment to the Northern Hemisphere via the pathways of external exposure and ingestion of radionuclides withd food. We developed a rural/urban model of external dose and we used the PATHWAY model for ingestion. The results are a collective dose commitment of 630,000 person-Gy over the first year and 1,200,000 person-Gy over 50 years. 13 refs., 1 tab

Dispersion modeling of accidental releases of toxic gases - utility for the fire brigades.

Science.gov (United States)

Stenzel, S.; Baumann-Stanzer, K.

2009-09-01

Several air dispersion models are available for prediction and simulation of the hazard areas associated with accidental releases of toxic gases. The most model packages (commercial or free of charge) include a chemical database, an intuitive graphical user interface (GUI) and automated graphical output for effective presentation of results. The models are designed especially for analyzing different accidental toxic release scenarios ("worst-case scenarios”), preparing emergency response plans and optimal countermeasures as well as for real-time risk assessment and management. The research project RETOMOD (reference scenarios calculations for toxic gas releases - model systems and their utility for the fire brigade) was conducted by the Central Institute for Meteorology and Geodynamics (ZAMG) in cooperation with the Viennese fire brigade, OMV Refining & Marketing GmbH and Synex Ries & Greßlehner GmbH. RETOMOD was funded by the KIRAS safety research program of the Austrian Ministry of Transport, Innovation and Technology (www.kiras.at). The main tasks of this project were 1. Sensitivity study and optimization of the meteorological input for modeling of the hazard areas (human exposure) during the accidental toxic releases. 2. Comparison of several model packages (based on reference scenarios) in order to estimate the utility for the fire brigades. For the purpose of our study the following models were tested and compared: ALOHA (Areal Location of Hazardous atmosphere, EPA), MEMPLEX (Keudel av-Technik GmbH), Trace (Safer System), Breeze (Trinity Consulting), SAM (Engineering office Lohmeyer). A set of reference scenarios for Chlorine, Ammoniac, Butane and Petrol were proceed, with the models above, in order to predict and estimate the human exposure during the event. Furthermore, the application of the observation-based analysis and forecasting system INCA, developed in the Central Institute for Meteorology and Geodynamics (ZAMG) in case of toxic release was

Model-based computer-aided design for controlled release of pesticides

DEFF Research Database (Denmark)

Muro Sunè, Nuria; Gani, Rafiqul; Bell, G.

2005-01-01

In the field of controlled release technology for pesticides or active ingredients (AI), models that can predict its delivery during application are important for purposes of design and marketing of the pesticide product. Appropriate models for the controlled release of pesticides, if available, ...... extended models have been developed and implemented into a computer-aided system. The total model consisting of the property models embedded into the release models are then employed to study the release of different combinations of AIs and polymer-based microcapsules.......In the field of controlled release technology for pesticides or active ingredients (AI), models that can predict its delivery during application are important for purposes of design and marketing of the pesticide product. Appropriate models for the controlled release of pesticides, if available...

Applications of energy-release-rate techniques to part-through cracks in plates and cylinders. Volume 2. ORVIRT: a finite element program for energy release rate calculations for 2-dimensional and 3-dimensional crack models

International Nuclear Information System (INIS)

Bass, B.R.; Bryson, J.W.

1983-02-01

Certain studies of fracture phenomena, such as pressurized-thermal-shock of cracked structures, require that crack tip parameters be determined for combined thermal and mechanical loads. A method is proposed here that modifies the isothermal formulation of deLorenzi to account for thermal strains in cracked bodies. The formulation has been implemented in the virtual-crack-extension program ORVIRT (Oak Ridge VIRTual-Crack-Extension). Program ORVIRT performs energy release rate calculations for both 2- and 3-dimensional nonlinear models of crack configurations in engineering structures. Two applications of the ORVIRT program are described. In the first, semielliptical surface cracks in an experimental test vessel are analyzed under elastic-plastic conditions using the finite element method. The second application is a thick-walled test vessel subjected to combined pressure and thermal shock loading

Recoil release of fission products from nuclear fuel

International Nuclear Information System (INIS)

Wise, C.

1985-01-01

An analytical approximation is developed for calculating recoil release from nuclear fuel into gas filled interspaces. This expression is evaluated for a number of interspace geometries and shown to be generally accurate to within about 10% by comparison with numerical calculations. The results are applied to situations of physical interest and it is demonstrated that recoil can be important when modelling fission product release from low temperature CAGR pin failures. Furthermore, recoil can contribute significantly in experiments on low temperature fission product release, particularly where oxidation enhancement of this release is measured by exposing the fuel to CO 2 . The calculations presented here are one way of allowing for this, other methods are suggested. (orig.)

Release model for black liquor droplet; Mustalipeaepisaran vapautumismalli

Energy Technology Data Exchange (ETDEWEB)

Saastamoinen, J. [VTT Energy, Espoo (Finland)

1997-10-01

The release of sodium, potassium, chlorine and sulphur from black liquor droplets during pyrolysis, combustion and gasification is studied by modelling work. A model for drying, pyrolysis and swelling of black liquor has been developed earlier. A submodel for the release of sulphur, which takes place at temperatures below 500 deg C has been incorporated to this model. A previous model for the combustion and gasification of char particles has been further developed to account for the effect of sodium, potassium and chlorine. A model for the release of these components as function of time has been developed. (orig.)

Calculation of gas release from DC and AC arc furnaces in a foundry

Science.gov (United States)

Krutyanskii, M. M.; Nekhamin, S. M.; Rebikov, E. M.

2016-12-01

A procedure for the calculation of gas release from arc furnaces is presented. The procedure is based on the stoichiometric ratios of the oxidation of carbon in liquid iron during the oxidation heat period and the oxidation of iron from a steel charge by oxygen in the period of solid charge melting during the gas exchange of the furnace cavity with the external atmosphere.

Correlation of recent fission product release data

International Nuclear Information System (INIS)

Kress, T.S.; Lorenz, R.A.; Nakamura, T.; Osborne, M.F.

1989-01-01

For the calculation of source terms associated with severe accidents, it is necessary to model the release of fission products from fuel as it heats and melts. Perhaps the most definitive model for fission product release is that of the FASTGRASS computer code developed at Argonne National Laboratory. There is persuasive evidence that these processes, as well as additional chemical and gas phase mass transport processes, are important in the release of fission products from fuel. Nevertheless, it has been found convenient to have simplified fission product release correlations that may not be as definitive as models like FASTGRASS but which attempt in some simple way to capture the essence of the mechanisms. One of the most widely used such correlation is called CORSOR-M which is the present fission product/aerosol release model used in the NRC Source Term Code Package. CORSOR has been criticized as having too much uncertainty in the calculated releases and as not accurately reproducing some experimental data. It is currently believed that these discrepancies between CORSOR and the more recent data have resulted because of the better time resolution of the more recent data compared to the data base that went into the CORSOR correlation. This document discusses a simple correlational model for use in connection with NUREG risk uncertainty exercises. 8 refs., 4 figs., 1 tab

The calculating methods of the release of airborne radionuclides to environment during the normal operation of a module high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Liu Yuanzhong

1993-01-01

The calculations of the release of radionuclides to environment are the basis of environmental impact assessment during the normal operation of a module high temperature gas-cooled reactor of the Institute of Nuclear Energy Technology, Tsinghua University, China. According to the features of the reactor it is pointed out that only five sources of the airborne radioactive materials released to environment are important. They are: (1) the activation of the air in the reactor cavity; (2) the escape from the primary coolant systems; (3) the release of radioactively contaminated helium from storage tanks; (4) the release of radioactively contaminated helium from the gas evacuation system of fuel load and unload system; (5) the leakage of the vapour from water-steam loop. In accordance with five release sources the calculating methods of radionuclides released to environment are worked out respectively and the respective calculating formulas are derived for the normal operation of the reactor

A model for short and medium range dispersion of radionuclides released to the atmosphere

International Nuclear Information System (INIS)

Clarke, R.H.

1979-09-01

A Working Group was established to give practical guidance on the estimation of the dispersion of radioactive releases to the atmosphere. The dispersion is estimated in the short and medium range, that is from about 100 m to a few tens of kilometres from the source, and is based upon a Gaussian plume model. A scheme is presented for categorising atmospheric conditions and values of the associated dispersion parameters are given. Typical results are presented for releases in specific meteorological conditions and a scheme is included to allow for durations of release of up to 24 hours. Consideration has also been given to predicting longer term average concentrations, typically annual averages, and results are presented which facilitate site specific calculations. The results of the models are extended to 100 km from the source, but the increasing uncertainty with which results may be predicted beyond a few tens of kilometres from the source is emphasised. Three technical appendices provide some of the rationale behind the decisions made in adopting the various models in the proposed dispersion scheme. (author)

The 2017 Release Cloudy

Science.gov (United States)

Ferland, G. J.; Chatzikos, M.; Guzmán, F.; Lykins, M. L.; van Hoof, P. A. M.; Williams, R. J. R.; Abel, N. P.; Badnell, N. R.; Keenan, F. P.; Porter, R. L.; Stancil, P. C.

2017-10-01

We describe the 2017 release of the spectral synthesis code Cloudy, summarizing the many improvements to the scope and accuracy of the physics which have been made since the previous release. Exporting the atomic data into external data files has enabled many new large datasets to be incorporated into the code. The use of the complete datasets is not realistic for most calculations, so we describe the limited subset of data used by default, which predicts significantly more lines than the previous release of Cloudy. This version is nevertheless faster than the previous release, as a result of code optimizations. We give examples of the accuracy limits using small models, and the performance requirements of large complete models. We summarize several advances in the H- and He-like iso-electronic sequences and use our complete collisional-radiative models to establish the densities where the coronal and local thermodynamic equilibrium approximations work.

Fission product release from nuclear fuel I. Physical modelling in the ASTEC code

International Nuclear Information System (INIS)

Brillant, G.; Marchetto, C.; Plumecocq, W.

2013-01-01

Highlights: • Physical modeling of FP and SM release in ASTEC is presented. • The release is described as solid state diffusion within fuel for high volatile FP. • The release is described as FP vaporisation for semi volatile FP. • The release is described as fuel vaporisation for low volatile FP. • ASTEC validation is presented in the second paper. - Abstract: This article is the first of a series of two articles dedicated to the mechanisms of fission product release from a degraded core as they are modelled in the ASTEC code. The ASTEC code aims at simulating severe accidents in nuclear reactors from the initiating event up to the radiological consequences on the environment. This code is used for several applications such as nuclear plant safety evaluation including probabilistic studies and emergency preparedness. To cope with the requirements of robustness and low calculation time, the code is based on a semi-empirical approach and only the main limiting phenomena that govern the release from intact rods and from debris beds are considered. For solid fuel, fission products are classified into three groups, depending on their degree of volatility. The kinetics of volatile fission products release depend on the rate-limiting process of solid-state diffusion through fuel grains. For semi-volatile fission products, the release from the open fuel porosities is assumed to be governed by vaporisation and mass transfer processes. The key phenomenon for the release of low volatile fission products is supposed to be fuel volatilisation. A similar approach is used for the release of fission products from a rubble bed. An in-depth validation of the code including both analytical and integral experiments is the subject of the second article

Derived release limits for airborne effluents at TRIGA - INR Reactor

International Nuclear Information System (INIS)

Toma, A.; Dulama, C.; Hirica, O.; Mihai, S.; Oprea, I.

2008-01-01

Beginning from fulfilling the purposes of dose limitation system recommended by ICRP, and now accepted in radiation protection, this paper presents an environmental transfer model to calculate derived release limits for airborne and gaseous radioactive effluents at TRIGA-INR, 14 MW Steady State Reactor, in function on INR-Pitesti site. The methodology consists in determination of the principal exposure pathways for different groups of population and dose calculations for each radionuclide. The characterization of radionuclides transfer to environment was made using the compartmental model. The parameter transfer concept was used to describe the distribution of radionuclides between the different compartments. Atmospheric dispersion was very carefully treated, because it is the primary mechanism of the transfer of radionuclides in the environment and it determines all exposure pathways. Calculation of the atmospheric dispersion was made using ORION-II computer code based on the Gaussian plume model which takes account of site's specific climate and relief conditions. Default values recommended by literature were used to calculate some of the parameters when specific site values were not available. After identification of all transfer parameters which characterize the most important exposure pathways, the release rate corresponding to the individual dose rate limit was calculated. This maximum release rate is the derived release limit for each radionuclide and source. In the paper, the derived release limits are calculated for noble gases, radioiodine and other airborne particulate radionuclides, which can be released on the TRIGA-INR reactor stack, and are important to radiation protection. (authors)

Selection of Hydrological Model for Waterborne Release

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

Following a request from the States of South Carolina and Georgia, downstream radiological consequences from postulated accidental aqueous releases at the three Savannah River Site nonreactor nuclear facilities will be examined. This evaluation will aid in determining the potential impacts of liquid releases to downstream populations on the Savannah River. The purpose of this report is to evaluate the two available models and determine the appropriate model for use in following waterborne release analyses. Additionally, this report will document the accidents to be used in the future study

Nuclide release from the near-field of a L/ILW repository

International Nuclear Information System (INIS)

Karlsson, L.G.; Hoeglund, L.O.; Pers, K.

1986-12-01

For Project Gewaehr 1985, the release of nuclides from a repository for low- and intermediate-level radioactive waste is calculated. The calculations are made for a reference design repository located in the marl host rock at the Oberbauen Stock reference site. The results are limited to the release of the nuclides from the waste through the engineered barriers into the surrounding host rock and will, therefore, constitute a source term for the far-field and biosphere calculations. The most probable nuclide transport mechanism is diffusion and releases are thus influenced by the nuclide diffusivities in the barriers, nuclide sorption and nuclide solubility limits. Degradation of the engineered concrete barriers is taken into account. The effects of convective flow through the barriers are described elsewhere. A near-field release model is presented. It consists of a set of computer programs suited to handel different repository designs, solubility limitations and the different waste categories. The release calculations were made for a base case in which best estimates of the parameters were used. Sensitivity to the choice of the most important parameters was tested by parameter variations. The numerical models used were checked by comparative calculations with different codes and similar data. The results of the base calculations show that near-field barriers will cause both a delay of the release to the far-field and a reduced rate of release. The sorbed nuclides, comprising the actinides and some activation and fission products, will be delayed by 10'000 years and have a maximum release rate of less than 10 -3 Ci/a each. The non-sorbed nuclides are delayed by only about 100 years and the maximum release rate is less than 10 -2 Ci per year and nuclide. The parameter variations and the design model tests gave only limited deviations from the base case results. (author)

A model to calculate exposure from radioactive discharges into the coastal waters of Northern Europe

International Nuclear Information System (INIS)

Clark, M.J.; Grimwood, P.D.; Camplin, W.C.

1980-11-01

A regional marine model is described which can be used to estimate the exposure of populations as a result of the discharge of radioactive effluents into the coastal waters of Northern Europe. The model simulates the dispersion of radionuclides in marine waters, Their interaction with marine sediments and the concentration mechanisms occurring in seafoods. There is a local/regional interface defined in the modelling approach whereby releases are assumed to first enter a local marine compartment prior to widespread dispersion in coastal waters. Depletion mechanisms operate within both the local and regional environments influencing the fraction of radionuclide release which contributes to collective exposure. General results of the regional marine model are presented in a form which can be combined with independent local marine models; collective intakes per unit release of various radionuclides into coastal waters are given for a series of integration times. For caesium-137 and plutonium-239 collective effective dose equivalent commitments have been calculated using a defined local marine model. Some general conclusions have been drawn from the results and there is some discussion of the various features of the modelling approach. (author)

The role of a detailed aqueous phase source release model in the LANL area G performance assessment

Energy Technology Data Exchange (ETDEWEB)

Vold, E.L.; Shuman, R.; Hollis, D.K. [Los Alamos National Lab., NM (United States)] [and others

1995-12-31

A preliminary draft of the Performance Assessment for the Los Alamos National Laboratory (LANL) low-level radioactive waste disposal facility at Area G is currently being completed as required by Department of Energy orders. A detailed review of the inventory data base records and the existing models for source release led to the development of a new modeling capability to describe the liquid phase transport from the waste package volumes. Nuclide quantities are sorted down to four waste package release categories for modeling: rapid release, soil, concrete/sludge, and corrosion. Geochemistry for the waste packages was evaluated in terms of the equilibrium coefficients, Kds, and elemental solubility limits, Csl, interpolated from the literature. Percolation calculations for the base case closure cover show a highly skewed distribution with an average of 4 mm/yr percolation from the disposal unit bottom. The waste release model is based on a compartment representation of the package efflux, and depends on package size, percolation rate or Darcy flux, retardation coefficient, and moisture content.

Evaporation release behavior of volatile fission products from liquid sodium pool to the inert cover gas

Energy Technology Data Exchange (ETDEWEB)

Nakagiri, T; Miyahara, S [Oarai Engineering Center, Power Reactor and Nuclear Fuel Development Corp., Oaraimachi, Ibaraki (Japan)

1996-12-01

In fuel failure of sodium cooled fast breeder reactors, released volatile fission products (VFPs) such as iodine and cesium from the fuel will be dissolved into the liquid sodium coolant and transferred to the cover vaporization. In the cover gas system of the reactor, natural convection occurs due to temperature differences between the sodium pool and the gas phase. The release rates of VFPs together with sodium vaporization are considered to be controlled by the convection. In this study, three analytical models are developed and examined to calculate the transient release rates using the equilibrium partition coefficients of VFPs. The calculated release rates are compared with experimental results for sodium and sodium iodide. The release rate of sodium is closest to the calculation by the heterogeneous nucleation theory. The release rate of sodium iodide obtained from the experiment is between the release rates calculated by the model based on heat-and-mass transfer analogy and the Hill`s theory. From this study, it is confirmed that the realistic release rate of sodium is able to be calculated by the model based on the heterogeneous nucleation theory. The conservative release rate of sodium iodide is able to be calculated by the model based on the Hill`s theory using the equilibrium partition coefficient of sodium iodide. (author) 7 figs., 1 tab., 3 refs.

Computational modeling and analysis of iron release from macrophages.

Directory of Open Access Journals (Sweden)

Alka A Potdar

2014-07-01

Full Text Available A major process of iron homeostasis in whole-body iron metabolism is the release of iron from the macrophages of the reticuloendothelial system. Macrophages recognize and phagocytose senescent or damaged erythrocytes. Then, they process the heme iron, which is returned to the circulation for reutilization by red blood cell precursors during erythropoiesis. The amount of iron released, compared to the amount shunted for storage as ferritin, is greater during iron deficiency. A currently accepted model of iron release assumes a passive-gradient with free diffusion of intracellular labile iron (Fe2+ through ferroportin (FPN, the transporter on the plasma membrane. Outside the cell, a multi-copper ferroxidase, ceruloplasmin (Cp, oxidizes ferrous to ferric ion. Apo-transferrin (Tf, the primary carrier of soluble iron in the plasma, binds ferric ion to form mono-ferric and di-ferric transferrin. According to the passive-gradient model, the removal of ferrous ion from the site of release sustains the gradient that maintains the iron release. Subcellular localization of FPN, however, indicates that the role of FPN may be more complex. By experiments and mathematical modeling, we have investigated the detailed mechanism of iron release from macrophages focusing on the roles of the Cp, FPN and apo-Tf. The passive-gradient model is quantitatively analyzed using a mathematical model for the first time. A comparison of experimental data with model simulations shows that the passive-gradient model cannot explain macrophage iron release. However, a facilitated-transport model associated with FPN can explain the iron release mechanism. According to the facilitated-transport model, intracellular FPN carries labile iron to the macrophage membrane. Extracellular Cp accelerates the oxidation of ferrous ion bound to FPN. Apo-Tf in the extracellular environment binds to the oxidized ferrous ion, completing the release process. Facilitated-transport model can

Model for incorporating fuel swelling and clad shrinkage effects in diffusion theory calculations (LWBR Development Program)

International Nuclear Information System (INIS)

Schick, W.C. Jr.; Milani, S.; Duncombe, E.

1980-03-01

A model has been devised for incorporating into the thermal feedback procedure of the PDQ few-group diffusion theory computer program the explicit calculation of depletion and temperature dependent fuel-rod shrinkage and swelling at each mesh point. The model determines the effect on reactivity of the change in hydrogen concentration caused by the variation in coolant channel area as the rods contract and expand. The calculation of fuel temperature, and hence of Doppler-broadened cross sections, is improved by correcting the heat transfer coefficient of the fuel-clad gap for the effects of clad creep, fuel densification and swelling, and release of fission-product gases into the gap. An approximate calculation of clad stress is also included in the model

Uncertainties in modeling of consequences of tritium release from fusion reactors. Plasma Fusion Center No. PFC/TR-79-5

International Nuclear Information System (INIS)

Piet, S.J.; Kazimi, M.S.

1979-07-01

The bases for various models concerned with all phases of estimating doses from routine tritium releases from fusion reactors have been examined. The implications of uncertainties in parameters and assumptions for the uncertainty of the calculated doses and resulting maximum permissible releases are presented. Global dispersion models are most affected by the assumptions made concerning movement, such as the role of the ocean as a sink. Dose models were generally found to agree within a factor of two, with the largest variation due to agricultural data. Plant tritium flow studies are the least developed and require substantial improvement in the data base. Based on two possible arbitrary global standards, the maximum allowable releases were found to range from 1.6 to 20,000 Ci/day. The local criteria imply releases between 5 and 20 Ci/day

Estimating risk using bounding calculations and limited data

International Nuclear Information System (INIS)

COWLEY, W.L.

1999-01-01

This paper describes a methodology for estimating the potential risk to workers and the public from igniting organic solvents in any of the 177 underground waste storage tanks at the Hanford Site in southeastern Washington state. The Hanford Site is one of the U.S. Department of Energy's former production facilities for nuclear materials. The tanks contain mixed radioactive wastes. Risk is measured by calculating toxicological and radiological accident consequences and frequencies and comparing the results to established regulatory guidelines. Available sample data is insufficient to adequately characterize the waste and solvent, so a model that maximizes releases from the tanks (bounding case) is used. Maximizing releases (and thus consequences) is a standard technique used in safety analysis to compensate for lack of information. The model predicts bounding values of fire duration, the time at which the fire extinguishes because of lack of oxygen, and a pressure history of a fire in a tank. The model output is used to calculate mass and volume release rates of material from the tanks. The mass and volume release rates permit calculation of radiological and toxicological consequences. The resulting consequence calculations demonstrate that risk from an organic solvent fire in the tanks is within regulatory guidelines

Selection of Hydrological Model for Waterborne Release

International Nuclear Information System (INIS)

Blanchard, A.

1999-01-01

This evaluation will aid in determining the potential impacts of liquid releases to downstream populations on the Savannah River. The purpose of this report is to evaluate the two available models and determine the appropriate model for use in following waterborne release analyses. Additionally, this report will document the Design Basis and Beyond Design Basis accidents to be used in the future study

LOFC fission product release and circulating activity calculations for gas-cooled reactors

International Nuclear Information System (INIS)

Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

1977-01-01

The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

Modelling of drug release from ensembles of aspirin microcapsules ...

African Journals Online (AJOL)

Purpose: In order to determine the drug release profile of an ensemble of aspirin crystals or microcapsules from its particle distribution a mathematical model that considered the individual release characteristics of the component single particles was developed. The model assumed that under sink conditions the release ...

Improvement of the model for surface process of tritium release from lithium oxide

International Nuclear Information System (INIS)

Yamaki, Daiju; Iwamoto, Akira; Jitsukawa, Shiro

2000-01-01

Among the various tritium transport processes in lithium ceramics, the importance and the detailed mechanism of surface reactions remain to be elucidated. The dynamic adsorption and desorption model for tritium desorption from lithium ceramics, especially Li 2 O was constructed. From the experimental results, it was considered that both H 2 and H 2 O are dissociatively adsorbed on Li 2 O and generate OH - on the surface. In the first model developed in 1994, it was assumed that either the dissociative adsorption of H 2 or H 2 O on Li 2 O generates two OH - on the surface. However, recent calculation results show that the generation of one OH - and one H - is more stable than that of two OH - s by the dissociative adsorption of H 2 . Therefore, assumption of H 2 adsorption and desorption in the first model is improved and the tritium release behavior from Li 2 O surface is evaluated again by using the improved model. The tritium residence time on the Li 2 O surface is calculated using the improved model, and the results are compared with the experimental results. The calculation results using the improved model agree well with the experimental results than those using the first model

Building an Efficient Model for Afterburn Energy Release

Energy Technology Data Exchange (ETDEWEB)

Alves, S; Kuhl, A; Najjar, F; Tringe, J; McMichael, L; Glascoe, L

2012-02-03

Many explosives will release additional energy after detonation as the detonation products mix with the ambient environment. This additional energy release, referred to as afterburn, is due to combustion of undetonated fuel with ambient oxygen. While the detonation energy release occurs on a time scale of microseconds, the afterburn energy release occurs on a time scale of milliseconds with a potentially varying energy release rate depending upon the local temperature and pressure. This afterburn energy release is not accounted for in typical equations of state, such as the Jones-Wilkins-Lee (JWL) model, used for modeling the detonation of explosives. Here we construct a straightforward and efficient approach, based on experiments and theory, to account for this additional energy release in a way that is tractable for large finite element fluid-structure problems. Barometric calorimeter experiments have been executed in both nitrogen and air environments to investigate the characteristics of afterburn for C-4 and other materials. These tests, which provide pressure time histories, along with theoretical and analytical solutions provide an engineering basis for modeling afterburn with numerical hydrocodes. It is toward this end that we have constructed a modified JWL equation of state to account for afterburn effects on the response of structures to blast. The modified equation of state includes a two phase afterburn energy release to represent variations in the energy release rate and an afterburn energy cutoff to account for partial reaction of the undetonated fuel.

Nuclide release calculation in the near-field of a reference HLW repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

2004-01-01

The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

Radiological impact from spanish coal power plants.2.- Dispersion model for deconcentration calculations

International Nuclear Information System (INIS)

Alvarez, M.C.; Garzon, L.

1990-01-01

In this paper a practical dispersion model is presented, which permits to calculate, in Spain, the concentration of natural radionuclides released to the atmosphere from coal power plants. To apply the model it is necessary to know the following data: emission rates, dry deposition velocity, scavenging coefficient, mixing layer height, together with climatological frequency data relating to wind speed and wind direction (to determinate trajectories from a given source) in the areas examined. Meteorological data can be obtained from meteorological stations across Spain. (Author)

HTR fuel modelling with the ATLAS code. Thermal mechanical behaviour and fission product release assessment

International Nuclear Information System (INIS)

Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent

2009-01-01

To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more

Relocation impacts of a major release from SRTC

International Nuclear Information System (INIS)

Blanchard, A.; Thompson, E.A.; Thompson, J.M.

1999-01-01

The relocation impacts of an accidental release, scenario 1-RD-3 , are evaluated for the Savannah River Technology Center. The extent of the area potentially contaminated to a level that would result in doses exceeding the relocation protective action guide(PAG)is calculated. The maximum calculated distance downwind from the accident at which the relocation PAG is exceeded is also determined. The consequences of the particulate portion of the release are evaluated using the HOTSPOT model and an EXCEL spreadsheet. The consequences of the tritium release are evaluated using UFOTRI

Precipitates/Salts Model Sensitivity Calculation

International Nuclear Information System (INIS)

Mariner, P.

2001-01-01

The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation on potential seepage waters within a potential repository drift. This work is developed and documented using procedure AP-3.12Q, ''Calculations'', in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The specific objective of this calculation is to examine the sensitivity and uncertainties of the Precipitates/Salts model. The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b). The calculation in the current document examines the effects of starting water composition, mineral suppressions, and the fugacity of carbon dioxide (CO 2 ) on the chemical evolution of water in the drift

POPFOOD - a computer code for calculating ingestion collective doses from continuous atmospheric releases

International Nuclear Information System (INIS)

Hotson, J.; Stacey, A.; Nair, S.

1980-07-01

The basic methodology incorporated in the POPFOOD computer code is described, which may be used to calculate equilibrium collective dose rates associated with continuous atmospheric releases and arising from consumption of a broad range of food products. The standard data libraries associated with the code are also described. These include a data library, based on the 1972 agricultural census, describing the spatial distribution of production, in England, Wales and Scotland, of the following food products: milk; beef and veal; pork bacon and ham; poultrymeat; eggs; mutton and lamb; root vegetables; green vegetables; fruit; cereals. Illustrative collective dose calculations were made for the case of 1 Ci per year emissions of 131 I, tritium and 14 C from a typical rural UK site. The calculations indicate that the ingestion pathway results in a greater collective dose than that via inhalation, with the contributions from consumption of root and green vegetables, and cereals being of comparable significance to that from liquid milk consumption, in all three cases. (author)

Extension and validation of ARTM (atmospheric radionuclide transportation model) for the application as dispersion calculation model in AVV (general administrative provision) and SBG (incident calculation bases); Erweiterung und Validierung von ARTM fuer den Einsatz als Ausbreitungsmodell in AVV und SBG

Energy Technology Data Exchange (ETDEWEB)

Martens, Reinhard; Bruecher, Wenzel; Richter, Cornelia; Sentuc, Florence; Sogalla, Martin; Thielen, Harald

2012-02-15

In the medium-term time scale the Gaussian plume model used so far for atmospheric dispersion calculations in the General Administrative Provision (AVV) relating to Section 47 of the Radiation Protection Ordinance (StrISchV) as well as in the Incident Calculation Bases (SGB) relating to Section 49 StrISchV is to be replaced by a Lagrangian particle model. Meanwhile the Atmospheric Radionuclide Transportation Model (ARTM) is available, which allows the simulation of the atmospheric dispersion of operational releases from nuclear installations. ARTM is based on the program package AUSTAL2000 which is designed for the simulation of atmospheric dispersion of nonradioactive operational releases from industrial plants and was adapted to the application of airborne radioactive releases. In the context of the research project 3608S05005 possibilities for an upgrade of ARTM were investigated and implemented as far as possible to the program system. The work program comprises the validation and evaluation of ARTM, the implementation of technical-scientific extensions of the model system and the continuation of experience exchange between developers and users. In particular, the suitability of the model approach for simulations of radiological consequences according to the German SBG and the representation of the influence of buildings typical for nuclear power stations have been validated and further evaluated. Moreover, post-processing modules for calculation of dose-relevant decay products and for dose calculations have been developed and implemented. In order to continue the experience feedback and exchange, a web page has been established and maintained. Questions by users and other feedback have been dealt with and a common workshop has been held. The continued development and validation of ARTM has strengthened the basis for applications of this model system in line with the German regulations AVV and SBG. Further activity in this field can contribute to maintain and

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

International Nuclear Information System (INIS)

Moiseyev, A.V.

2008-01-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

2008-07-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

A real-time PUFF-model for accidental releases in complex terrain

International Nuclear Information System (INIS)

Thykier-Nielsen, S.; Mikkelsen, T.; Larsen, S.E.; Troen, I.; Baas, A.F. de; Kamada, R.; Skupniewicz, C.; Schacher, G.

1990-01-01

LINCOM-RIMPUFF, a combined flow/puff model, was developed at Riso National Laboratory for the Vandenberg AFB Meteorology and Plume Dispersion Handbook and is suitable as is for real time response to emergency spills and vents of gases and radionuclides. LINCOM is a linear, diagnostic, spectral, potential flow model which extends the Jackson-Hunt theory of non-hydrostatic, adiabatic wind flow over hills to the mesoscale domain. It is embedded in a weighted objective analysis (WOA) of real-time Vandenberg tower winds and may be used in ultra-high speed lookup table mode. The mesoscale dispersion model RIMPUFF is a flexible Gaussian puff model equipped with computer-time effective features for terrain and stability-dependent dispersion parameterization, plume rise formulas, inversion and ground-level reflection capabilities and wet/dry (source) depletion. It can treat plume bifurcation in complex terrain by using a puff-splitting scheme. It allows the flow-model to compute the larger scale wind field, reserving turbulent diffusion calculations for the sub-grid scale. In diagnostic mode toxic exposure are well assessed via the release of a single initial puff. With optimization, processing time for RIMPUFF should be on the order of 2 CPU minutes or less on a PC-system. In prognostic mode with shifting winds, multiple puff releases may become necessary, thereby lengthening processing time

Implementation of a model of atmospheric dispersion and dose calculation in the release of radioactive effluents in the Nuclear Centre

International Nuclear Information System (INIS)

Cruz L, C. A.

2015-01-01

In the present thesis, the software DERA (Dispersion of Radioactive Effluents into the Atmosphere) was developed in order to calculate the equivalent dose, external and internal, associated with the release of radioactive effluents into the atmosphere from a nuclear facility. The software describes such emissions in normal operation, and not considering the exceptional situations such as accidents. Several tools were integrated for describing the dispersion of radioactive effluents using site meteorological information (average speed and wind direction and the stability profile). Starting with the calculation of the concentration of the effluent as a function of position, DERA estimates equivalent doses using a set of EPA s and ICRP s coefficients. The software contains a module that integrates a database with these coefficients for a set of 825 different radioisotopes and uses the Gaussian method to calculate the effluents dispersion. This work analyzes how adequate is the Gaussian model to describe emissions type -puff-. Chapter 4 concludes, on the basis of a comparison of the recommended correlations of emissions type -puff-, that under certain conditions (in particular with intermittent emissions) it is possible to perform an adequate description using the Gaussian model. The dispersion coefficients (σ y and σ z ), that using the Gaussian model, were obtained from different correlations given in the literature. Also in Chapter 5 is presented the construction of a particular correlation using Lagrange polynomials, which takes information from the Pasquill-Gifford-Turner curves (PGT). This work also contains a state of the art about the coefficients that relate the concentration with the equivalent dose. This topic is discussed in Chapter 6, including a brief description of the biological-compartmental models developed by the ICRP. The software s development was performed using the programming language Python 2.7, for the Windows operating system (the XP

Use of ELOCA.Mk5 to calculate transient fission product release from CANDU fuel elements

International Nuclear Information System (INIS)

Walker, J.R.; de Vaal, J.W.; Arimescu, V.I.; McGrady, T.G.; Wong, C.

1992-04-01

A change in fuel element power output, or a change in heat transfer conditions, will result in an immediate change in the temperature distribution in a fuel element. The temperature distribution change will be accompanied by concomitant changes in fuel stress distribution that lead, in turn, to a release of fission products to the fuel-to-sheath gap. It is important to know the inventory of fission products in the fuel-to-sheath gap, because this inventory is a major component of the source term for many postulated reactor accidents. ELOCA.Mk5 is a FORTRAN-77 computer code that has been developed to estimate transient releases to the fuel-to-sheath gap in CANDU reactors. ELOCA.Mk5 is an integration of the FREEDOM fission product release model into the ELOCA fuel element thermo-mechanical code. The integration of FREEDOM into ELOCA allows ELOCA.Mk5 to model the feedback mechanisms between the fission product release and the thermo-mechanical response of the fuel element. This paper describes the physical model, gives details of the ELOCA.Mkt code, and describes the validation of the model. We demonstrate that the model gives good agreement with experimental results for both steady state and transient conditions

Modeling Drug-Carrier Interaction in the Drug Release from Nanocarriers

Directory of Open Access Journals (Sweden)

Like Zeng

2011-01-01

Full Text Available Numerous nanocarriers of various compositions and geometries have been developed for the delivery and release of therapeutic and imaging agents. Due to the high specific surface areas of nanocarriers, different mechanisms such as ion pairing and hydrophobic interaction need to be explored for achieving sustained release. Recently, we developed a three-parameter model that considers reversible drug-carrier interaction and first-order drug release from liposomes. A closed-form analytical solution was obtained. Here, we further explore the ability of the model to capture the release of bioactive molecules such as drugs and growth factors from various nanocarriers. A parameter study demonstrates that the model is capable of resembling major categories of drug release kinetics. We further fit the model to 60 sets of experimental data from various drug release systems, including nanoparticles, hollow particles, fibers, and hollow fibers. Additionally, bootstrapping is used to evaluate the accuracy of parameter determination and validate the model in selected cases. The simplicity and universality of the model and the clear physical meanings of each model parameter render the model useful for the design and development of new drug delivery systems.

Evaluating risk using bounding calculations and limited data; TOPICAL

International Nuclear Information System (INIS)

COWLEY, W.L.

1999-01-01

This paper describes a methodology for estimating the potential risk to workers and the public from igniting organic solvents in any of the 177 underground waste storage tanks at the Hanford Site in southeastern Washington state. The Hanford Site is one of the U.S. Department of Energy's former production facilities for nuclear materials. The tanks contain mixed radioactive wastes. Risk is measured by calculating toxicological and radiological accident consequences and frequencies and comparing the results to established regulatory guidelines. Available sample data is insufficient to adequately characterize the waste and solvent, so a model that maximizes releases from the tanks (bounding case) is used. Maximizing releases (and thus consequences) is a standard technique used in safety analysis to compensate for lack of information. The model predicts bounding values of fire duration, the time at which the fire extinguishes because of lack of oxygen, and a pressure history of a fire in a tank. The model output is used to calculate mass and volume release rates of material from the tanks. The mass and volume release rates permit calculation of radiological and toxicological consequences. The resulting consequence calculations demonstrate that risk from an organic solvent fire in the tanks is within regulatory guidelines

ANDOSE: a computer code for calculating annual doses to man from routine releases of LWR effluents for the purpose of evaluating compliance with JAEC's guide for doseobjectives

International Nuclear Information System (INIS)

Iijima, Toshinori; Shiraishi, Tadao

1979-10-01

For environmental doses from routine releases of LWRs effluents to meet the Criterion 'As Low As is Practicable (ALAP)', Japan Atomic Energy Commission (JAEC) established a series of guides, the first for 'Dose Objectives' (May 1975), the second for models and parameters for calculating the environmental doses to compare with the 'Dose Objectives' (September 1976), and the third providing onsite meteorological programs, statistics of the data obtained and atmospheric dispersion models (June 1977). JAERI has developed a computer code, designated as ANDOSE, for calculating annual releases of radioactive gaseous and liquid effluents and, then, total body doses and thyroid doses to individuals around sites on the basis of these guides. The total body doses are from radioactive noble gases as well as from radioactive materials taken with marine food. For the calculation of thyroid doses are taken into account exposure pathways via inhalation and ingestion of leafy vegetables, cow's milk and marine food. The age-specific thyroid doses are evaluated. The doses are summed up when multisource or multisite conditions need to be evaluated (Nuclear Safety Bureau's requirement). In the present report, are described source-term models, environmental transport models and dose models used in the code, of which most are provided in the guides but some are complemented by the authors, the functions of ANDOSE and the manual for users of the code. The program lists and the latter two guides mentioned above are included in the appendices. (author)

Calculation method for gamma-dose rates from spherical puffs

International Nuclear Information System (INIS)

Thykier-Nielsen, S.; Deme, S.; Lang, E.

1993-05-01

The Lagrangian puff-models are widely used for calculation of the dispersion of atmospheric releases. Basic output from such models are concentrations of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on semi-infinite cloud model. This method is however only applicable for points far away from the release point. The exact calculation of the cloud dose using the volume integral requires significant computer time. The volume integral for the gamma dose could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor due to the fact that the same correction factors are used for all isotopes. The authors describe a more elaborate correction method. This method uses precalculated values of the gamma-dose rate as a function of the puff dispersion parameter (δ p ) and the distance from the puff centre for four energy groups. The release of energy for each radionuclide in each energy group has been calculated and tabulated. Based on these tables and a suitable interpolation procedure the calculation of gamma doses takes very short time and is almost independent of the number of radionuclides. (au) (7 tabs., 7 ills., 12 refs.)

The Chernobyl I-131 Release: Model Validation and Assessment of the Countermeasure Effectiveness. Report of the Chernobyl 131I Release Working Group of EMRAS Theme 1

International Nuclear Information System (INIS)

2012-01-01

The Chernobyl 131 I Release Working Group (IWG) which was established within the framework of the EMRAS Programme continues some of the more traditional work of previous international programmes that were aimed at increasing confidence in methods and models for the assessment of radiation exposure related to the environmental releases. There is still very little information regarding the quantitative relationship between radiation dose to the thyroid from Chernobyl and the risk of thyroid cancer. The uncertainty combined with individual estimates of radiation dose constitutes a crucial point in establishing this relationship, since, any release of radioiodine into environment creates wide range of uncertainty for internal dose assessments. The 131 I scenarios provide an excellent opportunity to compare a number of modelling approaches to a single assessment problem, in a dose reconstruction context. Nine experts in environmental modelling participated in the Plavsk Scenario dealing with areas of assessment modelling for which the capabilities are not yet well established. One could observes the remarkably improvement in models performance comparing with previous radioiodine scenarios. Predictions of the various models were within a factor of three of the observations, discrepancies between the estimates of average doses to thyroid produced by most participant not exceeded a factor of ten. The process of testing independent model calculations against independent data set also provided useful information to the originators of the test data.

Local-scale modelling of the releases of 137-Cs and 90-Sr from Fukushima NPP into the Pacific Ocean

International Nuclear Information System (INIS)

Raul, Perianez

2013-01-01

The dispersion of 137-Cs and 90-Sr released from Fukushima nuclear power plant to the sea after the March 11th 2011 tsunami has been studied using a numerical model. The 3D dispersion model consists of an advection/diffusion equation with terms describing uptake/release reactions between water and seabed sediments. The dispersion model has been fed with daily currents provided by JCOPE2 ocean model. Seabed sediment 137-Cs computed patterns have been compared with observations. The impact of tides and of atmospheric deposition on sediment contamination has been evaluated as well. First simulations carried out for Sr-90 are described. An evaluation of the amount of this radionuclide released to the sea has been made using a numerical model for the first time. Calculated vertical profiles of this radionuclide have been compared with measured ones. Finally, the variability in 90-Sr/137-Cs activity ratios has been analysed with the model

Calculation method for gamma dose rates from Gaussian puffs

Energy Technology Data Exchange (ETDEWEB)

Thykier-Nielsen, S; Deme, S; Lang, E

1995-06-01

The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E{sub {gamma}}, {sigma}{sub y}, the asymmetry factor - {sigma}{sub y}/{sigma}{sub z}, the height of puff center - H and the distance from puff center R{sub xy}. To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs.

Calculation method for gamma dose rates from Gaussian puffs

International Nuclear Information System (INIS)

Thykier-Nielsen, S.; Deme, S.; Lang, E.

1995-06-01

The Lagrangian puff models are widely used for calculation of the dispersion of releases to the atmosphere. Basic output from such models is concentration of material in the air and on the ground. The most simple method for calculation of the gamma dose from the concentration of airborne activity is based on the semi-infinite cloud model. This method is however only applicable for puffs with large dispersion parameters, i.e. for receptors far away from the release point. The exact calculation of the cloud dose using volume integral requires large computer time usually exceeding what is available for real time calculations. The volume integral for gamma doses could be approximated by using the semi-infinite cloud model combined with correction factors. This type of calculation procedure is very fast, but usually the accuracy is poor because only a few of the relevant parameters are considered. A multi-parameter method for calculation of gamma doses is described here. This method uses precalculated values of the gamma dose rates as a function of E γ , σ y , the asymmetry factor - σ y /σ z , the height of puff center - H and the distance from puff center R xy . To accelerate the calculations the release energy, for each significant radionuclide in each energy group, has been calculated and tabulated. Based on the precalculated values and suitable interpolation procedure the calculation of gamma doses needs only short computing time and it is almost independent of the number of radionuclides considered. (au) 2 tabs., 15 ills., 12 refs

Precipitates/Salts Model Sensitivity Calculation

Energy Technology Data Exchange (ETDEWEB)

P. Mariner

2001-12-20

The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation on potential seepage waters within a potential repository drift. This work is developed and documented using procedure AP-3.12Q, ''Calculations'', in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The specific objective of this calculation is to examine the sensitivity and uncertainties of the Precipitates/Salts model. The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b). The calculation in the current document examines the effects of starting water composition, mineral suppressions, and the fugacity of carbon dioxide (CO{sub 2}) on the chemical evolution of water in the drift.

Dynamic Model for the Stocks and Release Flows of Engineered Nanomaterials.

Science.gov (United States)

Song, Runsheng; Qin, Yuwei; Suh, Sangwon; Keller, Arturo A

2017-11-07

Most existing life-cycle release models for engineered nanomaterials (ENM) are static, ignoring the dynamics of stock and flows of ENMs. Our model, nanoRelease, estimates the annual releases of ENMs from manufacturing, use, and disposal of a product explicitly taking stock and flow dynamics into account. Given the variabilities in key parameters (e.g., service life of products and annual release rate during use) nanoRelease is designed as a stochastic model. We apply nanoRelease to three ENMs (TiO 2 , SiO 2 and FeO x ) used in paints and coatings through seven product applications, including construction and building, household and furniture, and automotive for the period from 2000 to 2020 using production volume and market projection information. We also consider model uncertainties using Monte Carlo simulation. Compared with 2016, the total annual releases of ENMs in 2020 will increase by 34-40%, and the stock will increase by 28-34%. The fraction of the end-of-life release among total release flows will increase from 11% in 2002 to 43% in 2020. As compared to static models, our dynamic model predicts about an order of magnitude lower values for the amount of ENM released from this sector in the near-term while stock continues to build up in the system.

A new method for calculating energy release rate in tunnel excavation subjected to high in situ stress

Directory of Open Access Journals (Sweden)

Xiao Qinghua

2016-03-01

Full Text Available Based on energy theory, energy release rate (EER and local energy release rate (LEER, a new index called FERR (Fractional Energy Release Rate is proposed, and this method can not only evaluate the risk of rock burst, but also can point out the location of high risk and the scale of rockburst. The FERR index is applied to the TBM assembling tunnel in Jinping Hydro Power Station II to evaluate the scale and intensity of rockburst, as well as the location where rockburst occurs. With FDM method adopted, the energy release rate of 3 excavation plans are calculated and the scale and risk of rockburst is evaluated, and the location of high risk of rockburst is also mapped. With FERR used in the evaluation, the rockburst is nicely controlled which ensured the safety and construction schedule of the project.

REITP3-Hazard evaluation program for heat release based on thermochemical calculation

Energy Technology Data Exchange (ETDEWEB)

Akutsu, Yoshiaki.; Tamura, Masamitsu. [The University of Tokyo, Tokyo (Japan). School of Engineering; Kawakatsu, Yuichi. [Oji Paper Corp., Tokyo (Japan); Wada, Yuji. [National Institute for Resources and Environment, Tsukuba (Japan); Yoshida, Tadao. [Hosei University, Tokyo (Japan). College of Engineering

1999-06-30

REITP3-A hazard evaluation program for heat release besed on thermochemical calculation has been developed by modifying REITP2 (Revised Estimation of Incompatibility from Thermochemical Properties{sup 2)}. The main modifications are as follows. (1) Reactants are retrieved from the database by chemical formula. (2) As products are listed in an external file, the addition of products and change in order of production can be easily conducted. (3) Part of the program has been changed by considering its use on a personal computer or workstation. These modifications will promote the usefulness of the program for energy hazard evaluation. (author)

Modelling isothermal fission gas release

International Nuclear Information System (INIS)

Uffelen, P. van

2002-01-01

The present paper presents a new fission gas release model consisting of two coupled modules. The first module treats the behaviour of the fission gas atoms in spherical grains with a distribution of grain sizes. This module considers single atom diffusion, trapping and fission induced re-solution of gas atoms associated with intragranular bubbles, and re-solution from the grain boundary into a few layers adjacent to the grain face. The second module considers the transport of the fission gas atoms along the grain boundaries. Four mechanisms are incorporated: diffusion controlled precipitation of gas atoms into bubbles, grain boundary bubble sweeping, re-solution of gas atoms into the adjacent grains and gas flow through open porosity when grain boundary bubbles are interconnected. The interconnection of the intergranular bubbles is affected both by the fraction of the grain face occupied by the cavities and by the balance between the bubble internal pressure and the hydrostatic pressure surrounding the bubbles. The model is under validation. In a first step, some numerical routines have been tested by means of analytic solutions. In a second step, the fission gas release model has been coupled with the FTEMP2 code of the Halden Reactor Project for the temperature distribution in the pellets. A parametric study of some steady-state irradiations and one power ramp have been simulated successfully. In particular, the Halden threshold for fission gas release and two simplified FUMEX cases have been computed and are summarised. (author)

Towards elucidation of the drug release mechanism from compressed hydrophilic matrices made of cellulose ethers. III. Critical use of thermodynamic parameters of activation for modeling the water penetration and drug release processes.

Science.gov (United States)

Ferrero, Carmen; Massuelle, Danielle; Jeannerat, Damien; Doelker, Eric

2013-09-10

The two main purposes of this work were: (i) to critically consider the use of thermodynamic parameters of activation for elucidating the drug release mechanism from hydroxypropyl methylcellulose (HPMC) matrices, and (ii) to examine the effect of neutral (pH 6) and acidic (pH 2) media on the release mechanism. For this, caffeine was chosen as model drug and various processes were investigated for the effect of temperature and pH: caffeine diffusion in solution and HPMC gels, and drug release from and water penetration into the HPMC tablets. Generally, the kinetics of the processes was not significantly affected by pH. As for the temperature dependence, the activation energy (E(a)) values calculated from caffeine diffusivities were in the range of Fickian transport (20-40 kJ mol⁻¹). Regarding caffeine release from HPMC matrices, fitting the profiles using the Korsmeyer-Peppas model would indicate anomalous transport. However, the low apparent E(a) values obtained were not compatible with a swelling-controlled mechanism and can be assigned to the dimensional change of the system during drug release. Unexpectedly, negative apparent E(a) values were calculated for the water uptake process, which can be ascribed to the exothermic dissolution of water into the initially dry HPMC, the expansion of the matrix and the polymer dissolution. Taking these contributions into account, the true E(a) would fall into the range valid for Fickian diffusion. Consequently, a relaxation-controlled release mechanism can be dismissed. The apparent anomalous drug release from HPMC matrices results from a coupled Fickian diffusion-erosion mechanism, both at pH 6 and 2. Copyright © 2013 Elsevier B.V. All rights reserved.

Recommendations on dose buildup factors used in models for calculating gamma doses for a plume

International Nuclear Information System (INIS)

Hedemann Jensen, P.; Thykier-Nielsen, S.

1980-09-01

Calculations of external γ-doses from radioactivity released to the atmosphere have been made using different dose buildup factor formulas. Some of the dose buildup factor formulas are used by the Nordic countries in their respective γ-dose models. A comparison of calculated γ-doses using these dose buildup factors shows that the γ-doses can be significantly dependent on the buildup factor formula used in the calculation. Increasing differences occur for increasing plume height, crosswind distance, and atmospheric stability and also for decreasing downwind distance. It is concluded that the most accurate γ-dose can be calculated by use of Capo's polynomial buildup factor formula. Capo-coefficients have been calculated and shown in this report for γ-energies below the original lower limit given by Capo. (author)

Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

Energy Technology Data Exchange (ETDEWEB)

Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

2010-11-15

Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

Applied exposure modeling for residual radioactivity and release criteria

International Nuclear Information System (INIS)

Lee, D.W.

1989-01-01

The protection of public health and the environment from the release of materials with residual radioactivity for recycle or disposal as wastes without radioactive contents of concern presents a formidable challenge. Existing regulatory criteria are based on technical judgment concerning detectability and simple modeling. Recently, exposure modeling methodologies have been developed to provide a more consistent level of health protection. Release criteria derived from the application of exposure modeling methodologies share the same basic elements of analysis but are developed to serve a variety of purposes. Models for the support of regulations for all applications rely on conservative interpretations of generalized conditions while models developed to show compliance incorporate specific conditions not likely to be duplicated at other sites. Research models represent yet another type of modeling which strives to simulate the actual behavior of released material. In spite of these differing purposes, exposure modeling permits the application of sound and reasoned principles of radiation protection to the release of materials with residual levels of radioactivity. Examples of the similarities and differences of these models are presented and an application to the disposal of materials with residual levels of uranium contamination is discussed. 5 refs., 2 tabs

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

International Nuclear Information System (INIS)

Weissfloch, R.

The fuel elements of High-Temperature Reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are present. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can serve as an interim state on the way to a complete theory. (U.S.)

Determination of a geometry-dependent parameter and development of a calculation model for describing the fission products transport from spherical fuel elements of graphite moderated gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Weissfloch, R

1973-07-15

The fuel elements of high-temperature reactors, coated with pyrolitic carbon and covered with graphite, release fission products like all other fuel elements. Because of safety reasons, the rate of this release has to be kept low and has also to be predictable. Measured values from irradiation tests and from post-irradiation tests about the actual release of different fission products are presented. The physical and chemical mechanism, which determines the release, is extraordinarily complex and in particular not clearly defined. Because of the mentioned reasons, a simplified calculation model was developed, which only considers the release-mechanisms phenomenologically. This calculation model coincides very well in its results with values received in experiments until now. It can be held as an interim state on the way to a complete theory.

Development, description and validation of a Tritium Environmental Release Model (TERM).

Science.gov (United States)

Jeffers, Rebecca S; Parker, Geoffrey T

2014-01-01

Tritium is a radioisotope of hydrogen that exists naturally in the environment and may also be released through anthropogenic activities. It bonds readily with hydrogen and oxygen atoms to form tritiated water, which then cycles through the hydrosphere. This paper seeks to model the migration of tritiated species throughout the environment - including atmospheric, river and coastal systems - more comprehensively and more consistently across release scenarios than is currently in the literature. A review of the features and underlying conceptual models of some existing tritium release models was conducted, and an underlying aggregated conceptual process model defined, which is presented. The new model, dubbed 'Tritium Environmental Release Model' (TERM), was then tested against multiple validation sets from literature, including experimental data and reference tests for tritium models. TERM has been shown to be capable of providing reasonable results which are broadly comparable with atmospheric HTO release models from the literature, spanning both continuous and discrete release conditions. TERM also performed well when compared with atmospheric data. TERM is believed to be a useful tool for examining discrete and continuous atmospheric releases or combinations thereof. TERM also includes further capabilities (e.g. river and coastal release scenarios) that may be applicable to certain scenarios that atmospheric models alone may not handle well. Copyright © 2013 Elsevier Ltd. All rights reserved.

Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

International Nuclear Information System (INIS)

Leech, N.A.; Smith, M.R.; Pearce, J.H.; Ellis, W.E.; Beatham, N.

1990-01-01

This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

TRADOS - an air trajectory dose model for long range transport of radioactive release to the atmosphere

International Nuclear Information System (INIS)

Rossi, J.; Valkama, I.

1985-01-01

A model for estimating radiation doses resulting from long range atmospheric transport of released radionuclides in accidents is precented. The model (TRADOS) is able to treat changing diffusion conditions. For example the plume can be exposed to temporary rain, changes in turbulence and mixing depth. This can result in considerable changes in individual doses. The method is applied to an example trajectory and the doses caused by a serious reactor accident are calculated

Double tracer experiments to investigate models for the calculation of gamma doses from a radioactive cloud

International Nuclear Information System (INIS)

Nielsen, S.P.; Gryning, S.E.; Thykier-Nielsen, S.; Karlberg, O.; Lyck, E.

1984-01-01

The paper presents work from a series of atmospheric dispersion experiments in May 1981 at the Ringhals nuclear power plant in Sweden. The aim of the project was to obtain short-term observations of concentrations and gamma-ray exposures from stack effluents and to compare these results with corresponding values calculated from computer models. Two tracers, sulphurhexafluoride (SF 6 ) and radioactive noble gases, were released from a 110-m stack and detected at ground level downwind at distances of 3-4 km. Calculations were made with two Gaussian plume models: PLUCON developed at Riso National Laboratory and UNIDOSE developed at Studsvik Energiteknik AB. (orig.)

Uncertainty estimates for predictions of the impact of breeder-reactor radionuclide releases

International Nuclear Information System (INIS)

Miller, C.W.; Little, C.A.

1982-01-01

This paper summarizes estimates, compiled in a larger report, of the uncertainty associated with models and parameters used to assess the impact on man radionuclide releases to the environment by breeder reactor facilities. These estimates indicate that, for many sites, generic models and representative parameter values may reasonably be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under such circumstances. However, even using site-specific information, inherent natural variability within human receptors, and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose following short-term releases

Study on radioactive release of gaseous and liquid effluents during normal operation of AP1000

International Nuclear Information System (INIS)

Gong Quan; Zhou Jing; Liu Yu

2014-01-01

The gaseous and liquid radioactive releases of pressurized water reactors plant during normal operation are an important content of environmental impact assessment and play a significant role in the design of nuclear power plant. According to the design characters of AP1OOO radioactive waste management system and the study on the calculation method and the release pathways, the calculation model of the gaseous and liquid radioactive releases during normal operation for AP1OOO are established. Base on the established calculation model and the design parameters of AP1000, the expected value of gaseous and liquid radioactive releases of AP1OOO is calculated. The results of calculation are compared with the limits in GB 6249-2011 and explain the adder that is included tu account for anticipated operational occurrences, providing a reference for environmental impact assessment of pressurized water reactor. (authors)

RSAC-6, Gamma doses, inhalation and ingestion doses, fission products inventory after fission products release

International Nuclear Information System (INIS)

Wenzel, Douglas R.; Schrader, Brad J.

2007-01-01

1 - Description of program or function: RSAC-6 is the latest version of the program RSAC (Radiological Safety Analysis Computer Program). It calculates the consequences of a release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory; decay and in-grow the inventory during transport through processes, facilities, and the environment; model the downwind dispersion of the activity; and calculate doses to downwind individuals. Internal dose from the inhalation and ingestion pathways is calculated. External dose from ground surface and plume gamma pathways is calculated. New and exciting updates to the program include the ability to evaluate a release to an enclosed room, resuspension of deposited activity and evaluation of a release up to 1 meter from the release point. Enhanced tools are included for dry deposition, building wake, occupancy factors, respirable fraction, AMAD adjustment, updated and enhanced radionuclide inventory and inclusion of the dose-conversion factors from FOR 11 and 12. 2 - Methods: RSAC6 calculates meteorological dispersion in the atmosphere using Gaussian plume diffusion for Pasquill-Gifford, Hilmeier-Gifford and Markee models. A unique capability is the ability to model Class F fumigation conditions, the meteorological condition that causes the highest ground level concentrations from an elevated release. Doses may be calculated for various pathways including inhalation, ingestion, ground surface, air immersion, water immersion pathways. Dose calculations may be made for either acute or chronic releases. Internal doses (inhalation and ingestion) are calculated using the ICRP-30 model with dose conversion factors from FOR 11. External factors are calculated using FOR 12. 3 - Unusual Features: RSAC6 calculates complete progeny in-growth and decay during all accident phases. The calculation of fission product inventories in particularly useful in the analysis of accidents where the

Calculation models for a nuclear reactor

International Nuclear Information System (INIS)

Tashanii, Ahmed Ali

2010-01-01

Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

Sensitivity analysis on a dose-calculation model for the terrestrial food-chain pathway

International Nuclear Information System (INIS)

Abdel-Aal, M.M.

1994-01-01

Parameter uncertainty and sensitivity were applied to the U.S. Regulatory Commission's (NRC) Regulatory Guide 1.109 (1977) models for calculating the ingestion dose via a terrestrial food-chain pathway in order to assess the transport of chronically released, low-level effluents from light-water reactors. In the analysis, we used the generation of latin hypercube samples (LHS) and employed a constrained sampling scheme. The generation of these samples is based on information supplied to the LHS program for variables or parameters. The actually sampled values are used to form vectors of variables that are commonly used as inputs to computer models for the purpose of sensitivity and uncertainty analysis. Regulatory models consider the concentrations of radionuclides that are deposited on plant tissues or lead to root uptake of nuclides initially deposited on soil. We also consider concentrations in milk and beef as a consequence of grazing on contaminated pasture or ingestion of contaminated feed by dairy and beef cattle. The radionuclides Sr-90 and Cs-137 were selected for evaluation. The most sensitive input parameters for the model were the ground-dispersion parameter, release rates of radionuclides, and soil-to-plant transfer coefficients of radionuclides. (Author)

Modeling of radionuclide transport through rock formations and the resulting radiation exposure of reference persons. Calculations using Asse II parameters

International Nuclear Information System (INIS)

Kueppers, Christian; Ustohalova, Veronika; Steinhoff, Mathias

2012-01-01

The long-term release of radioactivity into the ground water path cannot be excluded for the radioactive waste repository Asse II. The possible radiological consequences were analyzed using a radio-ecological scenario developed by GRS. A second scenario was developed considering the solubility of radionuclides in salt saturated solutions and retarding/retention effects during the radionuclide transport through the cap rock layers. The modeling of possible radiation exposure was based on the lifestyle habits of reference persons. In Germany the calculation procedure for the prediction of radionuclide release from final repositories is not defined by national standards, the used procedures are based on analogue methods from other radiation protection calculations.

Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

Energy Technology Data Exchange (ETDEWEB)

Katata, G.; Chino, M.; Kobayashi, T. [Japan Atomic Energy Agency (JAEA), Ibaraki (Japan); and others

2015-07-01

Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I{sub 2} and CH{sub 3}I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of

Sensitivity study of the Continuous Release Dispersion Model (CRDM) for radioactive pollutants

International Nuclear Information System (INIS)

Camacho, F.

1987-08-01

The Continuous Release Dispersion Model (CRDM) is used to calculate spatial distribution of pollutants and their radiation doses in the event of accidental releases of radioactive material from Nuclear Generation Stations. A sensitivity analysis of the CRDM was carried out to develop a method for quantifying the expected output uncertainty due to inaccuracies and uncertainties in the input values. A simulation approach was used to explore the behaviour of the sensitivity functions. It was found that the most sensitive variable is wind speed, the least sensitive is the ambient temperature, and that largest values of normalized concentrations are likely to occur for small values of wind speed and highly stable atmospheric conditions. It was also shown that an error between 10% and 25% should be expected in the output values for a 1% overall error in the input values, and this factor could be much larger in certain situations

Sensitivity of the engineered barrier system (EBS) release rate to alternative conceptual models of advective release from waste packages under dripping fractures

International Nuclear Information System (INIS)

Lee, J.H.; Atkins, J.E.; McNeish, J.A.; Vallikat, V.

1996-01-01

The first model assumed that dripping water directly contacts the waste form inside the ''failed'' waste package and radionuclides are released from the EBS by advection. The second model assumed that dripping water is diverted around the package (because of corrosion products plugging the perforations), thereby being prevented from directly contacting the waste form. In the second model, radionuclides were assumed to diffuse through the perforations, and, once outside the waste package, to be released from the EBS by advection. For the case with the second EBS release model, most radionuclides had lower peak EBS release rates than with the first model. Impacts of the alternative EBS release models were greater for the radionuclides with low solubility. The analysis indicated that the EBS release model representing advection through a ''failed'' waste package (the first model) may be too conservative; thus a ''failed'' waste package container with multiple perforations may still be an important barrier to radionuclide release

Calculations of Fission Gas Release During Ramp Tests Using Copernic Code

Energy Technology Data Exchange (ETDEWEB)

Tong, Liu [Nuclear Fuel R and D Center, China Nuclear Power Technology Research Institute (CNPRI) (China)

2013-03-15

The report performed under IAEA research contract No.15951 describes the results of fuel performance evaluation of LWR fuel rods operated at ramp conditions using the COPERNIC code developed by AREVA. The experimental data from the Third Riso Fission Gas Project and the Studsvik SUPER-RAMP Project presented in the IFPE database of the OECD/NEA has been utilized for assessing the code itself during simulation of fission gas release (FGR). Standard code models for LWR fuel were used in simulations with parameters set properly in accordance with relevant test reports. With the help of data adjustment, the input power histories are restructured to fit the real ones, so as to ensure the validity of FGR prediction. The results obtained by COPERNIC show that different models lead to diverse predictions and discrepancies. By comparison, the COPERNIC V2.2 model (95% Upper bound) is selected as the standard FGR model in this report and the FGR phenomenon is properly simulated by the code. To interpret the large discrepancies of some certain PK rods, the burst effect of FGR which is taken into consideration in COPERNIC is described and the influence of the input power histories is extrapolated. In addition, the real-time tracking capability of COPERNIC is tested against experimental data. In the process of investigation, two main dominant factors influencing the measured gas release rate are described and different mechanisms are analyzed. With the limited predicting capacity, accurate predictions cannot be carried out on abrupt changes of FGR during ramp tests by COPERNIC and improvements may be necessary to some relevant models. (author)

Use of realistic anthropomorphic models for calculation of radiation dose in nuclear medicine

International Nuclear Information System (INIS)

Stabin, Michael G.; Emmons, Mary A.; Fernald, Michael J.; Brill, A.B.; Segars, W.Paul

2008-01-01

Anthropomorphic phantoms based on simple geometric structures have been used in radiation dose calculations for many years. We have now developed a series of anatomically realistic phantoms representing adults and children using body models based on non-uniform rational B-spline (NURBS), with organ and body masses based on the reference values given in ICRP Publication 89. Age-dependent models were scaled and shaped to represent the reference individuals described in ICRP 89 (male and female adults, newborns, 1-, 5-, 10- and 15-year-olds), using a software tool developed in Visual C++. Voxel-based versions of these models were used with GEANT4 radiation transport codes for calculation of specific absorbed fractions (SAFs) for internal sources of photons and electrons, using standard starting energy values. Organ masses in the models were within a few % of ICRP reference masses, and physicians reviewed the models for anatomical realism. Development of individual phantoms was much faster than manual segmentation of medical images, and resulted in a very uniform standardized phantom series. SAFs were calculated on the Vanderbilt multi node computing network (ACCRE). Photon and electron SAFs were calculated for all organs in all models, and were compared to values from similar phantoms developed by others. Agreement was very good in most cases; some differences were seen, due to differences in organ mass and geometry. This realistic phantom series represents a possible replacement for the Cristy/Eckerman series of the 1980's. Both phantom sets will be included in the next release of the OLINDA/EXM personal computer code, and the new phantoms will be made generally available to the research community for other uses. Calculated radiation doses for diagnostic and therapeutic radiopharmaceuticals will be compared with previous values. (author)

A thermodynamic/mass-transport model for the release of ruthenium from irradiated fuel

International Nuclear Information System (INIS)

Garisto, F.; Iglesias, F.C.; Hunt, C.E.L.

1990-01-01

Some postulated nuclear reactor accidents lead to fuel failures and hence release of fission products into the primary heat transport system (PHTS). To determine the consequences of such accidents, it is important to understand the behavior of fission products both in the PHTS and in the reactor containment building. Ruthenium metal has a high boiling point and is nonvolatile under reducing conditions. However, under oxidizing conditions ruthenium can form volatile oxides at relatively low temperatures and, hence, could escape from failed fuel and enter the containment building. The ruthenium radioisotope Ru-106 presents a potentially significant health risk if it is released outside the reactor containment building. Consequently, it is important to understand the behavior of ruthenium during a nuclear reactor accident. The authors review the thermodynamic behavior of ruthenium at high temperatures. The qualitative behavior of ruthenium, predicted using thermodynamic calculations, is then compared with experimental results from the Chalk River Nuclear Laboratories (CRNL). Finally, a simple thermodynamic/mass-transport model is proposed to explain the release behavior of ruthenium in a steam atmosphere

Models for the evaluation of ingestion doses from the consumption of terrestrial foods following an atmospheric radioactive release

International Nuclear Information System (INIS)

Nair, S.

1984-04-01

Various methods are described which have been incorporated in the FOODWEB module of the CEGB's NECTAR environmental code and are currently being used within CEGB to assess ingestion doses from consumption of terrestrial foods following an atmospheric radioactive release. Four foodchain models which have been developed within CEGB are fully described and results of typical calculations presented. Also given are the results of a validation of the dynamic model against measured 90 Sr and 137 Cs levels in milk in the U.K. resulting from weapons fallout. Methods are also described for calculating individual and population doses from ingestion using the results of the model calculations. The population dose calculations utilise a data base describing the spatial distribution of production of a wide range of agricultural products. The development of such a data base for Great Britain is described, based on the 1972 land use and livestock census, and maps are presented for each agricultural product. (U.K.)

A fission gas release model

Energy Technology Data Exchange (ETDEWEB)

Denis, A; Piotrkowski, R [Argentine Atomic Energy Commission, Buenos Aires (Argentina)

1997-08-01

The hypothesis contained in the model developed in this work are as follows. The UO{sub 2} is considered as a collection of spherical grains. Nuclear reactions produce fission gases, mainly Xe and Kr, within the grains. Due to the very low solubility of these gases in UO{sub 2}, intragranular bubbles are formed, of a few nanometers is size. The bubbles are assumed to be immobile and to act as traps which capture gas atoms. Free atoms diffuse towards the grain boundaries, where they give origin to intergranular, lenticular bubbles, of the order of microns. The gas atoms in bubbles, either inter or intragranular, can re-enter the matrix through the mechanism of resolution induced by fission fragment impact. The amount of gas stored in intergranular bubbles grows up to a saturation value. Once saturation is reached, intergranular bubbles inter-connect and the gas in excess is released through different channels to the external surface of the fuel. The resolution of intergranular bubbles particularly affects the region of the grain adjacent to the grain boundary. During grain growth, the grain boundary traps the gas atoms, either free or in intragranular bubbles, contained in the swept volume. The grain boundary is considered as a perfect sink, i.e. the gas concentration is zero at that surface of the grain. Due to the spherical symmetry of the problem, the concentration gradient is null at the centre of the grain. The diffusion equation was solved using the implicit finite difference method. The initial solution was analytically obtained by the Laplace transform. The calculations were performed at different constant temperatures and were compared with experimental results. They show the asymptotic growth of the grain radius as a function of burnup, the gas distribution within the grain at every instant, the growth of the gas content at the grain boundary up to the saturation value and the fraction of gas released by the fuel element referred to the total gas generated

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)

Calculation of the inventory and near-field release rates of radioactivity from neutron-activated metal parts discharged from the high flux isotope reactor and emplaced in solid waste storage area 6 at Oak Ridge National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Kelmers, A.D.; Hightower, J.R.

1987-05-01

Emplacement of contaminated reactor components involves disposal in lined and unlined auger holes in soil above the water table. The radionuclide inventory of disposed components was calculated. Information on the composition and weight of the components, as well as reasonable assumptions for the neutron flux fueling use, the time of neutron exposure, and radioactive decay after discharge, were employed in the inventory calculation. Near-field release rates of /sup 152/Eu, /sup 154/Eu, and /sup 155/Eu from control plates and cylinders were calculated for 50 years after emplacement. Release rates of the europium isotopes were uncertain. Two release-rate-limiting models were considered and a range of reasonable values were assumed for the time-to-failure of the auger-hole linear and aluminum cladding and europium solubility in SWSA-6 groundwater. The bounding europium radionuclide near-field release rates peaked at about 1.3 Ci/year total for /sup 152,154,155/Eu in 1987 for the lower bound, and at about 420 Ci/year in 1992 for the upper bound. The near-field release rates of /sup 55/Fe, /sup 59/Ni, /sup 60/Co, and /sup 63/Ni from stainless steel and cobalt alloy components, as well as of /sup 10/Be, /sup 41/Ca, and /sup 55/Fe from beryllium reflectors, were calculated for the next 100 years, assuming bulk waste corrosion was the release-rate-limiting step. Under the most conservative assumptions for the reflectors, the current (1986) total radionuclide release rate was calculated to be about 1.2 x 10/sup -4/ Ci/year, decreasing by 1992 to a steady release of about 1.5 x 10/sup -5/ Ci/year due primarily to /sup 41/Ca. 50 refs., 13 figs., 8 tabs.

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1990-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous radionuclides. A modeling approach is outlined and sample calculations are provided that show the effects on release due to a limited perforation area, to decreasing temperature, and to the partial occlusion of the perforated area by corrosion products. 8 refs., 2 tabs

Dispersion modeling of accidental releases of toxic gases - Comparison of the models and their utility for the fire brigades.

Science.gov (United States)

Stenzel, S.; Baumann-Stanzer, K.

2009-04-01

Dispersion modeling of accidental releases of toxic gases - Comparison of the models and their utility for the fire brigades. Sirma Stenzel, Kathrin Baumann-Stanzer In the case of accidental release of hazardous gases in the atmosphere, the emergency responders need a reliable and fast tool to assess the possible consequences and apply the optimal countermeasures. For hazard prediction and simulation of the hazard zones a number of air dispersion models are available. The most model packages (commercial or free of charge) include a chemical database, an intuitive graphical user interface (GUI) and automated graphical output for display the results, they are easy to use and can operate fast and effective during stress situations. The models are designed especially for analyzing different accidental toxic release scenarios ("worst-case scenarios"), preparing emergency response plans and optimal countermeasures as well as for real-time risk assessment and management. There are also possibilities for model direct coupling to automatic meteorological stations, in order to avoid uncertainties in the model output due to insufficient or incorrect meteorological data. Another key problem in coping with accidental toxic release is the relative width spectrum of regulations and values, like IDLH, ERPG, AEGL, MAK etc. and the different criteria for their application. Since the particulate emergency responders and organizations require for their purposes unequal regulations and values, it is quite difficult to predict the individual hazard areas. There are a quite number of research studies and investigations coping with the problem, anyway the end decision is up to the authorities. The research project RETOMOD (reference scenarios calculations for toxic gas releases - model systems and their utility for the fire brigade) was conducted by the Central Institute for Meteorology and Geodynamics (ZAMG) in cooperation with the Vienna fire brigade, OMV Refining & Marketing GmbH and

MUDMAP: Simulation model for releases from offshore platforms

International Nuclear Information System (INIS)

Anon.

1994-01-01

The present article deals with a Norwegian developed simulation model dubbed MUDMAP. MUDMAP is a numerical model that simulates releases of drill muds and cuttings, produced water and other substances from offshore platforms. The model is envisioned as an advanced tool to assist in the rapid design and placement of intakes and release pipes on platforms, as well as in evaluating potential long-term impacts in the water and on the sea floor. MUDMAP allows rapid visual/graphical analysis of potential alternative solutions under various realistic environmental conditions, and for planning and executing platform monitoring projects. 4 figs

Modelling the release behaviour of cesium during severe fuel degradation

International Nuclear Information System (INIS)

Lewis, B.J.; Andre, B.; Morel, B.

1995-01-01

An analytical model has been applied to describe the diffusional release of fission product cesium from Zircaloy-clad fuel under high-temperature reactor accident conditions. The present treatment accounts for the influence of the atmosphere (i.e., changing oxygen potential) on the state of fuel oxidation and the release kinetics. The effects of fuel dissolution on the volatile release behaviour (under reducing conditions) is considered in terms of earlier crucible experiments and a simple model based on bubble coalescence and transport in metal pools. The model has been used to interpret the cesium release kinetics observed in steam and hydrogen experiments at the Vertical Irradiation (VI) Facility in the Oak Ridge National Laboratory and at the HEVA/VERCORS Facility in the Commissariat a l'Energie Atomique. (author)

Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

International Nuclear Information System (INIS)

1991-04-01

This standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The radionuclides considered are those associated with substances having the greatest potential for becoming airborne in reactor accidents: tritium (HTO), noble gases and their daughters, radioiodines, and certain radioactive particulates (Cs, Ru, Sr, Te). The standard focuses on the calculation of radiation doses for external exposures from radioactive material in the cloud; internal exposures for inhalation of radioactive material in the cloud and skin penetration of tritium; and external exposures from radionuclides deposited on the ground. It uses as modified Gaussian plume model to evaluate the time-integrated concentration downwind. (52 refs., 12 tabs., 21 figs.)

14C release from failed spent fuel containers

International Nuclear Information System (INIS)

Pescatore, C.

1989-01-01

Partially failed containers may provide a meaningful barrier to the release of gaseous 14 CO 2 . A modeling approach is outlined and sample calculations are provided that show the effect on release due to a limited perforation area, to decreasing temperature, and to partial occlusion of the perforated area by corrosion products. 5 refs., 4 tabs

Development of three-dimensional trajectory model for detecting source region of the radioactive materials released into the atmosphere

Energy Technology Data Exchange (ETDEWEB)

Suh, Kyung Suk; Park, Ki Hyun; Min, Byung Il; Kim, Sora; Yang, Byung Mo [Nuclear Environmental Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-05-15

It is necessary to consider the overall countermeasure for analysis of nuclear activities according to the increase of the nuclear facilities like nuclear power and reprocessing plants in the neighboring countries including China, Taiwan, North Korea, Japan and South Korea. South Korea and comprehensive nuclear-test-ban treaty organization (CTBTO) are now operating the monitoring instruments to detect radionuclides released into the air. It is important to estimate the origin of radionuclides measured using the detection technology as well as the monitoring analysis in aspects of investigation and security of the nuclear activities in neighboring countries. A three-dimensional forward/backward trajectory model has been developed to estimate the origin of radionuclides for a covert nuclear activity. The developed trajectory model was composed of forward and backward modules to track the particle positions using finite difference method. A three-dimensional trajectory model was validated using the measured data at Chernobyl accident. The calculated results showed a good agreement by using the high concentration measurements and the locations where was near a release point. The three-dimensional trajectory model had some uncertainty according to the release time, release height and time interval of the trajectory at each release points. An atmospheric dispersion model called long-range accident dose assessment system (LADAS), based on the fields of regards (FOR) technique, was applied to reduce the uncertainties of the trajectory model and to improve the detective technology for estimating the radioisotopes emission area. The detective technology developed in this study can evaluate in release area and origin for covert nuclear activities based on measured radioisotopes at monitoring stations, and it might play critical tool to improve the ability of the nuclear safety field.

Reaction dynamics of the four-centered elimination CH2OH + --> CHO + +H2: Measurement of kinetic energy release distribution and classical trajectory calculation

Science.gov (United States)

Lee, Tae Geol; Park, Seung C.; Kim, Myung Soo

1996-03-01

Mass-analyzed ion kinetic energy (MIKE) spectrum of CHO+ generated in the unimolecular dissociation of CH2OH+ was measured. Kinetic energy release distribution (KERD) was evaluated by analyzing the spectrum according to the algorithm developed previously. The average kinetic energy release evaluated from the distribution was extraordinarily large, 1.63 eV, corresponding to 75% of the reverse barrier of the reaction. A global analytical potential energy surface was constructed such that the experimental energetics was represented and that various features in the ab initio potential energy surface were closely reproduced. Classical trajectory calculation was carried out with the global analytical potential energy surface to investigate the causes for the extraordinarily large kinetic energy release. Based on the detailed dynamical calculations, it was found that the strained bending forces at the transition state and strengthening of the CO bond from double to triple bond character were mainly responsible for such a significant kinetic energy release. In addition, the dissociation products H2 and CHO+ ion were found to be rotationally excited in the trajectory calculations. This was attributed to the asymmetry of the transition state and the release of asymmetric bending forces. Also, the bending vibrational modes of CHO+ and the H2 stretching mode, which are coupled with the bending coordinates, were found to be moderately excited.

Thermophysical modeling of volatile fission product release from a debris pool

International Nuclear Information System (INIS)

Yun, J. I.; Suh, K. Y.; Kang, C. S.

1999-01-01

A model is described for fission product release from the debris pool in the lower plenum of the reactor pressure vessel. In the pool, turbulent natural convection flow is formed due to homogeneous internal heat generation. Using the best-known correlations, heat transfer at the curved bottom and the top of the pool may be calculated. Volatile fission product gases in the pool nucleate and diffuse to bubbles. Both the homogeneous nucleation and heterogeneous nucleation are considered. The bubble nucleation, growth, coalescence and loss due to rise is modeled pursuant to bubble dynamics. If the pressure and temperature of the pool are very high, homogeneous nucleation that accounts for effect of decrease in the pool pressure can occur. The effect of the bubble-to-pool interfacial tension and the pool pressure on the nucleation rate is investigated in this work

REDUCING AMMONIA CONCENTRATIONS IN ATMOSPHERE AFTER ITS UNPLANNED RELEASE

Directory of Open Access Journals (Sweden)

L. V. Amelina

2017-08-01

Full Text Available Purpose. The aim of this work is development of numerical model, which allows to calculate the efficiency of neutralizer supply for reduction of air pollution in case of unplanned ammonia emission at the territory of ammonia pump station. The numerical model should allow fast calculating, taking into account the meteorological parameters and buildings situated near the source of toxic chemical emission and equipment for neutralizer supply. Methodology. The developed model is based on the equation for potential flow and equation of pollutant dispersion. To simulate the chemical interaction between ammonia and neutralizer the stoichiometry equation is used. Equation of potential flow is used to compute flow pattern among buildings. To solve the equation for potential flow the Samarskii implicit difference scheme is used. The implicit change-triangle difference scheme is used to solve equation of mass transfer. While for the numerical integration the authors use the rectangular difference grid. Method of porosity technique («markers method» is applied to create the form of comprehensive computational region. Emission of ammonia is modeled using Delta function for point source. Findings. Developed numerical model belongs to the class of «diagnostic models». This model takes into account the main physical factors affecting the process of dispersion of ammonia and neutralizer in the atmosphere, as well as the influence of buildings on admixture dispersion. On the basis of the developed numerical models the authors carried out a computational experiment to estimate the efficiency of neutralizer supply for reduction of air pollution in case of unplanned ammonia release at ammonia pump station. Originality. Developed numerical model allows calculating the flow pattern among buildings and estimating the efficiency of neutralizer supply for reduction of air pollution in the case unplanned ammonia release. Practical value. Model allows performing fast

Study on the calculation method of source term from fission products

International Nuclear Information System (INIS)

Zhou Jing; Gong Quan; Qiu Haifeng

2014-01-01

As a major part of radioactive nuclides, fission products play an important role in nuclear power plant design. The paper analyzes the calculation model of core activity inventory, the model of fission products releasing from the pellets to RCS, the balance model of fission products in RCS, and then proves them by calculation of the typical pressurized water reactor. The model is proved applicable for calculating fission products of pressurized water reactors. (authors)

Relative Release-to-Birth Indicators for Investigating TRISO Fuel Fission Gas Release Models

International Nuclear Information System (INIS)

Harp, Jason M.; Hawari, Ayman I.

2008-01-01

TRISO microsphere fuel is the fundamental fuel unit for Very High Temperature Reactors (VHTR). A single TRISO particle consists of an inner kernel of uranium dioxide or uranium oxycarbide surrounded by layers of pyrolytic carbon and silicon carbide. If the silicon carbide layer fails, fission products, especially the noble fission gases Kr and Xe, will begin to escape the failed particle. The release of fission gas is usually quantified by measuring the ratio of the released activity (R) to the original birth activity (B), which is designated as the R/B ratio. In this work, relative Release-to-Birth indicators (I) are proposed as a technique for interpreting the results of TRISO irradiation experiments. By implementing a relative metric, it is possible to reduce the sensitivity of the indicators to instrumental uncertainties and variations in experimental conditions. As an example, relative R/B indicators are applied to the interpretation of representative data from the Advanced Gas Reactor-1 TRISO fuel experiment that is currently taking place at the Advanced Test Reactor of Idaho National Laboratory. It is shown that the comparison of measured to predicted relative R/B indicators (I) gives insight into the physics of release and helps validate release models. Different trends displayed by the indicators are related to the mechanisms of fission gas release such as diffusion and recoil. The current analysis shows evidence for separate diffusion coefficients for Kr and Xe and supports the need to account for recoil release. (authors)

COCO-1: model for assessing the cost of offsite consequences of accidental releases of radioactivity

International Nuclear Information System (INIS)

Haywood, S.M.; Robinson, C.A.; Heady, C.

1991-09-01

This report describes a new model, called COCO-1 (Cost Of Consequences Offsite), for assessing the offsite economic consequences of an accident involving the release of radioactive material. The costs calculated are a measure of the benefit foregone as a result of the accident, and in addition to tangible monetary costs the model attempts to include costs arising from the effect of the accident on individuals, for instance the disruption caused by the loss of homes. The approach has limitations, which are discussed, but offers a broadly applicable and robust technique for estimating the economic impact of most accidents. (author)

Novel Caffeic Acid Nanocarrier: Production, Characterization, and Release Modeling

Directory of Open Access Journals (Sweden)

Milad Fathi

2013-01-01

Full Text Available This paper deals with the development of novel nanocarriers using layer by layer carbohydrate coating of caffeic acid loaded solid lipid nanoparticles (SLNs to improve stability and colon delivery of the poorly water-soluble caffeic acid. Three biopolymers (chitosan, alginate, and pectin in different concentrations (0.1, 0.25, and 0.5% were electrostatically coated over the SLN surface. The size and zeta potential of produced nanocarriers were measured using photon correlation spectroscopy. Mathematical models (i.e., zero-order, first-order, Higuchi, Ritger-Peppas, reciprocal powered time, Weibull, and quadratic models were used to describe the release and kinetic modeling in gastrointestinal solution (GIS. Also, antioxidant activity of caffeic acid during the release in GIS was investigated using DPPH and reducing activity methods. The prepared treatments coated by alginate-chitosan as well as pectin-chitosan coated SLN at the concentration of 0.1% showed nanosized bead; the latter efficiently retarded the release of caffeic acid in gastric media up to 2.5 times higher than that of SLN. Zeta potential values of coated samples were found to significantly increase in comparison to SLN indicating the higher stability of produced nanocarriers. Antioxidant activity of caffeic acid after gastric release did not result in the same trend as observed for caffeic acid release from different treatments; however, in line with less caffeic acid release in the intestine solution by the effect of coating, lower antioxidant activity was determined at the end stage of the experiment.

Modelling and assessment of accidental oil release from damaged subsea pipelines.

Science.gov (United States)

Li, Xinhong; Chen, Guoming; Zhu, Hongwei

2017-10-15

This paper develops a 3D, transient, mathematical model to estimate the oil release rate and simulate the oil dispersion behavior. The Euler-Euler method is used to estimate the subsea oil release rate, while the Eulerian-Lagrangian method is employed to track the migration trajectory of oil droplets. This model accounts for the quantitative effect of backpressure and hole size on oil release rate, and the influence of oil release rate, oil density, current speed, water depth and leakage position on oil migration is also investigated in this paper. Eventually, the results, e.g. transient release rate of oil, the rise time of oil and dispersion distance are determined by above-mentioned model, and the oil release and dispersion behavior under different scenarios is revealed. Essentially, the assessment results could provide a useful guidance for detection of leakage positon and placement of oil containment boom. Copyright © 2017 Elsevier Ltd. All rights reserved.

IN-DRIFT MICROBIAL COMMUNITIES MODEL VALIDATION CALCULATIONS

Energy Technology Data Exchange (ETDEWEB)

D.M. Jolley

2001-12-18

The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN M09909SPAMINGl.003 using its replacement DTN M00106SPAIDMO 1.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 200 1) which includes controls for the management of electronic data.

In-Drift Microbial Communities Model Validation Calculations

Energy Technology Data Exchange (ETDEWEB)

D. M. Jolley

2001-09-24

The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data.

In-Drift Microbial Communities Model Validation Calculation

Energy Technology Data Exchange (ETDEWEB)

D. M. Jolley

2001-10-31

The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M&O 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M&O 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M&O 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M&O (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data.

In-Drift Microbial Communities Model Validation Calculations

International Nuclear Information System (INIS)

Jolley, D.M.

2001-01-01

The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS MandO 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS MandO 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN MO9909SPAMING1.003 using its replacement DTN MO0106SPAIDM01.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS MandO 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS MandO (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 2001) which includes controls for the management of electronic data

IN-DRIFT MICROBIAL COMMUNITIES MODEL VALIDATION CALCULATIONS

International Nuclear Information System (INIS)

D.M. Jolley

2001-01-01

The objective and scope of this calculation is to create the appropriate parameter input for MING 1.0 (CSCI 30018 V1.0, CRWMS M andO 1998b) that will allow the testing of the results from the MING software code with both scientific measurements of microbial populations at the site and laboratory and with natural analogs to the site. This set of calculations provides results that will be used in model validation for the ''In-Drift Microbial Communities'' model (CRWMS M andO 2000) which is part of the Engineered Barrier System Department (EBS) process modeling effort that eventually will feed future Total System Performance Assessment (TSPA) models. This calculation is being produced to replace MING model validation output that is effected by the supersession of DTN M09909SPAMINGl.003 using its replacement DTN M00106SPAIDMO 1.034 so that the calculations currently found in the ''In-Drift Microbial Communities'' AMR (CRWMS M andO 2000) will be brought up to date. This set of calculations replaces the calculations contained in sections 6.7.2, 6.7.3 and Attachment I of CRWMS M andO (2000) As all of these calculations are created explicitly for model validation, the data qualification status of all inputs can be considered corroborative in accordance with AP-3.15Q. This work activity has been evaluated in accordance with the AP-2.21 procedure, ''Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities'', and is subject to QA controls (BSC 2001). The calculation is developed in accordance with the AP-3.12 procedure, Calculations, and prepared in accordance with the ''Technical Work Plan For EBS Department Modeling FY 01 Work Activities'' (BSC 200 1) which includes controls for the management of electronic data

Incorporating water-release and lateral protein interactions in modeling equilibrium adsorption for ion-exchange chromatography.

Science.gov (United States)

Thrash, Marvin E; Pinto, Neville G

2006-09-08

The equilibrium adsorption of two albumin proteins on a commercial ion exchanger has been studied using a colloidal model. The model accounts for electrostatic and van der Waals forces between proteins and the ion exchanger surface, the energy of interaction between adsorbed proteins, and the contribution of entropy from water-release accompanying protein adsorption. Protein-surface interactions were calculated using methods previously reported in the literature. Lateral interactions between adsorbed proteins were experimentally measured with microcalorimetry. Water-release was estimated by applying the preferential interaction approach to chromatographic retention data. The adsorption of ovalbumin and bovine serum albumin on an anion exchanger at solution pH>pI of protein was measured. The experimental isotherms have been modeled from the linear region to saturation, and the influence of three modulating alkali chlorides on capacity has been evaluated. The heat of adsorption is endothermic for all cases studied, despite the fact that the net charge on the protein is opposite that of the adsorbing surface. Strong repulsive forces between adsorbed proteins underlie the endothermic heat of adsorption, and these forces intensify with protein loading. It was found that the driving force for adsorption is the entropy increase due to the release of water from the protein and adsorbent surfaces. It is shown that the colloidal model predicts protein adsorption capacity in both the linear and non-linear isotherm regions, and can account for the effects of modulating salt.

Methodology for calculating radiation doses from radioactivity released to the environment

International Nuclear Information System (INIS)

Killough, G.G.; McKay, L.R.

1976-03-01

This document represents a compilation of the principal environmental transport and dosimetry models developed, adapted, and implemented by the Radiological Analyses and Applications Group of the Environmental Sciences Division of the Oak Ridge National Laboratory. The transport of released radioactivity through the natural environment is discussed in four sections: atmospheric dispersion, resuspension of material by wind action, terrestrial transport, and movement of material in underground water seepage. The discussion of dose to man and biota is divided into internal and external exposure sections. And finally, a developmental model (CONDOS) which estimates the dose to a population resulting from the manufacture, storage, distribution, use, and disposal of consumer products which contain radioactivity is described. Numerous tables are included

Health effects associtated with unit radonuclide releases to the environment

International Nuclear Information System (INIS)

Helton, J.C.

1983-01-01

Simple models are presented for the estimation of individual and population health effects (i.e., latent cancer fatalities) for long-term radionuclide releases to the surface environment. These models were suggested by techniques employed by the Environmental Protection Agency in the development of a proposed standard for the disposal of high-level radioactive waste. The modeling approach is based on the use of asymptotic solutions to mixed-cell models in conjunction with appropriate usage rates, dose factors, risk factors, and population estimates. Although the models are simple, it is felt that they can be used in preliminary investigations of topics in high-level waste disposal such as potential importance of individual radionuclides, relative importance of different release patterns or exposure pathways, and relationships between individual and population exposures. The use of the models is illustrated by calculating the population health effects along various exposure pathways for the radionuclides considered in the proposed Environmental Protection Agency Standard. The results of these calculations are compared with the calculated population exposures on which the proposed Environmental Protection Agency Standard is based

Integrated environmental modeling system for noble gas releases at the Savannah River Plant

International Nuclear Information System (INIS)

Cooper, R.E.

1973-01-01

The Savannah River Plant (SRP) is a large nuclear complex engaged in varied activities and is the AEC's major site for the production of weapons material. As a result of these activities, there are continuous and intermittent releases of radioactive gases to the atmosphere. Of these releases, the noble gases constitute about 11 percent of the total man-rem exposure to the population out to a distance of 100 km. Although SRP has an extensive radiological monitoring program, an environmental modeling system is necessary for adequately estimating effects on the environment. The integrated environmental modeling system in use at SRP consists of a series of computer programs that generate and use a library of environmental effects data as a function of azimuth and distance. Annual average atmospheric dispersion and azimuthal distribution of material assumed to be released as unit sources is estimated from a 2-year meteorological data base--assuming an arbitrary point of origin. The basic library of data consists of: ground-level concentrations according to isotope, and whole body gamma dose calculations that account for the total spatial distribution at discrete energy levels. These data are normalized to tritium measurements, and are subsequently used to generate similar library data that pertain to specific source locations, but always with respect to the same population grid. Thus, the total additive effects from all source points, both on- and off-site, can be estimated. The final program uses the library data to estimate population exposures for specified releases and source points for the nuclides of interest (including noble gases). Multiple source points are considered within a single pass to obtain the integrated effects from all sources

Fission product release from fuel of water-cooled reactors

International Nuclear Information System (INIS)

Strupczewski, A.; Marks, P.; Klisinska, M.

1997-01-01

The report contains a review of theoretical models and experimental works of gaseous and volatile fission products from uranium dioxide fuel. The experimental results of activity release at low burnup and the model of fission gas behaviour at initial stage of fuel operational cycle are presented. Empirical models as well as measured results of transient fission products release rate in the temperature up to UO 2 melting point, with consideration of their chemical reactions with fuel and cladding, are collected. The theoretical and experimental data were used for calculations of gaseous and volatile fission products release, especially iodine and caesium, to the gas volume of WWER-1000 and WWER-440 type fuel rods at low and high burnup and their further release from defected rods at the assumed loss-of-coolant accident. (author)

A Novel Hybrid Similarity Calculation Model

Directory of Open Access Journals (Sweden)

Xiaoping Fan

2017-01-01

Full Text Available This paper addresses the problems of similarity calculation in the traditional recommendation algorithms of nearest neighbor collaborative filtering, especially the failure in describing dynamic user preference. Proceeding from the perspective of solving the problem of user interest drift, a new hybrid similarity calculation model is proposed in this paper. This model consists of two parts, on the one hand the model uses the function fitting to describe users’ rating behaviors and their rating preferences, and on the other hand it employs the Random Forest algorithm to take user attribute features into account. Furthermore, the paper combines the two parts to build a new hybrid similarity calculation model for user recommendation. Experimental results show that, for data sets of different size, the model’s prediction precision is higher than the traditional recommendation algorithms.

A review of mathematical modeling and simulation of controlled-release fertilizers.

Science.gov (United States)

Irfan, Sayed Ameenuddin; Razali, Radzuan; KuShaari, KuZilati; Mansor, Nurlidia; Azeem, Babar; Ford Versypt, Ashlee N

2018-02-10

Nutrients released into soils from uncoated fertilizer granules are lost continuously due to volatilization, leaching, denitrification, and surface run-off. These issues have caused economic loss due to low nutrient absorption efficiency and environmental pollution due to hazardous emissions and water eutrophication. Controlled-release fertilizers (CRFs) can change the release kinetics of the fertilizer nutrients through an abatement strategy to offset these issues by providing the fertilizer content in synchrony with the metabolic needs of the plants. Parametric analysis of release characteristics of CRFs is of paramount importance for the design and development of new CRFs. However, the experimental approaches are not only time consuming, but they are also cumbersome and expensive. Scientists have introduced mathematical modeling techniques to predict the release of nutrients from the CRFs to elucidate fundamental understanding of the dynamics of the release processes and to design new CRFs in a shorter time and with relatively lower cost. This paper reviews and critically analyzes the latest developments in the mathematical modeling and simulation techniques that have been reported for the characteristics and mechanisms of nutrient release from CRFs. The scope of this review includes the modeling and simulations techniques used for coated, controlled-release fertilizers. Copyright © 2017 Elsevier B.V. All rights reserved.

Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities

International Nuclear Information System (INIS)

Washington, K.E.; Carroll, D.E.

1988-01-01

Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab

Predicting red wolf release success in the southeastern United States

Science.gov (United States)

van Manen, Frank T.; Crawford, Barron A.; Clark, Joseph D.

2000-01-01

Although the red wolf (Canis rufus) was once found throughout the southeastern United States, indiscriminate killing and habitat destruction reduced its range to a small section of coastal Texas and Louisiana. Wolves trapped from 1973 to 1980 were taken to establish a captive breeding program that was used to repatriate 2 mainland and 3 island red wolf populations. We collected data from 320 red wolf releases in these areas and classified each as a success or failure based on survival and reproductive criteria, and whether recaptures were necessary to resolve conflicts with humans. We evaluated the relations between release success and conditions at the release sites, characteristics of released wolves, and release procedures. Although <44% of the variation in release success was explained, model performance based on jackknife tests indicated a 72-80% correct prediction rate for the 4 operational models we developed. The models indicated that success was associated with human influences on the landscape and the level of wolf habituation to humans prior to release. We applied the models to 31 prospective areas for wolf repatriation and calculated an index of release success for each area. Decision-makers can use these models to objectively rank prospective release areas and compare strengths and weaknesses of each.

Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

International Nuclear Information System (INIS)

Loesoenen, Pekka

2002-01-01

A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

IODES, Calculating the Estimation of Dose to the World Population from Releases of Iodine-129 to the Environment

International Nuclear Information System (INIS)

2002-01-01

1 - Description of program or function: IODES is a dynamic linear compartment model of the global iodine cycle which estimates long-term doses and dose commitments to the world population from releases of 129 I to the environment. The global environment is divided into different compartments comprising the atmosphere, hydrosphere, lithosphere, and terrestrial biosphere. The global transport of iodine is described by means of time-invariant fractional transfer rates between the environmental compartments. The fractional transfer rates for 129 I are determined primarily from available data on compartment inventories and fluxes for naturally occurring stable iodine and from data on the global hydrologic cycle. The dose to the world population is estimated from the calculated compartment inventories of 129 I, the known compartment inventories of stable iodine, a pathway analysis of the intake of iodine by a reference individual, dose conversion factors for inhalation and ingestion, and an estimate of the world population. For an assumed constant population of 12.21 billion beyond the year 2075, the estimated population dose commitment is 2 x 105 man-rem/Ci. 2 - Methods: IODES calculates 129 I inventories in the different environmental compartments and individual and population doses as a function of time after a release to the environment by solving a set of simultaneous first-order linear differential equations using numerical methods. 3 - Restrictions on the complexity of the problem: - The subroutine LSODE for solving the differential equations is provided online at the ORNL computer center and, thus, is not included in the IODES code package. If the LSODE routine is not available to the user, then an appropriate differential equation routine must be supplied by the user, and the initialization of parameters and the call statement for LSODE in the main program must be changed accordingly

Model for absorption and release of gaseous materials by forest canopies

International Nuclear Information System (INIS)

Murphy, C.E. Jr.

1976-01-01

A model of the physical processes defining the absorption and release of materials by a forest canopy has been developed. The model deals with the turbulent transport of gaseous materials in the surface boundary layer near the canopy, the turbulent transport in the canopy atmosphere, the transport through the boundary layer near the leaf and soil surface, and the solution of the gaseous materials in intracellular fluids and subsequent diffusion into the leaf cells. The model is used to simulate the uptake of molecular tritium by the forest canopy and the subsequent release of tritiated water. Results of dynamic simulations of tritium uptake and release are compared with data collected at the time of a release of molecular tritium to the atmosphere

Predictive property models for use in design of controlled release of pesticides

DEFF Research Database (Denmark)

Suné, Nuria Muro; Gani, Rafiqul; Bell, G.

2005-01-01

A model capable of predicting the release of an Active Ingredient (AI) from a specific device would be very useful in the field of pesticide controlled release technology for design purposes. For the release of an AI from a microcapsule a mathematical model is briefly presented here, as an introd...

Calculation of fission gases internal pressure in nuclear fuel rods

International Nuclear Information System (INIS)

Vasconcelos Santana, M. de.

1981-12-01

Models concerning the principal phenomena, particularly thermal expansion, fuel swelling, densification, reestructuring, relocation, mechanical strain, fission gas production and release, direct or indirectly important to calculate the internal pressure in nuclear fuel rods were analysed and selected. Through these analyses a computer code was developed to calculate fuel pin internal pressure evolution. Three different models were utilized to calculate the internal pressure in order to select the best and the most conservative estimate. (Author) [pt

Fuel behaviour calculations with version 2.0 of the code FUROM

International Nuclear Information System (INIS)

Kulacsy, K.

2011-01-01

The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. In 2010 the new version of the code, FUROM-2.0 was released. Calculations performed with this version and results are presented. (author)

Atmospheric release advisory capability

International Nuclear Information System (INIS)

Sullivan, T.J.

1981-01-01

The ARAC system (Atmospheric Release Advisory Capability) is described. The system is a collection of people, computers, computer models, topographic data and meteorological input data that together permits a calculation of, in a quasi-predictive sense, where effluent from an accident will migrate through the atmosphere, where it will be deposited on the ground, and what instantaneous and integrated dose an exposed individual would receive

The FLUKA atmospheric neutrino flux calculation

CERN Document Server

Battistoni, G.; Montaruli, T.; Sala, P.R.

2003-01-01

The 3-dimensional (3-D) calculation of the atmospheric neutrino flux by means of the FLUKA Monte Carlo model is here described in all details, starting from the latest data on primary cosmic ray spectra. The importance of a 3-D calculation and of its consequences have been already debated in a previous paper. Here instead the focus is on the absolute flux. We stress the relevant aspects of the hadronic interaction model of FLUKA in the atmospheric neutrino flux calculation. This model is constructed and maintained so to provide a high degree of accuracy in the description of particle production. The accuracy achieved in the comparison with data from accelerators and cross checked with data on particle production in atmosphere certifies the reliability of shower calculation in atmosphere. The results presented here can be already used for analysis by current experiments on atmospheric neutrinos. However they represent an intermediate step towards a final release, since this calculation does not yet include the...

Development of dose assessment code for accidental releases of activation products

International Nuclear Information System (INIS)

Noguchi, H.; Yokoyama, S.

2000-01-01

It is expected that activation products will be important radionuclides as well as tritium in the assessment of the public doses necessary for licensing of a future fusion reactor. In order to calculate the public doses due to the activation products released in cases of accidents, a code named ACUTAP (dose assessment code for ACUTe Activation Product releases) has been developed. Major characteristics of the code are as follows: (1) the transfer model reflects specific behavior of the activation products in the environment, (2) the doses are assessed based on ICRP dose models, (3) it is possible to calculate individual doses using annual meteorological data statistically according to the guide of the Nuclear Safety Commission of Japan, and (4) the code can calculate collective doses as well as individual doses. Individual doses are calculated for the following pathways: internal exposure by inhalation of activation products in a plume and those resuspended from the ground, external exposure from a plume (cloudshine), and external exposure from activation products deposited on the ground (groundshine). The inhalation in a plume and cloudshine pathways are included in the model for calculating collective doses. In addition to parent nuclides released from the facilities, progeny nuclides produced during the atmospheric dispersion are considered in calculating inhalation doses, and those during the deposition period in calculating groundshine doses. External doses from the cloudshine are calculated for 18 energy groups instead of individual energy of emitted gamma rays in order to save the computation time. Atmospheric concentrations are calculated using a Gaussian plume model with atmospheric dispersion parameters prescribed in the guide of the Nuclear Safety Commission of Japan. Data sets of parameters necessary for the dose assessment, such as internal dos coefficients, external dose rate conversion factors and half lives, are prepared for about 100 radionuclides

Transient modeling of electrochemically assisted CO2 capture and release

DEFF Research Database (Denmark)

Singh, Shobhana; Stechel, Ellen B.; Buttry, Daniel A.

2017-01-01

to analyze the time-dependent behavior of CO2 capture and electro-migration transport across the cell length. Given high nonlinearity of the system, we used a finite element method (FEM) to numerically solve the coupled mass transport equations. The model describes the concentration profiles by taking......The present work aims to develop a model of a new electrochemical CO2 separation and release technology. We present a one-dimensional transient model of an electrochemical cell for point source CO2 capture and release, which mainly focuses on the simultaneous mass transport and complex chemical...... reactions associated with the separation process. For concreteness, we use an ionic liquid (IL) with 2 M thiolate anion (RS−) in 1 M disulfide (RSSR) as an electrolyte in the electrochemical cell to capture, transport and release CO2 under standard operating conditions. We computationally solved the model...

Modeling the wind-fields of accidental releases with an operational regional forecast model

International Nuclear Information System (INIS)

Albritton, J.R.; Lee, R.L.; Sugiyama, G.

1995-01-01

The Atmospheric Release Advisory Capability (ARAC) is an operational emergency preparedness and response organization supported primarily by the Departments of Energy and Defense. ARAC can provide real-time assessments of atmospheric releases of radioactive materials at any location in the world. ARAC uses robust three-dimensional atmospheric transport and dispersion models, extensive geophysical and dose-factor databases, meteorological data-acquisition systems, and an experienced staff. Although it was originally conceived and developed as an emergency response and assessment service for nuclear accidents, the ARAC system has been adapted to also simulate non-radiological hazardous releases. For example, in 1991 ARAC responded to three major events: the oil fires in Kuwait, the eruption of Mt. Pinatubo in the Philippines, and the herbicide spill into the upper Sacramento River in California. ARAC's operational simulation system, includes two three-dimensional finite-difference models: a diagnostic wind-field scheme, and a Lagrangian particle-in-cell transport and dispersion scheme. The meteorological component of ARAC's real-time response system employs models using real-time data from all available stations near the accident site to generate a wind-field for input to the transport and dispersion model. Here we report on simulation studies of past and potential release sites to show that even in the absence of local meteorological observational data, readily available gridded analysis and forecast data and a prognostic model, the Navy Operational Regional Atmospheric Prediction System, applied at an appropriate grid resolution can successfully simulate complex local flows

Calculation model for predicting concentrations of radioactive corrostion products in the primary coolant of boiling water reactors

International Nuclear Information System (INIS)

Uchida, S.; Kikuchi, M.; Asakura, Y.; Yusa, H.; Ohsumi, K.

1978-01-01

A calculation model was developed to predict the shutdown dose rate around the recirculation pipes and their components in boiling water reactors (BWRs) by simulating the corrosion product transport in primary cooling water. The model is characterized by separating cobalt species in the water into soluble and insoluble materials and then calculating each concentration using the following considerations: (1) Insoluble cobalt (designated as crud cobalt is deposited directly on the fuel surface, while soluble cobalt (designated as ionic cobalt) is adsorbed on iron oxide deposits on the fuel surface. (2) Cobalt-60 activated on the fuel surface is dissolved in the water in an ionic form, and some is released with iron oxide as crud. The model can follow the reduction of 60 Co in the primary cooling water caused by the control of the iron feed rate into the reactor, which decreases the iron oxide deposits on the fuel surface and then reduces the cobalt adsorption rate. The calculated results agree satisfactorily with the measurements in several BWR plants

Hybrid reduced order modeling for assembly calculations

Energy Technology Data Exchange (ETDEWEB)

Bang, Youngsuk, E-mail: ysbang00@fnctech.com [FNC Technology, Co. Ltd., Yongin-si (Korea, Republic of); Abdel-Khalik, Hany S., E-mail: abdelkhalik@purdue.edu [Purdue University, West Lafayette, IN (United States); Jessee, Matthew A., E-mail: jesseema@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mertyurek, Ugur, E-mail: mertyurek@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2015-12-15

Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.

Hybrid reduced order modeling for assembly calculations

International Nuclear Information System (INIS)

Bang, Youngsuk; Abdel-Khalik, Hany S.; Jessee, Matthew A.; Mertyurek, Ugur

2015-01-01

Highlights: • Reducing computational cost in engineering calculations. • Reduced order modeling algorithm for multi-physics problem like assembly calculation. • Non-intrusive algorithm with random sampling. • Pattern recognition in the components with high sensitive and large variation. - Abstract: While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system.

Air modelling as an alternative to sampling for low-level radioactive airborne releases

International Nuclear Information System (INIS)

Morgenstern, M.Y.; Hueske, K.

1995-01-01

This paper describes our efforts to assess the effect of airborne releases at one DOE laboratory using air modelling based on historical data. Among the facilities affected by these developments is Los Alamos National Laboratory (LANL) in New Mexico. RCRA, as amended by the Hazardous and Solid Waste Amendments (HSWA) in 1984, requires all facilities which involve the treatment, storage, and disposal of hazardous waste obtain a RCRA/HSWA waste facility permit. LANL complied with CEARP by initiating a process of identifying potential release sites associated with LANL operations prior to filing a RCRA/HSWA permit application. In the process of preparing the RCRA/HSWA waste facility permit application to the U.S. Environmental Protection Agency (EPA), a total of 603 Solid Waste Management Units (SWMUs) were identified as part of the requirements of the HSWA Module VIH permit requirements. The HSWA Module VIII permit requires LANL to determine whether there have been any releases of hazardous waste or hazardous constituents from SWMUs at the facility dating from the 1940's by performing a RCRA Facility Investigation to address known or suspected releases from specified SWMUs to affected media (i.e. soil, groundwater, surface water, and air). Among the most troublesome of the potential releases sites are those associated with airborne radioactive releases. In order to assess health risks associated with radioactive contaminants in a manner consistent with exposure standards currently in place, the DOE and LANL have established Screening Action Levels (SALs) for radioactive soil contamination. The SALs for each radionuclide in soil are derived from calculations based on a residential scenario in which individuals are exposed to contaminated soil via inhalation and ingestion as well as external exposure to gamma emitters in the soil. The applicable SALs are shown

An evaluation of gas release modelling approaches as to their applicability in fuel behaviour models

International Nuclear Information System (INIS)

Mattila, L.J.; Sairanen, R.T.

1980-01-01

The release of fission gas from uranium oxide fuel to the voids in the fuel rod affects in many ways the behaviour of LWR fuel rods both during normal operating conditions including anticipated transients and during off-normal and accident conditions. The current trend towards significantly increased discharge burnup of LWR fuel will increase the importance of fission gas release considerations both from the design and safety viewpoints. In the paper fission gas release models are classified to 5 categories on the basis of complexity and physical sophistication. For each category, the basic approach common to the models included in the category is described, a few representative models of the category are singled out and briefly commented in some cases, the advantages and drawbacks of the approach are listed and discussed and conclusions on the practical feasibility of the approach are drawn. The evaluation is based on both literature survey and our experience in working with integral fuel behaviour models. The work has included verification efforts, attempts to improve certain features of the codes and engineering applications. The classification of fission gas release models regarding their applicability in fuel behaviour codes can of course be done only in a coarse manner. The boundaries between the different categories are vague and a model may be well refined in a way which transfers it to a higher category. Some current trends in fuel behaviour research are discussed which seem to motivate further extensive efforts in fission product release modelling and are certain to affect the prioritizing of the efforts. (author)

Sodium spray release accident analysis for fast reactor safety studies

International Nuclear Information System (INIS)

Shire, P.R.

1976-10-01

A computer code has been developed to model the effects of postulated sodium spray release from LMFBR piping. The calculation method utilizes gas convection and droplet combustion theory to calculate the pressure and temperature effects in a range of 0 to 21% oxygen. Droplet motion and large aggregate sodium surface area result in rapid release of combustion and sensible heat causing a nearly adiabtic pressure pulse peaking in a matter of seconds. This analytical tool has indicated good agreement with prototypic test data for a range of oxygen concentrations, cell volumes and droplet sizes

Analysis of an LPG accidental release

Energy Technology Data Exchange (ETDEWEB)

Demichela, M.; Piccinini, N.; Poggio, A.

2004-03-01

An industrial LPG storage accident due to the release of propane from a tanker is described in this paper and the sequence of events that led to the collapse of a storage tank is examined using simulation software. A model taking both released flows into account as a radiant thermal contribution in the presence of flames is presented. The progression of the accident is assessed through digitalization and processing of the film of the fireball produced by about 500 L of LPG, whose diameter was calculated as approximately 50 m. (author)

Assessment of particle size distribution in CO 2 accidental releases

NARCIS (Netherlands)

Hulsbosch-Dam, C.E.C.; Spruijt, M.P.N.; Necci, A.; Cozzani, V.

2012-01-01

A model was developed to calculate the particle size distribution following the release of pressurised supercritical CO 2. The model combines several sub-models for the different stages of jet break-up and specifically addresses the possible formation of solid particles, which is important for CO 2

Multi-compartment iodine calculations with FIPLOC/IMPAIR

Energy Technology Data Exchange (ETDEWEB)

Ewig, F; Allelein, H J [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Schwarz, S; Weber, G [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Garching (Germany)

1996-12-01

The multi-compartment containment code FIPLOC for the simulation of severe accidents in LWR plants was extended by the integration of the iodine model IMPAIR-3. The iodine model was changed for arbitrary compartment configurations and tightly coupled to the thermal hydraulic part. A main progress with the coupled version FIPLOC-3.0 is the sophisticated modelling of the aerosol iodine behaviour. In a PWR accident the mass of iodine is mainly released in form of CsI aerosol from the primary circuit. In IMPAIR-3 the aerosol behaviour of the species CsI, AgI and IO{sub 3}{sup -} is modelled in a very simplified way causing large uncertainties in the calculated distributions. The behaviour of these three aerosol species is treated by the aerosol model MAEROS/MGA. Agglomeration, particle growth by condensation and all deposition processes are calculated. The solubility effect for the hygroscopic species CsI and IO{sub 3}{sup -} are comprehended. Furthermore the impact of the iodine decay heat on the thermal hydraulic behaviour is considered. In order to test the code development a preliminary FIPLOC-3.0 calculation was done simulating a German PWR containment for the core melt scenario ND* according to the German risk study phase B. IN the calculation a contact of the core melt with the sump water was assumed and the containment vent line was opened after 70 hours. The result show that the different iodine species are distributed inhomogeneously within the containment. The CsI-aerosol concentrations differ by two orders of magnitude and the I{sub 2}-concentration even by three orders of magnitude. Most of the iodine is assumed to be released as CsI aerosol out of the primary circuit. Since it fastly deposits its contribution to the release into the environment is minor. CsI is however dissolved in the sump, where mainly the gaseous I{sub 2} is created which can react in the containment atmosphere to IO{sub 3}{sup -}. (author) 11 figs., 3 tabs., 12 refs.

Multi-compartment iodine calculations with FIPLOC/IMPAIR

International Nuclear Information System (INIS)

Ewig, F.; Allelein, H.J.; Schwarz, S.; Weber, G.

1996-01-01

The multi-compartment containment code FIPLOC for the simulation of severe accidents in LWR plants was extended by the integration of the iodine model IMPAIR-3. The iodine model which originally was only drafted for chains of compartments was changed for arbitrary compartment configurations and tightly coupled to the thermal hydraulic part. A main progress with the coupled version FIPLOC-3.0 is the sophisticated modelling of the aerosol iodine behaviour. In a PWR accident the mass of iodine is mainly released in form of CsI aerosol from the primary circuit. In IMPAIR-3 the aerosol behaviour of the species CsI, AgI and IO 3 - is modelled in a very simplified way causing large uncertainties in the calculated distributions. The behaviour of these three aerosol species is treated by the aerosol model MAEROS/MGA. Agglomeration, particle growth by condensation and all deposition processes are calculated. The solubility effect for the hygroscopic species CsI and IO 3 - are comprehended. Furthermore the impact of the iodine decay heat on the thermal hydraulic behaviour is considered. In order to test the code development a preliminary FIPLOC-3.0 calculation was done simulating a German PWR containment for the core melt scenario ND* according to the German risk study phase B. IN the calculation a contact of the core melt with the sump water was assumed and the containment vent line was opened after 70 hours. The result show that the different iodine species are distributed inhomogeneously within the containment. The CsI-aerosol concentrations differ by two orders of magnitude and the I 2 -concentration even by three orders of magnitude. Most of the iodine is assumed to be released as CsI aerosol out of the primary circuit. Since it fastly deposits its contribution to the release into the environment is minor. CsI is however dissolved in the sump, where mainly the gaseous I 2 is created which can react in the containment atmosphere to IO 3 - . (author) 11 figs., 3 tabs., 12

Modelling of Pb release during Portland cement alteration

Energy Technology Data Exchange (ETDEWEB)

Benard, A. [INERIS Mediterrannee, F-13545 Aix En Provence 04 (France); Rose, J.; Borschneck, D.; Bottero, J.Y. [Univ Paul Cezanne, CNRS, UMR 6635, CEREGE, IFR PMSE 112, F-13545 Aix En Provence, (France); Hazemann, J.L. [CNRS, Cristallog Lab, F-38042 Grenoble 09 (France); Proux, O. [Univ Grenoble 1, CNRS, UMR, LGIT, F-38400 St Martin Dheres (France); Trotignon, L. [CEA Cadarache, DTN, SMTM, Lab Modelisat Transferts Environm, 13 - Saint Paul lez Durance (France); Nonat, A. [Univ Bourgogne, CNRS, UMR 5613, Fac Sci Mirande, Lab Rech Reactivite Solides, F-21078 Dijon (France); Chateau, L. [ADEME, F-49004 Angers (France)

2009-07-01

Complex cementitious matrices undergo weathering with environmental exchange and can release metallic pollutants during alteration. The molecular mechanisms responsible for metal release are difficult to identify, though this is necessary if such processes are to be controlled. The present study determines and models the molecular mechanisms of Pb release during Portland cement leaching. As Pb release is strongly related to its speciation (i.e. atomic environment and the nature of bearing phases), the first objective of the present study was to investigate the evolution of Pb retention sites together with the evolution of the cement mineralogy during leaching. Complementary and efficient investigation tools were used, namely X-ray diffraction, micro-X-ray fluorescence and X-ray absorption fine structures. The second objective was to reproduce our results with a reactive transport code (CHESS/HYTEC) in order to test the proposed speciation model of Pb. Combined results indicate that in both the unaltered core and the altered layer of the leached cement, Pb(II) would be retained through C-S-H 'nano-structure', probably linked to a Q(1) or Q(2P) silicate tetrahedra. Moreover in the altered layer, the presence of Fe atoms in the atomic environment of Pb is highly probable. Unfortunately little is known about Fe phases in cement, which makes the interpretation difficult. Can Fe-substituted hydrogranet (C(3)AH(6)) be responsible for Pb retention? Modelling results were consistent with Pb retention through C-S-H in layers and also in an additional, possibly Fe-containing, Pb-retention phase in the altered layer. (authors)

Radionuclide release calculations for selected severe accident scenarios

International Nuclear Information System (INIS)

Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A.

1990-08-01

This report provides the results of source term calculations that were performed in support of the NUREG-1150 study. ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants.'' This is the sixth volume of a series of reports. It supplements results presented in the earlier volumes. Analyses were performed for three of the NUREG-1150 plants: Peach Bottom, a Mark I, boiling water reactor; Surry, a subatmospheric containment, pressurized water reactor; and Sequoyah, an ice condenser containment, pressurized water reactor. Complete source term results are presented for the following sequences: short term station blackout with failure of the ADS system in the Peach Bottom plant; station blackout with a pump seal LOCA for the Surry plant; station blackout with a pump seal LOCA in the Sequoyah plant; and a very small break with loss of ECC and spray recirculation in the Sequoyah plant. In addition, some partial analyses were performed which did not require running all of the modules of the Source Term Code Package. A series of MARCH3 analyses were performed for the Surry and Sequoyah plants to evaluate the effects of alternative emergency operating procedures involving primary and secondary depressurization on the progress of the accident. Only thermal-hydraulic results are provided for these analyses. In addition, three accident sequences were analyzed for the Surry plant for accident-induced failure of steam generator tubes. In these analyses, only the transport of radionuclides within the primary system and failed steam generator were examined. The release of radionuclides to the environment is presented for the phase of the accident preceding vessel meltthrough. 17 refs., 176 figs., 113 tabs

GAMMA-CLOUD: a computer code for calculating gamma-exposure due to a radioactive cloud released from a point source

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, O [Chugoku Electric Power Co. Inc., Hiroshima (Japan); Sawaguchi, Y; Kaneko, M

1979-03-01

A computer code, designated GAMMA-CLOUD, has been developed by specialists of electric power companies to meet requests from the companies to have a unified means of calculating annual external doses from routine releases of radioactive gaseous effluents from nuclear power plants, based on the Japan Atomic Energy Commission's guides for environmental dose evaluation. GAMMA-CLOUD is written in FORTRAN language and its required capacity is less than 100 kilobytes. The average ..gamma..-exposure at an observation point can be calculated within a few minutes with comparable precision to other existing codes.

H2S release rate assessment guidelines and audit forms

International Nuclear Information System (INIS)

Johnson, S.; Wilson, T.; Long, R.; Shewan, K.; Nchkalo, H.; Nelson, R.; Morand, M.

1998-01-01

Development of a process to evaluate and calculate potential hydrogen sulfide release has been recommended by the Canadian Association of Petroleum Producers' Drilling and Completion Committee, and the Alberta Energy and Utilities Board. To facilitate the process, CAPP has released guidelines describing a methodology and standard for the industry to calculate the potential hydrogen sulfide release rates of a well, and a standardized format for the documentation and retention of data. The Guidelines describe a four-step process, with each step having an increasing degree of complexity. Step One describes the zone hydrogen sulfide exclusion area maps, defining when reporting is to be used (based on well location and distance from inhabited areas). Step Two provides details of the recommended method for determining hydrogen sulfide release through a data search process. Step Three consists of a series of instructions in the use of geologic analogs, data editing and wellbore design to further refine the cumulative hydrogen sulfide release rate. Step Four contains information designed to assist in detailed geological and reservoir modeling. It is not necessary to use all four steps in all cases. The user, however, is advised to use sound engineering judgement and due diligence in the calculation decisions. Sample calculations are provided for a variety of different situations. Measurement techniques are described in an appendix. A completed example of an audit form is attached. 10 + 6 refs., tabs

Release and diffusional modeling of metronidazole lipid matrices.

Science.gov (United States)

Ozyazici, Mine; Gökçe, Evren H; Ertan, Gökhan

2006-07-01

In this study, the first aim was to investigate the swelling and relaxation properties of lipid matrix on diffusional exponent (n). The second aim was to determine the desired release profile of metronidazole lipid matrix tablets. We prepared metronidazole lipid matrix granules using Carnauba wax, Beeswax, Stearic acid, Cutina HR, Precirol ATO 5, and Compritol ATO 888 by hot fusion method and pressed the tablets of these granules. In vitro release test was performed using a standard USP dissolution apparatus I (basket method) with a stirring rate of 100 rpm at 37 degrees C in 900 ml of 0.1 N hydrochloric acid, adjusted to pH 1.2, as medium for the formulations' screening. Hardness, diameter-height ratio, friability, and swelling ratio were determined. Target release profile of metronidazole was also drawn. Stearic acid showed the highest and Carnauba wax showed the lowest release rates in all formulations used. Swelling ratios were calculated after the dissolution of tablets as 9.24%, 6.03%, 1.74%, and 1.07% for Cutina HR, Beeswax, Precirol ATO 5, and Compritol ATO 888, respectively. There was erosion in Stearic acid, but neither erosion nor swelling in Carnauba wax, was detected. According to the power law analysis, the diffusion mechanism was expressed as pure Fickian for Stearic acid and Carnauba wax and the coupling of Fickian and relaxation contributions for other Cutina HR, Beeswax, Compritol ATO 888, and Precirol ATO 5 tablets. It was found that Beeswax (kd=2.13) has a very close drug release rate with the target profile (kt=1.95). Our results suggested that swelling and relaxation properties of lipid matrices should be examined together for a correct evaluation on drug diffusion mechanism of insoluble matrices.

Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

International Nuclear Information System (INIS)

Verfondern, K.; Martin, R.C.; Moormann, R.

1993-01-01

The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.) [de

Characteristics of fission product release from a molten pool

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2001-01-01

The volatile fission products are released from the debris pool, while the less volatile fission products tend to remain as condensed phases because of their low vapor pressure. The release of noble gases and the volatile fission products is dominated by bubble dynamics. The release of the less volatile fission products from the pool can be analyzed based on mass transport through a liquid with the convection flow. The physico-numerical models were orchestrated from existing submodels in various disciplines of engineering to estimate the released fraction of fission products from a molten pool. It was assumed that the pool has partially filled hemispherical geometry. For the high pool pressure, the diameter of the bubbles at detachment was calculated utilizing the Cole and Shulman correlation with the effect of system pressure. Sensitivity analyses were performed and results of the numerical calculations were compared with analysis results for the TMI-2 accident. (author)

Effects of sorption hysteresis on radionuclide releases from waste packages

International Nuclear Information System (INIS)

Barney, G.S.; Reed, D.T.

1985-01-01

A one-dimensional, numerical transport model was used to calculate radionuclide releases from waste packages emplaced in a nuclear waste repository in basalt. The model incorporates both sorption and desorption isotherm parameters measured previously for sorption of key radionuclides on the packing material component of the waste package. Sorption hysteresis as described by these isotherms lowered releases of some radionuclides by as much as two orders of magnitude. Radionuclides that have low molar inventories (relative to uranium), high solubility, and strongly sorbed, are most affected by sorption hysteresis. In these cases, almost the entire radionuclide inventory is sorbed on the packing material. The model can be used to help optimize the thickness of the packing material layer by comparing release rate versus packing material thickness curves with Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) release limits

Factors controlling alkali salt deposition in recovery boilers. Release mechanisms

Energy Technology Data Exchange (ETDEWEB)

McKeough, P; Kurkela, M; Kylloenen, H; Tapola, E [VTT Energy, Espoo (Finland). Process Technology Group

1997-10-01

The research was part of an ongoing cooperative research effort aimed at developing a model to describe the behaviour of inorganic compounds in kraft recovery boilers. During 1996 experimental investigations of sulphur release were continued. Experiments at elevated pressures and employing larger particle sizes were performed in order to gain information about mass transfer effects. The first experiments yielding data on the rates of the sulphur-release reactions were performed. This data will be used as the basis of a drop model for sulphur release being developed in cooperation with another research group. The other part of the work during 1996 explored the possibility of using chemical equilibrium calculations to predict the release of sodium, potassium and chlorine in the recovery furnace. The approach is essentially different from that employed in earlier studies in that the effects of fume formation are taken into account. So far, the predictions of the chemical equilibrium release model have, in no way, conflicted with field measurements. (orig.)

Historical Doses from Tritiated Water and Tritiated Hydrogen Gas Released to the Atmosphere from Lawrence Livermore National Laboratory (LLNL). Part 5. Accidental Releases

Energy Technology Data Exchange (ETDEWEB)

Peterson, S

2007-08-15

Over the course of fifty-three years, LLNL had six acute releases of tritiated hydrogen gas (HT) and one acute release of tritiated water vapor (HTO) that were too large relative to the annual releases to be included as part of the annual releases from normal operations detailed in Parts 3 and 4 of the Tritium Dose Reconstruction (TDR). Sandia National Laboratories/California (SNL/CA) had one such release of HT and one of HTO. Doses to the maximally exposed individual (MEI) for these accidents have been modeled using an equation derived from the time-dependent tritium model, UFOTRI, and parameter values based on expert judgment. All of these acute releases are described in this report. Doses that could not have been exceeded from the large HT releases of 1965 and 1970 were calculated to be 43 {micro}Sv (4.3 mrem) and 120 {micro}Sv (12 mrem) to an adult, respectively. Two published sets of dose predictions for the accidental HT release in 1970 are compared with the dose predictions of this TDR. The highest predicted dose was for an acute release of HTO in 1954. For this release, the dose that could not have been exceeded was estimated to have been 2 mSv (200 mrem), although, because of the high uncertainty about the predictions, the likely dose may have been as low as 360 {micro}Sv (36 mrem) or less. The estimated maximum exposures from the accidental releases were such that no adverse health effects would be expected. Appendix A lists all accidents and large routine puff releases that have occurred at LLNL and SNL/CA between 1953 and 2005. Appendix B describes the processes unique to tritium that must be modeled after an acute release, some of the time-dependent tritium models being used today, and the results of tests of these models.

Atmospheric dispersion of radioactive releases: Computer code DIASPORA

International Nuclear Information System (INIS)

Synodinou, B.M.; Bartzis, J.M.

1982-05-01

The computer code DIASPORA is presented. Air and ground concentrations of an airborne radioactive material released from an elevated continuous point source are calculated using Gaussian plume models. Dry and wet deposition as well as plume rise effects are taken into consideration. (author)

Computer programs for developing source terms for a UF{sub 6} dispersion model to simulate postulated UF{sub 6} releases from buildings

Energy Technology Data Exchange (ETDEWEB)

Williams, W.R.

1985-03-01

Calculational methods and computer programs for the analysis of source terms for postulated releases of UF{sub 6} are presented. Required thermophysical properties of UF{sub 6}, HF, and H{sub 2}O are described in detail. UF{sub 6} reacts with moisture in the ambient environment to form HF and H{sub 2}O. The coexistence of HF and H{sub 2}O significantly alters their pure component properties, and HF vapor polymerizes. Transient compartment models for simulating UF{sub 6} releases inside gaseous diffusion plant feed and withdrawl buildings and cascade buildings are also described. The basic compartment model mass and energy balances are supported by simple heat transfer, ventilation system, and deposition models. A model that can simulate either a closed compartment or a steady-state ventilation system is also discussed. The transient compartment models provide input to an atmospheric dispersion model as output.

Cyclodextrin Controlled Release of Poorly Water-Soluble Drugs from Hydrogels

DEFF Research Database (Denmark)

Woldum, Henriette Sie; Madsen, Flemming; Larsen, Kim Lambertsen

2008-01-01

The effect of 2-hydroxypropyl- -cyclodextrin and -cyclodextrin on the release of ibuprofen, ketoprofen and prednisolone was studied. Stability constants calculated for inclusion complexes show size dependence for complexes with both cyclodextrins. Hydrogels were prepared by ultraviolet irradiation...... and release of each model drug was studied. For drugs formulated using cyclodextrins an increase in the achievable concentration and in the release from hydrogels was obtained due to increased solubility, although the solubility of all -cyclodextrin complexes was limited. The load also was increased...

Dispersion parameters: impact on calculated reactor accident consequences

Energy Technology Data Exchange (ETDEWEB)

Aldrich, D.C.

1979-01-01

Much attention has been given in recent years to the modeling of the atmospheric dispersion of pollutants released from a point source. Numerous recommendations have been made concerning the choice of appropriate dispersion parameters. A series of calculations has been performed to determine the impact of these recommendations on the calculated consequences of large reactor accidents. Results are presented and compared in this paper.

Statistical Model Calculations for (n,γ Reactions

Directory of Open Access Journals (Sweden)

Beard Mary

2015-01-01

Full Text Available Hauser-Feshbach (HF cross sections are of enormous importance for a wide range of applications, from waste transmutation and nuclear technologies, to medical applications, and nuclear astrophysics. It is a well-observed result that different nuclear input models sensitively affect HF cross section calculations. Less well known however are the effects on calculations originating from model-specific implementation details (such as level density parameter, matching energy, back-shift and giant dipole parameters, as well as effects from non-model aspects, such as experimental data truncation and transmission function energy binning. To investigate the effects or these various aspects, Maxwellian-averaged neutron capture cross sections have been calculated for approximately 340 nuclei. The relative effects of these model details will be discussed.

A new mechanistic and engineering fission gas release model for a uranium dioxide fuel

International Nuclear Information System (INIS)

Lee, Chan Bock; Yang, Yong Sik; Kim, Dae Ho; Kim, Sun Ki; Bang, Je Geun

2008-01-01

A mechanistic and engineering fission gas release model (MEGA) for uranium dioxide (UO 2 ) fuel was developed. It was based upon the diffusional release of fission gases from inside the grain to the grain boundary and the release of fission gases from the grain boundary to the external surface by the interconnection of the fission gas bubbles in the grain boundary. The capability of the MEGA model was validated by a comparison with the fission gas release data base and the sensitivity analyses of the parameters. It was found that the MEGA model correctly predicts the fission gas release in the broad range of fuel burnups up to 98 MWd/kgU. Especially, the enhancement of fission gas release in a high-burnup fuel, and the reduction of fission gas release at a high burnup by increasing the UO 2 grain size were found to be correctly predicted by the MEGA model without using any artificial factor. (author)

Equilibrium and kinetic models for colloid release under transient solution chemistry conditions

Science.gov (United States)

We present continuum models to describe colloid release in the subsurface during transient physicochemical conditions. Our modeling approach relates the amount of colloid release to changes in the fraction of the solid surface area that contributes to retention. Equilibrium, kinetic, equilibrium and...

A refined method for calculating equivalent effective stratospheric chlorine

Science.gov (United States)

Engel, Andreas; Bönisch, Harald; Ostermöller, Jennifer; Chipperfield, Martyn P.; Dhomse, Sandip; Jöckel, Patrick

2018-01-01

Chlorine and bromine atoms lead to catalytic depletion of ozone in the stratosphere. Therefore the use and production of ozone-depleting substances (ODSs) containing chlorine and bromine is regulated by the Montreal Protocol to protect the ozone layer. Equivalent effective stratospheric chlorine (EESC) has been adopted as an appropriate metric to describe the combined effects of chlorine and bromine released from halocarbons on stratospheric ozone. Here we revisit the concept of calculating EESC. We derive a refined formulation of EESC based on an advanced concept of ODS propagation into the stratosphere and reactive halogen release. A new transit time distribution is introduced in which the age spectrum for an inert tracer is weighted with the release function for inorganic halogen from the source gases. This distribution is termed the release time distribution. We show that a much better agreement with inorganic halogen loading from the chemistry transport model TOMCAT is achieved compared with using the current formulation. The refined formulation shows EESC levels in the year 1980 for the mid-latitude lower stratosphere, which are significantly lower than previously calculated. The year 1980 is commonly used as a benchmark to which EESC must return in order to reach significant progress towards halogen and ozone recovery. Assuming that - under otherwise unchanged conditions - the EESC value must return to the same level in order for ozone to fully recover, we show that it will take more than 10 years longer than estimated in this region of the stratosphere with the current method for calculation of EESC. We also present a range of sensitivity studies to investigate the effect of changes and uncertainties in the fractional release factors and in the assumptions on the shape of the release time distributions. We further discuss the value of EESC as a proxy for future evolution of inorganic halogen loading under changing atmospheric dynamics using simulations from

Comparison of analytic source models for head scatter factor calculation and planar dose calculation for IMRT

International Nuclear Information System (INIS)

Yan Guanghua; Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G

2008-01-01

The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity

Comparison of analytic source models for head scatter factor calculation and planar dose calculation for IMRT

Energy Technology Data Exchange (ETDEWEB)

Yan Guanghua [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL 32611 (United States); Liu, Chihray; Lu Bo; Palta, Jatinder R; Li, Jonathan G [Department of Radiation Oncology, University of Florida, Gainesville, FL 32610-0385 (United States)

2008-04-21

The purpose of this study was to choose an appropriate head scatter source model for the fast and accurate independent planar dose calculation for intensity-modulated radiation therapy (IMRT) with MLC. The performance of three different head scatter source models regarding their ability to model head scatter and facilitate planar dose calculation was evaluated. A three-source model, a two-source model and a single-source model were compared in this study. In the planar dose calculation algorithm, in-air fluence distribution was derived from each of the head scatter source models while considering the combination of Jaw and MLC opening. Fluence perturbations due to tongue-and-groove effect, rounded leaf end and leaf transmission were taken into account explicitly. The dose distribution was calculated by convolving the in-air fluence distribution with an experimentally determined pencil-beam kernel. The results were compared with measurements using a diode array and passing rates with 2%/2 mm and 3%/3 mm criteria were reported. It was found that the two-source model achieved the best agreement on head scatter factor calculation. The three-source model and single-source model underestimated head scatter factors for certain symmetric rectangular fields and asymmetric fields, but similar good agreement could be achieved when monitor back scatter effect was incorporated explicitly. All the three source models resulted in comparable average passing rates (>97%) when the 3%/3 mm criterion was selected. The calculation with the single-source model and two-source model was slightly faster than the three-source model due to their simplicity.

Comparison of Nordic dose models

International Nuclear Information System (INIS)

Thykier-Nielsen, S.

1978-04-01

A comparison is made between the models used in the four Nordic countries, Finland, Norway, Sweden and Denmark, for calculation of concentrations and doses from releases of radioactive material to the atmosphere. The comparison is limited to the near-zone models, i.e. the models for calculation of concentrations and doses within 50 km from the release point, and it comprises the following types of calculation: a. Concentrations of airborne material, b. External gamma doses from a plume, c. External gamma doses from radioactive material deposited on the ground. All models are based on the gaussian dispersion model (the gaussian plume model). Unit releases of specific isotopes under specific meteorological conditions are assumed. On the basis of the calculation results from the models, it is concluded that there are no essential differences. The difference between the calculation results only exceeds a factor of 3 in special cases. It thus lies within the known limits of uncertainty for the gaussian plume model. (author)

WRAITH, Internal and External Doses from Atmospheric Release of Isotopes

International Nuclear Information System (INIS)

1984-01-01

1 - Description of problem or function: WRAITH calculates the atmospheric transport of radioactive material to each of a number of downwind receptor points and the external and internal doses to a reference man at each of the receptor points. 2 - Method of solution: The movement of the released material through the atmosphere is calculated using a bivariate straight-line Gaussian distribution model with Pasquill values for standard deviations. The quantity of material in the released cloud is modified during its transit time to account for radioactive decay and daughter production. External doses due to exposure to the cloud can be calculated using a semi-infinite cloud approximation or a 'finite plume' three-dimensional point-kernel numerical integration technique. Internal doses due to acute inhalation are calculated using the ICRP Task Group Model and a four-segmented gastro- intestinal tract model. Translocation of the material between body compartments and retention in the body compartments are calculated using multiple exponential retention functions. Internal doses to each organ are calculated as sums of cross-organ doses with each target organ irradiated by radioactive material in a number of source organs. All doses are calculated in rads with separate values determined for high-LET and low-LET radiation. 3 - Restrictions on the complexity of the problem: - Doses to only three target organs (total body, red bone marrow, and the lungs) are considered and acute inhalation is the only pathway for material to enter the body. The dose response model is not valid for high-LET radiation other than alphas. The high-LET calculation ignores the contributions of neutrons, spontaneous fission fragments, and alpha recoil nuclei

Studies on Fragrance Delivery from Inorganic Nanocontainers: Encapsulation, Release and Modeling Studies

Science.gov (United States)

Ghodke, Shailesh Adinath; Sonawane, Shirish Hari; Bhanvase, Bharat Apparao; Mishra, Satyendra; Joshi, Kalpana Shrikant

2015-04-01

The present work deals with encapsulation of fragrance molecule in inorganic nanocontainers substrate and investigation of its prolonged release at different pH condition. The nanocontainers used were aluminosilicate clay (Halloysite) having cylindrical shape with outside diameter in the range of 30-50 nm, 15 nm lumen and length equal to 800 ± 300 nm. Rosewater absolute was used as a sample fragrance for loading in nanocontainer and delivery purpose. The fragrance loaded nanocontainers were coated with a thin layer of polyelectrolyte i.e. Polyacrylic Acid (PAA). The structural characteristics of prepared nanocontainers were determined by using Fourier Transform Intra-red Spectroscopy (FTIR), Thermal Gravimetric Analysis (TGA) and UV spectroscopy analysis. Release of fragrance molecules in the aqueous medium was monitored for 24 h. The fragrance release was found to be responsive as the amount of fragrance release increases with increase in pH value from 3 to 7. Fragrance release has been studied by using various permeation kinetic models such as zero order, first order, Hixson-Crowell, Higuchi, Korsmeyer-Peppas and Hopfenberg models. Korsemyer-Peppas shows the best fit (R2 = 0.9544) compared to other kinetic model for the release of fragrance from nanocontainers.

Source-receptor matrix calculation with a Lagrangian particle dispersion model in backward mode

Directory of Open Access Journals (Sweden)

P. Seibert

2004-01-01

Full Text Available The possibility to calculate linear-source receptor relationships for the transport of atmospheric trace substances with a Lagrangian particle dispersion model (LPDM running in backward mode is shown and presented with many tests and examples. This mode requires only minor modifications of the forward LPDM. The derivation includes the action of sources and of any first-order processes (transformation with prescribed rates, dry and wet deposition, radioactive decay, etc.. The backward mode is computationally advantageous if the number of receptors is less than the number of sources considered. The combination of an LPDM with the backward (adjoint methodology is especially attractive for the application to point measurements, which can be handled without artificial numerical diffusion. Practical hints are provided for source-receptor calculations with different settings, both in forward and backward mode. The equivalence of forward and backward calculations is shown in simple tests for release and sampling of particles, pure wet deposition, pure convective redistribution and realistic transport over a short distance. Furthermore, an application example explaining measurements of Cs-137 in Stockholm as transport from areas contaminated heavily in the Chernobyl disaster is included.

Atmospheric Dispersion Modeling of the February 2014 Waste Isolation Pilot Plant Release

Energy Technology Data Exchange (ETDEWEB)

Nasstrom, John [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Piggott, Tom [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Simpson, Matthew [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lobaugh, Megan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tai, Lydia [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Pobanz, Brenda [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Yu, Kristen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-07-22

This report presents the results of a simulation of the atmospheric dispersion and deposition of radioactivity released from the Waste Isolation Pilot Plant (WIPP) site in New Mexico in February 2014. These simulations were made by the National Atmospheric Release Advisory Center (NARAC) at Lawrence Livermore National Laboratory (LLNL), and supersede NARAC simulation results published in a previous WIPP report (WIPP, 2014). The results presented in this report use additional, more detailed data from WIPP on the specific radionuclides released, radioactivity release amounts and release times. Compared to the previous NARAC simulations, the new simulation results in this report are based on more detailed modeling of the winds, turbulence, and particle dry deposition. In addition, the initial plume rise from the exhaust vent was considered in the new simulations, but not in the previous NARAC simulations. The new model results show some small differences compared to previous results, but do not change the conclusions in the WIPP (2014) report. Presented are the data and assumptions used in these model simulations, as well as the model-predicted dose and deposition on and near the WIPP site. A comparison of predicted and measured radionuclide-specific air concentrations is also presented.

A comparison of alternative methods of calculating complementary cumulative distribution functions of health effects following an atmospheric radioactive release

International Nuclear Information System (INIS)

Ponting, A.C.; Nair, S.

1984-04-01

A concept extensively used in studying the consequences of accidental atmospheric radioactive releases is that of the Complementary Cumulative Distribution Function, CCDF. Various methods of calculating CCDFs have been developed with particular applications in putting degraded core accidents in perspective and in identifying release sequences leading to high risks. This note compares three methods with specific reference to their accuracy and computational efficiency. For two of the methods (that used in the US Reactor Safety Study code CRAC2 and extended version of that method), the effects of varying the sector width and considering site-specific population distributions have been determined. For the third method it is only necessary to consider the effects of site-specific population distributions. (author)

Hybrid reduced order modeling for assembly calculations

Energy Technology Data Exchange (ETDEWEB)

Bang, Y.; Abdel-Khalik, H. S. [North Carolina State University, Raleigh, NC (United States); Jessee, M. A.; Mertyurek, U. [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2013-07-01

While the accuracy of assembly calculations has considerably improved due to the increase in computer power enabling more refined description of the phase space and use of more sophisticated numerical algorithms, the computational cost continues to increase which limits the full utilization of their effectiveness for routine engineering analysis. Reduced order modeling is a mathematical vehicle that scales down the dimensionality of large-scale numerical problems to enable their repeated executions on small computing environment, often available to end users. This is done by capturing the most dominant underlying relationships between the model's inputs and outputs. Previous works demonstrated the use of the reduced order modeling for a single physics code, such as a radiation transport calculation. This manuscript extends those works to coupled code systems as currently employed in assembly calculations. Numerical tests are conducted using realistic SCALE assembly models with resonance self-shielding, neutron transport, and nuclides transmutation/depletion models representing the components of the coupled code system. (authors)

ECP evaluation by water radiolysis and ECP model calculations

Energy Technology Data Exchange (ETDEWEB)

Hanawa, S.; Nakamura, T.; Uchida, S. [Japan Atomic Energy Agency, Tokai-mura, Ibaraki (Japan); Kus, P.; Vsolak, R.; Kysela, J. [Nuclear Research Inst. Rez plc, Rez (Czech Republic)

2010-07-01

In-pile ECP measurements data was evaluated by water radiolysis calculations. The data was obtained by using an in-pile loop in an experimental reactor, LVR-15, at the Nuclear Research Institute (NRI) in Czech Republic. Three types of ECP sensors, a Pt electrode, an Ag/AgCl sensor and a zirconia membrane sensor containing Ag/Ag{sub 2}O were used at several levels of the irradiation rig at various neutron flux and gamma rates. For water radiolysis calculation, the in-pile loop was modeled to several nodes following their design specifications, operating conditions such as flow rates, dose rate distributions of neutron and gamma-ray and so on. Concentration of chemical species along the water flow was calculated by a radiolysis code, WRAC-J. The radiolysis calculation results were transferred to an ECP model. In the model, anodic and cathodic current densities were calculated with combination of an electrochemistry model and an oxide film growth model. The measured ECP data were compared with the radiolysis/ECP calculation results, and applicability the of radiolysis model was confirmed. In addition, anomalous phenomenon appears in the in-pile loop was also investigated by radiolysis calculations. (author)

Freedom: a transient fission-product release model for radioactive and stable species

International Nuclear Information System (INIS)

Macdonald, L.D.; Lewis, B.J.; Iglesias, F.C.

1989-05-01

A microstructure-dependent fission-gas release and swelling model (FREEDOM) has been developed for UO 2 fuel. The model describes the transient release behaviour for both the radioactive and stable fission-product species. The model can be applied over the full range of operating conditions, as well as for accident conditions that result in high fuel temperatures. The model accounts for lattice diffusion and grain-boundary sweeping of fusion products to the grain boundaries, where the fission gases accumulate in grain-face bubbles as a result of vacancy diffusion. Release of fission-gas to the free void of the fuel element occurs through the interlinkage of bubbles and cracks on the grain boundaries. This treatment also accounts for radioactive chain decay and neutron-induced transmutation effects. These phenomena are described by mass balance equations which are numerically solved using a moving-boundary, finite-element method with mesh refinement. The effects of grain-face bubbles on fuel swelling and fuel thermal conductivity are included in the ELESIM fuel performance code. FREEDOM has an accuracy of better than 1% when assessed against an analytic solution for diffusional release. The code is being evaluated against a fuel performance database for stable gas release, and against sweep-gas and in-cell fission-product release experiments at Chalk River for active species

RAFT: a computer program for fault tree risk calculations

International Nuclear Information System (INIS)

Seybold, G.D.

1977-11-01

A description and user instructions are presented for RAFT, a FORTRAN computer code for calculation of a risk measure for fault tree cut sets. RAFT calculates release quantities and a risk measure based on the product of probability and release quantity for cut sets of fault trees modeling the accidental release of radioactive material from a nuclear fuel cycle facility. Cut sets and their probabilities are supplied as input to RAFT from an external fault tree analysis code. Using the total inventory available of radioactive material, along with release fractions for each event in a cut set, the release terms are calculated for each cut set. Each release term is multiplied by the cut set probability to yield the cut set risk measure. RAFT orders the dominant cut sets on the risk measure. The total risk measure of processed cut sets and their fractional contributions are supplied as output. Input options are available to eliminate redundant cut sets, apply threshold values on cut set probability and risk, and control the total number of cut sets output. Hash addressing is used to remove redundant cut sets from the analysis. Computer hardware and software restrictions are given along with a sample problem and cross-reference table of the code. Except for the use of file management utilities, RAFT is written exclusively in FORTRAN language and is operational on a Control Data, CYBER 74-18--series computer system. 4 figures

Surface-wave generation by underground nuclear explosions releasing tectonic strain

International Nuclear Information System (INIS)

Patton, H.J.

1980-01-01

Seismic surface-wave generation by underground nuclear explosions releasing tectonic strain is studied through a series of synthetic radiation-pattern calculations based on the earthquake-trigger model. From amplitude and phase radiation patterns for 20-s Rayleigh waves, inferences are made about effects on surface-wave magnitude, M/sub s/, and waveform character. The focus of this study is a comparison between two mechanisms of tectonic strain release: strike-slip motion on vertical faults and thrust motion on 45 0 dipping faults. The results of our calculations show that Rayleigh-wave amplitudes of the dip-slip model at F values between 0.75 and 1.5 are significantly lower than amplitudes of the strike-slip model or of the explosion source alone. This effect translates into M/sub s/ values about 0.5 units lower than M/sub s/ of the explosion alone. Waveform polarity reversals occur in two of four azimuthal quadrants for the strike-slip model and in all azimuths of the dip-slip-thrust model for F values above about 3. A cursory examination of waveforms from presumed explosions in eastern Kazakhstan suggests that releases of tectonic strain are accompanying the detonation of many of these explosions. Qualitatively, the observations seem to favor the dip-slip-thrust model, which, in the case of a few explosions, must have F values above 3

Predicting patient exposure to nickel released from cardiovascular devices using multi-scale modeling.

Science.gov (United States)

Saylor, David M; Craven, Brent A; Chandrasekar, Vaishnavi; Simon, David D; Brown, Ronald P; Sussman, Eric M

2018-04-01

Many cardiovascular device alloys contain nickel, which if released in sufficient quantities, can lead to adverse health effects. However, in-vivo nickel release from implanted devices and subsequent biodistribution of nickel ions to local tissues and systemic circulation are not well understood. To address this uncertainty, we have developed a multi-scale (material, tissue, and system) biokinetic model. The model links nickel release from an implanted cardiovascular device to concentrations in peri-implant tissue, as well as in serum and urine, which can be readily monitored. The model was parameterized for a specific cardiovascular implant, nitinol septal occluders, using in-vitro nickel release test results, studies of ex-vivo uptake into heart tissue, and in-vivo and clinical measurements from the literature. Our results show that the model accurately predicts nickel concentrations in peri-implant tissue in an animal model and in serum and urine of septal occluder patients. The congruity of the model with these data suggests it may provide useful insight to establish nickel exposure limits and interpret biomonitoring data. Finally, we use the model to predict local and systemic nickel exposure due to passive release from nitinol devices produced using a wide range of manufacturing processes, as well as general relationships between release rate and exposure. These relationships suggest that peri-implant tissue and serum levels of nickel will remain below 5 μg/g and 10 μg/l, respectively, in patients who have received implanted nitinol cardiovascular devices provided the rate of nickel release per device surface area does not exceed 0.074 μg/(cm 2  d) and is less than 32 μg/d in total. The uncertainty in whether in-vitro tests used to evaluate metal ion release from medical products are representative of clinical environments is one of the largest roadblocks to establishing the associated patient risk. We have developed and validated a multi

TTS-Polttopuu - cost calculation model for fuelwood

International Nuclear Information System (INIS)

Naett, H.; Ryynaenen, S.

1999-01-01

The TTS-Institutes's Forestry Department has developed a computer based cost-calculation model, 'TTS-Polttopuu', for the calculation of unit costs and resource needs in the harvesting systems for wood chips and split firewood. The model enables to determine the productivity and device cost per operating hour by each working stage of the harvesting system. The calculation model also enables the user to find out how changes in the productivity and cost bases of different harvesting chains influence the unit cost of the whole system. The harvesting chain includes the cutting of delimbed and non-delimbed fuelwood, forest haulage, road transportation, chipping and chopping of longwood at storage. This individually operating software was originally developed to serve research needs, but it also serves the needs of the forestry and agricultural education, training and extension as well as individual firewood producers. The system requirements for this cost calculation model are at least 486- level processor with the Windows 95/98 -operating system, 16 MB of memory (RAM) and 5 MB of available hard-disk. This development work was carried out in conjunction with the nation-wide BIOENERGY-research programme. (orig.)

Analysis of railroad tank car releases using a generalized binomial model.

Science.gov (United States)

Liu, Xiang; Hong, Yili

2015-11-01

The United States is experiencing an unprecedented boom in shale oil production, leading to a dramatic growth in petroleum crude oil traffic by rail. In 2014, U.S. railroads carried over 500,000 tank carloads of petroleum crude oil, up from 9500 in 2008 (a 5300% increase). In light of continual growth in crude oil by rail, there is an urgent national need to manage this emerging risk. This need has been underscored in the wake of several recent crude oil release incidents. In contrast to highway transport, which usually involves a tank trailer, a crude oil train can carry a large number of tank cars, having the potential for a large, multiple-tank-car release incident. Previous studies exclusively assumed that railroad tank car releases in the same train accident are mutually independent, thereby estimating the number of tank cars releasing given the total number of tank cars derailed based on a binomial model. This paper specifically accounts for dependent tank car releases within a train accident. We estimate the number of tank cars releasing given the number of tank cars derailed based on a generalized binomial model. The generalized binomial model provides a significantly better description for the empirical tank car accident data through our numerical case study. This research aims to provide a new methodology and new insights regarding the further development of risk management strategies for improving railroad crude oil transportation safety. Copyright © 2015 Elsevier Ltd. All rights reserved.

Influence of FGR complexity modelling on the practical results in gas pressure calculation of selected fuel elements from Dukovany NPP

International Nuclear Information System (INIS)

Lahodova, M.

2001-01-01

A modernization fuel system and advanced fuel for operation up to the high burnup are used in present time in Dukovany NPP. Reloading of the cores are evaluated using computer codes for thermomechanical behavior of the most loaded fuel rods. The paper presents results of parametric calculations performed by the NRI Rez integral code PIN, version 2000 (PIN2k) to assess influence of fission gas release modelling complexity on achieved results. The representative Dukovany NPP fuel rod irradiation history data are used and two cases of fuel parameter variables (soft and hard) are chosen for the comparison. Involved FGR models where the GASREL diffusion model developed in the NRI Rez plc and standard Weisman model that is recommended in the previous version of the PIN integral code. FGR calculation by PIN2k with GASREL model represents more realistic results than standard Weisman's model. Results for linear power, fuel centre temperature, FGR and gas pressure versus burnup are given for two fuel rods

ARANO - a computer program for the assessment of radiological consequences of atmospheric radioactive releases

International Nuclear Information System (INIS)

Savolainen, I.; Vuori, S.

1980-09-01

A short description of the calculation possibilities, methods and of the structure of the computer code system ARANO is given, in addition to the input quide. The code can be employed in the calculation of environmental radiological consequences caused by radioactive materials released to atmosphere. Results can be individual doses for different organs at given distances from the release point, collective doses, numbers of persons exceeding given dose limits, numbers of casualties, areas polluted by deposited activity and losses of investments or production due to radioactive contamination. Both a case with a single release and atmospheric dispersion situation and a group of radioactive release and dispersions with discrete probability distributions can be considered. If the radioactive releases or the dispersion conditions are described by probability distributions, the program assesses the magnitudes of the specified effects in all combinations of the release and dispersion situations and then calculates the expectation values and the cumulative probability distributions of the effects. The vertical mixing in the atmosphere is described with a Ksub(Z)-model. In the lateral direction the plume is assumed to be Gaussian, and the release duration can be taken into account in the σsub(y)-values. External gamma dose from the release plume is calculated on the basis of a data file which has been created by 3-dimensional integration. Dose due to inhalation and due to gamma radiation from the contaminated ground are calculated by using appropriate dose conversion factors, which are collected into two mutually alternative block data subprograms. (author)

HOTSPOT, Field Evaluation of Radiation Release from Nuclear Accident

International Nuclear Information System (INIS)

2001-01-01

information. HOTSPOT 8.03 includes an improved geographical mapping module and runs on Pentium II and III correcting a problem which generated intermittent runtime errors on Pentium computers in the previous 8.0 release. 2 - Method of solution: HOTSPOT uses the well-established Gaussian Plume Model, which is widely used for initial emergency assessment or safety analysis planning of a radionuclide release. A detailed description of the HOTSPOT Gaussian model equations can be found in the HOTSPOT documentation. The dosimetric methods of ICRP Publication 30 have been used throughout the HOTSPOT programs. Individual doses (unweighted) are produced, along with the 50-year committed effective dose equivalent (CEDE). HOTSPOT supports both CLASSIC units such as rem, rad, curie, and SI units. The Gaussian model is still the basic workhorse for atmospheric dispersion calculations and has found its way into most governmental guidebooks. The Gaussian model has also been used and accepted by the Environmental Protection Agency. The adequacy of this model for making initial dispersion estimates has been tested and verified for many years. HOTSPOT strictly follows a well-established implementation of the Gaussian model and uses no 'black-box' techniques. All algorithms are presented and fully referenced in the HOTSPOT documentation. 3 - Restrictions on the complexity of the problem: This software is intended as a screening tool only for the initial assessment of emergency situations and is not a substitute for analysis of actual data. As with any atmospheric model, large errors are possible

Dose estimation models for environmental tritium released from fusion facilities

International Nuclear Information System (INIS)

Murata, Mikio

1993-01-01

Various mathematical models are being developed to predict the behavior of HT released to the natural environment and their consequent impact. This report outlines models and the major findings of HT field release studies in France and Canada. The models are constructed to incorporate the key processes thought to be responsible for the formation of atmospheric HTO from a release of HT. It has been established from the experiments that HT oxidized in surface soil is incorporated almost entirely into soil water as HTO. This tritium may be reemitted to the atmosphere in the form of HTO through exchange of soil and atmospheric moisture as well as through the bulk water mass flux from the soil the atmosphere due to evaporation and transpiration. The direct conversion of HT to HTO in air and direct uptake of HT by vegetation are expected to be negligible for the time and space scales of interest in considering short duration releases. HTO emitted to the atmosphere is can further exchange with soil and vegetation water. Validation of these models against experimental data is conducted to demonstrate their credibility. It may be concluded that further laboratory and field works are needed in order to develop a sufficiently good understanding of the dependence of the key processes on environmental factors (including diurnal cycling and seasonality) to allow the rates of the processes to be predicted from a knowledge of environmental conditions. (author)

batman: BAsic Transit Model cAlculatioN in Python

Science.gov (United States)

Kreidberg, Laura

2015-11-01

I introduce batman, a Python package for modeling exoplanet transit light curves. The batman package supports calculation of light curves for any radially symmetric stellar limb darkening law, using a new integration algorithm for models that cannot be quickly calculated analytically. The code uses C extension modules to speed up model calculation and is parallelized with OpenMP. For a typical light curve with 100 data points in transit, batman can calculate one million quadratic limb-darkened models in 30 seconds with a single 1.7 GHz Intel Core i5 processor. The same calculation takes seven minutes using the four-parameter nonlinear limb darkening model (computed to 1 ppm accuracy). Maximum truncation error for integrated models is an input parameter that can be set as low as 0.001 ppm, ensuring that the community is prepared for the precise transit light curves we anticipate measuring with upcoming facilities. The batman package is open source and publicly available at https://github.com/lkreidberg/batman .

Hindered disulfide bonds to regulate release rate of model drug from mesoporous silica.

Science.gov (United States)

Nadrah, Peter; Maver, Uroš; Jemec, Anita; Tišler, Tatjana; Bele, Marjan; Dražić, Goran; Benčina, Mojca; Pintar, Albin; Planinšek, Odon; Gaberšček, Miran

2013-05-01

With the advancement of drug delivery systems based on mesoporous silica nanoparticles (MSNs), a simple and efficient method regulating the drug release kinetics is needed. We developed redox-responsive release systems with three levels of hindrance around the disulfide bond. A model drug (rhodamine B dye) was loaded into MSNs' mesoporous voids. The pore opening was capped with β-cyclodextrin in order to prevent leakage of drug. Indeed, in absence of a reducing agent the systems exhibited little leakage, while the addition of dithiothreitol cleaved the disulfide bonds and enabled the release of cargo. The release rate and the amount of released dye were tuned by the level of hindrance around disulfide bonds, with the increased hindrance causing a decrease in the release rate as well as in the amount of released drug. Thus, we demonstrated the ability of the present mesoporous systems to intrinsically control the release rate and the amount of the released cargo by only minor structural variations. Furthermore, an in vivo experiment on zebrafish confirmed that the present model delivery system is nonteratogenic.

Sodium spray release accident analysis for fast reactor safety studies

International Nuclear Information System (INIS)

Shire, P.R.

1976-01-01

A computer code has been developed to model the effects of postulated sodium spray release from LMFBR piping, although this is an event of extremely low probability. The calculation method utilizes gas convection and droplet combustion theory to calculate the pressure and temperature effects in a range of 0 to 21 mole percent oxygen with humidity. Droplet motion and large aggregate sodium surface area result in rapid release of combustion and sensible heat causing a nearly adiabatic pressure rise which peaks in several seconds. This analytical tool has indicated reasonable agreement with prototypic test data for a range of oxygen and water vapor concentrations, cell volumes and droplet sizes

Atmospheric dispersion from releases in the vicinity of buildings

International Nuclear Information System (INIS)

Walsh, C.; Jones, J.A.

2002-01-01

The objective of this study is to advise FSA on the extent to which its current models for calculating air concentration and deposition for continuous releases close to sites with many buildings are adequate, whether there are circumstances for which the explicit modelling of building wake effects is required, and, if so, to recommend an appropriate model for this. The study considered the predictions of simple Gaussian models and the ADMS model. Results from the models are presented and compared, for a range of on-site building configurations and release locations. In addition, the extent to which details of the buildings on the site are required in ADMS is considered. The results indicate that buildings only affect the predicted concentration in a relatively small area around the site (less than 1 km from the site even for tall buildings). For dose calculations beyond 1 km, no allowance is required for modelling building effects. The results suggest that modelling the effects of buildings can be sensitive to a number of parameters and care should be used in interpreting results for locations within the region affected by buildings. However, because ADMS explicitly treats these factors, it is considered a better model for use than those based on a simple Gaussian approach. (author)

Flavor release measurement by atmospheric pressure chemical ionization ion trap mass spectrometry, construction of interface and mathematical modeling of release profiles

DEFF Research Database (Denmark)

Haahr, Anne-Mette; Madsen, Henrik; Smedsgaard, Jørn

2003-01-01

and the method can be used to measure breath from the nose. A mathematical model of the data was developed to give a quantitative method for description and characterization of the release of flavor compounds. The release profiles consisted of two sequences, one for a chewing period, and one for a phasing out...... process. The proposed method for modeling provided a reasonable description of the release process. In addition to flavor compounds, this new interface and mathematical application could provide information on chemicals in the human breath which could be interesting, for example, within medical diagnosis....... with that of the flavor detection threshold. An application study on the release of menthone and menthol from chewing gum by a group of six test persons was performed. Flavored chewing gum was used as a model matrix because of the long chewing periods and the simplicity of the system. It is concluded that the interface...

Extended latanoprost release from commercial contact lenses: in vitro studies using corneal models.

Directory of Open Access Journals (Sweden)

Saman Mohammadi

Full Text Available In this study, we compared, for the first time, the release of a 432 kDa prostaglandin F2a analogue drug, Latanoprost, from commercially available contact lenses using in vitro models with corneal epithelial cells. Conventional polyHEMA-based and silicone hydrogel soft contact lenses were soaked in drug solution (131 μg = ml solution in phosphate buffered saline. The drug release from the contact lens material and its diffusion through three in vitro models was studied. The three in vitro models consisted of a polyethylene terephthalate (PET membrane without corneal epithelial cells, a PET membrane with a monolayer of human corneal epithelial cells (HCEC, and a PET membrane with stratified HCEC. In the cell-based in vitro corneal epithelium models, a zero order release was obtained with the silicone hydrogel materials (linear for the duration of the experiment whereby, after 48 hours, between 4 to 6 μg of latanoprost (an amount well within the range of the prescribed daily dose for glaucoma patients was released. In the absence of cells, a significantly lower amount of drug, between 0.3 to 0.5 μg, was released, (p <0:001. The difference observed in release from the hydrogel lens materials in the presence and absence of cells emphasizes the importance of using an in vitro corneal model that is more representative of the physiological conditions in the eye to more adequately characterize ophthalmic drug delivery materials. Our results demonstrate how in vitro models with corneal epithelial cells may allow better prediction of in vivo release. It also highlights the potential of drug-soaked silicone hydrogel contact lens materials for drug delivery purposes.

Initial assessment of MHTGR confinement releases

International Nuclear Information System (INIS)

Maneke, J.L.; Lanning, D.D.; Lidsky, L.M.

1986-01-01

Initial investigation of Modular High Temperature Gas-Cooled Reactor (MHTGR) designs suggest that source terms during postulated accidents will be considerably lower than Light Water Reactor (LWR) source term estimates. These lower postulated accident releases are not only a safety incentive, but also an economic incentive for the development of this reactor type. For example, it is hoped that a filtered confinement building, rather than a more expensive LWR-like containment building, would adequately protect the public from radiological releases. The ability of a confinement building to satisfy safety requirements for the MHTGR depends on several reactor parameters, such as fuel quality, reactor design, and the design of the reactor building. SCIMCA, a Simple Code for Initial MHTGR Confinement Assessment has been developed for preliminary MHTGR building requirement calculations. The code is capable of modeling a decay chain with a maximum of five regions. Phenomena such as fission product decay and buildup, natural deposition, building filtration, and intercompartmental transport are incorporated. SCIMCA models reduction mechanisms, such as dispersion and decay, occurring as radioactivity is transported through the environment. A subroutine for calculating doses at specified distances has also been included

Center for Integrated Nanotechnologies (CINT) Chemical Release Modeling Evaluation

Energy Technology Data Exchange (ETDEWEB)

Stirrup, Timothy Scott [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-12-20

This evaluation documents the methodology and results of chemical release modeling for operations at Building 518, Center for Integrated Nanotechnologies (CINT) Core Facility. This evaluation is intended to supplement an update to the CINT [Standalone] Hazards Analysis (SHA). This evaluation also updates the original [Design] Hazards Analysis (DHA) completed in 2003 during the design and construction of the facility; since the original DHA, additional toxic materials have been evaluated and modeled to confirm the continued low hazard classification of the CINT facility and operations. This evaluation addresses the potential catastrophic release of the current inventory of toxic chemicals at Building 518 based on a standard query in the Chemical Information System (CIS).

Calculational model and code for corrosion products transfer in sodium systems

International Nuclear Information System (INIS)

Iizawa, Katsuyuki; Kikuchi, Takao; Nihei, Isao; Horie, Junnosuke

1987-01-01

The parameters to predict LMFBR radioactive CP behaviour are obtained by fitting the model to the release rates and distribution of deposits in experimental sodium loops. Analysis of the downstream effect for Co-60 deposition confirms that the mechanism controlling Co-60 transfer is diffusion through the boundary sublayer in the flowing sodium. On the other hand, analysis of Mn-54 deposition suggests that the controlling mechanism for this nuclide is probably an intermediate case between diffusion-control in the liquid and reaction-control at steel surfaces. The super-stoichiometric release of Mn-54 and the sub-stoichiometric release of Co-60 are interpreted from their model parameters as being likely related to their solubilities in sodium. The computer code developed for this model is named PSYCHE. (orig./HP)

Models and methods to evaluate consequences of the release of airborne radioactivity from NNPs

International Nuclear Information System (INIS)

Anh, T.H.

1989-01-01

To examine the radioactive contamination and possible consequences of a nuclear power plant on living organisms during its operation periodes, the computer programmes were elaborated for assessing its fluences on the environment. The authors have resolved the following problems: i) Calculation of fission product inventories in the reactor core; ii) Calculation of the atmospheric dispersion of the released radionuclides under the meteorological conditions as well as the deposition of the radioactive substances on the soil; iii) Calculation of the irradiation doses

Background and derivation of ANS-5.4 standard fission product release model. Technical report

International Nuclear Information System (INIS)

1982-01-01

ANS Working Group 5.4 was established in 1974 to examine fission product releases from UO2 fuel. The scope of ANS-5.4 was narrowly defined to include the following: (1) Review available experimental data on release of volatile fission products from UO2 and mixed-oxide fuel; (2) Survey existing analytical models currently being applied to lightwater reactors; and (3) Develop a standard analytical model for volatile fission product release to the fuel rod void space. Place emphasis on obtaining a model for radioactive fission product releases to be used in assessing radiological consequences of postulated accidents

Diffusive spatio-temporal noise in a first-passage time model for intracellular calcium release

KAUST Repository

Flegg, Mark B.

2013-01-01

The intracellular release of calcium from the endoplasmic reticulum is controlled by ion channels. The resulting calcium signals exhibit a rich spatio-temporal signature, which originates at least partly from microscopic fluctuations. While stochasticity in the gating transition of ion channels has been incorporated into many models, the distribution of calcium is usually described by deterministic reaction-diffusion equations. Here we test the validity of the latter modeling approach by using two different models to calculate the frequency of localized calcium signals (calcium puffs) from clustered IP3 receptor channels. The complexity of the full calcium system is here limited to the basic opening mechanism of the ion channels and, in the mathematical reduction simplifies to the calculation of a first passage time. Two models are then studied: (i) a hybrid model, where channel gating is treated stochastically, while calcium concentration is deterministic and (ii) a fully stochastic model with noisy channel gating and Brownian calcium ion motion. The second model utilises the recently developed two-regime method [M. B. Flegg, S. J. Chapman, and R. Erban, "The two-regime method for optimizing stochastic reaction-diffusion simulations," J. R. Soc., Interface 9, 859-868 (2012)] in order to simulate a large domain with precision required only near the Ca2+ absorbing channels. The expected time for a first channel opening that results in a calcium puff event is calculated. It is found that for a large diffusion constant, predictions of the interpuff time are significantly overestimated using the model (i) with a deterministic non-spatial calcium variable. It is thus demonstrated that the presence of diffusive noise in local concentrations of intracellular Ca2+ ions can substantially influence the occurrence of calcium signals. The presented approach and results may also be relevant for other cell-physiological first-passage time problems with small ligand concentration

Release of Fluconazole from Contact Lenses Using a Novel In Vitro Eye Model.

Science.gov (United States)

Phan, Chau-Minh; Bajgrowicz, Magdalena; Gao, Huayi; Subbaraman, Lakshman N; Jones, Lyndon W

2016-04-01

Rapid drug release followed by a plateau phase is a common observation with drug delivery from contact lenses (CLs) when evaluated in a vial. The aim of this study was to compare the release of fluconazole from seven commercially available daily disposable CLs using a conventional vial-based method with a novel in vitro eye model. An eye model was created using two 3-dimensional printed molds, which were filled with polydimethylsiloxane to obtain an inexpensive model that would mimic the eyeball and eyelid. The model was integrated with a microfluidic syringe pump, and the flow-through was collected in a 12-well microliter plate. Four commercial daily disposable conventional hydrogels (nelfilcon A, omafilcon A, etafilcon A, ocufilcon B) and three silicone hydrogels (somofilcon A, narafilcon A, delefilcon A) were evaluated. These CLs were incubated with fluconazole for 24 h. The drug release was measured in a vial containing 4.8 mL of phosphate-buffered saline and in the polydimethylsiloxane eye model with a 4.8-mL tear flow across 24 h. Overall, conventional hydrogel CLs had a higher uptake and release of fluconazole than silicone hydrogel CLs (p eye model (p eye model under low tear volume was sustained and did not reach a plateau across 24 h (p eyes with fungal keratitis may have increased tearing, which would significantly accelerate drug release.

Probabilistic consequence model of accidenal or intentional chemical releases.

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.-S.; Samsa, M. E.; Folga, S. M.; Hartmann, H. M.

2008-06-02

In this work, general methodologies for evaluating the impacts of large-scale toxic chemical releases are proposed. The potential numbers of injuries and fatalities, the numbers of hospital beds, and the geographical areas rendered unusable during and some time after the occurrence and passage of a toxic plume are estimated on a probabilistic basis. To arrive at these estimates, historical accidental release data, maximum stored volumes, and meteorological data were used as inputs into the SLAB accidental chemical release model. Toxic gas footprints from the model were overlaid onto detailed population and hospital distribution data for a given region to estimate potential impacts. Output results are in the form of a generic statistical distribution of injuries and fatalities associated with specific toxic chemicals and regions of the United States. In addition, indoor hazards were estimated, so the model can provide contingency plans for either shelter-in-place or evacuation when an accident occurs. The stochastic distributions of injuries and fatalities are being used in a U.S. Department of Homeland Security-sponsored decision support system as source terms for a Monte Carlo simulation that evaluates potential measures for mitigating terrorist threats. This information can also be used to support the formulation of evacuation plans and to estimate damage and cleanup costs.

Pathways to man for radionuclides released from disposal sites on land

International Nuclear Information System (INIS)

Hill, M.D.

1986-01-01

To predict the potential radiological impact on man of the disposal of radioactive wastes it is necessary to identify all the events and processes that could cause releases of radionuclides into the environment, to estimate their probabilities of occurrence and to calculate their consequences, for both individuals and populations. This paper briefly reviews the types of releases that have to be considered for land disposal sites and describes the mathematical models used to calculate rates of transport of radionuclides through the environment and doses to man. The difficulties involved in predicting environmental conditions in the far future are discussed, in the light of the ways in which the results of consequence calculations will be used. Assessments of land disposal of long-lived and highly radioactive wastes are briefly reviewed, with the aim of identifying the most important radionuclides and exposure pathways, and the areas where the models and their databases require improvement. (author)

Mechanistic modelling of drug release from polymer-coated and swelling and dissolving polymer matrix systems.

Science.gov (United States)

Kaunisto, Erik; Marucci, Mariagrazia; Borgquist, Per; Axelsson, Anders

2011-10-10

The time required for the design of a new delivery device can be sensibly reduced if the release mechanism is understood and an appropriate mathematical model is used to characterize the system. Once all the model parameters are obtained, in silico experiments can be performed, to provide estimates of the release from devices with different geometries and compositions. In this review coated and matrix systems are considered. For coated formulations, models describing the diffusional drug release, the osmotic pumping drug release, and the lag phase of pellets undergoing cracking in the coating due to the build-up of a hydrostatic pressure are reviewed. For matrix systems, models describing pure polymer dissolution, diffusion in the polymer and drug release from swelling and eroding polymer matrix formulations are reviewed. Importantly, the experiments used to characterize the processes occurring during the release and to validate the models are presented and discussed. Copyright © 2011 Elsevier B.V. All rights reserved.

Establishing credibility in the environmental models used for safety and licensing calculations in the nuclear industry

International Nuclear Information System (INIS)

Davis, P.A.

1997-01-01

Models that simulate the transport and behaviour of radionuclides in the environment are used extensively in the nuclear industry for safety and licensing purposes. They are needed to calculate derived release limits for new and operating facilities, to estimate consequences following hypothetical accidents and to help manage a real emergency. But predictions generated for these purposes are essentially meaningless unless they are accompanied by a quantitative estimate of the confidence that can be placed in them. For example, in an emergency where there has been an accidental release of radioactivity to the atmosphere, decisions based on a validated model with small uncertainties would likely be very different from those based on an untested model, or on one with large uncertainties. This paper begins with a discussion of some general methods for establishing the credibility of model predictions. The focus will be on environmental transport models but the principles apply to models of all kinds. Establishing the credibility of a model is not a trivial task, It involves a number of tasks including face validation, verification, experimental validation and sensitivity and uncertainty analyses. The remainder of the paper will present quantitative results relating to the credibility of environmental transport models. Model formation, choice of parameter values and the influence of the user will all be discussed as sources of uncertainty in predictions. The magnitude of uncertainties that must be expected in various applications of the models will be presented. The examples used throughout the paper are drawn largely from recent work carried out in BIOMOVS and VAMP. (DM)

Analysis of fuel centre temperatures and fission gas release data from the IFPE Database

International Nuclear Information System (INIS)

Schubert, A.; Lassmann, K.; Van Uffelen, P.; Van de Laar, J.; Elenkov, D.; Asenov, S.; Boneva, S.; Djourelov, N.; Georgieva, M.

2003-01-01

The present work has continued the analysis of fuel centre temperatures and fission gas release, calculated with standard options of the TRANSURANUS code. The calculations are compared to experimental data from the International Fuel Performance Experiments (IFPE) database. It is reported an analysis regarding UO 2 fuel for Western-type reactors: Fuel centre temperatures measured in the experiments Contact 1 and Contact 2 (in-pile tests of 2 rods performed at the Siloe reactor in Grenoble, France, closely simulating commercial PWR conditions); Fission gas release data derived from post-irradiation examinations of 9 fuel rods belonging to the High-Burnup Effects Programme, task 3 (HBEP3). The results allow for a comparison of predictions by TRANSURANUS for the mentioned Western-type fuels with those done previously for Russian-type WWER fuel. The comparison has been extended to include fuel centre temperatures as well as fission gas release. The present version of TRANSURANUS includes a model that calculates the production of Helium. The amount of produced Helium is compared to the measured and to the calculated release of the fission gases Xenon and Krypton

Performance assessment calculational exercises

International Nuclear Information System (INIS)

Barnard, R.W.; Dockery, H.A.

1990-01-01

The Performance Assessment Calculational Exercises (PACE) are an ongoing effort coordinated by Yucca Mountain Project Office. The objectives of fiscal year 1990 work, termed PACE-90, as outlined in the Department of Energy Performance Assessment (PA) Implementation Plan were to develop PA capabilities among Yucca Mountain Project (YMP) participants by calculating performance of a Yucca Mountain (YM) repository under ''expected'' and also ''disturbed'' conditions, to identify critical elements and processes necessary to assess the performance of YM, and to perform sensitivity studies on key parameters. It was expected that the PACE problems would aid in development of conceptual models and eventual evaluation of site data. The PACE-90 participants calculated transport of a selected set of radionuclides through a portion of Yucca Mountain for a period of 100,000 years. Results include analyses of fluid-flow profiles, development of a source term for radionuclide release, and simulations of contaminant transport in the fluid-flow field. Later work included development of a problem definition for perturbations to the originally modeled conditions and for some parametric sensitivity studies. 3 refs

UNIDOSE - a computer program for the calculation of individual and collective doses from airborne radioactive pollutants

International Nuclear Information System (INIS)

Karlberg, O.; Schwartz, H.; Forssen, B.-H.; Marklund, J.-E.

1979-01-01

UNIDOSE is a program system for calculating the consequences of a radioactive release to the atmosphere. The program is applicable for computation of dispersion in a rnage of 0 - 50 km from the release point. The Gaussion plume model is used for calculating the external dose from activity in the atmosphere, on the ground and the internal dose via inhalation. Radioactive decay, as well as growth and decay of daughter products are accounted for. The influence of dry deposition and wash-out are also considered. It is possible to treat time-dependent release-rates of 1 - 24 hours duration and constant release-rates for up to one year. The program system also contains routines for the calculation of collective dose and health effects. The system operates in a statistical manner. Many weather-situations, based on measured data, can be analysed and statistical properties, such as cumulative frequences, can be calculated. (author)

Numerical simulation of industrial and accidental release formation and transport

Energy Technology Data Exchange (ETDEWEB)

Piskunov, V.N.; Aloyan, A.A.; Gerasimov, V.M.; Pinaev, V.S.; Golubev, A.I.; Yanilkin, Yu.V.; Ivanov, N.V.; Nikonov, S.N.; Kharchenko, A.I. [and others

1995-05-01

Statement of work for contract 006 {open_quotes}Mathematical simulation of industrial and accidental release formation and transport{close_quotes} implies that the final result of the activity within this task will be VNIIEF developed techniques which will provide for the prediction of the post-accidental environment. Report [1] presents the description of physical models and calculation techniques which were chosen by VNIIEF to accomplish this task. These techniques were analysed for their capabilities, the areas of their application were defined, modifications within contract 006 were described, the results of test and methodical calculations were presented. Moreover, the experimental data were analysed over the source parameters and contamination measurements which can be used in the comparison with the calculation results. Based an these data this report compares the calculation results obtained with VNIIEF calculation techniques with the LANL-presented experimental results. The calculations which statements and results are given in section 1, included the following processes: explosion cloud ascent dynamics and jet release origin; aerosols kinetics in the release source including composite particle origin in the explosion cloud caused by radioactive substance sorption an the soil particles; contaminant transport in atmosphere and their in-site fallout due to the accidental explosions and fires; atmospheric flow dynamics and industrial contamination transfer over the complicated terrain. The calculated results were compared with the experimental data. Section 2 presents the parameters for a typical source in the explosion accidents based an the experimental results and calculated data from Section 1, as well as contamination patterns were calculated with basic technique {open_quotes}Prognosis{close_quotes}.

Precipitates/Salts Model Calculations for Various Drift Temperature Environments

International Nuclear Information System (INIS)

Marnier, P.

2001-01-01

The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation within a repository drift. This work is developed and documented using procedure AP-3.12Q, Calculations, in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The primary objective of this calculation is to predict the effects of evaporation on the abstracted water compositions established in ''EBS Incoming Water and Gas Composition Abstraction Calculations for Different Drift Temperature Environments'' (BSC 2001c). A secondary objective is to predict evaporation effects on observed Yucca Mountain waters for subsequent cement interaction calculations (BSC 2001d). The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b)

Mathematical models for calculating radiation dose to the fetus

International Nuclear Information System (INIS)

Watson, E.E.

1992-01-01

Estimates of radiation dose from radionuclides inside the body are calculated on the basis of energy deposition in mathematical models representing the organs and tissues of the human body. Complex models may be used with radiation transport codes to calculate the fraction of emitted energy that is absorbed in a target tissue even at a distance from the source. Other models may be simple geometric shapes for which absorbed fractions of energy have already been calculated. Models of Reference Man, the 15-year-old (Reference Woman), the 10-year-old, the five-year-old, the one-year-old, and the newborn have been developed and used for calculating specific absorbed fractions (absorbed fractions of energy per unit mass) for several different photon energies and many different source-target combinations. The Reference woman model is adequate for calculating energy deposition in the uterus during the first few weeks of pregnancy. During the course of pregnancy, the embryo/fetus increases rapidly in size and thus requires several models for calculating absorbed fractions. In addition, the increases in size and changes in shape of the uterus and fetus result in the repositioning of the maternal organs and in different geometric relationships among the organs and the fetus. This is especially true of the excretory organs such as the urinary bladder and the various sections of the gastrointestinal tract. Several models have been developed for calculating absorbed fractions of energy in the fetus, including models of the uterus and fetus for each month of pregnancy and complete models of the pregnant woman at the end of each trimester. In this paper, the available models and the appropriate use of each will be discussed. (Author) 19 refs., 7 figs

Hualien forced vibration calculation with a finite element model

International Nuclear Information System (INIS)

Wang, F.; Gantenbein, F.; Nedelec, M.; Duretz, Ch.

1995-01-01

The forced vibration tests of the Hualien mock-up were useful to validate finite element models developed for soil-structure interaction. In this paper the two sets of tests with and without backfill were analysed. the methods used are based on finite element modeling for the soil. Two approaches were considered: calculation of soil impedance followed by the calculation of the transfer functions with a model taking into account the superstructure and the impedance; direct calculation of the soil-structure transfer functions, with the soil and the structure being represented in the same model by finite elements. Blind predictions and post-test calculations are presented and compared with the test results. (author). 4 refs., 8 figs., 2 tabs

Example of emergency response model evaluation of studies using the Mathew/Adpic models

International Nuclear Information System (INIS)

Dickerson, M.H.; Lange, R.

1986-04-01

This report summarizes model evaluation studies conducted for the MATHEW/ADPIC transport and diffusion models during the past ten years. These models support the US Department of Energy Atmospheric Release Advisory Capability, an emergency response service for atmospheric releases of nuclear material. Field studies involving tracer releases used in these studies cover a broad range of meteorology, terrain and tracer release heights, the three most important aspects of estimating air concentration values resulting from airborne releases of toxic material. Results of these studies show that these models can estimate air concentration values within a factor of 2 20% to 50% of the time and a factor of 5 40% to 80% of the time. As the meterology and terrain become more complex and the release height of the tracer is increased, the accuracy of the model calculations degrades. This band of uncertainty appears to correctly represent the capability of these models at this time. A method for estimating angular uncertainty in the model calculations is described and used to suggest alternative methods for evaluating emergency response models

Improving containment mass and energy releases for CONTEMPT-LT/028 TU with RELAP5/MOD3

International Nuclear Information System (INIS)

DaSilva, H.C.; Choe, W.G.

1996-01-01

In order to obtain boundary conditions for RELAP5/MOD3 best estimate (BE) large break (LB) loss-of-coolant accident (LOCA) calculations, it is necessary to utilize a separate containment analysis code CONTEMPT-LT/028 TU, which in turn accepts mass and energy releases from the RELAP5/MOD3 calculation. When these boundary conditions are obtained, they are observed to be significantly lower than those reported in FSAR containment analyses. This motivates the present study, where RELAP5/MOD3 mass and energy releases are generated using the same assumptions listed in the FSAR containment calculations. Then CONTEMPT-LT/028 TU pressures and temperatures calculated with both sets of mass and energy releases are compared. It is seen that those obtained with the RELAP5/MOD3 input are still significantly lower, indicating a level of conservatism in the FSAR mass and energy releases that is even above that explicitly listed and also incorporated into the RELAP5/MOD3 calculation. An important conclusion from this finding is that Environmental Qualification (EQ) issues requiring containment re-analyses are likely to be easily resolved if new mass and energy releases are calculated with state-of-the-art LOCA codes modeling the entire reactor coolant system, even when conservative assumptions are incorporated

CMS Partial Releases Model, Tools, and Applications. Online and Framework-Light Releases

CERN Document Server

Jones, Christopher D; Meschi, Emilio; Shahzad Muzaffar; Andreas Pfeiffer; Ratnikova, Natalia; Sexton-Kennedy, Elizabeth

2009-01-01

The CMS Software project CMSSW embraces more than a thousand packages organized in subsystems for analysis, event display, reconstruction, simulation, detector description, data formats, framework, utilities and tools. The release integration process is highly automated by using tools developed or adopted by CMS. Packaging in rpm format is a built-in step in the software build process. For several well-defined applications it is highly desirable to have only a subset of the CMSSW full package bundle. For example, High Level Trigger algorithms that run on the Online farm, and need to be rebuilt in a special way, require no simulation, event display, or analysis packages. Physics analysis applications in Root environment require only a few core libraries and the description of CMS specific data formats. We present a model of CMS Partial Releases, used for preparation of the customized CMS software builds, including description of the tools used, the implementation, and how we deal with technical challenges, suc...

Model calculating annual mean atmospheric dispersion factor for coastal site of nuclear power plant

Institute of Scientific and Technical Information of China (English)

无

2001-01-01

This paper describes an atmospheric dispersion field experiment performed on the coastal site of nuclear power plant in the east part of China during 1995 to 1996. The three-dimension joint frequency are obtained by hourly observation of wind and temperature on a 100m high tower; the frequency of the “event day of land and sea breezes” are given by observation of surface wind and land and sea breezes; the diffusion parameters are got from measurements of turbulent and wind tunnel simulation test.A new model calculating the annual mean atmospheric dispersion factor for coastal site of nuclear power plant is developed and established.This model considers not only the effect from mixing release and mixed layer but also the effect from the internal boundary layer and variation of diffusion parameters due to the distance from coast.The comparison between results obtained by the new model and current model shows that the ratio of annual mean atmospheric dispersion factor gained by the new model and the current one is about 2.0.

CRRIS, Health Risk Assessment from Atmospheric Releases of Radionuclides

International Nuclear Information System (INIS)

1999-01-01

1 - Description of program or function: CRRIS consists of eight fully integrated computer codes which calculate environmental transport of atmospheric releases of radionuclides and resulting doses and health risks to individuals or populations. Each code may be used alone for various assessment applications. Because of its modular structure, CRRIS allows assessments to be tailored to the user's needs. Radionuclides are handled by CRRIS either in terms of the released radionuclides or the exposure radionuclides which consist of both the released nuclides and decay products that build up during environmental transport. Atmospheric dispersion calculations are performed by the ANEMOS computer code for distances less than 100 km and the RETADD-II computer code regional-scale distances. Both codes estimate annual-average air concentrations and ground deposition rates by location. SUMIT will translate and scale multiple ANEMOS runs onto a master grid. TERRA reads radionuclide air concentrations and deposition rates to estimate concentrations of radionuclides in food and surface soil. Radiologic decay and ingrowth, soil leaching, and transport through the food chain are included in the calculations. MLSOIL computes an effective radionuclide ground-surface concentration to be used in computing external health effects. The five-layer model of radionuclide transport through soil in MLSOIL provides an alternative to the single-layer model used in TERRA. DFSOIL computes dose factors used in MLSOIL to compute doses from the five soil layers and from the ground surface. ANDROS reads environmental concentrations of radionuclides computed by the other CRRIS codes and produces tables of doses and risks to individuals or populations from atmospheric releases of radionuclides. 2 - Method of solution: SUMIT performs geometric interpolation. TERRA and MLSOIL are terrestrial transport compartment models. DFSOIL computes soil-layer-specific dose factors based on the point-kernel method

Flavor perception and aroma release from model dairy desserts.

Science.gov (United States)

Lethuaut, Laurent; Weel, Koen G C; Boelrijk, Alexandra E M; Brossard, Chantal D

2004-06-02

Six model dairy desserts, with three different textures and two sucrose levels, were equally flavored with a blend of four aroma compounds [ethyl pentanoate, amyl acetate, hexanal, and (E)-2-hexenal] and evaluated by a seven person panel in order to study whether the sensory perception of the flavor and the aroma release during eating varied with the textural characteristics or the sweetness intensity of the desserts. The sensory perception was recorded by the time intensity (TI) method, while the in vivo aroma release was simultaneously measured by the MS-nose. Considering the panel as a whole, averaged flavor intensity increased with sucrose level and varied with the texture of the desserts. Depending on the aroma compound, the averaged profile of in vivo aroma release varied, but for each aroma compound, averaged aroma release showed no difference with the sucrose level and little difference with the texture of the desserts. Perceptual sweetness-aroma interactions were the main factors influencing perception whatever the texture of the desserts.

PROGNOSYS AND ANALYSIS OF DOSIMETRIC SITUATION AFTER EMERGENCY ATMOSPHERIC RELEASE OF NPP IN UKRAINE

Directory of Open Access Journals (Sweden)

Iu. V. Bonchuk

2009-01-01

Full Text Available The system for the operative analysis of dosimetric situation after radiation emergencies at nuclear power plants in Ukraine (KADO is developed for decision making support. Emergency release in KADO is modelled as sequence of the discrete elements which are emitted after small time intervals that allows to take into account in calculations heterogeneity of distribution of pollution in a released cloud. Calculated predicted and avertable doses are the base information for justification of emergency and urgent countermeasures.

Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

International Nuclear Information System (INIS)

Hamstra, J.; Kevenaar, J.W.A.M.

1978-06-01

A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

Simplified analytical model to simulate radionuclide release from radioactive waste trenches; Modelo simplificado para simulacao da liberacao de radionuclideos de repositorios de rejeitos radioativos

Energy Technology Data Exchange (ETDEWEB)

Sa, Bernardete Lemes Vieira de

2001-07-01

In order to evaluate postclosure off-site doses from low-level radioactive waste disposal facilities, a computer code was developed to simulate the radionuclide released from waste form, transport through vadose zone and transport in the saturated zone. This paper describes the methodology used to model these process. The radionuclide released from the waste is calculated using a model based on first order kinetics and the transport through porous media was determined using semi-analytical solution of the mass transport equation, considering the limiting case of unidirectional convective transport with three-dimensional dispersion in an isotropic medium. The results obtained in this work were compared with other codes, showing good agreement. (author)

Photometric Calibration of the Barium Cloud Image in a Space Active Experiment: Determining the Release Efficiency

International Nuclear Information System (INIS)

Xie Liang-Hai; Li Lei; Wang Jing-Dong; Tao Ran; Cheng Bing-Jun; Zhang Yi-Teng

2014-01-01

The barium release experiment is an effective method to explore the near-earth environment and to study all kinds of space physics processes. The first space barium release experiment in China was successfully carried out by a sounding rocket on April 5, 2013. This work is devoted to calculating the release efficiency of the barium release by analyzing the optical image observed during the experiment. First, we present a method to calibrate the images grey value of barium cloud with the reference stars to obtain the radiant fluxes at different moments. Then the release efficiency is obtained by a curve fitting with the theoretical evolution model of barium cloud. The calculated result is basically consistent with the test value on ground

DIGA/NSL new calculational model in slab geometry

International Nuclear Information System (INIS)

Makai, M.; Gado, J.; Kereszturi, A.

1987-04-01

A new calculational model is presented based on a modified finite-difference algorithm, in which the coefficients are determined by means of the so-called gamma matrices. The DIGA program determines the gamma matrices and the NSL program realizes the modified finite difference model. Both programs assume slab cell geometry, DIGA assumes 2 energy groups and 3 diffusive regions. The DIGA/NSL programs serve to study the new calculational model. (author)

Model cross section calculations using LAHET

International Nuclear Information System (INIS)

Prael, R.E.

1992-01-01

The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented

Glass operational file. Operational models and integration calculations

International Nuclear Information System (INIS)

Ribet, I.

2004-01-01

This document presents the operational choices of dominating phenomena, hypotheses, equations and numerical data of the parameters used in the two operational models elaborated for the calculation of the glass source terms with respect to the waste packages considered: existing packages (R7T7, AVM and CEA glasses) and future ones (UOX2, UOX3, UMo, others). The overall operational choices are justified and demonstrated and a critical analysis of the approach is systematically proposed. The use of the operational model (OPM) V 0 → V r , realistic, conservative and robust, is recommended for glasses with a high thermal and radioactive load, which represent the main part of the vitrified wastes. The OPM V 0 S, much more overestimating but faster to parameterize, can be used for the long-term behaviour forecasting of glasses with low thermal and radioactive load, considering today's lack of knowledge for the parameterization of a V 0 → V r type OPM. Efficiency estimations have been made for R7T7 glasses (OPM V 0 → V r ) and AVM glasses (OPM V 0 S), which correspond to more than 99.9% of the vitrified waste packages activity. The very contrasted results obtained, illustrate the importance of the choice of operational models: in conditions representative of a geologic disposal, the estimation of R7T7-type package lifetime exceeds several hundred thousands years. Even if the estimated lifetime of AVM packages is much shorter (because of the overestimating character of the OPM V 0 S), the release potential radiotoxicity is of the same order as the one of R7T7 packages. (J.S.)

Real-time modeling of complex atmospheric releases in urban areas

International Nuclear Information System (INIS)

Baskett, R.L.; Ellis, J.S.; Sullivan, T.J.

1994-08-01

If a nuclear installation in or near an urban area has a venting, fire, or explosion, airborne radioactivity becomes the major concern. Dispersion models are the immediate tool for estimating the dose and contamination. Responses in urban areas depend on knowledge of the amount of the release, representative meteorological data, and the ability of the dispersion model to simulate the complex flows as modified by terrain or local wind conditions. A centralized dispersion modeling system can produce realistic assessments of radiological accidents anywhere in a country within several minutes if it is computer-automated. The system requires source-term, terrain, mapping and dose-factor databases, real-time meteorological data acquisition, three-dimensional atmospheric transport and dispersion models, and experienced staff. Experience with past responses in urban areas by the Atmospheric Release Advisory Capability (ARAC) program at Lawrence Livermore National Laboratory illustrate the challenges for three-dimensional dispersion models

Real-time modelling of complex atmospheric releases in urban areas

International Nuclear Information System (INIS)

Baskett, R.L.; Ellis, J.S.; Sullivan, T.J.

2000-01-01

If a nuclear installation in or near an urban area has a venting, fire, or explosion, airborne radioactivity becomes the major concern. Dispersion models are the immediate tool for estimating the dose and contamination. Responses in urban areas depend on knowledge of the amount of the release, representative meteorological data, and the ability of the dispersion model to simulate the complex flows as modified by terrain or local wind conditions. A centralised dispersion modelling system can produce realistic assessments of radiological accidents anywhere in a country within several minutes if it is computer-automated. The system requires source-term, terrain, mapping and dose-factor databases, real-time meteorological data acquisition, three-dimensional atmospheric transport and dispersion models, and experienced staff. Experience with past responses in urban areas by the Atmospheric Release Advisory Capability (ARAC) program at Lawrence Livermore National Laboratory illustrate the challenges for three-dimensional dispersion models. (author)

Design and evaluation of nicorandil extended-release tablet

Directory of Open Access Journals (Sweden)

Ju-Young Kim

2015-04-01

Full Text Available The aim of this study was to design and evaluate extended-release formulations of a model drug, nicorandil, in order to achieve the desired steady-state plasma concentration of drug in vivo. Simulation was employed to estimate optimum dissolution and absorption rate of nicorandil. The dissolution test was employed using pH 1.2, 4.0, 6.8 buffer solution, or water, to measure the in vitro release behaviors of nicorandil formulations. A single dose (15 mg of each formulation was orally administered to four beagle dogs under fasted conditions, and the pharmacokinetic parameters were calculated. The in vitro/in vivo relationship of the extended-release formulation was confirmed using in vitro dissolution profiles and plasma concentrations of drug in beagle dogs. Nicorandil was released completely within 30 min from the immediate-release tablets and released for 24 h from the extended-release tablets. The nicorandil plasma concentration could be modified by adjusting the drug release rate from the extended-release formulation. The release rate of nicorandil was the rate-limiting step in the overall absorption of drug from the extended-release formulations. These results highlight the potential of a nicorandil extended-release formulation in the treatment of angina pectoris.

Atmospheric dispersion and deposition of 131I released from the Hanford Site

International Nuclear Information System (INIS)

Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.; Stage, S.A.

1996-01-01

Approximately 2.6 x 10 4 TBq (700,000 Ci) of 131 I were released to the air from reactor fuel processing plants on the Hanford Site in southcentral Washington State from December 1944 through December 1949. The Hanford Environmental Dose Reconstruction Project developed a suite of codes to estimate the doses that might have resulted from these releases. The Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET) computer code is part of this suite. The RATCHET code implements a Lagrangian-trajectory, Gaussian-puff dispersion model that uses hourly meterological and release rate data to estimate daily time-integrated air concentrations and surface contamination for use, in dose estimates. In this model, iodine is treated as a mixture of three species (inorganic gases, organic gases, and particles). Model deposition parameters are functions of the mixture and meterological conditions. A resistance model is used to calculate dry deposition velocities. Equilibrium between concentrations in the precipitation and the air near the ground is assumed in calculating wet deposition of gases, and irreversible washout of the particles is assumed. RATCHET explicitly treats the uncertainties in model parameters and meteorological conditions. Uncertainties in 131 I release rates and partitioning among the nominal species are treated by varying model input. The results of 100 model runs for December 1944 through December 1949 indicate that monthly average air concentrations and deposition have uncertainties ranging from a factor of two near the center of the time-integrated plume to more than an order of magnitude near the edge. These results indicate that ∼10% of the 131 I released to the atmosphere decayed during transit in the study area, ∼56% was deposited within the study area, and the remaining 34% was transported out of the study area while still in the air

Radionuclides and isotopes release of spent fuel matrix. Conceptual and mathematical models of wastes behaviour

International Nuclear Information System (INIS)

Cera, E.; Merino, J.; Bruno, J.

2000-01-01

We have developed a conceptual and numerical model to calculate release of selected radionuclides from spent fuel under repository condition. This has been done in the framework of the Enresa 2000 performance assessment exercise. The model has been developed based on kinetic mass balance equations in order to study the evolution of the spent fuel water interface as a function of time. Several processes have been kinetically modelled: congruent dissolution, radioactive decay, ingrowth and water turnover in the gap. The precipitation/redissolution of secondary solid phases has been taken into account from a thermodynamic point of view. Both approaches have been coupled and the resulting equations solved for a number of radionuclides in both, a conservative and realistic approach. The results show three distinct groups of radionuclides based on their release behaviour: a first group is composed of radioisotopes of highly insoluble elements (e. g., Pu, Am, Pd) whose concentration in the gap is mainly controlled by their solubility and therefore their evolution is identical in both cases. Secondly, a set of radionuclides from soluble elements under these conditions (e. g., I, Cs, Ra) show concentrations kinetically controlled, decreasing with time following the congruent dissolution trend. Their release concentrations are one order of magnitude larger in the conservative case than in the realistic case. Finally, a third group has been identified (e. g., Se, Th, Cm) where a mixed behaviour takes place: initially their solubility limiting phases control their concentration in the gap but the situation reverts to a kinetic control as the chemical conditions change and the secondary precipitates become totally dissolved. The fluxes of the different radionuclides are also given as an assessment of the source term in the performance assessment. (Author)

Continuous operation of RODOS in case of long lasting releases

International Nuclear Information System (INIS)

Raskob, W.; Paesler-Sauer, J.; Rafat, M.

2003-01-01

users who want to explore the impact of long term countermeasures in more detail. The users has also the option to set the time (in days after the first release) at which they consider the release to be ended or all significant deposition to have occurred. This will be entered interactively via the user interface. This will enable the user to start the consideration of countermeasures after the majority of the release has occurred, as some releases might have a very long, low level tail, and it would be unrealistic to expect decision makers to wait until deposition had completely stopped in such circumstances. This approach also recognizes that deposition will not end everywhere on the same day. The demonstration will exemplify the continuous automatic operation of RODOS using a release scenarios lasting over several days. On-line meteorological data measured at Forschungszentrum Karlsruhe, assumed to be the release point, will be used as input for the diagnostic calculations performed automatically within a distance range of 160 km x 160 km every 10 minutes. Numerical weather forecast from the German Weather Service, updated every 12 hours, will be used for prognostic calculations repeated in the same calculation area every hour. In parallel, the European wide numerical weather forecasts calculated by the ALADIN model, run by the Austrian Weather service and updated every 12 hours, are used to calculate in interactive RODOS runs the longer distance contamination. All functions of RODOS and its broad spectrum of results will be presented by the demonstration team. In particular, the necessity of emergency actions and countermeasure and their consequences in terms of areas affected, radiation doses and resources needed can be tracked with the ongoing activity release. (author)

Kinetics of corrosion products release from nickel-base alloys corroding in primary water conditions. A new modeling of release

International Nuclear Information System (INIS)

Carrette, F.; Guinard, L.; Pieraggi, B.

2002-01-01

The radioactivity in the primary circuit arises mainly from the activation of corrosion products in the core of pressurised water reactors; corrosion products dissolve from the oxide scales developed on steam generator tubes of alloy 690. The controlling and modelling of this process require a detailed knowledge of the microstructure and chemical composition of oxide scales as well as the kinetics of their corrosion and dissolution. Alloy 690 was studied as tubes and sheets, with three various surface states (as-received, cold-worked, electropolished). Corrosion tests were performed at 325 C and 155 bar in primary water conditions (B/Li - 1000/2 ppm, [H 2 ] 30 cm 3 .kg -1 TPN, [O 2 ] < 5 ppb); test durations ranged between 24 and 2160 hours. Corrosion tests in the TITANE loop provided mainly corrosion and oxidation kinetics, and tests in the BOREAL loop yielded release kinetics. This study revealed asymptotic type kinetics. Characterisation of the oxide scales grown in representative conditions of the primary circuit was performed by several techniques (SEM, TEM, SIMS, XPS, GIXRD). These analyses revealed the essential role of the fine grained cold-worked scale present on as-received and cold-worked materials. This scale controls the corrosion and release phenomena. The kinetic study and the characterisation of the oxide scales contributed to the modelling of the corrosion/release process. A growth/dissolution model was proposed for corrosion product scales grown in non-saturated dynamic fluid. This model provided the temporal evolution of oxide scales and release kinetics for different species (Fe, Ni, Cr). The model was validated for several surface states and several alloys. (authors)

Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

Science.gov (United States)

Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

2014-06-01

Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information), and simulations from the oceanic dispersion model SEA-GEARN-FDM, both developed by the authors. A sophisticated deposition scheme, which deals with dry and fogwater depositions, cloud condensation nuclei (CCN) activation and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The fallout to the ocean surface calculated by WSPEEDI-II was used as input data for the SEA-GEARN-FDM calculations. Reverse and inverse source-term estimation methods based on coupling the simulations from both models was adopted using air dose rates and concentrations, and sea surface concentrations. The results revealed that the major releases of radionuclides due to FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, the morning of 13 March after the venting event at Unit 3, midnight of 14 March when the SRV (Safely Relief Valve) at Unit 2 was opened three times, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of

Precipitates/Salts Model Calculations for Various Drift Temperature Environments

Energy Technology Data Exchange (ETDEWEB)

P. Marnier

2001-12-20

The objective and scope of this calculation is to assist Performance Assessment Operations and the Engineered Barrier System (EBS) Department in modeling the geochemical effects of evaporation within a repository drift. This work is developed and documented using procedure AP-3.12Q, Calculations, in support of ''Technical Work Plan For Engineered Barrier System Department Modeling and Testing FY 02 Work Activities'' (BSC 2001a). The primary objective of this calculation is to predict the effects of evaporation on the abstracted water compositions established in ''EBS Incoming Water and Gas Composition Abstraction Calculations for Different Drift Temperature Environments'' (BSC 2001c). A secondary objective is to predict evaporation effects on observed Yucca Mountain waters for subsequent cement interaction calculations (BSC 2001d). The Precipitates/Salts model is documented in an Analysis/Model Report (AMR), ''In-Drift Precipitates/Salts Analysis'' (BSC 2001b).

Calculation of radiological consequences of a radioactive matter release into the biosphere from a final repository using the code BioTREND; Berechnung radiologischer Konsequenzen der Freisetzung radioaktiver Stoffe aus einem Endlager in die Biosphaere mit dem Programm BioTREND

Energy Technology Data Exchange (ETDEWEB)

Reiche, Tatiana; Becker, Dirk-Alexander

2014-09-15

The aim of the work was the development of a module for the code RepoTREND that allows the evaluation of radiation exposure in the biosphere. The selection of the characteristics relevant for the risk assessment is described. The program module bioTREND is based on the results of fission product release and dispersion calculations and a separate biosphere modeling. Exposure data (annual effective doses and organ doses) can be calculated for individuals and collectives. Optional is the calculation of radiotoxicity concentrations and radiotoxicity fluxes. Several recommendations for the improvement of the calculation module are included.

On the mixing model for calculating the temperature fields in nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Mikhin, V.I.; Zhukov, A.V.

1985-01-01

One of the alternatives of the mixing model applied for calculating temperature fields in nuclear reactor fuel assemblies,including the fuel assemblies with nonequilibrium energy-release in fuel element cross section, is consistently described. The equations for both constant and variable values of coolant density and heat capacity are obtained. The mixing model is based on a set of mass, heat and longitudinal momentum balance equations. This set is closed by the ratios connecting the unknown values for gaps between fuel elements with the averaged values for neighbouring channels. The ratios to close momentum and heat balance equations, explaining, in particular, the nonequivalent heat and mass, momentum and mass transfer coefficients, are suggested. The balance equations with variable coolant density and heat capacity are reduced to the form coinciding with those of the similar equations with constant values of these parameters. Application of one of the main ratios of the mixing model relating the coolant transverse overflow in the gaps between fuel elements to the averaged coolant rates (flow rates) in the neighbouring channels is mainly limited by the coolant stabilized flow in the fuel assemblies with regular symmetrical arrangement of elements. Mass transfer coefficients for these elements are experimentally determined. The ratio in the paper is also applicable for calculation of fuel assembly temperature fields with a small relative shift of elements

Analysis of hydrogen sulfide releases in heavy water production facilities

International Nuclear Information System (INIS)

Croitoru, Cornelia; Dumitrescu, Maria; Preda, Irina; Lazar, Roxana

1996-01-01

Safety analyses conducted at ICIS concern primarily the heavy water production installations. The quantitative risk assessment needs the frequency calculation of accident sequences and consequences. In heavy water plants which obtain primary isotopic concentration of water by H 2 O - H 2 S exchange, large amounts of hydrogen sulfide which is a toxic, inflammable and explosive gas, are circulated. The first stage in calculating the consequences consists in potential analysis of H 2 S release. This work presents a study of this types of releases for pilot installations of the heavy water production at ICIS (Plant 'G' at Rm. Valcea). The installations which contain and maneuver large quantities of H 2 S and the mathematical models for different types of releases are presented. The accidents analyzed are: catastrophic column, container, spy-hole failures or gas-duct rupture and wall cracks in the installation. The main results are given as tables while the time variations of the flow rate and quantities of H 2 O released by stack disposal are plotted

Modeling the release, spreading, and burning of LNG, LPG, and gasoline on water

International Nuclear Information System (INIS)

Johnson, David W.; Cornwell, John B.

2007-01-01

Current interest in the shipment of liquefied natural gas (LNG) has renewed the debate about the safety of shipping large volumes of flammable fuels. The size of a spreading pool following a release of LNG from an LNG tank ship has been the subject of numerous papers and studies dating back to the mid-1970s. Several papers have presented idealized views of how the LNG would be released and spread across a quiescent water surface. There is a considerable amount of publicly available material describing these idealized releases, but little discussion of how other flammable fuels would behave if released from similar sized ships. The purpose of this paper is to determine whether the models currently available from the United States Federal Energy Regulatory Commission (FERC) can be used to simulate the release, spreading, vaporization, and pool fire impacts for materials other than LNG, and if so, identify which material-specific parameters are required. The review of the basic equations and principles in FERC's LNG release, spreading, and burning models did not reveal a critical fault that would prevent their use in evaluating the consequences of other flammable fluid releases. With the correct physical data, the models can be used with the same level of confidence for materials such as LPG and gasoline as they are for LNG

Source Release Modeling for the Idaho National Engineering and Environmental Laboratory's Subsurface Disposal Area

International Nuclear Information System (INIS)

Becker, B.H.

2002-01-01

A source release model was developed to determine the release of contaminants into the shallow subsurface, as part of the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) evaluation at the Idaho National Engineering and Environmental Laboratory's (INEEL) Subsurface Disposal Area (SDA). The output of the source release model is used as input to the subsurface transport and biotic uptake models. The model allowed separating the waste into areas that match the actual disposal units. This allows quantitative evaluation of the relative contribution to the total risk and allows evaluation of selective remediation of the disposal units within the SDA

A statistical model for deriving probability distributions of contamination for accidental releases

International Nuclear Information System (INIS)

ApSimon, H.M.; Davison, A.C.

1986-01-01

Results generated from a detailed long-range transport model, MESOS, simulating dispersal of a large number of hypothetical releases of radionuclides in a variety of meteorological situations over Western Europe have been used to derive a simpler statistical model, MESOSTAT. This model may be used to generate probability distributions of different levels of contamination at a receptor point 100-1000 km or so from the source (for example, across a frontier in another country) without considering individual release and dispersal scenarios. The model is embodied in a series of equations involving parameters which are determined from such factors as distance between source and receptor, nuclide decay and deposition characteristics, release duration, and geostrophic windrose at the source. Suitable geostrophic windrose data have been derived for source locations covering Western Europe. Special attention has been paid to the relatively improbable extreme values of contamination at the top end of the distribution. The MESOSTAT model and its development are described, with illustrations of its use and comparison with the original more detailed modelling techniques. (author)

Stochastic Modeling of Radioactive Material Releases

International Nuclear Information System (INIS)

Andrus, Jason; Pope, Chad

2015-01-01

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculates the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was

Stochastic Modeling of Radioactive Material Releases

Energy Technology Data Exchange (ETDEWEB)

Andrus, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pope, Chad [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-09-01

Nonreactor nuclear facilities operated under the approval authority of the U.S. Department of Energy use unmitigated hazard evaluations to determine if potential radiological doses associated with design basis events challenge or exceed dose evaluation guidelines. Unmitigated design basis events that sufficiently challenge dose evaluation guidelines or exceed the guidelines for members of the public or workers, merit selection of safety structures, systems, or components or other controls to prevent or mitigate the hazard. Idaho State University, in collaboration with Idaho National Laboratory, has developed a portable and simple to use software application called SODA (Stochastic Objective Decision-Aide) that stochastically calculates the radiation dose associated with hypothetical radiological material release scenarios. Rather than producing a point estimate of the dose, SODA produces a dose distribution result to allow a deeper understanding of the dose potential. SODA allows users to select the distribution type and parameter values for all of the input variables used to perform the dose calculation. SODA then randomly samples each distribution input variable and calculates the overall resulting dose distribution. In cases where an input variable distribution is unknown, a traditional single point value can be used. SODA was developed using the MATLAB coding framework. The software application has a graphical user input. SODA can be installed on both Windows and Mac computers and does not require MATLAB to function. SODA provides improved risk understanding leading to better informed decision making associated with establishing nuclear facility material-at-risk limits and safety structure, system, or component selection. It is important to note that SODA does not replace or compete with codes such as MACCS or RSAC, rather it is viewed as an easy to use supplemental tool to help improve risk understanding and support better informed decisions. The work was

PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

Energy Technology Data Exchange (ETDEWEB)

Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

1980-03-01

A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach.

PABLM: a computer program to calculate accumulated radiation doses from radionuclides in the environment

International Nuclear Information System (INIS)

Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

1980-03-01

A computer program, PABLM, was written to facilitate the calculation of internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. This report contains details of mathematical models used and calculational procedures required to run the computer program. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides in the environment after the releases have ended. Radioactive decay is considered during the release of radionuclides, after they are deposited on the plants or ground, and during holdup of food after harvest. The radiation dose models consider several exposure pathways. Doses may be calculated for either a maximum-exposed individual or for a population group. The doses calculated are accumulated doses from continuous chronic exposure. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radiological Protection (ICRP) for body burdens and MPC's of each radionuclide. The radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption that the contaminated medium is large enough to be considered an infinite volume or plane relative to the range of the emitted radiations. The equations for calculations of the radiation dose from external exposure to shoreline sediments include a correction for the finite width of the contaminated beach

A revised calculational model for fission

Energy Technology Data Exchange (ETDEWEB)

Atchison, F

1998-09-01

A semi-empirical parametrization has been developed to calculate the fission contribution to evaporative de-excitation of nuclei with a very wide range of charge, mass and excitation-energy and also the nuclear states of the scission products. The calculational model reproduces measured values (cross-sections, mass distributions, etc.) for a wide range of fissioning systems: Nuclei from Ta to Cf, interactions involving nucleons up to medium energy and light ions. (author)

Application of propagation calculations in air quality control

International Nuclear Information System (INIS)

Kuelske, S.

1993-01-01

This paper describes the development of a methodics of propagation calculation since the foundation of the 'Landesanstalt fuer Immissionsschutz' in 1963, and its practical use in air quality control. In this context, it deals with methods for calculating stack heights and the Gaussian propagation model adopted by the Technical Code on Clean Air, the propagation of flue gas emitted via cooling towers, the propagation of accidentally released substances, odours and automobile emissions, and with flow and propagation calculation for the proximity zone of buildings as well as for meteorological fields with space and time variations and topographically varied terrain. (orig.) [de

Modelling the dispersion of radionuclides following short duration releases to rivers

International Nuclear Information System (INIS)

Smith, J.T.; Bowes, M.; Denison, F.H.

2003-01-01

This project develops a model for assessing short duration liquid discharges of radionuclides to rivers. The assessment of doses arising from discharges to rivers is normally carried out by considering annual average discharge rates. Actual authorised discharges, however, may occur unevenly during the year or relatively high short-term discharges could occur in the unlikely event of an incident. Short term radionuclide releases could potentially result in temporary increases in radionuclide activity concentrations in water and fish which are greater than those resulting from a continuous discharge. The purpose of this project is to develop a model to assess short term releases from these sites, and where possible develop generic methods of assessing short term releases. An advection-dispersion model was developed to predict the concentrations of radionuclides in the river environment, ie in river water, river bed sediment and in predatory fish. Uptake of radionuclides to fish was modelled by estimating rates of uptake of radionuclides via the aquatic food chain or across the gill, as appropriate. The model was used to predict the concentrations of the radionuclides in the river Thames and its tributaries as a result of short duration discharges into stretches of the Thames and River Colne. Model output is given as a series of graphs of activity concentration and time integrated activity concentration resulting from a 1 MBq discharge for the following release durations: 5 minutes, 1 h, 3 h, 12 h and 24 h. The five locations for which predictions are given were 100m, 300m, 1000m, 3000m and 10000m downstream. The river volumetric flow rate was shown to be the most important environmental variable determining activity concentrations in water, fish and sediments following a release. In general, the maximum and integrated activity concentrations in water and fish will be in inverse proportion to the river volumetric flow rate, for a given amount and duration of release

Microbial Communities Model Parameter Calculation for TSPA/SR

International Nuclear Information System (INIS)

D. Jolley

2001-01-01

This calculation has several purposes. First the calculation reduces the information contained in ''Committed Materials in Repository Drifts'' (BSC 2001a) to useable parameters required as input to MING V1.O (CRWMS M and O 1998, CSCI 30018 V1.O) for calculation of the effects of potential in-drift microbial communities as part of the microbial communities model. The calculation is intended to replace the parameters found in Attachment II of the current In-Drift Microbial Communities Model revision (CRWMS M and O 2000c) with the exception of Section 11-5.3. Second, this calculation provides the information necessary to supercede the following DTN: M09909SPAMING1.003 and replace it with a new qualified dataset (see Table 6.2-1). The purpose of this calculation is to create the revised qualified parameter input for MING that will allow ΔG (Gibbs Free Energy) to be corrected for long-term changes to the temperature of the near-field environment. Calculated herein are the quadratic or second order regression relationships that are used in the energy limiting calculations to potential growth of microbial communities in the in-drift geochemical environment. Third, the calculation performs an impact review of a new DTN: M00012MAJIONIS.000 that is intended to replace the currently cited DTN: GS9809083 12322.008 for water chemistry data used in the current ''In-Drift Microbial Communities Model'' revision (CRWMS M and O 2000c). Finally, the calculation updates the material lifetimes reported on Table 32 in section 6.5.2.3 of the ''In-Drift Microbial Communities'' AMR (CRWMS M and O 2000c) based on the inputs reported in BSC (2001a). Changes include adding new specified materials and updating old materials information that has changed

Improved method for calculation of population doses from nuclear complexes over large geographical areas

International Nuclear Information System (INIS)

Corley, J.P.; Baker, D.A.; Hill, E.R.; Wendell, L.L.

1977-09-01

To simplify the calculation of potential long-distance environmental impacts, an overall average population exposure coefficient (P.E.C.) for the entire contiguous United States was calculated for releases to the atmosphere from Hanford facilities. The method, requiring machine computation, combines Bureau of Census population data by census enumeration district and an annual average atmospheric dilution factor (anti chi/Q') derived from 12-hourly gridded wind analyses provided by the NOAA's National Meteorological Center. A variable-trajectory puff-advection model was used to calculate an hourly anti chi/Q' for each grid square, assuming uniform hourly releases; seasonal and annual averages were then calculated. For Hanford, using 1970 census data, a P.E.C. of 2 x 10 -3 man-seconds per cubic meter was calculated. The P.E.C. is useful for both radioactive and nonradioactive releases. To calculate population doses for the entire contiguous United States, the P.E.C. is multiplied by the annual average release rate and then by the dose factor (rem/yr per Ci/m 3 ) for each radionuclide, and the dose contribution in man-rem is summed for all radionuclides. For multiple pathways, the P.E.C. is still useful, provided that doses from a unit release can be obtained from a set of atmospheric dose factors. The methodology is applicable to any point source, any set of population data by map grid coordinates, and any geographical area covered by equivalent meteorological data

TTS-Polttopuu - cost calculation model for fuelwood

International Nuclear Information System (INIS)

Naett, H.; Ryynaenen, S.

1998-01-01

The TTS-Institutes's Forestry Department has developed a computer based costcalculation model, 'TTS-Polttopuu', for the calculation of unit costs and resource needs in the harvesting systems for wood chips and split firewood. The model enables to determine the productivity and device cost per operating hour by each working stage of the harvesting system. The calculation model also enables the user to find out how changes in the productivity and cost bases of different harvesting chains influence the unit cost of the whole system. The harvesting chain includes the cutting of delimbed and non-delimbed fuelwood, forest haulage, road transportation chipping and chopping of longwood at storage. This individually operating software was originally developed to serve research needs, but it also serves the needs of the forestry and agricultural education, training and extension as well as individual firewood producers. The system requirements for this cost calculation model are at least 486-level processor with the Windows 95/98 -operating system, 16 MB of memory (RAM) and 5 MB of available hard-disk. This development work was carried out in conjunction with the nation-wide BIOENERGY Research Programme. (orig.)

Release Mathematical Model of Active Agent from Packaging Material into Food

Directory of Open Access Journals (Sweden)

Xiuling Huang

2013-01-01

Full Text Available Active packaging is an innovative packaging technology by which active compounds are released from the package to enhance the quality and microbial safety for a wide range of foods. The problem of active ingredient release through the bilayer packaging food system is studied from a theoretical viewpoint. A release model is built to provide predictions of concentration and amount of active ingredient. The equations are built based on Fickian diffusion, and numerical solutions are obtained through finite difference. Different diffusion coefficients DP and DC of active ingredient in different packaging layers, partition coefficient kCP at the interface of outer layer and inner layer, partition coefficient kFC at the interface of inner layer and food, and mass transfer coefficient hm at the interface of inner layer and food are considered in the model. The effects of kCP, thicknesses of outer layer and inner layer, CP0, DP, DC, kFC, and hm on the release are discussed. Corresponding conclusions and analysis are given.

SIMULATION OF VINPOCETINE RELEASE PROCESS FROM MICROCAPSULES WITH HYDROPHOBIC SHELL

Directory of Open Access Journals (Sweden)

Yu. A. Polkovnikova

2017-01-01

Full Text Available Nowadays microcapsules are widely spread in different industries. Microcapsules with vitamins, etheric and fatty oils are included into different cosmetics (creams, gels, serums. Microencapsulated probiotics are used in foods and fodder additives in veterinary. An important field of application of microencapsulation in pharmacy is the combination in the total dosage of drugs that are incompatible when mixed in free form.The aim of work is a comparative analysis of thermodynamic characteristics of vinpocetine release from the melt of beeswax and cacao butter 3:2 into water, solution of hydrochloric acid 0.01 M and ethanol.Materials and methods. For simulation of the process of vinpocetine release from the melt into different environments models component models of the studied systems were built and their atom charges were calculated by quantum-chemical method. Spatial models of the components were built in Hyper Chem 8.01. As an initial state for the thermodynamic characteristics of the calculation of vinpocetine release from the melt, a conformation of «melt-vinpocetine» system was used after thermodynamic equilibration by molecular dynamics simulation in Bioeurica program for 5 ns. For isolated systems a vibrational analysis was performed with the use of unrestricted Hartree-Fock method in STO-3G basis set in Orca 4.0 program.Results and discussion. Vinpocetine release from the melt of beeswax and cacao butter 3:2 into water with different pH values and to ethanol depends on its solubility in these environments, and also on solubility of the melt.Conclusion. The performed study of vinpocetine release from the melt of beeswax and cacao butter 3:2 by molecular dynamics simulation demonstrates the opportunity of vinpocetine release into water with pH=2 and into ethanol. The obtained results make it possible to assume a lower degree of vinpocetine release from the melt into ethanol compared with the solution of hydrochloric acid 0,01 M.

Computer-aided and predictive models for design of controlled release of pesticides

DEFF Research Database (Denmark)

Suné, Nuria Muro; Gani, Rafiqul

2004-01-01

In the field of pesticide controlled release technology, a computer based model that can predict the delivery of the Active Ingredient (AI) from fabricated units is important for purposes of product design and marketing. A model for the release of an M from a microcapsule device is presented...... in this paper, together with a specific case study application to highlight its scope and significance. The paper also addresses the need for predictive models and proposes a computer aided modelling framework for achieving it through the development and introduction of reliable and predictive constitutive...... models. A group-contribution based model for one of the constitutive variables (AI solubility in polymers) is presented together with examples of application and validation....

SU-F-19A-08: Optimal Time Release Schedule of In-Situ Drug Release During Permanent Prostate Brachytherapy

International Nuclear Information System (INIS)

Cormack, R; Ngwa, W; Makrigiorgos, G; Tangutoori, S; Rajiv, K; Sridhar, S

2014-01-01

Purpose: Permanent prostate brachytherapy spacers can be used to deliver sustained doses of radiosentitizing drug directly to the target, in order to enhance the radiation effect. Implantable nanoplatforms for chemo-radiation therapy (INCeRTs) have a maximum drug capacity and can be engineered to control the drug release schedule. The optimal schedule for sensitization during continuous low dose rate irradiation is unknown. This work studies the optimal release schedule of drug for both traditional sensitizers, and those that work by suppressing DNA repair processes. Methods: Six brachytherapy treatment plans were used to model the anatomy, implant geometry and calculate the spatial distribution of radiation dose and drug concentrations for a range of drug diffusion parameters. Three state partial differential equations (cells healthy, damaged or dead) modeled the effect of continuous radiation (radiosensitivities α,β) and cellular repair (time tr) on a cell population. Radiosensitization was modeled as concentration dependent change in α,β or tr which with variable duration under the constraint of fixed total drug release. Average cell kill was used to measure effectiveness. Sensitization by means of both enhanced damage and reduced repair were studied. Results: Optimal release duration is dependent on the concentration of radiosensitizer compared to the saturation concentration (csat) above which additional sensitization does not occur. Long duration drug release when enhancing α or β maximizes cell death when drug concentrations are generally over csat. Short term release is optimal for concentrations below saturation. Sensitization by suppressing repair has a similar though less distinct trend that is more affected by the radiation dose distribution. Conclusion: Models of sustained local radiosensitization show potential to increase the effectiveness of radiation in permanent prostate brachytherapy. INCeRTs with high drug capacity produce the greatest

Modeling fission gas release in high burnup ThO2-UO2 fuel

International Nuclear Information System (INIS)

Long, Y.; Yuan, Y.; Pilat, E.E.; Rim, C.S.; Kazimi, M.S.

2001-01-01

A preliminary fission gas release model to predict the performance of thoria fuel using the FRAPCON-3 computer code package has been formulated. The following modeling changes have been made in the code: - Radial power/burnup distribution; - Thermal conductivity and thermal expansion; - Rim porosity and fuel density; - Diffusion coefficient of fission gas in ThO 2 -UO 2 fuel and low temperature fission gas release model. Due to its lower epithermal resonance absorption, thoria fuel experiences a much flatter distribution of radial fissile products and radial power distribution during operation as compared to uranian fuel. The rim effect and its consequences in thoria fuel, therefore, are expected to occur only at relatively high burnup levels. The enhanced conductivity is evident for ThO 2 , but for a mixture the thermal conductivity enhancement is small. The lower thermal fuel expansion tends to negate these small advantages. With the modifications above, the new version of FRAPCON-3 matched the measured fission gas release data reasonably well using the ANS 5.4 fission gas release model. (authors)

Pest persistence and eradication conditions in a deterministic model for sterile insect release.

Science.gov (United States)

Gordillo, Luis F

2015-01-01

The release of sterile insects is an environment friendly pest control method used in integrated pest management programmes. Difference or differential equations based on Knipling's model often provide satisfactory qualitative descriptions of pest populations subject to sterile release at relatively high densities with large mating encounter rates, but fail otherwise. In this paper, I derive and explore numerically deterministic population models that include sterile release together with scarce mating encounters in the particular case of species with long lifespan and multiple matings. The differential equations account separately the effects of mating failure due to sterile male release and the frequency of mating encounters. When insects spatial spread is incorporated through diffusion terms, computations reveal the possibility of steady pest persistence in finite size patches. In the presence of density dependence regulation, it is observed that sterile release might contribute to induce sudden suppression of the pest population.

Emergency Doses (ED) - Revision 3: A calculator code for environmental dose computations

International Nuclear Information System (INIS)

Rittmann, P.D.

1990-12-01

The calculator program ED (Emergency Doses) was developed from several HP-41CV calculator programs documented in the report Seven Health Physics Calculator Programs for the HP-41CV, RHO-HS-ST-5P (Rittman 1984). The program was developed to enable estimates of offsite impacts more rapidly and reliably than was possible with the software available for emergency response at that time. The ED - Revision 3, documented in this report, revises the inhalation dose model to match that of ICRP 30, and adds the simple estimates for air concentration downwind from a chemical release. In addition, the method for calculating the Pasquill dispersion parameters was revised to match the GENII code within the limitations of a hand-held calculator (e.g., plume rise and building wake effects are not included). The summary report generator for printed output, which had been present in the code from the original version, was eliminated in Revision 3 to make room for the dispersion model, the chemical release portion, and the methods of looping back to an input menu until there is no further no change. This program runs on the Hewlett-Packard programmable calculators known as the HP-41CV and the HP-41CX. The documentation for ED - Revision 3 includes a guide for users, sample problems, detailed verification tests and results, model descriptions, code description (with program listing), and independent peer review. This software is intended to be used by individuals with some training in the use of air transport models. There are some user inputs that require intelligent application of the model to the actual conditions of the accident. The results calculated using ED - Revision 3 are only correct to the extent allowed by the mathematical models. 9 refs., 36 tabs

Off-gas treatment carbon footprint calculator : form and function

Energy Technology Data Exchange (ETDEWEB)

Kessell, L. [Good EarthKeeping Organization Inc., Corona, CA (United States); Squire, J.; Crosby, K. [Haley and Aldrich Inc., Boston, MA (United States)

2008-07-01

Carbon footprinting is the measurement of the impact on the environment in terms of the amount of greenhouse gases produced, measured in units of carbon dioxide released directly and indirectly by an individual, organization, process, event or product. This presentation discussed an off-gas treatment carbon footprint calculator. The presentation provided a review of off-gas treatment technologies and presented a carbon footprint model. The model included: form and function; parameters; assumptions; calculations; and off-gas treatment applications. Parameters of the model included greenhouse gases listed in the Kyoto Protocol to the United Nations Framework Convention on Climate Change, such as carbon dioxide, methane, nitrous oxide, sulfur hexafluoride, hydrofluorocarbons, and perfluorocarbons. Assumptions of the model included stationary combustion emissions; mobile combustion emissions; indirect emissions; physical or chemical processing emissions; fugitive emissions; and de minimus emissions. The presentation also examined resource conservation and discussed three greenhouse gas footprint case studies. It was concluded that the model involved a calculator with standard calculations with clearly defined assumptions with boundaries. tabs., figs.

Mathematical modeling of drug release from lipid dosage forms.

Science.gov (United States)

Siepmann, J; Siepmann, F

2011-10-10

Lipid dosage forms provide an interesting potential for controlled drug delivery. In contrast to frequently used poly(ester) based devices for parenteral administration, they do not lead to acidification upon degradation and potential drug inactivation, especially in the case of protein drugs and other acid-labile active agents. The aim of this article is to give an overview on the current state of the art of mathematical modeling of drug release from this type of advanced drug delivery systems. Empirical and semi-empirical models are described as well as mechanistic theories, considering diffusional mass transport, potentially limited drug solubility and the leaching of other, water-soluble excipients into the surrounding bulk fluid. Various practical examples are given, including lipid microparticles, beads and implants, which can successfully be used to control the release of an incorporated drug during periods ranging from a few hours up to several years. The great benefit of mechanistic mathematical theories is the possibility to quantitatively predict the effects of different formulation parameters and device dimensions on the resulting drug release kinetics. Thus, in silico simulations can significantly speed up product optimization. This is particularly useful if long release periods (e.g., several months) are targeted, since experimental trial-and-error studies are highly time-consuming in these cases. In the future it would be highly desirable to combine mechanistic theories with the quantitative description of the drug fate in vivo, ideally including the pharmacodynamic efficacy of the treatments. Copyright © 2011 Elsevier B.V. All rights reserved.

Model for Microcapsule Drug Release with Ultrasound-Activated Enhancement.

Science.gov (United States)

Tsao, Nadia H; Hall, Elizabeth A H

2017-11-14

Microbubbles and microcapsules of silane-polycaprolactone (SiPCL) have been filled with a fluorescent acridium salt (lucigenin) as a model for a drug-loaded delivery vehicle. The uptake and delivery were studied and compared with similar microbubbles and microcapsules of silica/mercaptosilica (S/M/S). Positively charged lucigenin was encapsulated through an electrostatic mechanism, following a Type I Langmuir isotherm as expected, but with an additional multilayer uptake that leads to a much higher loading for the SiPCL system (∼280 μg/2.4 × 10 9 microcapsules compared with ∼135 μg/2.4 × 10 9 microcapsules for S/M/S). Whereas the lucigenin release from the S/M/S bubbles and capsules loaded below the solubility limit is consistent with diffusion from a monolithic structure, the SiPCL structures show distinct release patterns; the Weibull function predicts a general trend for diffusion from normal Euclidean space at short times tending toward diffusion out of fractal spaces with increasing time. As a slow release system, the dissolution time (T d ) increases from 1 to 2 days for the S/M/S and for the low concentration, loaded SiPCl vehicles to ∼10 days for the high loaded microcapsules. However, T d can be reduced on insonation to 2 days, indicating the potential to gain control over the local enhanced release with ultrasound. This was tested for a docetaxel model and its effect on C4-2B prostate cancer cells, showing improved cell toxicity for concentrations below the normal EC 50 in solution.

Improved perturbative calculations in field theory; Calculation of the mass spectrum and constraints on the supersymmetric standard model; Calculs perturbatifs variationnellement ameliores en theorie des champs; Calcul du spectre et contraintes sur le modele supersymetrique standard

Energy Technology Data Exchange (ETDEWEB)

Kneur, J.L

2006-06-15

This document is divided into 2 parts. The first part describes a particular re-summation technique of perturbative series that can give a non-perturbative results in some cases. We detail some applications in field theory and in condensed matter like the calculation of the effective temperature of Bose-Einstein condensates. The second part deals with the minimal supersymmetric standard model. We present an accurate calculation of the mass spectrum of supersymmetric particles, a calculation of the relic density of supersymmetric black matter, and the constraints that we can infer from models.

The models of internal dose calculation in ICRP

International Nuclear Information System (INIS)

Nakano, Takashi

1995-01-01

There are a lot discussions about internal dose calculation in ICRP. Many efforts are devoted to improvement in models and parameters. In this report, we discuss what kind of models and parameters are used in ICRP. Models are divided into two parts, the dosimetric model and biokinetic model. The former is a mathematical phantom model, and it is mainly developed in ORNL. The results are used in many researchers. The latter is a compartment model and it has a difficulty to decide the parameter values. They are not easy to estimate because of their age dependency. ICRP officially sets values at ages of 3 month, 1 year, 5 year, 10 year, 15 year and adult, and recommends to get values among ages by linear age interpolate. But it is very difficult to solve the basic equation with these values, so we calculate by use of computers. However, it has complex shame and needs long CPU time. We should make approximated equations. The parameter values include much uncertainty because of less experimental data, especially for a child. And these models and parameter values are for Caucasian. We should inquire whether they could correctly describe other than Caucasian. The body size affects the values of calculated SAF, and the differences of metabolism change the biokinetic pattern. (author)

Modeling Inhalational Tularemia: Deliberate Release and Public Health Response

Science.gov (United States)

Hall, Ian M.; Leach, Steve

2011-01-01

Two epidemic modeling studies of inhalational tularemia were identified in the published literature, both demonstrating the high number of potential casualties that could result from a deliberate aerosolized release of the causative agent in an urban setting. However, neither study analyzed the natural history of inhalational tularemia nor modeled the relative merits of different mitigation strategies. We first analyzed publicly available human/primate experimental data and reports of naturally acquired inhalational tularemia cases to better understand the epidemiology of the disease. We then simulated an aerosolized release of the causative agent, using airborne dispersion modeling to demonstrate the potential number of casualties and the extent of their spatial distribution. Finally, we developed a public health intervention model that compares 2 mitigation strategies: targeting antibiotics at symptomatic individuals with or without mass distribution of antibiotics to potentially infected individuals. An antibiotic stockpile that is sufficient to capture all areas where symptomatic individuals were infected is likely to save more lives than treating symptomatic individuals alone, providing antibiotics can be distributed rapidly and their uptake is high. However, with smaller stockpiles, a strategy of treating symptomatic individuals alone is likely to save many more lives than additional mass distribution of antibiotics to potentially infected individuals. The spatial distribution of symptomatic individuals is unlikely to coincide exactly with the path of the dispersion cloud if such individuals are infected near their work locations but then seek treatment close to their homes. The optimal mitigation strategy will depend critically on the size of the release relative to the stockpile level and the effectiveness of treatment relative to the speed at which antibiotics can be distributed. PMID:22044315

Testing the behaviour of different kinetic models for uptake/release of radionuclides between water and sediments when implemented in a marine dispersion model

International Nuclear Information System (INIS)

Perianez, R.

2004-01-01

Three kinetic models for adsorption/release of 137 Cs between water and sediments have been tested when they are included in a previously validated dispersion model of the English Channel. Radionuclides are released to the Channel from La Hague nuclear fuel reprocessing plant (France). The kinetic models are a 1-step model consisting of a single reversible reaction, a 2-step model consisting of two consecutive reversible reactions and an irreversible model consisting of three parallel reactions: two reversible and one irreversible. The models have been tested under three typical situations that correspond to the source terms that can generally be found: instantaneous release, continuous release and redissolution of radionuclides from contaminated sediments. Differences between the models become more evident when contact times between water and sediments are larger (continuous release) and in the case of redissolution from sediments. Time scales for the redissolution process are rather different between the three models. The 1-step model produces a redissolution that is too fast when compared with experimental evidence. The irreversible model requires that saturation effects of the irreversible phase are included. Probably, the 2-step model represents the best compromise between ease and level of detail of the description of sorption/release processes

NUMERICAL SIMULATION OF AIR POLLUTION IN CASE OF UNPLANNED AMMONIA RELEASE

Directory of Open Access Journals (Sweden)

L. V. Amelina

2017-06-01

Full Text Available Purpose. Development fast calculating model which takes into account the meteorological parameters and buildings which are situated near the source of toxic chemical emission. Methodology. The developed model is based on the equation for potential flow and equation of pollutant dispersion. Equation of potential flow is used to compute wind pattern among buildings. To solve equation for potential flow Samarskii implicit difference scheme is used. The implicit change – triangle difference scheme is used to solve equation of mass transfer. Numerical integration is carried out using the rectangular difference grid. Method of porosity technique («markers method» is used to create the form of comprehensive computational region. Emission of ammonia is modeled using Delta function for point source. Findings. Developed 2D numerical model belongs to the class of «diagnostic models». This model takes into account the main physical factors affecting the process of dispersion of pollutants in the atmosphere. The model takes into account the influence of buildings on pollutant dispersion. On the basis of the developed numerical models a computational experiment was carried out to estimate the level of toxic chemical pollution in the case of unplanned ammonia release at ammonia pump station. Originality. Developed numerical model allows to calculate the 2D wind pattern among buildings and pollutant dispersion in the case unplanned ammonia release. Model allows to perform fast calculations of the atmosphere pollution. Practical value. The model can be used when developing the PLAS (Emergency Response Plan.

SPEEDI: a computer code system for the real-time prediction of radiation dose to the public due to an accidental release

International Nuclear Information System (INIS)

Imai, Kazuhiko; Chino, Masamichi; Ishikawa, Hirohiko

1985-10-01

SPEEDI, a computer code system for prediction of environmental doses from radioactive materials accidentally released from a nuclear plant has been developed to assist the organizations responsible for an emergency planning. For realistic simulation, have been developed a model which statistically predicts the basic wind data and then calculates the three-dimensional mass consistent wind field by interpolating these predicted data, and a model for calculation of the diffusion of released materials using a combined model of random-walk and PICK methods. These calculation in the system is carried out in conversational mode with a computer so that we may use the system with ease in an emergency. SPEEDI has also versatile files, which make it easy to control the complicated flows of calculation. In order to attain a short computation time, a large-scale computer with performance of 25 MIPS and a vector processor of maximum 250 MFLOPS are used for calculation of the models so that quick responses have been made. Simplified models are also prepared for calculation in a minicomputer widely used by local governments and research institutes, although the precision of calculation as same with the above models can not be expected to obtain. The present report outlines the structure and functions of SPEEDI, methods for prediction of the wind field and the models for calculation of the concentration of released materials in air and on the ground, and the doses to the public. Some of the diffusion models have been compared with the field experiments which had been carried out as a part of the SPEEDI development program. The report also discusses the reliability of the diffusion models on the basis of the compared results, and shows that they can reasonably simulate the diffusion in the internal boundary layer which commonly occurs near the coastal region. (J.P.N.)

Implementation of decommissioning materials conditional clearance process to the OMEGA calculation code

International Nuclear Information System (INIS)

Zachar, Matej; Necas, Vladimir; Daniska, Vladimir

2011-01-01

The activities performed during nuclear installation decommissioning process inevitably lead to the production of large amount of radioactive material to be managed. Significant part of materials has such low radioactivity level that allows them to be released to the environment without any restriction for further use. On the other hand, for materials with radioactivity slightly above the defined unconditional clearance level, there is a possibility to release them conditionally for a specific purpose in accordance with developed scenario assuring that radiation exposure limits for population not to be exceeded. The procedure of managing such decommissioning materials, mentioned above, could lead to recycling and reuse of more solid materials and to save the radioactive waste repository volume. In the paper an a implementation of the process of conditional release to the OMEGA Code is analyzed in details; the Code is used for calculation of decommissioning parameters. The analytical approach in the material parameters assessment, firstly, assumes a definition of radiological limit conditions, based on the evaluation of possible scenarios for conditionally released materials, and their application to appropriate sorter type in existing material and radioactivity flow system. Other calculation procedures with relevant technological or economical parameters, mathematically describing e.g. final radiation monitoring or transport outside the locality, are applied to the OMEGA Code in the next step. Together with limits, new procedures creating independent material stream allow evaluation of conditional material release process during decommissioning. Model calculations evaluating various scenarios with different input parameters and considering conditional release of materials to the environment are performed to verify the implemented methodology. Output parameters and results of the model assessment are presented, discussed and conduced in the final part of the paper

Models for recurrent gas release event behavior in hazardous waste tanks

International Nuclear Information System (INIS)

Anderson, D.N.; Arnold, B.C.

1994-08-01

Certain radioactive waste storage tanks at the United States Department of Energy Hanford facilities continuously generate gases as a result of radiolysis and chemical reactions. The congealed sludge in these tanks traps the gases and causes the level of the waste within the tanks to rise. The waste level continues to rise until the sludge becomes buoyant and ''rolls over'', changing places with heavier fluid on top. During a rollover, the trapped gases are released, resulting, in a sudden drop in the waste level. This is known as a gas release event (GRE). After a GRE, the wastes leading to another GRE. We present nonlinear time waste re-congeals and gas again accumulates leading to another GRE. We present nonlinear time series models that produce simulated sample paths that closely resemble the temporal history of waste levels in these tanks. The models also imitate the random GRE, behavior observed in the temporal waste level history of a storage tank. We are interested in using the structure of these models to understand the probabilistic behavior of the random variable ''time between consecutive GRE's''. Understanding the stochastic nature of this random variable is important because the hydrogen and nitrous oxide gases released from a GRE, are flammable and the ammonia that is released is a health risk. From a safety perspective, activity around such waste tanks should be halted when a GRE is imminent. With credible GRE models, we can establish time windows in which waste tank research and maintenance activities can be safely performed

Multiscale numerical modeling of Ce3+-inhibitor release from novel corrosion protection coatings

International Nuclear Information System (INIS)

Trenado, Carlos; Wittmar, Matthias; Veith, Michael; Strauss, Daniel J; Rosero-Navarro, Nataly C; Aparicio, Mario; Durán, Alicia; Castro, Yolanda

2011-01-01

A novel hybrid sol–gel coating has recently been introduced as an alternative to high toxic chromate-based corrosion protection systems. In this paper, we propose a multiscale computational model to estimate the amount and time scale of inhibitor release of the active corrosion protection coating. Moreover, we study the release rate under the influence of parameters such as porosity and viscosity, which have recently been implicated in the stability of the coating. Numerical simulations obtained with the model predicted experimental release tests and recent findings on the compromise between inhibitor concentration and the stability of the coating

Application of the HGSYSTEM/UF6 model to simulate atmospheric dispersion of UF6 releases from uranium enrichment plants

International Nuclear Information System (INIS)

Goode, W.D. Jr.; Bloom, S.G.; Keith, K.D. Jr.

1995-01-01

Uranium hexafluoride is a dense, reactive gas used in Gaseous Diffusion Plants (GDPs) to make uranium enriched in the 235 U isotope. Large quantities of UF 6 exist at the GDPs in the form of in-process gas and as a solid in storage cylinders; smaller amounts exist as hot liquid during transfer operations. If liquid UF 6 is released to the environment, it immediately flashes to a solid and a dense gas that reacts rapidly with water vapor in the air to form solid particles of uranyl fluoride and hydrogen fluoride gas. Preliminary analyses were done on various accidental release scenarios to determine which scenarios must be considered in the safety analyses for the GDPS. These scenarios included gas releases due to failure of process equipment and liquid/gas releases resulting from a breach of transfer piping from a cylinder. A major goal of the calculations was to estimate the response time for mitigating actions in order to limit potential off-site consequences of these postulated releases. The HGSYSTEM/UF 6 code was used to assess the consequences of these release scenarios. Inputs were developed from release calculations which included two-phase, choked flow followed by expansion to atmospheric pressure. Adjustments were made to account for variable release rates and multiple release points. Superpositioning of outputs and adjustments for exposure time were required to evaluate consequences based on health effects due to exposures to uranium and HF at a specific location

PABLM, Doses from Radioactive Releases to Atmosphere and Food Chain

International Nuclear Information System (INIS)

Napier, B.A.; Kennedy, W.E. Jr.; Soldat, J.K.

1983-01-01

1 - Description of problem or function: PABLM calculates internal radiation doses to man from radionuclides in food products and external radiation doses from radionuclides in the environment. Radiation doses from radionuclides in the environment may be calculated from deposition on the soil or plants during an atmospheric or liquid release, or from exposure to residual radionuclides after the releases have ended. Radioactive decay is considered during the release, after deposition, and during holdup of food after harvest. The radiation dose models consider exposure to radionuclides deposited on the ground or crops from contaminated air or irrigation water, radionuclides in contaminated drinking water, aquatic foods raised in contaminated water, and radionuclides in bodies of water and sediments where people might fish, boat, or swim. For vegetation, the radiation dose model considers both direct deposition and uptake through roots. Doses may be calculated for either a maximum-exposed individual or for a population group. The program is designed to calculate accumulated radiation doses from the chronic ingestion of food products that contain radionuclides and doses from the external exposure to radionuclides in the environment. A first-year committed dose is calculated as well as an integrated dose for a selected number of years. 2 - Method of solution: A chain decay scheme including branching for transitions to and from isomeric states is used for radioactive decay. The equations for calculating internal radiation doses are derived from those given by the International Commission on Radio- logical Protection (ICRP) for body burdens and the maximum possible concentration (MPC) for each radionuclide. These doses are calculated as a function of radionuclide concentration in food products, ingestion rates, and a radionuclide-specific dose- commitment factor. Radiation doses from external exposure to contaminated water and soil are calculated using the basic assumption

AEROS: a real-time emergency response system for atmospheric releases of toxic material

International Nuclear Information System (INIS)

Nasstrom, J.S.; Greenly, G.D.

1986-01-01

The Atmospheric Release Advisory Capability (ARAC) at the Lawrence Livermore National Laboratory has developed a sophisticated computer-based real-time emergency response system for radiotoxic releases into the atmosphere. The ARAC Emergency Response Operating System (AEROS) has a centralized computer facility linked to remote site computers, meteorological towers, and meteorological data sources. The system supports certain fixed sites, but has the ability to respond to accidents at arbitrary locations. Product quality and response time are optimized by using complex three-dimensional dispersion models; extensive on-line data bases; automated data processing; and an efficient user interface, employing graphical computer displays and computer-displayed forms. Upon notification, the system automatically initiates a response to an emergency and proceeds through preliminary calculations, automatically processing accident information, meteorological data, and model parameters. The model calculations incorporate mass-consistent three-dimensional wind fields, terrain effects, and particle-in-cell diffusion. Model products are color images of dose or deposition contours overlaid on a base map

Riola release report

Energy Technology Data Exchange (ETDEWEB)

Woodward, E.C.

1983-08-04

Eleven hours after execution of the Riola Event (at 0826 PDT on 25 September 1980) in hole U2eq of the Nevada Test Site (NTS), a release of radioactivity began. When the seepage stopped at about noon the following day, up to some 3200 Ci of activity had been dispersed by light variable winds. On 26 September, examination of the geophone records showed six hours of low-level, but fairly continuous, activity before the release. Electrical measurements indicated that most cables were still intact to a depth below the stemming platform. A survey of the ground zero area showed that the seepage came through cracks between the surface conductor and the pad, through cracks in the pad, and through a crack adjacent to the pad around the mousehole (a small hole adjacent to the emplacement hole). To preclude undue radiation exposure or injury from a surprise subsidence, safety measures were instituted. Tritium seepage was suffucient to postpone site activities until a box and pipeline were emplaced to contain and remove the gas. Radiation release modeling and calculations were generally consistent with observations. Plug-hole interaction calculations showed that the alluvium near the bottom of the plug may have been overstressed and that improvements in the design of the plug-medium interface can be made. Experimental studies verified that the surface appearance of the plug core was caused by erosion, but, assuming a normal strength for the plug material, that erosion alone could not account for the disappearance of such a large portion of the stemming platform. Samples from downhole plug experiments show that the plug may have been considerably weaker than had been indicted by quality assurance (QA) samples. 19 references, 32 figures, 10 tables.

Riola release report

International Nuclear Information System (INIS)

Woodward, E.C.

1983-01-01

Eleven hours after execution of the Riola Event (at 0826 PDT on 25 September 1980) in hole U2eq of the Nevada Test Site (NTS), a release of radioactivity began. When the seepage stopped at about noon the following day, up to some 3200 Ci of activity had been dispersed by light variable winds. On 26 September, examination of the geophone records showed six hours of low-level, but fairly continuous, activity before the release. Electrical measurements indicated that most cables were still intact to a depth below the stemming platform. A survey of the ground zero area showed that the seepage came through cracks between the surface conductor and the pad, through cracks in the pad, and through a crack adjacent to the pad around the mousehole (a small hole adjacent to the emplacement hole). To preclude undue radiation exposure or injury from a surprise subsidence, safety measures were instituted. Tritium seepage was suffucient to postpone site activities until a box and pipeline were emplaced to contain and remove the gas. Radiation release modeling and calculations were generally consistent with observations. Plug-hole interaction calculations showed that the alluvium near the bottom of the plug may have been overstressed and that improvements in the design of the plug-medium interface can be made. Experimental studies verified that the surface appearance of the plug core was caused by erosion, but, assuming a normal strength for the plug material, that erosion alone could not account for the disappearance of such a large portion of the stemming platform. Samples from downhole plug experiments show that the plug may have been considerably weaker than had been indicted by quality assurance (QA) samples. 19 references, 32 figures, 10 tables

Subsurface oil release field experiment - observations and modelling of subsurface plume behaviour

International Nuclear Information System (INIS)

Rye, H.; Brandvik, P.J.; Reed, M.

1996-01-01

An experiment was conducted at sea, in which oil was released from 107 metres depth, in order to study plume behaviour. The objective of the underwater release was to simulate a pipeline leakage without gas and high pressure and to study the behaviour of the rising plume. A numerical model for the underwater plume behaviour was used for comparison with field data. The expected path of the plume, the time expected for the plume to reach the sea surface and the width of the plume was modelled. Field data and the numerical model were in good agreement. 10 refs., 2 tabs., 9 figs

Maxdose-SR and popdose-SR routine release atmospheric dose models used at SRS

Energy Technology Data Exchange (ETDEWEB)

Jannik, G. T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trimor, P. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-28

MAXDOSE-SR and POPDOSE-SR are used to calculate dose to the offsite Reference Person and to the surrounding Savannah River Site (SRS) population respectively following routine releases of atmospheric radioactivity. These models are currently accessed through the Dose Model Version 2014 graphical user interface (GUI). MAXDOSE-SR and POPDOSE-SR are personal computer (PC) versions of MAXIGASP and POPGASP, which both resided on the SRS IBM Mainframe. These two codes follow U.S. Nuclear Regulatory Commission (USNRC) Regulatory Guides 1.109 and 1.111 (1977a, 1977b). The basis for MAXDOSE-SR and POPDOSE-SR are USNRC developed codes XOQDOQ (Sagendorf et. al 1982) and GASPAR (Eckerman et. al 1980). Both of these codes have previously been verified for use at SRS (Simpkins 1999 and 2000). The revisions incorporated into MAXDOSE-SR and POPDOSE-SR Version 2014 (hereafter referred to as MAXDOSE-SR and POPDOSE-SR unless otherwise noted) were made per Computer Program Modification Tracker (CPMT) number Q-CMT-A-00016 (Appendix D). Version 2014 was verified for use at SRS in Dixon (2014).

Laminated materials with plastic interfaces: modeling and calculation

International Nuclear Information System (INIS)

Sandino Aquino de los Ríos, Gilberto; Castañeda Balderas, Rubén; Diaz Diaz, Alberto; Duong, Van Anh; Chataigner, Sylvain; Caron, Jean-François; Ehrlacher, Alain; Foret, Gilles

2009-01-01

In this paper, a model of laminated plates called M4-5N and validated in a previous paper is modified in order to take into account interlaminar plasticity by means of displacement discontinuities at the interfaces. These discontinuities are calculated by adapting a 3D plasticity model. In order to compute the model, a Newton–Raphson-like method is employed. In this method, two sub-problems are considered: one is linear and the other is non-linear. In the linear problem the non-linear equations of the model are linearized and the calculations are performed by making use of a finite element software. By iterating the resolution of each sub-problem, one obtains after convergence the solution of the global problem. The model is then applied to the problem of a double lap, adhesively bonded joint subjected to a tensile load. The adhesive layer is modeled by an elastic–plastic interface. The results of the M4-5N model are compared with those of a commercial finite element software. A good agreement between the two computation techniques is obtained and validates the non-linear calculations proposed in this paper. Finally, the numerical tool and a delamination criterion are applied to predict delamination onset in composite laminates

Realistic shell-model calculations for Sn isotopes

International Nuclear Information System (INIS)

Covello, A.; Andreozzi, F.; Coraggio, L.; Gargano, A.; Porrino, A.

1997-01-01

We report on a shell-model study of the Sn isotopes in which a realistic effective interaction derived from the Paris free nucleon-nucleon potential is employed. The calculations are performed within the framework of the seniority scheme by making use of the chain-calculation method. This provides practically exact solutions while cutting down the amount of computational work required by a standard seniority-truncated calculation. The behavior of the energy of several low-lying states in the isotopes with A ranging from 122 to 130 is presented and compared with the experimental one. (orig.)

Simulation of helium release in the Battelle Model Containment facility using OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Wilkening, Heinz; Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.eu

2013-12-15

Highlights: • The HYJET Jx7 hydrogen release experiment at BMC facility is studied using OpenFOAM. • The SST model and 2nd order numerics for momentum and species concentration are used. • The behaviour is captured well but helium concentration is generally over-predicted. • OpenFOAM needs smaller time steps, higher resolution, more CPU time compared to CFX. • The study shows the potential of open source CFD codes in some nuclear application. - Abstract: The open source CFD code OpenFOAM has been validated against an experiment of jet release phenomena in the Battelle Model Containment facility (BMC), and benchmarked with the Ansys CFX5.7 results. In the selected test, HYJET Jx7, helium was released into the containment at a speed of 42 m/s over a time of 200 s. The SST turbulence model was applied to model helium release and dispersion with both codes. The overall behaviour is captured adequately. However, there are still some noticeable differences between the CFX and OpenFOAM solutions. The study confirms the potential of using open source codes like OpenFOAM in some nuclear applications. Nevertheless further investigations and improvements are needed.

Simulation of helium release in the Battelle Model Containment facility using OpenFOAM

International Nuclear Information System (INIS)

Wilkening, Heinz; Ammirabile, Luca

2013-01-01

Highlights: • The HYJET Jx7 hydrogen release experiment at BMC facility is studied using OpenFOAM. • The SST model and 2nd order numerics for momentum and species concentration are used. • The behaviour is captured well but helium concentration is generally over-predicted. • OpenFOAM needs smaller time steps, higher resolution, more CPU time compared to CFX. • The study shows the potential of open source CFD codes in some nuclear application. - Abstract: The open source CFD code OpenFOAM has been validated against an experiment of jet release phenomena in the Battelle Model Containment facility (BMC), and benchmarked with the Ansys CFX5.7 results. In the selected test, HYJET Jx7, helium was released into the containment at a speed of 42 m/s over a time of 200 s. The SST turbulence model was applied to model helium release and dispersion with both codes. The overall behaviour is captured adequately. However, there are still some noticeable differences between the CFX and OpenFOAM solutions. The study confirms the potential of using open source codes like OpenFOAM in some nuclear applications. Nevertheless further investigations and improvements are needed

Modeling the release of E. coli D21g with transients in water content

Science.gov (United States)

Transients in water content are well known to mobilize colloids that are retained in the vadose zone. However, there is no consensus on the proper model formulation to simulate colloid release during drainage and imbibition. We present a model that relates colloid release to changes in the air-water...